ML101590078: Difference between revisions

From kanterella
Jump to navigation Jump to search
(StriderTol Bot change)
(StriderTol Bot change)
 
Line 16: Line 16:


=Text=
=Text=
{{#Wiki_filter:76. Which
{{#Wiki_filter:76. Which one of the following correctly completes the statement below?
: 76. Which one one ofof the following correctly the following      correctly completes completes the the statement statement below?below?
Technical Specifications do NOT require the RWCU isolation from the SLC control switch in Mode (1) due to (2)
Technical Specifications Technical      Specifications do  do NOTNOT require require the the RWCU RWCU isolation isolation from from the the SLC SLC control control switch    in Mode switch in Mode                   due (1) due to (2)
A (1) 3 (2) control rods are not able to be withdrawn since the reactor mode switch must be in the shutdown position and a control rod block is applied B. (1)3 (2) the operability of each individual control rod scram accumulator is required which will ensure that the control rods can be inserted C. (1)5 (2) control rods are not able to be withdrawn since the reactor mode switch must be in the shutdown position and a control rod block is applied D. (1)5 (2) the operability of each individual control rod scram accumulator is required which will ensure that the control rods can be inserted Feedback K/A: 204000 G2.02.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.
(1)              to    (2)
Reactor Water Cleanup System (CFR: 41.5 / 41.7 /43.2)
(1) 33 A (1)
A!I (2) control rods are not    not able to be  be withdrawn since the reactor   reactor mode mode switch must  must be be in the shutdown position in                    position and and aa control control rod rod block block is  is applied (1)3 B. (1)   3 (2) the operability of each individual control rod scram accumulator is required which will ensure that the control rods can be inserted (1)5 C. (1)   5 (2) control rods are not able to be withdrawn since the reactor mode switch must be in the shutdown position and a control rod block is applied (1)5 D. (1)   5 (2) the operability of each individual control rod scram accumulator is required which will ensure that the control rods can be inserted Feedback K/A: 204000 G2.02.25 KIA:
Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.
Reactor Water Cleanup System 41.5 / 41.7 /43.2)
(CFR: 41.5/41.7/43.2)
There are no safety limits associated with RWCU system, so question is written directly to the TS.
There are no safety limits associated with RWCU system, so question is written directly to the TS.
ROISRO               3214.2 RO/SRO Rating: 3.2/4.2 Objective:
ROISRO Rating: 3214.2 Objective:


==Reference:==
==Reference:==
Cog Level: Low Explanation: Are no safety limits for RWCU. SLC is required in mode 3 (RO knowledge) and the bases for the mode 3 requirement is SRO knowledge.
Distractor Analysis:
Choice A: Correct answer from the bases document.
Choice B: Plausible becasue this is the bases for Mode 4/5 from the bases document.
Choice C:
Choice D: Plausible because the scram accumulators are capable of inserting the control rods with low reactor pressure conditions, but the accumulators are not required to be operable in Mode 3.
SRO Basis: 10 CFR 55.43(b)-2, Facility operating limitations in the TS and their bases.
Knowledge of TS bases that is required to analyze TS required actions and terminology.
: 76. Which one of the following correctly completes the statement below?
Technical Specifications do NOT require the RWCU isolation from the SLC control switch in Mode (1) due to (2)
A!I (1) 3 (2) control rods are not able to be withdrawn since the reactor mode switch must be in the shutdown position and a control rod block is applied B. (1) 3 (2) the operability of each individual control rod scram accumulator is required which will ensure that the control rods can be inserted C. (1) 5 (2) control rods are not able to be withdrawn since the reactor mode switch must be in the shutdown position and a control rod block is applied D. (1) 5 (2) the operability of each individual control rod scram accumulator is required which will ensure that the control rods can be inserted Feedback KIA: 204000 G2.02.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.
Reactor Water Cleanup System (CFR: 41.5/41.7/43.2)
There are no safety limits associated with RWCU system, so question is written directly to the TS.
RO/SRO Rating: 3.2/4.2 Objective:


Cog Level: Low Explanation: Are no safety limits for RWCU. SLC is required in mode 33 (RO knowledge) and the bases for the mode 33 requirement is SRO knowledge.
==Reference:==
Cog Level: Low Explanation: Are no safety limits for RWCU. SLC is required in mode 3 (RO knowledge) and the bases for the mode 3 requirement is SRO knowledge.
Distractor Analysis:
Distractor Analysis:
Choice Choice A:    Correct answer A: Correct   answer from the the bases bases document.
Choice A: Correct answer from the bases document.
document.
Choice B: Plausible becasue this is the bases for Mode 4/5 from the bases document.
Choice B:
B: Plausible Plausible becasue becasue this is  is the bases bases for Mode Mode 4/5 from the bases bases document.
document.
Choice C:
Choice C:
Choice Choice D:
Choice D: Plausible because the scram accumulators are capable of inserting the control rods with low reactor pressure conditions, but the accumulators are not required to be operable in Mode 3.
D: Plausible Plausible because because thethe scram scram accumulators accumulators are are capable capable ofof inserting inserting the the control control rods rods with with low low reactor pressure conditions, reactor pressure                but the conditions, but        accumulators are the accumulators    are not not required required toto be   operable in be operable    in Mode Mode 3.3.
SRO Basis: 10 CFR 55.43(b)-2, Facility operating limitations in the TS and their bases.
SRO SRO Basis:
Knowledge of TS bases that is required to analyze TS required actions and terminology.  
Basis: 1010 CFR CFR 55.43(b)-2, 55.43(b)-2, Facility Facility operating operating limitations limitations in in the the TS TS and and their their bases.
bases.
Knowledge Knowledge of  of TS TS bases bases that that is is required required to   analyze TS to analyze  TS required required actions actions and and terminology.
terminology.


Notes Notes RECTC MODE RE~.CTOR        MODE            AVERIo.GE AVERA3E REACTOR REACTOR MDDE MODE                TTLE TiTLE                    S1TCH POSfllON SWlTCH        POSfIICN         COOLANT EMPERTURE COOLANTIEMPERA            Th'RE
Notes RECTC MODE AVERA3E REACTOR MDDE TTLE S1TCH POSfIICN COOLANT EMPERTURE (F:
(=-l-t (F:
1 Pr Cperatbn Run 2
11    Pr Operatoo P'))VI,'Er Cperatbn             Run Run                                              N."'.
S1aiup Reftle? o-SatupHot NA Sanby 3
:22    S1aiup Slartup                          Reftle? ,or Refuer"')  o- S;,artup;'HD1 SatupHot                         Nt.!;.
Hot hutdon Sutcii
NA Sanby S1aoooji 33    Hot Shutdall.m'1I)
> 21.2 4
Hot   hutdon                   Sutcii Shutdb~\T1                                    >2'&2
Cold Sutdcn
                                                                                          > 21.2 44    Cold Silub:i'o\'ltn Cold    Sutdcn    t* ,        iuchr, Sh,utdb''ltn                                      21:2 S;212 5     Reje1g RefueEflg(ll1                    Sttubcii or R~ilel Silutdo''ltn      ReieI                             ..
: iuchr, 212 5
NI!.
Reje1g Sttubcii or ReieI From Bases 3.3.6.1 One channel of the SLC System Initiation Function is available and required to be OPERABLE only in MODES 1 and 2, since these are the only MODES where the reactor can be critical, and these MODES are consistent with the Applicability for the SLC System (LCO 3.1.7).
From Bases 3.3.6.1 One channel of the SLC System Initiation Function is available and OPERABLE only in MODES 'I1 and 2, since these are the required to be OPERA.BLE only MODES where the reactor can be critical, and these MODES are consistent with the Applicability for the SLC System (LCO 3.'1.7). 3.1.7).
From bases 3.1.7 APPLICABILITY In MODES I and 2, shutdown capability is required. In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in the shutdown position and a control rod block is applied. This provides adequate controls to ensure that the reactor remains subcritical. In MODE 6, only a single control rod can be withdrawn from a core cell containing fuel assemblies. Determination of adequate SDM (LCD 3.1.1, SHUTDOWN MARGIN (SDM)) ensures that the reactor will not become critical with the analytically determined strongest control rod withdrawn.
From bases 3.1.7 APPLICABILITY               In MODES 'II and 2, shutdown capability is required. In MODES 3 and 4,
Therefore, the SLC System is not required to be OPERABLE when only a single control rod can be withdrawn.
[n control rods are not able to be withdrawn since the reactor mode switch is in the shutdown position and a control rod block is applied. This provides adequate controls to ensure that the reactor remains sub              criticaL In subcritical.
MODE 5,  6, only a single control rod can be withdrawn from a core cell containing fuel assemblies. Determination of adequate SDM {LCO                    3. '1.'1 ,
(LCD 3.1.1, SHUTDOWN "SHUTDOWN MARGIN (SDM}")      (SDM)) ensures that the reactor ...willvill not become critical with the analytically determined strongest control rod withdrawn.
Therefore, the SLC System is not required to be OPERABLE OPERA.BLE when only aa single control rod can be withdrawn.
Categories K/A:
Categories K/A:
KIA:              204000 204000 G2.02.25                                     Tier / Group:
204000 G2.02.25 Tier / Group:
Group: T2G2 RO Rating:
T2G2 RO Rating:
RORating:         3.2 3.2                                                 SRO  Rating: 4.2 SRORating:    4.2 LP LP Obj:
3.2 SRO Rating:
Obj:          05-11                                               Source:
4.2 LP Obj:
Source:        NEW NEW Cog Cog Level:
05-11 Source:
Level:      LOW LOW                                                 Category Category 8:
NEW Cog Level:
8:
LOW Category 8:
: 77. Unit
Notes RE~.CTOR MODE AVERIo.GE REACTOR MODE TiTLE SWlTCH POSfllON COOLANTIEMPERA Th'RE
: 77. Unit One One isis operating operating at at full full power power when when the the following following plant plant conditions conditions occur:
(=-l-t,,'
occur:
1 P'))VI,'Er Operatoo Run N."'.
Load Reject
:2 Slartup Refuer"'),or S;,artup;'HD1 Nt.!;.
        - Load
S1aoooji 3
        -        Reject Signal Signal received received Line  31  (Whiteville
Hot Shutdall.m'1I)
        - Line 31 (Whiteville Line)
Shutdb~\\T1
        -                          Line) PCBs PCBs redred lights lights are are lit lit Line 31
>2'&2 4
        - Line
Cold Silub:i'o\\'ltnt*,
        -        31 (Whiteville (Whiteville Line)
Sh,utdb''ltn S; 21:2 5
Line) white white VOLT VOLT lightslights are are not not lit lit All other
RefueEflg(ll1 From Bases 3.3.6.1 From bases 3.1.7 APPLICABILITY Categories Silutdo''ltn or R~ilel NI!...
        - All
One channel of the SLC System Initiation Function is available and required to be OPERA.BLE only in MODES 'I and 2, since these are the only MODES where the reactor can be critical, and these MODES are consistent with the Applicability for the SLC System (LCO 3.'1.7).
        -    other line line PCBs PCBs green green lights lights are are lit lit 230 KV
[n MODES 'I and 2, shutdown capability is required. In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in the shutdown position and a control rod block is applied. This provides adequate controls to ensure that the reactor remains sub criticaL In MODE 5, only a single control rod can be withdrawn from a core cell containing fuel assemblies. Determination of adequate SDM {LCO 3. '1.'1,
        - 230
"SHUTDOWN MARGIN (SDM}") ensures that the reactor... vill not become critical with the analytically determined strongest control rod withdrawn.
        -        KV BUS BUS 1A    BUS POT IA BUS     POT UNDER UNDERVOL  VOLTAGETAGE is    is in in alarm alarm 230 KV
Therefore, the SLC System is not required to be OPERA.BLE when only a single control rod can be withdrawn.
        - 230
KIA:
        -        KV BUS BUS 1B lB BUS BUS POT POT UNDERVOLTAGE UNDER VOL TAGE is             is in in alarm alarm Which one Which     one of    the following of the following identifies identifies the the initial initial RPS RPS trip trip signal signal and and the the procedure procedure which which contains the contains     the guidance guidance to     trip the to trip  the Whiteville Whiteville Line Line PCBs?
204000 G2.02.25 Tier / Group: T2G2 RORating:  
PCBs?
 
A'I  Control Valve Fast A Control              Fast Closure; Closure; OAOP-36.1, Loss OAOP-36.1,        Loss ofof Any Any 4160V 4160V Buses Buses or       480V E-Buses.
===3.2 SRORating===
or 480V      E-Buses.
4.2 LP Obj:
Stop Valve Closure; B. Stop OAOP-36. 1, Loss OAOP-36.1,        Loss of Any 4160V41 60V Buses Buses or 480V E-Buses. E-Buses.
05-11 Source:
NEW Cog Level:
LOW Category 8:
: 77. Unit One is operating at full power when the following plant conditions occur:
- Load Reject Signal received
- Line 31 (Whiteville Line) PCBs red lights are lit
- Line 31 (Whiteville Line) white VOLT lights are not lit
- All other line PCBs green lights are lit
- 230 KV BUS IA BUS POT UNDERVOLTAGE is in alarm
- 230 KV BUS lB BUS POT UNDERVOL TAGE is in alarm Which one of the following identifies the initial RPS trip signal and the procedure which contains the guidance to trip the Whiteville Line PCBs?
A Control Valve Fast Closure; OAOP-36.1, Loss of Any 4160V Buses or 480V E-Buses.
B. Stop Valve Closure; OAOP-36. 1, Loss of Any 41 60V Buses or 480V E-Buses.
C. Control Valve Fast Closure; OAOP-22, Grid Instability.
C. Control Valve Fast Closure; OAOP-22, Grid Instability.
D. Stop Valve Closure; OAOP-22, Grid Instability.
D. Stop Valve Closure; OAOP-22, Grid Instability.
: 77. Unit One is operating at full power when the following plant conditions occur:
- Load Reject Signal received
- Line 31 (Whiteville Line) PCBs red lights are lit
- Line 31 (Whiteville Line) white VOLT lights are not lit
- All other line PCBs green lights are lit
- 230 KV BUS 1A BUS POT UNDER VOL TAGE is in alarm
- 230 KV BUS 1B BUS POT UNDERVOLTAGE is in alarm Which one of the following identifies the initial RPS trip signal and the procedure which contains the guidance to trip the Whiteville Line PCBs?
A'I Control Valve Fast Closure; OAOP-36.1, Loss of Any 4160V Buses or 480V E-Buses.
B. Stop Valve Closure; OAOP-36.1, Loss of Any 4160V Buses or 480V E-Buses.
C. Control Valve Fast Closure; OAOP-22, Grid Instability.
D. Stop Valve Closure; OAOP-22, Grid Instability.


Feedback Feedback K/A: 212000A2.12 KIA:    212000 A2.12 Ability to Ability    to (a)
Feedback K/A: 212000 A2.12 Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
(a) predict predict thethe impacts impacts ofof the the following following on  the REACTOR on the REACTOR PROTECTION PROTECTION SYSTEM;SYSTEM ; andand (b) based (b)  based on  on those those predictions, predictions, use use procedures procedures toto correct, correct, control, control, oror mitigate mitigate the the consequences consequences of those of  those abnormal abnormal conditions conditions or or operations:
Main turbine stop control valve closure (CFR: 41.5/ 45.6)
operations:
RO/SRO Rating: 4.0/4.1 Objective: CLS-LP-03, Obj. 8.
Main turbine Main     turbine stop stop control control valve valve closure closure (CFR: 41.5 (CFR:    41.5/   45.6) 145.6)
List the RPS trip signals, including setpoints and how/when each signal is bypassed.
RO/SRO Rating:
RO/SRO       Rating: 4.0/4.1 4.0/4.1 Objective: CLS-LP-03, Objective:     CLS-LP-03, Obj. Obj. 8.
8.
List the List   the RPS RPS trip  signals, including trip signals, including setpoints setpoints andand how/when how/when each each signal signal is is bypassed.
bypassed.


==Reference:==
==Reference:==
SD-03 Reactor Protection System, section 3.1 RPS Trips Cog Level High Explanation:
A load reject signal at any reactor power level will cause a turbine control valve fast closure scram. The load reject signal does not input into the turbine stop valve closure scram logic. During a loss of offsite power, if the grid is lost all PCBs are opened per OAOP-36. 1.
Distractor Analysis:
Choice A: Correct answer, see explanation Choice B: Incorrect Load reject initiates a TCV fast closure scram not a TSV. A misconception of the difference between TCV and TSV scrams may cause a student to select this answer.
Choice C: Incorrect. OAOP-22 does not have an action for loss of grid only for degraded.
Choice D: Incorrect. Load reject initiates a TCV fast closure scram not a TSV. A misconception of the difference between TCV and TSV scrams may cause a student to select this answer. OAOP-22 does not have an action for loss of grid only for degraded.
SRO Basis: 10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Feedback KIA: 212000A2.12 Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Main turbine stop control valve closure (CFR: 41.5 145.6)
RO/SRO Rating: 4.0/4.1 Objective: CLS-LP-03, Obj. 8.
List the RPS trip signals, including setpoints and how/when each signal is bypassed.


==Reference:==
==Reference:==
 
SD-03 Reactor Protection System, section 3.1 RPS Trips Cog Level High Explanation:
SD-03 Reactor SD-03      Reactor Protection Protection System, System, section section 3.1 3.1 RPS RPS Trips Trips Cog Level Cog    Level High High Explanation:
A load reject signal at any reactor power level will cause a turbine control valve fast closure scram. The load reject signal does not input into the turbine stop valve closure scram logic. During a loss of offsite power, if the grid is lost all PCBs are opened per OAOP-36.1.
Explanation:
load reject A load     reject signal at any reactor power level will cause a turbine control valve fast closure scram. The load reject signal does not input into the turbine stop valve closure scram logic. During      During a loss of offsite power, if the grid is lost all PCBs are opened per OAOP-36.1. OAOP-36. 1.
Distractor Analysis:
Distractor Analysis:
Choice A: Correct answer, see explanation Choice B: Incorrect Load reject initiates a TCV fast closure scram not a TSV. A misconception of the difference between TCV and TSV scrams may cause a student to select this answer.
Choice A: Correct answer, see explanation Choice B: Incorrect Load reject initiates a TCV fast closure scram not a TSV. A misconception of the difference between TCV and TSV scrams may cause a student to select this answer.
Choice C: Incorrect. OAOP-22 does not have an action for loss of grid only for degraded.
Choice C: Incorrect. OAOP-22 does not have an action for loss of grid only for degraded.
Choice D: Incorrect. Load reject initiates a TCV fast closure scram not a TSV. A misconception of the difference between TCV and TSV scrams may cause a student to select this answer. OAOP-22 does not have an action for loss of grid only for degraded.
Choice D: Incorrect. Load reject initiates a TCV fast closure scram not a TSV. A misconception of the difference between TCV and TSV scrams may cause a student to select this answer. OAOP-22 does not have an action for loss of grid only for degraded.
SRO Basis: 10    10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
SRO Basis: 10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.  
 
Notes An example of Turbine Control Valve Fast Closure is a load reject.
The definition of a load reject is greater than 40% mismatch between electrical output and mechanical input as sensed by generator stator amps and the Cross Around Piping pressure. This is to prevent excessive overspeed of the Turbine on loss of load.
A load reject signal will energize the fast acting Solenoid Valves on the control valve actuators, which removes hydraulic trip fluid pressure. The trip signal comes from pressure switches on the Vast acting trip control (FATC) supply to the control valve disc dumps (refer to EHC Hydraulics). Loss of this pressure will cause a rapid closure of the control valves. Circuitry is designed such that the pressure switch on either control valve 1 or 3 will trip RPS Trip System A. Either control valve 2 or 4 will trip RPS Trip System B.
These switches will also provide a Scram signal on loss of hydraulic trip fluid pressure when a load reject signal is not present - loss of hydraulic fluid pressure can result in a fast closure of the control valves SD-03 Rev. 9 Page 21 of 89 7.
IF the SAT was lost due to loss of power on the Progress Energy System, THEN PERFORM the following:
a.
PLACE AUTO RELOSE switches in MANUAL.
LI b.
PLACE transmission line PCB SUPERVISORY LOCAL/REMOTE switches in LOCAL.
c, TRIP all transmission line PCBs.
LI OAOP-36.2 Rev.
I Page 4 of 196 Categories K/A:
212000 A2.12 Tier / Group:
T2G1 RO Rating:
4.0 SRO Rating:
4.1 LP Obj:
03-08 Source:
PREV Cog Level:
HIGH Category 8:
Y Notes 1 SD-03 An example of Turbine Control Valve Fast Closure is a load reject.
The definition of a load reject is greater than 40% mismatch betvveen electrical output and mechanical input as sensed by generator stator amps and the Cross Around Piping pressLire. This is to prevent excessive overs peed of the Turbine on loss of load.
A load reject signal will energize the fast acting Solenoid Valves on the control valve actuators, which removes hydraulic trip fluid pressure. The trip signal comes from pressure switches on the fast acting trip control (FATC) supply to the control valve disc dumps (refer to EHC Hydraulics). Loss of this pressure will cause a rapid closure of the control valves. Circuitry is designed such that the pressure switch on either control valve 'lor 3 Will trip RPS Trip system A. Either control valve 2 or 4 will trip RPS Trip System B.
These switches will also provide a Scram signal on loss of hydraulic trip fluid pressure when a load reject signal is not present - loss of hydraulic fluid pressure can result in a fast closure of the control valves.
Rev. 9 Page 2'1 of 891 7,
IF tile SAT was lost due to loss of power on the Progress Energy System, THEN PERFORM the following:
: a.
.PLACE AUTO RECLOSE switches in MANUAL.
D
: b.
PLACE transmission line PCB SUPERVISORY D
LOCAUREMOTE switches in LOCAL
: c.
TRIP all transmission line PCBs.
D IOAOP-36.2 Rev. 4'1 Page 4 of 1961 Categories KIA:
212000 A2.12 RO Rating:
4.0 LP Obj:
03-08 Cog Level:
HIGH Tier / Group: T2G 1 SRO Rating:


Notes Notes An example An    example of of Turbine Turbine Control Control Valve Valve Fast Fast Closure Closure is  is aa load load reject.
===4.1 Source===
reject.
PREY Category 8:
The definition The    definition of of aa load load reject reject is is greater greater than than 40%
Y
40% mismatch mismatch betvveen between electrical  output and electrical output    and mechanical mechanical input input as as sensed sensed by  by generator generator stator stator amps and amps    and the the Cross Cross Around Piping Piping pressLire.
: 78. The Unit is at 7% power during reactor startup.
pressure. ThisThis isis to to prevent prevent excessive overs excessive          peed of overspeed    of the the Turbine Turbine on on loss loss ofof load.
The operator withdraws control rod 26-27 to position 48.
load.
The following indications are noted:
AA load reject reject signal will energize energize thethe fast fast acting Solenoid Valves on        on the control the   control valve valve actuators, which removes removes hydraulic hydraulic triptrip fluid fluid pressure. The trip signal pressure.                signal comes comes from from pressure switches switches on the   the fast Vast acting trip control (FATC) supply to the control valve disc dumps (refer to EHC Hydraulics). Loss of this pressure  pressure will causecause a rapidrapid closure of the control valves. Circuitry Circuitry is is designed such that the pressure switch on either control valve 'lor      1 or 3 Will will trip RPS Trip System A. Either system        Either control valve 2 or 4 will trip RPS  RPS TripTrip System B.
- ROD DRIFT alarm seals in
These switches will also provide a Scram signal on loss of hydraulic trip fluid pressure when a load reject signal is not present - loss of      -
- ROD OVER TRAVEL alarm seals in
hydraulic fluid pressure can result in a fast closure of the control valves.
- Rod 26-27 full core display red light out Which one of the following identifies:
valves 1 SD-03                                        Rev. 9                                        Page 2'1       891 21 of 89 7.
(1) the indication that would be displayed on the four-rod group display and (2) the required action for the inoperable control rod lAW Technical Specification 3.1.3, Control Rod Operability?
7,     IF tile the SAT was lost due to loss of power on the Progress Energy System, THEN PERFORM the following:
A. (1)48 (2) Fully insert control rod 26-27 and disarm the HCU B. (1)48 (2) Verify 1 2 control rods are withdrawn and implement GP-1 1, Second Operator Rod Sequence Checkoff Sheets C (1) Blank (2) Fully insert control rod 26-27 and disarm the HCU D. (1) Blank (2) Verify 1 2 control rods are withdrawn and implement GP-1 1, Second Operator Rod Sequence Checkoff Sheets
: a. PLACE AUTO RECLOSE
: 78. The Unit is at 7% power during reactor startup.
                          .PLACE              RELOSE switches in MANUAL.                                D LI
The operator withdraws control rod 26-27 to position 48.
: b. PLACE transmission line PCB SUPERVISORY                                      D LOCAL/REMOTE LOCAUREMOTE switches in LOCAL.        LOCAL c,
The following indications are noted:  
: c. TRIP all transmission line PCBs.                                              D LI OAOP-36.2 IOAOP-36.2                                      Rev. 4'1 I                    Page 44 of 1961961 Categories K/A:
- ROD DRIFT alarm seals in  
KIA:        212000 A2.12                                  Tier / Group:    T2G1 T2G 1 RO Rating:  4.0                                            SRO Rating:      4.1 LP Obj:      03-08                                          Source:
- ROD OVER TRAVEL alarm seals in
Source:          PREY PREV Cog Cog Level:
- Rod 26-27 full core display red light out Which one of the following identifies:
Level:  HIGH                                          Category Category 8:8:    Y
(1) the indication that would be displayed on the four-rod group display and (2) the required action for the inoperable control rod lAW Technical Specification 3.1.3, Control Rod Operability?
: 78. The Unit
A. (1) 48 (2) Fully insert control rod 26-27 and disarm the HCU B. (1) 48 (2) Verify ~12 control rods are withdrawn and implement GP-11, Second Operator Rod Sequence Checkoff Sheets C!I (1) Blank (2) Fully insert control rod 26-27 and disarm the HCU D. (1) Blank (2) Verify ~12 control rods are withdrawn and implement GP-11, Second Operator Rod Sequence Checkoff Sheets  
: 78. The    Unit isis at 7% power at 7%    power during    reactor startup.
during reactor     startup.
The operator The    operator withdraws withdraws control control rod rod 26-27 26-27 to    position 48.
to position 48.
The following The    following indications indications are are noted:
noted:
ROD DRIFT
        - ROD
        -          DRIFT alarm alarm seals seals in in ROD OVER
        - ROD
        -          OVER TRAVEL TRAVEL alarm alarm seals seals inin Rod 26-27
        - Rod
        -        26-27 full full core core display display red red light light out out Which one Which    one of    the following of the   following identifies:
identifies:
(1) the indication (1) the  indication thatthat would would be  displayed on be displayed     on the the four-rod four-rod group group display display and and (2) the required action for the inoperable control rod lAW Technical Specification 3.1.3, Control Rod Operability?
3.1.3, (1)48 A. (1)   48 Fully insert control rod 26-27 and disarm the HCU (2) Fully                                                      HCU B. (1)
B.  (1)4848 (2) Verify ~12 1 2 control rods are withdrawn and implement GP-11,      GP-1 1, Second Operator Rod Sequence Checkoff Sheets C (1) Blank C!I (2) Fully insert control rod 26-27 and disarm the HCU D. (1) Blank (2) Verify ~12 1 2 control rods are withdrawn and implement GP-11,      GP-1 1, Second Operator Rod Sequence Checkoff Sheets


Feedback Feedback K/A: 214000 KIA:    214000 A2.03 A2.03 Ability to Ability   to (a)   predict the (a) predict        impacts of the impacts         the following of the   following on on the the ROD ROD POSITION POSITION INFORMATION INFORMATION SYSTEM; SYSTEM; and (b) and    (b) based based on on those those predictions, predictions, use use procedures procedures to to correct, correct, control, control, or or mitigate mitigate the the consequences of consequences          of those those abnormal abnormal conditions conditions or or operations:
Feedback K/A: 214000 A2.03 Ability to (a) predict the impacts of the following on the ROD POSITION INFORMATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
operations:
Overtravel/in-out (CFR: 41.5 / 45.6)
Overtravel/in-out Overtravel/in-out (CFR: 41.5 (CFR:    41.5 /45.6)
RO/SRO Rating: 3.6/3.9 Objective: CLS-LP-07 Obj 5b Given plant conditions, determine if the following conditions exist: b. Indications of an uncoupled control rod.
                / 45.6)
RO/SRO Rating:
RO/SRO      Rating: 3.6/3.9 3.6/3.9 Objective: CLS-LP-07 Objective:      CLS-LP-07 Obj  Obj 5b 5b Given plant Given     plant conditions, conditions, determine determine ifif the the following following conditions conditions exist:
exist: b.
: b. Indications Indications of of an an uncoupled uncoupled control control rod.
rod.


==Reference:==
==Reference:==
SD-07 page 27 TS 3.1.3 Cog Level Low Explanation:
If the control rod is in the overtravel out position, the corresponding digital indicator will be blank since the magnet will not be near any of the 00 to 48 reed switches. lAW TS the rod is declared inoperable then inserted to 00 (within 3 hours) and disarmed (within 4 hours). TS 3.1.6 if the RWM is inoperable then if
>12 control rods are withdrawn GP-1 1 would be implemented, unless the rod is at 00 and is not intended to be moved.
Distractor Analysis:
Choice A: Plausible because the full in and 00 indications are at the same point or the exam inee may think that the rod may settle to the 48 position.
Choice B: Plausible because the full in and 00 indications are at the same point or the examinee may think that the rod may settle to the 48 position. These are TS actions for an inoperable RWM, not control rod.
Choice C: Correct answer, see explanation.
Choice D: Plausible because this is the correct indication but these are TS actions for an inoperable RWM, not control rod.
SRO Basis: 10 CFR 55.43(b)-2, Facility operating limitations in the technical specifications and their bases. Application of required actions statements.
Feedback KIA: 214000 A2.03 Ability to (a) predict the impacts of the following on the ROD POSITION INFORMATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
Overtravel/in-out (CFR: 41.5 /45.6)
RO/SRO Rating: 3.6/3.9 Objective: CLS-LP-07 Obj 5b Given plant conditions, determine if the following conditions exist: b. Indications of an uncoupled control rod.


==Reference:==
==Reference:==
 
SO-07 page 27 TS 3.1.3 Cog Level Low Explanation:
SD-07 page SO-07     page 27 27 TS 3.1.3 TS     3.1.3 Level Cog Level Cog Low Low Explanation:
If the control rod is in the overtravel out position, the corresponding digital indicator will be blank since the magnet will not be near any of the 00 to 48 reed switches. lAW TS the rod is declared inoperable then inserted to 00 (within 3 hours) and disarmed (within 4 hours). TS 3.1.6 if the RWM is inoperable then if  
Explanation:
.=::12 control rods are withdrawn GP-11 would be implemented, unless the rod is at 00 and is not intended to be moved.
If the control rod is in the overtravel out position, the corresponding corresponding digital indicator will be blank since the magnet will not be near any of the 00 to 48 reed switches. lAW TS the rod is declared inoperable then magnet inserted to 00 (within 3 hours) and disarmed (within 4 hours). TS 3.1.6 if the RWM is inoperable then if
Oistractor Analysis:
>12 control rods are withdrawn GP-11
Choice A: Plausible because the full in and 00 indications are at the same point or the examinee may think that the rod may settle to the 48 position.
.=::12                                GP-1 1 would be implemented, unless the rod is at 00 and is not intended to be moved.
Distractor Analysis:
Oistractor Choice A: Plausible because the full in and 00 indications are at the same point or the examinee         exam inee may think that the rod may settle to the 48 position.
Choice B: Plausible because the full in and 00 indications are at the same point or the examinee may think that the rod may settle to the 48 position. These are TS actions for an inoperable RWM, not control rod.
Choice B: Plausible because the full in and 00 indications are at the same point or the examinee may think that the rod may settle to the 48 position. These are TS actions for an inoperable RWM, not control rod.
Choice C: Correct answer, see explanation.
Choice C: Correct answer, see explanation.
Choice D:  0: Plausible because this is the correct indication but these are TS actions for an inoperable RWM, not control rod.
Choice 0: Plausible because this is the correct indication but these are TS actions for an inoperable RWM, not control rod.
SRO Basis:
SRO Basis: 10 CFR 55.43(b)-2, Facility operating limitations in the technical specifications and their bases. Application of required actions statements.  
Basis: 1010 CFR CFR 55.43(b)-2, 55.43(b)-2, Facility Facility operating limitations limitations in in the technical specifications and their bases. Application of bases.                    of required actions statements.


Notes Notes From SD-07 From  SD-07
Notes From SD-07 Coupling integrity of a control rod shall be checked anytime a control rod is fully withdrawn by verifying that the rod does not reach the overtravel position. An uncoupling check can be performed by maintaining the continuous withdraw signal for approximately 3 to 5 seconds when the control rod has reached position 48 and verifying the control rod does not retract beyond position 48. If the rod is uncoupled, then the four rod display indication will go out for the uncoupled rod and the Rod Over Travel Annunciator A-05 4-2 will illuminate.
                **      Coupling integrity Coupling    integrity of  of aa control control rod rod st"lall shall bebe checked checked anytime anytime aa control rod control  rod isis fully fully withdrawn withdrawn by    by verifying verifying that that the the rod rod does does not not reach the reacl)  the overtravel overtravel position.
SD-07 Rev. 6 Page 27 of 57 C.
position. An An uncoupling uncoupling check check can can bebe performed by performed      by maintaining maintaining the the continuous continuous withdrawwithdraw signal signal for for approximately 33 to approximately            to 55 seconds seconds whenwhen the  the control control rod rod t"las has reached reached position 48 position   48 and and verifying verifying thethe control control rodrod does does notnot retract retract beyOnd beyond position 48.
One or more control rods C. 1 inoperable for reasons other Inoperable control rod may than Condition A or 8.
position    48. IfIf the the rod rod isis uncoupled, uncoupled, then then thethe four four rod rod display display indication will indication  will go go out out for for the the uncoupled uncoupled rod    rod and and the the Rod Rod Over Over Travel  Annunciator Travel Annunciator A-05 4-2 A-05    4-2 will will illuminate.
be bypassed in the RWM or RWM may be bypassed as allowed by LCO 3.3.2.1, if required, to allow insertion of inoperable control rod and continued operation.
illuminate.
Fully insert inoperable 3 hours control rod.
I SD-07 SD-07                                                   Rev. 6 Rev.S                                          Page Page 2727 of 571 57 C. One or more control rods                 C. 1 C.*1        ----------NOTE--------
inoperable for reasons other                         Inoperable control rod may than Condition A or B. 8.                           be bypassed in the RWM or RWM may  may be bypassed bypassed as allowed bv  by LCO 3.3.2.1:
3.3.2.1, if required, to allow insertion of inoperable control rod and continued operation.
Fully insert inoperable               3 hours control rod.
AND (continued)
AND (continued)
C.   (continued)                           C.2         Disarm the associated                 4 hours CRD.
C.
(continued)
C.2 Disarm the associated 4 hours CRD.
From GP-11:
From GP-11:
This procedure provides aa method    method for aa second second licensed licensed operator or qualified qualified member member of  of the plant plant technical staff to      to verify control control rod movement movement and  and compliance compliance withwith the prescribed BPWS      BPWS control rod pattern    pattern with thethe rod worth minimizer    (RWM) inoperable minimizer (RWM)       inoperable in    in conformance conformance with   with the the requirements of    of Technical Technical Specification Specification 332.1.
This procedure provides a method for a second licensed operator or qualified member of the plant technical staff to verify control rod movement and compliance with the prescribed BPWS control rod pattern with the rod worth minimizer (RWM) inoperable in conformance with the requirements of Technical Specification 332.1. If the RWM is inoperable due to bypassed control rod(s) that will not be moved during the startup/shutdow, then this procedure is not required.
3.3.2.1. IfIf the the RWM RWM is  is inoperable inoperable due  due to  to bypassed bypassed control control rod(s) rod{s) that will Will not not be be moved moved during during the  the startup/shutdow, startup/shutdown, then      then this this procedure procedure is is not not required.
Categories K/A:
required.
214000 A2.03 Tier / Group:
Categories Categories K/A:
T2G2 RO Rating:
KIA:            214000 214000 A2.03 A2.03                                        Tier Tier // Group:
3.6 SRO Rating:
Group:    T2G2 T2G2 RO Rating:
3.9 LP Obj:
RORating:      3.6 3.6                                                    SRO   Rating:
07-5B Source:
SRORating:          3.9 3.9 LP Obj:
NEW Cog Level:
LPObj:          07-5B 07-SB                                                  Source:
HIGH Category 8:
Source:            NEW NEW Cog Cog Level:
Y Notes From SD-07 Coupling integrity of a control rod st"lall be checked anytime a control rod is fully withdrawn by verifying that the rod does not reacl) the overtravel position. An uncoupling check can be performed by maintaining the continuous withdraw signal for approximately 3 to 5 seconds when the control rod t"las reached position 48 and verifying the control rod does not retract beyOnd position 48. If the rod is uncoupled, then the four rod display indication will go out for the uncoupled rod and the Rod Over Travel Annunciator A-05 4-2 will illuminate.
Level:    HIGH HIGH                                                   Category Category 8:   8:   YY
I SD-07 Rev.S Page 27 of 571 C.
: 79. Given the
One or more control rods C.*1 inoperable for reasons other than Condition A or B.
: 79. Given the following following ATWS ATWS conditions conditions on    Unit Two:
AND C.
on Unit Two:
(continued)
2A CRD 2A  CRD Pump Pump                Overcurrent trip Overcurrent    trip 2B CRD 2B CRD PumpPump                Shaft  uncoupled Shaft uncoupled HPCI System HPCI    System                  Under Clearance Under   Clearance SLC SLC                            Both squib Both   squib valves valves failed failed to to fire fire RCIC RCIC                            Running   with Running with an  an unisolable unisolable steam steam supply supply leak leak Suppression Pool Suppression    Pool Level Level   -24 inches
C.2 From GP-11:  
                                      -24 inches Reactor Power Reactor    Power                10%
----------NOTE--------
10%
Inoperable control rod may be bypassed in the RWM or RWM may be bypassed as allowed bv LCO 3.3.2.1: if required, to allow insertion of inoperable control rod and continued operation.
Reactor Water Reactor  Water Level Level         160  inches 160 inches Which one Which  one of of the the following following identifies identifies the the action action that that would would be be taken taken concerning concerning the the RCIC  system    based RCIC system based on    on the conditions  above?
Fully insert inoperable control rod.
conditions above?
Disarm the associated CRD.
The RCIC The  RCIC system system would:
3 hours (continued) 4 hours This procedure provides a method for a second licensed operator or qualified member of the plant technical staff to verify control rod movement and compliance with the prescribed BPWS control rod pattern with the rod worth minimizer (RWM) inoperable in conformance with the requirements of Technical Specification 3.3.2.1. If the RWM is inoperable due to bypassed control rod{s) that Will not be moved during the startup/shutdown, then this procedure is not required.
would:
Categories KIA:
source of the steam leak lAW OAOP-05.0, Radioactive A. be isolated to secure the source                                                Radioactive Spills, High Spills, High Radiation, Radiation, and and Airborne Activity.
214000 A2.03 Tier / Group: T2G2 RORating:
B. have the high suppression pool water level transfer defeated and its suction transferred back to the CST lAW SEP-1  SEP-i 0, Circuit Alterations Procedure.
 
C C~ remain running because it is needed for boron injection lAW LEP-03, Alternate Boron Injection.
===3.6 SRORating===
===3.9 LPObj===
07-SB Source:
NEW Cog Level:
HIGH Category 8:
Y
: 79. Given the following ATWS conditions on Unit Two:
2A CRD Pump Overcurrent trip 2B CRD Pump Shaft uncoupled HPCI System Under Clearance SLC Both squib valves failed to fire RCIC Running with an unisolable steam supply leak Suppression Pool Level
-24 inches Reactor Power 10%
Reactor Water Level 160 inches Which one of the following identifies the action that would be taken concerning the RCIC system based on the conditions above?
The RCIC system would:
A. be isolated to secure the source of the steam leak lAW OAOP-05.0, Radioactive Spills, High Radiation, and Airborne Activity.
B. have the high suppression pool water level transfer defeated and its suction transferred back to the CST lAW SEP-i 0, Circuit Alterations Procedure.
C remain running because it is needed for boron injection lAW LEP-03, Alternate Boron Injection.
D. be terminated and prevented to reduce level to 90 inches lAW LPC.
D. be terminated and prevented to reduce level to 90 inches lAW LPC.
: 79. Given the following ATWS conditions on Unit Two:
2A CRD Pump 2B CRD Pump HPCI System SLC RCIC Suppression Pool Level Reactor Power Reactor Water Level Overcurrent trip Shaft uncoupled Under Clearance Both squib valves failed to fire Running with an unisolable steam supply leak
-24 inches 10%
160 inches Which one of the following identifies the action that would be taken concerning the RCIC system based on the conditions above?
The RCIC system would:
A. be isolated to secure the source of the steam leak lAW OAOP-05.0, Radioactive Spills, High Radiation, and Airborne Activity.
B. have the high suppression pool water level transfer defeated and its suction transferred back to the CST lAW SEP-1 0, Circuit Alterations Procedure.
C~ remain running because it is needed for boron injection lAW LEP-03, Alternate Boron Injection.
D. be terminated and prevented to reduce level to 90 inches lAW LPC.


Feedback Feedback K/A: 217000 KIA:  217000 G2.04.08 G2.04.08 Knowledge of Knowledge        how abnormal of how   abnormal operating operating procedures procedures are are used used in in conjunction conjunction with with EOPs.
Feedback K/A: 217000 G2.04.08 Knowledge of how abnormal operating procedures are used in conjunction with EOPs.
EOPs.
Reactor Core Isolation Cooling System (RCIC)
Reactor Core Reactor  Core Isolation Isolation Cooling  System (RCIC)
Cooling System     (RCIC)
(CFR: 41.10/43.5/45.13)
(CFR: 41.10/43.5/45.13)
(CFR:  41.10/43.5/45.13)
ROISRO Rating: 3.8/4.5 Objective: CLS-LP-300-J Obj 4 Given plant conditions and a copy of the LEPs, determine which method of alternate boron injection is appropriate.
ROISRO Rating:
RO/SRO    Rating: 3.8/4.5 3.8/4.5 Objective: CLS-LP-300-J Objective:  CLS-LP-300-J Obj Obj 44 Given plant Given        conditions and plant conditions  and aa copy copy of of the the LEPs, LEPs, determine determine which which method method ofof alternate alternate boron boron injection injection isis appropriate.
appropriate.


==Reference:==
==Reference:==
AOP-50/SCCP/LEP-03/LPC Cog Level high Explanation: EOP action that supercedes the AOP action is what the question is asking.
AOP-5 does have a step to isolate the system that is leaking, but SCCP overrides that if the system is required by EOPs. With the ATWS the RCIC system is required for alternate boron injection. The Suppression Pool level is high but this will only transfer the HPCI suction valves as RCIC only transfers on CST level. LEP-03 would first want to use CRD then RCIC as long as suction is from CST.
Distractor Analysis:
Choice A: Plausible because AOP-5 does have a step to isolate the system that is leaking, but SCCP overrides that if the system is required by EOPs.
Choice B: Plausible because the high suppression pool level would transfer HPCI and SEP-10 has a section for transferring the suction to the CST from the Suppression Pool.
Choice C: Correct answer, see explanation Choice D: Plausible becasue LPC does have a step for terminating and preventing but this does not address RCIC.
SRO Basis: 10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Feedback KIA: 217000 G2.04.08 Knowledge of how abnormal operating procedures are used in conjunction with EOPs.
Reactor Core Isolation Cooling System (RCIC)
(CFR: 41.10/43.5/45.13)
RO/SRO Rating: 3.8/4.5 Objective: CLS-LP-300-J Obj 4 Given plant conditions and a copy of the LEPs, determine which method of alternate boron injection is appropriate.


==Reference:==
==Reference:==
 
AOP-50 1 SCCP 1 LEP-03 1 LPC Cog Level high Explanation: EOP action that supercedes the AOP action is what the question is asking.
AOP-50/SCCP/LEP-03/LPC AOP-50 1 SCCP 1 LEP-03 1 LPC Cog Level Cog  Level high high Explanation: EOP Explanation:  EOP action that supercedes supercedes the AOP AOP action is is what the question is asking.
AOP-5 does have a step to isolate the system that is leaking, but SCCP overrides that if the system is required by EOPs. With the A TWS the RCIC system is required for alternate boron injection. The Suppression Pool level is high but this will only transfer the HPCI suction valves as RCIC only transfers on CST level. LEP-03 would first want to use CRD then RCIC as long as suction is from CST.
AOP-5 does have a step to isolate the system that is leaking, but SCCP overrides that if the system is required by EOPs. WithWith the AATWS TWS the RCIC system is required for alternate boron injection. The Suppression Pool level is high but this will only transfer the HPCI suction valves as RCIC only transfers on CST level. LEP-03 would first want to use CRD then RCIC as long as suction is from CST.
Distractor Analysis:
Distractor Analysis:
Choice A: Plausible because AOP-5 does have a step to isolate the system that is leaking, but SCCP overrides that if the system is required by EOPs.
Choice A: Plausible because AOP-5 does have a step to isolate the system that is leaking, but SCCP overrides that if the system is required by EOPs.
Choice B: Plausible because the high suppression pool level would transfer HPCI and SEP-10 has a section for transferring the suction to the CST from the Suppression Pool.
Choice B: Plausible because the high suppression pool level would transfer HPCI and SEP-10 has a section for transferring the suction to the CST from the Suppression Pool.
Choice C: Correct answer, see explanation Choice D: Plausible becasue LPC does have a step for terminating and preventing but this does not address RCIC.
Choice C: Correct answer, see explanation Choice D: Plausible becasue LPC does have a step for terminating and preventing but this does not address RCIC.
SRO Basis: 10  10 CFR 55.43(b)-5, 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency emergency situations.
SRO Basis: 10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.  


Notes Notes Actions from Actions      from AOP-05.0 AOP-05.O 1.
Notes Actions from AOP-05.O 1.
: 1.          INITIATE aa search INITIATE coolant or coolant search to or steam steam leak.
INITIATE a search to locate and isolate the source of any fl coolant or steam leak.
to locate leak.
2.
locate and and isolate isolate the the source source of  of any any    ofl 2.
IF radiography is in progress, AND personnel are in D
: 2.           IF radiography IF  radiography isis in danger of danger      of abnormal in progress, progress, AND abnormal exposure, exposure, THEN AND personnel personnel are THEN SECURE SECURE are in in            oD radiography.
danger of abnormal exposure, THEN SECURE radiography.
radiography.
3.
3.
: 3.          ENSURE all ENSURE dosimetry and dosimetry all personnel personnel in and report in the report unusual the area area monitor unusual exposure monitor their exposure to their to E&RC.
ENSURE all personnel in the area monitor their dosimetry and report unusual exposure to E&RC.
E&RC.
OAOP-05.C Rev. 23 Page 6 of 10 From SCCP has the actions to leave the system running:
o OAOP-05.C IOAOP-05.0                                                        Rev. 23 Rev. 23                                    Page Page 66 of"10 of 10 I From SCCP From      SCCP has   has thethe actions actions to to leave  the system leave the   system running:
ISOLATE ALL SYSTEMS DISCHARGING INTO THE AREA EXCEPT SYSTEMS REQUIRED:
running:
* TO BE OPERATED BYAN EMERGENCY OPERATING PROCEDURE
ISOLATE ALL ISOLATE          ALL SYSTEMS SYSTEMS DISCHARGING INTO DISCHARGING                INTO THE AREA EXCEPT  EXCEPT SYSTEMS  SYSTEMS REQUIRED:
* FOR DAMAGE CONTROL I
    ** TOEMERGENCY BE OPERATED BYAN TO BE OPERATED BY AN EMERGENCY OPERATING      OPERATING PROCEDURE
SCCP-14 From LEP-03 A
    **             DAMAGE CONTROL FOR DAMAGE FOR                        CONTROL I                 SCCP-14 From LEP-03 A   NOTE:               HPCI/RCIC should be used only it HPGIIRCIC                                        if suction is from the CST.
NOTE:
CST A
HPCI/RCIC should be used only if suction is from the CST A
From LPC, RCIC is not on list to Terminate and prevent (HPCI is):
From LPC, RCIC is not on list to Terminate and prevent (HPCI is):
LOWER REATQR LOWEll;    ReACTOR WATER WATER LEVEL LEVEL J!UEsrECTIvE IRlIE$PEC11VE OF  OF ANY REACTOR POWER OR REACTOR REACTOR
LOWER REATQR WATER LEVEL J!UEsrECTIvE OF ANY REACTOR POWER OR REACTOR WATER LEVELOSCILLATIONS ry TERMflIAJ1NO AND P4flTMflN tNJFCTIOU FRCM TIlE FOLLO%1NG SYSTE1S UNLESS THE SYSTEM IS SEINO USED IO NJEC1 CORON
      \VATER LEVELOSCILLATIONS WATER    I.EVELOSCILLAllONS ry BY TERMflIAJ1NO TI"~MIIIAliNG AND AND PRlilfliNTING tNJFCTIOU P4flTMflN      IN.I~cnOIJ FRCM FRQM TIlE THE FOLLO%1NG FOLlOVlIHG SYSTE1S SYSTEMS UNLESS UHLESSTHE THE SYSTEM   IS SEINO
* GQNDENSATEFEECWATER
          ~ YSTEM IS BEING USED USElJ IO INJECT CORON TO NJEC1    IIOItOH;
* NPCI
* GQNDENSATEFEECWATER CO.'lDENSATelFESOWATI!R
* RHR
* NPCI IIPCI
* CORE SPRAY
* RHR IIHR
* CORE     SPRAY CORESPRAY
* ALTERNATE COCL?.NT NJTCtIflN YTPM!
* ALTERNATE COCL?.NT NJTCtIflN YTPM!
RC)L- 17 Categories Categories KJA:
RC)L-17 Categories KJA:
KIA:                  217000 217000 G2.04.08 G2.04.08                                  Tier!
217000 G2.04.08 Tier! Group:
Tier / Group:
T2G1 RO Rating:
Group:    T2G1 T2G1 RO Rating:
3.8 SRO Rating:
RORating:              3.8 3.8                                                     SRO   Rating:
4.5 LP Obj:
SRORating:      4.5 4.5 LP Obj:
300J-4 Source:
LPObj:                300J-4 300J-4                                                  Source:
NEW Cog Level:
Source:          NEW NEW Cog Level:
HIGH Category 8:
Cog   Level:           HIGH HIGH                                                    Category Category 8:8:    Y Y
Y Notes Actions from AOP-05.0
: 80. Unit Two
: 1.
: 80. Unit   Two was was operating operating at at rated rated power power with with the  following conditions:
: 2.
the following   conditions:
: 3.
        - AA dual
INITIATE a search to locate and isolate the source of any coolant or steam leak.
          -  dual Unit Unit Loss Loss OfOf Offsite Offsite Power Power (LOOP)
IF radiography is in progress, AND personnel are in danger of abnormal exposure, THEN SECURE radiography.
(LOOP)
ENSURE all personnel in the area monitor their dosimetry and report unusual exposure to E&RC.
Spent  Fuel  Pool  level
o o
        - Spent Fuel Pool level is
o IOAOP-05.0 Rev. 23 Page 6 of"10 I From SCCP has the actions to leave the system running:
          -                                lowering is lowering rapidly rapidly due due toto aa dropped dropped test test weight weight RRCP has
ISOLATE ALL SYSTEMS DISCHARGING INTO THE AREA EXCEPT SYSTEMS REQUIRED:
        - RRCP
* TO BE OPERATED BY AN EMERGENCY OPERATING PROCEDURE
          -        has been      entered due been entered       due to to high high rad rad conditions conditions on on the the refuel refuel floor floor Which one Which      one of of the the following following isis the  first makeup the first makeup source source to to be be used used forfor filling filling the the fuel fuel pool pool and    identifies the and identifies     the procedure procedure to     perform the to perform      the action?
* FOR DAMAGE CONTROL From LEP-03 NOTE:
action?
HPGIIRCIC should be used only it suction is from the CST.
A. Emergency Diesel A. Emergency          Diesel Makeup Makeup Pump Pump via via hoses hoses lAW lAW OAOP-38.0, OAOP-38.0, Loss Loss of of Fuel Fuel Pool Pool Cooling Cooling B RHR B!'"  RHR BB Loop Loop via Fuel Fuel Pool Pool Cooling Cooling System System lAW lAW OAOP-38.0, OAOP-38.0, Loss Loss of  of Fuel Fuel Pool Pool Cooling Cooling C. Emergency Diesel   Diesel Makeup Makeup Pump via hoses lAW        lAW OEDMG-002, OEDMG-002, Spent Fuel     Fuel Pool Pool Makeup/Spray and Makeup/Spray        and Refuel FloorFloor Enhanced Enhanced Ventilation under Conditions of Extreme Damage Damage D. RHR B Loop via Fuel Pool Cooling System lAW OEDMG-002,            OEDMG-002, Spent Fuel Pool Makeup/Spra Makeup/Spray     y  and Refuel     Floor Enhanced Ventilation under Conditions of Extreme Damage
From LPC, RCIC is not on list to Terminate and prevent (HPCI is):
LOWEll; ReACTOR WATER LEVEL IRlIE$PEC11VE OF ANY REACTOR POWER OR REACTOR
\\VATER I.EVELOSCILLAllONS BY TI"~MIIIAliNG AND PRlilfliNTING IN.I~cnOIJ FRQM THE FOLlOVlIHG SYSTEMS UHLESSTHE
~ YSTEM IS BEING USElJ TO INJECT IIOItOH;
* CO.'lDENSATelFESOWATI!R
* IIPCI
*' IIHR
* CORESPRAY Categories KIA:
217000 G2.04.08 RORating:
 
===3.8 LPObj===
300J-4 Cog Level:
HIGH Tier / Group:
SRORating:
Source:
Category 8:
T2G1 4.5 NEW Y
: 80. Unit Two was operating at rated power with the following conditions:
- A dual Unit Loss Of Offsite Power (LOOP)
- Spent Fuel Pool level is lowering rapidly due to a dropped test weight
- RRCP has been entered due to high rad conditions on the refuel floor Which one of the following is the first makeup source to be used for filling the fuel pool and identifies the procedure to perform the action?
A. Emergency Diesel Makeup Pump via hoses lAW OAOP-38.0, Loss of Fuel Pool Cooling B RHR B Loop via Fuel Pool Cooling System lAW OAOP-38.0, Loss of Fuel Pool Cooling C. Emergency Diesel Makeup Pump via hoses lAW OEDMG-002, Spent Fuel Pool Makeup/Spray and Refuel Floor Enhanced Ventilation under Conditions of Extreme Damage D. RHR B Loop via Fuel Pool Cooling System lAW OEDMG-002, Spent Fuel Pool Makeup/Spray and Refuel Floor Enhanced Ventilation under Conditions of Extreme Damage
: 80. Unit Two was operating at rated power with the following conditions:
- A dual Unit Loss Of Offsite Power (LOOP)
- Spent Fuel Pool level is lowering rapidly due to a dropped test weight
- RRCP has been entered due to high rad conditions on the refuel floor Which one of the following is the first makeup source to be used for filling the fuel pool and identifies the procedure to perform the action?
A. Emergency Diesel Makeup Pump via hoses lAW OAOP-38.0, Loss of Fuel Pool Cooling B!'" RHR B Loop via Fuel Pool Cooling System lAW OAOP-38.0, Loss of Fuel Pool Cooling C. Emergency Diesel Makeup Pump via hoses lAW OEDMG-002, Spent Fuel Pool Makeup/Spray and Refuel Floor Enhanced Ventilation under Conditions of Extreme Damage D. RHR B Loop via Fuel Pool Cooling System lAW OEDMG-002, Spent Fuel Pool Makeup/Spray and Refuel Floor Enhanced Ventilation under Conditions of Extreme Damage  


Feedback Feedback K/A: 233000 KIA:  233000 G2.04.06 G2.04.06 Knowledge of Knowledge                mitigation strategies.
Feedback K/A: 233000 G2.04.06 Knowledge of EOP mitigation strategies.
EOP mitigation of EOP                strategies.
Fuel Pool Cooling and Clean-up (CFR: 41.10/43.5/45.13)
Fuel Pool Fuel  Pool Cooling Cooling and and Clean-up Clean-up (CFR: 41.10/43.5/45.13)
There are no direct EOP actions associated with FPC, a loss of level in the fuel pool will cause entty into RRCP which is an EOP. So these actions are mitigation strategies to RRCP.
(CFR:   41.10/43.5/45.13)
RO/SRO Rating: 3.7/4.7 Objective:
There are There  are no no direct direct EOP EOP actions actions associated associated with FPC, aa loss with FPC,      loss of oflevel level in in the the fuel fuel pool pool will will cause cause entry entty into into RRCP which RRCP    which is is an an EOP.
CLS-LP-1 3, Obj. 11. State the sources of makeup water for the Fuel Pool in order of preference.
EOP. So    these actions So these actions are are mitigation mitigation strategies strategies to    RRCP.
to RRCP.
RO/SRO Rating:
RO/SRO      Rating: 3.7/4.7 3.7/4.7 Objective:
Objective:
CLS-LP-1 3, Obj.
CLS-LP-13,    Obj. 11.
: 11. State State the the sources sources of of makeup makeup water water for for the the Fuel Fuel Pool Pool inin order order of of preference.
preference.


==Reference:==
==Reference:==
OAOP-38.0 Loss of Fuel Pool Cooling Cog Level High Explanation:
The order of the makeup sources is from the normal fill, Demin water hose stations, Fire protection hose stations, demin water through RHR keepfill, and then other sources that are not service water. For a high capacity water source and the gates installed RHR Loop B would be used via the FPC system. With a LOOP the demin pumps have no power. If no other sources are availble then the proceudre has injection from the EDMP.
Distractor Analysis:
Choice A: Plausible because although this is a makeup source it is not the preferred source (last resort per the procedure) and is performed per the EDMG procedures. Although upon entering the AOP there is a step to start lining this system up for injection because of the time required to get all of the hoses run in the procedure up to the fuel pool.
Choice B: Correct answer see explanation Choice C: Plausible because although this is a makeup source it is not the preferred source (last resort per the procedure). Although upon entering the AOP there is a step to start lining this system up for injection because of the time required to get all of the hoses run in the procedure up to the fuel pool.
Choice D: Plausible because RHR is the high capacity source that will need to be used, but the EDMG procedure does not provide this guidance.
SRO Basis: 10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Feedback KIA: 233000 G2.04.06 Knowledge of EOP mitigation strategies.
Fuel Pool Cooling and Clean-up (CFR: 41.10/43.5/45.13)
There are no direct EOP actions associated with FPC, a loss of level in the fuel pool will cause entry into RRCP which is an EOP. So these actions are mitigation strategies to RRCP.
RO/SRO Rating: 3.7/4.7 Objective:
CLS-LP-13, Obj. 11. State the sources of makeup water for the Fuel Pool in order of preference.


==Reference:==
==Reference:==
 
OAOP-38.0 Loss of Fuel Pool Cooling Cog Level High Explanation:
OAOP-38.0 OAOP-38.0 LossLoss of of Fuel Fuel Pool Pool Cooling Cooling Cog Level Cog  Level High High Explanation:
The order of the makeup sources is from the normal fill, Demin water hose stations, Fire protection hose stations, demin water through RHR keepfill, and then other sources that are not service water. For a high capacity water source and the gates installed RHR Loop B would be used via the FPC system. With a LOOP the demin pumps have no power. If no other sources are availble then the proceudre has injection from the EDMP.
Explanation:
sources is from the normal fill, Demin water hose stations, The order of the makeup sources                                                          stations, Fire protection hose stations, demin water through RHR keepfill, and then other sources that are not service water. For a high capacity water source and the gates installed RHR Loop B would be used via the FPC system. With a LOOP the demin pumps have no power. If no other sources are availble then the proceudre has injection from the EDMP.
Distractor Analysis:
Distractor Analysis:
Choice A: Plausible because although this is a makeup source it is not the preferred source (last resort per the procedure) and is performed per the EDMG procedures. Although upon entering the AOP there is a step to start lining this system up for injection because of the time required to get all of the hoses run in the procedure up to the fuel pool.
Choice A: Plausible because although this is a makeup source it is not the preferred source (last resort per the procedure) and is performed per the EDMG procedures. Although upon entering the AOP there is a step to start lining this system up for injection because of the time required to get all of the hoses run in the procedure up to the fuel pool.
Choice B: Correct answer see explanation Choice C: Plausible because although this is a makeup source it is not the preferred source (last resort per the procedure). Although upon entering the AOP there is a step to start lining this system up for injection because of the time required to get get all of the hoses hoses run in the procedure procedure up to the fuel pool.
Choice B: Correct answer see explanation Choice C: Plausible because although this is a makeup source it is not the preferred source (last resort per the procedure). Although upon entering the AOP there is a step to start lining this system up for injection because of the time required to get all of the hoses run in the procedure up to the fuel pool.
Choice D:0: Plausible because RHR is the high capacity source that will need to be used, but the EDMG procedure procedure does does not provide this guidance.
Choice 0: Plausible because RHR is the high capacity source that will need to be used, but the EDMG procedure does not provide this guidance.
SRO SRO Basis:    10 CFR Basis: 10   CFR 55.43(b)-5, 55.43(b)-5, Assessment of  of facility conditions and and selection selection of of appropriate procedures during during normal, normal, abnormal, abnormal, and and emergency situations.
SRO Basis: 10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.  
situations.


Notes Notes 2.
Notes 2.
: 2.        IMMEDIATELY ENTER IMMEDIATELY         ENTER OEDMG-002, OEDMG-002, Spent  Spent Fuel Fuel Pool Pool Makeup/Spray and Makeup/Spray       and Enhanced Enhanced Refuel Refuel Floor Floor Ventilation Ventilation D
IMMEDIATELY ENTER OEDMG-002, Spent Fuel Pool Makeup/Spray and Enhanced Refuel Floor Ventilation Under Conditions of Extreme Damage, AND make preparations to makeup to the fuel pool using the EDMP.
LI Under Conditions Under    Conditions of of Extreme Extreme Damage, Damage, AND AND make make preparations to preparations    to makeup makeup to to the the fuel fuel pool pool using using the the EDfvIP.
3.2.12 IF a high capacity makeup source of water through the RI-IR System is required to maintain fuel pool level AND the fuel pool gates are installed, THEN PERFORM the following:
EDMP.
CONFIRM one of the following ow paths available for use with the Fuel Pool Cooling System:
3.2.12 3.2:12        IF aa high IF     high capacity capacity makeup makeup source source of of water water through through the the RI-IR System RHR      System isis required required toto maintain maintain fuelfuel pool pool level level AND AND tt1e  fuel pool the fuel       gates are pool gates     are installed, installed, THEN      PERFORM tile THEN PERFORM           the following:
RHR Loop B only (RHR Loop B Shutdown Cooling must be secured)
following:
RHR Loop A through RHR Loop Cross-Tie to the RHR Loop B discharge. (Both RHR Loop A and Loop B Shutdown Cooling must be secured).
: 1.      CONFIRM one CONFIRM             of the one of   the following flow ow paths paths available available for for use with the Fuel use                    Pool Cooling Fuel Pool    Cooling System:
QAOP-38.O Rev. 22 Page 11 of 35 Actions for Emergency Diesel Makeup Pump:
RHR Loop B     only (RHR B only  (RHR LoopLoop 6B Shutdown Shutdown Cooling Cooling        D must be be secured)
3.2.19 IF no actions have been successful, THEN ENTER LI OEDMG-OD2, Spent Fuel Pool Makeup/Spray and Refuel Floor Enhanced Ventilation under Conditions of Extreme Damage.
RHR Loop A through RHR Loop Cross-Tie to the                           D Li RHR Loop 6B discharge.
discharge. (60th (Both RHR Loop  Loop A and LoopLoop 6B Shutdown Cooling must be secured).
QAOP-38.O IOAOP-38.0                                        Rev. 22                                   Page 1-1 11 of 351 35 Actions for Emergency Diesel Makeup Pump:
3.2.19 3.2:19      IF no actions have been successful, THEN ENTER                               D LI OEDMG-OD2, Spent Fuel Pool Makeup/Spray OEDMG-002,                            Makeup/Spray and Refuel Floor Enhanced Ventilation under Conditions of Extreme Damage.
From EMG-002:
From EMG-002:
3.3     Normal fuel pool makeup methods and the B.5.b          6.S.b requirement for aa diverse internal strategy (using installed plant equipment) eqUipment) are contained in OAOP-38.O, OAOP-38.0, Loss of Fuel Pool Cooling. OEDMG-002 is entered when the methods contained in    in OAOP-38.O OAOP-38.0 have proven to be inadequate or cannot be performed.
3.3 Normal fuel pool makeup methods and the B.5.b requirement for a diverse internal strategy (using installed plant equipment) are contained in OAOP-38.O, Loss of Fuel Pool Cooling. OEDMG-002 is entered when the methods contained in OAOP-38.O have proven to be inadequate or cannot be performed.
Categories Categories K/A:
Categories K/A:
KIA:          233000 233000 G2.04.06 G2.04.06                              Tier Tier / Group:
233000 G2.04.06 Tier / Group:
Group: T2G2 T2G2 RO Rating:
T2G2 RO Rating:
RORating:      3.7 3.7                                             SRO   Rating:
3.7 SRO Rating:
SRORating:      4.7


===4.7 LPObj===
===4.7 LPObj===
LPObj:          13-11 13-11                                           Source:
13-11 Source:
Source:          NEW NEW Cog Level:
NEW Cog Level:
Cog  Level:    HIGH HIGH                                            Category Category 8:8:
HIGH Category 8:
: 81. With Unit
LI Li Notes
: 81. With       Unit Two Two at at rated rated power, power, which which one one of  of the the following following identifies:
: 2.
identifie (1) the (1)        required number the reql:lired    number of    of operable operable SRVs SRVs for  for safety safety function function lAW  lAW Technical Technical Specification      3.4.3,  Safety/Relief Specification 3.4.3, Safety/Relief Valves and    Valves        and (2) the (2)    the bases bases for for this this number number of  of operable operable SRVs?  SRVs?
3.2:12
A. (1)
: 1.
A.      (1)99 (2) prevent overpressurization (2) prevent      overpressurization associated associated with   with anan ATWS ATWS event event B (1)
IMMEDIATELY ENTER OEDMG-002, Spent Fuel Pool Makeup/Spray and Enhanced Refuel Floor Ventilation Under Conditions of Extreme Damage, AND make preparations to makeup to the fuel pool using the EDfvIP.
B!'"  (1) 10 10 (2) prevent (2)  prevent overpressurization overpressurization associated associated with   with anan ATWS ATWS event  event C. (1)
IF a high capacity makeup source of water through the RHR System is required to maintain fuel pool level AND tt1e fuel pool gates are installed, THEN PERFORM tile following:
C.    (1)9 9 (2)  prevent overpressurization (2) prevent       overpressurization associated associated with     with anan MSIV MSIV closure closure D. (1)
CONFIRM one of the following flow paths available for use with the Fuel Pool Cooling System:
D.    (1)1010 (2)  prevent overpressurization (2) prevent       overpressurization associated associated with    with an an MSIV MSIV closure closure Feedback Feedback K/A: 239002 G2.02.25 KIA:              G2.02.25 Knowledge of the bases in Technical Specifications Specifications for limiting conditions for operations and safety limits.
D RHR Loop B only (RHR Loop 6 Shutdown Cooling D
Safety Relief Valves 41.5 / 41.7/43.2)
must be secured)
(CFR: 41.5/41.7/43.2)
RHR Loop A through RHR Loop Cross-Tie to the D
ROISRO Rating: 3.2/4.2 RO/SRO Objective: CLS-LP-25, Obj. 10 Given plant conditions and TS, including the Bases, TRM, ODCM, and COLR determine the required actions to be taken in accordance with TS associated with the Reactor Recirculation System. (SRO only)
RHR Loop 6 discharge. (60th RHR Loop A and Loop 6 Shutdown Cooling must be secured).
IOAOP-38.0 Rev. 22 Page 1-1 of 351 Actions for Emergency Diesel Makeup Pump:
3.2:19 From EMG-002:
IF no actions have been successful, THEN ENTER OEDMG-002, Spent Fuel Pool Makeup/Spray and Refuel Floor Enhanced Ventilation under Conditions of Extreme Damage.
D 3.3 Normal fuel pool makeup methods and the 6.S.b requirement for a diverse internal strategy (using installed plant eqUipment) are contained in OAOP-38.0, Loss of Fuel Pool Cooling. OEDMG-002 is entered when the methods contained in OAOP-38.0 have proven to be inadequate or cannot be performed.
Categories KIA:
233000 G2.04.06 Tier / Group: T2G2 RORating:  
 
===3.7 SRORating===
4.7 LPObj:
13-11 Source:
NEW Cog Level:
HIGH Category 8:
: 81. With Unit Two at rated power, which one of the following identifie (1) the required number of operable SRVs for safety function lAW Technical Specification 3.4.3, Safety/Relief Valves and (2) the bases for this number of operable SRVs?
A. (1)9 (2) prevent overpressurization associated with an ATWS event B (1) 10 (2) prevent overpressurization associated with an ATWS event C. (1)9 (2) prevent overpressurization associated with an MSIV closure D. (1)10 (2) prevent overpressurization associated with an MSIV closure Feedback K/A: 239002 G2.02.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.
Safety Relief Valves (CFR: 41.5 / 41.7/43.2)
ROISRO Rating: 3.2/4.2 Objective: CLS-LP-25, Obj. 10 Given plant conditions and TS, including the Bases, TRM, ODCM, and COLR determine the required actions to be taken in accordance with TS associated with the Reactor Recirculation System. (SRO only)


==Reference:==
==Reference:==
TS 3.4.3 and bases document Cog Level Low Explanation:
TS 3.4.3 states 10 must be operational for the safety function, the bases states the reason, ATWS.
Distractor Analysis:
Choice A: Plausible because the bases states that 9 are required for the MSIV closure.
Choice B: Correct answer, see explanation Choice C: Plausible because the bases states that 9 are required for the MSIV closure and the MSIV closure is not the binding failure mode.
Choice D: Plausible because 10 are required for the ATWS and the MSIV closure is not the binding failure mode.
SRO Basis: 10 CFR 55.43(b)-2, Facility operating limitations in the technical specifications and their bases. This is knowledge of tech spec bases to determine the reason 10 are required.
: 81. With Unit Two at rated power, which one of the following identifies:
(1) the reql:lired number of operable SRVs for safety function lAW Technical Specification 3.4.3, Safety/Relief Valves and (2) the bases for this number of operable SRVs?
A. (1) 9 (2) prevent overpressurization associated with an ATWS event B!'" (1) 10 (2) prevent overpressurization associated with an ATWS event C. (1) 9 (2) prevent overpressurization associated with an MSIV closure D. (1) 10 (2) prevent overpressurization associated with an MSIV closure Feedback KIA: 239002 G2.02.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.
Safety Relief Valves (CFR: 41.5/41.7/43.2)
RO/SRO Rating: 3.2/4.2 Objective: CLS-LP-25, Obj. 10 Given plant conditions and TS, including the Bases, TRM, ODCM, and COLR determine the required actions to be taken in accordance with TS associated with the Reactor Recirculation System. (SRO only)


==Reference:==
==Reference:==
TS 3.4.3 and bases document Cog Level Low Explanation:
TS 3.4.3 states 10 must be operational for the safety function, the bases states the reason, ATWS.
Distractor Analysis:
Choice A: Plausible because the bases states that 9 are required for the MSIV closure.
Choice B: Correct answer, see explanation Choice C: Plausible because the bases states that 9 are required for the MSIV closure and the MSIV closure is not the binding failure mode.
Choice D: Plausible because 10 are required for the ATWS and the MSIV closure is not the binding failure mode.
SRO Basis: 10 CFR 55.43(b)-2, Facility operating limitations in the technical speCifications and their bases. This is knowledge of tech spec bases to determine the reason 10 are required.


TS 3.4.3 3.4.3 and bases document Cog Level Low Low Explanation:
Notes 3.4.3 Safetv?Rehef Valves (SRVs)
Explanation:
LCO 3.4.3 The safety function of 10 SRVs shall be OPERASLE.
TS TS 3.4.3 3.4.3 states states 10 10 must must be be operational for the    safety function, the the safety                the bases bases states states the the reason, reason, ATWS.
ATWS.
Distractor Distractor Analysis:
Analysis:
Choice Choice A:  A: Plausible Plausible because because thethe bases bases states states that      are required that 99 are    required for for the the MSIV MSIV closure.
closure.
Choice Choice B:  B: Correct Correct answer,    see explanation answer, see    explanation Choice Choice C:  C: Plausible Plausible because because thethe bases bases states states that      are required that 99 are    required forfor the the MSIV MSIV closure closure andand the the MSIV MSIV closure      not the closure isis not  the binding binding failure failure mode.
mode.
Choice Choice D:  D: Plausible Plausible because because 10 10 are are required required forfor the the ATWS ATWS and and thethe MSIV MSIV closure closure isis not not the the binding binding failure failure mode.
mode.
SRO Basis:
SRO              10 CFR Basis: 10  CFR 55.43(b)-2, 55.43(b)-2, Facility Facility operating operating limitations          the technical limitations inin the  technical specification speCifications    and their s and  their bases.
bases. This This isis knowledge knowledge of oftech tech spec spec bases bases to to determine determine the  the reason reason 10  10 are are required.
required.
 
Notes Notes 3.4.3 Safety/Relief 3A.3    Safetv?RehefValves Valves (SRVs)
(SRVs)
LCO 3.4.3 LCO    3.4.3          The safety The safety function   of '10 function of  10 SRVs SRVs shall shall be be OPERABLE.
OPERASLE.
APPUCABILITY:
APPUCABILITY:
APPLICABILITY:        MODES 1,1. 2, MODES        2. and  3.
MODES 1. 2. and 3.
and 3.
From the Bases document:
From the From  the Bases Bases document:
APPLICABLE The overpressure protection system must accommodate the most SAFETY ANALYSES severe pressunzattan transient EvaIatioi aac1eterrnined that the rnoLseere4ransient icioit1flar&fres (MP7 fIod by readorscrt n
document:
1f1hr{ie fiIure of the direct scrani i+/-tdith MSIV position) (Ref. 1 For the purpose of the analyses, 9 SRVs are assumed to operate in the safety mode. The analysis results demonstrate that the design SRV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure (110% x 1250 psig 1375 psig). This LCO helps to ensure that the acceptance limit of 1375 psig is met during the Design Basis Event.
APPLICABLE APPLICABLE              The overpressure The  overpressure protection protection system system must must accommodate accommodate the   the most most SAFETY ANAL SAFETY    ANALYSES     severe pressurization YSES severe     pressunzattan transient.
(continued APPLICABLE For overpressurization associated with an ATWS event, 10 SRVs are SAFETY ANALYSES assumed to operate in the safety mode. The analysis (Ref. 2)
transient EvaIatioi aac1eterrnined that the rnoLseere4ransient icioit1flar&fres (MP7 fIod by readorscrt n                                   1f1hr{ie fiIure of   of the the direct scrani i+/-tdith MSIV position) (Ref. 1 For the purpose direct                                                                      purpose ofof the analyses;*
(continued) results demonstrate that the design capacity is capable of maintaining reactor pressure below the ASME Section Ill Code Serice Level C limits (1500 psig).
the              9 SRVs are assumed,tooperate.in,the analyses, 9SR\lsare        assumed to operate in the safety         mode. The safety mode. The analysis results results demonstrate demonstrate that the  the design design SRV SRV capacity capacity isis capable capable ofof maintaining reactor maintaining   reactor pressure pressure below below thethe ASME CodeCode limit limit of of 110%
From an overpressure standpoint, the design basis events are bounded by the overpressurization associated with the ATWS event described above. Reference 3 discusses additional events that are expected to actuate the SRVs.
110% of vessel design   pressure P'l design pressure      (110%
SRVs satisfy Criteiion 3 of 10 CER 50.36(cX2)(ii) (Ref. 4).
0% xx '1250   psig =1375 1250 psig      1375 psig).
Categories KJA:
psig). This This LCO LCO helps helps to ensure that the acceptance acceptance limit limit of of '1375 1375 psig is is met met during during the Design Design Basis Event.
239002 G2.02.25 Tier / Group:
Basis   Event.
T2G1 RO Rating:
3.2 SRO Rating:
4.2 LP Obj:
25-10 Source:
NEW Cog Level:
LOW Category 8:
Y Notes 3A.3 Safety/Relief Valves (SRVs)
LCO 3.4.3 The safety function of '10 SRVs shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3.
From the Bases document:
APPLICABLE The overpressure protection system must accommodate the most SAFETY ANAL YSES severe pressurization transient.
of the direct purpose of the analyses;* 9SR\\lsare assumed,tooperate.in,the safety mode. The analysis results demonstrate that the design SRV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure P'l 0% x '1250 psig = 1375 psig). This LCO helps to ensure that the acceptance limit of '1375 psig is met during the Design Basis Event.
(continued)
(continued)
(continued APPLICABLE           For overpressurization associated with an A         ATWS    event, 10 SRVs are nNS event.
APPLICABLE For overpressurization associated with an A nNS event. 10 SRVs are SAFETY ANALYSES assumed to operate in the safety mode. The analysis (Ref. 2)
SAFETY ANALYSES       assumed to operate in the safety mode. The analysis (Ref. 2)
(continued}
(continued)
results demonstrate that the design capacity is capable of maintaining reactor pressure below the ASME Section III Code Service Level C limits (1500 psig).
(continued}          results demonstrate that the design capacity is capable of maintaining reactor pressure below the ASME Section III         Ill Code Service Serice Level C limits (1500 psig).
Categories KIA:
From an overpressure standpoint, the design basis events are bounded by the overpressurization associated with the AnNS       ATWS event described above. Reference 3 discusses additional events that are expected to actuate the SRVs.
RORating:
Criteiion 33 of '10 SRVs satisfy Criterion           10 CFR CER 50.36(c){2)(ii) 50.36(cX2)(ii) (Ref. 4).
LP Obj:
Categories KJA:
Cog Level:
KIA:          239002 G2.02.25                                 Tier / Group: T2G1 RO  Rating:
From an overpressure standpoint, the design basis events are bounded by the overpressurization associated with the AnNS event described above. Reference 3 discusses additional events that are expected to actuate the SRVs.
RORating:      3.2
SRVs satisfy Criterion 3 of '10 CFR 50.36(c){2)(ii) (Ref. 4).
239002 G2.02.25 Tier / Group: T2G1  


===3.2 Rating===
===3.2 SRORating===
SRORating:
4.2 25-10 Source:
SRO              4.2 LP Obj:        25-10                                           Source:
NEW LOW Category 8:
Source:          NEW Cog Cog Level:    LOW LOW                                              Category Category 8:8:    Y
Y
: 82. Unit
: 82. Unit One is operating at full power when the Main Stack Rad Monitor lost its norm power supply.
: 82. Unit One One isis operating operating at       full power at full  power when when the  the Main Main Stack Stack Rad Rad Monitor Monitor lost lost its its normal norm power supply.
Which one of the following identifies the procedure that contains the steps to transfer the Main Stack Rad Monitor to its alternate power supply?
power    supply.
A. IOP-52, 120 Volt AC UPS, Emergency, and Conventional Electrical Systems Operating Procedure B 20P-52, 120 Volt AC UPS, Emergency, and Conventional Electrical Systems Operating Procedure C. 1APP UA-03 6-3, PROCESS SMPL OG VENT PIPE DNSC/INOP D. 2APP UA-03 6-3, PROCESS SMPL OG VENT PIPE DNSC/INOP Feedback K/A: 262002 G2.0l.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.
Which one Which    one of of the the following following identifies identifies the the procedure procedure that  that contains contains thethe steps steps to to transfer transfer the Main the  Main Stack Stack Rad      Monitor to Rad Monitor             its alternate to its   alternate powerpower supply?
Uninterruptable Power Supply (A.C.ID.C.)
supply?
(CFR: 41.10/43.5145.2/45.6)
A. IOP-52, 120 A. 10P-52,        120 Volt Volt ACAC UPS, UPS, Emergency, Emergency, and     and Conventional Conventional Electrical Electrical Systems Systems Operating Procedure Operating      Procedure B
ROISRO Rating: 4.3/4.4 Objective: CLS-LP-11.0, 15a Given plant conditions and a trip or failure of one of the following Radiation Monitors, determine appropriate plant response and use procedures to determine the actions required to control and/or mitigate the consequences of the event:
B~  20P-52, 120 20P-52,     120 Volt Volt ACAC UPS, UPS, Emergency, Emergency, and      and Conventional Conventional Electrical Electrical Systems Systems Operating Procedure Operating      Procedure C. 1APP UA-03 C. 1APP       UA-03 6-3,        PROCESS SMPL 6-3, PROCESS             SMPL OG   OG VENTVENT PIPE PIPE DNSCIINOP DNSC/INOP D. 2APP UA-03 D. 2APP      UA-03 6-3,        PROCESS SMPL 6-3, PROCESS             SMPL OG   OG VENTVENT PIPE PIPE DNSCIINOP DNSC/INOP Feedback Feedback K/A: 262002 KIA:  262002 G2.01.23 G2.0l.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.
Uninterruptable Power Supply (A.C'/D.C.)
(A.C.ID.C.)
41.10/43.514 (CFR: 41.10              5.2/45.6) 143.5 1 45.2  145.6)
ROISRO Rating: 4.3/4.4 RO/SRO Objective: CLS-LP-11.0, 15a Given plant conditions and a trip or failure of one of the following Radiation Monitors, determine appropriate plant response and use procedures to determine the actions required to control and/or mitigate the consequences consequences of the event:
: a. Main Stack.
: a. Main Stack.


==Reference:==
==Reference:==
20P-52, Section 8.7, Stack Radiation Monitor UPS Power Supply Transfer Cog Level High Explanation:
The normal power supply for the Main Stack Rad Monitor is from Unit Two. On a loss of power the from the normal power supply the operators will need to transfer to the alternate power supply. This direction is only in the U2 procedure. There is no directions to perform this in the Ui procedure or the APPs for either Unit.
Distractor Analysis:
Choice A: Plausible because the stem states this is Ui but the actions are in the U2 procedure.
Choice B: Correct answer, see explanation.
Choice C: Plausible because the downscale imp annunciator will be actuated on a loss of power but the APP5 do not address transfer of power to backup supply.
Choice D: Plausible because the downscale / mop annunciator will be actuated on a loss of power but the APPs do not address transfer of power to backup supply. U2 is the normal power supply to the rad monitor.
SRO Basis: 10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
: 82. Unit One is operating at full power when the Main Stack Rad Monitor lost its normal power supply.
Which one of the following identifies the procedure that contains the steps to transfer the Main Stack Rad Monitor to its alternate power supply?
A. 10P-52, 120 Volt AC UPS, Emergency, and Conventional Electrical Systems Operating Procedure B~ 20P-52, 120 Volt AC UPS, Emergency, and Conventional Electrical Systems Operating Procedure C. 1APP UA-03 6-3, PROCESS SMPL OG VENT PIPE DNSCIINOP D. 2APP UA-03 6-3, PROCESS SMPL OG VENT PIPE DNSCIINOP Feedback KIA: 262002 G2.01.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.
Uninterruptable Power Supply (A.C'/D.C.)
(CFR: 41.10 143.5 1 45.2 145.6)
RO/SRO Rating: 4.3/4.4 Objective: CLS-LP-11.0, 15a Given plant conditions and a trip or failure of one of the following Radiation Monitors, determine appropriate plant response and use procedures to determine the actions required to control and/or mitigate the consequences of the event:
: a. Main Stack.


==Reference:==
20P-52, Section 8.7, Stack Radiation Monitor UPS Power Supply Transfer Cog Level High Explanation:
20P-52, Section 8.7, Stack Radiation Monitor UPS Power Supply Transfer Cog Level High Explanation:
The normal power supply for the Main    Main Stack Rad MonitorMonitor is from Unit Unit Two. On aa loss of  of power the from the normal normal power supply the operators will need to transfer to the alternate power            power supply. This direction direction isis only only in in the the U2 U2 procedure. ThereThere is   no directions is no  directions to perform this in    in the U1 procedure the Ui  procedure or  or the the APPs APPs for either Unit.
The normal power supply for the Main Stack Rad Monitor is from Unit Two. On a loss of power the from the normal power supply the operators will need to transfer to the alternate power supply. This direction is only in the U2 procedure. There is no directions to perform this in the U1 procedure or the APPs for either Unit.
Unit.
Distractor Analysis:
Distractor Distractor Analysis:
Choice A: Plausible because the stem states this is U1 but the actions are in the U2 procedure.
Analysis:
Choice B: Correct answer, see explanation.
Choice Choice A: A: Plausible Plausible because because the     stem states the stem    states this this is is Ui U1 but but the the actions actions areare in in the the U2 U2 procedure.
Choice C: Plausible because the downscale 1 inop annunciator will be actuated on a loss of power but the APPs do not address transfer of power to backup supply.
procedure.
Choice D: Plausible because the downscale 1 inop annunciator will be actuated on a loss of power but the APPs do not address transfer of power to backup supply. U2 is the normal power supply to the rad monitor.
Choice Choice B: B: Correct Correct answer,     see explanation.
SRO Basis: 10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.  
answer, see    explanation.
Choice Choice C: C: Plausible Plausible because because the     downscale imp the downscale      1 inop annunciator annunciator will     be actuated will be  actuated onon aa loss loss of of power power but but the the APP5 APPs do do not address transfer not address    transfer of of power power toto backup backup supply.
supply.
Choice Choice D: D: Plausible Plausible because     the downscale because the    downscale /1mop  inop annunciator annunciator will will be be actuated actuated onon aa loss loss of of power power but but the the APPs APPs do do not not address address transfer transfer of of power power to   backup supply.
to backup    supply. U2 U2 is is the the normal normal power power supply supply to to the the rad rad monitor.
monitor.
SRO Basis:
SRO    Basis: 10 10 CFR CFR 55.43(b)-5, 55.43(b)-5, Assessment Assessment of   offacility facility conditions conditions andand selection selection of of appropriate appropriate procedures procedures during during normal, normal, abnormal, abnormal, and and emergency emergency situations.
situations.


Notes Notes 8.0 8.0      INFREQUENT OPERATIONS INFREQUENT         OPERATIONS..................................................................................
Notes 8.0 INFREQUENT OPERATIONS.
32 32 8.1 8.1       Transferring UPS Transferring     UPS Loads Loads From  From Alternate Alternate SourceSource to   to PrimalY Primary UPS  UPS Unit Unit 2A.....
32 8.1 Transferring UPS Loads From Alternate Source to Primary UPS Unit 2A 32 8.2 Transferring UPS Loads From Alternate Source to Standby UPS Unit 28 35 8.3 Transferring UPS Loads From Standby UPS Unit 2B to Alternate Source 39 8.4 Transferring UPS Loads From Primary UPS Unit 2A to Alternate Source 41 8.5 Alignment of Standby UPS Unit 2B After a Loss of Alternate Source Power 43 8.6 Returning Standby UPS Unit 2B to Normal Operating Condition Upon Regaining Alternate Source Power Supply 45 8.7 Stack Radiation Monitor UPS Power Supply Transfer 47 20P-52 Rev. 53 Page2of78 8.0 INFREQUENT OPERATIONS 34 8.1 Transferring UPS Loads From Alternate Source to Primary UPS Unit 1A 34 8.2 Transferring UPS Loads From Alternate Source to Standby UPS Unit 18 38 8.3 Transferring UPS Loads From Standby UPS Unit I B to Alternate Source 42 8.4 Transfening UPS Loads From Primary UPS Unit IA to Alternate Source 44 8.5 Alignment of Standby UPS Unit lB After a Loss of Alternate Source Power 46 8.6 Returning Standby UPS Unit 1 B to Normal Operating Condition Upon Regaining Alternate Source Power Supply 48 I OP-52 Rev. 35 Page 3 of 74 Unit2 APP UA-03 6-3 Page 1 of 1 PROCESS SMPL OG VENT PIPE DNSCIINOP (Process Sample Off-Gas Pipe Down-Inoperable)
2A         32 32 8.2 8.2       Transferring UPS Transferring     UPS Loads Loads From  From Alternate Alternate SourceSource to    to Standby Standby UPS   UPS Unit 28 Unit   28   .... 35 35 8.3 8.3        Transferring    UPS    Loads Transferring UPS Loads From Standby From    Standby UPS    UPS Unit Unit 2B2B to to Alternate Alternate SourceSource ....      39 39 8.4 8.4        Transferring UPS Transferring     UPS Loads Loads From  From Primary Primary UPS  UPS Unit Unit 2A2A to to Alternate Alternate SourceSource.....        4'1 41 8.5 8.5        Alignment of Alignment     of Standby Standby UPS    UPS Unit Unit 2B2B After After aa Loss Loss of  of Alternate Alternate SourceSource Power.Power        43 43 8.6 8.6       Returning  Standby      UPS      Unit  2B    to  Normal Returning Standby UPS Unit 2B to Normal Operating Condition      Operating        Condition Upon  Upon Regaining Altemate Regaining    Alternate SourceSource Power Power Supply Supply ............. ..... ........... ...... ........... ....... 45 45 8.7 8.7        Stack Radiation Stack   Radiation Monitor Monitor UPS   UPS PowerPower Supply Supply Transfer.
AUTO ACTIONS NONE CAUSE 1.
Transfer ...................................        47 47 20P-52 120P-52                                                      Rev. 53 Rev. 53                                                Page Page2of782 of 78      I 8.0 8.0      INFREQUENT OPERATIONS INFREQUENT          OPERATIONS .... ............ ................. ...... ................ ........... .............          34 34 8.1 8.1        Transferring UPS Transferring    UPS LoadsLoads From From Alternate Source   Source to    to PrimalY Primary UPS  UPS Unit Unit 1A.....
Off-gas vent pipe (stack) radiation monitor downscale or out of service.
1A         34 34 8.2 8.2      Transferring UPS Loads From Alternate Source to Standby UPS Unit 18                                         18  .... 38 8.3       Transferring UPSUPS Loads From      From Standby UPS Unit 'I1 B to Alternate Source ....                                42 8.4       Transfening UPS Loads From Primary UPS Unit 'lA Transferring                                                                IA to Alternate Source .....                44 8.5       Alignment of Standby UPS Unit 1lB                B After a Loss of Alternate Source Power ................................................................................................................
Power.                                                                                                                  46 8.6       Returning Standby UPS Unit 11 B to Normal Operating Condition Upon Regaining Alternate Source Power Supply.            Supply ........... ...... ..... ......... ................ ..... 48 I OP-52 l'lOP-52                                                    Rev. 35                                                   Page 3 of 74           I Unit22 Unit APP UA-03 6-3 Page 1       of'l1 1 of PROCESS SMPL OG VENT PIPE DNSCllNOP              DNSCIINOP (Process Sample Off-Gas Pipe Down-Inoperable)
Down-Inoperable)
AUTO ACTIONS NONE CAUSE CAUSE 1.
            'I. Off-gas vent Off-gas   vent pipe pipe (stack)
(stack) radiation radiation monitor monitor downscale downscale or           out of or out    of service.
service.
2.
2.
: 2.      Circuit Circuit malfunction.
Circuit malfunction.
malfunction.
3.
3.
: 3.      Change Change inin background background counts,  counts, possibly possibly from from unitunit power power reduction.
Change in background counts, possibly from unit power reduction.
reduction.
Categories K/A:
Categories Categories K/A:
262002 G2.0 1.23 Tier / Group:
KIA:              262002 262002 G2.0 G2.01.23 1.23                                       Tier Tier // Group:
T2G1 RO Rating:
Group:      T2G1 T2Gl RO   Rating:
4.3 SRO Rating:
RORating:        4.3 4.3                                                       SRO     Rating:
4.2 LP Obj:
SRORating:            4.2 4.2 LP Obj:
I 1-15A Source:
LPObj:            I11-15A 1-15A                                                   Source:
NEW Cog Level:
Source:              NEW NEW Cog Level:
HIGH Category 8:
Cog  Level:      HIGH HIGH                                                     Category Category 8:   8:     YY
Y Notes 8.0 INFREQUENT OPERATIONS.................................................................................. 32 8.1 Transferring UPS Loads From Alternate Source to PrimalY UPS Unit 2A..... 32 8.2 Transferring UPS Loads From Alternate Source to Standby UPS Unit 28.... 35 8.3 Transferring UPS Loads From Standby UPS Unit 2B to Alternate Source.... 39 8.4 Transferring UPS Loads From Primary UPS Unit 2A to Alternate Source..... 4'1 8.5 Alignment of Standby UPS Unit 2B After a Loss of Alternate Source Power. 43 8.6 Returning Standby UPS Unit 2B to Normal Operating Condition Upon Regaining Altemate Source Power Supply..................................................... 45 8.7 Stack Radiation Monitor UPS Power Supply Transfer.................................... 47 120P-52 Rev. 53 Page 2 of 78 I 8.0 INFREQUENT OPERATIONS............................................................................... 34 8.1 Transferring UPS Loads From Alternate Source to PrimalY UPS Unit 1A..... 34 8.2 Transferring UPS Loads From Alternate Source to Standby UPS Unit 18.... 38 8.3 Transferring UPS Loads From Standby UPS Unit '1 B to Alternate Source.... 42 8.4 Transferring UPS Loads From Primary UPS Unit 'lA to Alternate Source..... 44 8.5 Alignment of Standby UPS Unit 1 B After a Loss of Alternate Source Power................................................................................................................. 46 8.6 Returning Standby UPS Unit 1 B to Normal Operating Condition Upon Regaining Alternate Source Power Supply..................................................... 48 l'lOP-52 Rev. 35 Page 3 of 74 I PROCESS SMPL OG VENT PIPE DNSCllNOP (Process Sample Off-Gas Pipe Down-Inoperable)
: 83. The
AUTO ACTIONS NONE CAUSE Unit 2 APP UA-03 6-3 Page 1 of'l
: 83. The following following conditions conditions exist exist on on Unit Unit Two Two following following aa spurious spurious Main Main Turbine Turbine trip trip at at rated power:
'I.
rated    power:
Off-gas vent pipe (stack) radiation monitor downscale or out of service.
SDV HI-HI SDV   HI-HI WTR WTR LVL LVL TRIP TRIP BYPASS BYPASS             In alarm In  alarm OTBD NSSS OTBD          VALVES MTR NSSS VALVES      MTR OVERLOAD OVERLOAD          In alarm In alarm Reactor level Reactor level                                  185 185 inches inches and and steady Reactor Pressure Reactor  Pressure                                    psig with BPVs 900 psig        BPVs controlling Rods All Control Rods                              Fully inserted Fully  inserted Scram                                        Being reset reset lAW lAW LEP-02 RWCU System RWCU                                          Isolated Isolated by 2-G31-F001 2-G31 -FOOl The 2-G31-F004 (RWCU Outboard Isol    lsol Vlv)
: 2.
VIv) failed to automatically close on a valid isolation signal due to motor overload.
Circuit malfunction.
: 3.
Change in background counts, possibly from unit power reduction.
Categories KIA:
262002 G2.0 1.23 Tier / Group: T2Gl RORating:  
 
===4.3 SRORating===
===4.2 LPObj===
11-15A Source:
NEW Cog Level:
HIGH Category 8:
Y
: 83. The following conditions exist on Unit Two following a spurious Main Turbine trip at rated power:
SDV HI-HI WTR LVL TRIP BYPASS OTBD NSSS VALVES MTR OVERLOAD Reactor level Reactor Pressure All Control Rods Scram RWCU System In alarm In alarm 185 inches and steady 900 psig with BPVs controlling Fully inserted Being reset lAW LEP-02 Isolated by 2-G31 -FOOl The 2-G31-F004 (RWCU Outboard lsol VIv) failed to automatically close on a valid isolation signal due to motor overload.
Which one of the following identifies the Technical Specification requirements when the RSP is exited?
Which one of the following identifies the Technical Specification requirements when the RSP is exited?
The RSP can be exited to OGP-05, Unit Shutdown, provided an active LCO is implemented for Technical Specification         (1) .
The RSP can be exited to OGP-05, Unit Shutdown, provided an active LCO is implemented for Technical Specification (1)
The start time of the LCO action completion time is when the             (2)
The start time of the LCO action completion time is when the (2)
A. (1) 3.3.1.1, Reactor Protection System (RPS) Instrumentation (2) condition occurred B. (1) 3.3.1.1, Reactor Protection System (RPS) Instrumentation (2) RSP is exited C. (1) 3.6.1.3, Primary Containment Isolation Valves (PCIVs) (PCIV5)
A.
(2) condition occurred D (1) 3.6.1.3, Primary Containment Isolation Valves (PCIVs)
(1) 3.3.1.1, Reactor Protection System (RPS) Instrumentation (2) condition occurred B.
D!'"                                                        (PCIV5)
(1) 3.3.1.1, Reactor Protection System (RPS) Instrumentation (2) RSP is exited C. (1) 3.6.1.3, Primary Containment Isolation Valves (PCIV5)
(2) condition occurred D (1) 3.6.1.3, Primary Containment Isolation Valves (PCIV5)
(2) RSP is exited
(2) RSP is exited
: 83. The following conditions exist on Unit Two following a spurious Main Turbine trip at rated power:
SDV HI-HI WTR LVL TRIP BYPASS OTBD NSSS VALVES MTR OVERLOAD Reactor level Reactor Pressure All Control Rods Scram RWCU System In alarm In alarm 185 inches and steady 900 psig with BPVs controlling Fully inserted Being reset lAW LEP-02 Isolated by 2-G31-F001 The 2-G31-F004 (RWCU Outboard Isol Vlv) failed to automatically close on a valid isolation signal due to motor overload.
Which one of the following identifies the Technical Specification requirements when the RSP is exited?
The RSP can be exited to OGP-05, Unit Shutdown, provided an active LCO is implemented for Technical Specification (1)
The start time of the LCO action completion time is when the (2)
A. (1) 3.3.1.1, Reactor Protection System (RPS) Instrumentation (2) condition occurred B. (1) 3.3.1.1, Reactor Protection System (RPS) Instrumentation (2) RSP is exited C. (1) 3.6.1.3, Primary Containment Isolation Valves (PCIVs)
(2) condition occurred D!'" (1) 3.6.1.3, Primary Containment Isolation Valves (PCIVs)
(2) RSP is exited


Feedback Feedback K/A: S295006G KIA:  S295006G 2.02.22 2.02.22 Knowledge of Knowledge      of limiting limiting conditions conditions forfor operations operations andand safety safety limits.
Feedback K/A: S295006G 2.02.22 Knowledge of limiting conditions for operations and safety limits.
limits.
SCRAM (CFR: 41.5/43.2/45.2)
SCRAM SCRAM (CFR: 41.5 (CFR:   41.5/43.2/45    .2) 1 43.2 1 45.2)
RO/SRO Rating: 4.0/4.7 Objective: CLSLP300C*1 I
RO/SRO Rating:
: 11. Given plant conditions, the Unit Shutdown Procedure (GP-05), and the Reactor Scram Procedure, determine if conditions allow exiting the Reactor Scram Procedure.
RO/SRO    Rating: 4.0/4.7 4.0/4.7 CLSLP300C*1 I Objective: CLS-LP-300-C*11 Objective:
: 11. Given
: 11. Given plant plant conditions, conditions, the the Unit  Shutdown Procedure Unit Shutdown   Procedure (GP-05),
(GP-05), and and the the Reactor Reactor Scram Scram Procedure, Procedure, determine ifif conditions determine       conditions allow allow exiting exiting the the Reactor Reactor Scram Scram Procedure.
Procedure.


==Reference:==
==Reference:==
1 OCFR5O.36 OEOP-01-UG, Revision 55, Page 31, Section 3.5 Cog Level: High Explanation:
The EOPs authorize actions outside of technical specifications to mitigate the consequences of an emergency condition. The EOPs also provide for returning the system or component to service, If the system or component is not returned to its standby or operable condition prior to exiting the EOP5, then the appropriate limiting condition of operation shall be implemented in accordance with Technical Specifications The starting time for the limiting condition of operation is the time that the EOPs were exited.
In order to exit EOP, compatibilty with GP-05 along with active LCOs need to be implemented. PCIV G31-F004 is inoperable, TS 3.6.1.3 Condition A (A1)requires isolating the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve witf witj pursQ() jerify the affected penetration flow path is isolated Once per 31 days for isolation devices outside primary containment AND Prior to entering MODE 2 or 3 from MODE 4, if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside primary containment.
Distractor Analysis:
Choice A: Plausible because OGP-01 would be entered in order to restart the reactor and TS 3.3.1.1 would be correct if the unit was in Mode I or 2
- SDV Hi level is not required in Mode 3.
Choice B: Plausible because OGP-05 is correct and TS 3.3.1.1 would be correct if the unit was in Mode 1 or 2
- SDV Hi level is not required in Mode 3.
Choice C: Plausible because OGP-01 would be entered in order to restart the reactor and TS 3.6.1.3 is correct.
Choice D: Correct Answer SRO Only Basis: Requires assessment of plant conditions (RPS SDV Hi Level Bypass and Failed open PCIV) and prescribing a procedure with which to proceed (OGP-05).
Notes Feedback KIA: S295006G 2.02.22 Knowledge of limiting conditions for operations and safety limits.
SCRAM (CFR: 41.5 1 43.2 1 45.2)
RO/SRO Rating: 4.0/4.7 Objective: CLS-LP-300-C*11
: 11. Given plant conditions, the Unit Shutdown Procedure (GP-05), and the Reactor Scram Procedure, determine if conditions allow exiting the Reactor Scram Procedure.


==Reference:==
==Reference:==
 
10CFR50.36 OEOP-01-UG, Revision 55, Page 31, Section 3.5 Cog Level: High Explanation:
1 OCFR5O.36 10CFR50.36 OEOP-01-UG, Revision OEOP-01-UG,      Revision 55,   Page 31, Section 55, Page      Section 3.5 3.5 Cog Level:
The EOPs authorize actions outside of technical specifications to mitigate the consequences of an emergency condition. The EOPs also provide for returning the system or component to service. If the system or component is not returned to its standby or operable condition prior to exiting the EOPs, then the appropriate limiting condition of operation shall be implemented in accordance with Technical Specifications. The starting time for the limiting condition of operation is the time that the EOPs were exited.
Cog   Level: High High Explanation:
In order to exit EOP, compatibilty with GP-05 along with active LCOs need to be implemented. PCIV G31-F004 is inoperable, TS 3.6.1.3 Condition A (A1)requires isolating the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve wi1~q9Ll~l~J~q~~ilbiml&~l:Ii~(.~ify the affected penetration flow path is isolated Once per 31 days for isolation devices outside primary containment AND Prior to entering MODE 2 or 3 from MODE 4, if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside primary containment.
The EOPs authorize actions outside of technical specifications to mitigate the consequences consequences of an emergency condition. The EOPs  EOPs also provide provide for returning returning the system or component to service.
service, IfIf the system or component is    is not returned to its standby or operable condition prior to exiting the EOPs, EOP5, then the appropriate limiting condition of operation shall be implemented in accordance with Technical Specifications The starting time for the limiting condition of operation is the time that the EOPs were Specifications.
exited.
In order to exit EOP, compatibilty with GP-05 along with active LCOs need to be implemented. PCIV G31-F004 is inoperable, TS 3.6.1.3 Condition A (A1)requires isolating the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve check  valve wi1~q9Ll~l~J~q~~ilbiml&~l:Ii~(.~ify witf                                      witj    pursQ() jerify the affected penetration flow path is isolated Once per 31 days for isolation devices outside primary containment AND Prior to entering MODE 2 or 3 from MODE 4, if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside primary containment.
Distractor Analysis:
Distractor Analysis:
Choice A: Plausible because OGP-01 would be entered in order to restart the reactor and TS 3.3.1.1 would be correct if the unit was in Mode 1I or 2 - SDV Hi level is not required in Mode 3.
Choice A: Plausible because OGP-01 would be entered in order to restart the reactor and TS 3.3.1.1 would be correct if the unit was in Mode 1 or 2 - SDV Hi level is not required in Mode 3.
Choice B: Plausible because OGP-05 is correct and TS 3.3.1.1 would be correct if the unit was in Mode 1            1 or 2 - SDV Hi level is not required in Mode 3.
Choice B: Plausible because OGP-05 is correct and TS 3.3.1.1 would be correct if the unit was in Mode 1 or 2 - SDV Hi level is not required in Mode 3.
Choice C: Plausible because OGP-01 would be entered in order to restart the reactor and TS 3.6.1.3 is correct.
Choice C: Plausible because OGP-01 would be entered in order to restart the reactor and TS 3.6.1.3 is correct.
Choice D: Correct Answer SRO Only Basis: Requires assessment of plant conditions (RPS SDV Hi              Hi Level Bypass and Failed open PCIV) and prescribing a procedure procedure with which to proceed (OGP-05).
Choice D: Correct Answer SRO Only Basis: Requires assessment of plant conditions (RPS SDV Hi Level Bypass and Failed open PCIV) and prescribing a procedure with which to proceed (OGP-05).
Notes
Notes  


3.5 3.5      Technical Specifications Technical     Specifications The EOPs The    EOPs authorize authorize actions actions outside outside of of technical technical specifications specifications to  to mitigate mitigate thethe consequences of consequences         of an   emergency condition.
3.5 Technical Specifications The EOPs authorize actions outside of technical specifications to mitigate the consequences of an emergency condition. The EOPs also provide for returning the system or component to service.
an emergency        condition. TheThe EOPs EOPs also also provide provide for for returning the returning  the system system or    or component component to       service. IfIf the to service.        the system system or  or component component isis not returned not  returned to to itsits standby standby or  or operable operable condition condition prior prior to to exiting exiting the the EOPs, EOPs, then the then  the appropriate appropriate limiting limiting condition condition ofof operation operation shall shall be be implemented implemented in    in accordance with accordance    with Technical Technical Specifications.
If the system or component is not returned to its standby or operable condition prior to exiting the EOPs, then the appropriate limiting condition of operation shall be implemented in accordance with Technical Specifications. The starting time for the limiting condition of operation is the time that the EOPs were exited.
Specifications. The The starting starting time time for for the the limiting limiting condition of condition   of operation operation is    is the   time that the time  that the the EOPs EOPs were were exited.
OEOP-01-IJG Rev. 55 Page 31 of 151 Completion Times 1.3 1.0 USE AND APPLICATION 1.3 Completion Times PURPOSE The purpose of this section is to establish the Completion Time convention and to provide guidance for its use.
exited.
BACKGROUND Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe operation of the unit. The ACTIONS associated with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Times(s).
OEOP-01-IJG IOEOP-01-UG                                            Rev. 55 Rev. 55                                   Page 31 Page        of 'IS" 31 of  151 I Completion Completion Times Times 1.3 1.3 1.0 USE AND APPLICATION
DESCRIPTION The Completion Time is the amount of time allowed for completing a Required Action. it is referenced to the time of discovery of a situation (e.g, inoperable equipment or variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the unit is in a MODE or specified condition stated in the Applicability of the LCO. Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the unit is not within the LCO Applicability.
  '1.0           APPLICATION 1.3
3.5 Technical Specifications The EOPs authorize actions outside of technical specifications to mitigate the consequences of an emergency condition. The EOPs also provide for returning the system or component to service. If the system or component is not returned to its standby or operable condition prior to exiting the EOPs, then the appropriate limiting condition of operation shall be implemented in accordance with Technical Specifications. The starting time for the limiting condition of operation is the time that the EOPs were exited.
  '1.3 Completion Times Completion    Times PURPOSE PURPOSE                The purpose of this section is to establish the Completion Time convention and to provide guidance for its use.
IOEOP-01-UG Rev. 55
BACKGROUND BACKGROUND              Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe operation of the unit. The ACTIONS associated *with                with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Times(s).
'1.0 USE AND APPLICATION Page 31 of 'IS" I Completion Times 1.3  
DESCRIPTI DESCRIPTION  ON        The Completion Time is the amount of time allowed for completing a Required Action. itIt is referenced to the time of discovery of a situation (e.g., inoperable equipment or variable not within limits) that requires (e.g, entering entering an ACTIONS Condition  Condition unless otherwise specified, providing providing the unit is in in a MODE MODE or specified condition stated in the Applicability of          of the LCO. Required Required Actions must be completed prior to the        the expiration of  ofthe the specified specified Completion Completion Time.
'1.3 Completion Times PURPOSE BACKGROUND DESCRIPTION The purpose of this section is to establish the Completion Time convention and to provide guidance for its use.
Time. An An ACTIONS Condition Condition remains in    in effect and and the Required Actions apply until the ConditionCondition no  no longer longer exists or or the the unit unit isis not not within the LCO LCO Applicability Applicability..
Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe operation of the unit. The ACTIONS associated *with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Times(s).
The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the time of discovery of a situation (e.g., inoperable equipment or variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the unit is in a MODE or specified condition stated in the Applicability of the LCO. Required Actions must be completed prior to the expiration ofthe specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the unit is not within the LCO Applicability.  


WHITE 5-5 55 OTBD NSSS OTBD              VALVES MTR NSSS VALVES          MTR OVERLOAD OVERLOAD                        Page Page 11 of2 of 2 1.0 OPERATOR 1.0    OPERATOR ACTIONS: ACTIONS:
WHITE 55 OTBD NSSS VALVES MTR OVERLOAD Page 1 of 2 1.0 OPERATOR ACTIONS:
1.1   OBSERVE 1.1 OBSERVE Automatic Automatic Functions:
1.1 OBSERVE Automatic Functions:
Functions:
1.1.1 IF one of the affected valves was being operated. THEN:
1.1.1 IF 1.1.1    IF one one of of the  affected valves the affected   valves waswas being being operated, operated. THEN:
: 1. Valve motion will stop
THEN:
: 2. Valve will NOT respond to control signals
: 1. Valve motion
: 3. Valve position will still be indicated 1.2 PERFORM Corrective Actions:
: 1. Valve               will stop motion will   stop
NOTE: Resetting valve motor overload devices or manual operaon of tripped motor-operated valves should only be attempted in emergency situations as directed by the Unit SCO.
: 2. Valve will
CAUTION During manual oPeration of motor-operated valves, personnel should stand clear of the valve vhiIe either:
: 2. Valve     will NOT NOT respond respond to  to control control signals signals
: 1. Resetting the thermal 3verload device or 2.
: 3. Valve
Operating the valve remo:ely.
: 3. Valve position position will will still still be  indicated be indicated 1.2  PERFORM Corrective 1.2 PERfORM           Corrective Actions:
1.2.1 IF the affected valve is required for operation, THEN PERFORM the following steps:
Actions:
: 1. RESET the thermal overload device at the affected valve breaker compartment AND OPERATE the valve again.
NOTE: Resetting NOTE:     Resetting valve motor motor overload overload devices devices oror manual manual operation operaon of    tripped of tripped motor-operated valves sl10uld motor-operated                should onlyonly be be attempted attempted in in emergency Situations situations as directed by directed        the Unit by the        SCO.
2.
Unit SeQ.
IF the themal overload device actuates again. THEN MANUALLY OPERATE the valve.
CAUTION During manual During    manual operation oPeration of motor-operated motor-operated valves, personnel sh.ould should stand clear clear of the valve while vhiIe either:
: 3. WHEN the valve is broken off its closed or open seat, THEN RESET the themial overload device at the affected valve breaker compartment AND OPERATE the valve.
: 1. Resetting the thermal overload
1.2.2 REFERt0TS. 3.6.1.3 and TRMApp DTable 3.6.1.3-2.
: 1.                               3verload device device Ofor Operating the valve remotely.
24,PP-A-02 Rev. 32 Page 57 of 5-5 OTBD NSSS VALVES MTR OVERLOAD Page 1 of2 1.0 OPERATOR ACTIONS:
: 2. Operating                    remo:ely.
1.1 OBSERVE Automatic Functions:
1.2.1 IF  IF the affected valve is required for operation, THEN PERFORM the following steps:
1.1.1 IF one of the affected valves was being operated, THEN:
: 1. Valve motion will stop
: 2. Valve will NOT respond to control signals
: 3. Valve position will still be indicated 1.2 PERfORM Corrective Actions:
NOTE: Resetting valve motor overload devices or manual operation of tripped motor-operated valves sl10uld only be attempted in emergency Situations as directed by the Unit SeQ.
CAUTION During manual operation of motor-operated valves, personnel sh.ould stand clear of the valve while either:
: 1. Resetting the thermal overload device Of
: 2. Operating the valve remotely.
1.2.1 IF the affected valve is required for operation, THEN PERFORM the following steps:
: 1. RESET the thermal overload device at the affected valve breaker compartment AND OPERATE the valve again.
: 1. RESET the thermal overload device at the affected valve breaker compartment AND OPERATE the valve again.
themal overload device actuates again,
: 2. IF the themlal overload device actuates again, THEN MANUALLY OPERATE the valve.
: 2. IF the themlal                                         again. THEN MANUALLY OPERATE the valve.
: 3. WHEN the valve is broken off its closed or open seat, THEN RESET the themlal overload device at the affected valve breaker compartment AND OPERATE the valve.
: 3. WHEN the valve is broken off its closed or open seat, THEN RESET the themial overload device at the affected valve breaker compartment AND themlal OPERATE the valve.
1.2.2 REFER to T.S. 3.6.1.3 and TRM App. 0 Table 3.6.1.3-2.
REFERt0TS 1.2.2 REFER         to T.S.. 3.6.1.3 and TRM TRMApp App. 0DTable Table 3.6.1.3-2.
12APp-A-02 Page 57 of 791  
24,PP-A-02 12APp-A-02                                             Rev. 32                              Page Page 57  of 57 of 791


PCIVs 3.6.1.3 3.6 CONTAINMENT 3.6  CONTAINMENT SYSTEMS  SYSTEMS 3.6.1.3 3.6.'1.3          Primary Containment Primary Containment Isolation Isolation Valves Valves (PCIVs)
PCIVs 3.6.1.3 3.6 CONTAINMENT SYSTEMS 3.6.1.3 Primary Containment Isolation Valves (PCIVs)
(PCIVs)
LCO 16.1.3 Each PC1V, except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE.
LCO 3.6.
LCO     16.1.3'1.3        Each   PC1V, except Each PCIV,    except reactor reactor building-to-suppression building-to-suppression chamber chamber vacuum vacuum breakers, shall breakers,  shall bebe OPERABLE.
OPERABLE.
APPLICABILITY:
APPLICABILITY:
APPLICABILITY:            MODES 1, MODES      1, 2, 2, and and 3, 3, When associated VV'hen associated instrumentation instrumentation is is required required to to be  OPERABLE per be OPERABLE     per LCO   3.3.6.1, "Primary LCO 3.3.6:1,      Primary Containment Containment Isolation Isolation Instrumentation."
MODES 1, 2, and 3, When associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, Primary Containment Isolation Instrumentation.
Instrumentation.
CONDITION REQUIRED ACTION COMPLETION TIME A.
CONDITION                             REQUIRED ACTION REQUIRED      ACTION                COMPLETION COMPLETION TIME A.                   NOTE-------
    -------------NOTE-----------
A.l A:I          Isolate the affected Isolate                          8B hours Only applicable to                                penetration flow path by penetration flow paths 'Nithwith                  use of at least one closed two PCIVs.                                          and de-activated automatic valve, closed manual valve, blind flange, or check valve One or more penetration                            with flow through the valve flow flo ....' paths with one PCIV                      secured.
inoperable except for MSIV leakage not within limit.            AND


3.3.1 :1 3.3.1.1 3.3 3.3       INSTRUMENTATION INSTRUMENTATION 3.3.1 .1 3.3.*'.'1           Reactor Protection Reactor  Protection SystemSystem (RPS)  (RPS) Instrumentation Instrumentation LCO 3.3.-l.'I LCO      3.3.1.1           The RPS The  RPS instrumentation instrumentation for              for each each Function Function in  in Table       3.3.l.1I shall Table 3.3.'U-*1      shall be    be OPERABLE.
NOTE-------
OPERABLE APPLICABILITY:
A.l Isolate the affected B hours Only applicable to penetration flow path by penetration flow paths with use of at least one closed two PCIVs.
APPLICABILITY:              According to According          to Table Table 3.3.13.3.1 ...1-i. 1-*1.
and de-activated automatic
 
valve, closed manual valve, blind flange, or check valve One or more penetration with flow through the valve flow paths with one PCIV secured.
inoperable except for MSIV leakage not within limit.
3.6 CONTAINMENT SYSTEMS 3.6.'1.3 Primary Containment Isolation Valves (PCIVs)
LCO 3.6. '1.3 Each PCIV, except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3, VV'hen associated instrumentation is required to be OPERABLE per LCO 3.3.6:1, "Primary Containment Isolation Instrumentation."
CONDITION A.
-------------NOTE-----------
A:I Only applicable to penetration flow paths 'Nith two PCIVs.
One or more penetration flo.... ' paths with one PCIV inoperable except for MSIV leakage not within limit.
AND REQUIRED ACTION COMPLETION TIME Isolate the affected 8 hours penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.
 
3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.l.1I shall be OPERABLE.
APPLICABILITY:
According to Table 3.3.1.1-i.
ACTIONS
ACTIONS
      ---------------------------------------------------------- NOT NOTE--- E ---------------------------------------------------------
 
Separate Condition Condition entry is is allowed for each channel.             channel.
NOTE---
RPS Instrumentation 3.3.1.1 3.3.1.*'
Separate Condition entry is allowed for each channel.
Tt :t Tatfe    3~ 1.1'"~ (~3J~ .:.3 o~
RPS Instrumentation 3.3.1.1 Tt 3l.t 3 c R: Pn:r Oy: i:Irc
3l.t                  c 3t R: Pf\:t~,:.~or.
,%PPLICABLE OC:NOION MOD O REOAED REHCE OTHER CHANNEL2 FO 2PECWE TA QLLAE 2URVELLArCE ALLOWABLE FL4O2N OCNDI:ON 3OTE3 AOT2r D.
R~3-:'>>r Pn:r Sj':~m. Oy: !n:.~".m~r:~l!:':f:
REOLJIREVENTO VALUE 7.
i:Irc
Ccr CEcharc Vr 1,2 2
                                          ,%PPLICABLE
OR WLHh OR OR OR CONDITION REQUIRED ACTION COMPLETION TIME A.
                                          .'V'PLICASLE                                     OC:NOION CO~OI7fONS
One or more required Al Place channel in trip.
                                            !MOD
12 hours channels inoperable.
                                            ..~ooszc,:(O          REOAED RSQIJt:tED                    REHCE
OR A.2 NOTE 12 hours Not applicable for Functions 2.a, 2.b, 2.c, 24, or 2.f.
                                                                                        ,:t5FERENCEO OTHER C*"iHER.           CHANNEL2 CHANNeLS                      FO AAOM 2PECWE (lPECIFIE::l           i=E?t r,:w=
Place associated trip system in trip.
TA                    QLLAE FtEQIJ1,"iEO         aVR\'EII.LAJIoICE.
Categories KJA:
2URVELLArCE              "'1.L!:JW ALLOWABLE *.r..sLE FL4O2N FUNCttON              OCNDI:ON CC'NOI7'ICNS              3OTE3 3'V!l"iE\1              AOT2r 0.1 ACtiON    D.          RSQUIREAlEma REOLJIREVENTO                  VAL,VE VALUE 7.
S2950006G 2.02.22 Tier! Group:
: 7. Ccr CEcharc Vr ScramOtscharoeVO\\m""e                      1,2 1,;:                  2                        ~         OR ~.l.4.1";
T1GI RO Rating:
WLHh W3!er Le'o!et-HIgn                                                                                  a.=t OR :;.3:.~.4.:S OR 3.3."L~LH a,;:(
4.0 SRO Rating:
:a,=\
4.7 LP Obj:
OR 3.3:.il.L1':'
CLSLP300C*11 Source:
CONDITION                                          REQUIRED ACTION                                             COMPLETION TIME A. One or more required                          Al A.'I               Place channel in trip.                                   *12 12 hours channels inoperable.
NEW Cog Level:
OR A.2                                         NOTE
HIGH Category 8:
                                                                          ---------------NOTE-------------                         12 12 hours Not applicable for Functions 2.a, 2.b, 2.c, 24,            2.d, or 2.f.
3.3.1 :1 3.3 INSTRUMENTATION 3.3.*'.'1 Reactor Protection System (RPS) Instrumentation LCO 3.3.-l.'I The RPS instrumentation for each Function in Table 3.3.'U-*1 shall be OPERABLE APPLICABILITY:
According to Table 3.3.1.. 1-*1.
ACTIONS
----------------------------------------------------------NOT E ---------------------------------------------------------
Separate Condition entry is allowed for each channel.
FUNCttON
: 7.
ScramOtscharoeVO\\\\m""e W3!er Le'o!et-HIgn CONDITION A.
One or more required channels inoperable.
Categories KIA:
S2950006G 2.02.22 RORating:
 
==4.0 LPObj==
CLS-LP-300-C* 11 Cog Level:
HIGH RPS Instrumentation 3.3.1.*'
Tatfe :t 3~ 1.1'"~ (~3J~.:. o~ 3t R~3-:'>>r Pf\\:t~,:.~or. Sj':~m. !n:.~".m~r:~l!:':f:  
.'V'PLICASLE  
!.. ~ooszc,:(
C*"iHER.
(lPECIFIE::l CC'NOI7'ICNS 1,;:
CO~OI7fONS RSQIJt:tED
,:t5FERENCEO CHANNeLS AAOM i=E?t r,:w=
3'V!l"iE\\1 FtEQIJ1,"iEO aVR\\'EII.LAJIoICE.
ACtiON 0.1 RSQUIREAlEma  
~
OR ~.l.4.1";
a.=t
:;.3:.~.4.:S a,;:( 3.3."L~LH
:a,=\\ 3.3:.il.L1':'  
"'1.L!:JW *.r..sLE VAL,VE REQUIRED ACTION COMPLETION TIME A.'I Place channel in trip.  
*12 hours OR A.2  
---------------NOTE-------------
12 hours Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f.
Place associated trip system in trip.
Place associated trip system in trip.
Categories Categories KJA:
Tier / Group: TIG!
KIA:                  S2950006G S2950006G 2.02.22 2.02.22                                            Tier / Group:
SRORating: 4.7 Source:
Tier!        Group: T1GI  TIG!
NEW Category 8:
RO  Rating:
: 84. Following a scram on Unit Two, which one of the following correctly identifies:
RORating:            4.0 4.0                                                            SRO SRORating: Rating: 4.7    4.7 LP Obj:
(1) the initial response of reactor water level if an SRV is opened and (2) the procedure that contains the guidance to close the MSIVs due to water level?
LPObj:                CLSLP300C*11 CLS-LP-300-C* 11                                                Source:
A. (1) Shrink (2) Reactor Scram Procedure B. (1) Shrink (2) 2APP-A-07, REACTOR WATER LEVEL HIGH/LOW C (1) Swell (2) Reactor Scram Procedure D. (1) Swell (2) 2APP-A-07, REACTOR WATER LEVEL HIGH/LOW Feedback K/A: 295008 A2.05 Ability to determine andlor interpret the following as they apply to HIGH REACTOR WATER LEVEL:
Source:                    NEW NEW Cog Cog Level:
Swell (CFR: 41.10/43.5/45.13)
Level:          HIGH HIGH                                                            Category Category 8:        8:
ROISRO Rating: 2.9/3.1 Objective: CLS-LP-300-C, 10 Given plant conditions and the RSP, determine the required operator actions.
: 84. Following
: 84. Following aa scram scram onon Unit Unit Two, Two, which which one one of  of the the following following correctly correctly identifies:
identifies:
(1) the (1)  the initial initial response response of of reactor reactor water water level level ifif an an SRV SRV is is opened opened andand (2) the (2)  the procedure procedure thatthat contains contains the the guidance guidance to      close the to close   the MSIVs MSIVs due due toto water water level?
level?
A. (1)
A.        Shrink (1) Shrink (2) Reactor Scram Procedure (2) Reactor              Procedure B. (1)
B.  (1) Shrink Shrink (2)   2APP-A-07, (2) 2APP-A-07, REACTORREACTOR WATER  WATER LEVEL LEVEL HIGH/LOW HIGH/LOW C
c~  (1) Swell (1)
(2) Reactor Scram Procedure Procedure (1) Swell D. (1)     Swell (2) 2APP-A-07, REACTOR WA            WATER TER LEVEL HIGH/LOW Feedback K/A: 295008 A2.05 KIA:
andlor interpret the following as they apply to HIGH REACTOR WATER Ability to determine and/or LEVEL:
Swell 41.10/43.5/45.13)
(CFR: 41.10/43.5/45.13)
ROISRO Rating: 2.9/3.1 RO/SRO Objective: CLS-LP-300-C, 10 Given plant Given        conditions and plant conditions and the the RSP, RSP, determine determine the required operator actions.


==Reference:==
==Reference:==
RSP I 001-37.3 2APP-A-07, page 12 Cog Level: High Explanation:
Opening of the SRV will cause the reactor water level to swell up due to the reduction in pressure in the vessel and if level reaches the value in figure 1 on the RSP then closure of the MSIVs is directed. The MSIVs will close automatically but only on low level 3, not high level.
Distractor Analysis:
Choice A: Plausible if the examinee thinks that opening the SRV would reduce the water volume in the RPV, The RSP does contain the actions to close the MSIVs.
Choice B: Plausible if the examinee thinks that opening the SRV would reduce the water volume in the RPV, the examinee may think that the closure is an auto action, which are contained in the APP.
Choice C: Correct see explanation Choice D: Plausible because reactor water level will swell, and the examinee may think that the closure is an auto action, which are contained in the APP.
SRO Basis: 10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Notes
: 84. Following a scram on Unit Two, which one of the following correctly identifies:
(1) the initial response of reactor water level if an SRV is opened and (2) the procedure that contains the guidance to close the MSIVs due to water level?
A. (1) Shrink (2) Reactor Scram Procedure B. (1) Shrink (2) 2APP-A-07, REACTOR WATER LEVEL HIGH/LOW c~ (1) Swell (2) Reactor Scram Procedure D. (1) Swell (2) 2APP-A-07, REACTOR WA TER LEVEL HIGH/LOW Feedback KIA: 295008 A2.05 Ability to determine and/or interpret the following as they apply to HIGH REACTOR WATER LEVEL:
Swell (CFR: 41.10/43.5/45.13)
RO/SRO Rating: 2.9/3.1 Objective: CLS-LP-300-C, 10 Given plant conditions and the RSP, determine the required operator actions.


I 001-37.3 RSP 1001-37.3 2APP-A-07, page 12 Cog Level: High Explanation:
==Reference:==
Opening of the SRV will cause the reactor water level to swell up due to the reduction in pressure in the vessel and if level reaches the value in figure 1  1 on the RSP then closure of the MSIVs is directed. The MSIVs will close automatically but only on low level 3, not high level.
RSP 1001-37.3 2APP-A-07, page 12 Cog Level: High Explanation:
Opening of the SRV will cause the reactor water level to swell up due to the reduction in pressure in the vessel and if level reaches the value in figure 1 on the RSP then closure of the MSIVs is directed. The MSIVs will close automatically but only on low level 3, not high level.
Distractor Analysis:
Distractor Analysis:
Choice A: Plausible if the examinee thinks that opening the SRV would reduce the water volume in            in the RPV, The RSP does contain the actions to close the MSIVs. MSIVs.
Choice A: Plausible if the examinee thinks that opening the SRV would reduce the water volume in the RPV, The RSP does contain the actions to close the MSIVs.
Choice B: Plausible if the examinee thinks that opening the SRV would reduce the water volume in the RPV, the examinee may think that the  the closure is an auto action, which are contained in the APP.
Choice B: Plausible if the examinee thinks that opening the SRV would reduce the water volume in the RPV, the examinee may think that the closure is an auto action, which are contained in the APP.
Choice Choice C:C: Correct  see explanation Correct see explanation Choice Choice D:0: Plausible because because reactor water level level will will swell, and and the examinee examinee may may think think that that the the closure is is an auto an        action, which auto action,        are contained which are contained inin the the APP.
Choice C: Correct see explanation Choice 0: Plausible because reactor water level will swell, and the examinee may think that the closure is an auto action, which are contained in the APP.
APP.
SRO Basis: 10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
SRO Basis:
Notes  
Basis: 10 CFR 55.43(b)-5, 10 CFR  55.43(b)-5, Assessment of   of facility facility conditions and and selection of of appropriate procedures during during normal,   abnormal, and normal, abnormal,  and emergency emergency situations.
situations.
Notes Notes


REACTOR WATER REACTOR      WATER LEVELLEVEL HIGHlLOW I-IIGHFLOW                           PaOe 11 of Page        2 012 1.0   OPERATOR ACTIONS:
REACTOR WATER LEVEL I-IIGHFLOW PaOe 1 012 1.0 OPERATOR ACTIONS:
1.0 OPERATOR            ACTIONS:
1.1 CONFIRM by multiple indications actual high or low reactor water level:
1 .1 CONFIRM 1.1  CONFIRM by    by multiple multiple indications indications actual actual high high or or low low reactor reactor water water level:
1.1.1 Reactor water level indication on RTGB Panel P603 may be used ror verification of water level:
level:
1, Reactor Water Level A, C32-Ll-R606A.
1.1.1     Reactor  water level indication 1.1.1 Reactor water level indication on        on RTGB RTGB Panel Panel P603 P603 may may I)e be used used for ror verification of verification  of water  level:
: 2. Reactor Water Level B, C32-Ll-R6066.
water level:
: 3. Reactor Water Level C, C32-Ll-R606C.
: 1. Reactor Water 1, Reactor    Water Level Level .4" A, C32-Ll-R606.'\.
: 4. Reactor Level/Pressure Recorder, C32-R608.
C32-Ll-R606A.
1.2 OBSERVE Automatic Functions:
: 2. Reactor
1.2.1 IF reactor level decreases to 136 inches, THEN a reactor Scram results.
: 2. Reactor Water Water Level Level B, B, C32-Ll-R606B.
1.2.2 IF reactor level increases to 206 inches. THEN the Main Turbine, RFPTs, RCIC and HPCI turiines will trip.
C32-Ll-R6066.
1.2.3 IF either of the RFPs have tripped AND reactor water level is less than 182 inches. TI-lEN a Recirculation Pump runback will occur.
: 3. Reactor
2APP-A-07 Rev. 32 Page 12 of 45 From the Reactor Scram Procedure:
: 3. Reactor Water LevelLevel C, C, C32-U-R606C.
flWAN J_Yt PL.CLALLMIV lTCHE1OCLO o1 LOWER REAtOR WATER WH RWCU REACTOR WATER LEVEL HIGHlLOW Page 1 of 2 1.0 OPERATOR ACTIONS:
C32-Ll-R606C.
1.1 CONFIRM by multiple indications actual high or low reactor water level:
: 4. Reactor
1.1.1 Reactor water level indication on RTGB Panel P603 may I)e used for verification of water level:
: 4. Reactor Level/Pressure Level/Pressure Recorder, Recorder, C32-R608.
: 1. Reactor Water Level.4" C32-Ll-R606.'\\.
C32-R608.
: 2. Reactor Water Level B, C32-Ll-R606B.
1.2 OBSERVE Automatic 1.2                  Automatic Functions:
: 3. Reactor Water Level C, C32-U-R606C.
Functions:
: 4. Reactor Level/Pressure Recorder, C32-R608.
1.2.1 IF 1.2.1      reactor level IF reactor  level decreases decreases to to '166 136 inches, inches, THEN THEN aa reactor reactor Scram Scram results.
1.2 OBSERVE Automatic Functions:
results.
1.2.1 IF reactor level decreases to '166 inches, THEN a reactor Scram results.
1.2.2 IF 1.2.2    IF reactor reactor level level increases increases toto 206 206 inclles,   THEN the inches. THEN      the Main Main Turbine.
1.2.2 IF reactor level increases to 206 inclles, THEN the Main Turbine. RFPTs, RCIC and HPCI tUri)ines '.vill trip.
Turbine, RFPTs, RFPTs, RCIC and            turiines '.vill and HPCI tUri)ines     will trip.
1.2.3 IFeitller of tile RFPs have tripped AND reactor water level is less than 182 inches, THEN a Recirculation Pump runback will occur.
1.2.3   IF either of tile 1.2.3 IFeitller      the RFPs have tripped AND reactor water level is less than 182 inches. THEN inches,  TI-lEN a Recirculation Pump runback will occur.
12APP-A-07 Page 12 of 451 From the Reactor Scram Procedure:  
2APP-A-07 12APP-A-07                                         Rev. 32                                    Page 12 Page   12 of 451 of 45 From the Reactor Scram Procedure:
flWAN J_Yt                                                PL.CLALLMIV lTCHE1OCLO o1 LOWER REAtOR WATER WH RWCU


ATTACHMENT66 ATTACHMENT Page 19 Page      of 19 19 of 19 FIGURE 21 FIGURE    21 Reactor Water Reactor  Water Level Level at at MSL MSL (Main  Steam  Line  Flood (Main Steam Line Flood Level) Level>
Cl)
300 (Cl)
LUI (U)
J)
-J LU>
LU W
LU
MSL I
-j a
J:
z ATTACHMENT 6 Page 19 of 19 FIGURE 21 Reactor Water Level at MSL (Main Steam Line Flood Level>
WHEN REACTOR PRESSURE IS LESS THAN 60 PSIG, USE INDICATED LEVEL.
MSL IS+250 INCHES.
DEOP-Ol-UG I
Rev. 55 Page 106 of 151 300 250 REF LEG TEMP ABOVE OR EQUAL TO 200F REF LEG TEMP BELOW 2COF 200 11111 IIIIII lIIlllII!1lllIJIjI1,15o 100 300 500 700 900 13100 60 200 400 600 800 1,000 REACTOR PRESSURE (PSIG)
(J)
W J:
()
()
(U)
Z --
  --Z
..J W >
  .-J
W  
    .J REF lEG LU W                                                                       LEG 250                                                            TEMP LU W
..J C
ABQVEOR ABOVE OR EQUAL TO
W  
  ..J
~
    -j                                                                  200-F 200F Ca                                                                .r REF LEG
o -c z
                                                                    ,. TEMP TEMP W                                                                   BELOW
250 200 OEOP-01-UG ATTACHMENT 6 Page 19 of 19 FIGURE 21 Reactor Water Level at MSL (Main Steam Line Flood Level)
    ~
MSL 00 300 500 60 200 400 600 800 1,000 REACTOR PRESSURE (PSIG)
o 200"F 2COF cz z
WHEN REACTOR PRESSURE IS LESS THAN 60 PSIG, USE INDICATED LEVEL.
200 11111      IIIIII          lIIlllII!1lllIJIjI1,15o 100 00       300 300     500 500         700      900    13100 60 60    200 200      400 400      600 600        800 800    1,000 1,000 REACTOR REACTOR PRESSURE  PRESSURE (PSIG)    (PSIG)
MSL IS +250 INCHES.
WHEN WHEN REACTOR REACTOR PRESSURE PRESSURE IS    IS LESS LESS THAN THAN 60 60 PSIG,   USE INDICATED PSIG, USE  INDICATED LEVEL.
Rev. 55 REF lEG TEMP ABQVEOR EQUAL TO 200-F
LEVEL.
.r REF LEG
MSL MSL IS+250 IS +250 INCHES.
,. TEMP BELOW 200"F Page 106 of 151  
INCHES.
 
DEOP-Ol-UG OEOP-01-UG            I          Rev.
Categories KJA:
Rev. 55 55                        Page 106 Page   106 of of 151 151
295008 A2.05 Tier / Group:
T1G2 RU Rating:
2.9 SRO Rating:
3.1 LP Ubj:
300-C, 10 Source:
NEW Cog Level:
HIGH Category 8:
Y Categories KIA:
295008 A2.05 Tier / Group: TlG2 RORating:


Categories Categories KJA:
===2.9 SRORating===
KIA:        295008 A2.05 295008  A2.05 Tier// Group:
3.1 LP Obj:
Tier  Group: TlG2 T1G2 RU Rating:
300-C,1O Source:
RORating:    2.9 2.9          SRO Rating:
NEW Cog Level:
SRORating:    3.1 3.1 LP Obj:
HIGH Category 8:
LP  Ubj:    300-C, 10 300-C,1O     Source:
Y
Source:      NEW NEW Cog Level:
: 85. Unit Two is operating at 74% power when the FW-V120,FW HTRS 4 & 5 BYP VLV, is inadvertantly opened by mechanics. The valve is bound and can not be reclosed.
Cog  Level: HIGH HIGH         Category 8:8:
Initial Final Feedwater Temperature was 404°F.
Category      YY
Conditions are now stable with reactor power at 81% and Final Feedwater Temperature at 314°F.
: 85. Unit Two
: 85. Unit    Two isis operating operating at   74% power at 74%          when the power when    the FW-V120, FW-V120,FW         HTRS 44 && 55 BYP "FW HTRS            BYP VLV, VLV, isis inadvertantly opened inadvertantly     opened by  by mechanics.
mechanics. TheThe valve valve isis bound bound and and can can not not be be reclosed.
reclosed.
Initial Final Initial  Final Feedwater Feedwater Temperature Temperature was was 404°F.
404°F.
Conditions are Conditions      are now now stable stable with with reactor reactor power power atat 8181%% and and Final Final Feedwater Feedwater Temperature at Temperature        at 314°F.
314°F.
(Reference provided)
(Reference provided)
(Reference      provided)
Which one of the following identifies the required action based on the information above?
Which one Which    one of of the the following following identifies identifies the the required required action action based based onon the the information information above?
above?
Continued operation:
Continued operation:
Continued      operation:
Av is not allowed and reactor shutdown is required lAW OGP-05, Unit Shutdown.
Av is A'!  is not allowed and reactor not allowed          reactor shutdown is  is required required lAWlAW OGP-05, OGP-05, Unit Unit Shutdown.
Shutdown.
B. is not allowed and a manual reactor scram is required lAW 001-01.01, BNP Conduct of Operations Supplement.
B. is not allowed and a manual reactor scram is required lAW 001-01.01, BNP Conduct of Operations Supplement.
C. is allowed provided the FW Heaters 4 & 5 are isolated lAW 20P-32, Condensate and Feedwater Operating Procedure.
C. is allowed provided the FW Heaters 4 & 5 are isolated lAW 20P-32, Condensate and Feedwater Operating Procedure.
D. is allowed provided reduced thermal limits are established within 4 hours as required by Technical Specifications.
D. is allowed provided reduced thermal limits are established within 4 hours as required by Technical Specifications.
: 85. Unit Two is operating at 74% power when the FW-V120, "FW HTRS 4 & 5 BYP VLV, is inadvertantly opened by mechanics. The valve is bound and can not be reclosed.
Initial Final Feedwater Temperature was 404°F.
Conditions are now stable with reactor power at 81 % and Final Feedwater Temperature at 314°F.
(Reference provided)
Which one of the following identifies the required action based on the information above?
Continued operation:
A'! is not allowed and reactor shutdown is required lAW OGP-05, Unit Shutdown.
B. is not allowed and a manual reactor scram is required lAW 001-01.01, BNP Conduct of Operations Supplement.
C. is allowed provided the FW Heaters 4 & 5 are isolated lAW 20P-32, Condensate and Feedwater Operating Procedure.
D. is allowed provided reduced thermal limits are established within 4 hours as required by Technical Specifications.


Feedback Feedback K/A: 295014 KIA:  295014 G2.01.25 G2.01.25 Ability to Ability  to interpret interpret reference reference materials, materials, such such as as graphs, graphs, curves, curves, tables, tables, etc.
Feedback K/A: 295014 G2.01.25 Ability to interpret reference materials, such as graphs, curves, tables, etc.
etc.
Inadvertent Reactivity Addition (CFR: 41.10 / 43.5 / 45.12)
Inadvertent Reactivity Inadvertent    Reactivity Addition Addition (CFR: 41.10/43.5/45.12)
ROISRO Rating: 3.9/4.2 Objective: CLS-LP-34, Obj. 1 Ic Given plant conditions, describe the effect a loss/malfunction of the feedwater heaters will have on:
(CFR:   41.10 / 43.5 / 45.12)
: c. Feedwater Temperature
ROISRO Rating:
RO/SRO    Rating: 3.9/4.2 3.9/4.2 Objective: CLS-LP-34, Objective:   CLS-LP-34, Obj. 1 Ic Obj. 11c Given plant Given  plant conditions, conditions, describe describe the the effect effect aa loss/malfunction loss/malfunction of of the the feedwater feedwater heaters heaters wi" will have have on:
on:
: c. Feedwater
: c. Feedwater Temperature Temperature


==Reference:==
==Reference:==
20P-32, Attachment 4 (provided)
Cog Level HI Explanation:
From Attachment 4 of 20P-32 operation is outside of the allowable range (<1 10.3&deg;F) this wil require a Unit shutdown lAW GP-05.
Distractor Analysis:
Choice A: Correct see explanation Choice B: Plausible because the 01 has a table with the Selected Out-of-Service Equipment Contingencies. In this case the FW heater meets the definition of the heater OOS and operation is permitted.
Choice C: Plausible because operation is allowed if the FW heaters are isolated but not with the final temperature greater than the 110.3 limit.
Choice D: Plausible because if turbine bypass is inoperable with a FW Heater OOS then TS 3.7.6 requires this action.
SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations (43(b)(5) a. Assessement of the plant conditions and then prescribing the shutdown lAW the GP.
Notes Feedback KIA: 295014 G2.01.25 Ability to interpret reference materials, such as graphs, curves, tables, etc.
Inadvertent Reactivity Addition (CFR: 41.10/43.5/45.12)
RO/SRO Rating: 3.9/4.2 Objective: CLS-LP-34, Obj. 11c Given plant conditions, describe the effect a loss/malfunction of the feedwater heaters wi" have on:
: c. Feedwater Temperature


==Reference:==
==Reference:==
 
20P-32, Attachment 4 (provided)
20P-32,   Attachment 44 (provided) 20P-32, Attachment          (provided)
Cog Level HI Explanation:
Cog Level Cog  Level HIHI Explanation:
From Attachment 4 of 20P-32 operation is outside of the a"owable range <<11 0.3&deg;F) this wil require a Unit shutdown lAW GP-05.
Explanation:
From Attachment 4 of 20P-32 operation is outside of the a"owable From                                                              allowable range <<11   0.3&deg;F) this wil require a
(<1 10.3&deg;F)
Unit shutdown lAW GP-05.
Distractor Analysis:
Distractor Analysis:
Choice A: Correct see explanation Choice B: Plausible because the 01 has a table with the Selected Out-of-Service Equipment Contingencies. In this case the FW heater meets the definition of the heater OOS and operation is permitted.
Choice A: Correct see explanation Choice B: Plausible because the 01 has a table with the Selected Out-of-Service Equipment Contingencies. In this case the FW heater meets the definition of the heater OOS and operation is permitted.
Choice C: Plausible because operation is allowed if the FW heaters are isolated but not with the final temperature greater than the 110.3 limit.
Choice C: Plausible because operation is allowed if the FW heaters are isolated but not with the final temperature greater than the 110.3 limit.
Choice 0:D: Plausible because if turbine bypass is inoperable with a FW Heater OOS then TS 3.7.6 requires this action.
Choice 0: Plausible because if turbine bypass is inoperable with a FW Heater OOS then TS 3.7.6 requires this action.
SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations (43(b)(5) a. Assessement of the plant conditions and then prescribing the shutdown lAW the GP.
SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations (43(b)(5) a. Assessement of the plant conditions and then prescribing the shutdown lAW the GP.
Notes
Notes  
 
U 0
C 1)
C C C4 CJ cN C c C C4 C C c4 - - - - - -
C C) C C) C) C C C C C CrJ j cl)c


LL 0                      a                                                          (4          -c                                                                        LL        U                                                a)                        >
From OP-32, Attachment 4, Final Feedwater Temperature Vs Power E
2                                                                                      (5 C)      E                a)          C                                            (5  a)  ci)                                        IE  ci) 0    I-                              Co        0  ci) I RX  Ck                            x                                                                                                                                                                                                                        -          C                0 U
Nominal            -
Nominal )            C
                                                                                                                                                                                                                                                          *1*1O.3&deg;F U    :
PWR                                                                                  FVV Temp                                                                                      RlV Temp.                                                c  Reduced F'vV cl)                                          ;
Reduced&#xe7;L) 0 Temp              1)
LL
LL
                                                                                                                                                                                                      '10&deg;F in in in in cc cc cc cc r r-100 C C C4 CJ cN C 429.0                                                                                      4'19.0                                                          328.7
* r*
: 0) CO I- W [C) t C) CI, 99                                                                                           427.6                                                                                     4'17.6                                                         327.7 cc Lfl 98                                                                                           426.5                                                                                     4'16 . 6                 .                                    C 327.0 r cc cc cc in in
j iZ D
: 00) 0) I-. W [C)
C CJ c c CJ CJ 0
                                                    ) C1.
C)
97                                                                                           425.5                                                                                     4'15.6                                                         326.3
&#xe7;L)
: 00) CO I-. CO [C) 96                                                                                           424 . 4 4']4.6                                                         325.7 95 t 1 423.4                                                                                     4'13.6                                                         325.0 94 C) C 422.4                                                                                     4'12.6                                                         324.3   c 93                                                                                           42'1.4 4'I'L7                                                         323.7 9,2 C C4 C C c4 420.4                                                                                     4'10..7                                                         323.0 91                                                                                   419.5                                                                                   t  409.8
 
* 322.4 ci - cc L -r c cJ r*
t 1 t 1
90                                                                                   4-18.5                                                                                     408.8                                                           321.7 cc in                a c  cc 1
* t i t t J f f C) C C (V) C C3 Cr)
j 89                                                                                   4'17.5                                                                                     407.9                                                           32'1.'1
Cr) cc Lfl in in in in cc cc cc cc r r-r cc cc cc in in ci - cc L -r c cJ
* t i iZ 88                                                                                   416.5                                                                                     406 .. 9                                                       320.4
 
: 0) 0) 0) 0)0) 0) 0) 0) 0) 0) CO CO CO CO CO CO CO CO CO CO - I 1 I - I C) C) C C C 87                                                                                   415.6                                                                                     406.0                     D                                    319.8 86                                                                                   4'14.6                                                                                     405.0                                                           319:1         -
cc in
t t J f f C) C C 85                                                                                   413.6                                                                                     404.1                                                           318.5 84                                                                                   4-12.6                                                                                     403 .. 1                                                       317.8         -
 
COCCCQCO C CJ c c 83                                                                                   41'1.7                                                                                     402.2                                                           317.2         -
a c
82                                                                                   4-10.7                                                                                     40'1.2                                                         316.5         -
 
81                                                                                   409.7                                                                                     400.3                                                           315.8 C C CrJ 80                                                                                   408.7                                                                                     399.3                                                           315.2 79                                                                                   407.6                                                                                     398.3                                                           314.5 rzrr 78                                                                                   406.6                                                                                     397.3                                                           313.8 (V)                                                        j 77                                                                                   405.6                                                                                     396.3                                                           313:1 C C3 Cr) 76                                                                                   404.5                                                                                     395.3                                                           312.4 CJ CJ 75                                                                                   403.5                                                                                     394.2                                                           31'1.7 74                                                                                   402.4 393.2                                                           311.0         -                           
cc
        -                            C)                                                                                                                                                  Cr) 73                                                                                   40'1.3                                                                                     392.1                                                         310.3                                    
 
COCCCQCO rzrr U
x 0)
CO I-W
[C) t C) CI,
: 00) 0) I-. W
[C)
) C1.
00)
CO I-.
CO
[C)
C)
Ck
: 0) 0) 0) 0)0) 0) 0) 0) 0) 0)
CO CO CO CO CO CO CO CO CO CO -
I 1 I - I I
ci) 0 Co>
I-ci)0 Ea)
I ci)a)
U (5
LL Ca)
E
-c C)
(5 (4
a0 E
2 LL From OP-32, Attachment 4, Final Feedwater Temperature Vs Power RX Nominal Nominal
*1*1O.3&deg;F PWR FVV Temp RlV Temp Reduced F'vV Reduced Temp
'10&deg;F 100 429.0 4'19.0 328.7 99 427.6 4'17.6 327.7 98 426.5 4'16.. 6 327.0 97 425.5 4'15.6 326.3 96 424.. 4 4']4.6 325.7 95 423.4 4'13.6 325.0 94 422.4 4'12.6 324.3 93 42'1.4 4'I'L7 323.7 9,2 420.4 4'10..7 323.0 91 419.5 409.8 322.4 90 4-18.5 408.8 321.7 89 4'17.5 407.9 32'1.'1 88 416.5 406.. 9 320.4 87 415.6 406.0 319.8 86 4'14.6 405.0 319:1 85 413.6 404.1 318.5 84 4-12.6 403.. 1 317.8 83 41'1.7 402.2 317.2 82 4-10.7 40'1.2 316.5 81 409.7 400.3 315.8 80 408.7 399.3 315.2 79 407.6 398.3 314.5 78 406.6 397.3 313.8 77 405.6 396.3 313:1 76 404.5 395.3 312.4 75 403.5 394.2 31'1.7 74 402.4 393.2 311.0 73 40'1.3 392.1 310.3  


CAUTION CAUTION Unit operation Unit  operation outside outside tile the bounds bounds of      of the the Loss Loss of  of Feedwater Feedwater Heating Heating analysis analysis isis prohibited.
CAUTION Unit operation outside the bounds of the Loss of Feedwater Heating analysis is prohibited.
prohibited.
9.
9.
: 9.      IF   Step 8.7.2.8.c IF step    8.7.2.8c criteria criteria is    NOT met, is NOT       met. THENTHEN PERFORM PERFORM the following:
IF Step 8.7.2.8c criteria is NOT met. THEN PERFORM the following:
the  following:
a.
a.
: a.        IMMEDIATELY NOTIFY IMMEDIATELY                 NOTIFY the            Unit seo.
IMMEDIATELY NOTIFY the Unit SCO b
the Unit     SCO b
RESTORE unit operation within the bounds of the cycle Loss of Feedwater Heating analysis OR c
: b.        RESTORE unit RESTORE                     operation within unit operation         within the the bounds bounds of   of the cycle the cycle Loss   Loss of  of Feedvvater Feedwater Heating Heating analysis analysis OR OR c
COMMENCE unit shutdown in accordance with OGP-05.
: c.        COMMENCE unit shutdown COMMENCE                        shutdown in      in accordance 'ljVith with OGP-05.
20P-32 Rev. 165 Page 117 of 300 Permitted Condition Operation Comment 005 SingI (See NOTES)
20P-32 120P-32                                                      Rev. 165 165                                      Page 1-17  117 of 300       1 Permitted Condition Condition                    Operation Operation           Comment 005 Single DOS  Sin gI (See NOTES)
FWHOOS Yes Defined as a ID F or greater reduction n nominal feedwater temperature.
FWHOOS FWHOOS FWR I,FFTR)
FWR FFrR Yes Defined as a 10 F or greater reduction n feecwater temperature.
FW,R    FFrR Yes Yes Defined as a 10&deg;F ID F or greater reduction in Defined as a 10 'F Delined n nom  inal feedwater temperature.
Defined as a cycle extension strategy.
nominal F or greater reduction inn feedwater feecwater temperature.
MSIVOOS Yes-base MSIVOOS permits I MSIV to be inoperable.
                                                *
IF MSIVOOS. THEN thermal power shall be limited to 70% of rated.
* Defined as a cycle extension strategy.
TBPCCS Yes TSPOOS assumes a! turbine bypass valves (ThV) are incoerable.
MSIVOOS MSIIfOOS                      Yes-base
SLO Yes Permitted with a thermal limi: penalty.
* MSIVOOS permits 1      I MSIV to be inoperable.
005 Combination (See NOTES)
* MSIVOOS. THEN thermal power shall be limited to 70% of rated.
TBPOOS & WH003 Yes Combined 003 condition is permtted with a thermal limit penalty.
IF MSIVOOS_
TBPOOS Yes Combined 003 condtion s permitted with a thermal imit penatty.
TBPCCS TepOOS                          Yes
& FWTR (FFTR)
* TSPOOS assumes all      a! turbine bypass valves (I      BV) are inoperable.
Operefinq Power/Flow Map ICF Yes-base Permitted operations with thermal limits defined by 003 condition.
(ThV)        incoerable.
Power Coas:down Yes-base Permitted operations with thermal limits detned by OCS condition.
SLO SLO ODS Combination (See 005 Yes (See NOTES)
Turbine Control Mode Yes-base Partial arc cpera:ion is supported by safety analysis for all 005 conditions RCR Pump Per Source Yes-base Power source can be protiided by the UAT or SAT or all ODS conditions.
                                                  **  Permitted with a thermal       limi: penalty.
ihemlallimii TBPOOS TSPOOS & WH003 TBPOOS TBPOOS
  && FWTR FlNHOOS              Yes Yes       **
* Combined 003 Combined 003 OOS condition is permtted condition s 005 condtion permitted with a thermal limit penalty.
is permitted with a thermal limit      penalty.
imit penatty.
FltHR (FFTR)
(FFTR)
Operefinq Operating Power/Flow Power/Flow Map Power Map'" ICF Power Coas:down Coastdown ICF     Yes-base Yes-base Permitted Pem,iited operations Permitted Pemlitted operations with thermal operations with operations with thermal limits with thermal limits defined thermal limits defined by limits detned by 003 defined by OOS condition.
by OCS condition.
OOS condition.
condition.
Turbine Control Turbine RCR Control Mode Pump Per RCR Pump Mode Pwr Source Source Yes-base Yes-base Yes-base Yes-base
                                                  **  Partial Partial arc Power aro cpera:ion Power source operation isis supported source can supported by be protiided can be  provided by by safety by the saier{ analysis the UAT UAT or analysis for or SAT SAT or for all for all all 005 all ODS OOS conditions conditions.
OOS conditions.
conditions.
Yes:
Yes:
Yes:                Operations are Operations      permitted with are permitted     with restrictive  thermal limits.
Operations are permitted with restrictive thermal limits.
restrictive thermal   limits.
Yes-base:
Yes-base:
Ye.s-base:        Operations Operations are are permitted permitted with with base   thermal limits.
Operations are permitted with base thermal limits. fo thermal flnit changes are required.
base thermal    limits. fo No thermal   flnit changes thermallirnit  changes are are required.
001-01.01 Rev. 29 Page 121 of 177 Categories K/A:
required.
295014 G2.01.25 Tier/Group:
001-01.01 1001-01.01                                                      Rev. 29 Rev. 29                                        Page 121 Page              of '1771 12-1 of  177 Categories Categories K/A:
T1G2 RO Rating:
KIA:            295014 295014 G2.01.25 G2.01.25                                           Tier / Group: T1G2 Tier/Group:            T1 G2 RORating:
3.9 SRO Rating:
RO Rating:       3.9 3.9                                                          SRO SRO Rating:
 
Rating: 4.2  4.2 LPObj:
===4.2 LPObj===
LPObj:          34-11C 34-11C                                                        Source:
34-11C Source:
Source:              NEW NEW Cog Cog Level:
NEW Cog Level:
Level:      HIGH HIGH                                                         Category 8:8:
HIGH Category 8:
Category
CAUTION Unit operation outside tile bounds of the Loss of Feedwater Heating analysis is prohibited.
: 9.
IF step 8.7.2.8.c criteria is NOT met, THEN PERFORM the following:
120P-32 Condition
: a.
IMMEDIATELY NOTIFY the Unit seo.
: b.
RESTORE unit operation within the bounds of the cycle Loss of Feedvvater Heating analysis OR
: c.
COMMENCE unit shutdown in accordance 'ljVith OGP-05.
Rev. 165 Page 1-17 of 300 1 Permitted Operation Comment DOS Single (See NOTES)
FWHOOS Yes Defined as a 10&deg;F or greater reduction in nom inal feedwater temperature.
FW,R I,FFTR)
Yes Delined as a 10 'F or greater reduction in feedwater temperature.
Defined as a cycle extension strategy.
MSIIfOOS Yes-base
* MSIVOOS permits 1 MSIV to be inoperable.
IF MSIVOOS_ THEN thermal power shall be limited to 70% of rated.
TepOOS Yes TSPOOS assumes all turbine bypass valves (I BV) are inoperable.
SLO Yes Permitted with a ihemlallimii penalty.
ODS Combination (See NOTES)
TSPOOS & FlNHOOS Yes Combined OOS condition is permitted with a thermal limit penalty.
TBPOOS Yes Combined 005 condition is permitted with a thermal limit penalty.
& FltHR (FFTR)
Operating Power/Flow Map'" ICF Yes-base
* Pem,iited operations with thermal limits defined by OOS condition.
Power Coastdown Yes-base
* Pemlitted operations with thermal limits defined by OOS condition.
Turbine Control Mode Yes-base
* Partial aro operation is supported by saier{ analysis for all OOS conditions.
RCR Pump Pwr Source Yes-base
* Power source can be provided by the UA T or SAT for all OOS conditions.
Yes:
Operations are permitted with restrictive thermal limits.
Ye.s-base:
Operations are permitted with base thermal limits. No thermallirnit changes are required.
1001-01.01 Categories KIA:
RORating:
LPObj:
Cog Level:
295014 G2.01.25 3.9 34-11C HIGH Rev. 29 Tier / Group: T1 G2 SRO Rating:
 
===4.2 Source===
NEW Category 8:
Page 12-1 of '1771


8.7.2       Procedural Steps                                             Initials CAUTION Opening the Feedwater Heater tube side vents will release hot discharges under pressure to the drain trough.
8.7.2 Procedural Steps Initials CAUTION Opening the Feedwater Heater tube side vents will release hot discharges under pressure to the drain trough.
z.
z.
Z.      PERFORM the following to vent the tube side of the 4A(B) feed water heater:
PERFORM the following to vent the tube side of the 4A(B) feed water heater:
FEED WA TER HEATER
OPEN FEEDWA TER HEA TER 4A (B)
                          - OPEN FEEDWATER
CHANNEL INBOARD VENT VALVE, MVD-V69(V76).
                          -                        HEA TER 4A(B) 4A (B)
CRACK OPEN FEED WA TER HEA TER 4A(B) OUTBOARD CHANNEL VENT VALVE, MVD-V70(V75), to establish a vent path.
CHANNEL INBOARD VENT VAL    VALVE, VE, MVD- V69(V76).
aa.
MVD-V69(V76).
PERFORM the following to vent the tube side of the 5A(B) feed water heater:
FEED WA TER HEATER
OPEN FEED WA TER HEA TER 5A(B)
                          - CRACK OPEN FEEDWATER
CHANNEL INBOARD VENT VALVE, MVD-V8 I (V88).
                          -                                HEA TER 4A(B) OUTBOARD CHANNEL VENT VALVE, MVD-V70(V75), to establish a vent path.
CRACK OPEN FEED WA TER HEA TER 5A(B) CHANNEL OUTBOARD VENT VALVE, MVD-V82(V87), to establish a vent path.
aa. PERFORM the following to vent the tube side of the 5A(B) feed water heater:
NOTE:
HEA TER 5A(B)
Step 8.7.2.8 ensures unit operation with reduced feedwater temperature is bounded by the cycle Loss of Feedwater Heating analysis.
FEED WA TER HEATER
8.
                          - OPEN FEEDWATER
EVALUATE reduction in final feedwater temperature for compliance with Loss of Feedwater Heating analysis as follows:
                          -                                5A (B)
a.
CHANNEL INBOARD VENT VALVE, MVD- V8 I (V88).
RECORD current final feedwater temperature from PPC Display 825.
MVD-V81
20P-32 Rev. 166 Page 116 of 301 8.7.2 Procedural Steps CAUTION Opening the Feedwater Heater tube side vents will release hot discharges under pressure to the drain trough.
                          - CRACK OPEN FEEDWATER
Z.
                          -                                HEA TER FEED WA TER HEATER 5A(B) CHANNEL OUTBOARD VENT VALVE, MVD-V82(V87), to establish a vent path.
PERFORM the following to vent the tube side of the 4A(B) feed water heater:
NOTE:     Step 8.7.2.8 ensures unit operation with reduced feedwater temperature is bounded by the cycle Loss of Feedwater Heating analysis.
OPEN FEEDWATER HEATER 4A(B)
: 8. EVALUATE reduction in final feedwater temperature for compliance with Loss of Feedwater Heating analysis as follows:
CHANNEL INBOARD VENT VAL VE, MVD-V69(V76).
: a.     RECORD current final feedwater temperature from PPC Display 825.
CRACK OPEN FEEDWATER HEATER 4A(B) OUTBOARD CHANNEL VENT VALVE, MVD-V70(V75), to establish a vent path.
Sllj     of.
aa.
20P-32 120P-32                                  Rev. 166                     Page 116 of 301 I
PERFORM the following to vent the tube side of the 5A(B) feed water heater:
OPEN FEEDWATER HEATER 5A(B)
CHANNEL INBOARD VENT VALVE, MVD-V81 (V88).
CRACK OPEN FEEDWATER HEATER 5A(B) CHANNEL OUTBOARD VENT VALVE, MVD-V82(V87), to establish a vent path.
Initials NOTE:
Step 8.7.2.8 ensures unit operation with reduced feedwater temperature is bounded by the cycle Loss of Feedwater Heating analysis.
120P-32
: 8.
EVALUATE reduction in final feedwater temperature for compliance with Loss of Feedwater Heating analysis as follows:
: a.
RECORD current final feedwater temperature from PPC Display 825.
Sllj of.
Rev. 166 Page 116 of 301 I  


8.7.2 8.7.2      Procedural Steps Procedural                                                      Initials Initials b.
8.7.2 Procedural Steps Initials b.
: b.      RECORD 110.3&deg;F RECORD      110.3&deg;F Reduced Reduced FFWT FFWT value for current reactor current  reactor power power from Attachment 4.
RECORD 110.3&deg;F Reduced FFWT value for current reactor power from Attachment 4.
OF
c.
: c.     CONFIRM reduction in final feedwater temperature is less than 11 110.3&deg;F 0.3&deg;F by comparing the following:
CONFIRM reduction in final feedwater temperature is less than 110.3&deg;F by comparing the following:
3/)( .        OF    >
8.7.2.8.a 8.7.2.8.b CAUTION Unit operation outside the bounds of the Loss of Feedwater Heating analysis is prohibited.
B.7.2.B.a 8.7.2.8.a                     B.7.2.B.b 8.7.2.8.b CAUTION Unit operation outside the bounds of the Loss of Feedwater Heating analysis is prohibited.
9.
: 9. IF Step B.7.2.B.c 8.7.2.8.c criteria is NOT met, THEN PERFORM the following:
IF Step 8.7.2.8.c criteria is NOT met, THEN PERFORM the following:
: a.     IMMEDIATELY NOTIFY the Unit CRS.
a.
: b.     RESTORE unit operation within the bounds of the cycle Loss of Feedwater Heating analysis qj
IMMEDIATELY NOTIFY the Unit CRS.
: c.     COMMENCE unit shutdown in accordance with OGP-05.
b.
RESTORE unit operation within the bounds of the cycle Loss of Feedwater Heating analysis qj c.
COMMENCE unit shutdown in accordance with OGP-05.
10.
10.
: 10. IF feedwater temperature is more than 10&deg;F 10&deg;F below nominal (refer to Attachment 4), AND reactor power is greater than or equal to 30% of rated thermal power, THEN PERFORM the following:
IF feedwater temperature is more than 10&deg;F below nominal (refer to Attachment 4), AND reactor power is greater than or equal to 30% of rated thermal power, THEN PERFORM the following:
: a.     ENSURE reactor operation in accordance with applicable FWTR Power to Flow Map.
a.
: b.     REFER to 201-03.2 for required actions.
ENSURE reactor operation in accordance with applicable FWTR Power to Flow Map.
20P-32 120P-32                                     Rev. 166 166                    Page 117 117 of of 301 301 I
b.
REFER to 201-03.2 for required actions.
20P-32 Rev. 166 Page 117 of 301 8.7.2 Procedural Steps
: b.
RECORD 110.3&deg;F Reduced FFWT value for current reactor power from Attachment 4.
: c.
OF CONFIRM reduction in final feedwater temperature is less than 11 0.3&deg;F by comparing the following:
3/)(
OF B.7.2.B.a B.7.2.B.b CAUTION Unit operation outside the bounds of the Loss of Feedwater Heating analysis is prohibited.
: 9.
IF Step B.7.2.B.c criteria is NOT met, THEN PERFORM the following:
: a.
IMMEDIATELY NOTIFY the Unit CRS.
: b.
RESTORE unit operation within the bounds of the cycle Loss of Feedwater Heating analysis
: c.
COMMENCE unit shutdown in accordance with OGP-05.
: 10.
IF feedwater temperature is more than 10&deg;F below nominal (refer to Attachment 4), AND reactor power is greater than or equal to 30% of rated thermal power, THEN PERFORM the following:
: a.
ENSURE reactor operation in accordance with applicable FWTR Power to Flow Map.
: b.
REFER to 201-03.2 for required actions.
Initials 120P-32 Rev. 166 Page 117 of 301 I  


8.7.2     Procedural Steps                                               Initials
8.7.2 Procedural Steps Initials 11.
: 11. CONFIRM feedwater flow temperature compensation is accurate by performing the following:
CONFIRM feedwater flow temperature compensation is accurate by performing the following:
NOTE:     Feedwater Line A temperature can be obtained from any of the following:
NOTE:
U2CP_B050 PPC Point U2CP    B050 U2CP_B051 PPC Point U2CP    B051 Feedwater Lines Temperature Recorder, 821-B21-TR-5515 (20 el. Reactor TR-5515 (20' Building)
Feedwater Line A temperature can be obtained from any of the following:
: a. DETERMINE Feedwater Line A temperature AND RECORD temperature and instrument used below:
PPC Point U2CP_B050 PPC Point U2CP_B051 Feedwater Lines Temperature Recorder, B21-TR-5515 (20 el. Reactor Building) a.
OF -------------
DETERMINE Feedwater Line A temperature AND RECORD temperature and instrument used below:
FW Line A temp           Instrument NOTE:    Feedwater Line BB temperature can be obtained from any of the following:
FW Line A temp Instrument NOTE:
U2CP_B052 PPC Point U2CP   B052 U2CP_B053 PPC Point U2CP   B053 B21TR-5515, (20' Feedwater Lines Temperature Recorder, 821-TR-5515,     (20 el. Reactor Building)
Feedwater Line B temperature can be obtained from any of the following:
: b. DETERMINE Feedwater Line B temperature AND RECORD temperature and instrument used below:
PPC Point U2CP_B052 PPC Point U2CP_B053 Feedwater Lines Temperature Recorder, B21TR-5515, (20 el. Reactor Building) b.
                        ---------------OF -------------
DETERMINE Feedwater Line B temperature AND RECORD temperature and instrument used below:
FW Line BB temp         Instrument
FW Line B temp Instrument c.
: c. OBTAIN Feedwater Line A feedwater flow compensation value using Feedwater Line A temperature recorded in Step 8.7.2.11.a and Attachment 9 AND RECORD on Attachment 10, column 1.
OBTAIN Feedwater Line A feedwater flow compensation value using Feedwater Line A temperature recorded in Step 8.7.2.11.a and AND RECORD on Attachment 10, column 1.
20P-32 120P-32                                Rev. 166                             1 18 of 301 Page 118            I
20P-32 Rev. 166 Page 1 18 of 301 8.7.2 Procedural Steps Initials NOTE:
NOTE:
120P-32
: 11.
CONFIRM feedwater flow temperature compensation is accurate by performing the following:
Feedwater Line A temperature can be obtained from any of the following:
PPC Point U2CP B050 PPC Point U2CP B051 Feedwater Lines Temperature Recorder, 821-TR-5515 (20' el. Reactor Building)
: a.
DETERMINE Feedwater Line A temperature AND RECORD temperature and instrument used below:
OF FW Line A temp Instrument Feedwater Line B temperature can be obtained from any of the following:
PPC Point U2CP B052 PPC Point U2CP B053 Feedwater Lines Temperature Recorder, 821-TR-5515, (20' el. Reactor Building)
: b.
DETERMINE Feedwater Line B temperature AND RECORD temperature and instrument used below:
OF FW Line B temp Instrument
: c.
OBTAIN Feedwater Line A feedwater flow compensation value using Feedwater Line A temperature recorded in Step 8.7.2.11.a and AND RECORD on Attachment 10, column 1.
Rev. 166 Page 118 of 301 I  


8.7.2       Procedural Steps                                             Initials
8.7.2 Procedural Steps Initials d.
: d. OBTAIN Feedwater Line B feedwater flow compensation value using Feedwater Line B 8.7.2.11.b temperature recorded in Step 8.7 .2.11.b and Attachment 9 AND RECORD on Attachment 10, column 1.
OBTAIN Feedwater Line B feedwater flow compensation value using Feedwater Line B temperature recorded in Step 8.7.2.11.b and AND RECORD on Attachment 10, column 1.
NOTE:     Process Computer compensation values are located on the second page of the C32B022/B023 screen under HANDLING PARAMETERS, CORRECTION TYPE FACTOR O.        0.
NOTE:
: e. OBTAIN CORRECTION TYPE FACTOR 0 compensation value for PPC Point U2C32B022 (Feedwater Line A) AND RECORD on Attachment 10, column 2.
Process Computer compensation values are located on the second page of the C32B022/B023 screen under HANDLING PARAMETERS, CORRECTION TYPE FACTOR 0.
: f. OBTAIN CORRECTION TYPE FACTOR 0 compensation value for PPC Point U2C32B023 (Feedwater Line B) AND RECORD in Attachment 10, column 2.
e.
NOTE:     IF the values compared in the following step are within 0.002, THEN feedwater flow temperature compensation is accurate.
OBTAIN CORRECTION TYPE FACTOR 0 compensation value for PPC Point U2C32B022 (Feedwater Line A) AND RECORD on 0, column 2.
: g. VERIFY the values on Attachment 10, columns 1 and 2 for Feedwater Line A are within 0.002 1
f.
AND DOCUMENT on Attachment 10.
OBTAIN CORRECTION TYPE FACTOR 0 compensation value for PPC Point U2C32B023 (Feedwater Line B) AND RECORD in 0, column 2.
: h. VERIFY the values on Attachment 10, 10, columns I and 2 for Feedwater Line B 1                          B are within 0.002 AND DOCUMENT on Attachment 10. 10.
NOTE:
: i. IF feedwater flow temperature compensation is NOT accurate, THEN IMMEDIATELY NOTIFY the duty Reactor Engineer.
IF the values compared in the following step are within 0.002, THEN feedwater flow temperature compensation is accurate.
2OP-32
g.
\20P-32                                Rev. 166 166                      Page 119 119 of 301 \
VERIFY the values on Attachment 10, columns 1 and 2 for Feedwater Line A are within 0.002 AND DOCUMENT on Attachment 10.
: 86. While
h.
: 86. While inin Mode Mode 33 with with Shutdown Shutdown Cooling Cooling (SDC)
VERIFY the values on Attachment 10, columns I and 2 for Feedwater Line B are within 0.002 AND DOCUMENT on Attachment 10.
(SDC) inin service service onon Unit Unit One, One, aa complete complete Loss of Loss    of Off-site Off-site Power Power (LOOP)
i.
(LOOP) occurs.
IF feedwater flow temperature compensation is NOT accurate, THEN IMMEDIATELY NOTIFY the duty Reactor Engineer.
occurs.
2OP-32 Rev. 166 Page 119 of 301 8.7.2 Procedural Steps
The 1-E11-F009, The    1-E11-F009, RHR  RHR Shutdown Shutdown Cooling Cooling Inboard Inboard Isolation Isolation Valve, Valve, mechanically mechanically binds binds in aa mid-position in      mid-position andand cannot cannot be be fully fully opened.
: d.
opened.
OBTAIN Feedwater Line B feedwater flow compensation value using Feedwater Line B temperature recorded in Step 8.7.2.11.b and AND RECORD on Attachment 10, column 1.
Which one Which          of the one of   the following following isis the the minimum minimum levellevel required required toto support support natural natural circulation circulation and identifies and     identifies the the procedural procedural method method for      Decay Heat for Decay   Heat removal removal that that isis available?
Initials NOTE:
available?
Process Computer compensation values are located on the second page of the C32B022/B023 screen under HANDLING PARAMETERS, CORRECTION TYPE FACTOR O.
The minimum The    minimum Reactor Reactor Water Water Level Level toto support support Natural Natural Circulation Circulation is is     (1)
: e.
(1)    inches.
OBTAIN CORRECTION TYPE FACTOR 0 compensation value for PPC Point U2C32B022 (Feedwater Line A) AND RECORD on 0, column 2.
inches.
: f.
The available The    available method method of  of decay decay heat heat removal removal is is (2)(2)
OBTAIN CORRECTION TYPE FACTOR 0 compensation value for PPC Point U2C32B023 (Feedwater Line B) AND RECORD in 0, column 2.
A. (1)(1) 200 200 (2) Alternate (2) Alternate Decay Decay Heat Heat Removal Removal UsingUsing Natural Natural Circulation Circulation and and FPCCS FPCCS and and SSFPC     lAW   lOP-I  7,   Residual     Heat SSFPC lAW 1OP-17, Residual Heat Removal System    Removal    System Operating Operating Procedure Procedure B(I)
NOTE:
B~   (1) 200 (2)
IF the values compared in the following step are within 0.002, THEN feedwater flow temperature compensation is accurate.  
(2) Alternate Shutdown Cooling Cooling lAW lAW OAOP-15.0, OAOP-l 5.0, Loss Loss of Shutdown Shutdown Cooling Cooling C. (1) 254 (2) Alternate Decay Heat Removal Using Natural Circulation and FPCCS and SSFPC lAW 10P-17,lOP-i 7, Residual Heat Removal System Operating Procedure D. (1) 254 (2) Alternate Shutdown Cooling lAW OAOP-15.0, Loss of Shutdown Cooling
\\20P-32
: g.
VERIFY the values on Attachment 10, columns 1 and 2 for Feedwater Line A are within 0.002 AND DOCUMENT on Attachment 10.
: h.
VERIFY the values on Attachment 10, columns 1 and 2 for Feedwater Line B are within 0.002 AND DOCUMENT on Attachment 10.
: i.
IF feedwater flow temperature compensation is NOT accurate, THEN IMMEDIATELY NOTIFY the duty Reactor Engineer.
Rev. 166 Page 119 of 301 \\
: 86. While in Mode 3 with Shutdown Cooling (SDC) in service on Unit One, a complete Loss of Off-site Power (LOOP) occurs.
The 1-E11-F009, RHR Shutdown Cooling Inboard Isolation Valve, mechanically binds in a mid-position and cannot be fully opened.
Which one of the following is the minimum level required to support natural circulation and identifies the procedural method for Decay Heat removal that is available?
The minimum Reactor Water Level to support Natural Circulation is (1) inches.
The available method of decay heat removal is (2)
A.
(1) 200 (2) Alternate Decay Heat Removal Using Natural Circulation and FPCCS and SSFPC lAW lOP-I 7, Residual Heat Removal System Operating Procedure B(I) 200 (2) Alternate Shutdown Cooling lAW OAOP-l 5.0, Loss of Shutdown Cooling C. (1) 254 (2) Alternate Decay Heat Removal Using Natural Circulation and FPCCS and SSFPC lAW lOP-i 7, Residual Heat Removal System Operating Procedure D. (1) 254 (2) Alternate Shutdown Cooling lAW OAOP-15.0, Loss of Shutdown Cooling
: 86. While in Mode 3 with Shutdown Cooling (SDC) in service on Unit One, a complete Loss of Off-site Power (LOOP) occurs.
The 1-E11-F009, RHR Shutdown Cooling Inboard Isolation Valve, mechanically binds in a mid-position and cannot be fully opened.
Which one of the following is the minimum level required to support natural circulation and identifies the procedural method for Decay Heat removal that is available?
The minimum Reactor Water Level to support Natural Circulation is (1) inches.
The available method of decay heat removal is (2)
A. (1) 200 (2) Alternate Decay Heat Removal Using Natural Circulation and FPCCS and SSFPC lAW 1 OP-17, Residual Heat Removal System Operating Procedure B~ (1) 200 (2) Alternate Shutdown Cooling lAW OAOP-15.0, Loss of Shutdown Cooling C. (1) 254 (2) Alternate Decay Heat Removal Using Natural Circulation and FPCCS and SSFPC lAW 10P-17, Residual Heat Removal System Operating Procedure D. (1) 254 (2) Alternate Shutdown Cooling lAW OAOP-15.0, Loss of Shutdown Cooling  


Feedback Feedback KIA: S295021 KIA:  S295021 A2.03 A2.03 Ability to Ability    determine and/or to determine    andlor interpret interpret the following as the following    as they they apply apply to to LOSS LOSS OFOF SHUTDOWN SHUTDOWN COOLING:
Feedback KIA: S295021 A2.03 Ability to determine andlor interpret the following as they apply to LOSS OF SHUTDOWN COOLING:
COOLING:
Reactor water level (CFR: 41.10/43.5/45.13)
Reactor water Reactor          level water level (CFR: 41.10/43.5/45.13)
ROISRO Rating: 3.5/3.5 Objective: CLS-LP-1 20*06
(CFR:   41.10/43.5/45.13)
: 6. Describe how to determine when natural circulation exists within the Reactor Vessel.
ROISRO Rating:
RO/SRO    Rating: 3.5/3.5 3.5/3.5 CLS-LP-1 20*06 Objective: CLS-LP-120*06 Objective:
: 6. Describe
: 6. Describe how how to to determine determine when when natural natural circulation circulation exists exists within within the the Reactor Reactor Vessel.
Vessel.


==Reference:==
==Reference:==
OAOP-15, Revision 23, Page 11, Section 3.2.14 Cog Level: High Explanation:
During conditions in which there is no circulation, the reactor vessel water level, as read on B21-LI-R605A(B), should be maintained between 200 and 220, or as directed by the Shift Superintendent based on plant conditions, until forced circulation is restored. With a LOOP present and no actions taken to restore Off-site power (not provided in the question), the only available means of decay heat removal is alternate shutdown cooling utilizing SRVs.
Distractor Analysis:
Choice A: Plausible because 200 inches is correct and OP-17 contains actions for using FPC and SSFPC, but these would only be available if the reactor head is removed and fuel pool gates removed.
Choice B: Correct Answer Choice C: Plausible because 254 inches is the level of the MSLs and could be confused with Natural Circulation level due to the requirement to be at this level during alternate SDC and OP-17 contains actions for using FPC and SSFPC, but these would only be available if the reactor head is removed and fuel pool gates removed.
Choice D: Plausible because Plausible because 254 inches is the level of the MSLs and could be confused with Natural Circulation level due to the requirement to be at this level during alternate SDC SRO Only Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations (43(b)(5)a. Requires assessing plant conditions (LOOP, Mode 3, power availability, impact of power losses) and prescribing correct section of a procedure to provide DHR.
Notes Feedback KIA: S295021 A2.03 Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING:
Reactor water level (CFR: 41.10/43.5/45.13)
RO/SRO Rating: 3.5/3.5 Objective: CLS-LP-120*06
: 6. Describe how to determine when natural circulation exists within the Reactor Vessel.


OAOP-15, Revision 23, Page Page 11, 11, Section 3.2.14 3.2.14 Cog Level:
==Reference:==
Cog   Level: High High Explanation:
OAOP-15, Revision 23, Page 11, Section 3.2.14 Cog Level: High Explanation:
During conditions in which there is no circulation, the reactor vessel water level, as read on B21-LI-R605A(B), should be maintained between 200" 821-Ll-R605A(8),                                      200 and 220",
During conditions in which there is no circulation, the reactor vessel water level, as read on 821-Ll-R605A(8), should be maintained between 200" and 220", or as directed by the Shift Superintendent based on plant conditions, until forced circulation is restored. With a LOOP present and no actions taken to restore Off-site power (not provided in the question), the only available means of decay heat removal is alternate shutdown cooling utilizing SRVs.
220, or as directed by the Shift Superintendent based on plant conditions, until forced circulation is restored. With a LOOP present and no actions taken to restore Off-site power (not provided in the question), the only available means of decay heat removal is alternate shutdown cooling utilizing SRVs.
Distractor Analysis:
Distractor Analysis:
Choice A: Plausible because 200 inches is correct and OP-17 contains actions for using FPC and SSFPC, but these would only be available if the reactor head is removed and fuel pool gates removed.
Choice A: Plausible because 200 inches is correct and OP-17 contains actions for using FPC and SSFPC, but these would only be available if the reactor head is removed and fuel pool gates removed.
Choice B: Correct Answer Choice C: Plausible because 254 inches is the level of the MSLs and could be confused with Natural Circulation level due to the requirement to be at this level during alternate SDC and OP-17 contains actions for using FPC and SSFPC, but these would only be available if the reactor head is removed and fuel pool gates removed.
Choice B: Correct Answer Choice C: Plausible because 254 inches is the level of the MSLs and could be confused with Natural Circulation level due to the requirement to be at this level during alternate SDC and OP-17 contains actions for using FPC and SSFPC, but these would only be available if the reactor head is removed and fuel pool gates removed.
Choice D: Plausible because Plausible because 254 inches is the level of the MSLs and could be confused with Natural Circulation level due to the requirement to be at this level during alternate SDC SRO Only Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency emergency situations (43(b)(5)a. Requires assessing plant conditions (LOOP, Mode 3, power availability, impact of power losses) and prescribing correct section of a procedure to provide DHR.
Choice D: Plausible because Plausible because 254 inches is the level of the MSLs and could be confused with Natural Circulation level due to the requirement to be at this level during alternate SDC SRO Only Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations (43(b)(5)a. Requires assessing plant conditions (LOOP, Mode 3, power availability, impact of power losses) and prescribing correct section of a procedure to provide DHR.
Notes
Notes  
 
2.0 AUTOMATIC ACTIONS
 
Loop A(B) INBOARD INJECTION V4LVE E11-FO15.4iB, will close (Low Level One Only)


2.0 2.0       AUTOMATIC ACTIONS AUTOMATIC         ACTIONS Loop /\(8)
The RHR Pump in service for Shutdown Cooling will trip on a loss of suction path.
Loop    A(B) INBOARD INBOARD INJECTION INJECTION 1I.4L V4LVE VE,                       0 E11-FO15.4iB, will E11-F015A(B),       will close close (Low (Low Level Level One One Only)
3.0 OPERATOR ACTIONS 3,1 Immediate Actions None 3.2 Supplementary Actions ii 3.2.1 IF Shutdown Cooling has been lost due to a tripped RHR
Only)
[]
The RHR The    RHR Pump Pump inin service service for for Shutdown Shutdown Cooling will trip Cooling will trip        0 on aa loss on      loss of of suction suction path.
Pump, THEN START an RHR Pump in the loop being used for Shutdown Cooling.
path.
NOTE:
3.0       OPERATOR ACTIONS OPERATOR          ACTIONS 3,1 3.1       Immediate Actions Immediate None None 3.2       Supplementary Actions ii                                                  CAUTION CAUT.lON IfIf reactor coolant temperature ten,pera:ire is greater grea:er tllan   212rF and than 21.2&deg;F and reactor reac:or water level has been l:ieen raised to greater than than 218 212 inches foror 10 minutes minLtes or more, more, a false RPV low level signal could result when the reference leg condensing  condensing pot N12A(B)
During conditions in which there is no circulation, the reactor vessel water level, as read on 82f-LI-R6OA8, should be maintained between 200 and 220, or as directed by the Shift Superintendent based on plant conditions.
N12A(B: nozzle n3zzie is uncovered as subsequentl lowered below 218 level is subsequently                            21S inches.
until forced circulation is restored.
inches.
3.2.2 IF forced circulation has been lost, AND natural circulation has NOT been established, THEN RESTORE AND MAINTAIN reactor vessel water level.
3.2.1         IF Shutdown Cooling has been lost due to a tripped RHR Pump, THEN START an RHR Pump in the loop being o[]
OAOP-15.0 Rev. 23 Page 3 of 21 CAUTION If reactor coolant ten,pera:ire is grea:er than 212rF and reac:or water level has l:ieen raised to greater than 212 inches or 10 minLtes or more, a false RPV low level signal could result when the reference leg condensing pot N12A(B: n3zzie is uncovered as level is subsequentl lowered below 21S inches.
used for Shutdown Cooling.
2.0 AUTOMATIC ACTIONS Loop /\\(8) INBOARD INJECTION 1I.4L VE, 0
NOTE:           During conditions in which there is no circulation, the reactor vessel water 82f-LI-R6OA8, should be maintained between level, as read on B21-Ll-R605A(B),                                 between 200" 200 and 220, or as directed by the Shift Superintendent based on plant conditions, 220~,                                                                      conditions.
E11-F015A(B), will close (Low Level One Only)
until rorced unol  forced circulation is restored.
The RHR Pump in service for Shutdown Cooling will trip 0
3.2.2         IF forced circulation has been lost, AND natural                       0 circulation has NOT been established, THEN RESTORE AND MAINTAIN reactor vessel water level.
on a loss of suction path.
OAOP-15.0 IOAOP-15.0                                             Rev. 23                               Page Page 33 of 21 of21 I
3.0 OPERATOR ACTIONS 3.1 Immediate Actions None 3.2 Supplementary Actions CAUT.lON If reactor coolant temperature is greater tllan 21.2&deg;F and reactor water level has been raised to greater than 218 inches for 10 minutes or more, a false RPV low level signal could result when the reference leg condensing pot N12A(B) nozzle is uncovered as level is subsequently lowered below 218 inches.
3.2.1 IF Shutdown Cooling has been lost due to a tripped RHR Pump, THEN START an RHR Pump in the loop being used for Shutdown Cooling.
o NOTE:
During conditions in which there is no circulation, the reactor vessel water level, as read on B21-Ll-R605A(B), should be maintained between 200" and 220~, or as directed by the Shift Superintendent based on plant conditions, unol rorced circulation is restored.
3.2.2 IF forced circulation has been lost, AND natural 0
circulation has NOT been established, THEN RESTORE AND MAINTAIN reactor vessel water level.
IOAOP-15.0 Rev. 23 Page 3 of21 I  


10 3.0  OPERATOR ACTIONS OPERATOR          ACTIONS j.j,       IF the IF  the reactor reactor coolant coolant temperature temperature isis less less than than 212&deg;F, THEN 212&deg;F,    THEN ENSURE ENSURE the the following following valves valves are are open:
10 OPERATOR ACTIONS j,
open:
IF the reactor coolant temperature is less than 212&deg;F, THEN ENSURE the following valves are open:
                                    -    INBOARD RX iNBOARD      RX HE.4D HE4D VENT VENT VLV, VLV. B21-F003 821-F003             0
INBOARD RX HE4D VENT VLV. 821-F003 OUTBOARD PXHEAD VENT VLY. B21-F004.
                                    -    OUTBOARD RX OUTBOARD       PXHEAD HEAD VENT VENT VLV VLY. B21-F004.
k.
B21-F004.         0U k.k.       MAINTAIN RHR in MAINTAIN            in Shutdown Cooling accordance with 11(2)OP-17.
MAINTAIN RHR in Shutdown Cooling in accordance with 1(2)OP-17.
accordance           (2)OP-17.
IF RHR has NOT been restored in accordance with Step 3.2.11.5, THEN PLACE the RHR loop that was operating in Shutdown Cooling back in service in accordance with I (2)OP-1 7 as soon as conditions permit.
in Cooling in                   oU
U U
                  &       IFIF RHR RHR has has NOT 3.2.11.5, THEN 3.2.11.5, NOT been THEN PLACE been restored restored in PLACE the RHR in accordance RHR loop accordance with StepStep loop that was operating oU in   Shutdown Cooling in Shutdown              back in Cooling back  in service in in accordance with (2)OP-1 7 as soon as conditions permit.
U 3.2.12 IF necessary to minimize reactor coolant temperature rise, THEN PERFORM one of the following feed and I
1I (2)OP-17 3.2.12 IF necessary to minimize reactor coolant temperature IF rise,, THEN PERFORM one of the following feed and rise oU Not AvaD (LOOP)P) II bleed combinations:
bleed combinations:
I~ot Not Avail (RPS notJ not reset)
FEED BLEED CONDIFW in accordance with RWCU Reject in accordance 1(2)OP-32 with 1(2)OP-14 CRD in accordance with Reactor Water Level Control 1(2)OP-08 using Main Steam Lines in accordance with 1 (2)OP-32.
FEED                                   BLEED CONDIFW in accordance with RWCU Reject in accordance CONDJFW 1 (2)OP-32 1(2)OP-32                              with 1(2)OP-14 CRD in accordance with                   Reactor Water level Level Control 11(2)OP-08 (2)OP-08                               using Main Steam Lines in accordance with 11 (2)OP-32.
Core Spray in accordance Maintaining RPV Level Using with 1(2)OP-18 the Main Steam Line Drains LPCI in accordance with with 1(2)OP-25.
Core Spray in accordance                 Maintaining RPV Level Using with 1(2}OP-18 1(2)OP-18 LPCI in accordance with the Main Steam Line Drains with 1 (2)OP-25.
1(2)OP-1 7 IF NEITHER RHR loop can be placed in Shutdown Cooling, THEN PLACE the Condensate System in Condenser Cooling in accordance with 1(2)OP-32.
1(2)OP-25.
OAOP-15.0 I
I jvail (LOOP) I 1(2)OP-17 1(2)OP-1 7 3-23              IF NEITHER RHR loop can lJe       be placed in Shutdown                     0 U
Rev.23 Page 10 of 21 Not AvaD (LOOP) jvail (LOOP)
Cooling, THEN PLACE the Condensate System in Condenser Cooling in accordance with        "vith 1(2}OP-32.
I 3-23 U
1(2)OP-32.
Not Avail (RPS not reset)
OAOP-15.0 IOAOP-15.0                           I            Rev.23 Rev. 23                                 Page 10 Page   10 of of21 21 I
U 3.0 OPERATOR ACTIONS
: j.
IF the reactor coolant temperature is less than 212&deg;F, THEN ENSURE the following valves are open:
iNBOARD RX HE.4D VENT VLV, B21-F003 0
OUTBOARD RX HEAD VENT VLV B21-F004.
0
: k.
MAINTAIN RHR in Shutdown Cooling in accordance with 1 (2)OP-17.
IF RHR has NOT been restored in accordance with Step 3.2.11.5, THEN PLACE the RHR loop that was operating in Shutdown Cooling back in service in accordance with 1 (2)OP-17 as soon as conditions permit.
o o
3.2.12 IF necessary to minimize reactor coolant temperature rise THEN PERFORM one of the following feed and o
P) I bleed combinations:
I~ot Avail (RPS notJ reset)
FEED BLEED CONDJFW in accordance with RWCU Reject in accordance 1 (2)OP-32 with 1(2)OP-14 CRD in accordance with Reactor Water level Control 1 (2)OP-08 using Main Steam Lines in accordance with 1 (2)OP-32.
Core Spray in accordance Maintaining RPV Level Using I with 1(2}OP-18 the Main Steam Line Drains LPCI in accordance with with 1 (2)OP-25.
1(2)OP-17 IF NEITHER RHR loop can lJe placed in Shutdown 0
Cooling, THEN PLACE the Condensate System in Condenser Cooling in accordance "vith 1(2}OP-32.
IOAOP-15.0 Rev. 23 Page 10 of21 I  


3.0 3.0    OPERATOR ACTIONS OPERATOR      ACTIONS 3.2.14 3.2.14    IFIF ALL ALL of of the the allove above methods methods cancan NOTNOT maintain maintain reactor reactor vessel coolant vessel   coolant temperature temperature below below 212&deg;F, 212&deg;F, THEN THEN INITIATE alternate INITIATE      alternate Shutdown Shutdown Cooling Cooling with.
3.0 OPERATOR ACTIONS 3.2.14 IF ALL of the above methods can NOT maintain reactor vessel coolant temperature below 212&deg;F, THEN INITIATE alternate Shutdown Cooling with the SRVs as follows:
with the the SRVs SRVs asas follows:
follows:
1.
1.
: 1. ENSURE ALL ENSURE             control rods ALL control     rods are are funy fully inserted.
ENSURE ALL control rods are fully inserteth 2.
inserteth                     0 2.
CONFIRM reactor vessel head is installed and tensioned.
: 2. CONFIRM reactor CONFIRM       reactor vessel vessel head head isis installed installed and and                  0 tensioned.
3.
3.
: 3. IF the Reactor IF      Reactor Recirculation Recirculation Pumps Pumps are  are running:
IF the Reactor Recirculation Pumps are running, THEN PERFORM the following:
running, THEN PERFORM the PERFORM        the following:
a.
a.
: a.      RAISE AND MAINTAIN RAISE            MAINTAIN reactor reactor water water level level            0 between 200" 200 and 220" 220 as read on 321-LI-R6O548J. or as directed by Shift B21-U-R605A(8),
RAISE AND MAINTAIN reactor water level between 200 and 220 as read on 321-LI-R6O548J. or as directed by Shift Superintendent based on plant conditions.
Superintendent I)asedbased on plant conditions.
b.
b.
: b.                    running Reactor Recirculation Pumps STOP the running                                    Pumps in       0 accordance with 11(2)OP-02.
STOP the running Reactor Recirculation Pumps in accordance with 1(2)OP-02.
(2)OP-02.
4.
: 4. SHUT DOWN the RHR      RI-IR loop that was operating in                     0U Cooling in accordance with 1(2)OP-17.
SHUT DOWN the RI-IR loop that was operating in U
Shutdown Cooltng
Shutdown Cooling in accordance with 1(2)OP-17.
: 5. PLACE one RHR loop in the Suppression Pool Cooling                         0U mode in accordance with 11(2)OP-17.
5.
(2)OP-17.
PLACE one RHR loop in the Suppression Pool Cooling U
: 6. IF Suppression Pool temperature rises above 95                F, 95&deg;F.
mode in accordance with 1(2)OP-17.
Q 0U THEN GO TO OEOP-02-PCCP, Primary Containment Control Procedure AND PERFORM CONCURRENTLY with this procedure.
6.
J OAOP-15.O IOAOP-15.0                                    Rev. 23                                   Page Page 11 11 of 21 I Categories Categories K/A:
IF Suppression Pool temperature rises above 95&deg;F.
KIA:            S295021 S295021 A2.03 A2.03                                    Tier/Group:
U THEN GO TO OEOP-02-PCCP, Primary Containment Control Procedure AND PERFORM CONCURRENTLY with this procedure.
Tier / Group: T1G1T1 G 1 RO RO Rating:     3.5 3.5                                               SRO SRO Rating: 3.5   3.5 LP LP Obj:
J OAOP-15.O Rev. 23 Page 11 of 21 Categories K/A:
Obj:        CLSLP120*06 CLS-LP-120*06                                     Source:
S295021 A2.03 Tier/Group:
Source:          ~vv NEW Cog Level:
T1G1 RO Rating:
Cog  Level:    NIGH HIGH                                               Category Category 8:8:
3.5 SRO Rating:
: 87. While
3.5 LP Obj:
: 87. While performing performing refueling refueling activities activities on on Unit Unit Two, Two, aa spent spent fuel fuel bundle bundle was was dropped dropped and and the following the    following alarms alarms were were received:
CLSLP120*06 Source:
received:
NEW Cog Level:
AREA RAD AREA    RAD REFUEL REFUEL FLOOR FLOOR HIGH HIGH PROCESS RX PROCESS        FV( BLDG BLDG VENT VENT RADRAD HIGHHIGH Which one Which     one of the following of the  following identifies:
NIGH Category 8:
identifies:
3.0 OPERATOR ACTIONS 3.2.14 IF ALL of the allove methods can NOT maintain reactor vessel coolant temperature below 212&deg;F, THEN INITIATE alternate Shutdown Cooling with. the SRVs as follows:
(1) the (1)  the immediate immediate operator operator action action that that is is required required to to be be performed performed and and (2)  the  bases  for the  performance    of this (2) the bases for the performance of this action?    action?
: 1.
A. (1)
ENSURE ALL control rods are funy inserted.
A.   (1) Standby Standby Gas Gas Treatment Treatment (SBGT)
0
(SBGT)
: 2.
(2) Ensures   control room   operators (2) Ensures control room operators will      will receive receive .:5. 22 Rem Rem TEDE TEDE B. (1)
CONFIRM reactor vessel head is installed and 0
B.   (1) Standby Standby Gas Gas Treatment Treatment (SBGT)
tensioned.
(SBGT)
: 3.
(2) Ensures   control room   operators (2) Ensures control room operators will receive  receive .:5. 55 Rem
IF the Reactor Recirculation Pumps are running: THEN PERFORM the following:
                                                                    .      Rem TEDE TEDE C. (1) Control Room Emergency Ventilation (CREV)
: a.
(2) Ensures control roomroom operators will receive .:5.<2    2 Rem TEDE D
RAISE AND MAINTAIN reactor water level 0
D~   (1) Control Room Emergency Ventilation (CREV)
between 200" and 220" as read on B21-U-R605A(8), or as directed by Shift Superintendent I)ased on plant conditions.
(2) Ensures control room operators will receive .:5. 5 Rem TEDE
: b.
STOP the running Reactor Recirculation Pumps in 0
accordance with 1 (2)OP-02.
: 4.
SHUT DOWN the RHR loop that was operating in Shutdown Cooltng in accordance with 1(2)OP-17.
: 5.
PLACE one RHR loop in the Suppression Pool Cooling mode in accordance with 1 (2)OP-17.
: 6.
IF Suppression Pool temperature rises above 95QF, THEN GO TO OEOP-02-PCCP, Primary Containment Control Procedure AND PERFORM CONCURRENTLY with this procedure.
IOAOP-15.0 Rev. 23 Categories KIA:
S295021 A2.03 Tier / Group: T1 G 1 SRO Rating:
3.5 RO Rating:
3.5 LP Obj:
CLS-LP-120*06 Source:  
~vv Cog Level:
HIGH Category 8:
0 0
0 Page 11 of 21 I
: 87. While performing refueling activities on Unit Two, a spent fuel bundle was dropped and the following alarms were received:
AREA RAD REFUEL FLOOR HIGH PROCESS FV( BLDG VENT RAD HIGH Which one of the following identifies:
(1) the immediate operator action that is required to be performed and (2) the bases for the performance of this action?
A. (1) Standby Gas Treatment (SBGT)
(2) Ensures control room operators will receive 2 Rem TEDE B.
(1) Standby Gas Treatment (SBGT)
(2) Ensures control room operators will receive. 5 Rem TEDE C. (1) Control Room Emergency Ventilation (CREV)
(2) Ensures control room operators will receive <2 Rem TEDE D (1) Control Room Emergency Ventilation (CREV)
(2) Ensures control room operators will receive 5 Rem TEDE
: 87. While performing refueling activities on Unit Two, a spent fuel bundle was dropped and the following alarms were received:
AREA RAD REFUEL FLOOR HIGH PROCESS RX BLDG VENT RAD HIGH Which one of the following identifies:
(1) the immediate operator action that is required to be performed and (2) the bases for the performance of this action?
A. (1) Standby Gas Treatment (SBGT)
(2) Ensures control room operators will receive.:5. 2 Rem TEDE B. (1) Standby Gas Treatment (SBGT)
(2) Ensures control room operators will receive.:5. 5 Rem TEDE C. (1) Control Room Emergency Ventilation (CREV)
(2) Ensures control room operators will receive.:5. 2 Rem TEDE D~ (1) Control Room Emergency Ventilation (CREV)
(2) Ensures control room operators will receive.:5. 5 Rem TEDE  


Feedback Feedback K/A: S29S023G KIA:    S295023G 2.04.49 2.04.49 Ability to Ability   to perform perform without without reference reference to to procedures procedures those those actions actions thatthat require require immediate immediate operation operation of system of  system components components and   and controls.
Feedback K/A: S295023G 2.04.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.
controls.
Refueling Accidents (CFR: 41.10 /43.2/45.6)
Refueling Accidents Refueling    Accidents (CFR: 41.10 (CFR:    41.10 //43.2/45.6) 43.2 /4S.6)
ROISRO Rating: 4.6/4.4 Objective: CLSLP302..J*02
ROISRO Rating:
: 2. Given plant conditions with spent fuel damage and a high airborne activity problem in progress, determine if the appropriate automatic actions have occurred in accordance with OAOP-5.0, Radioactive Spills, High Radiation, and Airborne Activity.
RO/SRO      Rating: 4.6/4.4 4.6/4.4 CLSLP302..J*02 Objective: CLS-LP-302-J*02 Objective:
: 2. Given plant conditions with
: 2. Given plant conditions        with spent spent fuel fuel damage damage and and aa high high airborne airborne activity activity problem problem in in progress, progress, determine ifif the determine       the appropriate appropriate automatic automatic actions actions have have occurred occurred in in accordance accordance with with OAOP-S.O, OAOP-5.0, Radioactive Radioactive Spills, High Spills,  High Radiation, Radiation, and and Airborne Airborne Activity.
Activity.


==Reference:==
==Reference:==
OAOP-05, Revision 23, Page 2, Section 3.1 Cog Level: High Explanation:
OAOP-05 immediate action for a dropped or damaged fuel assembly is to ENSURE CREVS is in operation.
The dose consequence calculation for the fuel handling accident does not credit the secondary containment or automatic CREVS start, however, it does assume that CREVS is manually initiated within 20 minutes of a dropped/damaged fuel assembly. Based on this analysis, Technical Specifications do not require secondary containment or CREVS automatic initiation instrumentation except during Modes 1, 2, or 3 or during operations with the potential to drain the Reactor vessel. The CREV System is designed to maintain a habitable environment in the CRE for a 30 day continuous occupancy after a DBA without exceeding 5 rem total effective dose equivalent (TEDE).
Knowledge of DBA analysis initial conditions.
Distractor Analysis:
Choice A: Plausible because PROCESS RX BLDG VENT RAD HIGH annunciator is easily confused with the auto start for SBGT verifing Auto actions can be confused with Immediate Actions. SBGT start is a supplemental action which will reduce control room dose and 2 Rem TEDE is a site administrative dose limit and can be confused with the actual Dose Analysis from FHA of 2.69 rem TEDE.
Choice B: Plausible because PROCESS RX BLDG VENT RAD HIGH annunciator is easily confused with the auto start for SBGT verifing Auto actions can be confused with Immediate Actions. SBGT start is a supplemental action which will reduce control room dose and 5 Rem TEDE is correct.
Choice C: Plausible because CREV is correct and 2 Rem TEDE is a site administrative dose limit and can be confused with the actual Dose Analysis from FHA of 2.69 rem TEDE.
Choice D: Correct Answer.
SRO Only Basis: Conditions and limitations in the facility license (43(b)(1)
Notes Feedback KIA: S29S023G 2.04.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.
Refueling Accidents (CFR: 41.10 / 43.2 /4S.6)
RO/SRO Rating: 4.6/4.4 Objective: CLS-LP-302-J*02
: 2. Given plant conditions with spent fuel damage and a high airborne activity problem in progress, determine if the appropriate automatic actions have occurred in accordance with OAOP-S.O, Radioactive Spills, High Radiation, and Airborne Activity.


==Reference:==
==Reference:==
 
OAOP-OS, Revision 23, Page 2, Section 3.1 Cog Level: High Explanation:
OAOP-05, Revision OAOP-OS,     Revision 23,23, Page Page 2, 2, Section Section 3.1
OAOP-OS immediate action for a dropped or damaged fuel assembly is to ENSURE CREVS is in operation.
 
The dose consequence calculation for the fuel handling accident does not credit the secondary containment or automatic CREVS start, however, it does assume that CREVS is manually initiated within 20 minutes of a dropped/damaged fuel assembly. Based on this analysis, Technical Specifications do not require secondary containment or CREVS automatic initiation instrumentation except during Modes 1, 2, or 3 or during operations with the potential to drain the Reactor vessel. The CREV System is designed to maintain a habitable environment in the CRE for a 30 day continuous occupancy after a DBA without exceeding S rem total effective dose equivalent (TEDE).
===3.1 Level===
High Cog Level:     High Explanation:
OAOP-05 immediate action for a dropped or damaged fuel assembly is to ENSURE CREVS is in OAOP-OS operation.
The dose consequence calculation for the fuel handling accident does not credit the secondary containment or automatic CREVS start, however, it does assume that CREVS is manually initiated within 20 minutes of a dropped/damaged fuel assembly. Based on this analysis, Technical Specifications do not require secondary containment or CREVS automatic initiation instrumentation except during Modes 1, 2, or 3 or during operations with the potential to drain the Reactor vessel. The CREV System is designed to maintain a habitable environment in the CRE for a 30 day continuous occupancy after a DBA without exceeding S   5 rem total effective dose equivalent (TEDE).
Knowledge of DBA analysis initial conditions.
Knowledge of DBA analysis initial conditions.
Distractor Analysis:
Distractor Analysis:
Choice A: Plausible because PROCESS RX BLDG VENT RAD HIGH annunciator is easily confused with the auto start for SBGT verifing Auto actions can be confused with Immediate Actions. SBGT start is a supplemental action which will reduce control room dose and 2 Rem TEDE is a site administrative dose limit and can be confused with the actual Dose Analysis from FHA of 2.69 rem TEDE.
Choice A: Plausible because PROCESS RX BLDG VENT RAD HIGH annunciator is easily confused with the auto start for SBGT verifing Auto actions can be confused with Immediate Actions. SBGT start is a supplemental action which will reduce control room dose and 2 Rem TEDE is a site administrative dose limit and can be confused with the actual Dose Analysis from FHA of 2.69 rem TEDE.
Choice B: Plausible because PROCESS RX BLDG VENT RAD HIGH annunciator is easily confused with the auto start start for SBGT verifing Auto actions actions can be be confused with Immediate Immediate Actions. SBGT start is is aa supplemental supplemental action action which which will reduce control room dose    dose and 5S Rem TEDE is    is correct.
Choice B: Plausible because PROCESS RX BLDG VENT RAD HIGH annunciator is easily confused with the auto start for SBGT verifing Auto actions can be confused with Immediate Actions. SBGT start is a supplemental action which will reduce control room dose and S Rem TEDE is correct.
Choice Choice C:C: Plausible because because CREV CREV is is correct and and 22 Rem TEDETEDE is is aa site site administrative dose dose limit limit and and can be be confused confused withwith the the actual actual Dose Dose Analysis from from FHA     of 2.69 FHA of  2.69 rem TEDE.
Choice C: Plausible because CREV is correct and 2 Rem TEDE is a site administrative dose limit and can be confused with the actual Dose Analysis from FHA of 2.69 rem TEDE.
TEDE.
Choice D: Correct Answer.
Choice Choice D:   Correct Answer.
SRO Only Basis: Conditions and limitations in the facility license (43(b)(1)
D: Correct    Answer.
Notes  
SRO SRO Only Only Basis:
Basis: Conditions Conditions and and limitations limitations in in the the facility facility license license (43(b)(1)
(43(b)(1)
Notes Notes


Unit 22 Unit APP U.A.-03 APP    UA-03 3-73-7 Page 11 of Page      of 11 AREA RAD AREA  RAD REfUEL REFUEL FLOOR FLOOR HIGHHIGH AUTO ACTIONS AUTO  ACTIONS NONE NONE CAUSE
Unit 2 APP UA-03 3-7 Page 1 of 1 AREA RAD REFUEL FLOOR HIGH AUTO ACTIONS NONE CAUSE 1.
        '1.1. High   radiation level High radiation    level in in the the cask cask wash wash area.
High radiation level in the cask wash area.
area.
2.
Circuit malfunction.
3.
Refueling cavity water seal failure.
OBSERVATIONS 1.
ARM indicator and ip unit Upscale light illuminated on Panel H 12-P600.
ACTIONS 1.
Refer to EOP-03-SCCP, Table 3; enter EOP-03-SCCP as appropriate.
2.
2.
: 2. Circuit malfunction.
Refer to AOP-05.0, Radioacte Spills, High Radiation, and Airborne Activity.
Circuit  malfunction.
3.
3.
: 3. Refueling cavity Refueling    cavity water water sealseal failure.
Suspend refueling operation if due to fuel pool low level from refueling cavity water seal leakage.
failure.
4.
OBSERVATIONS OBSERV.A.TIONS 1.
If a circuit malfunction is suspected, ensure that a Trouble Tag is prepared.
        '1. ARM indicator ARM    indicator and and trip ip unitunit Upscale Upscale light   illuminated on light illuminated  on Panel Panel HH 12-P600.
DEVICEISETPOINTS ARM Channel 29 1<2 40 rnRPhr POSSIBLE PLANT EFFECTS 1.
12-P600.
Suspension of refuel floor activities.
ACTIONS ACTIONS 1.
REFERENCES 1.
: 1. Refer to EOP-03-SCCP, EOP-03-SCCP, Table 3;        3; enter EOP-03-SCCP EOP-03-SCCP as appropriate.
LL-9353-39 2.
appropriate.
AOP-05.0 3.
: 2. Refer to Refer    to AOP-OS.O, AOP-05.0, Radioactive Radioacte Spills, Spills, High   Radiation, and High Radiation,    and Airborne Activity.
EOP-03-SCCP 2APP-LIA-03 Rev. 46 Page 34 of AREA RAD REfUEL FLOOR HIGH AUTO ACTIONS NONE
: 3. Suspend refueling Suspend      refueling operation ifif due to fuel pool pool low low !evel level from refueling refueling cavity cavity water seal leakage.
'1.
: 4. If a circuit malfunction is suspected, ensure that a Trouble Tag is prepared.
High radiation level in the cask wash area.
DEVICEISETPOINTS DE\iICE/SETPOINTS ARM Channel 29 K2     1<2                                                    rnRPhr 40 mRlhr POSSIBLE PLANT EFFECTS 1.
: 2.
: 1. Suspension of refuel floor activities.
Circuit malfunction.
REFERENCES REFERENCES 1.
: 3.
: 1. LL-9353-39 LL-9353    - 39
Refueling cavity water seal failure.
: 2. AOP-05.0
OBSERV.A.TIONS Unit 2 APP U.A.-03 3-7 Page 1 of 1  
: 3. EOP-03-SCC EOP-03-SCCP       P 2APP-LIA-03 12APP-uA-03                                           Rev. 46 Rev. 46                                 Page 34 Page    34 ofof 631
'1.
ARM indicator and trip unit Upscale light illuminated on Panel H 12-P600.
ACTIONS
: 1.
Refer to EOP-03-SCCP, Table 3; enter EOP-03-SCCP as appropriate.
: 2.
Refer to AOP-OS.O, Radioactive Spills, High Radiation, and Airborne Activity.
: 3.
Suspend refueling operation if due to fuel pool low !evel from refueling cavity water seal leakage.
: 4.
If a circuit malfunction is suspected, ensure that a Trouble Tag is prepared.
DE\\iICE/SETPOINTS ARM Channel 29 K2 40 mRlhr POSSIBLE PLANT EFFECTS
: 1.
Suspension of refuel floor activities.
REFERENCES
: 1.
LL-9353 - 39
: 2.
AOP-05.0
: 3.
EOP-03-SCCP 12APP-uA-03 Rev. 46 Page 34 of 631  


Unit Unit 22 APP APP U.A.-03 UA-03 4-54-S Page'l Page 1 of of 11 PROCESS RX PROCESS       BLDG VENT RX BLDG       VENT RADRAD HIGHHIGH AUTO ACTIONS AUTO ACTIONS NONE NONE CAUSE 1.
Unit 2 APP UA-03 4-S Page 1 of 1 PROCESS RX BLDG VENT RAD HIGH AUTO ACTIONS NONE CAUSE 1.
: 1. High airborne activity High              activity in in Reactor Reactor Building Suiding ventilation exhaust exhaust plenum.
High airborne activity in Reactor Suiding ventilation exhaust p!enum.
p!enum.
2.
2.
: 2. Circuit malfunction.
Circuit malfunction.
Circuit malfunction.
OBSERVATIONS 1.
OBSERVATIONS OBSERV.A.TiONS 1.
Reactor Building Vent Rad Recorder D12-RR-R605 ChannelA or B indicates high radiation level.
      '1. Reactor Building Reactor    Building Vent Rad  Rad Recorder Recorder D12-RR-R605 D12-RR-R605 Channel ChannelAA or or B B indicates indicates high high radiation level.
2.
: 2. Reactor Building Exhaust Plenum Rad      Rad Monitor Channel A or B indicates greater rnRlhr on Panel than 3 mRlhr         Panel H12-P606.
Reactor Building Exhaust Plenum Rad Monitor Channel A or B indicates greater than 3 rnRlhr on Panel H12-P6G6.
H12-P6G6.
ACTIONS 1.
Enter EOP-03.SCCP. Secondary Containment Conti-o.
2.
Refer to AOP-05.O, Radioacte SpiIs, High Radiation, and Airborne Activity.
3.
If a circuit malfunction is suspected, ensure that a Troube Tag is prepared.
DEVICEISETPOINTS D12-RR-R605 red or black pen 3 mRihr POSSIBLE PLANT EFFECTS 1.
Possible release to environs.
2.
If airborne activity increases to 4 niRihr. Reactor Building HVAC isolation, a Group 6 isolation, drjwell purge isolation, and initiation of the Standby Gas Treatment System ifl1 occur.
REFERENC ES 1.
LL-9353 - 35 2.
AOP-O5.O 3.
EOP-03-SCCP 4.
Plant Modification 85-081 2APP-UA-03 Rev. 46 Page 41 of 63 PROCESS RX BLDG VENT RAD HIGH AUTO ACTIONS NONE Unit 2 APP U.A.-03 4-5 Page'l of 1
: 1.
High airborne activity in Reactor Building ventilation exhaust plenum.
: 2.
Circuit malfunction.
OBSERV.A.TiONS  
'1.
Reactor Building Vent Rad Recorder D12-RR-R605 Channel A or B indicates high radiation level.
: 2.
Reactor Building Exhaust Plenum Rad Monitor Channel A or B indicates greater than 3 mRlhr on Panel H12-P606.
ACTIONS
ACTIONS
: 1.            EOP-03.SCCP. Secondary Containment Control.
: 1.
Enter EOP-03-SCCP,                                     Conti-o.
Enter EOP-03-SCCP, Secondary Containment Control.
: 2. Refer to AOP-OS.O,       Radioacte Spills, AOP-05.O, Radioactive        SpiIs, High Radiation, and Airborne Activit'!.
: 2.
Activity.
Refer to AOP-OS.O, Radioactive Spills, High Radiation, and Airborne Activit'!.
: 3. If a circuit malfunction is suspected, ensure that a Trouble Troube Tag is prepared.
: 3.
DEVICEISETPOINTS DEVICE/SETPOINTS D12-RR-R605 red or black pen D'12-RR-R605                                                        3 mRlhr mRihr POSSIBLE PLANT EFFECTS
If a circuit malfunction is suspected, ensure that a Trouble Tag is prepared.
: 1. Possible release to environs.
DEVICE/SETPOINTS D'12-RR-R605 red or black pen 3 mRlhr POSSIBLE PLANT EFFECTS
: 2.                                           niRihr. Reactor Building HVAC isolation, a Group If airborne activity increases to 4 mRlhr, 6 isolation, dr'Jwell drjwell purge isolation, and initiation of the Standby Gas Treatment System ifl1
: 1.
                      \'Iill occur.
Possible release to environs.
REFERENC REFERENCESES
: 2.
: 1. LL-9353 - 35
If airborne activity increases to 4 mRlhr, Reactor Building HVAC isolation, a Group 6 isolation, dr'Jwell purge isolation, and initiation of the Standby Gas Treatment System \\'Iill occur.
: 2. AOP-O5.O AOP-05.0
REFERENCES
: 3. EOP-03-SCCP
: 1.
: 4. Plant Modification 85-081 2APP-UA-03 12APP-uA-03                                     Rev. 46                                Page 41 Page    41 ofof 63 631
LL-9353 - 35
: 2.
AOP-05.0
: 3.
EOP-03-SCCP
: 4.
Plant Modification 85-081 12APP-uA-03 Page 41 of 631
 
1.0 SYMPTOMS 1.1 AREA RAD REFUEL FLOOR HIGH (UA-03 3-7) is in alarm.
1.2 AREA RAD NEW FUEL STORAGE HIGH (UA-03 4-7) is in alarm.
1.3 PROCESS RX BLDG VENT RAD i-il (UA-03 4-5) is in alarm.
1.4 TVRB BLDG VENT RAD HIGH (U.4-03 3-3 is in alarm.
1.5 Area Radiation Monitor (ARM) is in alarm.
1.6 Continuous Air Monitor (CAM) is in alarm.
1.7 Turbine Building once-through effluent monitor indicates elevated (higher than expected or an unanticipated increase) activity.
1.8 Routine surveys indicate high radiation, contamination andlor airborne activity.
1.9 Report of spill. leak. or potential damage to new or spent fuel.
2.0 AUTOMATIC ACTIONS 2.1 IF PROCESS RX BLDG VENT RAD HI-HI (UA-03 3-5) is in alam, THEN the following actions occur:
 
Reactor Building Ventilation isolation


1.0 1.0        SYMPTOMS SYMPTOMS 1.1 1.1        AREA RAD RAD REFUEL REFUEL FLOOR FLOOR HIGHHIGH (UA-03 (UA-03 3-7) 3-7) is    in alarm.
SBGTSautostart U
is in  alarm.
1.2 1.2        AREA RAD AREA    RAD NEWNEW FUELFUEL STORAGE STORAGE HIGH  HIGH (UA-03 (UA-03 4-7) 4-7) isis in in alarm.
alarm.
1.3 1.3        PROCESS RX PROCESS              BLDG VENT RX BLDG      VENT RAD RAD HI  i-il (UA-03 (UA-03 4-5) 4-5) isis in in alarm.
alarm.
1.4 1.4        TVRB BLDG TURB    BLDG VENTVENT RADRAD H.IGH HIGH (U.4-03 (U.4-03 3-3)3-3 is is in in alarm.
alarm.
1.5 1.5        Area Radiation Area    Radiation Monitor Monitor (ARM)
(ARM) is is in  alarm.
in alaml.
1 .6 1.6        Continuous ,",ir Continuous          Monitor (CAM)
Air Monitor  (CAM) is inin alarm.
1.7 1.7        Turbine Building Turbine  Building once-through once-through effluent effluent monitor monitor indicates indicates elevated elevated (higher (higher than expected than  expected or  or an an unanticipated unanticipated increase) increase) activity.
activity.
1.8        Routine surveys indicate high radiation, contamination and/or          andlor airi)ome airborne activity.
1.9                    spill. leak, Report of spill,    leak. or potential damage to ne'.vnew or spent fuel.
2.0      AUTOMATIC ACTIONS 2.1        IF PROCESS RX BLDG VENT RAD HI-HI (UA-03 3-5) is in alam, THEN the following actions occur:
alam1, Reactor Building Ventilation isolation                                            0 SBGTSautostart SBGTS      auto start                                                            0U Group 6 Isolation.                                                                0 3.0 3,0      OPERATOR ACTIONS 3.1        Immediate Actions
))            3.1.1        IF aa fuel assembly was dropped or damaged, THEN ENSURE the Control Room Emergency Ventilation U0 System (CREVS) is in operation.
OAOP-05.O IOAOP-OS.O                                          Rev. 24 24                                        Page Page 22 of of 10 10 I


UPDATED UPDA      TED FSAR  FSAR                           evision:
Group 6 Isolation.
Revision:                21 21 Cr&Ll Pmm, ENGInEERED SAFETY ENGINEERED          SAFETYFEA    FruREs TURES Chapter:
3.0 OPERATOR ACTIONS 3.1 Immediate Actions
Chapter:
]
Page:
3.1.1 IF a fuel assembly was dropped or damaged, THEN U
Page:            108 oi 108   o 121 66 121 6.44.12 Fuel 6.4.4.1.2         Fuel Handling Handling Accident  Accident - Control
ENSURE the Control Room Emergency Ventilation System (CREVS) is in operation.
                                                            - Control RoomRoom Dose  Dose Section 15.7.1 Section     15.7.1 discusses discusses the  the release release ofof aotivity and its activity and  its transport transpoi to to the the environment environment following following aa postulated postulated fuel handling fuel  handling aoodent accident (FH (FHA).
OAOP-05.O Rev. 24 Page 2 of 10 1.0 SYMPTOMS 1.1 AREA RAD REFUEL FLOOR HIGH (UA-03 3-7) is in alarm.
                                      ..o,).
1.2 AREA RAD NEW FUEL STORAGE HIGH (UA-03 4-7) is in alarm.
The design The              inpiAs utilized design inputs       utilized to to ellaluate evalua:e the tne intake   & this ntake of  ths acjjvjt~*   into the acv,ty into     the conlrol control room room and and to  to assess assess thethe resultant dose resultant   dose to      the control to the   control room room operators operators are are tabulated tabulated in     Table 6-28.
1.3 PROCESS RX BLDG VENT RAD HI (UA-03 4-5) is in alarm.
in Table  6-22. AA sensitivity sensitivity study study of of unfiltered unuttered outsde air outside   air inleakage inleakage into   nto the the oontrol control room room 'lias   performed ellaluating was performed        evaluating ininleakage    rates of leakage rates       of 10.000 10,030 ofmofm (bouxidirig case).
1.4 TURB BLDG VENT RAD H.IGH (U.4-03 3-3) is in alarm.
(bounding     case). 30003000 cfrncfrn {control
1.5 Area Radiation Monitor (ARM) is in alaml.
{confrol rcom room design).
1.6 Continuous,",ir Monitor (CAM) is in alarm.
design), and and 00 cfm.
1.7 Turbine Building once-through effluent monitor indicates elevated (higher than expected or an unanticipated increase) activity.
cfm, Accident Acodent XiQ XIQ values values are are developed developed as  as discussed in discussed          Section 15.9.2.
1.8 Routine surveys indicate high radiation, contamination and/or airi)ome activity.
in Secticn      15..2. SectionSection '15.9.3   descr bes the 1.9.S describes       the parameters parameters utilized utilized in in conjunction conjunction with   with the the AQTRAO computer RADTRAD         computer code           Reterence 8-35}
1.9 Report of spill, leak, or potential damage to ne'.v or spent fuel.
code {Reference                  to con~'ert 8-35} to  convert the     Alternative Source the Alternative    Source Tern, Temi activity activity drawn drawn into into the contrcl the  control room room during during the      postulatec accident the postulated       accident into into aa total  effective dose total effeclive   dose equivalent equivalent (TEDE)
2.0 AUTOMATIC ACTIONS 2.1 IF PROCESS RX BLDG VENT RAD HI-HI (UA-03 3-5) is in alam1, THEN the following actions occur:
TEDE) dose. dose.
Reactor Building Ventilation isolation 0
Th 30-day The   30-day FHA FHA dose dose to to the the conirol control room room operator operator from from thethe internal internal cloud cloud associated associated with  with the    FHA is the FHA   is calculated to c,3lculated   to be   2.69 rem be 2.69    rem TEOE.
SBGTS auto start 0
TEQE.
Group 6 Isolation.
The onsile The    onsi:e control control room room operator operator dosedose criterion    established by criterion established       by Reference Reference 8-368-36 for for this this accident accident is is that that the the total  control room total contro!!   room cperator operator dose dose should should be be less         the to than the less than          10 CFR CFR 50.67 50.67 guidelines; guidelines: i.ei.e.,
0 3,0 OPERATOR ACTIONS 3.1 Immediate Actions
                                                                                                                        .* that   the total that the  total dose dose should be should    be less   than 5S rem less than        rem TEDE.
]
TEDE.
3.1.1 IF a fuel assembly was dropped or damaged, THEN 0
ENSURE the Control Room Emergency Ventilation System (CREVS) is in operation.
IOAOP-OS.O Rev. 24 Page 2 of 10 I
 
UPDATED FSAR evision:
21 r&L ENGInEERED SAFETYFruREs Chapter:
6
: Pmm, Page:
108 o 121 6.44.12 Fuel Handling Accident
- Control Room Dose Section 15.7.1 discusses the release of activity and its transpoi to the environment following a postulated fuel handling accident (FHA).
The design inpiAs utilized to evalua:e tne ntake & ths acv,ty into the control room and to assess the resultant dose to the control room operators are tabulated in Table 6-22. A sensitivity study of unuttered outsde air inleakage nto the control room was performed evaluating inleakage rates of 10,030 ofm (bouxidirig case). 3000 cfrn {confrol room design), and 0 cfm, Acodent XIQ values are developed as discussed in Section 15..2. Section 1.9.S descr bes the parameters utilized in conjunction with the AQTRAO computer code Reterence 8-35} to convert the Alternative Source Temi activity drawn into the control room during the postulatec accident into a total effective dose equivalent TEDE) dose.
Th 30-day FHA dose to the control room operator from the internal cloud associated with the FHA is calculated to be 2.69 rem TEQE.
The onsi:e control room operator dose criterion established by Reference 8-36 for this accident is that the total control room operator dose should be less than the 10 CFR 50.67 guidelines: i.e., that the total dose should be less than S rem TEDE.
C l UPDA TED FSAR ENGINEERED SAFETY FEA TURES 6.4.4.1.2 Fuel Handling Accident - Control Room Dose Revision:
21 Chapter:
6 Page:
108 oi 121 Section 15.7.1 discusses the release of aotivity and its transport to the environment following a postulated fuel handling aoodent (FH..o,).
The design inputs utilized to ellaluate the intake of this acjjvjt~* into the conlrol room and to assess the resultant dose to the control room operators are tabulated in Table 6-28. A sensitivity study of unfiltered outside air inleakage into the oontrol room 'lias performed ellaluating in leakage rates of 10.000 ofm (bounding case). 3000 cfrn {control rcom design). and 0 cfm. Accident XiQ values are developed as discussed in Secticn 15.9.2. Section '15.9.3 describes the parameters utilized in conjunction with the RADTRAD computer code {Reference 8-35} to con~'ert the Alternative Source Tern, activity drawn into the contrcl room during the postulated accident into a total effeclive dose equivalent (TEDE) dose.
The 30-day FHA dose to the conirol room operator from the internal cloud associated with the FHA is c,3lculated to be 2.69 rem TEOE.
The onsile control room operator dose criterion established by Reference 8-36 for this accident is that the total contro!! room cperator dose should be less than the to CFR 50.67 guidelines; i.e.* that the total dose should be less than 5 rem TEDE.  


3.0 3.0  OPERATOR ACTIONS OPERATOR      ACTIONS 3.2.3 3.2.3    IFIF new new oror spent    fuel damage spent fuel   damage isis suspected, suspected, THEN THEN PERFORM the PERFORM               following:
3.0 OPERATOR ACTIONS 3.2.3 IF new or spent fuel damage is suspected, THEN PERFORM the following:
the following:
1.
1.
: 1. PLACE any PLACE      any fuel fuel that that is is being  moved in being moved     in aa safe safe condition.
PLACE any fuel that is being moved in a safe condition.
condition.      D 2.
2.
: 2. SECURE further SECURE       further fuel fuel movement.
SECURE further fuel movement.
movement.                                    D 3.
3.
: 3. EVACUATE          personnel trom EVACUATE personnel             from the the following following areas:
EVACUATE personnel from the following areas:
areas:
Refueling Floor 0
                -    Refueling Floor Refueling    Floor                                                    D0
Drywell, if occupied Reactor Building, -17 Elev., if Shutdown Cooling in service.
                --    Drywell, ifif occupied Drywell,                                                              D
ECCS Pipe Tunnel U
                --    Reactor Building, Reactor  Building, -17'-17 Elev.,
Any area determined to have the potential for high radiation.
Elev., ifif Shutdown Shutdown Cooling Cooling in in      D service.
4.
                --    ECCS Pipe Tunnel                                                       DU
ISOLATE Secondary Containment.
                -    Any area determined to have the potential for high                     D radiation.
5.
: 4. ISOLATE Secondary Containment  Containment.                                 D
START Standby Gas Trains.
: 5. START Standby Gas Trains.                                                   D U
U 3.2.4 NOTIFY E&RC to perform the following as necessary:
3.2.4   NOTIFY E&RC to perform the following as necessary:
Area radiation survey U
                -    Area radiation survey                                                   D U
Air sampling U
                -    Air sampling                                                           D U
Smear survey U
                -    Smear survey                                                           D U
Posttheaffectedareaasnecessary U
                -    Posttheaffect Post  the affected edareaasnece area as necessary ssary                          D U
Control access to reduce exposure and U
Control access to reduce exposure and                                   D U
contamination.
contamination.
DAD P-05.0 IOAOP-05.0                                     Rev Rev. 23                                 Page 44 of 10 10 I
DADP-05.0 Rev 23 Page 4 of 10 3.0 OPERATOR ACTIONS 3.2.3 IF new or spent fuel damage is suspected, THEN PERFORM the following:
: 1.
PLACE any fuel that is being moved in a safe condition.
D
: 2.
SECURE further fuel movement.
D
: 3.
EVACUATE personnel trom the following areas:
Refueling Floor D
Drywell, if occupied D
Reactor Building, -17' Elev., if Shutdown Cooling in D
service.
ECCS Pipe Tunnel D
Any area determined to have the potential for high D
radiation.
: 4.
ISOLATE Secondary Containment D
: 5.
START Standby Gas Trains.
D 3.2.4 NOTIFY E&RC to perform the following as necessary:
Area radiation survey D
Air sampling D
Smear survey D
Post the affected area as necessary D
Control access to reduce exposure and D
contamination.
IOAOP-05.0 Rev. 23 Page 4 of 10 I
 
4.0 GENERAL DISCUSSION Liquid radioactive spills may be caused by valve packing leaks, leaky fittings, system leaks, or system draining evolutions. Liquids spills should be covered with an absorbent material to minimize the spread of contamination. Solid spills may be caused by leaks from the containers or process streams which handle radioactive material or by an accident during the transport of new or spent fuel, radioactive sources, or other solid radioactive materials. Solid spills should be covered by a damp material to minimize the spread of airborne contamination. A spill of highly radioactive solid materials such as spent resin, filter sludge, neutron sources, or irradiated reactor internal components may create a serious personnel exposure problem and should be handled with extreme caution. In addition, high radiation and high airborne activity may accompany a spill.
High airborne activity may occur from reactor coolant leaks, coolant spills, radwaste leaks, sampling, grinding, draining, and other maintenance. High airborne activity in the turbine buildings may require ventilation shutdown or realignment to the recirculation lineup if the ventilation systems are operating in the once-through lineup.
High radiation levels may be caused by radiation streaming, loss of or degraded shielding, fuel element damage, high airborne activity, coolant spills, or radiography.
New or spent fuel damage may occur within the plant during fuel handling operations. Fuel may be damaged if it is inadvertently dropped or allowed to collide with objects. Damage may also be sustained if heavy objects (shipping casks, reactor vessel head, drywell head, etc.) are allowed to fall on the fuel. These accidents may release a substantial amount of radioactive noble gases, halogens, and other fission products into the secondary containment. The secondary containment will be automatically isolated due to high radiation at its ventilation exhaust plenum. Although Standby Gas Treatment (SBGT) System will reduce the activity released to the environs, there is a chance that technical specification limits may be exceeded.
The dose consequence calculation for the fuel handling accident does not credit the secondary containment or automatic CREVS start, however, it does assume that CREVS is manually initiated within 20 minutes of a dropped(damaged fuel assembly. Based on this analysis, Technical Specifications do not require secondary containment or CREVS automatic initiation instrumentation except during Modes 1, 2, or 3 or during operations with the potential to drain the Reactor vessel.
OAOP-05.0 Rev. 23 Page 8 of 10 4.0 GENERAL DISCUSSION Liquid radioactive spills may be caused by valve packing leaks, leaky fittings, system leaks, or system draining evolutions. Liquids spills should be covered with an absorbent material to minimize the spread of contamination. Solid spills may I)e caused by leaks from the containers or process streams which handle radioactive material or by an accident during the transport of new or spent fuel, radioactive sources, or other solid radioactive materials. Solid spills should be covered by a damp material to minimize the spread of airborne contamination. A spill of highly radioactive solid materials such as spent resin, filter sludge, neutron sources, or irradiated reactor internal components may create a serious personnel exposure problem and should be handled with extreme caution. In addition, high radiation and higl1 airborne acUvity may accompany a spill.
High airborne activity may occur from reactor coolant leaks, coolant spills, radwaste leal,s, sampling, grinding, draining, and other maintenance. High airborne activity in the turbine buildings may require ventilation shutdown or realignment to the recirculation lineup if the ventilation systems are operating in the once-through lineup.
High radiation [evels may be caused by radiation "streaming," loss of or degraded shielding, fuel element damage, high airi)orne activity, coolant spillS, or radiography.
New or spent fuel damage may occur within the plant during fuel handling operations. Fuel may be damaged if it is inadvertently dropped oral[owed to collide with objects. Damage may also be sustained if heavy objects (shipping casks, reactor vessel head, dlY'Nell head, etc.) are allowed to fall on the fuel. These accidents may release a substantial amount of radioactive noble gases, halogens, and other fission products into the secondary containment The secondary containment will be automatically isolated due to high radiation at its ventilation exhaust plenum. Although Standby Gas Treatment (SBGT) System will reduce the activir; released to the environs, there is a chance that technical speCification limits may be exceeded.
The dose consequence calculation for the fuel handling accident does not credit the secondary containment or automatic CREVS start, however, it does assume that CREVS is manually initiated within 20 minutes of a droppedfdamaged fuel assembly. Based on this analysis, Technical Specifications do not require secondary containment or CREVS automatic initiation instrumentation except during Modes 1, 2, or 3 or during operations with the potential to drain the Reactor vessel.
IOAOP-05.0 Rev. 23 Page B of 10 I  
 
UPDATED FSAR Revion:
21 CP&L EzIGIIIEERED SAFETY FEA TURES Chapter:
3 CHAPTER 6 TABLES 1 of 1 TABLE 6-28 Control Room Design Inputs Design Basis Accidents Control Room 1.
Control room habitabi!ity vo ume


4.0 4.0    GENERAL DISCUSSION GENERAL      DISCUSSION Liquid radioactive Liquid  radioactive spills spills may may be be caused caused by by valve valve packing packing leaks, leaks, leaky leaky fittings, fittings, system leaks, system    leaks, or or system system draining draining evolutions.
2g8,650 f 2.
evolutions. Liquids Liquids spills spills should should bebe covered covered with with an absorbent an  absorbent material material to    minimize the to minimize  the spread spread of of contamination.
Assumed unfiltered inIeakage 10.000 crn Control Room Ventilation 1.
contamination. Solid Solid spills spills may may I)e be caused by caused    by leaks leaks from from the the containers containers oror process process streams streams which which handle handle radioactive radioactive material  or by material or  by an an accident accident during during the  transport of the transport    of new new or or spent spent fuel, radioactive radioactive sources, or sources,    or other other solid solid radioactive radioactive materials.
Normal mode operation outside air intake 2,1C0 cm 2.
materials. Solid Solid spills should should bebe covered covered by by aa damp material damp    material toto minimize minimize the the spread spread ofof airborne airborne contamination.
Normal mode roughing filter, aerosol removal 0%
contamination. A    A spill spill of of highly highly radioactive solid radioactive    solid materials materials such such as as spent spent resin, resin, filter filter sludge, sludge, neutron neutron sources, sources, oror irradiated reactor internal irradiated            internal components may create aa serious personnel  personnel exposure problem and should be handled with extreme caution. In addition,          addition, high high radiation high airborne acUvity and higl1              activity may accompany aa spill.
3.
High airborne High  airborne activity    may occur activity may    occur from reactor reactor coolant coolant leaks, leaks, coolant coolant spills, radwaste leal,s, radwaste    leaks, sampling, grinding, draining, draining, and other maintenance.
Normal mode roughing filter, elemental iod:ne removal 0%
maintenance. High  High airborne activity in the turbine buildings may require ventilation shutdown or realignment to the recirculation lineup if the ventilation systems are operating in the once-through lineup.
4.
levels may be caused High radiation [evels              caused by radiation "streaming,"
Normal mode roughing filter, organic iodine removal 0%
streaming, loss of or degraded element damage, high airi)orne shielding, fuel element                        airborne activity, coolant spillS, spills, or radiography.
5.
New or spent fuel damage may occur within the plant during fuel handling operations. Fuel may be damaged if it is inadvertently dropped oral[owed        or allowed to collide with objects. Damage may also be sustained if heavy objects (shipping casks, reactor vessel head, dlY'Nell drywell head, etc.) are allowed to fall on the fuel. These accidents may release a substantial amount of radioactive noble gases, halogens, and other fission products into the secondary containmentcontainment. The secondary containment will be automatically isolated due to high radiation at its ventilation exhaust plenum. Although Standby Gas Treatment (SBGT) System will reduce the activity released to the environs, there is a activir;                                          a chance that technical speCification specification limits may be exceeded.
Time of manual switchover from normal to radiation mode 20 minutes 0.
The dose consequence calculation for the fuel handling accident does not credit the secondary containment or automatic CREVS start, however, it does assume that CREVS is manually initiated within 20 minutes of a droppedfdamageddropped(damaged fuel assembly. Based on this analysis, Technical Specifications do not require secondary containment or CREVS automatic initiation instrumentation except during Modes 1,  1, 2, or 33 or during operations with the potential to drain the Reactor vessel.
Radiation mode operation outsde sir ntake I.500 cfm 7.
OAOP-05.0 IOAOP-05.0                                      Rev.
Radiation mode HEPA ft ter. aerosol removal Radiation mode charcoal filter, elemental iodine removal Radiation mode charcoal fiter, organic iodine removal 10.
Rev. 23 23                                    Page Page 8B of of 10 10 I
Radiation train charcoal depth 2 inches 11.
Radiation mode filtered recrculated airflow 40 cfm 12.
Radiation mode aerosol iodine removal 13.
Radiation mode elemental iodine removal 14.
Radiation mode organic iodne removal NOTES


UPDATED UPDA     TED FSARFSAR                      Revision:
Sensitivity cases using 3.00 cfm and 0 cThi unfiltered outside air inleakage into the control room were also evaluated. The 10,000 cfm unfiltered inleakage case is bounding for the LOCA, the FHA, and the CRDA events.
Revion:          21 21 CP&L                              EzIGIIIEEREDSAFETY ENGmEERED         SAFETYFEA CHAPTER 66 TABLES FEATURES TABLES TURES Chapter:
For the MSLB event, 0 cfm unfiltered outside air nleakage represents the bounding vaiue.
 
For the MSLB event, a 5ensitlvity study was performed, isolating the control room at various times between 5.0 seconds and 30 days.
TABLE 6-28 UPDA TED FSAR ENGmEERED SAFETY FEA TURES CHAPTER 6 TABLES Control Room Design Inputs - Design Basis Accidents e - roughing filter. aerosol removal
* organic iodine removal
- aerosol iodine removal
- elemental iodine remollal NOTES Revision:
Chapter:
Chapter:
Page:         11 of 63 of 1 CHAPTER TABLE 6-28 TABLE  6-28    Control Room Control    Room Design Design Inputs Inputs - Design Basis Accidents Design Basis    Accidents Control Room
Page:
: 1. Control room habitabi!ity vo ume                                                        2g8,650 flf 2gS.650
2gS.650 fl 10.000 cim' 2.100 cfm 0%
: 2. Assumed unfiltered inIeakage                                                            10.000    crn 10.000 cim' Control Room Ventilation
: 1. Normal mode operation          outside air intake                                        2,1C0 cfm 2.100 cm
: 2. Normal mode      roughing filter.
e - roughing      filter, aerosol aerosol removal removal                                      0%
0%
0%
: 3. Normal mode      roughing filter, elemental iod:ne removal                                  0%
0%
0%
: 4. Normal mode      roughing filter, organic iodine removal                                    0%
20 minutes'"
0%
1.500 cfm 95%
: 5. Time of manual switchover from normal to radiation mode                                      minutes 20 minutes'"
90%
20
90%
: 0. Radiation mode operation          outsde sir ntake                                      I .500 cfm 1.500 cfm
2inehes 400 cfm 95%
: 7. Radiation mode      HEPA ft ter. aerosol removal                                            95%
90%
  . Radiation mode      charcoal filter, elemental iodine removal                              90%
90%
  . Radiation mode      charcoal fiter,* organic iodine removal                                90%
Sensitivity cases using 3.000 cfm a,nd 0 dm unfiliered outside air inleakage into the control room were also evaluated. The 10.000 cim unfiltered inleakage case is boundi,ng for the lOCA. the FHA. and the eRDA ellents.
: 10. Radiation train    charcoal depth                                                        2inehes 2 inches
For the MSLB ellent. 0 cfm unfiltered outside air inleakage represents the bounding value.
: 11. Radiation mode      filtered recrculated airflow                                        400 40 cfm
For the MSLB ellent. a sensitivity study was performed. isolating the contrel room at various times between 5.5 seconds and 30 days.
: 12. Radiation mode - aerosol iodine removal 95%
21 6
: 13. Radiation mode - elemental iodine remollal removal                                          90%
1 of
: 14. Radiation mode      organic iodne removal                                                  90%
 
NOTES NOTES Sensitivity cases using 3.000  3.00 cfm a,nd     cThi unfiliered and 0 dm   unfiltered outside air inleakage into the control room were also into                              also evaluated. The 10,000 10.000 cfm cim unfiltered inleakage case is bounding boundi,ng for the LOCA, lOCA. the FHA, FHA. and the CRDA eRDA events.
CREV System B 37.3 BASES BACKGROUND The CREV System is designed to maintain a habitable environment in the (connued}
ellents.
CRE for a 30 day connuous occupancy after a DBA without exceeding 5 rem total effective dose equivalent (TEDE), A single CREV subsystem operating at a flow rate of 2200 cfm will slightly pressurize the CRE relative to outside atmosphere to minimize infiltration of air from surrounding areas adjacent to the CRE bounday. CREV System operation in maintaining CRE habitability is discussed in the UFSAR, Sections 6.4 and 9.4, (Refs. 1 and 2. respectively).
For the MSLB MSLB event, ellent. 0 cfm unfiltered outside air inleakage nleakage represents the bounding vaiue.
APPLICABLE The ability of the CREV System to maintain the habitability of the CRE SAFETY ANALYSES is an explicit assumption for the design basis accident presented in the UFSAR (Ref. 3). The radiation!smoke protection. mode of the CREV System is assumed (explicitly or implicitly) to operate following a DBA.
value.
The radiological doses to the CRE occupants as a result of a DBA are summarized in Reference 3. Postulated single active failures that may cause the loss of outside or recirculated air from the CRE are bounded by BNP radiological dose calculations for CRE occupants.
For For the MSLB MSLB event, ellent. aa 5ensitlvity sensitivity study study was performed,    isolating the control performed. isolating        contrel room room at at various various times between between 5.05.5 seconds seconds and and 30 30 days.
Brunswick Unit 2 S 3.7,3-2 Revision No. 61 Categories K/A:
days.
S295023G2.04.49 Tier/Group:
T1G1 RO Rating:
4.6 SRO Rating:
4.4 LP Obj:
CLSLP.3O2J*O2 Source:
NEW Cog Level:
HIGH Category 8:
BASES BACKGROUND (continued}
CREV System B 3.7.3 The CREV System is designed to maintain a habitable environment in the CRE for a 30 day continuous occupancy after a DBA without exceeding 5 rem total effecti.... e dose equivalent {TED E). A single CREV subsystem operating at a f10Vl rate of s: 2200 cfm ',viii sligh!ly pressurize the CRE relative to outside atmosphere to minimize infiltration of air from surrounding areas adjacent to the CRE boundary. CREV System operation in maintaining CRE habitability is discussed in the UFSAR, Sections 6.4 am! 9.4, (Refs. *1 and 2, respectively).
APPLICABLE The ability of the CREV System to maintain the habitability of the CRE SAFETY.ANAL YSES is an explicit assumption ror the design basis accident presented in the UfSAR (Ref. 3}. The radiation/smoke protection mode ofthe CREV System is assumed (explicitly or implicitly) to operate fol!owing a DBA.
The radiological doses to the CRE occupants as a result of a DBA are summarized in Reference 3. Postulated single active failures that may cause the loss of outside or recirculated air from the CRE are bounded by BNP radiological dose calculations for CRE occupants.
Brunswick Unit 2 B 3.7.3-2 Re.... ision No. 61 Categories KIA:
S295023G 2.04.49 Tier / Group: TIGl RORating:


CREV CREV System  System BB 3.7.3 37.3 BASES BASES BACKGROUND BACKGROUND            The CREV The    CREV System System is      is designed designed to  to maintain maintain aa habitable habitable environment environment in        in the the (connued}
===4.6 SRORating===
(continued}        CRE for CRE    for aa 30 30 day      connuous occupancy day continuous          occupancy after  after aa DBA DBA without without exceeding exceeding rem total 55 rem  total effecti effective      dose equivalent
4.4 LP Obj:
                                                .... e dose    equivalent {TED(TEDE), E). A A single single CREV CREV subsystemsubsystem operating at operating      at aa f10Vl flow raterate of of s: 2200 2200 cfmcfm ',viii will sligh!ly slightly pressurize pressurize the the CRE  CRE relative to relative        outside atmosphere to outside        atmosphere to    to minimize minimize infiltration infiltration of of air air from from surrounding areas surrounding        areas adjacent adjacent to  to the the CRE CRE boundary.
CLS-LP-302-J*02 Source:
bounday. CREV    CREV System System operation in operation          maintaining CRE in maintaining        CRE habitability habitability is is discussed discussed in in the the UFSAR, UFSAR, Sections 6.4 Sections      6.4 am!
NEW Cog Level:
and 9.4,9.4, (Refs.
HIGH Category 8:
(Refs. *11 and and 2,2. respectively).
: 88. An event on Unit One has resulted in the following plant conditions:
respectively).
Reactor pressure 1000 psig Reactor Water Level 120 inches Control Rod Positions All unknown APRMs Downscale Drywell pressure 3 psig Supp. Pool pressure 2 psig Supp. Pool water temp 150&deg; F Supp. Pool water level
APPLICABLE APPLICABLE                  ability of The ability    of the the CREV CREV System to maintain  maintain the habitability habitability ofof the CRE  CRE SAFETY .ANAL SAFETY    ANALYSES    is an YSES is  an explicit explicit assumption assumption ror          the design for the  design basis basis accident accident presented presented in      in the the UFSAR (Ref.
-4 feet (Reference provided)
UfSAR      (Ref. 3}.3). The The radiation/smoke radiation!smoke protection protection. modemode ofthe of the CREV  CREV System is System    is assumed assumed (explicitly (explicitly or or implicitly) implicitly) to to operate operate fol!owing following aa DBA. DBA.
Which one of the following identifies the status of the Heat Capacity Temperature Limit (HCTL) and the required procedure for reactor pressure control?
The radiological doses to the CRE occupants as aa result of a DBA                            DBA are summarized in ReferenceReference 3. Postulated single        single active failures that may        may cause the loss of outside or recirculated air from the CRE                  CRE are bounded by BNP radiological dose calculations for CRE occupants.
HCTL Pressure Control Leg of Procedure A. has been exceeded RVCP B has been exceeded LPC C. has NOT been exceeded RVCP D. has NOT been exceeded LPC
BNP Brunswick Unit 2                                      S 3.7.3-2 B 3.7,3-2                                        Re  .... ision No. 61 Revision Categories K/A:
: 88. An event on Unit One has resulted in the following plant conditions:
KIA:            S295023G2.
Reactor pressure Reactor Water Level Control Rod Positions APRMs Drywell pressure Supp. Pool pressure Supp. Pool water temp Supp. Pool water level (Reference provided) 1000 psig 120 inches All unknown Downscale 3 psig 2 psig 150 0 F
S295023G      04.49 2.04.49                                          Tier/Group:
-4 feet Which one of the following identifies the status of the Heat Capacity Temperature Limit (HCTL) and the required procedure for reactor pressure control?
Tier / Group: T1G1      TIGl RO Rating:
HCTL Pressure Control Leg of Procedure A. has been exceeded RVCP By has been exceeded LPC C. has NOT been exceeded RVCP D. has NOT been exceeded LPC  
RORating:      4.6                                                          SRO      Rating: 4.4 SRORating:
LP Obj:
LP              CLSLP.3O2J*O2 CLS-LP-302-J*02                                               Source:               NEW Cog Level:     HIGH                                                           Category 8:
: 88. An event
: 88. An event on on Unit Unit One One has has resulted resulted in in the the following following plant plant conditions:
conditions:
Reactor pressure Reactor    pressure                        1000 psig 1000   psig Reactor  Water Level Reactor Water     Level                   120 inches 120   inches Control Rod Control  Rod Positions Positions                  All unknown All  unknown APRMs APRMs                                     Downscale Downscale Drywell pressure Drywell  pressure                        33 psig psig Supp. Pool pressure Supp. Pool    pressure                  22 psig psig Supp. Pool Supp. Pool water water temp temp                  150&deg;0 FF 150 Supp. Pool Supp. Pool water    level water level                 -4 feet
                                                    -4 feet (Reference provided)
(Reference provided)
Which one   of the following one of      following identifies identifies the the status status ofof the the Heat Heat Capacity Capacity Temperature Temperature Limit Limit (HCTL)
(HCTL) and the required  procedure required procedure     for reactor reactor pressure control?
control?
HCTL                                       Pressure Control Leg of Procedure A. has been exceeded                                                     RVCP B has been exceeded By                                                                      LPC C. has NOT been exceeded                                                 RVCP D. has NOT been exceeded                                                 LPC


Feedback Feedback K/A: S295026 KIA:              A2.03 S295026 A2.03 Ability to Ability  to determine determine and/or andlor interpret interpret the the following following as as they they apply apply to to SUPPRESSION SUPPRESSION POOL  POOL HIGH HIGH WATER TEMPERATURE:
Feedback K/A: S295026 A2.03 Ability to determine andlor interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:
WATER      TEMPERATURE:
Reactor pressure (CFR: 41.10/43.5/45.13)
Reactor pressure Reactor   pressure (CFR:   41.10/43.5/4  5.13)
RO/SRO Rating: 3.9/4.0 Objective: CLSLP300L*05a
(CFR: 41.10 143.5 145.13)
: 05. Given the PCCP, determine the appropriate actions if any of the following limits are approached or exceeded:
RO/SRO Rating:
: a. Heat Capacity Temperature Limit.
RO/SRO      Rating: 3.9/4.0 3.9/4.0 Objective:     CLSLP300L*05a Objective: ClS-lP-300-l    *05a
: 05. Given   the PCCP, determine
: 05. Given the PCCP,      determine the the appropriate appropriate actions actions ifif any any of of the the following following limits limits are are approached approached or or exceeded:
exceeded:
: a. Heat
: a. Heat Capacity Capacity Temperature Temperature Limit.
Limit.


==Reference:==
==Reference:==
Heat Capacity Temperature Graph only is given to examinee PCCP.
Cog Level: High Explanation:
HCTL has been exceeded. With rods unknown the operator would be in LPC.
Distractor Analysis:
Choice A: Plausible because rods are unknown, would be in LPC.
Choice B: Correct Answer Choice C: Plausible because HCTL has been exceeded. rods are unknown, would be in LPC Choice D: Plausible because HCTL has been exceeded.
SRO Only Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations (43(b)(5)
Notes Feedback KIA: S295026 A2.03 Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:
Reactor pressure (CFR: 41.10 143.5 145.13)
RO/SRO Rating: 3.9/4.0 Objective: ClS-lP-300-l *05a
: 05. Given the PCCP, determine the appropriate actions if any of the following limits are approached or exceeded:
: a. Heat Capacity Temperature Limit.


==Reference:==
==Reference:==
 
Heat Capacity Temperature Graph only is given to examinee PCCP.
Heat Capacity Heat   Capacity Temperature Temperature Graph Graph only only isis given given to to examinee examinee PCCP.
Cog level: High Explanation:
PCCP.
HCTl has been exceeded. With rods unknown the operator would be in lPC.
Cog level:
Cog  Level: High High Explanation:
Explanation:
HCTL has been exceeded. With rods unknown the operator would be in HCTl                                                                              in lPC.
LPC.
Distractor Analysis:
Distractor Analysis:
Distractor Choice A: Plausible because rods are Choice                                are unknown, would be in lPC. LPC.
Choice A: Plausible because rods are unknown, would be in lPC.
Choice B: Correct Answer Choice C: Plausible because HCTl Choice                          HCTL has been exceeded. rods are unknown, would be in lPC           LPC Choice D: Plausible because HCTl HCTL has been exceeded.
Choice B: Correct Answer Choice C: Plausible because HCTl has been exceeded. rods are unknown, would be in lPC Choice D: Plausible because HCTl has been exceeded.
SRO Only Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations (43(b)(5)
SRO Only Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations (43(b)(5)
Notes Notes
Notes  
 
I I,,C CJ I
j) ti mm c
a 8 8 SUPPRESSION POOL WATER TEMPERATURE (&deg;F)
C)
&#xe7;)&#xe7;)QQ C)
-4
 
0 rnti Z
0
--9 0
Z 0
rn rn cn r>
rn C) rn-n 0
0 om m
mm UI I
T1
-o 00 COO m
r-
.rngm CO rn C
 
0 It rn
 
-Ui Cl) o
 
11[
m.,..
D IE[
r rn U
0 TEl W
 
liii I
I I 1 z <


ATTACHMENT 5
i-rn tit-r tW mm z
                                                                                                                                .rngm I                                                                                                            Page '18 of 27 FIGURE 3 T1
titt llttVtt rmi m
                      -                                                                        Heat Capacity Temperature Limit 0          ..                              r SUPPRESSION POOL WATER TEMPERATURE (&deg;F)
F
L1.
 
                      ~ 220 W
o ID 0
                      $                                         210                   ~                                    UNSAFE ABOVE ~
C Cl C
r-                    W m.,..rn z<
LII Cl 0 Cl UI CD fl 1flfll1
                      !;;c                                                  =f::l=Il
-I
                                                                            =~I                                             SELECTED LINE ~
---4 g
rn                        mm              rmi n::                                       200
ATTACHMENT 5 Page '18 of 27 FIGURE 3 Heat Capacity Temperature Limit L1.  
                      ~ 190 5j                                         180 l-n::                                       170
~ 220 W  
                                                                  -    -4
$ 210  
(-) 0.25 FT W
~  
1flfll1 Cl UI
=f::l=Il  
                      ~                                          160
!;;c
(-) 1.25 FT s:                                        150
=~I n:: 200 UNSAFE ABOVE ~
(-) 2.50 FT                           0
SELECTED LINE ~
                                                                                                                                                                                                                ---4
~ 190 5j 180 l-n:: 170 W
                      ..J
~ 160 s:
                                                                                                                                                            ~~~ (-) 3.25 FT o
..J 150 140 o o D..
Cl
Z o
:S:i~
=f:: f::
o                                                                                                                                    : (-)
SAFE BELOW 130 =~~ SELECTED LINE (I) 120 (I)
I I 1 tit-r tW titt llttVtt 140 D..                                                  =f:: f::            SAFE BELOW                                                =t=~     4.25 FT                       LII          fl cnr>  om      mm                                    IE[  ;=I=~
~ 110 D..
                                                                                                                                                                                                                -I
D.. 100
                                                                            =~~
::J (I)
It Z                                                                            rn-n                                      11[    TEl St=~
(-) 0.25 FT
o
(-) 1.25 FT
( I)
(-) 2.50 FT  
(I) 130 120 SELECTED LINE  I              m                          ,-p.,~
~~~ (-) 3.25 FT
:=I::~ (-) 5.50 FT
:S:i~  
=t=~ (-) 4.25 FT  
;=I=~
St=~  
,-p.,~
: =I::~ (-) 5.50 FT
:=~~
:=~~
I
:=~I=  
:=~I=
=1=1=
Cl C      -
                      ~ 110                                      -    0                                                                                      liiii-rn
                                                                                                                                                                =1=1=
:=I=~
:=I=~
D..                                                                                                                                  :=I=~
:=I=~  
D..                                        100                                                                                      ,-I-f-
,-I-f-I
::J                                                              -                                                -                                    -1,150 F
-1,150 100 300 500 700 900 1,100 o
0 (I)                                                                                                                                                                          C o
200 400 600 800 1,000 REACTOR PRESSURE (PSIG)
100 200 300 0
SUPPRESSION POOL WATER TEr.... 1PERATURE IS DETERMINED BY:
400 500 UI 600 700 800 900 1,000 I
CAC-TR-4426-*1A, POINT WTR AVG OR CAC-TR-4426-2A, POINTWTRAVG OR COrvlPUTER POINT G050 OR COMPUTER POINT G051 OR CAC-TY -4426-1 OR CAC-TY -4426-2 SELECT GRAPH LINE IMMEDIATELY BELOW SUPPRESSION POOL WATER LEVEL AS THE LIMIT.
1,100 0
I OEOP-01-UG Rev. 55 Page 78 of 151 I  
rn 0
COO REACTOR PRESSURE (PSIG)
C)  -      -o          CO            o      U SUPPRESSION POOL WATER TEr. .1PERATURE IS DETERMINED BY:
c j)                                                            Z        0    0        rn      m        -
Cl)      rn D                          m      ID CAC-TR-4426-*1A, POINT WTR AVG OR 8 8
                                                                                      --9 CAC-TR-4426-2A, POINTWTRAVG OR                                                                           00 I        COrvlPUTER POINT G050 OR
                        &#xe7;)&#xe7;)QQ COMPUTER POINT G051 OR CAC-TY-4426-1 OR CAC-TY -4426-2 SELECT GRAPH LINE IMMEDIATELY BELOW SUPPRESSION POOL WATER CJ                                                  C)                    Z rnti
                                                                                                                        -<                  C        -                    z        -  o    0                  -
mm
                                                                                                                                              -Ui LEVEL AS THE LIMIT.                                                         rn I
OEOP-01-UG I,,
C        a ti                                              C)                                                Rev. 55                                                                     Page 78 of 151 CD              g  -
I


YES NO MAiNTAIN REACTOR MAINTAIN  IIEACTORPRE5S PRESS BELOW THE BELOW    TIlE HEAT CAPACITY CAPACfY TEMP Ur.tIT UM1 IRRESPECTIVE 1RRESPECTPJE OF THE RESULTING OFTHERESULTING tOOLDOWN RATE COOLOOWN RCWI1 BNP VOLV1 IEOP-Ot- LPC REldlStONNO 9 UNIT i ONLY
YES BNP VOLV1 IEOP-Ot-LPC REldlStONNO 9 NO MAiNTAIN IIEACTORPRE5S BELOW TIlE HEAT CAPACfY TEMP UM1 1RRESPECTPJE OF THE RESULTING tOOLDOWN RATE RCWI1 UNIT i ONLY MAINTAIN REACTOR PRESS BELOW THE HEAT CAPACITY TEMP Ur.tIT IRRESPECTIVE OFTHERESULTING COOLOOWN RATE


INITIATEAREACTOR INITIATEA           REACTORSCRAM SCRAM AND ENTER AND      ENTER EOP-01 EOP-9i
INITIATEA REACTOR SCRAM AND ENTER EOP-9i SPIT-99 REDUCE REACTOR PRE PER THE RCIP SECTION OF EOP-91 AS NECESSARY TO REMAIN IN SAFE REGION OF HEAT CAPACITf TEMP UMIT SPIT-b
                        ~    ______-L______                  ~
/
SPIT-99 SP/T-09 REDUCEREACTOR REDUCE PERTHE PER REACTOR PRESS THE RCIP RCIP SECTION PRE SECTION OFOF I
CONSIDERANTICIRATIONOF EMERGENCY DEPRESSURIZA11ON
EOP- 01 EOP-      91 AS AS NECESSARY NECESSARY TO   TO REMAIN IN REMAIN       IN SAFE SAFE REGION REGION OF HEAT OF  HEAT CAPACITY CAPACITf TEMP UMIT TEMP     UMIT SPIT-b SPIT-10
\\
                /~~~~'~~'~~1iiii&#xa3;"'~~~~~"'''''\
PERRCIPSECTIONOF
CONSIDERANTICIRATIONOF I/ CONSIDER                 ANTICIPATION OF \
\\
EMERGENCY DEPRESSURlZAll0N
REACTOR VESSEL CONTROL
( EMERGENCY                    DEPRESSURIZA11ON \
\\ PROCEDURE{EOP.01. RVCP)
              \\               PERRCIPSECTIONOF PER   RCIP SECTION OF                   /
HCTL (MERGENCY DEPRESSURI THE REACTOR PER THE RCIP SECTION OF EOP-91 SPIT-IS BNPVOL-VI OEOP-02-PCCP REVISION NO 10 Categories K/A:
                \
S295026 A2.03 Tier / Group:
                \ ~REACTOR REACTOR VESSEL      VESSEL CONTROL/
T1G1 RO Rating:
CONTROL
3.9 SRO Rating:
                  \
4.0 LP Obj:
                  \~~~,:~~~~~:,,~~~~
CLSLP300L*05A Source:
PROCEDURE{EOP.01. RVCP)
PREV Cog Level:
                                              .1.                SPIT-11 HCTL HCTL                        ~/".r(;AN~,
HIGH Category 8:
                        ~//           THE HEAT       ---..........
Y SPIT-12 INITIATEA REACTOR SCRAM AND ENTER EOP-01 SP/T-09  
            </"               CAPACITY TEMP UMIT                   '-.,.-!Ii~L.".,
~
                  -.,...,. BE MAINTAINED IN THE /"-~
______ -L ______
                          -.,...,., SAFEREGION /""'"
~
                                    ---'::e~   NO SPIT- 12 SPIT-12 f
REDUCE REACTOR PRESS I PER THE RCIP SECTION OF EOP- 01 AS NECESSARY TO REMAIN IN SAFE REGION OF HEAT CAPACITY TEMP UMIT SPIT-10  
                          ;;;';EN;-~EP;;~~IZE (MERGENCY DEPRESSURI THE REACTOR PER THE RCIP SECTION OF EOP.01   EOP-91
/~~~~'~~'~~1iiii&#xa3;"'~~~~~"'''''\\
                        ~~~~~~~~.
I CONSIDER ANTICIPATION OF \\
SPIT-IS SPIT-13 BNPVOL-VI BNP VOL- VI OEOP-02-PCCP OEOP PCCP REVISION NO          NO: 10  10 Categories K/A:
( EMERGENCY DEPRESSURlZAll0N \\  
KIA:             S295026 A2.03                                                      Tier / Group: TIG!
\\
T1G1 RO  Rating:
PER RCIP SECTION OF  
RORating:       3.9                                                                 SRO  Rating:
/  
SRORating:    4.0 LP LP Obj:
\\  
Obj:        CLSLP300L*05A CLS-LP-300-L *05A                                                   Source:
~REACTOR VESSEL CONTROL/  
Source:       PREV PREY Cog Cog Level:
\\~~~,:~~~~~:,,~~~~  
Level:    HIGH HIGH                                                                Category Category 8:
.1.
8:  Y
SPIT-11 HCTL  
: 89. Unit
~/".r(;AN~,  
: 89.        Two isis operating Unit Two         operating at    at rated rated power  when half power when       half of of the the Orywell Drywell (OW)(DW) Coolers Coolers are are lost.
~//
lost.
THE HEAT  
Which one Which      one of    the following of the   following correctly correctly completes completes the the statements statements below?
</"
below?
CAPACITY TEMP UMIT  
(Assume (Assume initialinitial OW DW andand Suppression Suppression Pool Pool pressures pressures are are equal) equal)
'-.,.-!Ii~L.".,  
As OW As    DW temperature temperature rises,  rises, Suppression Suppression Pool Pool pressure pressure willwill rise rise at   (1) OW at (1)      DW pressure.
-.,...,. BE MAINTAINED IN THE /"-~  
pressure.
-.,...,., SAFEREGION /""'"
If DW If OW Air Air Temperature Temperature is      is not not restored restored to to within within the the LCO LCO limit limit in    (2) hours, in (2)       hours, the the Unit Unit is required is  required to to be be in   Mode 33 within in Mode        within the the following following 1212 hours hours per per TS TS 3.6.1.4 3.6.1.4 (Orywell (Drywell Air Air Temperature).
SPIT-12 NO
Temperature).
---'::e~
A. (1 A.   (1)) the the same same rate rate as as (2)
f
(2) 8 8 B. (1 B.   (1)) the the same same rate rate as as (2)
;;;';EN;-~EP;;~~IZE THE REACTOR PER THE RCIP SECTION OF EOP.01  
(2) 1212 C (1) a slower rate Cy                      rate than (2) 8 D. (1 O.   (1)) a slower rate than (2) 12
~~~~~~~~.
SPIT-13 BNP VOL-VI OEOP PCCP Categories KIA:
RORating:
LP Obj:
Cog Level:
REVISION NO: 10 S295026 A2.03 3.9 CLS-LP-300-L *05A HIGH Tier / Group: TIG!
SRORating: 4.0 Source:
PREY Category 8:
Y
: 89. Unit Two is operating at rated power when half of the Drywell (DW) Coolers are lost.
Which one of the following correctly completes the statements below?
(Assume initial DW and Suppression Pool pressures are equal)
As DW temperature rises, Suppression Pool pressure will rise at (1)
DW pressure.
If DW Air Temperature is not restored to within the LCO limit in (2) hours, the Unit is required to be in Mode 3 within the following 12 hours per TS 3.6.1.4 (Drywell Air Temperature).
A. (1) the same rate as (2) 8 B. (1) the same rate as (2) 12 C (1) a slower rate than (2) 8 D. (1) a slower rate than (2) 12
: 89. Unit Two is operating at rated power when half of the Orywell (OW) Coolers are lost.
Which one of the following correctly completes the statements below?
(Assume initial OW and Suppression Pool pressures are equal)
As OW temperature rises, Suppression Pool pressure will rise at (1)
OW pressure.
If OW Air Temperature is not restored to within the LCO limit in (2) hours, the Unit is required to be in Mode 3 within the following 12 hours per TS 3.6.1.4 (Orywell Air Temperature).
A. (1 ) the same rate as (2) 8 B. (1 ) the same rate as (2) 12 Cy (1) a slower rate than (2) 8 O. (1 ) a slower rate than (2) 12  


Feedback Feedback K/A: S295028 KIA:  S295028 A2.05 A2.05 Ability to Ability  to determine determine and/or andlor interpret interpret thethe following following as     they apply as they  apply toto HIGH HIGH ORYWELL DRYWELL TEMPERATURE:
Feedback K/A: S295028 A2.05 Ability to determine andlor interpret the following as they apply to HIGH DRYWELL TEMPERATURE:
TEMPERATURE:
Torus/suppression chamber pressure: Plant-Specific (CFR: 41.10 /43.5 /45.13)
Torus/suppression chamber Torus/suppression       chamber pressure:
RO/SRO Rating: 3.6/3.8 Objective: CLSLP004A*1 5a
pressure: Plant-Specific Plant-Specific (CFR: 41.10/43.5/45.13)
: 15. Given plant conditions, determine the effects that the following will have on the Primary Containment, Primary Containment Ventilation and Primary Containment Monitoring:
(CFR:   41.10 /43.5 /45.13)
: a. Loss of Drywell cooling.
RO/SRO Rating:
RO/SRO      Rating: 3.6/3.8 3.6/3.8 Objective:     CLSLP004A*1 Objective: CLS-LP-004-A          5a
                                *15a
: 15. Given   plant conditions, determine
: 15. Given plant conditions,        determine thethe effects effects that that the following will the following  will have have on on the the Primary Primary Containment, Containment, Primary Containment Primary    Containment Ventilation Ventilation and Primary Primary Containment Containment Monitoring:
Monitoring:
Loss of
: a. Loss       Drywell cooling.
of Drywell


==Reference:==
==Reference:==
SD-04, Revision 5, Page 25 TS Cog Level: High Explanation:
Reduced DW cooling or rising DW temperature results in DW pressure increases whose severity is dependent upon plant conditions. OAOP-14.0, Abnormal Primary Containment Conditions provides guidance on indications to be monitored and actions to be taken which include verification of cooling system lineups and reductions in power to maintain average temperature below 150&deg;F. Failure to accomplish this may require entry into the OEOP-02-PCCP Primary Containment Control.
Elevated DW temperature causes DW pressure to rise. As DW pressure rises, SP water level rises causing a rise in SP pressure. Due to the downcomers extending 3 feet below the surface of the SP water level a differential pressure will always exist. Temperature response is different from LOCA response due to steam AND non-condensibles being forced into the SP
- steam condensing and non-condensibles collecting in SP air space.
TS 3.6.1.4 (DW Air Temperature) limit of < 150&deg;F, CONDITION A - Drywell average air temperature not within limit, REQUIRED ACTION A.1 Restore drywell average air temperature to within limit has a COMPLETION TIME of 8 hours.
If temperature is not restored to.
150&deg;F, CONDITION B, REQUIRED ACTION 8.1 Be in MODE 3 has a COMPLETION TIME of 12 hours.
Distractor Analysis:
Choice A: Plausible because SP pressure is changed by the change in SP level only vs pressure, steam, and non-condensibles during a LOCA. The SP air space temperature is in equilibrium with SP water temperature (95&deg;Fmax during normal operations) rising DW pressure would have a direct impact on SP level. However during temperature only (no steam), the DW pressure increase is cushioned by SP water, small changes in SP water level provides small change in SP pressure.
8 hours to restore temperature is correct.
Choice B: Plausible because SP pressure is changed by the change in SP level only vs pressure, steam, and non-condensibles during a LOCA. The SP air space temperature is in equilibrium with SP water temperature (95&deg;Fmax during normal operations) rising DW pressure would have a direct impact on SP level. However during temperature only (no steam), the DW pressure increase is cushioned by SP water, small changes in SP water level provides small change in SP pressure.
12 hours is the time required to get to MODE 3 if not restored within the required Completion Time.
Choice C: Correct Answer Choice D: Plausible because rising at a slower rate is correct and 12 hours is the time required to get to MODE 3 if not restored within the required Completion Time.
SRO Only Basis: Application of required actions (Section 3) and surveillance requirements (Section 4) in accordance with rules of application requirements (Section 1). (43(b)(2)
Feedback KIA: S295028 A2.05 Ability to determine and/or interpret the following as they apply to HIGH ORYWELL TEMPERATURE:
Torus/suppression chamber pressure: Plant-Specific (CFR: 41.10/43.5/45.13)
RO/SRO Rating: 3.6/3.8 Objective: CLS-LP-004-A *15a
: 15. Given plant conditions, determine the effects that the following will have on the Primary Containment, Primary Containment Ventilation and Primary Containment Monitoring:
: a. Loss of Drywell cooling.


==Reference:==
==Reference:==
 
SD-04, Revision 5, Page 25 TS Cog Level: High Explanation:
SD-04, Revision SD-04,   Revision 5,5, Page Page 25 25 TS TS Cog Level: High Explanation:
Reduced OW cooling or rising DW temperature results in DW pressure increases whose severity is dependent upon plant conditions. OAOP-14.0, Abnormal Primary Containment Conditions provides guidance on indications to be monitored and actions to be taken which include verification of cooling system lineups and reductions in power to maintain average temperature below 150&deg;F. Failure to accomplish this may require entry into the OEOP-02-PCCP Primary Containment Control.
Explanation:
Elevated DW temperature causes DW pressure to rise. As DW pressure rises, SP water level rises causing a rise in SP pressure. Due to the downcomers extending 3 feet below the surface of the SP water level a differential pressure will always exist. Temperature response is different from LOCA response due to steam AND non-condensibles being forced into the SP - steam condensing and non-condensibles collecting in SP air space.
Reduced OW Reduced    DW cooling or rising DW  DW temperature results in      in DW pressure increases increases whose severity is dependent upon plant conditions. OAOP-14.0, Abnormal Primary Containment Conditions provides guidance on indications to be monitored and actions to be taken which include verification of cooling system lineups and reductions in power to maintain average temperature below 150&deg;F. Failure to accomplish this may require entry into the OEOP-02-PCCP Primary Containment Control.
TS 3.6.1.4 (DW Air Temperature) limit of ~ 150&deg;F, CONDITION A - Drywell average air temperature not within limit, REQUIRED ACTION A.1 Restore drywell average air temperature to within limit has a COMPLETION TIME of 8 hours. If temperature is not restored to ~
Elevated DW temperature causes DW pressure to rise. As DW pressure rises, SP water level rises causing a rise in SP pressure. Due to the downcomers extending 3 feet below the surface of the SP water level a differential pressure will always exist. Temperature response is different from LOCA response due to steam AND non-condensibles being forced into the SP - steam condensing and-non-condensibles collecting in SP air space.
TS 3.6.1.4 (DW Air Temperature) limit of ~         < 150&deg;F, CONDITION A - Drywell average air temperature not within limit, REQUIRED ACTION A.1 Restore drywell average air temperature to within limit has a COMPLETION TIME of 8 hours. If temperature is not restored to ~                       .
150&deg;F, CONDITION B, REQUIRED ACTION 8.1 Be in MODE 3 has a COMPLETION TIME of 12 hours.
150&deg;F, CONDITION B, REQUIRED ACTION 8.1 Be in MODE 3 has a COMPLETION TIME of 12 hours.
Distractor Analysis:
Distractor Analysis:
Choice A: Plausible because SP pressure is changed by the change in SP level only vs pressure, steam, and non-condensibles during a LOCA. The SP air space temperature is in equilibrium with SP water temperature (95&deg;Fmax during normal operations) rising DW pressure would have a direct impact on SP level. However during temperature only (no steam), the DW pressure increase is cushioned by SP water, small changes in SP water level provides small change in SP pressure.
Choice A: Plausible because SP pressure is changed by the change in SP level only vs pressure, steam, and non-condensibles during a LOCA. The SP air space temperature is in equilibrium with SP water temperature (95&deg;Fmax during normal operations) rising DW pressure would have a direct impact on SP level. However during temperature only (no steam), the DW pressure increase is cushioned by SP water, small changes in SP water level provides small change in SP pressure.
88 hours to restore temperature is correct.
8 hours to restore temperature is correct.
Choice Choice B: Plausible because SP pressure is changed by the change in                     SP level only vs pressure, steam, in SP and non-condensibles non-condensibles duringduring aa LOCA.
Choice B: Plausible because SP pressure is changed by the change in SP level only vs pressure, steam, and non-condensibles during a LOCA. The SP air space temperature is in equilibrium with SP water temperature (95&deg;Fmax during normal operations) rising DW pressure would have a direct impact on SP level. However during temperature only (no steam), the DW pressure increase is cushioned by SP water, small changes in SP water level provides small change in SP pressure.
LOCA. The SP air space temperature is in          in equilibrium with SP water temperature (95&deg;Fmax during             normal operations) rising DW during normal                          DW pressure pressure would have have a direct direct impact on SP     level. However during SP level.                during temperature only (no  (no steam),
12 hours is the time required to get to MODE 3 if not restored within the required Completion Time.
steam), the DW pressure increase is cushioned by  by SP   water, small changes SP water,            changes in    SP water in SP   water level level provides provides small small change change in in SP SP pressure.
Choice C: Correct Answer Choice D: Plausible because rising at a slower rate is correct and 12 hours is the time required to get to MODE 3 if not restored within the required Completion Time.
pressure.
SRO Only Basis: Application of required actions (Section 3) and surveillance requirements (Section 4) in accordance with rules of application requirements (Section 1). (43(b)(2)  
12  hours is 12 hours   is the time time required required to getget to MODE MODE 33 ifif not not restored within within the required Completion Completion Time.
Time.
Choice Choice C:   Correct Answer C: Correct    Answer Choice Choice D:D: Plausible Plausible because because rising   at aa slower rising at    slower rate  is correct rate is correct and and 12 12 hours  is the hours is the time time required required to to get get to to MODE 33 ifif not MODE              restored within not restored    within the the required required Completion Completion Time.
Time.
SRO SRO Only Only Basis:
Basis: Application Application ofof required required actions actions (Section (Section 3)3) and and surveillance surveillance requirements requirements (Section (Section 4)
: 4) in in accordance accordance withwith rules rules ofof application application requirements requirements (Section (Section 1).
1). (43(b)(2)
(43(b)(2)


Notes Notes Drywell  Air Temperature Drvwell Air Temperature 3.6.1.4 3.6.1.4 3.6 CONTAINMENT 3.6   CONTAINMENT SYSTEMS SYSTEMS 3.6.1.4 3.6.'1.4       Drvwetl Air Drywell      Temperature Air Temperature LCO 3.6.
Notes 3.6 CONTAINMENT SYSTEMS 3.6.1.4 Drvwetl Air Temperature Drvwell Air Temperature 3.6.1.4 LCO 3.6.1.4 Drywell average air temperature shall be 150&deg;F.
LCO     3.6.1.4
              '1.4        Drywell  average air Drywell average     air temperature temperature shall    be s '150&deg;F.
shall be     150&deg;F.
APPLICABIUTY:
APPLICABIUTY:
APPLICABILITY:          MODES '1,2, MODES      1,2, and and 3.3.
MODES 1,2, and 3.
ACTIONS ACTIONS CONDITION CONDITION                            REQUIRED ACTION REQUIRED                              COMPLETION TIME COMPLETION     TIME A.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
A      Drywell average air DrY'Nell                          A. 1 A.-I        Restore drywell average air 8 hours Restore                              hours temperature not within limit.
Drywell average air A. 1 Restore drywell average air 8 hours temperature not within limit, temperature to within limit.
limit,             temperature to within limitlimit.
B.
B.
B. Required Action and               B.l B.-I        Be in MODE 3.                       12 hours associated Completion Time not met not  met.                         AND B.2         Be in MODE 4.                       36 hours
Required Action and B.l Be in MODE 3.
: 4. Drywell Orywell Temperature A loss of RBCCW to the drywell results in drywell temperature and pressure increases whose severity is dependent upon plant conditions. OAOP-1 OAOP-14.0, 4.0, Abnormal Primary Containment Conditions provides guidance on indications to be monitored and actions to be taken which include verification of cooling system lineups and reductions in power to maintain average temperature below 150&deg;F.              '150&deg;F.
12 hours associated Completion Time not met.
Failure to accomplish this may require entry into the OEOP-02-PC    OEOP-02-PCCP   CP Primary Containment Containment Control.
B.2 Be in MODE 4.
Loss Loss of of RBCCW to the Drywell           due to all Orywell due        all RBCCW pumps tripping 1 SD-21 SO-21                                                Rev.
36 hours 4.
Rev. 55                                    Page Page 25 25 of  421 of 42 Categories Categories KJA:
Drywell Temperature A loss of RBCCW to the drywell results in drywell temperature and pressure increases whose severity is dependent upon plant conditions. OAOP-14.0, Abnormal Primary Containment Conditions provides guidance on indications to be monitored and actions to be taken which include verification of cooling system lineups and reductions in power to maintain average temperature below 150&deg;F.
KIA:             S295028 S295028 A2.05 A2.05                              Tier Tier // Group:
Failure to accomplish this may require entry into the OEOP-02-PCCP Primary Containment Control.
Group:   T1G1 T1 G 1 RO  Rating:
Loss of RBCCW to the Drywell due to all RBCCW pumps tripping SD-21 Rev. 5 Page 25 of 42 Categories KJA:
RORating:         3.6 3.6                                         SRO SRO Rating:
RO Rating:
Rating:  3.8 3.8 LP Obj:
LP Obj:
LP Obj:          CLSLPOO4A*    I 5A CLS-LP-004-A*15A                            Source:
Cog Level:
Source:         NEW NEW Cog Level:
S295028 A2.05 CLSLPOO4A* I 5A HIGH Tier / Group:
Cog Level:       HIGH mGH                                        Category Category 8:8:    YF YF
SRO Rating:
: 90. The
Source:
: 90. The following following plant plant conditions conditions existexist on on Unit  Two:
Category 8:
Unit Two:
T1G1 3.8 NEW YF 3.6 Notes Drywell Air Temperature 3.6.1.4 3.6 CONTAINMENT SYSTEMS 3.6.'1.4 Drywell Air Temperature LCO 3.6. '1.4 Drywell average air temperature shall be s '150&deg;F.
An ATWS
APPLICABILITY:
      - An
MODES '1,2, and 3.
        -    ATWS with with aa spurious spurious GroupGroup II Isolation Isolation has    occurred has occurred HPCI isis injecting
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A
      - HPCI
DrY'Nell average air A.-I Restore drywell average air 8 hours temperature not within limit.
        -            injecting to to the  RPV to the RPV        maintain RPV to maintain            level RPV level SUPPRESSION CHAMBER
temperature to within limit B.
      - SUPPRESSION
Required Action and B.-I Be in MODE 3.
        -                      CHAMBER LLVL          HI-HI isis inin alarm VL HI-HI            alarm Which one Which    one of   the following of the  following identifies identifies the the action action required required for    long term for long  term HPCI HPCI system system operation and operation  and thethe reason reason forfor this this action?
12 hours associated Completion Time not met AND 1 SO-21 Categories KIA:
action?
RORating:
When suppression When    suppression pool pool temperature temperature reaches reaches 140&deg;F, 140&deg;F,      (1)
LP Obj:
(1)    to to prevent prevent    (2)
Cog Level:
(2)  .
B.2 Be in MODE 4.
A. (1)
36 hours
A.      lower HPCI (1) lower    HPCI flow flow to to less less than than 2000 2000 gpmgpm lAW lAW LPC LPC (2) pump (2)  pump bearing bearing damage damage B. (1)
: 4.
B.       lower HPCI (1) lower    HPCI flow to to less less than than 2000 2000 gpm lAW    lAW LPC LPC (2) a loss of NPSH C
Orywell Temperature A loss of RBCCW to the drywell results in drywell temperature and pressure increases whose severity is dependent upon plant conditions. OAOP-14.0, Abnormal Primary Containment Conditions provides guidance on indications to be monitored and actions to be taken which include verification of cooling system lineups and reductions in power to maintain average temperature below '150&deg;F.
C~  (1) defeat the automatic suction transfer logic      logic and transfer HPCI  HPCI suction suction to the CST lAW SEP-10 (2) pump bearing damage D. (1) defeat the automatic suction transfer logic and transfer HPCI suction to the CST lAW SEP-10 (2) a loss of NPSH
Failure to accomplish this may require entry into the OEOP-02-PCCP Primary Containment Control.
Loss of RBCCW to the Orywell due to all RBCCW pumps tripping S295028 A2.05 3.6 CLS-LP-004-A*15A mGH Rev. 5 Tier / Group: T1 G 1 SRO Rating:


Feedback Feedback K/A: 295029 KIA:  295029 G2.01.07 G2.01.07 Ability to  evaluate  plant performance Ability to evaluate plant     performance and   and make make operational operational judgments judgments based based on  on operating operating characteristics, reactor characteristics,    reactor behavior, behavior, and and instrument instrument interpretation.
===3.8 Source===
interpretation.
NEW Category 8:
High Suppression High  Suppression PoolPool Water Water Level Level (CFR: 41.5/43.5/45.12/45.13)
YF Page 25 of 421
(CFR:  41.5/43.5/45.12/45.13)
: 90. The following plant conditions exist on Unit Two:
ROISRO Rating:
- An ATWS with a spurious Group I Isolation has occurred
RO/SRO      Rating: 4.4/4.7 4.4/4.7 Objective:
- HPCI is injecting to the RPV to maintain RPV level
Objective:
- SUPPRESSION CHAMBER LVL HI-HI is in alarm Which one of the following identifies the action required for long term HPCI system operation and the reason for this action?
LOl-CLS-LP-0l 9-A, 26g:
When suppression pool temperature reaches 140&deg;F, (1) to prevent (2)
LOI-CLS-LP-019-A,            Given plant 26g: Given           conditions and plant conditions   and one one of of the the following following events, events, use use plant plant procedures procedures toto determine the determine   the actions  required to actions required       control and/or to control         mitigate the and/or mitigate   the consequences consequences of  of the the event:
A. (1) lower HPCI flow to less than 2000 gpm lAW LPC (2) pump bearing damage B. (1) lower HPCI flow to less than 2000 gpm lAW LPC (2) a loss of NPSH C (1) defeat the automatic suction transfer logic and transfer HPCI suction to the CST lAW SEP-10 (2) pump bearing damage D. (1) defeat the automatic suction transfer logic and transfer HPCI suction to the CST lAW SEP-10 (2) a loss of NPSH
event:
: 90. The following plant conditions exist on Unit Two:
High Suppression High  Suppression Pool Pool water water level.
- An ATWS with a spurious Group I Isolation has occurred
level.
- HPCI is injecting to the RPV to maintain RPV level
- SUPPRESSION CHAMBER L VL HI-HI is in alarm Which one of the following identifies the action required for long term HPCI system operation and the reason for this action?
When suppression pool temperature reaches 140&deg;F, (1) to prevent (2)
A. (1) lower HPCI flow to less than 2000 gpm lAW LPC (2) pump bearing damage B. (1) lower HPCI flow to less than 2000 gpm lAW LPC (2) a loss of NPSH C~ (1) defeat the automatic suction transfer logic and transfer HPCI suction to the CST lAW SEP-10 (2) pump bearing damage D. (1) defeat the automatic suction transfer logic and transfer HPCI suction to the CST lAW SEP-10 (2) a loss of NPSH
 
Feedback K/A: 295029 G2.01.07 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.
High Suppression Pool Water Level (CFR: 41.5/43.5/45.12/45.13)
ROISRO Rating: 4.4/4.7 Objective:
LOl-CLS-LP-0l 9-A, 26g: Given plant conditions and one of the following events, use plant procedures to determine the actions required to control and/or mitigate the consequences of the event:
High Suppression Pool water level.


==Reference:==
==Reference:==
001-37.5 SUPPRESSION CHAMBER LVL HI-HI APP Cog Level
- High Explanation: HPCI system is normally aligned to the CST, with the torus high water level this transfers to the torus. this meets the KA by having to evaluate the suction path has transferred to the torus and the operational implications of the high torus tempeature on continued operation of the HPCI system. this requires the suction to be transferred back to the CST lAW SEP-I 0.
From : The lube oil and control oil for both HPCI and RCIC are cooled by the water being pumped. Very high lube oil temperatures can result in loss of lubricating qualities in the oil and thus cause damage to the bearings. Suction for HPCI and RCIC is aligned to the Condensate Storage Tank (CST) if it is available. The HPCI automatic suction transfer logic can be defeated to allow this lineup if necessary provided suppression pool temperature is approaching 140&deg;F.
Distractor Analysis:
Choice A: Plausible because reducing flow would be a correct action if HPCI NPSH was the concern. At high temperatures with low level in the torus this could be a correct action. Pump bearing damage is a correct statement.
Choice B: Plausible because reducing flow would be a correct action if HPCI NPSH was the concern. At high temperatures with low level in the torus this could be a correct action. a loss of NPSH would be correct for the reason to reduce flow.
Choice C: Correct answer, see explanation Choice D: Plausible because transferring the suction is correct but the concern is for pump bearing damage.
SRO Basis: 10 CFR 55.43(b)-S Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
Feedback KIA: 295029 G2.01.07 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.
High Suppression Pool Water Level (CFR: 41.5/43.5/45.12/45.13)
RO/SRO Rating: 4.4/4.7 Objective:
LOI-CLS-LP-019-A, 26g: Given plant conditions and one of the following events, use plant procedures to determine the actions required to control and/or mitigate the consequences of the event:
High Suppression Pool water level.


==Reference:==
==Reference:==
 
001-37.5 SUPPRESSION CHAMBER LVL HI-HI APP Cog Level - High Explanation: HPCI system is normally aligned to the CST, with the torus high water level this transfers to the torus. this meets the KA by having to evaluate the suction path has transferred to the torus and the operational implications of the high torus tempeature on continued operation of the HPCI system. this requires the suction to be transferred back to the CST lAW SEP-10.
001-37.5 001-37.5 SUPPRESSION CHAMBER SUPPRESSION        CHAMBER LVL          HI-HI APP LVL HI-HI   APP Cog Level Cog  Level - High
From: The lube oil and control oil for both HPCI and RCIC are cooled by the water being pumped. Very high lube oil temperatures can result in loss of lubricating qualities in the oil and thus cause damage to the bearings. Suction for HPCI and RCIC is aligned to the Condensate Sto(age Tank (CST) if it is available. The HPCI automatic suction transfer logic can be defeated to allow this lineup if necessary provided suppression pool temperature is approaching 140&deg;F.
            - High Explanation: HPCI Explanation:    HPCI system system isis normally normally aligned aligned to  the CST, to the CST, with with the the torus high water torus high  water level level this this transfers transfers to to meets the KA by having the torus. this meets                 having to evaluate the suction path has transferred to the torus and the operational implications of the high torus tempeature on continued operation of the HPCI system. this requires the suction to be transferred back to the CST lAW SEP-10.      SEP-I 0.
From : The lube oil and control oil for both HPCI and RCIC are cooled by the water being From:
pumped. Very high lube oil temperatures can result in loss of lubricating qualities in the oil and thus cause damage to the bearings. Suction for HPCI and RCIC is aligned to the Condensate Sto(age                 Storage Tank (CST) if it is available. The HPCI automatic suction transfer logic can be defeated to allow this lineup if necessary provided suppression pool temperature is approaching 140&deg;F.
Distractor Analysis:
Distractor Analysis:
Choice A: Plausible because reducing flow would be a correct action if HPCI NPSH was the concern. At high temperatures with low level in the torus this could be a correct action. Pump bearing damage is a correct statement.
Choice A: Plausible because reducing flow would be a correct action if HPCI NPSH was the concern. At high temperatures with low level in the torus this could be a correct action. Pump bearing damage is a correct statement.
Choice B: Plausible because reducing flow would be a correct action if HPCI NPSH was the concern. At high temperatures with low level in the torus this could be a correct action. a loss of NPSH would be correct for the reason to reduce flow.
Choice B: Plausible because reducing flow would be a correct action if HPCI NPSH was the concern. At high temperatures with low level in the torus this could be a correct action. a loss of NPSH would be correct for the reason to reduce flow.
Choice C: Correct answer, see explanation Choice D:0: Plausible because transferring the suctionsuction isis correct but the concern is    is for pump pump bearing bearing damage.
Choice C: Correct answer, see explanation Choice 0: Plausible because transferring the suction is correct but the concern is for pump bearing damage.
damage.
SRO Basis: 10 CFR 55.43(b )-5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.  
SRO SRO Basis:
 
Basis: 10 10 CFR CFR 55.43(b)-S 55.43(b )-5 Assessment Assessment of of facility facility conditions conditions and and selection selection ofof appropriate procedures procedures during during normal, abnormal, abnormal, and emergency emergency situations.
Notes MAXIMUM INJECTION
.,VSTEM PRESSURE (P5I)
CONDENSATE1FEEDWATER 1250 CRt FLOW NAY OE 1490 MAXIMIZED PER EOP0i. SEP-00 RCIC WITH SUCTION FROM CST IF AVAILABLE. DEFEAT LOW REACTOR 1190 PRESS AND HIGH AREA TEMPERATURE ISOLATION LOGIC IF NECESSARY PER CIRCUW ALTERATION PROC EDURE c,.oISEP-io HPCI WITH SUCTION FROM 1250 OST IF AVAILAbLE, OFIAT 1IPCI NI SUPPRESSION POOL I.EVEL SUCflO1 TRANSFER AND I-IGt-I AREA TEMPERATURE ISOLATION LOGIC IF NECESSARY PER CIRCUIt ALTERATION PROCEDURE EOP.0I-SEP-10)
IPUI-ESTALI5H RHR SERVICE 200 WATER FLOW AS SOON AS POSSIBLE CAUTION OPERATION OF HPGI OR ROIG WITH SUCTION TEMPERATURES ABOVE 140? MAY RESULT IN EQUIPMENT DAMAGE RC1L-23 Distractor plausibility:
CAUTION I
HPCI FLOW ABOVE 2000 GPM WITH SUCTION FROM CST AND GST LEVEL BELOW 5 FEET MAY RESULT IN VORTEXNG AND EQUIPMENT DAMAGE I
RCFL Categories K/A:
295029 G2.01.07 Tier/Group:
T1G2 RO Rating:
4.4 SRO Rating:
4.7 LP Obj:
19-A 26G Source:
BANK Cog Level:
HIGH Category 8:
TABLE I M1MUM SYSTEM INJECTION PRESSURES Notes 1
MAXIMUM SYSTEM INJECTlD.'I PRESSURES SYSTEM CONDENSATEIFEEDWA.TER CRD FI.OW NAY DIl:
MAXIMIZED PER eOP.Of. SEP. 09 RCiC WITH SUCTION FROM CST IF AVAILABLE. DEFEAT LOW REA.CTOR PRESS AND HIGH AREA TEMPERATURE ISOLATION LOGIC IF NECESSARY PER *CIRCUIT ALTERATION PROCEDURE" leop* Ot* sep* 10)
HPCI WITH SUCTION FROM CST II' AVAILABLe. OeFeAT HP(;I HI SUI"PRI!$$ION POOt. I.I!Yf!1.
SUCTION TAANSFSR ANI) HIGH ARIOA TIOMPERATURS ISOLATION lOGIC If: NeCeSSARY PER *CIRCUIT ALTeRATION PROCEDURe"
~EOP* 01* SEP* 10) lPCI* ESTABLISH RHR SERVICE WATER FI.OW AS 500H AS POSSIBLE CAUTION ERATION OF HPCI OR RCI H SUCTION TEMPERATURE OVE 140*' MAY RESULT IN EQUIPMENT DAMAGE Distractor plausibility:
CAUTION HPCI FLOW ABOVE 2000 GPM WITH SUCTION FROM CST AND CST LEVEL BELOW 5 FEeT MAY RESULT IN VORTEX<<NG AND EQUIPMENT DAMAGE Categories KIA:
RORating:
LPObj:
Cog Level:
295029 G2.01.07 4.4 19-A 26G HIGH MAXIM LIM INJECTION PRESSURE (PSIG) 1250 141)0 1190 12110 200 Tier / Group: T1 G2 SRO Rating:


Notes Notes TABLE 1I M1MUM SYSTEM MAXIMUM        SYSTEMINJECTlD.'I INJECTION PRESSURES PRESSURES MAXIM MAXIMUMLIM
===4.7 Source===
                                .,VSTEM INJECTION INJECTION SYSTEM PRESSURE PRESSURE (P5I)
BANK Category 8:
(PSIG)
: 91. Which one of the following identifies the controlling document and the required action to be taken if SJAE Offgas Radiation monitor readings increase 50% during steady state rated power operation?
CONDENSATE1FEEDWATER CONDENSATEIFEEDWA.TER                                        1250 1250 CRD    FLOW NAY CRt FI.OW            OE NAY DIl:                                    1490 141)0 MAXIMIZED PER MAXIMIZED      PER EOP0i.      SEP-00 eOP.Of. SEP. 09 RCIC WITH RCiC          SUCTION FROM WITH SUCTION          FROM CST IF IF AVAILABLE.
Notify E&RC to perform the Surveillance I Test Requirement (SRITR) required by (1)
AVAILABLE. DEFEAT DEFEAT LOW      REACTOR LOW REA.CTOR          1190 1190 CST PRESS AND PRESS      AND HIGH HIGHAREA        TEMPERATURE AREA TEMPERATURE ISOLATION LOGIC ISOLATION      LOGIC IF  IF NECESSARY NECESSARY PER *CIRCUIT PER  CIRCUW ALTERATION ALTERATION PROC    EDURE PROCEDURE" leop*  Ot* sep* 10) c,.oISEP-          io HPCI WITH HPCI  WITH SUCTION SUCTION FROM  FROM                            1250 12110 OST II' CST  IF AVAILABLe.
, which confirms the SJAE release rate is within limits within (2) following the monitor reading increase.
AVAILAbLE, OeFeAT OFIAT HP(;I 1IPCI HI NI SUPPRESSION POOt.
A. (1) ODCM 7.3.2, Radioactive Gaseous Effluent Monitoring Instrumentation (2) 4 hours B. (1) ODCM 7.3.2, Radioactive Gaseous Effluent Monitoring Instrumentation (2) 12 hours C (1) T.S. 3.7.5, Main Condenser Offgas (2) 4 hours D. (1) T.S. 3.7.5, Main Condenser Offgas (2) 12 hours
SUI"PRI!$$ION      POOL I.I!Yf!1.
: 91. Which one of the following identifies the controlling document and the required action to be taken if SJAE Offgas Radiation monitor readings increase 50% during steady state rated power operation?
I.EVEL SUCflO1 TAANSFSR SUCTION      TRANSFER ANI)   AND HIGH I-IGt-I AREA TIOMPERATURS ARIOA    TEMPERATURE ISOLATION ISOLATION LOGIC    IF NECESSARY PER lOGIC If: NeCeSSARY            PER *CIRCUIT CIRCUIt ALTERATION PROCEDURe" ALTeRATION      PROCEDURE EOP.0I-SEP      -
Notify E&RC to perform the Surveillance I Test Requirement (SRITR) required by (1)  
        ~EOP* 01* SEP* 10)  10)
,which confirms the SJAE release rate is within limits within (2) following the monitor reading increase.
IPUI- ESTABLISH lPCI*    ESTALI5H RHR    RHR SERVICE SERVICE WATER FI.OW 200 WATER      FLOW AS 500HSOON AS POSSIBLE CAUTION CAUTION OPERATION ERATION OF HPCI    HPGI OR  OR RCI ROIG WITH  SUCTION TEMPERAT H SUCTION        TEMPERATURE      URES ABOVE      140? MAY OVE 140*'      MAY RESULT IN EQUIPMEN EQUIPMENT        T DAMAGE RC1L- 23 Distractor Distractor plausibility:
A. (1) ODCM 7.3.2, Radioactive Gaseous Effluent Monitoring Instrumentation (2) 4 hours B. (1) ODCM 7.3.2, Radioactive Gaseous Effluent Monitoring Instrumentation (2) 12 hours C~ (1) T.S. 3.7.5, Main Condenser Offgas (2) 4 hours D. (1) T.S. 3.7.5, Main Condenser Offgas (2) 12 hours  
plausibility:
CAUTION CAUTION I    HPCI HPCI FLOW WITH FLOW ABOVE WITH SUCTION ABOVE 2000 SUCTION FROM    FROM CST 2000 GPM GPM CST AND AND CST LEVEL GST    LEVEL BELOWBELOW 55 FEET  FEeT MAY MAY RESULT RESULT IN    IN VORTEXNG VORTEX<<NG AND AND EQUIPMEN EQUIPMENT        T DAMAGE DAMAGE Categories I                      RCFL Categories K/A:
KIA:              295029 295029 G2.01.07 G2.01.07                                    Tier/Group:
Tier / Group: T1G2 T1 G2 RO  Rating:
RORating:        4.4 4.4                                                    SRO SRO Rating:
Rating: 4.7 4.7 LP Obj:
LPObj:            19-A 19-A 26G26G                                            Source:
Source:      BANK BANK Cog Cog Level:
Level:      HIGH HIGH                                                    Category Category 8:8:
: 91. Which
: 91. Which one one of of the the following following identifies identifies the the controlling controlling document document and and the the required required action action to be to  be taken taken ifif SJAE SJAE Offgas Offgas Radiation Radiation monitor monitor readings readings increase increase 50%
50% during during steady steady state    rated power state rated   power operation?
operation?
Notify E&RC Notify    E&RC to  to perform perform the the Surveillance Surveillance II Test Test Requirement Requirement (SRITR)
(SRITR) required required byby (1) ,which (1)    , which confirms confirms the the SJAE    release rate SJAE release     rate isis within within limits limits within within (2) (2)    following following the monitor the    monitor reading reading increase.
increase.
A. (1)
A.    (1) ODCM ODCM 7.3.2,7.3.2, Radioactive Radioactive Gaseous Gaseous Effluent Effluent Monitoring Monitoring Instrumentation Instrumentation (2)  4  hours (2) 4 hours B. (1)
B.        ODCM 7.3.2, (1) ODCM       7.3.2, Radioactive Radioactive Gaseous Gaseous Effluent Effluent Monitoring Monitoring Instrumentation Instrumentation (2) 12 (2)   12 hours hours C
C~   (1)  T.S. 3.7.5, (1) T.S. 3.7.5, Main Main Condenser Condenser Offgas Offgas (2) 4 hours D. (1) T.S. 3.7.5, Main Condenser Offgas (2) 12 hours


Feedback Feedback K/A: S29503BG KIA:    S295038G 2.02.42 2.02.42 Ability to Ability   to recognize recognize system system parameters parameters thatthat are are entry-level entry-level conditions conditions for for Technical Technical Specifications.
Feedback K/A: S295038G 2.02.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications.
Specifications.
High Off-Site Release Rate (CFR: 41.7/41.10 / 43.2/43.3/45.3)
High Off-Site High    Off-Site Release Release Rate Rate (CFR:   41.7/41.10 / 43.2/43.3/45.3)
RO/SRO Rating: 3.9/4.6 Objective: CLSLP30*08
(CFR: 41.7/41.10/43.2/43.3/45.3)
: 08. Given plant conditions and Technical Specifications, including the Bases, TRM, ODCM, and COLR, determine whether given plant conditions meet minimum Technical Specifications, TRM, or ODCM requirements associated with the Condenser Air Removal/Augmented Offgas System.
RO/SRO Rating:
RO/SRO       Rating: 3.9/4.6 3.9/4.6 CLSLP30*08 Objective: CLS-LP-30*OB Objective:
: 08. Given     plant conditions and OB. Given plant conditions      and Technical Technical Specifications, Specifications, including including the the Bases, Bases, TRM, TRM, ODCM, ODCM, andand COLR, COLR, determine whether determine      whether given given plant plant conditions conditions meet meet minimum minimum Technical Technical Specifications, Specifications, TRM, TRM, or or ODCM ODCM requirements associated requirements      associated with with the the Condenser Condenser AirAir Removal/Augmented Removal/Augmented Offgas Offgas System.
System.


==Reference:==
==Reference:==
101-03.1, Revision 10, Page 44, Item #57 (CODSR)
Cog Level: High Explanation:
NOTIFY E&RC to confirm release rate is within limits within 4 hours following a monitor reading increase of greater than or equal to 50% without an accompanying increase in thermal power. SR 3.7.5.1 Distractor Analysis:
Choice A: Plausible because the SJAE Rad Monitor operability is required by ODCM 7.3.2 and 4 hours is correct.
Choice B: Plausible because the SJAE Rad Monitor operability is required by 00CM 7.3.2 and 12 hours is a timeframe for another Required Action in this spec.
Choice C: Correct Answer Choice D: Plausible because TS 3.5.7 is correct and 12 hours is a timeframe for another Required Action in this spec.
SRO Only Basis: Application of Surveillance Requirements and timeframe greater than 1 hour.
Notes Feedback KIA: S29503BG 2.02.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications.
High Off-Site Release Rate (CFR: 41.7/41.10/43.2/43.3/45.3)
RO/SRO Rating: 3.9/4.6 Objective: CLS-LP-30*OB OB. Given plant conditions and Technical Specifications, including the Bases, TRM, ODCM, and COLR, determine whether given plant conditions meet minimum Technical Specifications, TRM, or ODCM requirements associated with the Condenser Air Removal/Augmented Offgas System.


==Reference:==
==Reference:==
101-03.1, Revision 10, Page 44, Item #57 (CODSR)
Cog Level: High Explanation:
NOTIFY E&RC to confirm release rate is within limits within 4 hours following a monitor reading increase of greater than or equal to 50% without an accompanying increase in thermal power. SR 3.7.5.1 Distractor Analysis:
Choice A: Plausible because the SJAE Rad Monitor operability is required by ODCM 7.3.2 and 4 hours is correct.
Choice B: Plausible because the SJAE Rad Monitor operability is required by ODCM 7.3.2 and 12 hours is a timeframe for another Required Action in this spec.
Choice C: Correct Answer Choice D: Plausible because TS 3.5.7 is correct and 12 hours is a timeframe for another Required Action in this spec.
SRO Only Basis: Application of Surveillance Requirements and timeframe greater than 1 hour.
Notes
ATTACHMENT 1 Page 39 of 67 ITEM SI-ilF CHECK LIST NOTES OPER FREQ TIME TSOPER NO.
MODE LIMITS RECORD SJAE OFF(4S RAD MONiTOR DD
: 1. 2. 3 b
07-13 012-RM-K60tA. NOTIFY E&RC to confirm release rate is within lmits within 4 hours oIIowing a monitor reading increase of greater than or equal to 50% without an l3-1 accompanying increase in thermal pcwer.
SR 3.7.5.1 RECORD S4E OFFGAS R40 MON.TOR CD
: 1. 2. 3 07-12 D12-RM-K6018. NOTIFY E&RC to confirm reLease rate is within limits within 4 hours following a nionitor reading increase of greater than or equal to 50% without an accompanying increase in thermal pcf.ver.
SR 3.7.5.1 PERFORM channel check utilizing the
07-13 Reference calculabon on Table I SJAE OFF-GAS RAE) calcufaton on MONITORS 012-RM-keoIA and B 00CM Table I TR 7.3.2-1 FunctionS, TR 7.3.2.1 PERFORM channel check on SERViCE R
c 07-13 channel WATER EFFLLJENTRAD MONITOR operabe 012-RM-K605. 00CM Tabie 7,3. 1-i Function 3, TR 7.3.1.1 i
PERFORM channel check on RAE)WASTE 8
c 07-13 channel EFFLUENT RAD MONITOR D12-RM-K6C4 operabe on Control Room Panel 2-H12-P804 with recorder D12-ROO1 on XU-2. CCCM Tab!e 7.3.1-i, Item 1. TR7.3.l.I During operation of the main condenser air ejector.
SHIFT Dayshift BRUNSWICK STEAM ELECTRIC PL4NT DAILY SURVE1LLANCE REPORT CONTROL OPERATORS 101-03.1 Rev. 101 ITEM SHIFT CHECK LIST NOlES NO.
57 RECORD SJAE OFFGAS RADMONITOR DD D12-RM-K6Cl"IA. NOTIFY E&RC to confilm rel~ase rate is within limits,... ijhin 4 hrurs follO\\lotng a monitor reading increase o*f greater than or equal to 50% without an accompanying increase in thermal pow~r.
SR3.7.5.1 58 RECORD SJAE OFFGAS RAD MONITOR DD D12-RM-K601B. NOTIFY E&RC to confirm re!~ase rate is within limits within 4 hours follo',\\;ng a monitor reading increase of greater than or equal to 50% without an accompanying increase in thermal pow~r.
SR 3.7.5.1 59 PERFORM ohannel check utilizing the calculation on Table SJAE OFF-GAS RAD MONITORS D'12-RM-K601A and B ODCM TR 7.3.2-1 Function 6, rn 7.3.2.1 00 PERFORM channel cheok on SERVICE R
~\\-:41ER EFFLUENT RAD MONITOR D12-RM-K605, ODCM Table 7.3.1-1, Function 3, TR 7.3.1.1 61 PERFORM ohannel check on P.ADI*lIASTE EFFLUENT RAD MOM TOR D12-RM-K604 en Control Room Pan~12-H12-Pa04 'Qi1h recorder D'12-ROOt on XU-3, o!)CM Table 7.3.1-1, !t~m I, TR 7.3.1.1
'During operation of the main oondenser air ejector.
SHIFT Davshift 1'101-03.1 OPER MODE 1,2' 3' 1,2',3'
'6 6
6 ATTACHMENT 'I Page 39 of6?
FREQ TIME':
b 07-13 13-19 b
07-13 13-19 0
07-13 e
07-13 e
07-13 TSiOPER LIMITS Refer~nce calculation on Table 1 channel operable channel operable BRUNSWICK STEAM ELECTRIC PL4.NT DAilY SURVEIllANCE REPORT CONTROL OPERATORS Rev. 101


101-03.1, Revision 101-03.1,   Revision 10, 10, Page Page 44, 44, Item  #57 (CODSR)
Main Condenser Offgas 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Main Condenser Offgas LCO 3.7.5 The gross gamma activity rate of the noble gases measured at the main condenser air ejector shall be 243,600 pCiisecond after decay of 30 minutes.
Item #57  (CODSR)
APPLICABILITY:
Cog  Level: High Cog Level:      High Explanation:
MODE 1, MODES 2 and 3 with any main steam line not isolated and steam jet air ejector (SJAE) in operation.
Explanation:
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
NOTIFY E&RC NOTIFY      E&RC to confirm release rate rate is within limits within 4 hours hours following a monitor monitor reading reading increase of greater than or equal to 50% without an accompanying accompanying increase in thermal power. SR    SR 3.7.5.1 Distractor Analysis:
Gross gamma activity rate of A.1 Restore gross gamma 72 hours the noble gases not within activity rate of the nobie
Distractor Choice A: Plausible because the SJAE Rad Monitor operability is required by ODCM 7.3.2 and 4 hours is correct.
: limit, gases to within limit.
Choice B: Plausible because the SJAE Rad Monitor operability is required by ODCM        00CM 7.3.2 and 12 hours is a timeframe for another Required Action in this spec.
B.
Choice C: Correct Answer Choice D: Plausible because TS 3.5.7 is correct and 12 hours is a timeframe for another Required Action in this spec.
Required Action and B.1 Isolate all main steam lines. 2 hours associated Completion Time no met.
SRO Only Basis: Application of Surveillance Requirement Requirements  s and timeframe greater than 11 hour.
B.2 Isolate SJAE.
Notes
12 hours OR B.3.1 Be in MODE 3.
12 hours AND B.3.2 Be in MODE 4.
36 hours Brunswick Unit 1 3.7-18 Amendment No. 203 3.7 PLANT SYSTEMS 3.7.5 Main Condenser Offgas Main Condenser Offgas 3.7.5 LCO 3.7.5 The gross gamma activity mte of the noble gases measured at the main condenser air ejector shall be ::; 243,600 jJCilsecond after decay of 30 minutes.
APPLICABILITY:
: MODEl, MODES 2 and 3 with any main steam line not isolated and steam jet air ejector (SJAE) in operation.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Gross gamma activity rate of.1l.* 1 Restore gross gamma 72 hours the noble gases not within activity mte of the noble limit.
gases to within limit B.
Required Action and B.1 Isolate all main steam lines. 12 hours associated Completion Time not met.
OR B.2 Isolate SJAE.
12 hours OR B.3.1 Be In MODE 3.
12 hours AND B.3.2 Be in MODE4.
313 hours Brunswick Unit 1 3.7-18 Amendment No. 203


ATTACHMENT ATTACHMENT 'I1 Page Page 3939 of6?
Main Condenser Offgas 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1,
of 67 ITEM ITEM                      SI-il FCHECK SHIFT      CHECKLIST    LIST                NOTES NOlES  OPER OPER            FREQ FREQ    TIME':
TIME      TSiOPER TSOPER NO.
NO.                                                                          MODE MODE                                    LIMITS LIMITS RECORD SJAE RECORD        SJAEOFFGASOFF(4SRADMONITOR RAD MONiTOR 57                                                                    DD DD  1. 2. 3' 1,2'    3        bb      07-13 07-13 012-RM-K60tA. NOTIFY D12-RM-K6Cl"IA.            NOTIFYE&RC  E&RCto  toconfilm confirm release rate rel~ase    rate isis within withinlimits lmits,...within    hours ijhin 44 hrurs oIIowing aamonitor follO\lotng    monitorreading reading increase increase o*f of greaterthan greater    than or      equal to orequal      50%without to 50%    withoutan  an 13-19 l3-1 accompanying increase accompanying          increase inin thermal thermal pow~r.
pcwer.
SR 3.7.5.1 SR3.7.5.1 RECORD SJAE RECORD        S4E OFFGASOFFGASRAD    R40 MONITOR MON.TOR 58                                                                    CD DD  1. 2. 3 1,2',3'          b      07-13 07-12 D12-RM-K6018. NOTIFY D12-RM-K601B.            NOTIFY E&RC  E&RC to    confirm to confirm reLease rate re!~ase    rate isis within within limits limits within within 44 hours hours following aa monitor follo',\;ng    nionitor reading reading increase increase of of greater than greater    than or or equal equal to to 50%    without an 50% without        an 13-19 accompanying increase accompanying          increase in    thermal pow~r.
in thermal    pcf.ver.
3.7.5.1 SR 3.7.5.1 SR PERFORM ohannel PERFORM        channel check          utilizing the check utilizing      the 59                                                                            '6 0      07-13 07-13      Refer~nce Reference calculabon on calculation          Table I SJAE on Table        SJAE OFF-GAS OFF-GAS RAD  RAE) calculation calcufaton onon MONITORS D'12-RM-K601A MONITORS          012-RM-keoIA and        and BB ODCM 00CM Table Table 1I TR  7.3.2-1 Function TR 7.3.2-1      FunctionS,   6, rn TR 7.3.2.1 7.3.2.1 PERFORM channel PERFORM        channel cheokcheck on  on SERVICE SERViCE 00                                                                    R R        6            ec      07-13 07-13        channel channel WATER EFFLUENT
            ~\-:41ER    EFFLLJENTRAD            MONITOR RAD MONITOR operable operabe 012-RM-K605. ODCM D12-RM-K605,          00CM TableTabie 7.3.1-1, 7,3. 1-i Function 3, Function    3, TR TR 7.3.1.1 7.3.1.1 i      PERFORM ohannel PERFORM        channel check on      on P.ADI*lIASTE RAE)WASTE 61                                                                              8 6            ec      07-13        channel EFFLUENT RAD EFFLUENT        RAD MOM  MONITOR        D12-RM-K6C4 TOR D12-RM-K604                                                          operable operabe on Control en  Control Room Room Pan~12-H12-Pa04 Panel 2-H12-P804 'Qi1h    with recorder D'12-ROOt recorder    D12-ROO1 on XU-3,  XU-2. o!)CM CCCM Tab!eTable 7.3.1-i, !t~m 7.3.1-1,  Item I,  1. TR7.3.l.I TR 7.3.1.1 operation of During operation
'During              of the mainmain oondenser condenser air ejector. ejector.
SHIFT SHIFT        Dayshift Davshift                                                    BRUNSWIC BRUNSWICK                        PL4.NT K STEAM ELECTRIC PL4NT DAILY SURVE1LL DAilY SURVEIllANCE ANCE REPORT CONTROL OPERATOR CONTROL  OPERATORS  S 101-03.1 1'101-03.1                                                                              Rev. 101 101


Main Condenser Main  Condenser Offgas Offgas 3.7.5 3.7.5 3.7 PLANT 3.7  PLANT SYSTEMS SYSTEMS 3.7.5 Main 3.7.5    Main Condenser Condenser Offgas Offgas LCO 3.7.5 LCO    3.7.5            The gross The  gross gamma gamma activity activity mte rate of    the noble of the  noble gases gases measured measured at at the the main main condenser air condenser    air ejector ejector shall shall be      243,600 jJCilsecond be ::; 243,600    pCiisecond after after decay decay of of minutes.
NOTE
30 minutes.
30 APPLICABILITY:
APPLICABILITY:          MODE 1, MODEl, MODES 22 and MODES        and 33 with with any any main      steam line main steam      line not not isolated isolated and and steam steam jetjet air air ejector (SJAE) ejector  (SJAE) in  in operation.
operation.
ACTIONS ACTIONS CONDITION CONDITION                              REQUIRED ACTION REQUIRED                              COMPLETION TIME COMPLETION          TIME A. Gross gamma activity rate of .1l  A.1
                                          .* 1        Restore gross gamma                  72 hours gases not within the noble gases                                activity mte activity  rate of the noble nobie limit, limit.                                          gases to within limitlimit.
B. Required Action and                B.1          Isolate all main steam lines. 2      12 hours associated Completion Time no met.
not                                OR B.2          Isolate SJAE.                        12 hours OR B.3.1        Be In in MODE 3.                      12 12 hours AND B.3.2 B.3.2        Be in in MODE MODE4. 4.                  313 hours 36 Brunswick    Unit 11 Brunswick Unit                                      3.7-18 3.7-18                                Amendment No.
Amendment        No. 203 203


Main Condenser Main  Condenser OffgasOffgas 3.7.5 3.7.5 SURVEILLANCE REQUIREMENTS SUR'.,.'EILLANCE   REQUIREMENTS SURVEILLANCE SURVEILLANCE                                              FREQUENCY FREQUENCY SR 3.7.5.1 SR 3.7.5.1,                                    NOTE
Not required to be performed until 31 days after any main steam line not isolated and SJAE in operation.
                    -----------------------------NOTE----------------------------
Verify the gross gamma activity rate of the noble 31 days gases is 243600 pCiisecond after decay of 30 minutes.
Not required Not  required to  to be be performed performed until until 31 31 days days after after any any main steam main  steam line line not not isolated isolated and and SJAE SJAE in in operation.
AND Once within 4 hours afteraa5O%
operation.
increase in the nominal steady state fission gas release after factoring out increases due to changes in THERMAL POWER level 8runswick Unit 1 3.7-19 Amendment No. 203 SUR'.,.'EILLANCE REQUIREMENTS SURVEILLANCE Main Condenser Offgas 3.7.5 FREQUENCY SR 3.7.5.1  
Verify the Verify  the gross     gamma activity gross gamma      activity rate rate of of the the noble noble        31 days 31  days gases is gases    is ::; 243,600 243600 ~Cilseoond pCiisecond after after decay decay of of 30 minutes.
-----------------------------NOTE----------------------------
30  minutes.                                                         AND Once within 4 hours Once              hours after a :::50%
Brunswick Unit 1 Not required to be performed until 31 days after any main steam line not isolated and SJAE in operation.
afteraa5O%
Verify the gross gamma activity rate of the noble gases is ::; 243,600 ~Cilseoond after decay of 30 minutes.
increase in increase    in the the nominal steady nominal    steady state fission gas slate release after factoring oul out increases due to changes in THERMAL POWER level 8runswick Brunswick Unit Unit 11                                  3.7-19 3.7-19                            Amendment Amendment No. 1\10. 203 2113
3.7-19 31 days Once within 4 hours after a :::50%
increase in the nominal steady slate fission gas release after factoring oul increases due to changes in THERMAL POWER level Amendment 1\\10. 2113  


Radioactive Gaseous Radioactive              Gaseous Effluent Effluent Monitoring               Instrumentation Monitoring Instrumentatlon 7.3.2 7.3.2 Table 7.:?.2-1 T<ible  7.12- ipage          of4)4) aage 22 off Radactie Radioacti    **-e C-,aSECUS    Effluent~,''''itocn!!
Radioactive Gaseous Effluent Monitoring Instrumentation 7.3.2 Table 7.12-aage 2 of 4)
I3asecusEffluent        Mitang1nslrurr,entat'crI reburrentatcn FJNC.'"
Radactie I3asecus Effluent Mitang reburrentatcn FJNC.
FUNCTlml                            PJCA6_E
PJCA6_E aL:iEt ccrilyr1cr4 TE&T ftp tLC0S 3ft C%NLS 9EFEfNCO aECuIRE,lETS 3TPCNT OThER ZER ft REJIRD S,CIFIED FJt2TDt CCt.FEriSArOv CCt4DrIC?3 tE4.SLL9E3 Al 2.
                                                      ,.P?LlCA8:'E                aL:iEt REaLiI.~EO              ccrilyr1cr4 CCNDlTICNS                    TEST TE&T             Al.ARWiRiP ftp tLC0S 3ft fb:::O:SOR                C%NLS CH>.NNELS                9EFEfNCO REfERIENCEO              aECuIRE,lETS REQUIREMENTS              S:=TPO!NT 3TPCNT OThER OThER                        ZER
Rea,tsor Buildng Ventilatcn Ffonhtorir System (continueo}
                                                                                      ?ER            fROM    REJIRD ft REQUIRED                                          V,.LUE S,CIFIED S"=:CIFIEO                FJt2TDt F;Jr..;:::TiO:-I      CCt.FEriSArOv CCI/PEN!l"'TO.RY CCt4DrIC?3 CCN:)liIC/,"S                                      tE4.SLL9E3 Al MEASU~ES"      *. l 2.
e.
: 2. Re:actor    Buildng Ventilation Rea,tsorBuild\!1g    Ventilatcn Ffonhtorir Systsm Monitoring        System (continueo}
&#xa3;airplerFlcwRaie At at tme; 1
(rontinuec}
0 TR 7.3.2.1 srernent Device TR 7.12.6 TR 7.3.2. C 3.
e.e.    &#xa3;airplerFlcwRaie Sarrpler      Row Rate                    at fmas At all At          tme;                   1                     D0                  TR 7.3.2.1 TR    7.3.2.1              (0;1 srernent Device; Measurerr,ent      Device                                                                                        TR 7.3.2.6 TR   7.12.6 TR 7.3.2.10 TR  7.3.2. C 3.
Turbine Buildng Venblatcn Monito,ir System a.
: 3. Turbine Buildillg TLiltine    Buildng Venlilaton Venblatcn Monito,ir Systam Monitorir"        System
NobieOasMtWiy Atalmes 1
: 3. a. NobieOasMtWiy Notle      Gas Acti'.i.ly            Atalmes
3 TR 7.a2.
                                                      .<\t all tim;;s                    1                      3                 TR 7.32.;
(b Atnikr TR 7.3.2.3 TR 7.3.15 TR 7.3.2.10 b.
TR   7.a2.                 (b)
bineSanpIer Ata[tmes 1
(b Atnikr M::fIitcr                                                                                                          TR 7.3.2.3 TR   7.3.2.3 7.3.15 TR 7.3.2.5 TR TR 7.3.2.10 TR  7.3.2.10 b.b. bineSanpIer looine    Sarrpler                  Ata[tmes At  'II! limss                    1                     C C                  TR 7.3.22 TR   7.3.12               NA NA Cartroge Cmr:dge c.o. Parclate S<1mj:ler PartiCIIlate    Sampler             Ata[ lim;;s AtaN      bmes                   1                     C C                  TR 7.3.2.2 TR   7.3.2.2               NA NA Ftter Fiiter d.
C TR 7.3.12 NA Cartroge o.
: d. System       Effluent Flem System Effiuem        Fbiw         At alla lim;;s tines                   1                     00                  TR 7.3.2 7.3.2i .*            NA NA Rate Measurement Rate                                                                                                              TR 7.3.2.0 7.3.2.
Parclate Sampler Ata[ bmes 1
Device De-~ice                                                                                                          TR 7.3.2.10 e.
C TR 7.3.2.2 NA Ftter d.
: e. low    Rar,e Sarrpler Lcv Range      Lairpler           At aU      tines al lim;;s                    1                     00                        7.3.21 TR 7.3.2.1                    (e)
System Effluent Fbiw At a tines 1
(Cl F Rate FbwRate                                                                                                          TR 7.3.2.0 7.3.2.5 easarernent Device; Measurement        Desice                                                                                         TR 7.3.2.10 f.f. Md:High Rang;;
0 TR 7.3.2i NA Rate Measurement TR 7.3.2.
MeiHigh      Rante                         (rn:
Device TR 7.3.2.10 e.
(m)                        1                     0                   TR 77.3.2.10
Lcv Rar,e Lairpler At al tines 1
                                                                                                                                          ..3.2.10             NA Sampler ReI'.'
0 TR 7.3.21 (Cl F Rate TR 7.3.2.5 easarernent Desice TR 7.3.2.10 f.
Sarrpler      F1cv Rate Measurenrent Device; ME-<1SlJrement    0eioe 4.
Md:High Rante (rn:
: 4. ManCondenserOtf-Gas Main    Condenser Off-Gas                          (e)
1 0
(ei                       1                       BB                  TR 1.3.2.1 7.32.1                 (bJ Treatment Treatment Systam System Ncble Noble                                                                                              7.3..2.3 TR 7.3.2.3 Gsa Gas Acbiity       Mcnhtor Activity Monitor    '"                                                                                                7.12.6 TR 7.3..2.6 (0onseam (Dovmstream of AOG                                                                                                       R 7.3.2.10 TR Treatment Treafment Systam)System)
TR 7.3.2.10 NA Sampler F1cv Rate Measurenrent 0eioe 4.
(CQntinued)
ManCondenserOtf-Gas (ei 1
B TR 7.32.1 Treatment System Ncble TR 7.3.2.3 Gsa Acbiity Mcnhtor TR 7.12.6 (0onseam of AOG R 7.3.2.10 Treatment System)
(continued)
(continued)
(a)
(a)
(a)  Specif Speci/a  a inatn.anentatii instrumentatioo dentlicaban idE!1~licalion nuntere numbers are   <ire provided provided in    in Appendix   E.
Specif a inatn.anentatii dentlicaban nuntere are provided in Appendix 3.
AppEIldix 3.
b)
b)
(b)  Alarnscrp AlarJl1l~rip setpcints setpoinfs ahat shall be be detemiine determined in    in accocdaie acoocdance teth\";th 00CM ODCM methodacgy melhodo:cgy and  and set    10 ensure set ia ensure thethe limta lim(s of oi 000MS ODCMS 7.3.7,         'Dose RateGaseous 7.3.7. Dose    Rate-Gaseous Effluents, Effluents; are  are nt no! eteeeded.
Alarnscrp setpcints ahat be detemiine in accocdaie teth 00CM methodacgy and set ia ensure the limta of 000MS 7.3.7, Dose RateGaseous Effluents, are nt eteeeded.
elQOeeaed.
Ic)
Ic)
(c}  Alarmr AJarll1/tr,p seoints setpoints shall shall be be deterntined detemlined in         oY..o::roance eith in aconntiartce    \\;th asscctsed assccia:;ed desi9n     specitcatiDn(5) and desig\!1specifoation{s)      and setto set to ensure anS!!re the  limits at the ltnit.s of COCMS COCMS 7.3.7,7.:'..7. Dose
Alarmr seoints shall be deterntined in aconntiartce eith asscctsed desi9n specitcatiDn(5) and setto ensure the ltnit.s at COCMS 7.3.7, Dose RateGaseous Effluents, are na exceeded.
                              'Dose RateGaseous           Effluents; are Rate---Gaseous Effluents,           are na not exceeded.
exceec'ed.
d)
d)
(d)  Provides Provides alarm.
Provides alarm.
alaml.
ie)
ie)
(e)  During Main Doring   Main Condenser Coodenser 0ff-Gas Oft-Gas Tresurent Treatment System   Systemoperatcn operation irn)
Doring Main Condenser 0ff-Gas Tresurent System operatcn irn)
(m)  During During MWHig MdiHigll RareSymaac Rang;; System opeaIion      n Brunswidc Brunswick Units Units 11 and and 22                                      7.3.2-18 7.3.2-10                                                                      Rev. 32 Rev.      32 I1 Categories Categories K/A:
During MWHig RareSymaacn Brunswidc Units 1 and 2 7.3.2-18 Rev. 32 I Categories K/A:
KIA:                  S295038G2.
S295038G2.02.42 Tier/Group:
S295038G 2.02.42      02.42                                                   Tier/Group:
T1G1 RO Rating:
Tier / Group:              TIGl T1G1 RO   Rating:
3.9 SRO Rating:
RORating:              3.9 3.9                                                                            SRO     Rating:
4.6 LP Obj:
SRORating:                 4.6 4.6 LP  Obj:
CLSLP3O*O8 Source:
LPObj:                 CLSLP3O*O CLS-LP-30*08           8                                                      Source:
NEW Cog Level:
Source:                  NEW NEW Cog Cog Level:
HIGH Category 8:
Level:          HIGH HIGH                                                                            Category Category 8:8:              YY
Y Radioactive Gaseous Effluent Monitoring Instrumentatlon T<ible 7.:?.2-1 ipage 2 off 4)
: 92. The
Radioacti ** -e C-,aSECUS Effluent ~,''''itocn!! 1nslrurr,entat'crI FUNCTlml '"
: 92. The following following plant plant conditions conditions exist exist on on Unit Unit Two Two due due to to aa malfunction malfunction of of the the Air Air Dryer:
,.P?LlCA8:'E REaLiI.~EO CCNDlTICNS TEST fb:::O:SOR CH>.NNELS REfERIENCEO REQUIREMENTS OThER
Dryer:
?ER fROM REQUIRED S"=:CIFIEO F;Jr..;:::TiO:-I CCI/PEN!l"'TO.RY CCN:)liIC/,"S MEASU~ES"
SERVICE AIR
*. l
        - SERVICE
: 2.
        -            AIR PRESS-LOW PRESS-LOW isis in    in alarm alarm
Re:actor Build\\!1g Ventilation Monitoring Systsm (rontinuec}
        - RB
: e.
        -     INSTR AIR RB INSTR     AIR RECEIVER RECEIVER 2A  2A PRESS PRESS LOW LOWis     in alarm is in  alarm RB INSTR
Sarrpler Row Rate At all fmas D
        - RB
TR 7.3.2.1 Measurerr,ent Device; TR 7.3.2.6 TR 7.3.2.10
        -      INSTR AIR     RECEIVER 2B AIR RECEIVER              PRESS LOW 2B PRESS      LOW isis inin alarm alarm Instrument Air
: 3.
        - Instrument
TLiltine Buildillg Venlilaton Monitorir" Systam
        -                Air pressure pressure isis 93 93 psig psig and and recovering recovering Based on Based    on the the above above indications, indications, which which one one of of the the following following correctly correctly identifies:
: 3.
identifies:
Notle Gas Acti'.i.ly
(1)  the  status  of the  Service (1) the status of the Service Air  Air Dryer Dryer Bypass Bypass Valve, Valve, SA-PV-5067, SA-PV-5067, and and (2) the (2)  the procedure procedure that that contains contains the the steps steps to to close close the the Reactor Reactor Building Building Inboard Inboard and and Outboard Isolation Outboard    Isolation Valves Valves (BFIVs)?
.<\\t all tim;;s TR 7.32.;
(BFIVs)?
M::fIitcr TR 7.3.2.3 TR 7.3.2.5 TR 7.3.2.10
A'I  (1) open A (1)    open (2) OAOP-20.O, Pneumatic (2) OAOP-20.0,       Pneumatic (Air/Nitrogen)
: b.
(Air/Nitrogen) System System Failures Failures B. (1) open (2) 2APP-UA-01, 2APP-UA-O1, Service Air Press-Low C. (1) closed (2) OAOP-20.0, OAOP-20.O, Pneumatic (Air/Nitrogen) System Failures D. (1) closed (2) 2APP-UA-01, 2APP-UA-O1, Service Air Press-Low
looine Sarrpler At 'II! limss C
TR 7.3.22 Cmr:dge
: c.
PartiCIIlate S<1mj:ler AtaN lim;;s C
TR 7.3.2.2 Fiiter
: d.
System Effiuem Flem At all lim;;s 0
TR 7.3.2.*
Rate Measurement TR 7.3.2.0 De-~ice TR 7.3.2.10
: e.
low Range Sarrpler At aU lim;;s 0
TR 7.3.2.1 FbwRate TR 7.3.2.0 Measurement Device; TR 7.3.2.10
: f.
MeiHigh Rang;;
(m) 0 TR 7.. 3.2.10 Sarrpler ReI'.' Rate ME-<1SlJrement Device;
: 4.
Main Condenser Off-Gas (e)
B TR 1.3.2.1 Treatment Systam Noble TR 7.3..2.3 Gas Activity Monitor '"
TR 7.3..2.6 (Dovmstream of AOG TR 7.3.2.10 T reafment Systam)
(a)
Speci/a instrumentatioo idE!1~licalion numbers <ire provided in AppEIldix E.
(b)
AlarJl1l~rip setpoinfs shall be determined in acoocdance \\";th ODCM melhodo:cgy and set 10 ensure the lim(s oi ODCMS 7.3.7. 'Dose Rate-Gaseous Effluents; are no! elQOeeaed.
7.3.2 Al.ARWiRiP S:=TPO!NT V,.LUE (0;1 (b)
NA NA NA (e)
NA (bJ (CQntinued)
(c}
AJarll1/tr,p setpoints shall be detemlined in oY..o::roance \\\\;th assccia:;ed desig\\!1specifoation{s) and set to anS!!re the limits of COCMS 7.:'..7. 'Dose Rate---Gaseous Effluents; are not exceec'ed.
(d)
Provides alaml.
(e)
During Main Coodenser Oft-Gas Treatment System operation (m)
During MdiHigll Rang;; System opeaIion Brunswick Units 1 and 2 Categories KIA:
RORating:
LPObj:
Cog Level:
S295038G 2.02.42 3.9 CLS-LP-30*08 HIGH 7.3.2-10 Rev. 32 1 Tier / Group: TIGl SRORating:  
 
===4.6 Source===
NEW Category 8:
Y
: 92. The following plant conditions exist on Unit Two due to a malfunction of the Air Dryer:
- SERVICE AIR PRESS-LOW is in alarm
- RB INSTR AIR RECEIVER 2A PRESS LOWis in alarm
- RB INSTR AIR RECEIVER 2B PRESS LOW is in alarm
- Instrument Air pressure is 93 psig and recovering Based on the above indications, which one of the following correctly identifies:
(1) the status of the Service Air Dryer Bypass Valve, SA-PV-5067, and (2) the procedure that contains the steps to close the Reactor Building Inboard and Outboard Isolation Valves (BFIVs)?
A (1) open (2) OAOP-20.O, Pneumatic (Air/Nitrogen) System Failures B. (1) open (2) 2APP-UA-O1, Service Air Press-Low C. (1) closed (2) OAOP-20.O, Pneumatic (Air/Nitrogen) System Failures D. (1) closed (2) 2APP-UA-O1, Service Air Press-Low
: 92. The following plant conditions exist on Unit Two due to a malfunction of the Air Dryer:
- SERVICE AIR PRESS-LOW is in alarm
- RB INSTR AIR RECEIVER 2A PRESS LOW is in alarm
- RB INSTR AIR RECEIVER 2B PRESS LOW is in alarm
- Instrument Air pressure is 93 psig and recovering Based on the above indications, which one of the following correctly identifies:
(1) the status of the Service Air Dryer Bypass Valve, SA-PV-5067, and (2) the procedure that contains the steps to close the Reactor Building Inboard and Outboard Isolation Valves (BFIVs)?
A'I (1) open (2) OAOP-20.0, Pneumatic (Air/Nitrogen) System Failures B. (1) open (2) 2APP-UA-01, Service Air Press-Low C. (1) closed (2) OAOP-20.0, Pneumatic (Air/Nitrogen) System Failures D. (1) closed (2) 2APP-UA-01, Service Air Press-Low  


Feedback Feedback K/A: 300000 KiA:    300000 A2.01A2.01 Ability to Ability      to (a)   predict the (a) predict        impacts of the impacts        the following of the  following on on the the INSTRUMENT INSTRUMENT AIR         SYSTEM and AIR SYSTEM      and (b) based (b)    based on  on those those predictions, predictions, use use procedures procedures to    to correct, correct, control, control, or or mitigate mitigate the the consequences consequences ofof those those abnormal abnormal operation:
Feedback K/A: 300000 A2.01 Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
operation:
Air dryer and filter malfunctions (CFR: 41.5/45.6)
Air dryer Air   dryer and and filter filter malfunctions malfunctions (CFR: 41.5 (CFR:      41.5/45.6)
RO/SRO Rating: 2.9/2.8 Objective:
                  /45.6)
CLS-LP-46, 07i: Given plant conditions, determine if the following automatic actions should occur: Air Dryer is bypassed.
RO/SRO Rating:
CLS-LP-037.1, 8b: State how the RBHVAC is affected by the following: Loss of Instrument Air.
RO/SRO        Rating: 2.9/2.8 2.9/2.8 Objective:
Objective:
CLS-LP-46, 07i:
CLS-LP-46,         07i: Given Given plant plant conditions, conditions, determine determine ifif the the following following automatic automatic actions actions should should occur:
occur: Air Air Dryer is Dryer      is bypassed.
bypassed.
CLS-LP-037.1, 8b:
CLS-LP-037.1,          8b: State State how how the the RBHVAC RBHVAC is  is affected affected by by the the following:
following: Loss Loss of of Instrument Instrument Air.
Air.


==Reference:==
==Reference:==
RB INSTR AIR RECEIVER 28 PRESS LOW (UA-01 1-2)
SERVICE AIR PRESS LOW (UA-01 5-4)
OAOP-20, Pneumatic (Air/Nitrogen) System Failures Cog Level: High Explanation:
The air dryer malfunction has caused air pressure to lower. The Service Air low pressure alarms comes in at 107 psig. At 105# decreasing the Service Air system isolates, thus the 0 psig indication on Service Air.
The alarms for the receivers low pressure come in at 95# and are located in the Reactor Building. With these alarms in the operators are required to close the BFIVs while there is still sufficient air pressure remaining to make the secondary containment isolation valves close in accordance with the AOP supplemental actions.
Distractor Analysis:
Choice A: Correct answer; The air dryer bypass valve opens at 98# and dropping and the steps are in the AOP for closing the BFIVs.
Choice B: Plausible because the air dryer bypass valve is open, but the guidance for closure of the BFIVs is contained in the AOP or RB INSTRAIR RECEIVER 2A(B) PRESS LOWAPP.
Choice C: Plausible because the AOP is the correct procedure for closure of the BFIVs, but the air dryer bypass valve would be open (requires system knowledge to know the setpoint for the bypass opening).
Choice D: Plausible because the student may not know the setpoint of the bypass valve opening and the guidance for closure of the BFIVs is contained in the AOP or RB INSTR AIR RECEIVER 2A(B) PRESS LOWAPP.
SRO Basis: 10 CFR 55.43(b)-5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
The first part of the question is RO knowledge (setpoint for the auto opening of the air dryer bypass valve the second part is Assessing plant conditions (normal, abnormal, or emergency) and then prescribing a procedure to mitigate, recover, or with which to proceed.
Notes Feedback KiA: 300000 A2.01 Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:
Air dryer and filter malfunctions (CFR: 41.5 /45.6)
RO/SRO Rating: 2.9/2.8 Objective:
CLS-LP-46, 07i: Given plant conditions, determine if the following automatic actions should occur: Air Dryer is bypassed.
CLS-LP-037.1, 8b: State how the RBHVAC is affected by the following: Loss of Instrument Air.


==Reference:==
==Reference:==
 
RB INSTR AIR RECEIVER 2B PRESS LOW (UA-01 1-2)
RB INSTR RB     INSTR AIR       RECEIVER 2B AIR RECEIVER        28 PRESS PRESS LOW LOW (UA-01 (UA-01 1-2)1-2)
SERVICE AIR PRESS LOW (UA-01 5-4)
SERVICE AIR PRESS SERVICE              PRESS LOW  LOW (UA-01 (UA-01 5-4) 5-4)
OAOP-20, Pneumatic (Air/Nitrogen) System Failures Cog Level: High Explanation:
OAOP-20, Pneumatic OAOP-20,        Pneumatic (Air/Nitrogen)
(Air/Nitrogen) System      Failures System Failures Cog Level:
Cog      Level: High High Explanation:
The air dryer malfunction has caused air pressure to lower. The Service Air low pressure alarms comes in at 107 psig. At 105# decreasing the Service Air system isolates, thus the 0 psig indication on Service Air.
The air dryer malfunction has caused air pressure to lower. The Service Air low pressure alarms comes in at 107 psig. At 105# decreasing the Service Air system isolates, thus the 0 psig indication on Service Air.
The alarms for the receivers low pressure come in at 95# and are located in the Reactor Building. With these alarms in the operators are required to close the BFIVs while there is still sufficient air pressure remaining to make the secondary containment isolation valves close in accordance with the AOP supplemental actions.
The alarms for the receivers low pressure come in at 95# and are located in the Reactor Building. With these alarms in the operators are required to close the BFIVs while there is still sufficient air pressure remaining to make the secondary containment isolation valves close in accordance with the AOP supplemental actions.
supplemental Distractor Analysis:
Distractor Analysis:
Choice A: Correct answer; The air dryer bypass valve opens at 98# and dropping and the steps are in the AOP for closing the BFIVs.
Choice A: Correct answer; The air dryer bypass valve opens at 98# and dropping and the steps are in the AOP for closing the BFIVs.
Choice B: Plausible because the air dryer bypass valve is open, but the guidance for closure of the BFIVs is contained in the AOP or RB INSTR       INS TRAIRAIR RECEIVER 2A(B) PRESS LOW           LOWAPP.
Choice B: Plausible because the air dryer bypass valve is open, but the guidance for closure of the BFIVs is contained in the AOP or RB INSTR AIR RECEIVER 2A(B) PRESS LOW APP.
APP.
Choice C: Plausible because the AOP is the correct procedure for closure of the BFIVs, but the air dryer bypass valve would be open (requires system knowledge to know the setpoint for the bypass opening).
Choice C: Plausible because the AOP is the correct procedure for closure of the BFIVs, but the air dryer bypass valve would be open (requires system knowledge to know the setpoint for the bypass opening).
Choice D: Plausible because  because the student may not know the setpoint of the bypass valve opening and the guidance guidance for closure of the BFIVs is contained in the AOP or RB INSTR AIR RECEIVER 2A(B) PRESS LOWAPP.
Choice D: Plausible because the student may not know the setpoint of the bypass valve opening and the guidance for closure of the BFIVs is contained in the AOP or RB INSTR AIR RECEIVER 2A(B) PRESS LOWAPP.
SRO SRO Basis:Basis: 10 10 CFR CFR 55.43(b)-5 55.43(b)-5 Assessment of    of facility conditions conditions andand selection selection of of appropriate procedures procedures during during normal, normal, abnormal, and   and emergency emergency situations.
SRO Basis: 10 CFR 55.43(b)-5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
situations.
The first part of the question is RO knowledge (setpoint for the auto opening of the air dryer bypass valve the second part is Assessing plant conditions (normal, abnormal, or emergency) and then prescribing a procedure to mitigate, recover, or with which to proceed.
The first The           part of first part   of the the question question isis RO RO knowledge knowledge (setpoint (setpoint forfor the the auto auto opening opening of of the the air air dryer dryer bypass bypass valve valve the    second part the second         part is is Assessing Assessing plant plant conditions conditions (normal, (normal, abnormal, abnormal, or  or emergency) emergency) andand then then prescribing prescribing aa procedure procedure to     to mitigate, mitigate, recover, recover, oror with with which which toto proceed.
Notes  
proceed.
Notes Notes


4.
4.
: 4.          IF RB IF        INSTR AIR RB INSTR        AiR RECEIVER RECEIVER 1A(2A)    IA(2A) PRESS PRESS LOWLOW (UA-O1 1-1)
IF RB INSTR AiR RECEIVER IA(2A) PRESS LOW (UA-O1 1-1) OR RB IWSTR AIR RECEiVER IB(2B,)
(UA-01      1-1) OR OR RB RB INSTR IWSTR AIR   AIR RECElVER RECEiVER 1B(2B) IB(2B,)
PRESS LOW (UA-Ol 1-2) alarm is received, THEN PERFORM the following:
PRESS LOW PRESS        LOW (UA-01 (UA-Ol 1-2)1-2) alarm alarm isis received, received, THEN PERFORM THEN      PERFORM the          following:
the following:
NOTE:
NOTE:
NOTE:      Isolation or Isolation      of the the Reactor Reactor Building Building supply supply andand exhaust exhaust dampers dampers will will render render the the building ventilation building       ventilation system system inoperable.
Isolation of the Reactor Building supply and exhaust dampers will render the building ventilation system inoperable. Consideration should be given for starting the Standby Gas Treatment System to ensure the Reactor Building differential pressure remains negative.
inoperable. Consideration Consideration should should be be given given ror for starting the starting      the Standby Standby Gas Gas Treatment Treatment System System to  to ensure ensure methe Reactor Reactor Building Building differential pressure differential     pressure remains remains negative.
negative.
a.
a.
: a.      IFIF necessary, necessary. THEN Treatment System.
IF necessary. THEN START the Standby Gas 0
Treatment THEN START System.
Treatment System.
START the  the Standby Standby Gas Gas            o0 NOTE:
NOTE:
NOTE:    Local "Tee Local       Tee Handles" Handles may  may be be used used to    close the to close   the Reactor Reactor Building Building Isolation Isolation Dampers ifif insufficient Dampers           insufficient control control airair is is available.
Local Tee Handles may be used to close the Reactor Building Isolation Dampers if insufficient control air is available. 1 (2)OP-37.1 provide L
available. 11 (2)OP-37.1 (2)OP-37.1 provide provide L           instructions for manual instructions            manual operation operation of    of Reactor Reactor Building Building Isolation Isolation Valves.
instructions for manual operation of Reactor Building Isolation Valves.
b.
b.
: b.        CLOSE the CLOSE            following dampers:
CLOSE the following dampers:
the following      dampers:
RB VENT 1NBO VALVES, 1A2A)-BFiV-RB and iCi2C)-BF1V-RB RB VENT OUTBD VALVES, 1B(2B)-BF1V-RB and ID(2D)-BF1 V-RB
                                  -        ,~B  VENT 1NBO VALVES, RB VENTJNBD              VALVES,                                  0 1A2A)-BFiV-RB and tC(2C)-BFIV-RB 1A(2A)-BFlV-RB                  iCi2C)-BF1 V-RB
) OAOP-20.O Rev. 35 Page 5 of 18 Unit 2 APP UA-O1 5-3 Page 1 of 2 AIR DRYER 2A TROUBLE AUTO ACTIONS 1.
                                          .~B RB VENTOUTBD VENT OUTBD VALVES,                                           0 1B(2B)-BF1V-RB and 1B(2B)-BFlV-RB            and 1D(2D)-BFIV-RB ID(2D)-BF1 V-RB
Service air dryer bypass valve SA-PV-5067 will begin to open if service air header pressure decreases to 98 psig.
) OAOP-20.O IOAOP-20.0                                              Rev. 35                                       Page 5 of 18 Page50f18      I Unit 2 APP UA-01 UA-O1 5-3 Page*lof2 Page 1 of 2 AIR DRYER .2A  2A TROUBLE AUTO ACTIONS 1.
        *1.      Service air dryer bypass valve SA-PV-5067 Service                                              SA-PV-5067 will begin to open ifif service air header pressure decreases to 98                      98 psig.
2.
2.
: 2.      IfIf control     power is control power        is lost lost oror interrupted interrupted the dryer    dryer will fail safe, providing providing continued continued air      air flow through through one  one tower.
If control power is lost or interrupted the dryer will fail safe, providing continued air flow through one tower.
3.
3.
: 3.       IfIf aa dryer dryer tower tower moisture moisture sensing probe      probe related fault or    or malfunction occurs, occurs, the the dryer dryer control        system will control system           will default default to to aa 44 hour hour drying drying cycle.
If a dryer tower moisture sensing probe related fault or malfunction occurs, the dryer control system will default to a 4 hour drying cycle.
cycle.
: 4.
IF RB INSTR AIR RECEIVER 1A(2A) PRESS LOW (UA-01 1-1) OR RB INSTR AIR RECElVER 1B(2B)
PRESS LOW (UA-01 1-2) alarm is received, THEN PERFORM the following:
NOTE:
Isolation or the Reactor Building supply and exhaust dampers will render the building ventilation system inoperable. Consideration should be given ror starting the Standby Gas Treatment System to ensure me Reactor Building differential pressure remains negative.
: a.
IF necessary, THEN START the Standby Gas Treatment System.
o NOTE:
Local "Tee Handles" may be used to close the Reactor Building Isolation Dampers if insufficient control air is available. 1 (2)OP-37.1 provide instructions for manual operation of Reactor Building Isolation Valves.
: b.
CLOSE the following dampers:
IOAOP-20.0 AIR DRYER.2A TROUBLE AUTO ACTIONS
,~B VENTJNBD VALVES, 0
1A(2A)-BFlV-RB and tC(2C)-BFIV-RB
.~B VENTOUTBD VALVES, 0
1B(2B)-BFlV-RB and 1D(2D)-BFIV-RB Rev. 35 Page50f18 I Unit 2 APP UA-01 5-3 Page*lof2
*1.
Service air dryer bypass valve SA-PV-5067 will begin to open if service air header pressure decreases to 98 psig.
: 2.
If control power is lost or interrupted the dryer will fail safe, providing continued air flow through one tower.
: 3.
If a dryer tower moisture sensing probe related fault or malfunction occurs, the dryer control system will default to a 4 hour drying cycle.  


RB INSTR RB  INSTR AIR   RECEIVER 2B AIR RECEIVER            PRESS LOW 2B PRESS     LOW AUTO ACTIONS AUTO    ACTIONS 1.
RB INSTR AIR RECEIVER 2B PRESS LOW AUTO ACTIONS 1.
        'I. Standby Instrument standby      Instrument Air Compressor Compressor 2B  2B starts starts and and loads.
Standby Instrument Air Compressor 2B starts and loads.
loads.
2.
2.
: 2. High   Pressure BoUIe High Pressure    Bottle Rack Rack Isolation Isolation Valve, RNA-SV-548'1 RNA-SV-5481 opens, supplying SRV's supplying    SRVs and CAC-V17 AC-V17 with aa pneumatic pneumatic source.
High Pressure Bottle Rack Isolation Valve, RNA-SV-5481 opens, supplying SRVs and AC-V17 with a pneumatic source.
source.
CAUSE 1.
CAUSE CAUSE 1.
Low air pressure (95 psig) in instrument air receiver 2B.
        'I. Low air pressure Low        pressure (95 psig) psig) in in instrument instrument air receiver receiver 2B.
2.
: 2. Loss of plant Loss      plant air compressors.
Loss of plant air compressors.
: 3. Instrument air pipe Instrument        pipe rupture or air leak.
3.
: 4. Circuit malfunction.
Instrument air pipe rupture or air leak.
4.
Circuit malfunction.
OBSERVATIONS 1.
OBSERVATIONS 1.
        '1. Standby compressor starts automatically and loads (it will unload at '105      105 psig).
Standby compressor starts automatically and loads (it will unload at 105 psig).
: 2. Service air header may have isolated.
2.
: 3. Pressure Indicator 2-RNA-PI-5268 2-RNA-Pl-5268 (XU-51) indicates approximately 100 psig.
Service air header may have isolated.
3.
Pressure Indicator 2-RNA-Pl-5268 (XU-51) indicates approximately 100 psig.
ACTIONS 1.
ACTIONS 1.
        '1.            that standby compressor is running.
Check that standby compressor is running.
Check tllat
2.
: 2. Check to see if instrument air pressure is maintaining or increasing above 95 psig.
Check to see if instrument air pressure is maintaining or increasing above 95 psig.
: 3. Check plant compressors.
3.
: 4. Check for instrument air ruptures.
Check plant compressors.
Cl1eck
4.
: 5. Isolate any instrument air piping leaks or ruptures.
Check for instrument air ruptures.
: 6. Isolate nonessential air supplies in order to maintain more than 95 psig on instrument header.
5.
heade[
Isolate any instrument air piping leaks or ruptures.
: 7. Ensure the High Pressure Bottle Rack Isolation Valve, RNA-SV-548'1RNA-SV-5481 (XU-51) opens.
6.
: 8. If aa circuit malfunction is suspected, ensure that a WRIJO WRfJO is prepared.
Isolate nonessential air supplies in order to maintain more than 95 psig on instrument heade[
: 9. If secondary containment isolation is required, close secondary containment isolation valves 2B-BFI        V-RB and 2D-BFI 2B-BFIV-RB          20-BFIV-RB V-RB prior to accumulator air pressure bleedoff.
7.
2APP-UA-0i
Ensure the High Pressure Bottle Rack Isolation Valve, RNA-SV-5481 (XU-51) opens.
!2APP-UA-O'l                                  Rev. 64                               Page 66 of 102 1021 Categories Categories K/A:
8.
KIA:        300000 A2.01 300000    A2.0 1                              Tier / Group:
If a circuit malfunction is suspected, ensure that a WRIJO is prepared.
Tier/Group:    T2G 1 T2G1 RO RO Rating:   2.9 2.9                                           SRO Rating:
9.
SRO            2.8 2.8 LP Obj:
If secondary containment isolation is required, close secondary containment isolation valves 2B-BFIV-RB and 2D-BFIV-RB prior to accumulator air pressure bleedoff.
Obj:    46-71 46-71                                         Source:
2APP-UA-0i Rev. 64 Page 6 of 102 Categories K/A:
Source:        ~vv NEW Cog Cog Level:
300000 A2.01 Tier/Group:
Level:  HIGH HIGH                                         Category Category 8:
T2G1 RO Rating:
8:    Y Y
2.9 SRO Rating:
: 93. Unit
2.8 LP Obj:
: 93.      Two isis operating Unit Two      operating at  at power power withwith Reactor Reactor Recirculation Recirculation Loop Loop AA isolated isolated due due to to abnormal seal abnormal    seal leakage.
46-71 Source:
leakage. AA fire fire inin the  reactor building the reactor  building occurs occurs and and the the Site Site Incident Incident Commander has Commander        has requested requested thatthat MCCMCC 2XA-2 2XA-2 bebe de-energized de-energized forfor fire fire suppression.
NEW Cog Level:
suppression.
HIGH Category 8:
Which one Which    one ofof the  following identifies the following    identifies thethe impact impact that that deenergizing deenergizing MCC  MCC 2XA-2 2XA-2 hashas on on RHR    Loop  A  availability    and  the    procedure RHR Loop A availability and the procedure which          which provides provides this this guidance guidance underunder the the above plant above          conditions?
Y RB INSTR AIR RECEIVER 2B PRESS LOW AUTO ACTIONS
plant conditions?
'I.
Deenergizing MCC Deenergizing      MCC 2XA-2 2XA-2 will will make make RHR  RHR Loop Loop A A Inoperable Inoperable but but Available Available provided provided that the that  the 2-E11-F015A, 2-El l-FO15A, InboardInboard Injection Injection Vlv, Vlv,    (1)
standby Instrument Air Compressor 2B starts and loads.
(1) to   support LPCIIAW to support  LPCI lAW (2)    (2)
: 2.
A. (1)
High Pressure BoUIe Rack Isolation Valve, RNA-SV-548'1 opens, supplying SRV's and CAC-V17 with a pneumatic source.
A.      is maintained (1) is  maintained (de-energized)
CAUSE
(de-energized) openedopened (2)  OAP-025, BNP (2) OAP-025,      BNP Integrated Integrated Scheduling Scheduling B. (1)
'I.
B.       is maintained (I) is  maintained (de-energized)
Low air pressure (95 psig) in instrument air receiver 2B.
(de-energized) opened opened (2) 001-01.08, Control of Equipment and System Status (2)
: 2.
C C~  (1) has a Dedicated Operator is assigned for manual operation (2) OAP-025, BNP Integrated Scheduling D. (1) has a Dedicated Operator is assigned for manual operation (2) 001-01.08, Control of Equipment and System Status
Loss of plant air compressors.
: 3.
Instrument air pipe rupture or air leak.
: 4.
Circuit malfunction.
OBSERVATIONS
'1.
Standby compressor starts automatically and loads (it will unload at '105 psig).
: 2.
Service air header may have isolated.
: 3.
Pressure Indicator 2-RNA-PI-5268 (XU-51) indicates approximately 100 psig.
ACTIONS
'1.
Check tllat standby compressor is running.
: 2.
Check to see if instrument air pressure is maintaining or increasing above 95 psig.
: 3.
Check plant compressors.
: 4.
Cl1eck for instrument air ruptures.
: 5.
Isolate any instrument air piping leaks or ruptures.
: 6.
Isolate nonessential air supplies in order to maintain more than 95 psig on instrument header.
: 7.
Ensure the High Pressure Bottle Rack Isolation Valve, RNA-SV-548'1 (XU-51) opens.
: 8.
If a circuit malfunction is suspected, ensure that a WRfJO is prepared.
: 9.
If secondary containment isolation is required, close secondary containment isolation valves 2B-BFIV-RB and 20-BFIV-RB prior to accumulator air pressure bleedoff.  
!2APP-UA-O'l Categories KIA:
300000 A2.0 1 RO Rating:
2.9 LP Obj:
46-71 Cog Level:
HIGH Rev. 64 Tier / Group: T2G 1 SRO Rating:


Feedback Feedback K/A: S600000G KIA:  S600000G 2.02.37 2.02.37 Ability to Ability       determine operability to determine     operability and/or andlor availability availability of of safety safety related related equipment.
===2.8 Source===
equipment.
~vv Category 8:
Plant Fire Plant  Fire OnOn Site Site (CFR: 41.7/43.5/45.12)
Y Page 6 of 1021
(CFR:  41.7 /43.5 / 45.12)
: 93. Unit Two is operating at power with Reactor Recirculation Loop A isolated due to abnormal seal leakage. A fire in the reactor building occurs and the Site Incident Commander has requested that MCC 2XA-2 be de-energized for fire suppression.
ROISRO Rating:
Which one of the following identifies the impact that deenergizing MCC 2XA-2 has on RHR Loop A availability and the procedure which provides this guidance under the above plant conditions?
RO/SRO      Rating: 3.6/4.6 3.6/4.6 Objective: CLS-LP-Objective:     CLS-LP
Deenergizing MCC 2XA-2 will make RHR Loop A Inoperable but Available provided that the 2-El l-FO15A, Inboard Injection Vlv, (1) to support LPCI lAW (2)
A.
(1) is maintained (de-energized) opened (2) OAP-025, BNP Integrated Scheduling B. (I) is maintained (de-energized) opened (2) 001-01.08, Control of Equipment and System Status C (1) has a Dedicated Operator is assigned for manual operation (2) OAP-025, BNP Integrated Scheduling D. (1) has a Dedicated Operator is assigned for manual operation (2) 001-01.08, Control of Equipment and System Status
: 93. Unit Two is operating at power with Reactor Recirculation Loop A isolated due to abnormal seal leakage. A fire in the reactor building occurs and the Site Incident Commander has requested that MCC 2XA-2 be de-energized for fire suppression.
Which one of the following identifies the impact that deenergizing MCC 2XA-2 has on RHR Loop A availability and the procedure which provides this guidance under the above plant conditions?
Deenergizing MCC 2XA-2 will make RHR Loop A Inoperable but Available provided that the 2-E11-F015A, Inboard Injection Vlv, (1) to support LPCIIAW (2)
A. (1) is maintained (de-energized) opened (2) OAP-025, BNP Integrated Scheduling B. (1) is maintained (de-energized) opened (2) 001-01.08, Control of Equipment and System Status C~ (1) has a Dedicated Operator is assigned for manual operation (2) OAP-025, BNP Integrated Scheduling D. (1) has a Dedicated Operator is assigned for manual operation (2) 001-01.08, Control of Equipment and System Status
 
Feedback K/A: S600000G 2.02.37 Ability to determine operability andlor availability of safety related equipment.
Plant Fire On Site (CFR: 41.7 /43.5 / 45.12)
ROISRO Rating: 3.6/4.6 Objective: CLS-LP


==Reference:==
==Reference:==
OAP-025, Revision 39, Page 9, Section 3.1 Cog Level: High Explanation:
Requires knowledge of equipment powered from MCC 2XA-2 (opposite Unit power E5). With the RR Loop A isolated (RR Discharge and Disch Bypass valves will be close
- required for LPCI) and FO15A (located in ECCS Pipe Tunnel - RB 20) can be manually opened by a dedicated operator. 001-01.08 has recently been revised to support implementation of OPS-NGGC-1 000, Fleet Conduct of Operations. Risk assessment and equipment removal from service guidance has been removed from 01-01.08.
Distractor Analysis:
Choice A: Plausible because if the valve is de-energized in its intended state (such as a PCIV in the closed dierection) this could be considered correct although for this case the open direction for this normally closed valve would have to have interlocks defeated to have both the F015 and F017 both open (potential to pressurize low pressure piping). OAP-025 is correct.
Choice B: Plausible because if the valve is de-energized in its intended state (such as a PCIV in the closed dierection) this could be considered correct although for this case the open direction for this normally closed valve would have to have interlocks defeated to have both the F015 and F017 both open (potential to pressurize low pressure piping). 001-01.08 no longer provides guidance for evaluating MR/PSA system availability.
Choice C: Correct Answer Choice D: Plausible because RHR Loop is available and 001-01.08 no longer provides guidance for evaluating MRIPSA system availability.
SRO Only Basis: Knowledge of administrative procedures that specify implementation, and/or coordination of plant normal procedures.
Notes Feedback KIA: S600000G 2.02.37 Ability to determine operability and/or availability of safety related equipment.
Plant Fire On Site (CFR: 41.7/43.5/45.12)
RO/SRO Rating: 3.6/4.6 Objective: CLS-LP-


==Reference:==
==Reference:==
 
OAP-025, Revision 39, Page 9, Section 3.1 Cog Level: High Explanation:
OAP-025, Revision OAP-025,     Revision 39, 39, Page Page 9, 9, Section Section 3.1 3.1 Cog  Level: High Cog Level:     High Explanation:
Requires knowledge of equipment powered from MCC 2XA-2 (opposite Unit power E5). With the RR Loop A isolated (RR Discharge and Disch Bypass valves will be close - required for LPCI) and F015A (located in ECCS Pipe Tunnel - RB 20') can be manually opened by a dedicated operator. 001-01.08 has recently been revised to support implementation of OPS-NGGC-1000, Fleet Conduct of Operations. Risk assessment and equipment removal from service guidance has been removed from 01-01.08.
Explanation:
Requires knowledge Requires     knowledge of of equipment equipment powered powered from from MCC MCC 2XA-2 2XA-2 (opposite (opposite UnitUnit power power E5).
E5). With With the the RR RR Loop A isolated Loop      isolated (RR (RR Discharge Discharge and and Disch Disch Bypass Bypass valves valves will be be close close - required
                                                                                  -  required for for LPCI)
LPCI) and and F015A FO15A (located in (located  in ECCS ECCS Pipe Pipe Tunnel Tunnel - RB
                                    - RB 20')
: 20) can can be be manually manually opened opened by     dedicated operator.
by aa dedicated    operator. 001-01.08 001-01.08 has has recently been recently          revised to support been revised      support implementation implementation ofof OPS-NGGC-1000, OPS-NGGC-1 000, Fleet  Fleet Conduct of of Operations.
Operations. Risk Risk assessment and assessment     and equipment equipment removal removal from from service service guidance guidance has has been been removed removed from from 01-01.08.
01-01.08.
Distractor Analysis:
Distractor Analysis:
Choice A: Plausible because if the valve is de-energized in its intended state (such as a PC               IV in the PCIV closed dierection) this could be considered correct although for this case the open direction for this normally closed valve would have to have interlocks defeated to have both the F015 and F017 both open (potential to pressurize low pressure piping). OAP-025 is correct.
Choice A: Plausible because if the valve is de-energized in its intended state (such as a PC IV in the closed dierection) this could be considered correct although for this case the open direction for this normally closed valve would have to have interlocks defeated to have both the F015 and F017 both open (potential to pressurize low pressure piping). OAP-025 is correct.
Choice B: Plausible because if the valve is de-energized in its intended state (such as a PCIV in the closed dierection) this could be considered correct although for this case the open direction for this normally closed valve would have to have interlocks defeated to have both the F015 and F017 both open (potential to pressurize low pressure piping). 001-01.08 001-01 .08 no longer provides guidance for evaluating MRIPSA MR/PSA system availability.
Choice B: Plausible because if the valve is de-energized in its intended state (such as a PCIV in the closed dierection) this could be considered correct although for this case the open direction for this normally closed valve would have to have interlocks defeated to have both the F015 and F017 both open (potential to pressurize low pressure piping). 001-01.08 no longer provides guidance for evaluating MRIPSA system availability.
Choice C: Correct Answer Choice D: Plausible because RHR Loop is available and 001-01.08 no longer provides guidance for evaluating MRIPSA system availability.
Choice C: Correct Answer Choice D: Plausible because RHR Loop is available and 001-01.08 no longer provides guidance for evaluating MRIPSA system availability.
SRO Only Basis: Knowledge of administrative procedures that specify implementation, implementation, and/or coordination of plant normal procedures.
SRO Only Basis: Knowledge of administrative procedures that specify implementation, and/or coordination of plant normal procedures.
Notes
Notes  


3.0 3.0      DEFINITIONS DEFINITIONS 3.1 3.1     Available (Availability)
3.0 DEFINITIONS 3.1 Available (Availability)
Available       (Availability)
The status of a system, structure or component (SSC) that is OPERABLE, in service or can be placed in a FUNCTIONAL state within a reasonably short period of time consistent with its intended need. The SSC must be capable of meeting all of its most limiting requirements for the plant mode under consideration Using a manual means for placing an SSC in service requires a dedicated operator assigned to be cognizant of the SSC along with a written procedure for its restoration. A dedicated operator for the purpose of this definition is one who is specifically assigned the task and 2vailable, as necessary, to perform the required actions.
The status The    status of       system, structure of aa system,    structure or  or component component (SSC)  (SSC) that that isis OPERABLE, OPERABLE, inin service or service   or can can be be placed placed inin aa FUNCTIONAL FUNCTIONAL state      state within within aa reasonably reasonably short short period of period    of time time consistent consistent with      its intended with its   intended need.
3.2 Backbone Schedule A preliminary schedule consisting of work items that are either required to he performed or have been designated by management as high priority items.
need. The  The SSC SSC must must be be capable capable of meeting of  meeting all       of its all of  its most most limiting limiting requirements requirements for    for the the plant plant mode mode under under consideration Using consideration.         Using aa manual manual means means for  for placing placing an an SSC SSC in in service service requires aa dedicated requires        dedicated operator operator assigned assigned to   to bebe cognizant cognizant of of the the SSC SSC along along with aa written with                procedure tor written procedure          for its   restoration. A its restoration.        A "dedicated" dedicated operator operator for for the the purpose of purpose    of this     definition is this definition    is one one whowho is is specifically specifically assigned assigned the the task task and and 2vailable, as available,         necessary, to as necessary,       to perform perform the  the required required actions.
The following items would nomally comprise the backbone schedule:
actions.
Implementing Supervisor recommendations KeyI(a)(1 Equipment priority action items Required SurveillancesiPMs System Outages Committed Items Priority 1 & 2 CAPRs, CORRs. and regulatory committed items Modification ECs determined a priority by Engineering representative or Scheduler (must be ready to work with work orders in ready or approved status)
3.2 3.1       Backbone Schedule Backbone        Schedule AA preliminary preliminary schedule schedule consisting consisting of   of work work items items that that are are either either required required to to be he performed or performed            have been or have      been deSignated designated by    by management management as      as high high priority priority items.
Engineering recommendations Reactivity Management flagged Work Orders 13 Compensatory Actions Measures that are used to niitigate the impact and minimize the duration of an ELEVATED RISK activity. These measures may include CONTINGENCY PLANS or procedural controls.
items.
3.4 Contingency Planning A look ahead process whereby potential problems are systematically identified, assessed, and addressed by adding plans or mitigating actions.
The following items  items would nom1ally nomally comprise the backbone schedule:
The necessity for a contingency plan is based on the potential consequences as well as the probability of a problem occurring.
            **    Implementing Supervisor recommendations Implementing
DAP-025 Rev 39 I
            **    KeyI(a)(1 Equipment priority Key/(a)(1)                       priority action items
Page 9 of 121 3.0 DEFINITIONS 3.1 Available (Availability)
* SurveillancesiPMs Required Surveillances/PMs
The status of a system, structure or component (SSC) that is OPERABLE, in service or can be placed in a FUNCTIONAL state within a reasonably short period of time consistent with its intended need. The SSC must be capable of meeting all of its most limiting requirements for the plant mode under consideration. Using a manual means for placing an SSC in service requires a dedicated operator assigned to be cognizant of the SSC along with a written procedure tor its restoration. A "dedicated" operator for the purpose of this definition is one who is specifically assigned the task and available, as necessary, to perform the required actions.
            **    System Outages
3.1 Backbone Schedule A preliminary schedule consisting of work items that are either required to be performed or have been deSignated by management as high priority items.
            **    Committed Items - Priority 11 & 2 C.4.PR's, CAPRs, CORR's, CORRs. and regulatory committed items
The following items would nom1ally comprise the backbone schedule:
* Modification EC's   ECs determined a priority by Engineering repre.sentative   representative or Scheduler (must be ready to work with work orders in ready or approved status)
Implementing Supervisor recommendations Key/(a)(1) Equipment priority action items Required Surveillances/PMs System Outages Committed Items - Priority 1 & 2 C.4.PR's, CORR's, and regulatory committed items Modification EC's determined a priority by Engineering repre.sentative or Scheduler (must be ready to work with work orders in ready or approved status)
* Engineering recommendations
Engineering recommendations Reactivity Management flagged Work Orders 3.3 Compensatory Actions Measures that are used to mitigate the impact and minimize the duration of an ELEVATED RISK activity. These measures may include CONTINGENCY PLANS or procedural controls.
* Reactivity Management flagged Work Orders 13 3.3     Compensatory Actions Measures that are used to mitigate     niitigate the impact and minimize the duration of an ELEVATED RISK activity. These measures may include CONTINGENCY PLANS or procedural controls.
3.4 Contingency Planning IOAP-025 A look ahead process whereby potential problems are systematically identified, assessed, and addressed by adding plans or mitigating actions.
3.4     Contingency Contingency Planning A look ahead process whereby potential problems are systematically identified, assessed, assessed, and addressed by adding plans or mitigating              mitigating actions.
The necessity for a contingency plan is based on the potential consequences as well as the probabllity of a problem occurring.
actions.
Rev. 39 Page 9 of 121 I  
The necessity necessity for aa contingency contingency plan plan is is based based on the   the potential potential consequences as       as well well as the probability as the    probabllity of   of aa problem occurring.
DAP-025 IOAP-025                                                Rev Rev. 39 39                I                    Page 99 of Page       of 121 121  I


ATTACHMENT 33 ATTACHMENT Page 15 Page      15 of  19 of 19 480V Substation 4BOV    Substation E5/MCC/Panei E5?MCCIPaneI Load     Load Summary Summary 480V Motor Control Load: 480V Motor Control Center Load:                                      Center2-2XA-222XA-2 Locadc9: Unit Location:  Ur t   ReaotorBuilding 2 Reactor    3T. ldin 20'  NE ZY NE Draw ng  
ATTACHMENT 3 Page 15 of 19 480V Substation E5?MCCIPaneI Load Summary Load: 480V Motor Control Center 22XA-2 Locadc9: Ur t Reaotor 3T. ldin ZY NE Draw ng  


==Reference:==
==Reference:==
-2 3c-9 Upstream Power Source: dBOV Substation E5 COMPT LOAD DESCRIPTION EFFECTS ON LOSS OF POWER DF5 F?lR Cuibaard lniection Valve Loss of load 2-El1-F3l7ATS 3.5.1, 36.1.3, 3.5.2, 3.3.2.1:
D3 RHR Inboard lnection Valve 2-El -FO15A Loss of load ITS 3.51, 3.8.1.3, 3.5.2,3.3.3.1)
DGO HR Tor,s Spray Valve 2-El i-F028A Loss of load tTS3.8.1,3.8.1.3.3.&2.3. 3.3.3.1) 0D7 Rx ecrculation Pump 2A Discharge Valve Losa of load 2-B22-FO31ATS 3.4.1, 3.5.1) 008 Rx Recrculation Pump 2A Discharge Loss of load Sypass Valve 2-232-F032A (TS 3.4.1 3.5.11 001-50.1 Rev. 42 Page 24 of 55 ATTACHMENT 3 Page 15 of 19 4BOV Substation E5/MCC/Panei Load Summary Load: 480V Motor Control Center 2-2XA-2 Location: Unit 2 Reactor Building 20' NE Drawing


Drawing 
==Reference:==
F-03D49 Upstream Power Source: 4S0V SUbstation E5 COMPT LOAD DESCRIPTION EFfECTS ON LOSS OF POWER DFS RHR Outboard Injection Valve Less of load 2-E11-F017A(TS 3.5.1. 3.6.1.3, 3.5.2, 3.3.3.1}
DF3 RHR Inboard Injecticn Val'.e 2-E11-FOt5A Loss of load
{IS 3.5.1. 3.8.1.3, 3.5.2, 3.3.3.1}
DGO R HR Torus Spray Valve 2-E 11-1"028.11, Loss of load (IS 3.5.1. 3.8.1.3, 3.6.2.3, 3.3.3. t) 007 Rx Recirculation Pump 2A Discharge Valve Loss of load 2-B;?2-F031A iTS 3.4.1. 3.S.1}
008 Rx Recirculation Pump 2A Discharge Loss of load Bypass Valve 2-B32-F032A (is 3.4.1, 3.5.1)
\\ 001-50.1 Rev. 42 Page 24 of 55\\


==Reference:==
ATTACHMENT 2A Page 2 of 3 Residual Hear Removal System Loop A Panel Lineup Number Description Positicn Checked Venfied Indication Loop A Control Room Panel H12 P801 El lFcOeC Pump C Shuown Cooling CLOSED Suction V El l-FCOeA Pump A Shutdovn Cooling CLOSED Suction V El l-V32 Check Vae Bypass VIv CLOSED El l-FO17A Ouoard lrec:ion Vlv OPEN El l-F0le4 Drywell Spray Otbd Isol Vlv CLOSED Ell-F1O4A HX 2A Inboard Vent Vlv CLOSED Ell-FOI5A Inboard InjeciionVlv CLOSED El l-FO2IA Orywell Spray Inbd led Vlv LOSEO Ell-FIO3A HX 2A Outboard Vent /1v CLOSES El 1-F024A Torus Cooling Isol Vhi CLOSED Ell-FC4A HX24Bypass1v OPEN El l-F027A Torus Spray led Vi CLOSES El 1-FOl IA HX 2 Drain To Torus Vl CLOSES El 1-FcO4C Pump C Torus Suction Viv OPEN Eil-F028A Torus Discharge IsolVlv CLOSED El I-F026A
F-03D49
-IX 2A Drain To RCIC VIv CLOSES El I-FcO4A Pump A Torus Sucbon VIv OPEN E1l-FcO2A HX 2A Cutlev OPEN El I-FC7A Mn Flow Sypass V CLOSED El 1-FO2GA Pump A&C Torus Suction Vlv CPEN El i-Fc47A HX Z InletV OPEN El l-FE8OA Manual lnection ilv OPEN El l-PDV-Ffl38A HX Z SW Disch VIv CLOSED CS-517A Containment Spray Valve Control OFF Think Swilch 20P-17 Rev. 155 Page 244 of 297 Categories K/A:
                          -2 3c-9 Upstream Power Upstream  PowerSource:
S600000G2.02.37 Tier/Group:
Source: 4S0V dBOV SUbstation Substation E5 E5 COMPT COMPT        LOAD DESCRIPTION LOAD      DESCRIPTION                                    EFFECTS ON EFfECTS      ON LOSS LOSS OF OF POWER POWER DF5 DFS        F?lR Outboard RHR    Cuibaard Injection    Valve lniection Valve                      Loss of Less  ofload load 2-El1-F3l7ATS 3.5.1.
T1G1 RO Rating:
2-E11-F017A(TS        3.5.1, 3.6.1.3, 36.1.3, 3.5.2, 3.5.2, 3.3.2.1:
3.6 SRO Rating:
3.3.3.1}
4.6 LP Obj:
D3 DF3      RHR    Inboard Injecticn RHR Inboard      lnection Val'.e Valve 2-E11-FOt5A 2-El -FO15A          Loss of Loss  of load load
Source:
{IS  3.51, 3.8.1.3, ITS 3.5.1. 3.8.1.3, 3.5.2, 3.5.2,3.3.3.1) 3.3.3.1}
NEW Cog Level:
DGO DGO      RHR    Tor,s Spray HR Torus   Spray Valve Valve 2-E    i-F028A 2-El11-1"028.11,          Loss of Loss  of load load tTS3.8.1,3.8.
HIGH Category 8:
(IS              1.3.3.&2.3.
Numb:r El1-FOOeC Ell-FOOM Ell-V32 Ell-F017.4 El1-F01M Ell-Fl04A Ell-FOl5A El1-F021A El1-Fl03A El1-F024A Ell-F048A Ell-F027A Ell-FOllA Ell-FOO4C Ell-F028A E11-F02M El1-FOO4A E1l-FOO3.4 E11-FOO7A Ell-F020A E11-F047A E11-FOOOA E11-PDV-F068A CS-S17A ATTACHMENT 2.4-Page 2 of 3 Residual Heat Removal System Loop A Panel Lineup Description Positionl Checked Indication Loop A Control Room - Panel H12-P801 Pump C Shulx:lc'IIn Cooling CLOSED SuctionV'N Pump A Shutdown Cooling CLOSED SuclionV'N Check Va'Ne 8}'Jlass VIII CLOSED Outboard Injection VIII OPEN Drywell Spray Otbd Isol VIII CLOSED
3.5.1. 3.8.1.3,            3.3.3.1) 3.6.2.3, 3.3.3. t) 0D7 007        Rx Recirculation Rx  ecrculation Pump Pump 2A2A Discharge Discharge Valve Valve    Losa of Loss  of load load 2-B22-FO31ATS 2-B;?2-F031A          3.4.1, 3.S.1}
!-IX 2A Inboard Vent V'N CLOSED Inboard Injection V'N CLOSED Drywell Spray lnbd (501 Iflv CLOSED
iTS 3.4.1. 3.5.1) 008 008      Rx  Recrculation Pump Rx Recirculation    Pump 2A   Discharge 2A Discharge               Loss of Loss  of load load Sypass Valve Bypass    Valve 2-B32-F032A 2-232-F032A (is      3.4.1 (TS 3.4.1, 3.5.11 3.5.1)
!-IX 2.4 Outboard Vent Vlv CLOSED Torus Cooling 1501 VIII CLOSED
\ 001-50.1 001-50.1                                              Rev.
!-IX 2A Bypass 'v1v OPEN Torus Spray lsol V'N CLOSED
Rev. 42 42                                Page Page 24 24 of 55\
!-IX 2A Drain To Torus Vlv CLOSED Pump C Torus Suction V'N OPEN Torus Discharge lsol V'N CLOSED
of 55
!-IX 2A Drain To RCIC VI" CLOSED Pump A Torus Suction VIII OPEN
!-IX 2A Outlet Vlv OPEN Min Flow Bypass IJI~'
CLOSED Pump A&C Torus Suction Ifl...
OPEN HX 2A Inlet V'N OPEN M.anuallnieciion Vlv OPEN
!-IX 24 SW Disch VI" CLOSED Containment Spray Va'Ne Control OFF Think Switch Verified 1 20P-17 Rev. 155 Page 244 of 2971 Categories KIA:
RORating:
LP Obj:
Cog Level:
S600000G 2.02.37 3.6 HIGH Tier / Group: T1 G 1 SRO Rating:


ATTACHMENT2.4-ATTACHMENT                2A Page 22 of Page          of 33 Residual Heat Residual    Hear Removal Removal System    System Loop  Loop AA Panel Panel Lineup Lineup Number Numb:r                      Description Description                          Positicn Positionl      Checked Checked    Verified Venfied Indication Indication Loop AA Control Loop      Control Room          Panel H12-P801 Room - Panel    H12 P801 El lFcOeC El1-FOOeC              Pump CC Shulx:lc'IIn Pump      Shuown Cooling Cooling                CLOSED CLOSED Suction V SuctionV'N El l-FCOeA Ell-FOOM              Pump AA Shutdown Pump      Shutdovn Cooling Cooling                CLOSED CLOSED Suction V SuclionV'N El l-V32 Ell-V32                Check Va'Ne Check    Vae 8}'Jlass Bypass VIII VIv                  CLOSED CLOSED El l-FO17A Ell-F017.4            Ouoard Injection Outboard    lrec:ion VIII Vlv                        OPEN OPEN El l-F0le4 El1-F01M              Drywell Spray Drywell  Spray Otbd Otbd IsolIsol VIII Vlv              CLOSED CLOSED Ell-F1O4A Ell-Fl04A              HX 2A
===4.6 Source===
                          !-IX    Inboard Vent 2A Inboard    Vent V'NVlv                  CLOSED CLOSED Ell-FOI5A Ell-FOl5A            Inboard  InjeciionVlv Inboard Injection  V'N                        CLOSED CLOSED El l-FO2IA El1-F021A            Drywell  Spray lnbd Orywell Spray  Inbd (501 led Iflv Vlv              LOSEO CLOSED Ell-FIO3A El1-Fl03A              HX 2.4
NEW Category 8:
                          !-IX 2A Outboard Outboard Vent Vlv  /1v                CLOSES CLOSED El 1-F024A El1-F024A                    Cooling 1501 Torus Cooling    Isol VIII Vhi                      CLOSED CLOSED Ell-FC4A Ell-F048A              HX24Bypass1v
: 94. What action is required to be taken if Alternate Safe Shutdown (ASSD) Staffing drops below minimum complement due to an emergent on-shift AC illness and what procedure provides the guidance for this action?
                          !-IX 2A Bypass 'v1v                              OPEN El l-F027A Ell-F027A                            led V'N Torus Spray lsol    Vi                          CLOSES CLOSED El 1-FOl IA Ell-FOllA              HX 2A
The guidance for establishing an (1) if ASSD staffing composition is less than the minimum required is provided by (2)
                          !-IX 2 Drain To Torus Vl    Vlv                CLOSES CLOSED El 1-FcO4C Ell-FOO4C            Pump C Torus Suction V'N      Viv                OPEN Eil-F028A Ell-F028A            Torus Discharge lsol IsolVlv V'N                  CLOSED El I-F026A E11-F02M              -IX 2A Drain To RCIC VI"
A (1) ASSD Impairment (2) OASSD-00, User Guide B.
                          !-IX                        VIv                CLOSES CLOSED El I-FcO4A El1-FOO4A            Pump A Torus Suction Sucbon VIII VIv                OPEN E1l-FcO2A E1l-FOO3.4            HX 2A Outlet
(1) Active LCO for T.S. 5.2.2, Facility Staff Organization, (2) OASSD-00, User Guide C. (1) ASSD Impairment (2) 001-01.01, BNP Conduct of Operations Supplement D. (1) Active LCO forT.S. 5.2.2, Facility Staff Organization, (2) 001-01.01, BNP Conduct of Operations Supplement
                          !-IX    Cutlev  Vlv                              OPEN El I-FC7A E11-FOO7A            Mn Flow Bypass Min        Sypass V  IJI~'                      CLOSED El 1-FO2GA Ell-F020A            Pump A&C Torus Suction Vlv        Ifl...         CPEN OPEN El i-Fc47A E11-F047A            HX Z 2A InletV Inlet V'N                                OPEN El l-FE8OA E11-FOOOA            Manual  lnection ilv M.anuallnieciion    Vlv                          OPEN OPEN El l-PDV-Ffl38A E11-PDV-F068A        HX  24 SW Disch
: 94. What action is required to be taken if Alternate Safe Shutdown (ASSD) Staffing drops below minimum complement due to an emergent on-shift AO illness and what procedure provides the guidance for this action?
                          !-IX Z      Disch VIv VI"                        CLOSED CS-517A CS-S17A              Containment Containment Spray Spray Valve Va'Ne Control Control          OFF OFF Think Think Swilch Switch 20P-17 1 20P-17                                            Rev.
The guidance for establishing an (1) if ASSD staffing composition is less than the minimum required is provided by (2)
Rev. 155  155                            Page Page 244 244 of of 297 2971 Categories Categories K/A:
A'I (1) ASSD Impairment (2) OASSD-OO, User Guide B. (1) Active LCO for T.S. 5.2.2, Facility Staff Organization, (2) OASSD-OO, User Guide C. (1) ASSD Impairment (2) 001-01.01, BNP Conduct of Operations Supplement D. (1) Active LCO for T.S. 5.2.2, Facility Staff Organization, (2) 001-01.01, BNP Conduct of Operations Supplement  
KIA:              S600000G2.02.37 S600000G 2.02.37                                            Tier/Group:
Tier / Group: T1G1 T1 G 1 RO  Rating:
RORating:        3.6 3.6                                                        SRO Rating:
SRO    Rating: 4.6 4.6 LP LP Obj:
Obj:                                                                      Source:
Source:          NEW NEW Cog Cog Level:
Level:      HIGH HIGH                                                      Category Category 8:8:
: 94. What
: 94. What action action isis required required to to be   taken ifif Alternate be taken      Alternate Safe Safe Shutdown Shutdown (ASSD)
(ASSD) Staffing Staffing drops drops below minimum below     minimum complement complement due due to   an emergent to an    emergent on-shift on-shift AO AC illness illness and and what what procedure provides procedure     provides the the guidance guidance for for this this action?
action?
The guidance The  guidance for  for establishing establishing an an    (1)
(1)     ifif ASSD ASSD staffing staffing composition composition isis less less than than the the minimum required minimum      required isis provided provided by by    (2)
(2)
(1) ASSD A (1)
A'I      ASSD Impairment Impairment (2) OASSD-OO, (2)  OASSD-00, User  User Guide Guide B. (1)
B.   (1) Active Active LCOLCO forfor T.S.
T.S. 5.2.2, 5.2.2, Facility Facility Staff Staff Organization, Organization, (2)  OASSD-00,        User (2) OASSD-OO, User Guide    Guide C. (1)
C.        ASSD Impairment (1) ASSD       Impairment (2)  001-01.01, (2) 001-01.01, BNP    BNP Conduct Conduct of of Operations Operations Supplement Supplement D. (1) Active LCO for     forT.S. 5.2.2, Facility Staff T.S. 5.2.2,              Staff Organization, Organization, (2)  001-01.01, (2) 001-01 .01 , BNP BNP  Conduct     of Operations      Supplement Operations Supplement


Feedback Feedback K/A: SG2.01.05 KIA:    SG2.O1 .05 Conduct of Conduct          Operations of Operations Ability to Ability   to use   procedures related use procedures    related to   shift staffing, to shift  staffing, such such asas minimum minimum crew crew complement, complement, overtime overtime limitations, etc.
Feedback K/A: SG2.O1.05 Conduct of Operations Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.
limitations,    etc.
(CFR: 41.10 / 43.5/45.12)
(CFR: 41.10/43.5/45.12)
ROISRO Rating: 2.9/3.9 Objective: CLSLP304M*1 3m
(CFR:   41.10 / 43.5/45.12)
: 13. Given ASSD procedures and plant conditions that require use of ASSD procedures, determine the following:
ROISRO Rating:
: m. The manpower required to support the ASSD actions.
RO/SRO      Rating: 2.9/3.9 2.9/3.9 CLSLP304M*1 3m Objective: CLS-LP-304-M*13m Objective:
: 13. Given
: 13. Given ASSD ASSD procedures procedures and and plant plant conditions conditions that that require require use use of of ASSD ASSD procedures, procedures, determine determine the the following:
following:
: m. The
: m. The manpower manpower required required to  support the to support the ASSD ASSD actions.
actions.


==Reference:==
==Reference:==
OASSD-00, Revision 37, Page 30, Section 5.3.3 Cog Level: High Explanation:
The ASSD staffing composition may be less than the minimum requirements for a period of time not to exceed two hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore requirements to within the minimum requirements of the shift ASSD staffing.
If the ASSD staffing composition is less than the minimum required, establish an Alternative Safe Shutdown Impairment in accordance with OPLP-1.5, Alternative Shutdown Capability Controls, and OFPP-020, Impairment Notification.
Distractor Analysis:
Choice A: Correct Answer Choice B: Plausible because an impairment is the same as LCO (001-01.01) but impairments are not established against TS 5.2.2 and ASSD User Guide is correct.
Choice C: Plausible because ASSD impairment is correct and 001-01.01 provides staffing requirements for TS 5.2.2 but directs use of OASSD-00 procedure use for required staffing.
Choice D: Plausible because Plausible because an impairment is the same as LCO (001-01.01) but impairments are not established against TS 5.2.2 and 001-01.01 provides staffing requirements for TS 5.2.2 but directs use of OASSD-00 procedure use for required staffing.
SRO Only Basis: Requires knowledge of TS 5.2.2 Facility Staff Organization
- and prescribes the procedure required for guidance during periods of ASSD minimum complement not maintained.
Notes Feedback KIA: SG2.01.05 Conduct of Operations Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.
(CFR: 41.10/43.5/45.12)
RO/SRO Rating: 2.9/3.9 Objective: CLS-LP-304-M*13m
: 13. Given ASSD procedures and plant conditions that require use of ASSD procedures, determine the following:
: m. The manpower required to support the ASSD actions.


==Reference:==
==Reference:==
 
OASSD-OO, Revision 37, Page 30, Section 5.3.3 Cog Level: High Explanation:
OASSD-00, Revision OASSD-OO,       Revision 37, 37, Page Page 30, 30, Section Section 5.3.3 5.3.3 Cog Level:
The ASSD staffing composition may be less than the minimum requirements for a period of time not to exceed two hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore requirements to within the minimum requirements of the shift ASSD staffing. If the ASSD staffing composition is less than the minimum required, establish an Alternative Safe Shutdown Impairment in accordance with OPLP-1.5, Alternative Shutdown Capability Controls, and OFPP-020, Impairment Notification.
Cog            High Level: High Explanation:
Explanation:
The ASSD The  ASSD staffing staffing composition composition may may be  less than be less   than the the minimum minimum requirements requirements for for a period period of time time not not to to accommodate unexpected absence of on-duty shift crew members provided exceed two hours in order to accommodate immediate action is taken to restore requirements to within the minimum requirements of the shift ASSD staffing. If the ASSD staffing composition is less than the minimum  minimum required, establish an Alternative Safe Shutdown Impairment in accordance with OPLP-1.5,  OPLP-1 .5, Alternative Shutdown Capability Controls, and OFPP-020, Impairment Notification.
Distractor Analysis:
Distractor Analysis:
Choice A: Correct Answer Choice B: Plausible because an impairment is the same as LCO (001-01.01) but impairments are not established against TS 5.2.2 and ASSD User Guide is correct.
Choice A: Correct Answer Choice B: Plausible because an impairment is the same as LCO (001-01.01) but impairments are not established against TS 5.2.2 and ASSD User Guide is correct.
Choice C: Plausible because ASSD impairment is correct and 001-01.01 provides staffing requirements OASSD-00 procedure use for required staffing.
Choice C: Plausible because ASSD impairment is correct and 001-01.01 provides staffing requirements for TS 5.2.2 but directs use of OASSD-OO procedure use for required staffing.
for TS 5.2.2 but directs use of OASSD-OO Choice D: Plausible because Plausible because an impairment is the same as LCO (001-01.01) but impairments are not established against TS 5.2.2 and 001-01.01 provides staffing requirements for TS 5.2.2 but directs use of OASSD-00 OASSD-OO procedure use for required staffing.
Choice D: Plausible because Plausible because an impairment is the same as LCO (001-01.01) but impairments are not established against TS 5.2.2 and 001-01.01 provides staffing requirements for TS 5.2.2 but directs use of OASSD-OO procedure use for required staffing.
SRO Only Basis: Requires knowledge of TS 5.2.2 Facility Staff        Staff Organization - and prescribes the procedure procedure required for guidance guidance during during periods of ASSD minimum minimum complement complement notnot maintained.
SRO Only Basis: Requires knowledge of TS 5.2.2 Facility Staff Organization - and prescribes the procedure required for guidance during periods of ASSD minimum complement not maintained.
Notes
Notes  


5.0 5.0    INSTRUCTIONS INSTRUCTIONS 5.3 General 5.3    General GUidelines Guidelines for   ASSD Staff for ASSD        Staff 6.31 5.3.1    All ASSD All  ASSD Staffing Staffing Roster Roster members members must must be  capable of be capable   of prompt prompt response when response   when events events are   are inin progress progress that that may may require     entry into require entrl    into ASSO procedures.
5.0 INSTRUCTIONS 5.3 General Guidelines for ASSD Staff 6.31 All ASSD Staffing Roster members must be capable of prompt response when events are in progress that may require entry into ASSO procedures.
ASSD    procedures.
5.12 All ASSD members shall obtain a designated radio at the beginning of shift and ensure that it is charged.
5.12 5.3.2    All ASSD All  ASSD members members shall shall obtain obtain aa deSignated designated radio radio at at the the beginning beginning of  of shift and shift      ensure that and ensure   that itit is is charged.
charged.
rNol-E:
rNol-E:
NOTE:      Planned reduction Planned    reduction of of ASSn ASSD personnel personnel below below the the minimum minimum number number required required permitted.
Planned reduction of ASSD personnel below the minimum number required is NOT permitted.
NOT permitted.
5.3.3 The ASSD staffing composition may be less than the minimum requirements for a period of time not to exceed two hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore requirements to within the minimum requirements of the shift ASSD staffing.
is NOT is
5.3.4 If the ASSD staffing composition is less than the minimum required.
* 5.3.3   The ASSD staffing composition composition may          be less may be   less than the minimum minimum requirements for aa period requirements          period of  of time not toto exceed exceed twotwo hours hours in order order toto accommodate unexpected accommodate      unexpected absenceabsence of  of on-duty on-duty shift shift crew crew members members provided immediate provided  immediate action is      is taken to restore restore requirements requirements to within the minimum requirements of the shift ASSn        ASSD staffing.
establish an Alternative Safe Shutdown Impairment in accordance with OPLP-1.5, Alternative Shutdown Capability Controls, and OFPP-020, Impairment Notification.
5.3.4           ASSD staffing composition is less than the minimum required, If the ASSn                                                                 required.
5.3.5 If an impairment exceeds two hours, initiate a Condition Report.
establish an Alternative Safe Shutdown Impairment in accordance '.'1ith               with OPLP-1.5, Alternative Shutdown Capability Controls, and OFPP-020, Impairment Notification.
5.3.6 With both units in Mode 4 or 5, ASSD staffing is not required.
5.3.5   If an impairment exceeds two hours, initiate a Condition Report.
DASSO-QO Rev. 37 Page 30 of 53 5.0 INSTRUCTIONS 5.3 General GUidelines for ASSD Staff 5.3.1 All ASSD Staffing Roster members must be capable of prompt response when events are in progress that may require entrl into ASSD procedures.
5.3.6   With both units in Mode 4 or 5, ASSD staffing is not required.
5.3.2 All ASSD members shall obtain a deSignated radio at the beginning of shift and ensure that it is charged.
DASSO-QO 10ASSD-OO                                       Rev. 37 Rev. 37                               Page Page 30    of 53 30 of   531
NOTE:
Planned reduction of ASSn personnel below the minimum number required is NOT permitted.
5.3.3 The ASSD staffing composition may be less than the minimum requirements for a period of time not to exceed two hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore requirements to within the minimum requirements of the shift ASSn staffing.
5.3.4 If the ASSn staffing composition is less than the minimum required, establish an Alternative Safe Shutdown Impairment in accordance '.'1ith OPLP-1.5, Alternative Shutdown Capability Controls, and OFPP-020, Impairment Notification.
5.3.5 If an impairment exceeds two hours, initiate a Condition Report.
5.3.6 With both units in Mode 4 or 5, ASSD staffing is not required.
10ASSD-OO Rev. 37 Page 30 of 531  


5.0 5.0    INSTRUCTIONS INSTRUCTIONS 5.4 Minimum 5.4     Minimum ASSD ASSD Nuclear Nuclear Shift Shift Staffing/Assignments Staffing/Assignments 5.4.1 5.4.*1  Senior Reactor Senior    Reactor Operators:
5.0 INSTRUCTIONS 5.4 Minimum ASSD Nuclear Shift Staffing/Assignments 5.4.1 Senior Reactor Operators:
Operators:
1 Unit 1 SCO: Unit 1 Remote Shutdown Panel 1
1   Unit 11 seo:
Unit 2 SCO:
Unit    SCO: Unit Unit 11 Remote Remote Shutdown Shutdown Panel Panel 11  Unit 22 seo:
Unit 2 Remote Shutdown Panel 5.4.2 Auxiliary Operators:
Unit   SCO: Unit Unit 22 Remote Remote Shutdown Shutdown Panel Panel 5.4.2 5.4.2  Auxiliary Operators:
1 Unit 1 Reactor BuildinglMCC Operator or as directed by the Unit SCO 1
Auxiliary 1   Unit 11 Reactor Unit   Reactor Building/MCe BuildinglMCC Operator Operator or as   directed by as directed by the Unit seo Unit SCO 11    Unit 22 Reactor Unit   Reactor BuildingJMCe BuildinglMCC Operator Operator or or as directed by as directed by the the Unit seo Unit  SCO 11    Diesel Generator Operator or as directed by the Unit seoSCO 11                Switchgear Operator Emergency Switchgear     Operator or as directed by the as directed Unit seo Unit  SCO 11    Service Water Building Operator or as directed by the Unit SCO UnitSeO
Unit 2 Reactor BuildinglMCC Operator or as directed by the Unit SCO 1
\ QASSD-00 OASSD-OO                                Rev. 37                           Page Page 3131 ofj of 53\
Diesel Generator Operator or as directed by the Unit SCO 1
Emergency Switchgear Operator or as directed by the Unit SCO 1
Service Water Building Operator or as directed by the Unit SCO QASSD-00 Rev. 37 Page 31 ofj 5.0 INSTRUCTIONS 5.4 Minimum ASSD Nuclear Shift Staffing/Assignments 5.4.*1 Senior Reactor Operators:
Unit 1 seo: Unit 1 Remote Shutdown Panel 1
Unit 2 seo: Unit 2 Remote Shutdown Panel 5.4.2 Auxiliary Operators:
\\ OASSD-OO Unit 1 Reactor Building/MCe Operator or as directed by the Unit seo 1
Unit 2 Reactor BuildingJMCe Operator or as directed by the Unit seo 1
Diesel Generator Operator or as directed by the Unit seo 1
Emergency Switchgear Operator or as directed by the Unit seo 1
Service Water Building Operator or as directed by the UnitSeO Rev. 37 Page 31 of 53\\  


9.4 9.4    Operations Leadership Operations     Leadership RoleRole in in Station Station Activities        (continued)
9.4 Operations Leadership Role in Station Activities (continued) 5.
Activities (continued) 5.
Operators work closely with station support personnel to establish appropriate priorities for resoMng plant equipment and station program deficiencies. Being aware of the integrated effect of equipment out of service and establishing priorities for equipment return-to-service consistent with plant impact are key components of this philosophy.
: 5.      Operators work Operators   work closely closely with with station station support support personnel personnel to  to establish establish appropriate priorities appropriate   priorities for   resoMng plant for resolving    plant equipment equipment and        station program and station     program deficiencies. Being deficiencies. Being aware aware of  of the the integrated integrated effect effect of of equipment equipment out  out of of service and service   and establishing establishing priorities priorities for for equipment equipment return-to-service return-to-service consistent with consistent   with plant   impact are plant impact      are key key components components of    of this this philosophy.
philosophy.
6.
6.
: 6.      Operations pursues OperatiOns    pursues the     root cause(s}
Operations pursues the root cause(s) of problems; provides direction to implement corrective actions and hold department and station personnel accountable for achieving expected levels of perfomance.
the root    cause(s) of of problems; problems; provides provides direction direction to implement to  implement corrective corrective actions actions and      hold department and hold     department and  and station station personnel accountable personnel    accountable for for achieving achieving expected expected levels levels ofof perfornlance.
perfomance.
9.5.
9.5.
9.5. Operations Shift Operations     Shift Slaffing StaflinO Standards Standards Operations ensures Operations   ensures that the Control Room   Room is  is adequately adequately staffed for    for plant plant operations with appropriately with appropriately qualified qualified individuals.
Operations Shift StaflinO Standards Operations ensures that the Control Room is adequately staffed for plant operations with appropriately qualified individuals. Additionally, Operations ensures staffing is adequate to meet regulatory and programmatic requirements.
individuals. Additionally, Additionally, Operations Operations ensures ensures staffing staffing is is adequate to meet adequate      meet regulatory regulatory andand programmatic programmatic requirements.
requirements.
Expectations 1.
Expectations 1.
: 1. General
General a.
: a.     The CRS and Shift Manager are responsible for ensuring that watchstanders hold required positions. Personnel only qualified watchstanders should verify they are qualified for the position to be held prior to assuming the watch.
The CRS and Shift Manager are responsible for ensuring that only qualified watchstanders hold required positions. Personnel should verify they are qualified for the position to be held prior to assuming the watch.
: b.       Individual qualifications qualifications for specific specifc positions can be found in REG-NGGC-0012, Confirmation of Personnel Qualifications REG-NGGC-0012,                                                Qualifications Associated with Commitments Commitments to Regulatory Guide 1.8.
b.
: c.     The shift complement may be one less than the minimum requirement for a  a period of time not to exceed 2 hours in order to accommodate accommodate unexpected absence of on-duty            onduty shift meml)ers members provided immediate action is taken to restore the shift complement to within the minimum requirements. requirements. This provision does not permit any shift member position to be unmanned upon shift change due to an oncoming shift member         meml)er beingI)eing late or or absent.
Individual qualifications for specifc positions can be found in REG-NGGC-0012, Confirmation of Personnel Qualifications Associated with Commitments to Regulatory Guide 1.8.
absent.
c.
The shift complement may be one less than the minimum requirement for a period of time not to exceed 2 hours in order to accommodate unexpected absence of onduty shift members provided immediate action is taken to restore the shift complement to within the minimum requirements. This provision does not permit any shift member position to be unmanned upon shift change due to an oncoming shift member being late or absent.
d.
d.
: d.      Shift staffing shall shall meet the  the requirements of the indMdual individual plant license/Tech licenseffech Specs Specs and and other regulatory regulatory andand programmati programmatic     c required   positions at required positions     at all all times. Required Required staff staff numbers numbers and  and positions positions can be be found in  in Attachment I1 "Shift Shift Staffing.
Shift staffing shall meet the requirements of the indMdual plant license/Tech Specs and other regulatory and programmatic required positions at all times. Required staff numbers and positions can be found in Attachment I Shift Staffing.
Staffing".
OPS-NGGC-1000 Rev.2 I
I OPS-NGGC   -1000 OPS-NGGC-1000                                   Rev.2 Rev. 2                 I                                1491 Page 46 of 1491 Page46of
Page46of 1491 9.4 Operations Leadership Role in Station Activities (continued)
: 5.
Operators work closely with station support personnel to establish appropriate priorities for resolving plant equipment and station program deficiencies. Being aware of the integrated effect of equipment out of service and establishing priorities for equipment return-to-service consistent with plant impact are key components of this philosophy.
: 6.
OperatiOns pursues the root cause(s} of problems; provides direction to implement corrective actions and hold department and station personnel accountable for achieving expected levels of perfornlance.
9.5.
Operations Shift Slaffing Standards Operations ensures that the Control Room is adequately staffed for plant operations with appropriately qualified individuals. Additionally, Operations ensures staffing is adequate to meet regulatory and programmatic requirements.
Expectations
: 1.
General
: a.
The CRS and Shift Manager are responsible for ensuring that only qualified watchstanders hold required positions. Personnel should verify they are qualified for the position to be held prior to assuming the watch.
: b.
Individual qualifications for specific positions can be found in REG-NGGC-0012, Confirmation of Personnel Qualifications Associated with Commitments to Regulatory Guide 1.8.
: c.
The shift complement may be one less than the minimum requirement for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift meml)ers provided immediate action is taken to restore the shift complement to within the minimum requirements. This provision does not permit any shift member position to be unmanned upon shift change due to an oncoming shift meml)er I)eing late or absent.
: d.
Shift staffing shall meet the requirements of the individual plant licenseffech Specs and other regulatory and programmatic required positions at all times. Required staff numbers and positions can be found in Attachment 1 "Shift Staffing".
I OPS-NGGC-1000 Rev. 2 Page 46 of 1491
 
Attachment I
- Shift Staffing Shift Manning BNP Position Minimum Note staffing SM I
CRS 2
SRO/STA 1
RO 3
AO 9
9.5 Operations Shift Staffing In addition to the requirements of OPS-NGGC-1000. the following requirements apply:
9.5.1 General The following table outlines the administrative guideline for the normal Operations shift complement Any deviation from the normal shift complement must remain in accordance with Section 5.2.2 of Technical Specifications, and apphcable sections of IJASSD-OO.
OFPP-031. and Attachment 13. (Attachment 13 contains a listing of required ERO Watch Stations and qualifications for each and ASSD position& This attachment may be used as a tool to support determining shift staffing requirements.)
BNP Watchstations BNP Shift Complement License Shift Manager (SM) 1 Shift Manager SRO Control Room Supervisor (CRS) 2 CRSs (1 for each unit)
SRO Reactor Operator (RO) 4 Reactor Operators rtypically. 2 ROISRO for each unit)
Auxiliarj Operator (AO) 9 (includes 2 in Radwaste)
N/A Operations Center/Field SRO 1 Operations Center/Field SRO SRO STA 1 STA STA Qualified


Attachment 1I - Shift Attachment          -  Shift Staffing Staffing Sheet 11 of2 Sheet        of 2 Shift Manning Shift Manning BNP BNP Position Position              Minimum Minimum            Note Note staffing staffing SM SM                   1I CRS CRS                     22 SRO/STA SRO/STA                   11 RO RO                   33 AO AO                    99 9.5 9.5       Operations Shift Operations         Shift Staffing Staffing In addition In  addition to to the the requirements requirements of    of OPS-NGGC-1000, OPS-NGGC-1000. the      the following following requirements apply:
The STA may stand watch as a CRS or Reactor Operator provided the following requirements are met:
requirements        apply:
At least 4 SROs are available on shift (this includes the STA hut does NOT include the Fire Brigade Advisor which may be filled by an RO licensed individual).
9.5.1         General The following table outlines the administrative guideline for the normal Operations normal      Operations shift complement A.flY         Any deviation from the normal shift complement must remain in accordance with Section 5.2.2 of Technical Specifications, and applicable apphcable sections of OASSO-OO, IJASSD-OO.
Another Licensed Operator is designated to relieve the STA as Unit CRS or RO.
OFPP-031. and Attachment 13. (Attachment 13 contains a listing of OFPP-031, required ERO Watch Stations and qualifications for each and ASSD position& This attachment may be used as a tool to support positions.
(Relief as Reactor Operator is required if only one operator is assigned to a unit Relief as CR5 shall he filled from the CR5 position on the shift staffing roster.)
determining shift staffing requirements.)
The designated relief must NOT be assigned as the Fire Brigade Advisor.
Watchstations BNP Watchstations                                   BNP Shift Com~lement Complement                    License Shift Manager (SM)                                   1 Shift Manager 1                                            SRO Control Room Supervisor (CRS)                       2 CRSs (1 for each unit)                       SRO Reactor Operator (RO)                               4 Reactor Operators (typically, rtypically. 2   ROISRO RO/SRO for each unit)
The designated relief has taken turnover on the affected unit.
Auxiliarj Auxiliary Operator (AO)                             9 (includes 2 in Radwaste)                     N/A Operations Center/Field SRO                         11 Operations Center/Field Center/Freid SRO           SRO STA STA'                                               11 STA STA                                        STA Qualified STA  Qualified The
The designated relief must he able to relieve the STA within 10 minutes.
        'The STASTA may may stand stand watch watch as as aa CRS CRS or or Reactor Reactor Operator Operator provided the    the following requirements are  are met:
001-Ui.131 Rev.
met:
Page 14 of 177 Sheet 1 of 2 Shift Manning BNP Position SM CRS SRO/STA RO AO - Shift Staffing Sheet 1 of2 Minimum Note staffing 1
* At At least least 44 SROs SROs are are available available on  on shift shift (this (thiS includes includes thethe STA STA hut but does  NOT does .NOT include include the the Fire Fire Brigade Brigade Advisor Advisor which which may may bebe filled  by an filled by an RO RO licensed licensed individual).
2 1
individual).
3 9
* Another Another Licensed Licensed Operator Operator is is designated designated to    to relieve relieve the   STA as the STA    as Unit Unit CRS CRS or or RO.
9.5 Operations Shift Staffing In addition to the requirements of OPS-NGGC-1000, the following requirements apply:
RO.
9.5.1 General The following table outlines the administrative guideline for the normal Operations shift complement A.flY deviation from the normal shift complement must remain in accordance with Section 5.2.2 of Technical Specifications, and applicable sections of OASSO-OO, OFPP-031, and Attachment 13. (Attachment 13 contains a listing of required ERO Watch Stations and qualifications for each and ASSD positions. This attachment may be used as a tool to support determining shift staffing requirements.)
(Relief as (Relief   as Reactor Reactor Operator Operator is is required      only one required ifif only   one operator operator isis assigned assigned to to aa unit unit Relief Relief as as CR5 CRS shall shall he be filled filled from from thethe CR5 CRS position position on on the the shift shift staffing staffing roster.)
BNP Watchstations BNP Shift Com~lement License Shift Manager (SM) 1 Shift Manager SRO Control Room Supervisor (CRS) 2 CRSs (1 for each unit)
roster.)
SRO Reactor Operator (RO) 4 Reactor Operators (typically, 2 RO/SRO for each unit)
* The The designated deSignated relief relief must must NOTNOT be  be assigned assigned as       the Fire as the    Fire Brigade Brigade Advisor.
Auxiliary Operator (AO) 9 (includes 2 in Radwaste)
Advisor.
N/A Operations Center/Field SRO 1 Operations Center/Freid SRO SRO STA' 1 STA ST A Qualified  
            **    The The designated deSignated relief relief has  taken turnover has taken     turnover on on the the affected affected unit.
'The STA may stand watch as a CRS or Reactor Operator provided the following requirements are met:
unit.
At least 4 SROs are available on shift (thiS includes the STA but does.NOT include the Fire Brigade Advisor which may be filled by an RO licensed individual).
* The designated The   deSignated relief relief must must he    able to be able   to relieve relieve the the STA      within 10 STA within   10 minutes.
Another Licensed Operator is designated to relieve the STA as Unit CRS or RO.
minutes.
(Relief as Reactor Operator is required if only one operator is assigned to a unit Relief as CRS shall be filled from the CRS position on the shift staffing roster.)
001-Ui .131 1001-01.01                                             Rev.
The deSignated relief must NOT be assigned as the Fire Brigade Advisor.
Rev. 29                                   Page Page 1414 of   1771 of 177
The deSignated relief has taken turnover on the affected unit.
The deSignated relief must be able to relieve the STA within 10 minutes.
1001-01.01 Rev. 29 Page 14 of 1771  


ATTACHMENT 13 ATTACHMENT      13 of 22 Page 11 of Page Operations Staffing Operations   Staffing Roster Roster Date: _ _ _ _ _ __
ATTACHMENT 13 Page 1 of 2 Operations Staffing Roster Date:__________________
Date:__________________ Shift _______
Shift:________________
Shift:________________
STA     P613                                       STA SRO       PB11                                       Unit 1 CRS/U-1 RSD Panel SRO     PB1I                                       Unit 2 CRS1U-2 RSD Panel ARO       PB12                                       Unit 1 RB MCC Operator ARC       P812                                       Unit 2 RB MCC Operator ARC       PB12                                       *FB Advisor CREC     PBI7                                       CREC AC       PB14                                       SW Operator AC       P614                                       DC Operator AC       P814                                       Emergency Switchgear Operator FB (SIC) FBO2FBD3                                   FB (SIC)
STA P613 STA SRO PB11 Unit 1 CRS/U-1 RSD Panel SRO PB1I Unit 2 CRS1U-2 RSD Panel ARO PB12 Unit 1 RB MCC Operator ARC P812 Unit 2 RB MCC Operator ARC PB12
FB       FBO2                                       F8 F8       FBO2                                       FB FB       FGO2                                       FB FB       F802                                       FB Security Contact E&RC Contact Maintenance Contact A
*FB Advisor CREC PBI7 CREC AC PB14 SW Operator AC P614 DC Operator AC P814 Emergency Switchgear Operator FB (SIC) FBO2FBD3 FB (SIC)
FB FBO2 F8 F8 FBO2 FB FB FGO2 FB FB F802 FB Security Contact E&RC Contact Maintenance Contact A
May hold an RO OR SRO license.
Security Key Accountability iiiui; ii*;
Unit I RB AC Unit2 RBAO Outside AC Unit 1 CRS Un1t2CRS Shift Manager 001-01.01 Rev 29 Page 97 of 177 Date: ______
ATTACHMENT 13 Page 1 of 2 Operations Staffing Roster Shift ______ _
May hold an RO OR SRO license.
May hold an RO OR SRO license.
Security Key Accountabili Accountability   ty Unit I RB AC iiiui;                                    ii*;
Security Key Accountability 1001-01.01 Rev. 29 Page 97 of 1771  
Unit2 RBAO Outside AC Unit 1 CRS Un1t2CRS Shift Manager 001-01.01 1001-01.01                           Rev. 29 Rev  29                        Page Page 97 97 of 1771 of 177


5.2.2 5.2.2            Facility Staff Facility  Staff The facility The  facility staff staff organization organization shall shall include include the the fol[owing:
5.2.2 Facility Staff The facility staff organization shall include the following:
following:
a.
a.
: a.        AA total total of   three non-licensed of three   non-licensed operators operators shall shall bebe assigned assigned for  for Brunswick Brunswick Units 1I and Units              at all and 22 at  all times.
A total of three non-licensed operators shall be assigned for Brunswick Units I and 2 at all times.
times.
Organization 5.2 5.2 Organization 5.2.2 Facility Staff continued) b.
At east one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, when either unit is in MODE 1, 2, or 3. at least one licensed Senior Reactor Operator (SRO) shall be present in the control room. With one unit in MODE 1, 2, or 3 and the other unit defueled, the minimum shift crew shall include a total of two SROs and two ROs.
c Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and Specifications 5.2.2.a and 5.2.2.g for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.
d.
An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.
e.
Deleted.
f.
The operations manager or assistant operations manager shall hold an.
SRO license.
g.
The shift technical advisor shall serve in an advisory capacity to the shift superintendent on matters pertaining to the engineering aspects assuring safe operation of the unit when either unit is in MODE 1.2, or 3.
Eiunswick Unit 1 5.0-3 Amendment No. 253 I
5.2.2 Facility Staff The facility staff organization shall include the fol[owing:
: a.
A total of three non-licensed operators shall be assigned for Brunswick Units 1 and 2 at all times.
5.2 Organization 5.2.2 Facility Staff (continued)
( continued)
( continued)
Organization Organization 5.2 5.2 5.2 Organization 5.2  Organization 5.2.2 5.2.2          Facility Staff Facility  Staff (continued) continued) b.
Organization 5.2
: b.             east one At least     one licensed licensed Reactor Reactor Operator Operator (RO)
: b.
(RO) shall shall be be present present in in the control room control    room when fuel is      is in in the reactor.
At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, when either unit is in MODEl, 2, or 3. at least one licensed Senior Reactor Operator (SRO) shall be present in the control room. V'I'ith one unit in MODEl, 2, or 3 and the other unit defueled, the minimum shift crew shall include a total of two SROs and two ROs.
reactor. InIn addition, when eithereither unit unit is is in in MODE 1, 2, or MODEl,            or 3. at at least least one licensed licensed Senior Senior Reactor Reactor Operator Operator (SRO)(SRO) shall be    present in be present     in the control room.        With one unit room. V'I'ith       unit in in MODEl, MODE 1, 2, or 3 and the other unit defueled, the minimum shift crew shall include a total of two SROs and two ROs.
: c.
c
Shift crew composition may be lee.s than the minimum requirement of 10 CFR SO.54(m){2)(i) and Specifications 5.2.2.a and 5.2.2.g for a peliod of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.
: c.       Shift crew composition may be lee.s           less than the minimum requirement of 10 50.54(m)(2)(i) and Specifications 5.2.2.a and 5.2.2.g for a peliod CFR SO.54(m){2)(i)                                                                      period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.
: d.
: d.       An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours, in order to provide for unexpected absence:         absence, provided immediate action is taken to fill the required pOSition. position.
An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours, in order to provide for unexpected absence: provided immediate action is taken to fill the required pOSition.
: e.       Deleted.
: e.
: f.         The The operations manager manager or    or assistant assistant operations operations manager manager shall shall hold hold an.
Deleted.
an SRO SRO license.
: f.
license.
The operations manager or assistant operations manager shall hold an SRO license.
g.
: g.
: g.        The The shift shift technical technical advisor advisor shall       serve in shall serve  in an an advisory advisory capacity capacity to to the the shift shift superintendent superintendent on     on matters        pertaining to matters pertaining       to the the engineering engineering aspects aspects assuring assuring safe safe operation operation of of the the unit unit when    either unit when either     unit is is in in MODE MODE 1.2, 1, 2, or or 3.
The shift technical advisor shall serve in an advisory capacity to the shift superintendent on matters pertaining to the engineering aspects assuring safe operation of the unit when either unit is in MODE 1, 2, or 3.
3.
Brunswick Unit '1 5.0-3 Amendment No. 253 I
Eiunswick Brunswick Unit Unit 1'1                                         5.0-3 5.0-3                                   Amendment No.
 
Amendment          No. 253 253  II
Categories K/A:
SG2.0L05 Tier/Group:
T3 RO Rating:
2.9 SRO Rating:
3.9 LP Obj:
CLSLP3O4M*13M Source:
NEW Cog Level:
HIGH Category 8:
Y Categories KIA:
SG2.01.05 Tier / Group: T3 RORating:


Categories Categories K/A:
===2.9 SRORating===
KIA:        SG2.0L05 SG2.01.05        Tier/Group:
3.9 LP Obj:
Tier / Group:  T3 T3 RO  Rating:
CLS-LP-304-M* 13M Source:
RORating:  2.9 2.9              SRO Rating:
NEW Cog Level:
SRORating:    3.9 3.9 LP Obj:
HIGH Category 8:
LP Obj:      CLSLP3O4M*13M CLS-LP-304-M* 13M Source:
Y
Source:      NEW NEW Cog Level:
: 95. OFH-1 1, Refueling, prohibits control rod withdrawal during the core load sequence until a neutronic bridge is established.
Cog  Level:  HIGH HIGH             Category 8:8:
Which one of the following core loading sequences establishes a neutronic bridge as described in OFH-1 1?
Category      YY
Four fuel bundles are loaded around (1)
: 95. OFH-11,
,then fuel is loaded in all fuel cells in a line between SRMs (2)
: 95. OFH-1 1, Refueling, Refueling, prohibits prohibits control control rod rod withdrawal withdrawal during during the  the core core load load sequence sequence untiluntil neutronic bridge aa neutronic   bridge isis established.
A. (1) SRMsAand Donly (2) AandD B.
established.
(1) SRMs B and D only (2) B and D C. (1) each of the four SRMs (2) A and D D (1) each of the four SRMs (2) B and D
Which one Which    one of of the the following following core core loading loading sequences sequences establishes establishes aa neutronic neutronic bridge bridge as as described inin OFH-11?
: 95. OFH-11, Refueling, prohibits control rod withdrawal during the core load sequence until a neutronic bridge is established.
described      OFH-1 1?
Which one of the following core loading sequences establishes a neutronic bridge as described in OFH-11?
Four fuel Four        bundles are fuel bundles          loaded around are loaded  around    (1)
Four fuel bundles are loaded around (1)  
(1)  ,then fuel
,then fuel is loaded in all fuel cells in a line between SRMs (2)
                                                          ,then  fuel isis loaded loaded inin all all fuel fuel cells cells in in aa line between line  between SRMs SRMs (2)     (2) .
A. (1) SRMs A and D only (2) A and D B. (1) SRMs Band D only (2) Band D
A.  (1) SRMs A. (1)   SRMsAandA and D   Donly only (2) A (2) AandD and D B. (1)
: c. (1) each of the four SRMs (2) A and D D~ (1) each of the four SRMs (2) Band D
B.   (1) SRMs SRMs BandB and D D only only (2) Band (2) B and D  D c.
C.  (1) each of the four SRMs and D (2) A and D
D~   (1) each of the four SRMs B and D (2) Band


Feedback Feedback K/A: SG2.01.42 KIA:  SG2.01.42 Conduct of Conduct   of Operations Operations Knowledge of Knowledge        new and of new and spent spent fuel fuel movement movement procedures.
Feedback K/A: SG2.01.42 Conduct of Operations Knowledge of new and spent fuel movement procedures.
procedures.
(CFR: 41.10/43.7/45.13)
(CFR: 41.10/43.7/45.13)
(CFR:  41.10/43.7/45.13)
RO/SRO Rating: 2.5/3.4 Objective: CLSLP305C*
RO/SRO Rating:
RO/SRO     Rating: 2.5/3.4 2.5/3.4 CLSLP305C*
Objective: CLS-LP-305-C*
Objective:


==Reference:==
==Reference:==
OFH-1, Revision 93, Page 9, Section 4.37 Cog Level: High Explanation:
Provide ENP-24-12, Figure 1 as a reference From FH-11, 4.37 To help ensure that an unmonitored criticality will not occur, control rod withdrawal is not allowed during the core reload sequence until a neutronic bridge is established. The neutronic bridge ensures that two SRMs are neutronically coupled, thus monitoring the loaded area of the core. The reload sequence has three basic steps. Four fuel bundles are loaded around each of the four SRMs, the neutronic bridge is established and a spiral reload of the other fuel bundles completes the sequence. The neutronic bridge is established by loading fuel in all fuel cells in a line between two SRMs. These SRMs must be on opposite sides of the core and the line of loaded fuel cells must intersect the center of the core (A to D would not intersect the center, B to D would).
Distractor Analysis:
Choice A: Plausible because loading fuel around 2 SRMs and a line between them would establish a neutron bridge (between those 2 SRM5) but not PAW OFH-1 1 and A&D are adjacent.
Choice B: Plausible because loading fuel around 2 SRMs and a line between them would establish a neutron bridge (between those 2 SRM5) but not JAW OFH-1 I and A&D are adjacent.
Choice C: Plausible because loading fuel around all SRMs is correct but A&D are adjacent.
Choice D: Correct Answer SRO Only Basis: IOCFR55.43.6 Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity.
IOCFR55.43.7 Fuel handling facilities and procedures.
Notes Feedback KIA: SG2.01.42 Conduct of Operations Knowledge of new and spent fuel movement procedures.
(CFR: 41.10/43.7/45.13)
RO/SRO Rating: 2.5/3.4 Objective: CLS-LP-305-C*


==Reference:==
==Reference:==
 
OFH-1, Revision 93, Page 9, Section 4.37 Cog Level: High Explanation:
Revision 93, Page OFH-1, Revision       Page 9, Section 4.37 High Cog Level: High Explanation:
Provide ENP-24-12, Figure 1 as a reference From FH-11, 4.37 To help ensure that an unmonitored criticality will not occur, control rod withdrawal is not allowed during the core reload sequence until a neutronic bridge is established. The neutronic bridge ensures that two SRMs are neutronically coupled, thus monitoring the loaded area of the core. The reload sequence has three basic steps. Four fuel bundles are loaded around each of the four SRMs, the neutronic bridge is established and a spiral reload of the other fuel bundles completes the sequence. The neutronic bridge is established by loading fuel in all fuel cells in a line between two SRMs. These SRMs must be on opposite sides of the core and the line of loaded fuel cells must intersect the center of the core (A to D would not intersect the center, B to D would).
Explanation:
Provide ENP-24-12, Provide ENP-24-12, Figure Figure 11 as a reference reference From FH-11, 4.37 To help ensure that an unmonitored criticality will not occur, control rod withdrawal is not allowed during the core reload sequence until a neutronic bridge is established. The neutronic bridge ensures that two SRMs are neutronically coupled, thus monitoring the loaded area of the core. The reload sequence has three basic steps. Four fuel bundles are loaded around each of the four SRMs, the neutronic bridge is established and a spiral reload of the other fuel bundles completes the sequence. The neutronic bridge is established by loading fuel in all fuel cells in a line between two SRMs. These SRMs must be on opposite sides of the core and the line of loaded fuel cells must intersect the center of the core (A to D would not intersect the center, B to D would).
Distractor Analysis:
Distractor Analysis:
Choice A: Plausible because loading fuel around 2 SRMs and a line between them would establish a neutron bridge (between those 2 SRMs)SRM5) but not lAW PAW OFH-11 OFH-1 1 and A&D are adjacent.
Choice A: Plausible because loading fuel around 2 SRMs and a line between them would establish a neutron bridge (between those 2 SRMs) but not lAW OFH-11 and A&D are adjacent.
Choice B: Plausible because loading fuel around 2 SRMs and a line between them would establish a neutron bridge (between those 2 SRM5)                JAW OFH-1 SRMs) but not lAW   OFH-11I and A&D are adjacent.
Choice B: Plausible because loading fuel around 2 SRMs and a line between them would establish a neutron bridge (between those 2 SRMs) but not lAW OFH-11 and A&D are adjacent.
Choice C: Plausible because loading fuel around all SRMs is correct but A&D are adjacent.
Choice C: Plausible because loading fuel around all SRMs is correct but A&D are adjacent.
Choice D: Correct Answer SRO Only Basis: IOCFR55.43.6 10CFR55.43.6 Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity.
Choice D: Correct Answer SRO Only Basis: 10CFR55.43.6 Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity.
IOCFR55.43.7 Fuel handling 10CFR55.43.7         handling facilities and procedures.
10CFR55.43.7 Fuel handling facilities and procedures.
Notes
Notes
 
4.0 PRECAUTIONS AND LIMITATIONS 4.34 RPS shorting links SHALL be removed for control rod withdrawal (except for control rods removed in accordance with Technical Specifications) in the refuel mode when core verification AND subsequent strongest rod out verification have NOT been performed. Control rods may be withdrawn with the shorting links installed, provided core verification (QENP-24.13),
subsequent strongest rod out verification (single control rod suhcriticality test in accordance with OFH-1 1) have been performed, and the one-rod-out refuel interlocks have been demonstrated to be operable.
4.35 An SRO with no other concurrent duties shall directly supervise all core alterations.
4.36 Members of fuel handling crew, scheduled for consecutive daily duty, should NOT normally work more than 12 hours out of each 24 hours.
4.37 To help ensure that an unmonitored criticality will NOT occur, control rod withdrawal is NOT allowed during the core reload sequence until a neutronic bridge is established. The neutronic bridge ensures that two SRM5 are neutronically coupled, thus monitoring the loaded area of the core. The reload sequence has three basic steps. Four fuel bundles are loaded around each of the four SRMs, the neutronic bridge is established and a spiral reload of the other fuel bundles completes the sequence. The neutronic bridge is established by loading fuel in all fuel cells in a line between two SRM& These SRMs must be on opposite sides of the core and the line of loaded fuel cells must intersect the center of the core.
4.38 With fuel removed, if a control rod is withdrawn without blade guides installed, the insertion capability shall be removed for the control rod.
4.39 The Bridge Operator should immediately push the STOP button if the bridge fails to respond to Operator commands, such as speed changes or jogs. The STOP button will prevent all bridge movement 4.40 If attaching tools, such as ajet pump grapple or control blade latching tool, to either the monorail or frame mounted hoist, verify proper thread engagement/size by ensuring there is no play in the connection prior to thread engagement of three (3) full tum& The correct tool and coupling thread size is 7116-14 UNC. Additionally, a 112-13 UNC bolt will not fit into a proper size tool (7/16-14 U NC); thus, this check may be performed if practical. Failure to detect mis-matched thread sizes will significantly reduce the load capacity of the tool/hoist.
4.41 Indication of criticality observed on the SRM indicators during functional, subcritical, or shutdown margin rod checks shall be reason to teminate fuel loading until a complete evaluation of the cause of the criticality indication is determined.
OFH-1 1 Rev. 93 Page 9 of 55 4.0 PRECAUTIONS AND LIMITATIONS 4.34 RPS shorting links SHALL be removed for control rod withdrawal (except for control rods removed in accordance with Technical Specifications) in the refuel mode when core verification AND subsequent strongest rod out verification have NOT been performed. Control rods may be withdrawn with the shorting links installed, provided core verification (OENP-24.13),
subsequent strongest rod out verification (single control rod sub criticality test in accordance with OFH-11) have IJeen performed, and the one-rod-out refuel interlocks have been demonstrated to IJe operable.
4.35 An SRO with no other concurrent duties shall directly super"ise all core alterations.
4.36 Members of fuel handling crew, scheduled for consecutive dail~' duty, should NOT nom13l1y work more than 12 hours out of each 24 hours.
4.37 To help ensure that an unmonitored criticality will NOT occur, control rod withdrawal is NOT allowed during the core reload sequence until a neutronic bridge is established. The neutronic bridge ensures that two SRMs are neutronically coupled, thus monitoring the loaded area of the core. The reload seQuence has three basic steps. Four fuel bundles are loaded around each ofthe four SRMs, the neutronic bridge is established and a spiral reload of the other fuel bundles completes the sequence. The neutronic bridge is established by loading fuel in all fuel cells in a line between two SRMs. These SRMs must be on OPPOsite sides of the core and the line of loaded fuel cells must intersect the center of the core.
4.38 With fuel removed, if a control rod is withdrawn without blade guides installed, the insertion capabiliPf shall be removed for the control rod.
4.39 The Bridge Operator should immediately push the STOP button if the bridge failS to respond to Operator commands, such as speed changes or jogs. The STOP button will prevent an bridge movement.
4.40 If attaching tools, such as a jet pump grapple or controll)lade latching tOOl, to either the monorail or frame mounted hoist, verify proper thread engagement/size by ensuring there is no play in the connection prior to thread engagement of three (3) full turns. The correct tool and coupling thread size is 7116-14 UNC. Additionally, a 112-13 UNC IlOlt wiI[ not fit into a proper size tool {7/16-14 UNC); thus, this check may be performed if practical. Failure to detect mis-matched thread sizes will significantly reduce the load callacity of the tOOl/hoist.
4.41 Indication of criticality observed on the SRM indicators during functional, subcritical, or shutdown margin rod checks shall be reason to temlinate fuel loading until a complete evaluation of the cause oftM criticality indication is determined.
IOFH-11 Rev. 93 Page90f55 I
 
FIGURE 09.1-2 IN-Core Instrurnenation Location Diagram 00 LI IRM DETECTOR LOCATION DETECTOR LOCATION PLANT NORTH CORE RM LOCATION A
12.41 B
36-41 C
20-33 0
283.3 CORE IRM LOCATION E
28.25 F
20-26 G
36-09 H
12-09 SD-091 Rev. 6 Page 49 of 61 Categories K/A:
SG2.O1.42 RORating:
2.5 LP Obj:
Tier/Group:
T3 SRO Rating:
 
===3.4 Source===
BANK CORE SRM LOCATION A
12.33 B
2641 C
36-26 0
20.17 Cog Level:
HIGH Category 8:
Y FIGURE 09.1-2 IN-Core Instrumentation Location Diagram r
r r!,
LA..J r
(, ~
~ ~
L L
rHl I I
L 12 CORE SRM LOCATION A
12.33 a
28-41 C
36-25 0
20*11 I SD-09.1 Categories KIA:
SG2.01.42 RO Rating:


4.0 4.0      PRECAUTIONS AND PRECAUTIONS            AND LIMITATIONS LIMITATIONS 4.34 4.34      RPS shorting RPS    shorting links links SHALL SHALL be  be removed removed for for control control rod rod withdrawal withdrawal (except (except forfor control rods control    rods removed removed inin accordance accordance withwith Technical Technical Specifications)
===2.5 LPObj===
Specifications) inin thethe refuel mode refuel    mode whenwhen corecore verification verification AND AND subsequent subsequent strongest strongest rod rod out out verification have verification    have NOT NOT been been performed.
Cog Level:
performed. Control Control rods rods may may be be withdrawn withdrawn with with the shorting the    shorting links links installed, installed, provided provided core core verification verification (OENP-24.13),
HIGH PLANT ~
(QENP-24.13),
NORTH,..".-
subsequent strongest subsequent        strongest rodrod out out verification verification (single (single control control rod rod sub  criticality test suhcriticality    test in accordance in  accordance with  with OFH-11)
n h
OFH-1 1) have have IJeen been performed, performed, andand the the one-rod-out one-rod-out refuel interlocks refuel    interlocks havehave been been demonstrated demonstrated to  to IJe be operable.
/'" ~
operable.
;; l/
4.35 4.35    An SRO An    SRO withwith nono other other concurrent concurrent duties duties shall shall directly directly super"ise supervise all  all core core alterations.
B,-l
alterations.
--41
4.36 4.36      Members of Members      of fuel fuel handling handling crew,      scheduled for crew, scheduled          consecutive dail~'
~
for consecutive      daily duty, duty, should should NOT nom13l1y NOT    normally work work more more than than 12 12 hours hours outout of of each each 2424 hours.
ffr
hours.
-33 4:f. ~
4.37          help ensure that an unmonitored To help                          unmonitored criticality will NOT  NOT occur, control rod    rod withdrawal is  is NOT allowed allowed during the  the core core reload reload sequence until  until aa neutronic neutronic bridge is bridge        established. The is established. The neutronic neutronic bridge bridge ensures ensures that two  two SRMs SRM5 are  are neutronically coupled, neutronically      coupled, thus monitoring monitoring thethe loaded loaded area area of the the core.
t:l.f; ~
core. The reload seQuence reload    sequence has three basic steps. Four        Four fuel bundles are loaded  loaded around each ofthe of the four SRMs, the neutronicneutronic bridge is established and a spiral reload of the other fuel bundles completes the sequence. The neutronic bridge is established by loading fuel in all fuel cells in a line between two SRM& These SRMs must be on OPPOsite SRMs.                                        opposite sides of the core and the line of loaded fuel cells must intersect the center of the core.
'( -:;
4.38    With fuel removed, if a control rod is withdrawn without blade guides capability shall be removed for the control rod.
-25
installed, the insertion capabiliPf 4.39    The Bridge Operator should immediately push the STOP button if the bridge fails to respond to Operator commands, such as speed changes or jogs. The failS STOP button will prevent all        an bridge movement.
~ ~
movement 4.40    If attaching tools, such as a        ajet jet pump grapple or controll)lade control blade latching tool,  tOOl, to either the monorail or frame mounted hoist, verify proper thread engagement/
~ :J w
engagement/size    size by ensuring there is no play in the connection prior to thread engagement of three (3) full tum&        turns. The correct tool and coupling thread size is 7116-14 UNC. AdditionallyAdditionally,, aa 112-13 112-13 UNC bolt IlOlt will wiI[ not fit into a a proper size tool (7/16-14
-11 L.~
{7/16-14 U    NC); thus, this check may be performed if UNC);
w
practical.
--09 W
practical. Failure Failure to detect detect mis-matched mis-matched thread sizes will significantly significantly reduce the the load load capacity callacity of of the the tool/hoist.
20 28 D IRM DETECTOR LOCATION ZSIO SRM DETECTOR LOCATION CORE CORE IRM LOCATION IRM LOCATION A
tOOl/hoist.
12*41 E
4.41 4.41    Indication Indication of criticality observed observed on the SRM    SRM indicators indicators during during functional, subcritical, subcritical, or or shutdown shutdown marginmargin rod rod checks checks shall shall be be reason reason to to teminate temlinate fuel loading loading until until aa complete evaluation evaluation of of the the cause of oftM    criticality indication the criticality    indication is is determined.
28*25 B
determined.
36-41 F
OFH-1 1 IOFH-11                                              Rev.
20*25 C
Rev. 93 93                                      Page90f55 Page    9 of 55    I
20*33 G
36*09 0
28.33 H
12*09 Rev. 6 Page490f61 I Tier / Group: T3 SRORating:


FIGURE 09.1-FIGURE        09.1-22 IN-Core Instrumentation IN-Core    Instrurnenation Location Location Diagram Diagram 00 PLANT ~
===3.4 Source===
PLANT NORTH NORTH ,..".-
BANK Category 8:
r                                  n r                                                      h r!,                     /'"
Y
                                                        ~
: 96. With Unit Two operating at power, Annunciator RCIC TURBINE STM LINE DRN POT LEVEL HI alarms and the RO observes the E51-F054, F025, & F026 indicate closed on Panel P601.
r LA..J
Which one of the following identifies the cause of the above indications and the operability status of RCIC?
                                                  ;; l/            B,-l            --41
(Reference provided)
(, ~
These valves are closed due to loss of (1) and after taking the appropriate actions in the annunciator procedure the system would be declared Inoperable and(but)
                          ~~      ~            ffr                                    -33 4:f.~ '( -:;          t:l.f; ~                    -25
(2)
                                      ~~                                                -11 L                    ~ :J                                        w rHl L    I                                    L.~
A.
w    --09 L                                W I
(1) pneumatics (2) Unavailable B (1) pneumatics (2) Available C. (1) DC Power (2) Unavailable D. (1) DC Power (2) Available
12          20            28 D
: 96. With Unit Two operating at power, Annunciator RCIC TURBINE STM LINE DRN POT LEVEL HI alarms and the RO observes the E51-F054, F025, & F026 indicate closed on Panel P601.
LI      IRM DETECTOR LOCATION ZSIO  SRM DETECTOR LOCATION CORE                                CORE                          CORE SRM  LOCATION                  RM IRM      LOCATION                IRM    LOCATION A   12.33                        A      12.41 12*41                    E    28.25 28*25 a
Which one of the following identifies the cause of the above indications and the operability status of RCIC?
B   2641 28-41                        BB    36-41                    FF  20-26 20*25 C  36-26 36-25                        C     20-33 20*33                    G    36-09 36*09 0
0  20.17 20*11                        0      283.3 28.33                    H    12-09 12*09 ISD-091 SD-09.1                                       Rev.
Rev. 66                                  Page 49 of 61 Page490f61    I Categories K/A:
KIA:          SG2.O1.42 SG2.01.42                                              Tier/Group:
Tier / Group:  T3 T3 RORating:
RO Rating:    2.5 2.5                                                    SRO SRORating:
Rating:    3.4 3.4 LP Obj:
LPObj:                                                              Source:
Source:        BANK BANK Cog Cog Level:
Level:  HIGH HIGH                                                  Category 8:
Category  8:  YY
: 96. With Unit
: 96. With   Unit Two Two operating operating atat power, power, Annunciator Annunciator RCIC RCIC TURBINE TURBINE STMSTM LINE LINE DRN DRN POT POT LEVEL HI LEVEL         alarms and HI alarms     and the the RO RO observes observes the the E51-F054, E51-F054, F025, F025, && F026 F026 indicate indicate closed closed on Panel on   Panel P601.
P601.
Which one Which    one ofof the the following following identifies identifies the the cause cause of of the the above above indications indications and and the the operability status operability    status ofof RCIC?
RCIC?
(Reference provided)
(Reference provided)
(Reference      provided)
These valves are closed due to loss of (1) and after taking the appropriate actions in the annunciator procedure the system would be declared Inoperable and (but)
These valves These    valves areare closed closed due due to to loss loss of of (1) (1)    and after and         taking the after taking the appropriate appropriate actions in actions    in the the annunciator annunciator procedure procedure the the system system would would bebe declared declared Inoperable Inoperable and(but) and   (but)     (2) .
(2)
(2)
(1) pneumatics A. (1)   pneumatics (2)  Unavailable (2) Unavailable B (1) pneumatics B:-"
A. (1) pneumatics (2) Unavailable B:-" (1) pneumatics (2) Available C. (1) DC Power (2) Unavailable D. (1) DC Power (2) Available  
(2) Available C. (1) DC Power (2) Unavailable D. (1) DC Power (2) Available


Feedback Feedback K/A: SG2.02.15 KIA:    SG2.02.15 EQUIPMENT CONTROL EQUIPMENT         CONTROL Ability to Ability  to determine determine the  the expected expected plantplant configuration configuration using using design design andand configuration configuration control control documentation, such documentation,        such as as drawings, drawings, line-ups, line-ups, tag-outs, tag-outs, etc.
Feedback K/A: SG2.02.15 EQUIPMENT CONTROL Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc.
etc.
(CFR: 41.10 143.3/45.13)
(CFR: 41.10/43.3/45.13)
ROISRO Rating: 3.9/4.3 Objective: CLSLP016*15e
(CFR:   41.10 143.3/45.13)
: 15. Given plant conditions, predict the RCIC System response to the following conditions:
ROISRO Rating:
: s. Loss of instrument air.
RO/SRO      Rating: 3.9/4.3 3.9/4.3 CLSLP016*15e Objective: CLS-LP-016*15e Objective:
: e. DC power failure.
: 15. Given
: 15. Given plant plant conditions, conditions, predict predict the the RCIC   System response RCIC System    response toto the the following following conditions:
conditions:
: s. Loss
: s. Loss ofof instrument instrument air.
air.
: e. DC power
: e. DC    power failure.
failure.


==Reference:==
==Reference:==
2APP A-03 3-5, Revision 49, Page 44 Cog Level: High Explanation:
Valves fail closed on loss of DC power or Pneumatics, however with a loss of power, position indication on P601 will also be lost. Per APP A-03, 3-5
- If either E51-F025 or E51-F026 has been failed closed for more than 5 minutes, perform the following:
: a. Close Turbine Trip and Throttle Valve, E51 -V8, to prevent water hammer damage from a RCIC auto start.
: b. If RCIC must be started, proceed to OP-16.
this would still make RCIC available for use per the procedure but it is inoperable because it will not auto start as required.
This will make the RCIC system inoperable but available to be restarted per the procedure.
Distractor Analysis:
Choice A: Plausible because loss of pneumatics only is correct and the system will not start in auto when required, but could be manually started.
Choice B: Correct Answer Choice C: Plausible because a loss of power will cause valves to fail closed, but with loss of power position indication will be lost and the system will not start in auto when required.
Choice D: Plausible because pneumatics and power will cause valves to fail closed, but with loss of power position indication will be lost and it is available to start per the procedure which makes it available.
SRO Only Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, emergency conditions.
Notes Feedback KIA: SG2.02.15 EQUIPMENT CONTROL Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc.
(CFR: 41.10/43.3/45.13)
RO/SRO Rating: 3.9/4.3 Objective: CLS-LP-016*15e
: 15. Given plant conditions, predict the RCIC System response to the following conditions:
: s. Loss of instrument air.
: e. DC power failure.


==Reference:==
==Reference:==
 
2APP A-03 3-5, Revision 49, Page 44 Cog Level: High Explanation:
2APP A-03 2APP     A-03 3-5, 3-5, Revision Revision 49, 49, Page Page 44 44 Cog Level: High Explanation:
Valves fail closed on loss of DC power or Pneumatics, however with a loss of power, position indication on P601 will also be lost. Per APP A-03, 3-5 -If either E51-F025 or E51-F026 has been failed closed for more than 5 minutes, perform the following:
Explanation:
: a. Close Turbine Trip and Throttle Valve, E51-V8, to prevent water hammer damage from a RCIC auto start.
Valves fail closed on loss of DC    DC power or Pneumatics, however with a loss      loss of power, position indication on P601 will also be lost. Per APP A-03, 3-5 -IfIf either E51-F025 or E51-F026 has been failed closed for more than 5 minutes, perform the following:
: a. Close Turbine Trip and Throttle Valve, E51-V8,  E51 -V8, to prevent water hammer damage from a RCIC auto start.
: b. If RCIC must be started, proceed to OP-16.
: b. If RCIC must be started, proceed to OP-16.
this would still make RCIC available for use per the procedure but it is inoperable because it will not auto start as required.
this would still make RCIC available for use per the procedure but it is inoperable because it will not auto start as required.
Line 2,108: Line 2,936:
Choice A: Plausible because loss of pneumatics only is correct and the system will not start in auto when required, but could be manually started.
Choice A: Plausible because loss of pneumatics only is correct and the system will not start in auto when required, but could be manually started.
Choice B: Correct Answer Choice C: Plausible because a loss of power will cause valves to fail closed, but with loss of power position indication will be lost and the system will not start in auto when required.
Choice B: Correct Answer Choice C: Plausible because a loss of power will cause valves to fail closed, but with loss of power position indication will be lost and the system will not start in auto when required.
Choice D: Plausible because pneumatics and power          power will cause valves to to fail closed, but with loss loss of of power power position position indication indication will be be lost and itit is is available available to start per the procedure procedure which makes makes itit available.
Choice D: Plausible because pneumatics and power will cause valves to fail closed, but with loss of power position indication will be lost and it is available to start per the procedure which makes it available.
available.
SRO Only Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, emergency conditions.
SRO SRO Only      Basis: Assessment Only Basis:     Assessment of  of facility facility conditions conditions and and selection   of appropriate procedures selection of                procedures during during normal, normal, abnormal, abnormal, emergency emergency conditions.
Notes  
conditions.
Notes Notes


RCIC STEAM POT Partial P&ID QRAIN!
RCIC STEAM POT Partial P&ID RCIC STEAM POT Partial P&ID
          ~I
~
    .f.&#xa3;lI .
QRAIN!
F04S
I  
.f.&#xa3;lI.
F04S  


Unit Unit 22 APP APP A-03 A-03 3-5 3-S Page Page 22 ofof 22 (Cont a)
Unit 2
ACTIONS (Cont'd)
APP A-03 3-S Page 2 of 2
ACTIONS CAUTION CAUTION If Main Steam If r~in  Steam Line Line Drain Drain v~v, Vlv, MVD-F021, NVD-F021, fails fails toto close, close, then then the the Main Main Steam Steam Line Drain Line   Drain Inboard Inboard and and Outboard Outboard Isolation Isolation valves valves must must bebe closed.
ACTIONS (Cont a)
closed.
CAUTION If Main Steam Line Drain Vlv, NVD-F021, fails to close, then the Main Steam Line Drain Inboard and Outboard Isolation valves must be closed.
: 6. If required,
: 6. If reguii-ed, then close Main Steam Line Drain Inbd lad
: 6. If   reguii-ed, then then close close [*lain Main Stearn Steam Line Line Drain Drain Inbd Inbd Iso1 lad V1v, Vlv, B21-F0l6, and B21-F016,            Main Steam and Main    Steam LineLine Drain Drain Otbd Otbd Iso1 Isol V1v, Vlv, B21-F019.
: Vlv, B21-F0l6, and Main Steam Line Drain Otbd Isol Vlv, B2l-FOl.
B2l-FOl.
: 7. If alarm fails to clear within five minutes after completion of actions 1 2,3, 5, or 6, then dispatch an Auxiliary Operator to the Drwell access roof to determine if the HPCI/RCIC Condensate Drain Line Back Pressure Orifice is plugged or the drain line isolated.
: 7. If
NOTE Greater than 500 psig on HPCI/RCIC Back Pressure Orifice Inlet Pressure Gauge, 2-MVD-PI-7146 would be an indication of a plugged orifice.
: 7. If alarm alarm fails fails to to clear clear within within five five minutes minutes after after completion completion of  of actions 11 2,3, actions        2,3, 5,5, oror 6, 6, then then dispatch dispatch an an Auxiliary Auxiliary Operator Operator to to thethe Drwell access Drywe11    access roofroof to to determine determine if if the the HPCI/RCIC HPCI/RCIC Condensate Condensate Drain Drain Line Back Line   Back Pressure Pressure Orifice Orifice is  is plugged plugged oror the the drain drain line line isolated.
8.
isolated.
IF back pressure orifice is plugged, a,
NOTE NOTB:    Greater than 500 psig on Greater                          on HPCI/RCIC BackBack Pressure Orifice Inlet   Inlet Pressure    Gauge, 2-1*lYD-PI-7146 Pressure Gauge,         2-MVD-PI-7146 would would be be an an indication indication of aa plugged orifice.
Open HPCI/RCIC (ond Drn Line Back Press Orifice Bypass Valve, 2-MVD-VSOO2.
orifice.
b.
: 8. IF back pressure orifice is plugged, a,
Close BPCI/RCIC Cond Drn Line Back Press Orifice Inlet Isol
: a. Open HPCI/RCIC Cond    (ond Drn Line Back Press Orifice Bx~ss    Bypass Valve, 2-MVD-VSOO2.
: Valve, 2-MVD-VS000.
2-['lVD-V5002.
c.
: b.              BPCI/RCIC Cond Drn Line Back Press Orifice Inlet Iso1 Close HPCI/RCIC                                                              Isol Valve, 2-t-1VD-V5000.
Close HPCI/RCIC Cond Drn Line Back Press Orifice Outlet Isol
2-MVD-VS000.
: Valve, 2-MVD-VSOOl.
: c. Close HPCI/RCIC Cond Drn Line Back Press Orifice Outlet Isol Valve, 2-MVD-VSO 2-t-r.rD-V5001.
d.
Ol.
Place valves under proper administrative control.
: d. Place valves val,,'es under proper adrninistrat administrativ  ive e control.
9.
: 9. If NPCI/RCIC HPCI/RCIC Cond Drain Line is isolateth   isolated:
If NPCI/RCIC Cond Drain Line is isolateth a.
: a. Open HPCI/RIC HPCI/RCIC (ondCond Drn Line Back Press Orifice Inlet Isol Valve, 2-MVD-VS0 2-M'~~-V5000. 00.
Open HPCI/RIC (ond Drn Line Back Press Orifice Inlet Isol
: b. Open HPCI/RCIC HPCI/RCIC Cond Drn Line Back Press Orifice      Orifice Outlet Isol Valve, 2-MVD-VSO 2-M'.rD-V5001.Ol.
: Valve, 2-MVD-VS000.
b.
Open HPCI/RCIC Cond Drn Line Back Press Orifice Outlet Isol
: Valve, 2-MVD-VSOOl.
10.
10.
: 10. IF IF a circuit circuit malfunction malfm1ction is    is suspected, suspected, ensure ensure thatthat a WE/JO WR/JO is is prepared.
IF a circuit malfunction is suspected, ensure that a WE/JO is prepared.
prepared.
DEVICE/SETPOINTS Level Switch ES1-LSH-NO1O-l Instrument failure in the Switch Point tl dry condition/1980 my.
DEVICE/SE  TPOINTS D~vlCE/SETPOINTS Level Level Switch Switch ES1-LSH-NO E51-LSH-NOIO-l 1O-l                Instrument failure in Instrument                in the the Switch Switch Point Pc,int tl
Level Switch ESl-LSH-NOl0-l Also detects instrument failure in the Switch Point #2/0 +/- 2 wet condition.
                        #1                                  dry condition/19 dry  condition/1980     mV.
Incox-porates water.
80 my.
100 sec time delay in annunciator circuitry.
Level Level Switch Switch ESl-LSH-NO E51-LSH-NOIO-l   l0-l              .~lso detects Also   detects instrument instrument failure fa.ilure in ill the Switch the   Switch Point     #2/0" +/-+/- 2 Point #2/0          2"              wet condition.
POSSIBLE PLANT EFFECTS Damage to the RCIC turbine due to high moisture carryover on the steam.
wet                  Inc01~orates condition. Incox-porate        s water.
REFERENCES 1.
water.                                                100 sec 100  sec time time delay delay in    ~1nunciator in annunciator circuitry.
LL-9364 50 2.
circuitry.
OP-iS, RCIC System Operating Procedure 2APP-A-03 Rev. 49 Page 45 of 102 Unit 2 APP A-03 3-5 Page 2 of 2 ACTIONS (Cont'd)
POSSIBLE POSSIBLE PLANT PLANT EFFECTS EFFECTS Damage Damage to to the the RCIC RCIC turbine turbine due due to to high high moisture moisture carryover carryover on on the the steam.
CAUTION If r~in Steam Line Drain v~v, MVD-F021, fails to close, then the Main Steam Line Drain Inboard and Outboard Isolation valves must be closed.
steam.
: 6. If required, then close [*lain Stearn Line Drain Inbd Iso1 V1v, B21-F016, and Main Steam Line Drain Otbd Iso1 V1v, B21-F019.
REFERENC REFERENCESES 1.
: 7. If alarm fails to clear within five minutes after completion of actions 1 2,3, 5, or 6, then dispatch an Auxiliary Operator to the Drywe11 access roof to determine if the HPCI/RCIC Condensate Drain Line Back Pressure Orifice is plugged or the drain line isolated.
: 1. LL-9364 LL-9364 - 50 50 2.
NOTB:
: 2. OP-iS, OP-16, RCIC RCIC System System Operating Opera.t ing Procedure Procedure 2APP-A-03 I2APP-A-03                                         Rev.
Greater than 500 psig on HPCI/RCIC Back Pressure Orifice Inlet Pressure Gauge, 2-1*lYD-PI-7146 would be an indication of a plugged orifice.
Rev. 49 49                              Page 45 Page   45 ofof 102
: 8.
                                                                                                          '1021
IF back pressure orifice is plugged,
: a.
Open HPCI/RCIC Cond Drn Line Back Press Orifice Bx~ss Valve, 2-['lVD-V5002.
: b.
Close HPCI/RCIC Cond Drn Line Back Press Orifice Inlet Iso1 Valve, 2-t-1VD-V5000.
: c.
Close HPCI/RCIC Cond Drn Line Back Press Orifice Outlet Isol Valve, 2-t-r.rD-V5001.
: d.
Place val,,'es under proper adrninistrat ive control.
: 9.
If HPCI/RCIC Cond Drain Line is isolated:
: a.
Open HPCI/RCIC Cond Drn Line Back Press Orifice Inlet Isol
: Valve, 2-M'~~-V5000.
: b.
Open HPCI/RCIC Cond Drn Line Back Press Orifice Outlet Isol Valve, 2-M'.rD-V5001.
: 10. IF a circuit malfm1ction is suspected, ensure that a WR/JO is prepared.
D~vlCE/SETPOINTS Level Switch E51-LSH-NOIO-l Switch Pc,int #1 Level Switch E51-LSH-NOIO-l the Switch Point #2/0" +/- 2" water.
POSSIBLE PLANT EFFECTS Instrument failure in the dry condition/1980 mV.
.~lso detects instrument fa.ilure ill wet condition.
Inc01~orates 100 sec time delay in ~1nunciator circuitry.
Damage to the RCIC turbine due to high moisture carryover on the steam.
REFERENCES
: 1.
LL-9364 - 50
: 2.
OP-16, RCIC System Opera.t ing Procedure I2APP-A-03 Rev. 49 Page 45 of '1021  


Unit Unit 22 APP A-03 3-S APP "~.-03  3-S Page Page 11 of of 22 RCIC RCIC      TIJREINE STI-l TURBINE  STh LINE LINE DRN DRN POT POT LEVEL LEVEL HI HI RCIC iRCIC      Turbine  Steam Line Turbine Steam       Line Water Water Drain Drain Pot Pot High High L=vel)
Unit 2
Level)
APP A-03 3-S Page 1 of 2
AUTO
RCIC TIJREINE STh LINE DRN POT LEVEL HI RCIC Turbine Steam Line Water Drain Pot High Level)
  "~UTO    ACTIONS ACTIONS 1.
AUTO ACTIONS 1.
: 1. Supply Drain Supply   Drain Pot Pot Drain Drain Bypass Bypass Valve, Valve, ESI-F054, ES1-F054, opens.
Supply Drain Pot Drain Bypass Valve, ES1-F054, opens.
opens.
CATJSE 1.
CCATJSE
Heavy condensate load dul-ing steam line warmup.
    ....USE 1.
: 1. Heavy condensate Heavy   condensate load load during dul-ing steam steam line line warmup.
warmup.
2.
2.
: 2. Normal orifioe Normal   orifice clogged.
Normal orifice clogged.
clogged.
3, HPCIJRCIC ond.
3,
Drain Line Back Pressure Orifice is plugged.
: 3. HPCIJRCIC Cond.
4.
HPCI!RCIC    ond. Drain Drain Line Line Baok Back Pressure Pressure Orifioe Orifice is is plugged.
Drain line isolation valves to main condenser closed.
plugged.
: 4. Drain line Drain         isolation valves line isolation     valves to main oondenser condenser olosed.
closed.
5.
5.
S. Drain pot level Drain        level instrument instrument failure failure oror loss loss of instrument instrument, po;;er.
Drain pot level instrument failure or loss of instrument, power.
power.
6.
6.
: 6. Circuit malfunction.
Circuit malfunction.
Cirouit OBSERVATIONS 1.
OBSERVATIONS 1.
: 1. RCIC Supply Drain Pot Drain Byp Valve, ES1-F054,    ESl-F054, opened.
RCIC Supply Drain Pot Drain Byp Valve, ESl-F054, opened.
NOTE:
NOTE:
NOTE:    If alarm ooours If          occurs and the E51-FOS4 ES1-F054 valve does not automatioally automatically open, the the  most   probable   cause is instrument failure or loss of instrument oause power (Panel 2B-Rx "HIO"  PH1O1T CKT 14/.
If alarm occurs and the ES1-F054 valve does not automatically open, the most probable cause is instrument failure or loss of instrument power (Panel 2B-Rx PH1O1T CKT 14),
14),
NOTE:
NOTE:                       indications are available inside the level element Additional LED indioations control box device H5E    HSE (P&#xa3; (RB 20' 20 elevation) as follo;;s:
Additional LED indications are available inside the level element control box device HSE (RB 20 elevation) as follows:
follows:
Normal status No annunciator No LEDs illuminated High Water level Green LED on Instrument failure Red LED on ACTIONS 1.
Normal Normal status       No annunciator                 No LEDs illuminated High Water level                 Green LED on Instrument failure             Red LED on011.
Ensure Supply Drain Pot Drain Byp Vlv, ESl-F054, is open.
ACTIONS ACTIONS 1.
: 1. Ensure   Supply Drain Pot Drain Byp Vlv,    "lIlv, ESl-F054, ESI-FOS4, is open.
2.
2.
: 2. Ensure Ensure    RCIC RCIC Supp Supp Pot Pot Inbd IOOd Isolation Isolation Valve, Valve, ESi-F025, ESI-F025, is  is open.
Ensure RCIC Supp Pot Inbd Isolation Valve, ESi-F025, is open.
open.
3.
3.
: 3. Ensure Thlsure  RCIC RCIC Supp Supp Pot Pot Outbd Outbd Isolation Isolation Valve, Valve, ES1-F026, ESI-F026, is  is open.
Ensure RCIC Supp Pot Outbd Isolation Valve, ES1-F026, is open.
open.
NOTE:
NOTE:
NOTE:  Valves Valves E51-F025 E51-F025 andand E51-F026 ESI-F026 will will close close on on loss loss ofof instrument instrument airair and   will also and will    also close close if if E5l-F045 ESI-F045 is is not not fully fully closed.
Valves E51-F025 and E51-F026 will close on loss of instrument air and will also close if E5l-F045 is not fully closed.
olosed. Valves Valves ES1-F025 ESI-F025 and and E51-F026 E51-F026 cannot oannot bebe opened opened in in either either of of these these conditions.
Valves ES1-F025 and E51-F026 cannot be opened in either of these conditions.
conditions.
4.
4.
: 4. If If either either ESl-Ft25 ESI-F025 oror ESl-F026 ES1-F026 ha has been been failed failed closed olosed forfor more more than than 55 minutes, minutes, perform perform the the following:
If either ESl-Ft25 or ESl-F026 ha been failed closed for more than 5 minutes, perform the following:
following:
a.
a.
: a. Close Close Turbine Turbine Trip Trip and and Throttle Throttle Valve, Valve, ESl-V8, ESI-V8, to to prevent prevent water water hammer hammer damage damage frcm from aa RCIC RCIC auto auto start.
Close Turbine Trip and Throttle Valve, ESl-V8, to prevent water hammer damage frcm a RCIC auto start.
start.
b.
b.
: b. If If RCIC RCIC must must hebe started, started, proceed proceed toto OP-16.
If RCIC must he started, proceed to OP-16.
OP-16.
S.
S.
S. Ensure Ensure Main Main Steam Steam Drain Drain Lme Line Vlv, Vlv, MVD-F02l, MVD-F02l, is is cload.
Ensure Main Steam Drain Lme Vlv, MVD-F02l, is cload.
olosed.
2APP-A-03 Rev. 49 Page 44 of 102 Unit 2 APP "~.-03 3-S Page 1 of 2 RCIC TURBINE STI-l LINE DRN POT LEVEL HI iRCIC Turbine Steam Line Water Drain Pot High L=vel)
2APP-A-03 I2APp-A-03                                         Rev.
"~UTO ACTIONS C.... USE
Rev. 4949                              Page 44 Page      of '1021 44 of  102
: 1.
Supply Drain Pot Drain Bypass Valve, ESI-F054, opens.
: 1.
Heavy condensate load during steam line warmup.
: 2.
Normal orifioe clogged.
: 3.
HPCI!RCIC Cond. Drain Line Baok Pressure Orifioe is plugged.
: 4.
Drain line isolation valves to main oondenser olosed.
S.
Drain pot level instrument failure or loss of instrument po;;er.
: 6.
Cirouit malfunction.
OBSERVATIONS
: 1.
RCIC Supply Drain Pot Drain Byp Valve, ES1-F054, opened.
NOTE:
If alarm ooours and the E51-FOS4 valve does not automatioally open, the most probable oause is instrument failure or loss of instrument power (Panel 2B-Rx "HIO" CKT 14/.
NOTE:
Additional LED indioations are available inside the level element control box device H5E (P&#xa3; 20' elevation) as follo;;s:
Normal status No annunciator No LEDs illuminated Green LED on Red LED 011.
ACTIONS High Water level Instrument failure
: 1.
Ensure Supply Drain Pot Drain Byp "lIlv, ESI-FOS4, is open.
: 2.
Ensure RCIC Supp Pot IOOd Isolation Valve, ESI-F025, is open.
: 3.
Thlsure RCIC Supp Pot Outbd Isolation Valve, ESI-F026, is open.
NOTE:
Valves E51-F025 and ESI-F026 will close on loss of instrument air and will also close if ESI-F045 is not fully olosed.
Valves ESI-F025 and E51-F026 oannot be opened in either of these conditions.
: 4.
If either ESI-F025 or ES1-F026 has been failed olosed for more than 5 minutes, perform the following:
: a.
Close Turbine Trip and Throttle Valve, ESI-V8, to prevent water hammer damage from a RCIC auto start.
: b.
If RCIC must be started, proceed to OP-16.
S.
Ensure Main Steam Drain Line Vlv, MVD-F02l, is olosed.
I2APp-A-03 Rev. 49 Page 44 of '1021  


8.4 8.4       Isolating the Isolating     the RCIC RCIC System System Stearn Steam Supply Supply                                          RR Rijlefijr R3r9rr Use Uae 8.4.1 8.4.1          Initial Conditions Initial  Conditions 1.
I IF rapid isolation is NOT desired, THEN PERFORM the following to isolate and de-pressurize the RCIC steam supply line:
: 1.      All applicable All  applicable prerequisites prerequisites listed  in listed in Section Section 4.0 4.0 are are met.
a.
met.      o 8.4.2 8.4.2         Procedural Steps Procedural      Steps 1.
CLOSE STEAM SUPPLY INBOARD 1SOL VLV, E51-F007.
: 1.      IF rapid IF   rapid isolation isolation of of RCIC steam line RCIC steam    line is is desired, desired, THEN THEN PERFORM the PERFORM              following:
b.
the following:
OPEN HPCLRCIC COND DRN LINE SACK PRESS ORiFiCE BYPASS VALVE MVD-V5002.
c.
OPEN TURBiNE STEAM SUPPLY VL E1-FO45, AND MONITOR turbine response.
d.
CLOSE SUPPLYDRAIN POT1N8D DRAIN VLV, E1-FO25.
e.
CLOSE SUPPLY DRAiN POT OTBD DRAIN VLV, E51-FO26.
20P-16 Rev. 107 Page 34 oi1 R
R3r9rr Uae 8.4 Isolating the RCIC System Steam Supply 8.4.1 Initial Conditions 1.
All applicable prerequisites listed in Section 4.0 are met.
8.4.2 Procedural Steps 1.
IF rapid isolation of RCIC steam line is desired, THEN PERFORM the following:
a.
a.
: a.       CLOSE STEAM CLOSE E1-FOO7.
CLOSE STEAM SUPPLYIWBOARD ISOL VLV, E1-FOO7.
b.
CLOSE STEAMSUPPLYOUTBOARDISOL VLV, E1-FOO8.
8.4 Isolating the RCIC System Stearn Supply 8.4.1 Initial Conditions
: 1.
All applicable prerequisites listed in Section 4.0 are met.
8.4.2 Procedural Steps
: 1.
IF rapid isolation of RCIC steam line is desired, THEN PERFORM the following:
: a.
: b.
CLOSE STEAM SUPPL Y INBOARD /SOL VL V.
E51-F007.
E51-F007.
STEAM SUPPL SUPPLYIWBOARD Y INBOARD /SOL  ISOL VL V.
CLOSE STEAM SUPPL Y OUTBOARD ISOL VL 11, E5'1-FOOB.
VLV,       o b.
CAUTION o
: b.      CLOSE STEAM CLOSE E1-FOO8.
o o
E5'1-FOOB.
Opening the TURBINE STEAM SUPPL Y VL V. E51-F045, to de-pressurize the RCIC steam line will roU the RCIC turbrne,
STEAMSUPPLYOUTBOARDISOL SUPPL Y OUTBOARD ISOL VL            11, VLV,      o CAUTION Opening the TURBINE STEAM SUPPL Y VL V. E51-F045, to de-pressurize the RCIC steam line will roU the RCIC turbrne, I
: 2.
: 2.     IF rapid isolation is NOT desired, THEN PERFORM the following to isolate and de-pressurize the RCIC    RCIC steam supply line:
IF rapid isolation is NOT desired, THEN PERFORM the following to isolate and de-pressurize the RCIC steam supply line:
: a.       CLOSE STEAM SUPPL   SUPPLY  Y INBOARD ISOL 1SOL VL V.
: a.
VLV,        0 E51-F007.
CLOSE STEAM SUPPL Y INBOARD ISOL VL V.
E51-FOO7,
0 E51-FOO7,
: b.       OPEN HPCLRCIC HPCllf?CIC COND DRN LINE SACK      BACK               0 ORiFiCE BYPASS VALVE, PRESS ORIFICE                    VALVE MVD-V5002.
: b.
: c.               TURBiNE STEAM SUPPLY OPEN TURBINE                   SUPPL Y VL V,                   0 E1-FO45, E51-F045, AND MONITOR turbine response.
OPEN HPCllf?CIC COND DRN LINE BACK 0
: d.       CLOSE SUPPLY SUPPLYDRAIN DRAIN POT1N8D POT lNBD DRAIN VLV;   VLV,      0 E1-FO25.
PRESS ORIFICE BYPASS VALVE, MVD-V5002.
: c.
OPEN TURBINE STEAM SUPPL Y VL V, 0
E51-F045, AND MONITOR turbine response.
: d.
CLOSE SUPPLY DRAIN POT lNBD DRAIN VLV; 0
E51-F025.
E51-F025.
e.
: e.
: e.      CLOSE SUPPLY DRAiN  DRAIN POT OTBD  OTaD DRAIN VLV, VL1I,     0 E51-FO26.
CLOSE SUPPLY DRAIN POT OTaD DRAIN VL1I, 0
E51-F026.
E51-F026.
20P-16 12oP-16                                         Rev.
R Rijlefijr Use 12oP-16 Rev. 107 Page 34 of 891  
Rev. 107 107                              Page     oi1 34 of Page 34    891


8.12 8.12    Controlled Manual Controlled   Manual Start Start of of the the RCIC RCIC System System With With Turbine Turbine Steam Steam Line Line        RR Reference Refrence Drain Pot Drain  Pot High High Level Level oror RCIC RCIC Pump Pump LowLow Discharge Discharge Pressure Pressure                Use Indicated Indicated 8.12.1 8.12.1       Initial Conditions Initial  Conditions
8.12 Controlled Manual Start of the RCIC System With Turbine Steam Line R
: 1.     IF RCIC IF  RCIC is   being operated for aa planned is being                  planned evolution evolution (non-(non       o emergency operation), THEN THEN Health Health Physics Physics (HPs)
Drain Pot High Level or RCIC Pump Low Discharge Pressure Refrence Indicated 8.12.1 Initial Conditions 1.
(HPs) shall shall be notified be  notified to attend  the pre~ob attend the pre-job briefing AND aa log briefing AND    log entry entry made to identify the individual made                    individual contacted.
IF RCIC is being operated for a planned evolution (non emergency operation), THEN Health Physics (HPs) shall be notified to attend the pre-job briefing AND a log entry made to identify the individual contacted.
: 2. One of the follol;1,1ng following conditions exist exist:
2.
: a.       The RCIC turbine has been shutdown or tripped and annunciator RCICRC!C TURBINE STM LINE DRN o
One of the following conditions exist:
POT LEVEL HI (A-03 3-5) sealed in.
a.
: b.       The RCIC turbine has been shutdown or tripped and the RCIC PUMP DISCH  DISH PRESS LOW o
The RCIC turbine has been shutdown or tripped and annunciator RC!C TURBINE STM LINE DRN POT LEVEL HI (A-03 3-5) sealed in.
annunciator (A-02, 1-6) is sealed in.
b.
: 3. A controlled manual start of RCIC is desired.                         o 8.12.2       Procedural Steps CAUTION The RCIC turbine has the potential for failures that could cause personnel injuries. The potential is most significant when the system is initially started after control system maintenance, or after an extended period of being idle. Announcing turbine starts and clearing of all personnel from the RCIC area are required during this period of risk.
The RCIC turbine has been shutdown or tripped and the RCIC PUMP DISH PRESS LOW annunciator (A-02, 1-6) is sealed in.
3.
A controlled manual start of RCIC is desired.
8.12.2 Procedural Steps CAUTION The RCIC turbine has the potential for failures that could cause personnel injuries. The potential is most significant when the system is initially started after control system maintenance, or after an extended period of being idle. Announcing turbine starts and clearing of all personnel from the RCIC area are required during this period of risk.
Permission to access this area during initial RCIC roll requires the approval of the Unit sco.
1.
EVACUATE all personnel f 0111 he RCIC turbine area.
20P-16 Rev. 107 Page 52 of 89 8.12 Controlled Manual Start of the RCIC System With Turbine Steam Line Drain Pot High Level or RCIC Pump Low Discharge Pressure Indicated 8.12.1 Initial Conditions IF RCIC is being operated for a planned evolution (non-emergency operation), THEN Health Physics (HPs) shall be notified to attend the pre~ob briefing AND a log entry made to identify the individual contacted.
o
: 2.
One of the follol;1,1ng conditions exist
: 3.
8.12.2
: a.
: b.
The RCIC turbine has been shutdown or tripped and annunciator RCIC TURBINE STM LINE DRN POT LEVEL HI (A-03 3-5) sealed in.
The RCIC turbine has been shutdown or tripped and the RCIC PUMP DISCH PRESS LOW annunciator (A-02, 1-6) is sealed in.
A controlled manual start of RCIC is desired.
Procedural Steps CAUTION o
o o
The RCIC turbine has the potential for failures that could cause personnel injuries. The potential is most significant when the system is initially started after control system maintenance, or after an extended period of being idle. Announcing turbine starts and clearing of all personnel from the RCIC area are required during this period of risk.
Permission to access this area during initial RCIC roll requires the approval of the Unit SCO.
Permission to access this area during initial RCIC roll requires the approval of the Unit SCO.
sco.
EVACUATE all personnel from the RCIC turbine area.
: 1. EVACUATE all personnel ffrom    0111 the he RCIC turbine area.           0 20P-16 120p--.16                                     Rev. '107 Rev. 107                              Page 52 Page 52 of 891 of 89
0 R
Reference Use 120p--.16 Rev. '107 Page 52 of 891  


ATTACHMENT ATTACHMENT2A                2A Page Page5of 5 of 30  30 PANEL 4A                                 LOCATION                                         NORMAL NORMAL SUPPLY   SUPPLY Switchboard 21\
ATTACHMENT 2A Page5of 30 PANEL 4A LOCATION NORMAL SUPPLY Reference Drawing LL-3024-6 Control Building 49 ft East Switchboard 2A CIRCUIT LOAD EFFECT Rx.Annunciator Logic, 2-H12-P630 1.
Control Building 49 ft East Reference Reference Drawin Drawing     LL-3024-6                                                                 Switchboard       2A CIRCUIT CIRCUIT                        LOAD lOAD                                                                                EFFECT EFFECT 1      Rx.Annunciator R>:.                Logic, 2-H12-P630 Annunciator logic,    2-H12-P630      i.1. Auto   transfers toto alternate Auto transiers          alternatesoursource, ce, Panel4B PaneI4B circuit circuit 11.
Auto transfers to alternatesource, PaneI4B circuit 1.
Panels 60 Panels    80t/503
Panels 80t/503 2.
                      !fe03                            2. Receive
Receive annunciator A5-&.
: 2. Receive annunciator annunciator Ae-5-S.
2 HPCI Flow controler 1.
A5-&.
Controller fails downscale.
22      HPCI Flow HPCI   Flow controller controler                   [11. Controller Controller fails fails downsca.ie.
E41-FIC-K800 (24 VDC;.
downscale.
2.
E41-FIC-K800 (24 E41-FIC-K600      (24\lDC)
Loss of f;ow ndicatcn.
VDC;.                 2.
: 2. Loss of loss    offlo'l/
f;ow indication.
ndicatcn.
3.
3.
: 3. Receive annunciator Reoeive      annunciator AA1-2-5. 1-2-5.
Receive annunciator A 1-2-5.
4.
4.
: 4. Loss of loss    of HPCr HPCI 5.
Loss of HPCI 5.
: 5. Loss of loss    of ASSD ASSD function.
Loss of ASSD function.
function.
HPC1 Supervisory Lights 1.
HPC1 Super~'isory HPCI    Supervisory Lights Lights                1.1. Loss of loss    of E41-1/S E41-VS and       E4l-V indication and E41-V9      indicaticn 2.
Loss of E41-VS and E4l-V indicaticn 2.
: 2. Loss HPCI loss    HPCI oiloil lank tank le ... el HillO level   HilLO alam  *.
Loss HPCI oil tank level HilLO alarm.
alarm.
HPCI Vertical Board meters 1.
HPCI  Vertical Board HPCI Vertical   Board meters meters            [1.1. Loss of loss        pressure transmitters/meters of pressure      transmitters/meters Re01,        R32, Re~3, R6OI R602.      R503, R60!
Loss of pressure transmitters/meters R6OI R32, R503, R80(
R80(
VDC)
(52.5 VDC)
(52.5 VDC)
(52.5 HPCI Turbine HPCl  Turbine Speed Speed Control Control              1. loss
HPCI Turbine Speed Control 1.
: i. Loss ofof speed speed control, control, EGM 2GM and and speed soeed sensor.
Loss of speed control, 2GM and soeed sensor.
sensor.
2.
: 2. Loss of
Loss of speed indicaon on veitcal board.
: 2. loss    of speed speed indication indicaon on   on vertioal veitcal board.
E41-F053, E41-FOE4, E41-F026 1.
board.
Fail closed.
E41-F053, E41-F054.
2.
E41-F053,     E41-FOE4, E41-F026 E41-F026            1. Faif
E41-F054 and 241-F026 loss of indication.
: 1. Fail closed.
251-F006,E51-F025 1.
closed.
Failclosed.
: 2. E41-F054 and
2.
: 2. E41-F054         and E41-F026 241-F026 loss  loss ofindication.
Loss of indication, 001-50 Rev. 45 I
of indication.
Reference Drawin CIRCUIT lOAD 1
251-F006,E51-F025 E51-F005,   E51-F025                       1. Failclosed.
R>:. Annunciator logic, 2-H12-P630
: 1. Fail  clcsed.
: i.
: 2. loss
Panels 60 !fe03
: 2. Loss of     indication, of indication.
: 2.
001-50 1001-50                                                                  Rev.
2 HPCI Flow controller
Rev. 45 I
[1 E41-FIC-K600 (24 \\lDC)
: 2.
: 3.
: 4.
: 5.
HPCI Super~'isory Lights
: 1.
: 2.
HPCI Vertical Board meters
[1.
(52.5 VDC)
HPCl Turbine Speed Control
: i.
: 2.
E41-F053, E41-F054. E41-F026
: 1.
: 2.
E51-F005, E51-F025
: 1.
: 2.
1001-50 ATTACHMENT 2A Page 5 of 30 NORMAL SUPPLY Switchboard 21\\
EFFECT Auto transiers to alternate source, Panel4B circuit 1 Receive annunciator Ae-5-S.
Controller fails downsca.ie.
loss of flo'l/ indication.
Reoeive annunciator A 1-2-5.
loss of HPCr loss of ASSD function.
loss of E41-1/S and E41-V9 indication loss HPCI oil lank le... el HillO alam *.
loss of pressure transmitters/meters Re01, R602. Re~3, R60!
loss of speed control, EGM and speed sensor.
loss of speed indication on vertioal board.
Faif closed.
E41-F054 and E41-F026 loss ofindication.
Fail clcsed.
loss of indication.
Rev. 45  


ATTACHMENT ATTACHMENT 29                    25 Page Page 9        of 32 90132 PANEl:4B PANEL: 4B                                         LOCATION:
ATTACHMENT 25 Page 90132 PANEL: 4B LOCATION:
LOCATION:                                                        NORMAL SUt>,P:L NOBMAt              SUPPLY:   Y:            J Reference Reference Drawing:
NORMAL SUPPLY:
Drawing: lL..,3024-7 LL-3024-7           Contra!
Reference Drawing: LL-3024-7 Control Building 49 ft South Switchboard 2B Cku:,
Control Building Building 49   49 ftSouth ft South                                     .. Switchboard Switchboard 2B      2B Cku:,
LOAD EFFECT E
Ckt#                              LOAD LOAD                                                                                                 EFFECT EFFECT 5E          Recrc Pump Recirc                Auxiliary Equipment Pump 83 AuxillaI'I    EqtLpment       1.1. Loss Loss ofof alternate aJternate control           power to:
Recrc Pump 3 Auxiliary EqtLpment 1.
control power        to:
Loss of aJternate control power to:
Aternate Control Alternate    Control PeweI Power                         ** Recilc Recirc B  B Gen.
Aternate Control Power Recirc B Gen. Fielo Breaker, control, frp and indicaior.
Gen. FieldFielo Breaker, Breaker, oontre!,
Recirc B Scoop Thbe Power Failure Look & Reset Recirc Lube Oil Pumps S-i ano 8-2, control and ndication.
control, trip frp and and indication.
Reciro B Lock out lTrp Logic ATWS Trip Logic S 2.
indicaior.
Normal power is from Panel 1OA, ckt 2.
                                                                  ** Recilc Recirc BB Scoop Scoop Tube  Thbe Power Power Failure Failure Lock Look && Reset Reset
Backup Scram valve. 2-C12-1 lflS 1.
                                                                  ** Recife Recirc LubeLube Oil  Oil Pumps Pumps 8-1          ano B-2.
Backup Scram valve fails closed; Div I Backup Scram valve can still fun Div II Backup Scram Logic 1.
S-i and    8-2, control control andand indicatien.
Sc-ram Discharge Volume Vent and 2ran Vaves wil no receive a close valves will sti I function with Div I.
ndication.
2.
                                                                  ** Recifc Reciro BB lockLock out out ITrip lTrp logicLogic
DFWLCS wD rot receive auto set down from Div II. Digital eedwater w 2.
                                                                  ** AATWS TWS Trip Trip logic Logic 8S
Ten-second tme delay por to scram reset, will not function for B RPS 7
: 2. Normal power
Spare Spare RCIC Flow controller 1.
: 2. Normal       power is   is irom from Panel Panel lOA,  1OA, ckt ckt 3.2.
Controller sails cownecale.
t)        Backup Scram Backup                valve. 2-C12-i=11 Scram valve,    2-C12-1 lflS08      1.
E51 -FlC-KDO (24 VOC) 2.
: 1. Backup Backup ScramScram val'Jevalve fails fails dosed; closed; Div Div II Backup Backup ScramScram valve valve can can still still fun, fun Div IIII Backup Div                Scram Logic Backup Scram       Logic               1. Sc-ram Discharge
Loss of flow indication.
: 1. Sc*ram      Discharge Volume  Volume Vent    Vent and and Drain 2ran Va.lves Vaves will wil nOl no receive receive aa close close valves wilf lIallies  will still sti I function function wilhwith DillDiv I.I.
3.
: 2. DFWLCS w~1
Receive annunciator A3-6-5.
: 2. DFWLCS           wD not rot receive receive auto,  auto set set down down from     Div rI.
RCIC Supervisory Lights 1.
from Dill    II. Digital eedwater w Digital Feedwater      w
Loss of E51-V8 and 251-VQ ndic-ation.
: 2. Ten-second
RCIC Vertical Board meters 1.
: 3. Ten-second time    tme delav delay prior por to       scram reset, to scram      reset, will will not not function function for for B B RPS RPS !
Loss of pressure :ransmittersmeters ROOl, RO2, RD3, RCIJ4 on the P i2.5 VDC1 E51-F02& E51-FCD4. E51-F054 1.
77        Spare Spare                                                 Spare Spare 8          RCIC Flo'll RCIC     Flow contrcller controller                       1. Controller fails
Fail closed.
: 1. Conlroller              cownecale.
2.
sails downscale.
Loss of indicaon.
E51 -FlC-KDO (24 E5l-FIC-K800        (24 VDq VOC)                         Loss of
HPCI E41-FC25 1.
: 2. Loss     of flow flow indication.
indication.
: 3. Receive
: 3. Receive annunciatcr annunciator />'3-6-5.
A3-6-5.
RCIC Supervisory RCtC    Supervisory lights Lights                       Loss of
: 1. Loss
: 1.          of E51-V8 E51-V8 and E51-V9                  ndic-ation.
251-VQ indicatien.
RCIC Vertical Board meters RCIC                        meters                1. Loss
: 1. Loss of of pressure         :ransmittersmeters ReOl, pressure transmitters/meters                    ROOl, R602,RO2, R60S.
RD3, R804 RCIJ4 on thethe R P i2.5 VDC1 (52.5VDCi E51-F026.
E51-F02& E51-F004.
E51-FCD4. E51-F054               1. Fai.!closed.
Fail closed.
Fail closed.
: 2. Loss of of indication.
2.
indicaon.
Loss of indicaon.
HPCI E41-F025 HPCr    E41-FC25                                1. Fail closed.
RCIC 2GM 1.
Loss of speed control.
2.
Loss of speed indication on RIGS RCIC Initiation and Control Logic 1.
RCIC will not auto initiate. Cannot be manually operated.
2.
Receive annunciator A3-1-4.
2.
Mm flow valve will not auto open.
4.
Barometric condenser vacuum tank auto level control nop.
001-50 I
Rev. 45 r
PIPING HY QLlo PUi tC 1i4 13 tLJRB.VNDQR h1 REv)SL PLR C 6t164 55 R-Vt5fD PIR rc 64l2 A
PROGRESS ENERGY li-(MQRMATIOtI ON HS oNAw:1o CDMPI IFS WllI Ctk %l
 
tJtT 2 5i S/A 2-FP46 (l PO 729E48B SH 2)
REACIOR BL.i1DING RLACJOR CORE ISOLATION NOTE RLVISONS TO THIS ORAWING MUST ALSO 1E COOLiNG SYSII M P4CORPORATO ON lE CORRESPONDING ORAW14S:
PIPING I)IAG AM C) 1476, C) 0427, 0 04219, I) 04220 &
1) 04221 D&#xf8;229 PANEl:4B Reference Drawing: lL..,3024-7 Ckt#
LOAD 5
Recirc Pump 8 AuxillaI'I Equipment Alternate Control PeweI t)
Backup Scram valve, 2-C12-i=11 08 Div II Backup Scram Logic 7
Spare 8
RCIC Flo'll contrcller E5l-FIC-K800 (24 VDq RCtC Supervisory lights RCIC Vertical Board meters (52.5VDCi E51-F026. E51-F004. E51-F054 HPCr E41-F025 It RC1CEGM RCIC Initiation and Contrel Logic 1001-50 ATTACHMENT 29 Page 9 of 32 LOCATION:
J Contra! Building 49 ftSouth NOBMAt SUt>,P:L Y:
.. Switchboard 2B EFFECT
: 1. Loss of alternate control power to:
Recilc B Gen. Field Breaker, oontre!, trip and indication.
Recilc B Scoop Tube Power Failure Lock & Reset Recife Lube Oil Pumps 8-1 and B-2. control and indicatien.
Recifc B lock out ITrip logic A TWS Trip logic 8
: 2. Normal power is irom Panel lOA, ckt 3.
: 1. Backup Scram val'Je fails dosed; Div I Backup Scram valve can still fun,
: 1. Sc*ram Discharge Volume Vent and Drain Va.lves will nOl receive a close lIallies wilf still function wilh Dill I.
: 2. DFWLCS w~1 not receive auto, set down from Dill rI. Digital Feedwater w
: 3. Ten-second time delav prior to scram reset, will not function for B RPS !
Spare
: 1. Conlroller fails downscale.
: 2. Loss of flow indication.
: 3. Receive annunciatcr />'3-6-5.
: 1. Loss of E51-V8 and E51-V9 indicatien.
: 1. Loss of pressure transmitters/meters ReOl, R602, R60S. R804 on the R
: 1. Fai.!closed.
: 2. Loss of indication.
: 1. Fail closed.
: 2. Loss of indication.
: 2. Loss of indication.
indicaon.
: 1. Loss of speed control.
It        RCIC 2GM RC1CEGM                                          1. Loss of speed control.
: 2. Loss of speed indication on RTG8
: 2. Loss of speed indication on RTG8             RIGS Initiation and Contrel RCIC Initiation          Control Logic          1.           will noi RCIC 'liill   not auto iniliate.
: 1. RCIC 'liill noi auto iniliate. Cannot be manuati'l operated.
initiate. Cannot be manuati'l manually operated.
: 2. Receive annunciator />'3-14.
: 2. Receive annunciator />'3-14. A3-1-4.
2,_
2,_
: 2. Mm flow Min   flc'II valve will not auto open.      open.
Min flc'II valve will not auto open.
: 4. Barometric condenser lIacuum       vacuum tank auto level control        oontrol inop.
: 4. Barometric condenser lIacuum tank auto level oontrol inop.
nop.
Rev. 45 56 55  
001-50 1001-50                                                  I                          Rev. 45 r  ,
~~~'-L-------------~---~~~----------------;A C3 PROGRESS ENERGY  
PIPING HY tLJRB.VNDQR                                                                  -
REv)SL PLR C 6t164 56           -          QLlo PUi tC 1i4 13 h1 55 55                         R-Vt5fD PIR rc 64l2
                                                                                            ~~~'-L-------------~---~~~----------------;A                                      A C3 PROGRESS                                     ENERGY li-  (MQRMATIOtI ON HS oNAw:1o CDMPI IFS WllI                                                                        Ctk %l            tJtT 2    5i    S/A 2-FP46 (l PO 729E48B SH                                  & 2)                                          REACIOR BL.i1DING
                                                    -                                                  RLACJOR CORE ISOLATION NOTE    RLVISONS TO THIS ORAWING MUST ALSO 1E                                                                      COOLiNG SYSII M P4CORPORATO ON lE CORRESPONDING ORAW14S:
C) 1476, C) 0427, 0 04219, I) 04220 & 1) 04221                                                            PIPING I)IAG AM D&#xf8;229


Q 1 R NIT S:
Q 1 R NIT S:
I   QUPPJ
I QUPPJ
* NS RUM1S & PIPING AR REIXII IY UNlf                        JNII &:
* NS RUM1S &
SSIM         LMiS 2 th NLSS OTHWS N0ED UEFNCE LRWIIN(S SEC D &#xd8;1q.
PIPING AR REIXII IY JNII &
      .       511 IMEiAT ION
SSIM LMiS 2 th NLSS OTHWS N0ED UEFNCE LRWIIN(S SEC D &#xd8;1q.
      .), IINSmUI,lENtAl    lfl\ PENElRt'\nONS
. I 511 IMEiAT lfl\\
                                    - N RA1 ICNS AREAf* MULTI-LINES MJI I; L,IHFS THROUGH IR)JH ONl: QN S Iy 4.
- N RA1 ICNS Af*
: 4. ALL      INSRUAENT RACKS A...L INSTRUMENT      R.CKS M~E AFE PREnXED PREIXED H21" 2H2l.
MJI I; L,IHFS IR)JH QN S Iy
S. Ki REFER
: 4. A...L INSRUAENT R.CKS AFE PREIXED 2H2l.
: 5.          *FR O OG1C INTERLOCK.[NrR1QC<,
S. Ki *FR O OG1C [NrR1QC<,
      &A.L
&A.L ANUNCLA0 ALAQS R[ PRE-IXED 2 W2 Pfl XX, 7.
: 6.        ANUNCLA0 ALARMS ALL ANNUNCIATOR          ALAQS N~E    PRE-IXED 2 HI2**P60 R[ PI~EFlXED        W2 Pfl       XX,
-. OENDTS VALVE LEAKOFT WHICH WILL E NORM&L OPEN WILL 3E PIPED TO CRW, UNDER CASS 6e.
                                                                          ,**xX".
: 8. VENDOR FURNIS ED.
7.
9 < )oio f,S MASIR EQUIPMN LISt NUAt*R
: 7.    - . OENDTS W"LVE DENOTES    VALVE LEAKOFF LEAKOFT 'NHICl-I WHICH VaLL    E NORMAllY WILL BE  NORM&L OPEN OPEN
. SC-CS24 IS USt) fl Sl Ci filHfR IC RkI O FtC 3S? AI[)
          & WIll WILL BE3E PIPED PIPED TO TO C.R.W.
FO NSR OWR JIY SOUCf 10 1
CRW, UNDER     CASS 160",
NiL.
UNDER CLASS    6e.
U, x=:s: CLASS
: 8. VENDOR B.              FURNIS ED.
=Qc)..LTY Cj.,S EE 0- &#xf8;2i9 FQF ADDI IIQNA NOTES t2 HIGH POThT UTNT CAE S NORMALY MQVEO AND ORAIN tOSF NSTAI D FOR iS1FM VENTING ii SE ECHNICAL
VENDOR fURN1StIED.
*PoRr PI52 FOR APPUCADLE ASPE SCrION XI RFQUIRMN7S.
9   <                 MASIR EQU!PMENI
LF Categories KJA:
              )oio f,S MA$rEI~        EQUIPMN LIsr      NUAt*R LISt NUMBER,
SG2.02.15 Tier! Group:
    .       SC- CS24 IS USt) fl Sl Ci filHfR IC RkI O FtC 3S? AI[) FO           NSR OWR           JIY SOUCf 10 1 NiL.
T3 RO Rating:
U,                   x=:s: CLASS             -
3.9 SRO Rating:
                            =Qc)..LTY Cj.,S (1,2.3.M.- )
4.3 LP Obj:
EE 0- &#xf8;2i9 FQF AOantONAl ADDI IIQNA NOTES.
CLSLPO16*15E Source:
NOTES t2 HtGH
NEW Cog Level:
: 12. HIGH POiNTPOThT VENT UTNT CAP CAE !S NORMALY R(MOVEO S NORMAlt,Y    MQVEO ANO    ORAIN tOSF INSTALLED AND DRAIN                          FOR iS1FM VENrtNG.
HIGH Category 8:
NSTAI D r.OR            VENTING ii SE ECHNICAL           *PoRr PI52 F"OR R(PORf                    FOR APPLICABLE APPUCADLE ASPE SCrION         XI RFQUIRMN7S.
YF UNlf &:
XI LF Categories Categories KJA:
.), I NSmUI,lENtAl ION PENElRt'\\nONS ARE MULTI-LINES THROUGH ONl:
KIA:                SG2.02.15                                   Tier!/ Group:
: 4. ALL INSTRUMENT RACKS M~E PREnXED H21"
Tier            T3 RO Rating:
: 5.
RORating:            3.9 3.9                                          SRO Rating:
REFER INTERLOCK.
SRORating:      4.3 LP Obj:
: 6. ALL ANNUNCIATOR ALARMS N~E PI~EFlXED HI2**P60,**xX".
LPObj:              CLSLPO16*15E CLS-LP-016*15E                               Source:         NEW Cog Level:           HIGH                                         Category 8:     YF
: 7.
: 97. The
DENOTES W"LVE LEAKOFF 'NHICl-I VaLL BE NORMAllY OPEN WIll BE PIPED TO C.R.W. UNDER CLASS 160",
: 97. The following following conditions conditions exist exist on on Unit Unit One One after after aa transient:
B. VENDOR fURN1StIED.
transient:
MA$rEI~ EQU!PMENI LIsr NUMBER, (1,2.3.M.-)
Jet Pump Flow Jet Pump    Flow Loop Loop AA                              22 22 Mlbs/hr Mlbs/hr Jet Pump Jet  Pump FlowFlow Loop Loop BB                              33 Mlbs/hr 33 Mlbs/hr Recirc Pump Recirc   Pump AA Percent Percent Speed Speed                     47%
AOantONAl NOTES.
47%
: 12. HtGH POiNT VENT CAP !S NORMAlt,Y R(MOVEO ANO DRAIN INSTALLED r.OR VENrtNG.
Recirc   Pump Recirc Pump BB Percent Percent Speed Speed                      66%
R(PORf F"OR APPLICABLE XI Categories KIA:
SG2.02.15 Tier / Group:
RORating:
 
===3.9 SRORating===
LPObj:
CLS-LP-016*15E Source:
Cog Level:
HIGH Category 8:
T3 4.3 NEW YF
: 97. The following conditions exist on Unit One after a transient:
Jet Pump Flow Loop A 22 Mlbs/hr Jet Pump Flow Loop B 33 Mlbs/hr Recirc Pump A Percent Speed 47%
Recirc Pump B Percent Speed 66%
Total Core Flow (UICPWTCF) 55 MIbs/hr Which one of the following identifies the Required Action lAW T.S. 3.4.1, Recirculation Loops Operating, and the bases for this action?
Recirculation (1) mismatch is exceeded requiring Recirculation Loop A to be considered out of service (2)
A (1) Loop Flow (2) to ensure that assumptions of the LOCA analysis are satisfied B. (1) Loop Flow (2) due to the inability to detect significant degradation in jet pump performance C. (1) Pump Speed (2) to ensure that assumptions of the LOCA analysis are satisfied D. (1) Pump Speed (2) due to the inability to detect significant degradation in jet pump performance
: 97. The following conditions exist on Unit One after a transient:
Jet Pump Flow Loop A Jet Pump Flow Loop B Recirc Pump A Percent Speed Recirc Pump B Percent Speed Total Core Flow (U1CPWTCF) 22 Mlbs/hr 33 Mlbs/hr 47%
66%
66%
Total Core Total  Core Flow Flow (U1CPWTCF)
55 Mlbs/hr Which one of the following identifies the Required Action lAW T.S. 3.4.1, Recirculation Loops Operating, and the bases for this action?
(UICPWTCF)                          55 55 Mlbs/hr MIbs/hr Which one Which    one of   the following of the  following identifies identifies the the Required Required Action Action lAW lAW T.S.
Recirculation (1) mismatch is exceeded requiring Recirculation Loop A to be considered out of service (2)
T.S. 3.4.1, 3.4.1, Recirculation Recirculation Loops Operating, Loops    Operating, andand the the bases bases forfor this this action?
A'! (1) Loop Flow (2) to ensure that assumptions of the LOCA analysis are satisfied B. (1) Loop Flow (2) due to the inability to detect significant degradation in jet pump performance C. (1) Pump Speed (2) to ensure that assumptions of the LOCA analysis are satisfied D. (1) Pump Speed (2) due to the inability to detect significant degradation in jet pump performance  
action?
Recirculation Recirculation        (1)    mismatch is (1) mismatch       is exceeded exceeded requiring requiring Recirculation Recirculation Loop Loop AA to be be considered out of considered          of service service      (2)
(2)
(1) Loop A (1)
A'!      Loop Flow Flow (2) to ensure that assumptions of the LOCA analysis are satisfied B. (1) Loop FlowFlow (2) due to the inability to detect significant degradation in jet pump performance C. (1) Pump Speed (2) to ensure that assumptions of the LOCA analysis are satisfied D. (1) Pump Speed (2) due to the inability to detect significant degradation in jet pump performance


Feedback Feedback K/A: SG2.02.22 KIA:    SG2.02.22 Equipment Control Equipment       Control Knowledge of Knowledge            limiting conditions of limiting     conditions for    for operations operations andand safety safety limits.
Feedback K/A: SG2.02.22 Equipment Control Knowledge of limiting conditions for operations and safety limits.
limits.
(CFR: 41.5 / 43.2 I 45.2)
(CFR:   41.5 / 43.2 (CFR: 41.5 143.2 1 45.2) I 45.2)
RO/SRO Rating: 4.0/4.7 Objective: CLSLP002*34
RO/SRO Rating:
: 27. Explain why there is a limit for mismatch between total Jet Pump Loop flows
RO/SRO                  4.0/4.7 Rating: 4.0/4.7 CLSLP002*34 Objective: CLS-LP-002*34 Objective:
: 34. Given plant conditions and Technical Specifications, including the Bases, TRM, ODCM, and COLR determine the required action(s) to be taken in accordance with Technical Specifications associated with the Reactor Recirculation System. (SROISTA only)
: 27. Explain
: 27. Explain why       there isis aa limit why there            limit for  mismatch between for mismatch     between total total Jet Jet Pump Pump LoopLoop flows flows
: 34. Given
: 34. Given plant plant conditions conditions and  and Technical Technical Specifications, Specifications, including including the the Bases, Bases, TRM, TRM, ODCM, ODCM, and and COLR COLR determine the determine     the required required action(s) action(s) to  to be   taken in be taken  in accordance accordance with with Technical Technical Specifications Specifications associated associated with with the Reactor the  Reactor Recirculation Recirculation System.
System. (SRO/STA (SROISTA only)only)


==Reference:==
==Reference:==
Unit 1 Technical Specification 3.4.1 and BASES Cog Level: High Explanation:
Two recirculation loops are normally required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied.
Jet pump loop flow mismatch should be maintained within the following limits:
- jet pump loop flows within 10% (maximum indicated difference 7.5 xl 0 lbslhr) with total core flow less than 58 x10 lbs/hr
- jet pump loop flows within 5% (maximum indicated difference 3.5 xl 06 lbslhr) with total core flow greater than or equal to 58 xl 06 lbs/hr Distractor Analysis:
Choice A: Correct Answer Choice B: Plausible because Loop flow mismatch is correct and vibrations would be a result of low or reverse flow.
Choice C: Plausible because Pump Speed used to be the indication utilized and LOCA analysis is correct.
Choice D: Plausible because Pump Speed used to be the indication utilized and vibrations would be a result of low or reverse flow.
SRO Only Basis: Application of Required Actions and Knowledge of TS Bases.
Notes Feedback KIA: SG2.02.22 Equipment Control Knowledge of limiting conditions for operations and safety limits.
(CFR: 41.5 143.2 1 45.2)
RO/SRO Rating: 4.0/4.7 Objective: CLS-LP-002*34
: 27. Explain why there is a limit for mismatch between total Jet Pump Loop flows
: 34. Given plant conditions and Technical Specifications, including the Bases, TRM, ODCM, and COLR determine the required action(s) to be taken in accordance with Technical Specifications associated with the Reactor Recirculation System. (SRO/STA only)


==Reference:==
==Reference:==
 
Unit 1 Technical Specification 3.4.1 and BASES Cog Level: High Explanation:
Unit 11 Technical Unit    Technical Specification Specification 3.4.13.4.1 andand BASES BASES Cog Level:
Two recirculation loops are normally required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied.
Cog    Level: High High Explanation:
Jet pump loop flow mismatch should be maintained within the following limits:  
Explanation:
- jet pump loop flows within 10% (maximum indicated difference 7.5 x1 06 Ibs/hr) with total core flow less than 58 X102 Ibs/hr  
Two recirculation recirculation loops are are normally normally required required to be     operation with their flows matched within the limits be in operation specified inin SR 3.4.1.1 to ensure that that during a LOCA          LOCA caused by         break of the piping of one by a break recirculation loop the assumptions of the LOCA          LOCA analysis are satisfied.
- jet pump loop flows within 5% (maximum indicated difference 3.5 x106 Ibs/hr) with total core flow greater than or equal to 58 x106 Ibs/hr Distractor Analysis:
Jet pump loop flow mismatch should be maintained within the following limits:
  - jet pump loop flows within 10% (maximum indicated difference 7.5 x1 xl 006 Ibs/hr) lbslhr) with total core flow less than 58 X102     lbs/hr x10 Ibs/hr
  - jet pump loop flows within 5% (maximum indicated difference 3.5 x10 xl 066 Ibs/hr) lbslhr) with total core flow greater than or equal to 58 x10      xl 066 Ibs/hr lbs/hr Distractor Analysis:
Choice A: Correct Answer Choice B: Plausible because Loop flow mismatch is correct and vibrations would be a result of low or reverse flow.
Choice A: Correct Answer Choice B: Plausible because Loop flow mismatch is correct and vibrations would be a result of low or reverse flow.
Choice C: Plausible because Pump Speed used to be the indication utilized and LOCA analysis is correct.
Choice C: Plausible because Pump Speed used to be the indication utilized and LOCA analysis is correct.
Choice D: Plausible because Pump Speed used to be the indication utilized and vibrations would be a result of low or reverse flow.
Choice D: Plausible because Pump Speed used to be the indication utilized and vibrations would be a result of low or reverse flow.
SRO Only Basis:
SRO Only Basis: Application of Required Actions and Knowledge of TS Bases.
Basis: Application of Required Actions and Knowledge of TS Bases.
Notes  
Notes


Recirculation Loops Recirculation      Loops Operating Operating 3.4."1 3.4.1 3.4 3.4      REACTOR COOL.A.NT RE.A.CTOR    COOLANT SYSTEM SYSTEM (RCS)(RCS) 3.4.1 3.4 ..1   Recirculation Loops Recirculation Loops Operating Operating LCD 3.4.1 LCO    3.4.1          Two  recirculation loops Two recirculation   loops with with matched matched lIows flows shall shall be be inin operation.
Recirculation Loops Operating 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCD 3.4.1 Two recirculation loops with matched flows shall be in operaon, OR One recirculalion oop may be in operation provided the following limits are applied when the associated LCD is applicable:
operaon, OR One recirculalion loop One recirculation        may be oop may   be in operation provided in operation  provided thethe following following limits limits are applied when applied  when thethe associated associated LCO LCD is  applicable:
is applicable:
a.
a.
: a. LCO 3.2.1, LCO    3.2.1. "AVERAGE AVERAGE PLANAR PLANAR LINEAR LINEAR HEAT HEAT GENERATION GENERATION (APLHGR), single loop operation limits specified in the COLR; RATE (APLHGR),
LCO 3.2.1.
: b.               MINIMUM CRITICAL POWER RATIO {MCPR},
AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR), single loop operation limits specified in the COLR; b.
LCO 3.2.2, "MINIMUM                                           (MCPR), single loop operation limits specified in the COLR;COLR; c.
LCO 3.2.2, MINIMUM CRITICAL POWER RATIO (MCPR), single loop operation limits specified in the COLR; c.
: c.        3.2.3. "LINEAR LCO 3.2.3,    LINEAR HEAT HEAT GENERATION GENERATION RATE  RATE (LHGR),"
LCO 3.2.3. LINEAR HEAT GENERATION RATE (LHGR) single loop operation limits specified in the COLR: and d.
(LHGR) single specified in the COLR; loop operation limits specified             COLR: and
LCO 3.3.1.1, Reactor Protection System RPS) Insfrumentation, Function 2.b (Average Power Range Monitors Simulated Thermal PowerHigh), Allowable Value of Tabe 3.3.1.1-1 is reset for single loop operation.
: d.                   Reactor Protection System (RPS)
APPLICABILITY:
LCO 3.3.1.1, "Reactor                             RPS) Instrumentation, Insfrumentation, Function 2.b (Average Power Range Monitors Simulated Thermal PowerHigh), Allowable Value Power-High),                  Value of Table Tabe 3.3.
MODES I and 2.
3.3.1.1-1
ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A.
                                                                                '1.1-1 is resetror reset for single loop operation.
Requirements of the LCD A.1 Satist the requirements of 6 hours not met, the LCD.
APPLICABILITY:         MODES 1  I and 2.
ACTIONS COMPLETION CONDITION                           REQUIRED ACTION                               TIME A. Requirements of the LCO LCD       A.1         Satist the requirements of SatisPj                            6 hours not met, nolmet                                      the LCO.
LCD.
( continued)
(continued)
(continued)
Brunswick Unit Siunswick     Unit 11                              3.4-1 3.4-1                               Amendment Amendment No.        246 No. 246   I
Siunswick Unit 1 3.4-1 Amendment No. 246 Recirculation Loops Operating 3.4."1 3.4 RE.A.CTOR COOL.A.NT SYSTEM (RCS) 3.4.. 1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched lIows shall be in operation.
One recirculation loop may be in operation provided the following limits are applied when the associated LCO is applicable:
: a.
LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR), single loop operation limits specified in the COLR;
: b.
LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO {MCPR}, single loop operation limits specified in the COLR;
: c.
LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," single loop operation limits specified in the COLR; and
: d.
LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation, Function 2.b (Average Power Range Monitors Simulated Thermal Power-High), Allowable Value of Table 3.3. '1.1-1 is resetror single loop operation.
APPLICABILITY:
MODES 1 and 2.
ACTIONS CONDITION A.
Requirements of the LCO nolmet Brunswick Unit 1 A.1 REQUIRED ACTION COMPLETION TIME SatisPj the requirements of 6 hours the LCO.
( continued) 3.4-1 Amendment No. 246 I  


Recirculation Loops Recirculation  Loops Operating Operating 3A:l 3.4.1 ACTIONS (continued)
Recirculation Loops Operating ACTIONS continuedi 3.4.1 COMPLETION CONDITION REQUIRED ACTION TIME 5.
ACTIONS    continuedi COMPLETION COMPLETION CONDITION CONDITION                                    REQUIRED .A.CTlON REQUIRED      ACTION                   TIME TIME 5.
Required Action and 8.1 Se n MODE 3.
B. Required Action Required   Action and and                8.1 B.1            Be  n MODE Se in MODE 3.3.              12 12 hours hours associated Completion associated  Completion Time  Time of Condition of Condition .A.
12 hours associated Completion Time of Condition A not met.
A not  met.
OR No recirculation loops in operation.
not met.
SURVEILLANCE_REQUIREMENTS SURVEILLANCE FREQUENCY SR 34.i.1
OR No recirculation No  recirculation loops loops in in operation.
------------------NOTE----
operation.
Not required to be performed until 24 hours after both recrcuation toops are in operation.
SURVEILLANCE_REQU SURVEILLANCE                IREMENTS REQUIREMENTS SURVEILLANCE SUR\lEILLANCE                                            FREQUENCY FREQUENCY SR 3.4:1.1 SR 34.i.1                           ------------------NOTE----
Verify recirculation loop jet pump flow mismatch with 24 hours both recirculation loops in operation:
                    --------------------------NOTE--------------------------
10% of rated core flow when operating at
Not required to be performed until 24 hours after both recrcuation lOops recirculation    toops are in operation.
< 75% of rated core flow; and b.
Verify recirculation loop jet pump flow mismatch with               24 hours both recirculation loops in operation:
5% of rated core flow when operating at 75% of rated core flow.
: a.     5: -t10%
Snjnswjck Unit 1 3.4-2 Amendment No. 244 Recirculation Loops Operating 3A:l ACTIONS (continued)
0% of rated core flow when operating at
CONDITION REQUIRED.A.CTlON B.
                              <: 75% of rated core flow;
Required Action and B.1 Be in MODE 3.
                              <                            flow; and
associated Completion Time of Condition.A. not met.
: b.     5: 5% of rated core flow    flO'A' when operating at
No recirculation loops in operation.
                            ;:: 75% of rated core flow.
SURVEILLANCE REQUIREMENTS SR 3.4:1.1 Brunswick Unit 1 SUR\\lEILLANCE
Snjnswjck Brunswick Unit Unit 11                                      3.4-2 3.4-2                        Amendment Amendment No.No. 244 244
--------------------------NOTE--------------------------
Not required to be performed until 24 hours after both recirculation lOops are in operation.
Verify recirculation loop jet pump flow mismatch with both recirculation loops in operation:
: a.
5: -t 0% of rated core flow when operating at  
<: 75% of rated core flow; and
: b.
5: 5% of rated core flO'A' when operating at  
;:: 75% of rated core flow.
3.4-2 COMPLETION TIME 12 hours FREQUENCY 24 hours Amendment No. 244  


Recirculation Loops Recirculation     Loops Operating Operating BB 3.4.1 3.4.1 BASES B.A.SES APPLICABLE APPLICABLE              For AREVA For    AREVA fuel, ftiel, the the COLR COLR presents presents single single loop loop operation operation APLHGR APLHGR limits limits SAFETY AN.A.L SAFETY    ANALYSES YSES in the in   the form form of of aa multiplier multiplier that that is is applied applied to to the the two two loop loop operation operation (continued)
Recirculation Loops Operating B 3.4.1 BASES APPLICABLE For AREVA ftiel, the COLR presents single loop operation APLHGR limits SAFETY ANALYSES in the form of a multiplier that is applied to the two loop operation (continued)
(continued}         APLHGR limits.
APLHGR limits.
APLHGR         limits.
The transient analyses of Chapter 15 of the UFSAR have also been evaluated for single recirculation loop operation. The evaluation concludes that results of the transient analyses are not significantly affected by the single recirculation loop operation. There is, however, an impact on the fuel cladding integritj SL since some of the uncertainties for the parameters used in the critical power determination are higher in single loop operation. The net result is an increase in the MCPR operating limit.
The transient The    transient analyses analyses of  of Chapter Chapter 15 15 of of the the UFSAR UFSAR have have also also been been evaluated for evaluated       for single   recirculation loop single recirculation      loop operation.
During single recircutation loop operation, modification to the Reactor Pratection System (RPS) average power range monitor APRM)
operation. The The evaluation evaluation concludes that concludes       that results results ofof the the transient transient analyses analyses are are not not significantly significantly affected by affected      by the the single    recirculation loop single recirculation       loop operation.
Simulated Themial PowerHigh Allowable Value is required to account for the different analyzed limits between two-recirculation drive now loop operation and operation with only one loop. The APRM channel subtracts the W value from the measured recirculation drive flow to effectivej shift the limits and uses the adjusted recirculation drive flow value to determine the APRM Simulated Themial PowerHigh Function trip setpoint.
operation. There There is,   however, an is, however,    an impact on impact     on the   fuel cladding the fuel    cladding integrity integritj SL SL since since some some of of the the uncertainties uncertainties forfor the parameters the    parameters used   used inin the the critical critical power power determination determination are  are higher higher in in single loop single     loop operation.
Recirculation loops operating satisfies Criterion 2 of 10 CFR S0.36(cX2)(ii) (Ref. 4).
operation. The  The net     result is net result    is an an increase increase in in the the MCPR MCPR operating limit.
LCO Two recirculation loops are normally required to be in operation with their flows matched within the limits specified in SR 3.4.1,1 to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. Alternately, with only one recirculation loop in operation, modifications to the required APLHGR limits LCO 3.2.1, AVERAGE PLANAR L1NEAR HEAT GENERATION RATE (APLHGRy), MCPR limits (LCO 3.2.2.
operating       limit.
MINIMUM CRITICAL POWER RATIO MCPRfl, LHGR limits (LCO 3.2.3, LINEAR HEAT GENERATION RATE (LHGR)), and APRM Simulated Thermal Power High Allowable Value (LCO 3.3.1.1), as applicable, must be applied to allow continued operation. The COLR defines adjustments or modifications required for the APLHGR, MCPR, and LHGR limits for the current operating cycle.
During single During                recircutation loop single recirculation       loop operation, operation, modification to     to the  Reactor the Reactor Pratection System Protection       System (RPS)(RPS) average average power power range range monitor monitor (APRM)
continued)
APRM)
Brunswick Unit 1 6 3.4.1-3 Revision No.58 B.A.SES Recirculation Loops Operating B 3.4.1 APPLICABLE For AREVA fuel, the COLR presents single loop operation APLHGR limits SAFETY AN.A.L YSES in the form of a multiplier that is applied to the two loop operation (continued}
Simulated Themlal Simulated        Themial Power-High PowerHigh Allowable 'Value       Value isis required required to to account for the different analyzed limits bet'A'een   between two-recirculation two-recirculation drive tlow now loop loop operation and operation with only one loop.            loop. The APRM channel subtracts W value from the measured the l!.W                                          recirculation drive flow to effective~1 measured recirculation                            effectivej the limits shift the     limits and uses the  the adjusted recirculation drive flow value to determine the APRM Simulated Themlal           Themial Power-High PowerHigh Function trip setpoint.
APLHGR limits.
Recirculation loops operating satisfies Criterion 2 of Recirculation S0.36(cX2)(ii) (Ref. 4}.
lCO Brunswick Unit 1 The transient analyses of Chapter 15 of the UFSAR have also been evaluated for single recirculation loop operation. The evaluation concludes that results of the transient analyses are not significantly affected by the single recirculation loop operation. There is, however, an impact on the fuel cladding integrity SL since some of the uncertainties for the parameters used in the critical power determination are higher in single loop operation. The net result is an increase in the MCPR operating limit.
10 CFR 50.36(c)(2)(ii)                   4).
During single recirculation loop operation, modification to the Reactor Protection System (RPS) average power range monitor (APRM)
LCO lCO                  Two recirculation loops are normally required to be in operation with their flows matched within the limits specified in SR 3.4.1.1            3.4.1,1 to ensure !hat that dunng a LOCA caused by a break of the piping of one recirculation loop during the assumptions of the LOCA analysis are satisfied. Alternately, with only
Simulated Themlal Power-High Allowable 'Value is required to account for the different analyzed limits bet'A'een two-recirculation drive tlow loop operation and operation with only one loop. The APRM channel subtracts the l!.W value from the measured recirculation drive flow to effective~1 shift the limits and uses the adjusted recirculation drive flow value to determine the APRM Simulated Themlal Power-High Function trip setpoint.
                      !he one recirculation loop in operation, modifications to the required APLHGR LCO 3.2.1, "AVERAGE limits (LCO                AVERAGE PLANAR LINEAR         L1NEAR HEAT GENERATION (APLHGRy), MCPR limits (LCO RATE (APLHGR)"),                                         3.2.2. "MINIMUM (LeO 3.2.2,       MINIMUM CRITICAL POWER RATIO MCPRfl, POWER                    (MCPR)"), LHGR limits (LCO 3.2.3, LINEAR          "LINEAR HEAT HEAT GENERATION RATE (LHGR)),        (LHGR)"), and APRM Simulated    Simulated Thermal Power  Power-High Allowable Value (LCO 3.3.1.1), as applicable, must be applied to allo\'/ continued operation. The COLR defines adjustments or allow modifications modifications required for the APLHGR,      APLHGR, MCPR,  MCPR, and LHGR   LHGR limits limits for the current operating cycle.
Recirculation loops operating satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii) (Ref. 4}.
current (continued) continued)
Two recirculation loops are normally required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure !hat dunng a LOCA caused by a break of the piping of one recirculation loop  
Brunswick Brunswick Unit Unit 11                                  6B 3.4.1-3 3.4.1-3                                          Revision No.58 Revision     No. 58
!he assumptions of the LOCA analysis are satisfied. Alternately, with only one recirculation loop in operation, modifications to the required APLHGR limits (LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), MCPR limits (LeO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"), LHGR limits (LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)"), and APRM Simulated Thermal Power-High Allowable Value (LCO 3.3.1.1), as applicable, must be applied to allo\\'/ continued operation. The COLR defines adjustments or modifications required for the APLHGR, MCPR, and LHGR limits for the current operating cycle.
( continued)
B 3.4.1-3 Revision No. 58  


Recirculation Loops Recirculation      Loops Operating Operating B 3.4.1 3.4.1 BASES BASES APPLICABILITY APPLICABILITY    InIn MODES MODES 1I and   and 2,2, requirements requirements ior    for operation operation ofthe of the Reactor Reactor Coolant Coolant Recirculation System Recirculation      System are  are necessary' necessary since since there there isis considerable considerable energy energy in   in the  reactor core the reactor             and the core and     the limiting liniiting design design basis basis transients transients andand accidents accidents are    are assumed to assumed        to occur.
Recirculation Loops Operating 3.4.1 APPLICABILITY In MODES I and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the liniiting design basis transients and accidents are assumed to occur.
occur.
In MODES 3,4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recircuation loops are not important.
InIn MODES MODES 3,    3,4,    and 5, 4, and     5, the the consequences consequences of    of an   accident are an accident      are reduced reduced and    and the coastdown the  coastdown characteristics characteristics of    of the   recircuation loops the recirculation    loops areare not not important.
ACTIONS Al With the requirements of the LCO not met, the recirculation loops must be restored to operation with matched flows within 6 hours. A recirculation loop is considered not in operation when the pump in that loop is idle or when the mismatch between total jet pump flows of the two loops is greater than the required limits. The loop with the lower flow must be considered not in operation. Should a LOCA occur with one recirculation loop not in operation, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses. Therefore, only a limited time is allowed to restore the inoperable loop to operating status.
important.
Alternatively, if the single loop requirements of the LCO are applied to operating limits and RPS setpoints, as applicable, operation with only one recirculation loop would satisfy the requirements of the LCO and the initial conditions of the accident sequence.
ACTIONS ACTIONS            Al With  the requirements With the     requirements of    of the   LCO not the LCO            met, the not met,        recirculation loops the recirculation    loops mustmust be  be restored to restored      to operation operation with with matched       flows within matched flows      within 66 hours.
The 6 hour Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable time to complete the Required Action (i.e.. reset the applicable limits or setpoints for single recirculation loop operation), and on frequent care monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected.
hours. AA recirculation recirculation loop is loop   is considered considered not   not inin operation operation when        the pump when the     pump in in that that loop loop is   idle or is idle   or when the when    the mismatch mismatch between between total  total jet pump    flows of pump flows     of the the two two loops loops is  is greater than greater    than thethe required required limits.
This Required Action does not require tripping the recirculation pump in the lowest flow loop when the mismatch between the total jet pump flaws of the two loops is greater than the required limits. However, in cases where large flow mismatches occur, low flow or reverse flow can occur in the low flaw loop jet pumps. causing vibration of the jet pumps.
limits. The The loop loop with the lower lower flow must must be  be considered not considered        not in operation. Should in operation.        Should aa LOCA LOCA occur occur with with one one recirculation recirculation loop no!
If zero or reverse flaw is detected, the condition should be alleviated by changing pump speeds to re-establish forward flow.
loop  not inin operation, operation, the the core core flow flow coastdown coastdown and        resultant core and resultant     core response may not    not be bounded bounded by the LOCA analyses. Therefore, only a limited time limited  time is allowed allowed to      restore the to restore     the inoperable loop loop toto operating operating status.
I continued)
status.
BASES Brunswick Unit 1 B 3.4.1-4 Revision No. SB BASES APPLICABILITY ACTIONS Brunswick Unit 1 Recirculation Loops Operating B 3.4.1 In MODES 1 and 2, requirements ior operation ofthe Reactor Coolant Recirculation System are necessary' since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur.
Alternatively, ifif the single Altematively,              single loop       requirements of the lCO loop requirements                  LCO are applied applied to operating limits and RPS setpoints, as applicable, operation with only one recirculation loop would satisfl/   satisfy the requirements of the LCO and the initial conditions of the accident sequence.
In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important.
The 6 hour Completion Time is bae.ed           based on the the low probability of an accident occurring during this time period, on a reae.onable         reasonable time to complete the Required Action (I.e.,         (i.e.. reset the applicable limits or setpoints for single recirculation loop operation), and on frequent core                 care monitoring by operators allowing abrupt changes in core flow conditions to be quickiy                     quickly detected.
With the requirements of the LCO not met, the recirculation loops must be restored to operation with matched flows within 6 hours. A recirculation loop is considered not in operation when the pump in that loop is idle or when the mismatch between total jet pump flows of the two loops is greater than the required limits. The loop with the lower flow must be considered not in operation. Should a LOCA occur with one recirculation loop no! in operation, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses. Therefore, only a limited time is allowed to restore the inoperable loop to operating status.
This Required Action does not require tripping the recirculation pump in the lowest flow loop when the mismatch between the total jet pump flaws                          flows of the two loops is greater than the required limits. However, in cases where large flow mismatches occur, low flow or reverse flow can occur in the low flaw flow loop jet pumps.
Altematively, if the single loop requirements of the lCO are applied to operating limits and RPS setpoints, as applicable, operation with only one recirculation loop would satisfl/ the requirements of the LCO and the initial conditions of the accident sequence.
pumps, causing          vibration of the jet pumps. If causingllibralion                                If zero or reverse flawflow is detected, the condition condition should should be alleviated by changing pump speeds to re-establish forward flow.
The 6 hour Completion Time is bae.ed on the low probability of an accident occurring during this time period, on a reae.onable time to complete the Required Action (I.e., reset the applicable limits or setpoints for single recirculation loop operation), and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickiy detected.
pump I(continued) continued)
This Required Action does not require tripping the recirculation pump in the lowest flow loop when the mismatch between the total jet pump flows of the two loops is greater than the required limits. However, in cases where large flow mismatches occur, low flow or reverse flow can occur in the low flow loop jet pumps, causingllibralion of the jet pumps. If zero or reverse flow is detected, the condition should be alleviated by changing pump speeds to re-establish forward flow.
Brunswick Brunswick Unit Unit 11                              BB 3.4.1-4 3.4.1-4                                         Revision No.
(continued)
Revision       No. SB58
B 3.4.1-4 Revision No. 58


Recirculation Loops Recirculation      Loops Operating Operating BB 3.4:1 3.4.1 BASES BASES ACTIONS ACTIONS             .!U.
Recirculation Loops Operating B 3.4.1 ACTIONS (continued)
(continued)
With no recirculation loops in operation or the Required Action and associated Completion Time oi Condition A not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the pant must be brought to MODE 3 within 12 hours. In this condition, the recirculation loops are not required to be operating because of the reduced severity of DBAs and minimal dependence on the recirculation loop coastdown characteristics. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
(continued)
SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This SR ensures the recirculation loops are within the allowable limits for mismatch. At low core flow (i.e.. < 76% of rated core flow), the MCPR requirements provide larger margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of early boiling transition during a LOCA is reduced. A larger flow mismatch can, therefore, be allowed when core flow is < 75% of rated core flow. The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop.
With no With   no recirculation recirculation loopsloops in in operation operation or    or the the Required Required Action Action and and associated Completion associated       Completion Time   Time of oi Condition Condition AA not  not met, met, the the plant plant must must be  be brought to brought    to aa MODE MODE in    in which which the the LCO LCO does does not not apply_
The mismatch is measured in terms of the percent of rated core flow.
apply. ToTo achieve achieve this  this status, th.e status,    the plant pant must must be  be brought brought to    to MODE MODE 33 within within 12 hours_
If the flow mismatch exceeds the specified limits, the loop with the lower flow is considered not in operation. The SR is not required when both loops are not in operation since the mismatch limits are meaningless during single loop or natural circulation operation. The Surveillance must be performed within 24 hours after both loops are in operation. The 24 hour Frequency is consistent with the Surveillance Frequency for jet pump OPERABILITY verification and has been shown by operating experience to be adequate to detect off normal jet pump loop flows in a timely manner.
hours. In  In this this condition, the condition,   the recirculation recirculation loopsloops areare not not required required to to be be operating operating because because of the of the reduced reduced severity severity of  of DBAs DBAs and  and minimal minimal dependence dependence on    on the the recirculation loop recirculation    loop coastdown coastdown characteristics.
REFERENCES 1.
characteristics. The    The allowed allowed Completion Completion Time of Time    of 12 12 hours hours is is reasonable, reasonable, based  based on    on operating operating experience, experience, to     to reach reach MODE 33 from MODE        from full full power power conditions conditions in    in an   orderly manner an orderly    manner and and *...without
UFSAR. Section 5.4.1.3.
                                                                                                                    'ithout challenging piant challenging      plant systems.
systems.
SURVEILLANCE SURVEILLANCE          SR 3.4.U SR   3.4.1.1 REQUIREMENTS REQUIREMENTS This SRSR ensures the recirculation loops          loops are within the allowable limits for mismatch. At low mismatch_          low core t/ow flow (Le.,
(i.e.. -<< 75%
76% of rated rated core flow), the MCPR requirements provide larger    larger margins to the fuel cladding integrity Safety Limit such that the potential Limit                      potential adverse effect     effect ofof early boiling transition during a LOCA is reduced_reduced. A larger flow mismatch can, therefore, be allowed when core tlow aI/owed                    flow is < 75% of rated core flow. The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop_                        loop.
The mismatch is measured in terms of the percent of rated core flow. If the flow mismatch exceeds the specified limits, the loop wiltl                    with the lower flow is considered not in operation. The SR is not required when both loops are not in operation since the mismatch limits are meaningless during single looploop or natural circulation operation. The Surveillance must be performed within 24 hours after both loops are in operation.             operation_ The 24 hour frequency Frequency is consistent with the Surveillance Frequency for jet pump OPERABILITY OPERA.BILITY verification and has been shown by operating experience to be adequate to          to detect off normal jet      jet pump pump loop loop flows in a timely manner.
manner.
REFERENCES REFERENCES           1.
: 1.        UFSAR.
UFSAR, Section Section 5.4.1.3.
SA 1.3.
2.
2.
: 2.        UFSAR, UFSAR, Chapter Chapter 15. 15.
UFSAR, Chapter 15.
3.
3.
: 3.        NEDC-31776P, NEDC-31776P, Brunswick  Brunswick Steam  Steam Electric Electric Plant     Units 1 Plan! Units    I andand 22 Single Single Loop Loop Operation, Operation, February February 1990. 1990.
NEDC-31776P, Brunswick Steam Electric Plant Units I and 2 Single Loop Operation, February 1990.
4.
4.
: 4.        10 10 CFR CFR E0.36(c)t2)(ii).
10 CFR E0.36(c)t2)(ii).
SD_36(c)(2)(ii)_
Brunswick Unit 1 B 3.4.1-6 Revision No.58 Categories KJA:
Brunswick   Unit 11 Brunswick Unit                                      B63A1-5 3.4.1-6                                           Revision No.58 Revision        No. 58 Categories Categories KJA:
SG2.02.22 Tier / Group:
KIA:            SG2.02.22 SG2.02.22                                                      Tier Tier // Group:
T3 RO Rating:
Group:      T3 T3 RO Rating:
4.0 SRO Rating:
RORating:        4.0 4_0                                                            SRO     Rating:
4.7 LP Obj:
SRORating:            4.7 4.7 LP LP Obj:
CLSLP0O2*34 Source:
Obj:        CLSLP0O2*34 CLS-LP-002*34                                                  Source:
NEW Cog Level:
Source:              NEW NEW Cog Cog Level:
HIGH Category 8:
Level:    HIGH mGH                                                            Category Category 8:8:         YY
Y BASES BASES Recirculation Loops Operating B 3.4:1 ACTIONS
: 98. Which
.!U.
: 98. Which one,Qf    the following oneQf the   following identifies identifies the the procedure procedure required required to to control control drywell drywell pressure pressure within PCPL-A lAW PCCP withinPCPL-A:IAW       PCCP and    the release and the   release rate rate restrictions, restrictions, ifif any, any, inin effect effect during during the the venting?
(continued)
venting?
With no recirculation loops in operation or the Required Action and associated Completion Time of Condition A not met, the plant must be brought to a MODE in which the LCO does not apply_ To achieve this status, th.e plant must be brought to MODE 3 within 12 hours_ In this condition, the recirculation loops are not required to be operating because of the reduced severity of DBAs and minimal dependence on the recirculation loop coastdown characteristics. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and *... 'ithout challenging piant systems.
A A~ SEP-01 Section SEP-01    Section 11 'j Venting Venting Primary Primary Containment Containment irrespective irrespective of of Off Off Site Site Release Release rate rate B. SEP-01, B. SEP-01, Section Section 2.,.
SURVEILLANCE SR 3.4.U REQUIREMENTS This SR ensures the recirculation loops are within the allowable limits for mismatch_ At low core t/ow (Le., -< 75% of rated core flow), the MCPR requirements provide larger margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of early boiling transition during a LOCA is reduced_ A larger flow mismatch can, therefore, be aI/owed when core tlow is < 75% of rated core flow. The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop_
2 Venting Venting Primary Primary Containment Containment via via the the Suppression Suppression Chamber Chamber within Site  Release    Rate within Site Release Rate Limit Limit C. SEP-01, Section C. SEP-01,   Section 3,3, Venting Venting Primary Primary Containment Containment via  via the the Drywell Drywell within within Site Site Release   Rate Release Rate LimitLimit D. OEDMG-003, D. OLDMG-003, Containment Containment Venting Under Under Conditions Conditions of of Extreme Extreme DamageDamage irrespective  of Off  Site Release irrespective of Off Site Release RatesRates
The mismatch is measured in terms of the percent of rated core flow. If the flow mismatch exceeds the specified limits, the loop wiltl the lower flow is considered not in operation. The SR is not required when both loops are not in operation since the mismatch limits are meaningless during single loop or natural circulation operation. The Surveillance must be performed within 24 hours after both loops are in operation_ The 24 hour frequency is consistent with the Surveillance Frequency for jet pump OPERA.BILITY verification and has been shown by operating experience to be adequate to detect off normal jet pump loop flows in a timely manner.
REFERENCES
: 1.
: 2.
: 3.
UFSAR, Section SA 1.3.
UFSAR, Chapter 15.
: 4.
NEDC-31776P, Brunswick Steam Electric Plan! Units 1 and 2 Single Loop Operation, February 1990.
10 CFR SD_36(c)(2)(ii)_
Brunswick Unit 1 63A1-5 Revision No. 58 Categories KIA:
RORating:
LP Obj:
Cog Level:
SG2.02.22 4_0 CLS-LP-002*34 mGH Tier / Group: T3 SRORating: 4.7 Source:
NEW Category 8:
Y
: 98. Which oneQf the following identifies the procedure required to control drywell pressure within PCPL-A lAW PCCP and the release rate restrictions, if any, in effect during the venting?
A SEP-01 Section 1 Venting Primary Containment irrespective of Off Site Release rate B. SEP-01, Section 2 Venting Primary Containment via the Suppression Chamber within Site Release Rate Limit C. SEP-01, Section 3, Venting Primary Containment via the Drywell within Site Release Rate Limit D. OLDMG-003, Containment Venting Under Conditions of Extreme Damage irrespective of Off Site Release Rates
: 98. Which one,Qf the following identifies the procedure required to control drywell pressure withinPCPL-A:IAW PCCP and the release rate restrictions, if any, in effect during the venting?
A~ SEP-01 Section 1 'j Venting Primary Containment irrespective of Off Site Release rate B. SEP-01, Section 2.,. Venting Primary Containment via the Suppression Chamber within Site Release Rate Limit C. SEP-01, Section 3, Venting Primary Containment via the Drywell within Site Release Rate Limit D. OEDMG-003, Containment Venting Under Conditions of Extreme Damage irrespective of Off Site Release Rates


Feedback Feedback SG2.03.11 K/A: SG2.03.11 KIA:
Feedback K/A: SG2.03.11 Radiation Control Ability to control radiation releases.
Radiation Control Radiation       Control Ability to Ability    to control control radiation radiation releases.
(CFR: 41.11 /43.4/45.10)
releases.
ROISRO Rating: 3.8/4.3 Objective: CLSLP300L*08d
(CFR:   41.11   /43.4/45.10)
: 8. Given the Primary Containment Control Procedure and plant conditions, determine if the following actions are required:
: c. Venting the primary containment while staying within radioactivity release rate limits
: d. Venting the primary containment IRRESPECTIVE of radioactivity release rate limits
 
==Reference:==
001-37.8, Revision 4, Page 33, Step PC/P-18 Cog Level: High Explanation:
Action to vent the primary containment is taken before drywell pressure rises to Primary Containment Pressure Limit A to assure that the integrity of the primary containment is maintained and to prevent core damage that might be caused by the inability to vent the reactor, as necessary, to permit injection of water to cool the core. Venting of the primary containment is performed irrespective of the off-site radioactivity release rate that will occur, and defeating isolation interlocks if necessary, because the consequences of not doing so may be either severe core damage or loss of primary containment integrity and uncontrolled radioactive release much greater than might otherwise occur. Note that primary containment venting is performed only, as necessary, to restore and then maintain pressure below the limit.
Distractor Analysis:
Choice A: Correct Answer.
Choice B: Plausible because within ODCM limits is utilized during SEP-01 section 1 when venting the torus due to containment Hydrogen/Oxygen concentration concerns and before exceeding PCPL-A Choice C: Plausible because within ODCM limits is utilized during SEP-01 section 1 when venting the drywell due to containment Hydrogen/Oxygen concentration concerns and before exceeding PCPL-A.
Choice D: Plausible because irrespective is correct and after exceeding PCPL-A is wrong SRO Only Basis: Detailed knowledge of diagnostic steps and decision points in the EOPs that involve transitions to emergency contingency procedures.
Notes PRIMARY CONTAINMENT PRESSURE LIMIT-A The lesser of the pressure capability of the primary containment, pressure at which containment vent valves sized to reject all decay heat from the containment can be opened and closed, or pressure at which SRVs can be opened and will remain open (Figure 2).
DE0P-01-UG Rev. 55 Page 71 of 151 Feedback KIA: SG2.03.11 Radiation Control Ability to control radiation releases.
(CFR: 41.11/43.4/45.10)
(CFR: 41.11/43.4/45.10)
ROISRO Rating:
RO/SRO Rating: 3.8/4.3 Objective: CLS-LP-300-L *08d
RO/SRO       Rating: 3.8/4.3 3.8/4.3 Objective:      CLSLP300L*08d Objective: CLS-LP-300-L         *08d
: 8. Given the Primary Containment Control Procedure and plant conditions, determine if the following actions are required:
: 8. Given the
: c. Venting the primary containment while staying within radioactivity release rate limits
: 8. Given    the Primary Primary Containment Containment Control Control Procedure Procedure and  and plant plant conditions, conditions, determine determine ifif thethe following following actions are actions    are required:
: d. Venting the primary containment IRRESPECTIVE of radioactivity release rate limits  
required:
: c. Venting
: c. Venting thethe primary primary containment containment while while staying staying within within radioactivity radioactivity release release rate rate limits limits
: d. Venting
: d. Venting thethe primary primary containment containment IRRESPECTIVE IRRESPECTIVE of            radioactivity release of radioactivity   release rate rate limits limits


==Reference:==
==Reference:==
001-37.8, Revision 4, Page 33, Step PC/P-18 Cog Level: High Explanation:
Action to vent the primary containment is taken before drywell pressure rises to Primary Containment Pressure Limit A to assure that the integrity of the primary containment is maintained and to prevent core damage that might be caused by the inability to vent the reactor, as necessary, to permit injection of water to cool the core. Venting of the primary containment is performed irrespective of the off-site radioactivity release rate that will occur, and defeating isolation interlocks if necessary, because the consequences of not doing so may be either severe core damage or loss of primary containment integrity and uncontrolled radioactive release much greater than might otherwise occur. Note that primary containment venting is performed only, as necessary, to restore and then maintain pressure below the limit.
Distractor Analysis:
Choice A: Correct Answer.
Choice B: Plausible because within ODCM limits is utilized during SEP-01 section 1 when venting the torus due to containment Hydrogen/Oxygen concentration concerns and before exceeding PCPL-A Choice C: Plausible because within ODCM limits is utilized during SEP-01 section 1 when venting the drywell due to containment Hydrogen/Oxygen concentration concerns and before exceeding PCPL-A.
Choice D: Plausible because irrespective is correct and after exceeding PCPL-A is wrong SRO Only Basis: Detailed knowledge of diagnostic steps and decision points in the EOPs that involve transitions to emergency contingency procedures.
Notes PRIMARY CONTAINMENT PRESSURE LlMIT-A The lesser of the pressure capability of the primary containment, pressure at which containment vent valves sized to reject all decay heat from the containment can be opened and closed, or pressure at which SRVs can be opened and will remain open (Figure 2).
IOEOP-01-UG Rev. 55 Page 71 of 151 I


==Reference:==
-rT C
Cl) z H
C) 0
 
-1


001-37.8, Revision 001-37.8,      Revision 4,4, Page Page 33, 33, Step Step PC/P-18 PC/P-18 Cog Level:
C
Cog    Level: High High Explanation:
-4 m
Explanation:
Inl cci
Action to Action    to vent the the primary primary containment is    is taken before before drywell pressure pressure rises rises to Primary Containment Containment Pressure Limit Pressure      Limit AA to assure that the integrity of the primary containment is maintained    maintained and to prevent core damage that might be caused by the inability to vent the reactor, as necessary, to permit injection of damage                                                                                                                        of water cool the core. Venting of the primary containment is performed irrespective of the off-site radioactivity to cool release rate that will occur, and defeating isolation interlocks if necessary, because the consequences release                                                                                                          consequences of not doing so may be either severe core damage or loss of primary containment integrity and uncontrolled not radioactive release much greater than might otherwise occur. Note that primary containment venting is performed performed only, as necessary, to restore and then maintain pressure below the limit.
-o Cl)cl) 0 o
Distractor Analysis:
mmo
Distractor Choice A: Correct Answer.
-w
Choice        Plausible because within ODCM limits is utilized during SEP-01 section 11 when venting the Choice B: Plausible torus due to containment Hydrogen/Oxygen concentration concerns and before exceeding PCPL-A PCPL-A Choice Choice C: Plausible because within ODCM limits is utilized during SEP-01 section 1                          1 when venting the drywell due to containment Hydrogen/Oxygen concentration concerns and before exceeding PCPL-A.
Choice D: Plausible becausebecause irrespective is correct and after exceeding PCPL-A is                  is wrong SRO OnlyOnly Basis: Detailed knowledge of        of diagnostic steps and decision decision points in the EOPs that involve transitions transitions to emergency emergency contingency procedures.
Notes Notes PRIMARY PRIMARY CONTAINMENT CONTAINMENT PRESSUREPRESSURE LIMIT-ALlMIT-A The lesser The  lesser of  the pressure of the pressure capability capability of of the  primary containment, the primary    containment, pressure pressure atat which containment which    containment vent vent valves valves sized sized toto reject reject all all decay decay heat heat from  the containment from the  containment can be can  be opened opened and and closed,  or pressure closed, or  pressure at  at which which SRVs SRVs can    be opened can be  opened and and will will remain remain open open (Figure (Figure 2).
2).
DE0P-01-UG IOEOP-01-UG                                      Rev.
Rev. 5555                              Page 71 Page      of 151 71 of    151  I


ATTACHMENT 5 G)CDQ Page 17 of 27      4
HH r
                                                                                                                .mgm 01 i                  k) N) Z D C -.x FIGURE 2 Primary Containment Pressure Limit-A
ma m
                                                                                          =                          -H          I 100  V I
:t;x
1'10 H
)OO
-tJ-U-o
.- C.,
DRYWELL PRESSURE (PSIG)
DRYWELL PRESSURE (PSIG)
( !)
N)
80 (j)
W ai 0)
Il.
V H
III .11 70 W
Hi ii.
a:
iii III.11 H
:::::I                         60 0)
I m0
CJ)
-o C)
U) w 50                            ai t:t:                                             iii
C,,
: a.                             40
0,
      ...I
-v 1
      ..J                                               ii.
-b 01
W 30 W
-s
      ~
= i.
G)CDQ D C -.x 4
.mgm k) N) Z
-H 01 I
(!)
(j)
Il. -
W a:
:::::I CJ)
U) w t:t:
: a.  
...I  
..J W  
~
il2 C
il2 C
20 N)
100 1'10 80 70 60 50 40 30 20 10 0
Hi H
ATTACHMENT 5 Page 17 of 27 FIGURE 2 Primary Containment Pressure Limit-A
10 0
*10 0
                                                        *10               0   10   20                   30               40   50       GO       70           8Q PRIMARY CONTAINMENT WATER LEVEL (FEET)                             -4                                    ma I
10 20 30 40 50 GO 70 8Q PRIMARY CONTAINMENT WATER LEVEL (FEET)
nl                    m IF USING THE FOLLOWING INSTRUMENT:
IF USING THE FOLLOWING INSTRUMENT:
                                -rT C
CAC-PI-1230 CAC-PI-4176 CAC-PR-1257-1 PCPl-A IS:
Cl) z                          H                  C)            0
70 PSIG USE THE GRAPH USE THE GR.f\\PH IOEOP-01-UG Rev. 55 Page 77 of 151 I
                                                                                                  -1            C    m                  PCPl-A IS:
 
                                                                                                                                          -o o
PCPL-A j DRYWELL PHESS PCPL.A VENT ThE PRMARV CTT RRSPCTIV& OF OFcsIrE NZUASE RAXF
                                                                                                                                          -w      r CAC-PI-1230                                                                                               70 PSIG
$ETtON I OF pcjp. i STEP BASES:
                                )OO C.,
Action to vent the primary containment is taken before drywell pressure rises to Primary Containment Pressure Limit A, defined to be the lesser of either:
                                              -tJ-U-o cci CAC-PI-4176                       .-
a.
0 Cl)cl)
The pressure capability of the containment, or b.
USE THE GRAPH
The maximum containment pressure at which vent valves sized to reject all decay heat from the containment can be opened and closed, or C.
:t;x mmo CAC-PR-1257-1                                                                                             USE THE GR.f\PH HH IOEOP-01-UG m
The maximum containment pressure at which SRVs can be opened and will remain open, or d.
0
The maximum containment pressure at which reactor vent valves can be opened and closed.
-o                              C)                                                      Rev. 55          C,,
This action is taken to assure that the integrity of the primary containment is maintained and to prevent core damage that might be caused by the inability to vent the reactor, as necessary, to permit injection of water to cool the core.
0,                                         Page 77 of 151
The directions Lo vent before drywell pressure reaches PCPL-A allows, but does not require, venting at significantly lower pressures Early or extended venting can permit primary containment pressure reductions before significant fuel damage occurs, thereby increasing the capacity of the containment to retain lission products and reducing the radioactivity released to the environment, If the primary containment has failed, venting may also reduce the offsite dose by directing fission products through an elevated release point.
                                                                                                                                                        -v      1   -b    01
001-37.8 Rev.
                                                                                                                                                                              -s  I
I Page 33 of 58 PCPL-A STEP BASES:
rmVWllll. PRESS R~AClfeS PCPL*A VENT THE PRIMARY elMT IRRESPECTIVE OF OFFSITE "IELEASe RAtE PER SECTION 1 OF 1001"*01* SEP. 01 "CiP*HI Action to vent the primary containment is taken before drywell pressure rises to Primary Containment Pressure Limit A,defined to be tile lesser of either:
: a.
The pressure capability of the containment, or
: b.
The maximum containment pressure at whicll vent valves sized to reject all decay heat from tile containment can be opened and closed, or
: c.
The maximum containment pressure at which SRVs can be opened and will remain open, or
: d.
The maximum containment pressure at which reactor vent valves can be opened and closed.
This action is taken to assure that the integrity of the primary containment is maintained and to prevent core damage that might be caused by tile inability to vent the reactor, as necessary, to permit injection of water to cool tile core.
The directions to vent "before drywell pressure reaches pePl-A" allows, but does not require, venting at significantly lower pressures. Early or extended venting can permit primary containment pressure reductions before significant fuel damage occllrs, thereby increasing the capacity of the containment to retain fission products and reducing the radioactivity released to the environment. If the primary containment has failed, venting may also reduce the offsile dose by directing fission prodllcts throllgh an elevated release point.
1001-37.8 Rev. 4 Page 33 of 581


PCPL-A PCPL-A j
STEP PCIP.18 (continued)
DRYWELL PRESS rmVWllll. PHESS R~AClfeS PCPL.A PCPL*A VENT THE VENT  ThE PRIMARY PRMARV elMT CTT RRSPCTIV& OF IRRESPECTIVE    OF OFFSITE OFcsIrE NZUASE RAtE "IELEASe    RAXF PER
Venting of the primary containment is performed irrespective of the off-site radioactivity release rate that will occur, and defeating isolation interlocks if necessary, because the consequences of not doing so may be either severe core damage or loss of primary containment integrity and uncontrolled radioactive release much greater than might otherwise occur. Note that primary containment venting is performed only, as necessary, to restore and then maintain pressure below the limiL Primary containment venting is performed using Primary Containment Venting, EDP-O1 -SEP-01.
                                                  $ETtON 1I OF SECTION    OF 1001"*01* SEP. 01 pcjp. i "CiP*HI STEP BASES:
001-37.8 Rev. 4 Pane 34 of 58 STEP PC/P*18 (continued) venting of the primary containment is petiormed irrespective of the off-site radioactivity release rate that will occur, and defeating isolation interlocks if necessary, because the consequences of not doing so may be either severe core damage or loss of primary containment integrity and uncontrolled radioactive release much greater than might otherwise occur. Note that primary containment venting is petiormed only, as necessary, to restore and then maintain pressure below the limit Primary containment venting is performed using Primary Containment Venting, EOP-01-SEP-0'I.
Action to vent the primary containment containment isis taken before drywell pressure    pressure rises rises to Primary Primary Containment Pressure Containment    Pressure Limit Limit A,defined A, defined toto be be tile the lesser lesser of either:
1001-37.8 Rev. 4 Page 34 of 581  
: a.      The pressure capability of the containment, or
: b.     The maximum containment pressure at whicll    which vent valves sized to reject all decay heat from tile the containment can be opened and closed, or c.
C.     The maximum containment pressure at which SRVs can be opened and will remain open, or
: d.     The maximum containment pressure at which reactor vent valves can be opened and closed.
This action is taken to assure that the integrity of the primary containment is maintained and to prevent core damage that might be caused by tile          the inability to vent the reactor, as necessary, to permit injection of water to cool tile the core.
Lo vent "before The directions to      before drywell pressure reaches pePl-A"    PCPL-A allows, but does not require, venting at significantly lower pressures.
pressures Early or extended venting can permit primary containment pressure reductions before significant fuel damage occurs,           occllrs, thereby increasing the capacity of the containment to retain lission      fission products and reducing the radioactivity released to the environment.
environment, If the primary containment has failed, venting may also reduce the offsile offsite dose by directing fission prodllcts  products throllgh through an elevated release point.
001-37.8 1001-37.8                                     Rev. 4 I                      33 of 58 Page 33    581


STEP PC/P*18 STEP      PCIP.18 (continued)
PRIMARY CONTAINMENT VENTING 1.0 ENTRY CONDITIONS As dhrected by the PC/P section of Primary Contaniment Control Procedure, EOP-O2-PCCP OR As directed by the PC/H section of Primary Containment Control Procedure, EOP-02-PCCP 2.0 OPERATOR ACTIONS CO:
(continued)
2.1 IF while executing this procedure, it is recognized the actions can NOT be performed, OR will NOT be effective, THEN GO TO Containment Venting Under Conditions of Extreme Damage, OEDMG-003, if directed by the Unit SCO.
Venting of venting  of the the primary primary containment containment isis petiormed performed irrespective irrespective of of the the off-site off-site radioactivity radioactivity release rate  that release rate that will    occur, and will occur,         defeating isolation and defeating    isolation interlocks interlocks ifif necessary, necessary, because because the the consequences of consequences      of not not doing doing so so may may bebe either either severe severe core core damage damage or  or loss loss of primary of primary containment integrity containment    integrity and and uncontrolled uncontrolled radioactive radioactive release release much much greater greater than than might might otherwise occur.
CO:
otherwise  occur. NoteNote that that primary primary containment containment venting venting isis petiormed performed only,only, as as necessary, to necessary,   to restore restore andand then then maintain maintain pressure pressure below below the the limit limiL Primary containment Primary  containment venting venting is  performed using is performed    using Primary Primary Containment Containment Venting, Venting, EDP-O1 -SEP-01.
2.2 IF venting for pressure control, THEN PERFORM Section 1, on page 3.
EOP-01-SEP-0'I.
CO:
001-37.8 1001-37.8                                    Rev. 44 Rev.                                  Pane 34 Page  34 of 581 of 58
2.3 IF venting for H2/02 control, THEN PERFORM section of procedure directed by SCO.
I OEOP-Oi-SEP-Oi Rev. 24 Page 2 of 22 Categories KJA:
SG2.03.I 1 Tier / Group:
T3 RU Rating:
3.8 SRU Rating:
4.3 LP Ubj:
CLSLP3OOL*O8D Source:
NEW Cog Level:
HIGH Category 8:
PRIMARY CONTAINMENT VENTING 1.0 ENTRY CONDITIONS As directed by the PC/P section of Primary Containment Control Procedure, EOP-02-PCCP OR As directed by the PC/H section of Primary Containment Control Procedure, EOP-02-PCCP 2.0 OPERA TOR ACTIONS CO:
2.1 IF while executing this procedure, it is recognized the actions can NOT be performed, OR will NOT be effective, THEN GO TO Containment Venting Under Conditions of Extreme Damage, OEDMG-003, if dlrected by the Unit SCO.
CO:
2.2 IF venting for pressure control, THEN PERFORM Section oJ, on page 3.
CO:
2.3 IF venting for H2/02 control.. THEN PERFORM section of procedure directed by SCO.
IOEOP-ool-SEP-O'l Rev. 24 Categories KIA:
SG2.03.11 RO Rating:
3.8 LP Obj:
CLS-LP-300-L*08D Cog Level:
HIGH Tier / Group: T3 SRO Rating:


PRIMARY CONTAINMENT PRIMARY    CONTAINMENT VENTINGVENTING 1.0 1.0     ENTRY CONDITIONS ENTRY    CONDITIONS
===4.3 Source===
                    -        dhrected by As directed  by the PC/P PC/P section section ofof Primary Primary Containment Contaniment Control Control Procedure, Procedure, EOP-O2-PCCP EOP-02-PCCP OR OR
~vv Category 8:
                    -    As directed directed by by the the PC/H section of of Primary Containment Control Procedure, EOP-02-PCCP 2.0     OPERATOR OPERA  TOR ACTIONS CO:          2.1   IF while executing this procedure, IF                      procedure, it is recognized the actions        0 can NOT be performed, OR will NOT be effective, THEN GO TO Containment Venting Under Conditions of Extreme Damage, OEDMG-003, if dlrected directed by the Unit SCO.
0 0
CO:         2.2    IF venting for pressure control, THEN PERFORM Section oJ,    1,        0 on page 3.
0 Page 2 of 22 1
CO:          2.3    IF venting for H2/02 control control,. THEN PERFORM section of                 0 procedure directed by SCO.
: 99. During non-ATWS emergency conditions on Unit Two, Emergency Depressurization is required with reactor pressure at 1100 psig.
IIOEOP-ool-SEP-O'l OEOP-Oi-SEP-Oi                              Rev. 24 Rev. 24                          Page 22 of Page    of 22 22 1 Categories Categories KJA:
Which one of the following identifies the bases for the Minimum Number of SRV5 Required for Emergency Depressurization and the required procedure utilized if this number of SRVs open cannot be achieved?
KIA:            SG2.03.I SG2.03.111                                Tier Tier // Group:
The Minimum Number of SRVs Required for Emergency Depressurization is based on the low pressure ECCS system with the lowest head being capable of making up the SRV steam flow at the Minimum (1)
Group: T3 T3 RU RO Rating:
(2)
Rating:    3.8 3.8                                      SRO Rating:
Procedure is required if the minimum number of SRVs cannot be opened.
SRU    Rating: 4.3 4.3 LP Ubj:
A.
LP Obj:        CLSLP3OOL*O8D CLS-LP-300-L*08D                          Source:
(1) Reactor Flooding Pressure (2) Primary Containment Flooding B. (1) Reactor Flooding Pressure (2) Alternate Emergency Depressurization C. (1) Alternate Reactor Flooding Pressure (2) Primary Containment Flooding D (1) Alternate Reactor Flooding Pressure (2) Alternate Emergency Depressurization
Source:        ~vv NEW Cog Cog Level:
: 99. During non-ATWS emergency conditions on Unit Two, Emergency Depressurization is required with reactor pressure at 1100 psig.
Level:    HIGH HIGH                                      Category 8:
Which one of the following identifies the bases for the Minimum Number of SRVs Required for Emergency Depressurization and the required procedure utilized if this number of SRVs open cannot be achieved?
Category  8:
The Minimum Number of SRVs Required for Emergency Depressurization is based on the low pressure ECCS system with the lowest head being capable of making up the SRV steam flow at the Minimum (1)
: 99. During non-ATWS
: 99. During   non-ATWS emergency emergency conditions conditions on on Unit Unit Two, Two, Emergency Emergency Depressurization Depressurization isis required with required   with reactor reactor pressure pressure at at 1100 1100 psig.
psig.
Which one Which         of the one of   the following following identifies identifies the the bases bases for for the the Minimum Minimum Number Number of of SRVs SRV5 Required for Required     for Emergency Emergency Depressurization Depressurization and  and the the required required procedure procedure utilized utilized ifif this this number of number       SRVs open of SRVs     open cannot cannot be be achieved?
achieved?
The Minimum The    Minimum Number Number of of SRVs SRVs Required Required forfor Emergency Emergency Depressurization Depressurization is is based based on    on the  low  pressure    ECCS    system    with  the the low pressure ECCS system with the lowest head    lowest  head being being capable capable of of making making upup the the SRV steam SRV     steam flow flow at at the the Minimum Minimum (1)    (1)
(2)
(2)
(2)      Procedure is Procedure     is required required ifif the the minimum minimum number number of of SRVs SRVs cannot cannot be be opened.
Procedure is required if the minimum number of SRVs cannot be opened.
opened.
A. (1) Reactor Flooding Pressure (2) Primary Containment Flooding B. (1) Reactor Flooding Pressure (2) Alternate Emergency Depressurization C. (1) Alternate Reactor Flooding Pressure (2) Primary Containment Flooding D:' (1) Alternate Reactor Flooding Pressure (2) Alternate Emergency Depressurization  
(1) Reactor A. (1)
A.        Reactor Flooding Flooding Pressure Pressure (2)  Primary    Containment (2) Primary Containment Flooding Flooding B. (1) Reactor Flooding Pressure (2) Alternate Emergency (2)              Emergency Depressurization Depressurization C. (1) Alternate Reactor Flooding Pressure (2) Primary Containment Flooding D (1) Alternate Reactor Flooding Pressure D:'
(2) Alternate Emergency Depressurization Depressurization


Feedback Feedback KJA: SG2.04.17 KJA:  SG2.04.17 Emergency Procedures Emergency     Procedures II PlanPlan Knowledge of Knowledge     of EOP EOP terms terms andand definitions.
Feedback KJA: SG2.04.17 Emergency Procedures I Plan Knowledge of EOP terms and definitions.
definitions.
(CFR: 41.10 /45.13)
(CFR:  41.10 /45.13)
ROISRO Rating: 3.9/4.3 Objective: CLSLP300H*002
(CFR: 41.10/45.13)
: 2. Given plant conditions and the Emergency Operating Procedures, determine if execution of the Alternate Emergency Depressurization Procedure is required.
ROISRO Rating:
RO/SRO    Rating: 3.9/4.3 3.9/4.3 CLSLP300H*002 Objective: CLS-LP-300-H*002 Objective:
: 2. Given
: 2. Given plant plant conditions conditions and and the the Emergency Emergency Operating Operating Procedures, Procedures, determine determine ifif execution execution of of the the Alternate Emergency Alternate  Emergency Depressurization Depressurization Procedure Procedure isis required.
required.


==Reference:==
==Reference:==
OEOP-01-UG, Revision 55, Page 70, Attachment 5 (Definitions)
RVCP Cog Level: High Explanation:
The Minimum Number of SRVs Required for Emergency Depressurization (5) is defined to be the least number of SRVs which correspond to a Minimum Alternate Reactor Flooding Pressure sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow at the corresponding Minimum Alternate Reactor Flooding Pressure.
If the number of SRVs specified cannot be opened, the reactor must be depressurized by other means. A list of alternate systems that can be used for depressurizing the reactor is included in the Alternate Emergency Depressurization Procedure, EOP-01 -AEDP.
Distractor Analysis:
Choice A: Plausible because Minimum Reactor Flooding Pressure is easily confused with Minimum Alternate Reactor Flooding Pressure and Primary Containment Flooding requires exiting all EOPs which is wrong for the given conditions.
Choice B: Plausible because Minimum Reactor Flooding Pressure is easily confused with Minimum Alternate Reactor Flooding Pressure and AEDP is correct.
Choice C: Plausible because Minimum Alternate Reactor Flooding Pressure is correct and Primary Containment Flooding requires exiting all EOPs which is wrong for the given conditions.
Choice D: Correct Answer SRO Only Basis: Detailed knowledge of diagnostic steps and decision points in the EOPs that involve transitions to emergency contingency procedures.
Notes MINIMUM ALTERNATE FLOODING PRESSURE The lowest reactor pressure at which steam 110w through open SRVs is sufficient to preclude any clad temperature from exceeding 1500F even if the reactor core is not completely covered OEOP-Oi-UG Rev. 55 Page 69 of 151 Feedback KJA: SG2.04.17 Emergency Procedures I Plan Knowledge of EOP terms and definitions.
(CFR: 41.10/45.13)
RO/SRO Rating: 3.9/4.3 Objective: CLS-LP-300-H*002
: 2. Given plant conditions and the Emergency Operating Procedures, determine if execution of the Alternate Emergency Depressurization Procedure is required.


==Reference:==
==Reference:==
 
OEOP-01-UG, Revision 55, Page 70, Attachment 5 (Definitions)
OEOP-01-UG,      Revision 55, OEOP-01-UG, Revision       55, Page Page 70, 70, Attachment Attachment 55 (Definitions)
RVCP Cog Level: High Explanation:
(Definitions)
The Minimum Number of SRVs Required for Emergency Depressurization (5) is defined to be the least number of SRVs which correspond to a Minimum Alternate Reactor Flooding Pressure sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow at the corresponding Minimum Alternate Reactor Flooding Pressure. If the number of SRVs specified cannot be opened, the reactor must be depressurized by other means. A list of alternate systems that can be used for depressurizing the reactor is included in the Alternate Emergency Depressurization Procedure, EOP-01-AEDP.
RVCP RVCP Cog Level:
Cog          High Level: High Explanation:
Explanation:
The Minimum     Number of Minimum Number      of SRVs SRVs Required Required forfor Emergency Emergency Depressurization Depressurization (5)
(5) is is defined defined toto be be the the least least number of number      SRVs which of SRVs    which correspond correspond to to aa Minimum Minimum Alternate Alternate Reactor Reactor Flooding Flooding Pressure Pressure sufficiently low  low that the ECCS with the lowest head will be capable of making up the SRV steam flow at the corresponding    corresponding Minimum Alternate Reactor Flooding Pressure. If the number of SRVs specified cannot be opened, the be depressurized by other means. A list of alternate systems that can be used reactor must be                                                                                    used for depressurizing the reactor is included in the Alternate Emergency Depressurization depressurizing                                                              Depressurization Procedure, EOP-01 -AEDP.
EOP-01-AEDP.
Distractor Analysis:
Distractor Analysis:
Choice A: Plausible because Minimum Reactor Flooding Pressure is easily confused with Minimum Alternate Reactor Flooding Pressure and Primary Containment Flooding requires exiting all EOPs which is wrong for the given conditions.
Choice A: Plausible because Minimum Reactor Flooding Pressure is easily confused with Minimum Alternate Reactor Flooding Pressure and Primary Containment Flooding requires exiting all EOPs which is wrong for the given conditions.
Choice B: Plausible because Minimum Reactor Flooding Pressure is easily confused with Minimum Alternate Reactor Flooding Pressure and AEDP is correct.
Choice B: Plausible because Minimum Reactor Flooding Pressure is easily confused with Minimum Alternate Reactor Flooding Pressure and AEDP is correct.
Choice C: Plausible because Minimum Alternate Reactor Flooding Pressure is correct and Primary Containment Flooding requires exiting all EOPs which is wrong for the given conditions.
Choice C: Plausible because Minimum Alternate Reactor Flooding Pressure is correct and Primary Containment Flooding requires exiting all EOPs which is wrong for the given conditions.
Choice D: Correct Answer SRO Only Basis: Detailed knowledge of diagnosticdiagnostic steps and decision decision points in in the EOPs that involve involve transitions to emergency contingency procedures.
Choice D: Correct Answer SRO Only Basis: Detailed knowledge of diagnostic steps and decision points in the EOPs that involve transitions to emergency contingency procedures.
Notes Notes MINIMUM MINIMUM ALTERNATE ALTERNATE FLOODING FLOODING PRESSURE PRESSURE The    lowest reactor The lowest     reactor pressure pressure atat which  steam 110w which steam           through open flow through   open SRVs SRVs is is sufficient sufficient to to preclude preclude anyany clad clad temperature temperature from from exceeding exceeding 1500F 1500&deg;F even even ifif the the reactor reactor core core is is not not completely completely covered covered IOEOP-Oi-UG OEOP-O*!-UG                                     Rev.
Notes MINIMUM ALTERNATE FLOODING PRESSURE The lowest reactor pressure at which steam flow through open SRVs is sufficient to preclude any clad temperature from exceeding 1500&deg;F even if the reactor core is not completely covered I
Rev. 55 55                              Page 69 Page    69 of of *15"1 151  I
OEOP-O*!-UG Rev. 55 Page 69 of *15"1 I  


ATTACHMENT ATTACHMENT 55 Page Page 1010 ofof 2727 Definitions Definitions MINI MUM CORE MINIMUM      CORE FLOODING FLOODING INTERVAL INTERVAL The greatest The    greatest amount amount of of time time required required to to flood flood the the reactor reactor toto the the top top of of the the active active Tuel with   reactor pressure fuel with reactor     pressure at at the the minimum minimum reactor reactor flooding flooding pressure pressure and and atat least least the minimum the   minimum number number of     SRVs required of SRVs        required for for emergency emergency depressurization depressurization open. open.
ATTACHMENT 5 Page 10 of 27 Definitions MINI MUM CORE FLOODING INTERVAL The greatest amount of time required to flood the reactor to the top of the active Tuel with reactor pressure at the minimum reactor flooding pressure and at least the minimum number of SRVs required for emergency depressurization open.
MINIMUM INDICATED MINIMUM      INDICATED LEVEL LEVEL The highest The    highest reactor reactor water level level instrument instrument indication indication which which results results from from off-calibration instrument off-calibration     instrument run temperature conditions  conditions when reactor water level is          is actually at at the elevation of  of the instrument instrument variable leg    leg tap.
MINIMUM INDICATED LEVEL The highest reactor water level instrument indication which results from off-calibration instrument run temperature conditions when reactor water level is actually at the elevation of the instrument variable leg tap.
M1NIMUM NUMBER OF SRVS REQUIRED FOR EMERGENCY MINIMUM                                                              EMERGENCY DEPRESSURIZATION DEPRESSURIZATION The least number of SRVs which correspond to a minimum altemate                   alternate reactor flooding pressure sufficiently low that tile       the ECCS witll with the 101,vest lowest head will be capable of making up the SRV steam flow at the corresponding        corresponding minimum alternate reactor flooding pressure.
M1NIMUM NUMBER OF SRVS REQUIRED FOR EMERGENCY DEPRESSURIZATION The least number of SRVs which correspond to a minimum alternate reactor flooding pressure sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow at the corresponding minimum alternate reactor flooding pressure.
MINIMUM REACTOR FLOODING PRESSURE The minimum SRV reopening pressure; 50 psid witl)                 with 5 5 SRVs open. This pressure is utilized to assure sufficient liquid injection into the reactor to maintain SRVs open and to flood the reactor to the elevation of the main steam lines during the flooding evolution when the reactor is shutdown.
MINIMUM REACTOR FLOODING PRESSURE The minimum SRV reopening pressure; 50 psid with 5 SRVs open. This pressure is utilized to assure sufficient liquid injection into the reactor to maintain SRVs open and to flood the reactor to the elevation of the main steam lines during the flooding evolution when the reactor is shutdown.
MINIMUM SRV REOPENING PRESSURE The lowest reactor pressure at which an SRV will lufly              fully open and remain fully opened when its control switch is placed in the OPEN position.
MINIMUM SRV REOPENING PRESSURE The lowest reactor pressure at which an SRV will lufly open and remain fully opened when its control switch is placed in the OPEN position.
MINIMUM STEAM MINIMUM      STEAM COOLING COOLING REACTOR REACTOR WATER LEVEL The lowest reactor water level at which the covered          covered portion of  of the reactor core will generate generate sufficient sufficient steam steam to  to preclude preclude any any clad clad temperature in   in the the uncovered uncovered portion portion of   the core of the  core from exceeding exceeding 150DF. 1500&deg;F. This  This limit limit is is used used during during anan ATWS ATWS event event to to prevent prevent fuel fuel damage damage when  when level level isis lowered lowered below below TAFTAF (Unit (Unit 1'I only:
MINIMUM STEAM COOLING REACTOR WATER LEVEL The lowest reactor water level at which the covered portion of the reactor core will generate sufficient steam to preclude any clad temperature in the uncovered portion of the core from exceeding 150DF. This limit is used during an ATWS event to prevent fuel damage when level is lowered below TAF (Unit 1 only:
only:
Figure 18; Unit 2 only: Figure 18A).
Figure Figure 18;18; Unit Unit 22 only:
OEOP-01-UG Rev. 55 Page 70 of 151 ATTACHMENT 5 Page 10 of 27 Definitions MINIMUM CORE FLOODING INTERVAL The greatest amount of time required to flood the reactor to the top of the active fuel with reactor pressure at the minimum reactor flooding pressure and at least the minimum number of SRVs required for emergency depressurization open.
only: Figure Figure'18A).18A).
MINIMUM INDICATED LEVEL The highest reactor water level instrument indication which results from off-calibration instrument run temperature conditions when reactor water level is actually at the elevation of the instrument variable leg tap.
IOEOP-01-UG OEOP-O'l-UG                                           Rev. 55 Rev. 55                                   Page 70 Page     70 of of 151
MINIMUM NUMBER OF SRVS REQUIRED FOR EMERGENCY DEPRESSURIZATION The least number of SRVs which correspond to a minimum altemate reactor flooding pressure sufficiently low that tile ECCS witll the 101,vest head will be capable of making up the SRV steam flow at the corresponding minimum alternate reactor flooding pressure.
                                                                                                                    '15'1 I
MINIMUM REACTOR FLOODING PRESSURE The minimum SRV reopening pressure; 50 psid witl) 5 SRVs open. This pressure is utilized to assure sufficient liquid injection into the reactor to maintain SRVs open and to flood the reactor to the elevation of the main steam lines during the flooding evolution when the reactor is shutdown.
MINIMUM SRV REOPENING PRESSURE The lowest reactor pressure at which an SRV will fully open and remain fully opened when its control switch is placed in the OPEN position.
MINIMUM STEAM COOLING REACTOR WATER LEVEL The lowest reactor water level at which the covered portion of the reactor core will generate sufficient steam to preclude any clad temperature in the uncovered portion of the core from exceeding 1500&deg;F. This limit is used during an ATWS event to prevent fuel damage when level is lowered below T AF (Unit 'I only:
Figure 18; Unit 2 only: Figure'18A).
I OEOP-O'l-UG Rev. 55 Page 70 of '15'1 I  


STEPS RC/P*23 STEPS        RCIP-23 through    through RC/P*25       RC/P-25 PERFORM "ALTERNATE PERFORM     ALTERNATE EFi.ERGE.NCY EMERGENCY DEPRESSURI2.ATION DEPRESSURIZATION PROCEDURE (EOP.
STEPS RCIP-23 through RC/P-25 The Minimum Number of SRVs Required for Emergency Depressurizatiori (5) is defined to be the least number of SRVs which correspond to a Minimum Alternate Reactor Flooding Pressure sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow at the corresponding Minimum Alternate Reactor Flooding Pressure.
PROCEDURE"     (EOP. 01.
The Minimum SRV Re-opening Pressure is the lowest reactor pressure at which an SRV will remain fully open with its control switch in the open position. The accuracy of the re-opening pressure and the indication available to determine reactor pressure result in conditions such that the SRVs are not always open when the pressure indicated is 50 psig. One hundred psig has been selected as a value which can be used to determine the SRV5 have failed to function. When reactor pressure is below this value, depressurization is considered complete and reactor pressure reduction need not be augmented by use of additional systems even if less than the minimum number of SRVs are open.
: 01. AEDPI AEDP IRRSPECTNE OF IRRESPECTIVE        OFFSITE OI OPPSIlE RADIOACTIVITY REll:
If the number of SRVs specified cannot be opened, the reactor must be depressurizeci by other means. A list of alternate systems that can be used for depressurizing the reactor is included in the Alternate Emergency Depressurization Procedure, EOP-O1-AEDP. However, since event independence must be maintained and specific plant conditions cannot be presumed, no priority regarding system use is indicated. This approach provides an operator the flexibility of being able to use whatever system(s) may be most appropriate under current plant conditions.
RADIOACTIVITY    RELEASE      1 RATE
001-37.4 Rev. 8 Page 59 of 78 PERFORM ALTERNATE EFi.ERGE.NCY DEPRESSURI2.ATION PROCEDURE (EOP. 01. AEDP IRRSPECTNE OI OFFSITE RADIOACTIVITY RELEASE RATE 1
                                    .... SE RAT!:
STEP BASES:
STEPS RC/P*23 through RC/P*25 PERFORM "ALTERNATE EMERGENCY DEPRESSURIZATION PROCEDURE" (EOP. 01. AEDPI IRRESPECTIVE OF OPPSIlE RADIOACTIVITY REll:.... SE RAT!:
RCIP-ZS STEP BASES:
RCIP-ZS STEP BASES:
The Minimum Number of SRVs Required for Emergency Depressurization               Depressurizatiori (5) is defined to be the least number of SRVs which correspond to a Minimum Alternate Reactor Flooding Pressure sufficiently low that the ECCS with the lowest !lead                     head will be capable of making up the SRV steam flow at the correspondin          corresponding   g Minimum Alternate Reactor Flooding Pressure.
The Minimum Number of SRVs Required for Emergency Depressurization (5) is defined to be the least number of SRVs which correspond to a Minimum Alternate Reactor Flooding Pressure sufficiently low that the ECCS with the lowest !lead will be capable of making up the SRV steam flow at the corresponding Minimum Alternate Reactor Flooding Pressure.
The Minimum SRV Re-opening Pressure is the lowest reactor pressure at which an                                an SRV will remain fully open with its control switch in the open position. The accuracy of the re-opening pressure and the indication available to determine              detelmine reactor pressure result in conditions such that the SRVs are not always open when the pressure indicated indicated is 50 psig. One hundred  hundred psig has        has been selected as aa value which can be used    used to determine determine the SRV5  SRVs have failed to function. When reactor pressure is below                  below this value, depressurizat depressurization   ion is considered considered complete and reactor pressure reduction need             need notnot be be augmented augmented by    by use use of of additional additional systems systems even even ifif less less than than the the minimum minimum number number ofof SRVs SRVs are are open.
The Minimum SRV Re-opening Pressure is the lowest reactor pressure at which an SRV will remain fully open with its control switch in the open position. The accuracy of the re-opening pressure and the indication available to detelmine reactor pressure result in conditions such that the SRVs are not always open when the pressure indicated is 50 psig. One hundred psig has been selected as a value which can be used to determine the SRVs have failed to function. When reactor pressure is below this value, depressurization is considered complete and reactor pressure reduction need not be augmented by use of additional systems even if less than the minimum number of SRVs are open. If the number of SRVs specified cannot be opened, the reactor must be depressurized by other means. A list of alternate systems that can be used for depressurizing the reactor is included in the Alternate Emergency Depressurization Procedure, EOP-01-AEDP. However, since event independence must be maintained and specific plant conditions cannot be presumed, no priority regarding system use is indicated. This approach provides an operator the flexibility of being able to use whatever system(s) may be most appropriate under current plant conditions.
open. IfIf the number number of      of SRVs SRVs specified cannot be        be opened, the reactor mustmust be  depressurizec    i by  other be depressurized by other means. A list  means.          list of of alternate alternate systems systems that that can be be used used for depressurizin depressurizing     the reactor is g the                    is included included in  in the the Alternate Emergency Depressuriza Depressurization tion Procedure, Procedure, EOP-O1-AE EOP-01-AEDP.      DP. However,However, since since event event independenc independence       must be e must be maintained maintained and specific plant and specific  plant conditions conditions cannot  cannot be  be presumed, presumed, no   no priority priority regarding regarding system system use use is is indicated. This approach indicated. This     approach provides provides an    an operator operator the the flexibility flexibility of of being being able able to to use use whatever whatever system(s) system(s) may   may bebe most  most appropriate appropriate underunder current current plant plant conditions.
1001-37.4 Rev. 8 Page 59 of 781  
conditions.
001-37.4 1001-37.4                                                     Rev. 88 Rev.                                     Page 59 Page      of 78 59 of 781


ALTERNATE EMERGENCY ALTERNATE      EMERGENCY DEPRESSURIZATION DEPRESSURIZATION PROCEDURE      PROCEDURE 1.0 1.0      ENTRY CONDITIONS ENTRY    CONDITIONS As directed by As directed    by the the RC/P RCIP section section ofof Reactor Reactor Vessel Vessel Control Control Procedure, Procedure, EOP-Ol-RVCP EOP-01-RVCP OR
ALTERNATE EMERGENCY DEPRESSURIZATION PROCEDURE 1.0 ENTRY CONDITIONS As directed by the RCIP section of Reactor Vessel Control Procedure, EOP-Ol-RVCP OR As directed by the RC/P section of Level/Power Control, EOP-O1-LPC OR As directed by SAMG Primary Containment Flooding, SAMG-Ol 2.0 OPERATOR ACTIONS NOTE:
                -    As directed As  directed byby the the RC/P RC/P section section ofof Level/Power Level/Power Control, Control, EOP-01-LPC EOP-O1-LPC OR OR As directed by As directed       SAMG Primary by SAMG     Primary Containment Containment Flooding, Flooding, SAMG-O'l SAMG-Ol 2.0   OPERATOR ACTIONS NOTE:     Manpower:                                     '1I Control Operator
Manpower:
                                                                  '1I Auxiliary Operator
I Control Operator I Auxiliary Operator I Independent Verifier Special equipment:
                                                                  'II Independent Verifier equipment:
4 jumpers (32, 33, 34, and 35:
Special eqUipment:                            4 jumpers (32, 33, 34, and 35}      35:
1 Ilathead screwdriver I locking screwdriver tape NOTE:
11 flathead Ilathead screwdriver 1I locking screwdriver tape NOTE:       Performance of this procedure will affect any main steam line leakage control pathways established by EOP-O'l-SEP-'!'L EOP-Oi -SEP-i t 2.1 2,1      EVACUATE tile    the Unit 1 1 and 2 Turbine Buildings using the following actions:
Performance of this procedure will affect any main steam line leakage control pathways established by EOP-Oi -SEP-i t 2.1 EVACUATE the Unit 1 and 2 Turbine Buildings using the following actions:
CO:           2.11 2.1:1        SOUND the Unit 1    1 and Unit 22 Turbine Building evacuation alarms AND ANNOUNCE the o
CO:
evacuation.
2.11 SOUND the Unit 1 and Unit 2 Turbine Building evacuation alarms AND ANNOUNCE the evacuation.
CO:           2.1.2         REQUEST the SCO to notify the TSC that the Turbine Building Building is being evacuated due      due to to potential o
CO:
high high radiation conditions conditions during during the the alternate alternate emergency clepressuriza depressurization.
2.1.2 REQUEST the SCO to notify the TSC that the Turbine Building is being evacuated due to potential high radiation conditions during the alternate emergency clepressurization.
tion.
CO:
2.2 IF either Unit 1 or Unit 2 Turbine Building ventilation is in service in the once-through lineup, THEN SECURE that units turbine building ventilation (OP-37.3).
OEOP-Oi-AEDP Rev. 18 Page 2 of 16 Categories K/A:
SG2.04.17 Tier/Group:
T3 RO Rating:
3.9 SRO Rating:
4.3 LP Obj:
CLSLP3OOH*OO2 Source:
NEW Cog Level:
HIGH Category 8:
Y ALTERNATE EMERGENCY DEPRESSURIZATION PROCEDURE 1.0 ENTRY CONDITIONS As directed by the RC/P section of Reactor Vessel Control Procedure, EOP-01-RVCP OR As directed by the RC/P section of Level/Power Control, EOP-01-LPC OR As directed by SAMG Primary Containment Flooding, SAMG-O'l 2.0 OPERATOR ACTIONS NOTE:
Manpower:
Special eqUipment:
'1 Control Operator
'1 Auxiliary Operator
'I Independent Verifier 4 jumpers (32, 33, 34, and 35}
1 flathead screwdriver 1 locking screwdriver tape NOTE:
Performance of this procedure will affect any main steam line leakage control pathways established by EOP-O'l-SEP-'!'L 2,1 EVACUATE tile Unit 1 and 2 Turbine Buildings using the following actions:
CO:
CO:
CO:       2.2 2.2     IF either IF either Unit service service in Unit 11 or or Unit Unit 22 Turbine the once-through in the Turbine Building once-througlliineup, Building ventilation lineup, THEN ventilation is THEN SECURE SECURE that is in that in          o units units' turbine turbine building building ventilation ventilation (OP-37.3).
CO:
(OP-37.3).
CO:
IOEOP-Oi-AE OEOP-O'I-AEDP DP                              Rev. 18 Rev.   '18                                     Page 22 of Page        161 of 16 Categories Categories K/A:
2.2 2.1:1 2.1.2 SOUND the Unit 1 and Unit 2 Turbine Building evacuation alarms AND ANNOUNCE the evacuation.
KIA:          SG2.04.17 SG2.04.17                                   Tier/Group:
REQUEST the SCO to notify the TSC that the Turbine Building is being evacuated due to potential high radiation conditions during the alternate emergency depressurization.
Tier / Group:      T3 T3 RO RO Rating:
IF either Unit 1 or Unit 2 Turbine Building ventilation is in service in the once-througlliineup, THEN SECURE that units' turbine building ventilation (OP-37.3).
Rating:  3.9 3.9                                         SRO SRO Rating:
I OEOP-O'I-AEDP Rev. '18 Categories KIA:
Rating:    4.3 4.3 LP Obj:
SG2.04.17 RO Rating:
LP Obj:      CLSLP3OOH    *OO2 CLS-LP-300-H*002                             Source:
3.9 LP Obj:
Source:            ~VV NEW Cog Cog Level:
CLS-LP-300-H*002 Cog Level:
Level:  HIGH mGH                                         Category Category 8:8:     YY
mGH Tier / Group: T3 SRO Rating:  


100. An 100.        ATWS has An ATWS        has occurred occurred on  on Unit Unit Two:
===4.3 Source===
Two:
~VV Category 8:
ARI has ARI    has been been actuated.
Y o
actuated.
o o
No  blue  lights  are No blue lights are lit  lit on on the  Full Core the Full  Core Display.
Page 2 of 161
Display.
Suppression Pool Suppression        Pool Temperature Temperature isis 112  112&deg;0 F.
F.
The 2A The    2A SLC SLC pump pump has has aa red red light light indication.
indication.
The 2B The    2B SLC SLC pump pump has has aa green green light    indication light indication The SLC The    SLC AA Squib Squib Valve Valve Continuity Continuity white white light light isis lit lit The SLC The    SLC BB Squib Squib Valve Valve Continuity Continuity white white light light isis extinguished.
extinguished.
Which one Which    one ofof the the following following identifies identifies the the procedure procedure that  that an an AO AC would would bebe directed directed to to perform    based    on  the  above  conditions    and perform based on the above conditions and the resultant      the  resultant effect effect of of those those actions?
actions?
A'!  Perform LEP-02, A Perform      LEP-02, Section Section 22 toto insert insert control control rods rods in    in order order to to shutdown shutdown thethe reactor reactor by by venting  the  Scram    Air venting the Scram Air Header. Header.
B. Perform LEP-02, B. Perform      LEP-02, Section Section 66 toto insert insert control control rods rods in      order to in order to shutdown shutdown thethe reactor reactor by by venting the overpiston area of the control rods.
C. Perform LEP-03, Section 2 to inject      inject boron to shutdown the reactor using RCIC.
D. Perform LEP~03, D.              LEP03, Section 3 to inject boron to shutdown the reactor using RWCU via the SLC tank.


Feedback Feedback K/A: SG2.04.35 KIA:    SG2.04.35 Emergency Procedures Emergency        Procedures II PlanPlan Knowledge of Knowledge       of local local auxiliary auxiliary operator operator tasks tasks during during an an emergency emergency and and thethe resultant resultant operational operational effects effects.
100. An ATWS has occurred on Unit Two:
(CFR: 41.10/43.5/45.13)
ARI has been actuated.
(CFR:     41.10 /43.5/45.13)
No blue lights are lit on the Full Core Display.
RO/SRO Rating:
Suppression Pool Temperature is 112&deg; F.
RO/SRO      Rating: 3.8/4.0 3.8/4.0 CLSLP300J*005 Objective: CLS-LP-300-J*005 Objective:
The 2A SLC pump has a red light indication.
: 5. Given plant
The 2B SLC pump has a green light indication The SLC A Squib Valve Continuity white light is lit The SLC B Squib Valve Continuity white light is extinguished.
: 5. Given           conditions and plant conditions   and the the Local Local Emergency Emergency Procedures, Procedures, determine determine which which sections sections ofof the the Alternate Control Alternate  Control Rod Rod Insertion Insertion Procedure Procedure should should be be utilized utilized for for Control Control RodRod Insertion Insertion (EOP-01 -LEP-02).
Which one of the following identifies the procedure that an AC would be directed to perform based on the above conditions and the resultant effect of those actions?
(EOP-01-LEP-02).
A Perform LEP-02, Section 2 to insert control rods in order to shutdown the reactor by venting the Scram Air Header.
: 4. Given
B. Perform LEP-02, Section 6 to insert control rods in order to shutdown the reactor by venting the overpiston area of the control rods.
: 4. Given plant plant conditions conditions andand the the Local Local Emergency Emergency Procedures, Procedures, determine determine which which method method of of the the Alternate Alternate Boron Injection Boron   Injection is  appropriate (EOP-01-LEP-03) is appropriate     (EOP-01-LEP-03)
C. Perform LEP-03, Section 2 to inject boron to shutdown the reactor using RCIC.
D. Perform LEP03, Section 3 to inject boron to shutdown the reactor using RWCU via the SLC tank.
100. An ATWS has occurred on Unit Two:
ARI has been actuated.
No blue lights are lit on the Full Core Display.
Suppression Pool Temperature is 112 0 F.
The 2A SLC pump has a red light indication.
The 2B SLC pump has a green light indication The SLC A Squib Valve Continuity white light is lit The SLC B Squib Valve Continuity white light is extinguished.
Which one of the following identifies the procedure that an AO would be directed to perform based on the above conditions and the resultant effect of those actions?
A'! Perform LEP-02, Section 2 to insert control rods in order to shutdown the reactor by venting the Scram Air Header.
B. Perform LEP-02, Section 6 to insert control rods in order to shutdown the reactor by venting the overpiston area of the control rods.
C. Perform LEP-03, Section 2 to inject boron to shutdown the reactor using RCIC.
D. Perform LEP~03, Section 3 to inject boron to shutdown the reactor using RWCU via the SLC tank.
 
Feedback K/A: SG2.04.35 Emergency Procedures I Plan Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects (CFR: 41.10 /43.5/45.13)
RO/SRO Rating: 3.8/4.0 Objective: CLSLP300J*005 5.
Given plant conditions and the Local Emergency Procedures, determine which sections of the Alternate Control Rod Insertion Procedure should be utilized for Control Rod Insertion (EOP-01 -LEP-02).
: 4. Given plant conditions and the Local Emergency Procedures, determine which method of the Alternate Boron Injection is appropriate (EOP-01-LEP-03)


==Reference:==
==Reference:==
OEOP-01 -LEP-02 Cog Level: High Explanation:
Based on the conditions given, determines that scram valves have not opened (no blue lights on full core display) and that Boron is injecting with A pump running (red light on) and B squib valve opened (white light extinguished) so LEP-03 is not required. The pumps discharge into a c9rflmoIheader before going to the squib valves. Requires assessment of alternate control rod insertion seOtions anddetermines venting the scram air header is appropriate.
Distractor Analysis:
Choice A: Correct Answer Choice B: Plausible because venting of the over piston area will insert the control rods but would be the inappropriate decision for rod insertion given the conditins. The operational effect is reactor shutdown with control rod insertion.
Choice C: Plausible because suppression pool temperature is greater than 1100 F and boron injection is required. With A pump running but the A squib valve not open and no B pump a common misconception is that SLC flow will not occur to the Reactor. this would be correct under different conditions in the stem. The operational effect is reactor shutdown with boron injection.
Choice D: Plausible because suppression pool temperature is greater than 1100 F and boron injection is required. With A pump running but the A squib valve not open and no B pump a common misconception is that SLC flow will not occur to the Reactor. this would be correct under different conditions in the stem. The operational effect is reactor shutdown with boron injection.
SRO Only Basis: Assessing plant conditions and prescribing a section of a procedure with which to proceed.
Notes Feedback KIA: SG2.04.35 Emergency Procedures I Plan Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.
(CFR: 41.10/43.5/45.13)
RO/SRO Rating: 3.8/4.0 Objective: CLS-LP-300-J*005
: 5. Given plant conditions and the Local Emergency Procedures, determine which sections of the Alternate Control Rod Insertion Procedure should be utilized for Control Rod Insertion (EOP-01-LEP-02).
: 4. Given plant conditions and the Local Emergency Procedures, determine which method of the Alternate Boron Injection is appropriate (EOP-01-LEP-03)


==Reference:==
==Reference:==
OEOP-01-LEP-02 Cog Level: High Explanation:
Based on the conditions given, determines that scram valves have not opened (no blue lights on full core display) and that Boron is injecting with A pump running (red light on) and B squib valve opened (white light extinguished) so LEP-03 is not required. The pumps discharge into a c9ffi..rnQ!lJl~~der before going to the squib valves. Requires assessment of alternate control rod insertion seCtions and"determines venting the scram air header is appropriate.
Distractor Analysis:
Choice A: Correct Answer Choice B: Plausible because venting of the over piston area will insert the control rods but would be the inappropriate decision for rod insertion given the conditins. The operational effect is reactor shutdown with control rod insertion.
Choice C: Plausible because suppression pool temperature is greater than 110&deg; F and boron injection is required. With A pump running but the A squib valve not open and no B pump a common misconception is that SLC flow will not occur to the Reactor. this would be correct under different conditions in the stem. The operational effect is reactor shutdown with boron injection.
Choice D: Plausible because suppression pool temperature is greater than 110&deg; F and boron injection is required. With A pump running but the A squib valve not open and no B pump a common misconception is that SLC flow will not occur to the Reactor. this would be correct under different conditions in the stem. The operational effect is reactor shutdown with boron injection.
SRO Only Basis: Assessing plant conditions and prescribing a section of a procedure with which to proceed.
Notes


OEOP-01 -LEP-02 OEOP-01-LEP-02 Cog Level: High Cog Explanation:
21 INSERT control rods by one or more of the following methods:
Explanation:
2.71 DE-ENERGIZE the scram pilot valve solenoids AND VENT the scram air header Section 2 on Page 9 2.7.2 RESET RPS AND INITIATE a manual scram. Section 3 on Page 14.
Based on Based    on the conditions given, determines that scram valves have not opened (no blue lights on full core                  core display)  and that Boron is injecting with A pump running (red light on) and B squib valve opened (white display) and light extinguished) light  extinguished) so so LEP-03 is not required. The pumps discharge into a c9ffi..rnQ!lJl~~der c9rflmoIheader before going to  the squib to the  squib valves.
2.7.3 SCRAM indMdual rods with the scram test switches, Section 4 on Page 17.
valves. Requires assessment of alternate control rod insertion seCtions      seOtions and"determines anddetermines venting the scram air header is appropriate.
2.7.4 INSERT control rods with the Reactor Manual Control LI System, Section 5 on Page 21.
venting Distractor Analysis:
2.7.5 VENT the over piston area of control rods, Section 6 on Page 22.
Distractor  Analysis:
OEOP-O1-LEP-02 Rev. 26 Page 3 of 29 2.2 INJECT boron with one or more of the following systems:
Choice A: Correct Answer Choice Choice Choice B: Plausible because venting of the over piston area will insert the control rods but would be the inappropriate decision for rod insertion given the conditins. The operational effect is reactor shutdown with control rod insertion.
Choice Choice C: Plausible because because suppression suppression pool pool temperature is    greater than 1100 is greater          110&deg; FF and boron injection injection isis required.
required. With  A  pump    running  but  the A squib  valve  not open and not        and no B  B pump a common misconceptio misconception  n is is that SLC flow will not  not occur to the Reactor. this would be    be correct under under different different conditions conditions in in the the stem. The operational stem. The    operational effect effect is is reactor shutdown shutdown with with boron boron injection.
injection.
Choice Choice D: D: Plausible Plausible because because suppression suppression pool pool temperature temperature is is greater  than 1100 greater than    110&deg; FF and and boron boron injection injection isis required.
required. With With A pump pump running      but the running but    the A squib squib valve valve not not open open and and no no BB pump pump aa common common misconceptio misconception    is that n is  that SLC SLC flow flow will will not not occur occur to to the the Reactor.
Reactor. this this would would bebe correct correct under under different different conditions conditions in in the the stem. The operational stem. The    operational effect effect is is reactor reactor shutdown shutdown with with boron boron injection.
injection.
SRO SRO OnlyOnly Basis:
Basis: Assessing Assessing plant plant conditions conditions and and prescribing prescribing aa section section ofof aa procedure procedure with with which which to to proceed.
proceed.
Notes Notes
 
21 2.7      INSERTcontrol INSERT    controlrods rodsbybyoneoneorormore moreofofthe  thefollowing following metho ds:
methods:
2.71 2.7.1      DE-ENERGIZEthe DE-ENERGIZE VENT    thescram thescram scramair scrampilot airheader, pilotvalve headerSection valvesolenoids Section22on solenoidsAND  AND       o VENT the                                                 Page9.9 on Page 2.7.2 2.7.2      RESETRPS RESET 33on on Page RPSAND Page '14.
AND INITIATE 14.
INITIATEaamanual manual scram,scram. Section Section    o 2.7.3 2.7.3       SCRAM individual SCRAM Sectio n  4 indMdual rods on Page Section 4 on Page H.
rods with 17.
with the the scram scram testtest switches, switches,      o 2.7.4 2.7.4       INSERT control INSERT System control rods System,, Section rods with Section 55 on with the on Page the Reactor Page 21.
21.
Reactor ManualManual Control Control    oLI 2.7.5 2.7.5     VENT the VENT on Page on the over Page 22.
over piston 22.
piston area area of of control control rods, rods, Section Section 66      o IOEOP-01-LEP-02 OEOP-O1-LEP-02                                 Rev. 26 Rev. 26                                  Page 3 of Page  of 29 I 2.2 2.2    INJECT boron INJECT     boron with one or more of the following following systems:
systems:
ENOTE:
ENOTE:
NOTE:      System (s) should System(s)    should be selecte selected d in order listed and based upon system I            availab ility  and access availability and           ibility.
System(s) should be selected in order listed and based upon system II availability and accessibility.
accessibility.
CO:
CO:
CO:            -        CRD, CRD, Sectio Section n 11 on on page page 33                                          o LI NOTE NOTE::    HPCI/ RCIC should HPCIIRCIC      should bebe used used only only ifjf suction suction is is from from the the CST.
CRD, Section 1 on page 3 LI NOTE:
CST.
HPCI/RCIC should be used only if suction is from the CST.
CO:
CO:
CO:            -        HPCl/     RCICS HPCIIRCIC,        ection2 Section      onpage 2 on    page14  14                            o LI CO:
HPCl/RCICSection2onpage14 LI CO:
CO:          -        RWCU RWCU via  via SLC SLC tank, tank, Section Section 33 on  on page page 21 21                o LI CO:
RWCU via SLC tank, Section 3 on page 21 LI CO:
CO:          -        RWCU RWCU with with borax, borax, Section Section 44 on  on page page31  31                   o LI IOEOR  -Di-LEP-03 OEOP-01-LEP-03                                  Rev.
RWCU with borax, Section 4 on page 31 LI OEOR-Di-LEP-03 Rev. 27 Page 2 of 41 2.7 INSERT control rods by one or more of the following methods:
Rev. 27 27                                  Page22of Page     4-1 of41  I
2.7.1 2.7.2 2.7.3 2.7.4 2.7.5 DE-ENERGIZE the scram pilot valve solenoids AND VENT the scram air header, Section 2 on Page 9.
RESET RPS AND INITIATE a manual scram, Section 3 on Page '14.
SCRAM individual rods with the scram test switches, Section 4 on Page H.
INSERT control rods with the Reactor Manual Control System, Section 5 on Page 21.
VENT the over piston area of control rods, Section 6 on Page 22.
I OEOP-01-LEP-02 Rev. 26 2.2 INJECT boron with one or more of the following systems:
o o
o o
o Page 3 of 29 I NOTE:
System(s) should be selected in order listed and based upon system availability and accessibility.
CO:
CRD, Section 1 on page 3 NOTE:
HPCIIRCIC should be used only jf suction is from the CST.
CO:
HPCIIRCIC, Section 2 on page 14 CO:
RWCU via SLC tank, Section 3 on page 21 CO:
RWCU with borax, Section 4 on page 31 I OEOP-01-LEP-03 Rev. 27 o
o o
o Page 2 of 4-1 I  


f002A SQUIB F004A PUMP COO1A F002B PUMP
K/A:
~026 C001B SQUIB F004B Categories Categories K/A:
SG2.04.35 RORating:
KIA:          SG2.04.35 SG2.04.35        Tier / Group: T3 RORating:
3.8 LP Obj:
RORating:    3.8 3.8              SRO Rating: 4.0 SRORating:
CLSLP3OOJ*OO5 Cog Level:
LP Obj:
NIGH Tier / Group:
LPObj:       CLSLP3OOJ  *OO5 CLS-LP-300-J*005 Source:       NEW Cog Cog Level:
T3 SRO Rating:
Level:  NIGH HIGH            Category 8:}}
 
==4.0 Source==
NEW Category 8:
Categories
~026 Categories KIA:
RORating:
LPObj:
Cog Level:
SQUIB F004A SQUIB F004B SG2.04.35 3.8 CLS-LP-300-J*005 HIGH Tier / Group: T3 SRORating:  
 
==4.0 Source==
NEW Category 8:
f002A F002B PUMP COO1A PUMP C001B}}

Latest revision as of 04:39, 14 January 2025

Initial Exam 2010-301 Draft SRO Written Exam
ML101590078
Person / Time
Site: Brunswick  Duke Energy icon.png
Issue date: 05/03/2010
From:
NRC/RGN-II
To:
Progress Energy Carolinas
References
50-324/10-301, 50-325/10-301
Download: ML101590078 (122)
v • d • e


Text

76. Which one of the following correctly completes the statement below?

Technical Specifications do NOT require the RWCU isolation from the SLC control switch in Mode (1) due to (2)

A (1) 3 (2) control rods are not able to be withdrawn since the reactor mode switch must be in the shutdown position and a control rod block is applied B. (1)3 (2) the operability of each individual control rod scram accumulator is required which will ensure that the control rods can be inserted C. (1)5 (2) control rods are not able to be withdrawn since the reactor mode switch must be in the shutdown position and a control rod block is applied D. (1)5 (2) the operability of each individual control rod scram accumulator is required which will ensure that the control rods can be inserted Feedback K/A: 204000 G2.02.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

Reactor Water Cleanup System (CFR: 41.5 / 41.7 /43.2)

There are no safety limits associated with RWCU system, so question is written directly to the TS.

ROISRO Rating: 3214.2 Objective:

Reference:

Cog Level: Low Explanation: Are no safety limits for RWCU. SLC is required in mode 3 (RO knowledge) and the bases for the mode 3 requirement is SRO knowledge.

Distractor Analysis:

Choice A: Correct answer from the bases document.

Choice B: Plausible becasue this is the bases for Mode 4/5 from the bases document.

Choice C:

Choice D: Plausible because the scram accumulators are capable of inserting the control rods with low reactor pressure conditions, but the accumulators are not required to be operable in Mode 3.

SRO Basis: 10 CFR 55.43(b)-2, Facility operating limitations in the TS and their bases.

Knowledge of TS bases that is required to analyze TS required actions and terminology.

76. Which one of the following correctly completes the statement below?

Technical Specifications do NOT require the RWCU isolation from the SLC control switch in Mode (1) due to (2)

A!I (1) 3 (2) control rods are not able to be withdrawn since the reactor mode switch must be in the shutdown position and a control rod block is applied B. (1) 3 (2) the operability of each individual control rod scram accumulator is required which will ensure that the control rods can be inserted C. (1) 5 (2) control rods are not able to be withdrawn since the reactor mode switch must be in the shutdown position and a control rod block is applied D. (1) 5 (2) the operability of each individual control rod scram accumulator is required which will ensure that the control rods can be inserted Feedback KIA: 204000 G2.02.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

Reactor Water Cleanup System (CFR: 41.5/41.7/43.2)

There are no safety limits associated with RWCU system, so question is written directly to the TS.

RO/SRO Rating: 3.2/4.2 Objective:

Reference:

Cog Level: Low Explanation: Are no safety limits for RWCU. SLC is required in mode 3 (RO knowledge) and the bases for the mode 3 requirement is SRO knowledge.

Distractor Analysis:

Choice A: Correct answer from the bases document.

Choice B: Plausible becasue this is the bases for Mode 4/5 from the bases document.

Choice C:

Choice D: Plausible because the scram accumulators are capable of inserting the control rods with low reactor pressure conditions, but the accumulators are not required to be operable in Mode 3.

SRO Basis: 10 CFR 55.43(b)-2, Facility operating limitations in the TS and their bases.

Knowledge of TS bases that is required to analyze TS required actions and terminology.

Notes RECTC MODE AVERA3E REACTOR MDDE TTLE S1TCH POSfIICN COOLANT EMPERTURE (F:

1 Pr Cperatbn Run 2

S1aiup Reftle? o-SatupHot NA Sanby 3

Hot hutdon Sutcii

> 21.2 4

Cold Sutdcn

iuchr, 212 5

Reje1g Sttubcii or ReieI From Bases 3.3.6.1 One channel of the SLC System Initiation Function is available and required to be OPERABLE only in MODES 1 and 2, since these are the only MODES where the reactor can be critical, and these MODES are consistent with the Applicability for the SLC System (LCO 3.1.7).

From bases 3.1.7 APPLICABILITY In MODES I and 2, shutdown capability is required. In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in the shutdown position and a control rod block is applied. This provides adequate controls to ensure that the reactor remains subcritical. In MODE 6, only a single control rod can be withdrawn from a core cell containing fuel assemblies. Determination of adequate SDM (LCD 3.1.1, SHUTDOWN MARGIN (SDM)) ensures that the reactor will not become critical with the analytically determined strongest control rod withdrawn.

Therefore, the SLC System is not required to be OPERABLE when only a single control rod can be withdrawn.

Categories K/A:

204000 G2.02.25 Tier / Group:

T2G2 RO Rating:

3.2 SRO Rating:

4.2 LP Obj:

05-11 Source:

NEW Cog Level:

LOW Category 8:

Notes RE~.CTOR MODE AVERIo.GE REACTOR MODE TiTLE SWlTCH POSfllON COOLANTIEMPERA Th'RE

(=-l-t,,'

1 P'))VI,'Er Operatoo Run N."'.

2 Slartup Refuer"'),or S;,artup;'HD1 Nt.!;.

S1aoooji 3

Hot Shutdall.m'1I)

Shutdb~\\T1

>2'&2 4

Cold Silub:i'o\\'ltnt*,

Sh,utdbltn S; 21:2 5

RefueEflg(ll1 From Bases 3.3.6.1 From bases 3.1.7 APPLICABILITY Categories Silutdoltn or R~ilel NI!...

One channel of the SLC System Initiation Function is available and required to be OPERA.BLE only in MODES 'I and 2, since these are the only MODES where the reactor can be critical, and these MODES are consistent with the Applicability for the SLC System (LCO 3.'1.7).

[n MODES 'I and 2, shutdown capability is required. In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in the shutdown position and a control rod block is applied. This provides adequate controls to ensure that the reactor remains sub criticaL In MODE 5, only a single control rod can be withdrawn from a core cell containing fuel assemblies. Determination of adequate SDM {LCO 3. '1.'1,

"SHUTDOWN MARGIN (SDM}") ensures that the reactor... vill not become critical with the analytically determined strongest control rod withdrawn.

Therefore, the SLC System is not required to be OPERA.BLE when only a single control rod can be withdrawn.

KIA:

204000 G2.02.25 Tier / Group: T2G2 RORating:

3.2 SRORating

4.2 LP Obj:

05-11 Source:

NEW Cog Level:

LOW Category 8:

77. Unit One is operating at full power when the following plant conditions occur:

- Load Reject Signal received

- Line 31 (Whiteville Line) PCBs red lights are lit

- Line 31 (Whiteville Line) white VOLT lights are not lit

- All other line PCBs green lights are lit

- 230 KV BUS IA BUS POT UNDERVOLTAGE is in alarm

- 230 KV BUS lB BUS POT UNDERVOL TAGE is in alarm Which one of the following identifies the initial RPS trip signal and the procedure which contains the guidance to trip the Whiteville Line PCBs?

A Control Valve Fast Closure; OAOP-36.1, Loss of Any 4160V Buses or 480V E-Buses.

B. Stop Valve Closure; OAOP-36. 1, Loss of Any 41 60V Buses or 480V E-Buses.

C. Control Valve Fast Closure; OAOP-22, Grid Instability.

D. Stop Valve Closure; OAOP-22, Grid Instability.

77. Unit One is operating at full power when the following plant conditions occur:

- Load Reject Signal received

- Line 31 (Whiteville Line) PCBs red lights are lit

- Line 31 (Whiteville Line) white VOLT lights are not lit

- All other line PCBs green lights are lit

- 230 KV BUS 1A BUS POT UNDER VOL TAGE is in alarm

- 230 KV BUS 1B BUS POT UNDERVOLTAGE is in alarm Which one of the following identifies the initial RPS trip signal and the procedure which contains the guidance to trip the Whiteville Line PCBs?

A'I Control Valve Fast Closure; OAOP-36.1, Loss of Any 4160V Buses or 480V E-Buses.

B. Stop Valve Closure; OAOP-36.1, Loss of Any 4160V Buses or 480V E-Buses.

C. Control Valve Fast Closure; OAOP-22, Grid Instability.

D. Stop Valve Closure; OAOP-22, Grid Instability.

Feedback K/A: 212000 A2.12 Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Main turbine stop control valve closure (CFR: 41.5/ 45.6)

RO/SRO Rating: 4.0/4.1 Objective: CLS-LP-03, Obj. 8.

List the RPS trip signals, including setpoints and how/when each signal is bypassed.

Reference:

SD-03 Reactor Protection System, section 3.1 RPS Trips Cog Level High Explanation:

A load reject signal at any reactor power level will cause a turbine control valve fast closure scram. The load reject signal does not input into the turbine stop valve closure scram logic. During a loss of offsite power, if the grid is lost all PCBs are opened per OAOP-36. 1.

Distractor Analysis:

Choice A: Correct answer, see explanation Choice B: Incorrect Load reject initiates a TCV fast closure scram not a TSV. A misconception of the difference between TCV and TSV scrams may cause a student to select this answer.

Choice C: Incorrect. OAOP-22 does not have an action for loss of grid only for degraded.

Choice D: Incorrect. Load reject initiates a TCV fast closure scram not a TSV. A misconception of the difference between TCV and TSV scrams may cause a student to select this answer. OAOP-22 does not have an action for loss of grid only for degraded.

SRO Basis: 10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Feedback KIA: 212000A2.12 Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Main turbine stop control valve closure (CFR: 41.5 145.6)

RO/SRO Rating: 4.0/4.1 Objective: CLS-LP-03, Obj. 8.

List the RPS trip signals, including setpoints and how/when each signal is bypassed.

Reference:

SD-03 Reactor Protection System, section 3.1 RPS Trips Cog Level High Explanation:

A load reject signal at any reactor power level will cause a turbine control valve fast closure scram. The load reject signal does not input into the turbine stop valve closure scram logic. During a loss of offsite power, if the grid is lost all PCBs are opened per OAOP-36.1.

Distractor Analysis:

Choice A: Correct answer, see explanation Choice B: Incorrect Load reject initiates a TCV fast closure scram not a TSV. A misconception of the difference between TCV and TSV scrams may cause a student to select this answer.

Choice C: Incorrect. OAOP-22 does not have an action for loss of grid only for degraded.

Choice D: Incorrect. Load reject initiates a TCV fast closure scram not a TSV. A misconception of the difference between TCV and TSV scrams may cause a student to select this answer. OAOP-22 does not have an action for loss of grid only for degraded.

SRO Basis: 10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Notes An example of Turbine Control Valve Fast Closure is a load reject.

The definition of a load reject is greater than 40% mismatch between electrical output and mechanical input as sensed by generator stator amps and the Cross Around Piping pressure. This is to prevent excessive overspeed of the Turbine on loss of load.

A load reject signal will energize the fast acting Solenoid Valves on the control valve actuators, which removes hydraulic trip fluid pressure. The trip signal comes from pressure switches on the Vast acting trip control (FATC) supply to the control valve disc dumps (refer to EHC Hydraulics). Loss of this pressure will cause a rapid closure of the control valves. Circuitry is designed such that the pressure switch on either control valve 1 or 3 will trip RPS Trip System A. Either control valve 2 or 4 will trip RPS Trip System B.

These switches will also provide a Scram signal on loss of hydraulic trip fluid pressure when a load reject signal is not present - loss of hydraulic fluid pressure can result in a fast closure of the control valves SD-03 Rev. 9 Page 21 of 89 7.

IF the SAT was lost due to loss of power on the Progress Energy System, THEN PERFORM the following:

a.

PLACE AUTO RELOSE switches in MANUAL.

LI b.

PLACE transmission line PCB SUPERVISORY LOCAL/REMOTE switches in LOCAL.

c, TRIP all transmission line PCBs.

LI OAOP-36.2 Rev.

I Page 4 of 196 Categories K/A:

212000 A2.12 Tier / Group:

T2G1 RO Rating:

4.0 SRO Rating:

4.1 LP Obj:

03-08 Source:

PREV Cog Level:

HIGH Category 8:

Y Notes 1 SD-03 An example of Turbine Control Valve Fast Closure is a load reject.

The definition of a load reject is greater than 40% mismatch betvveen electrical output and mechanical input as sensed by generator stator amps and the Cross Around Piping pressLire. This is to prevent excessive overs peed of the Turbine on loss of load.

A load reject signal will energize the fast acting Solenoid Valves on the control valve actuators, which removes hydraulic trip fluid pressure. The trip signal comes from pressure switches on the fast acting trip control (FATC) supply to the control valve disc dumps (refer to EHC Hydraulics). Loss of this pressure will cause a rapid closure of the control valves. Circuitry is designed such that the pressure switch on either control valve 'lor 3 Will trip RPS Trip system A. Either control valve 2 or 4 will trip RPS Trip System B.

These switches will also provide a Scram signal on loss of hydraulic trip fluid pressure when a load reject signal is not present - loss of hydraulic fluid pressure can result in a fast closure of the control valves.

Rev. 9 Page 2'1 of 891 7,

IF tile SAT was lost due to loss of power on the Progress Energy System, THEN PERFORM the following:

a.

.PLACE AUTO RECLOSE switches in MANUAL.

D

b.

PLACE transmission line PCB SUPERVISORY D

LOCAUREMOTE switches in LOCAL

c.

TRIP all transmission line PCBs.

D IOAOP-36.2 Rev. 4'1 Page 4 of 1961 Categories KIA:

212000 A2.12 RO Rating:

4.0 LP Obj:

03-08 Cog Level:

HIGH Tier / Group: T2G 1 SRO Rating:

4.1 Source

PREY Category 8:

Y

78. The Unit is at 7% power during reactor startup.

The operator withdraws control rod 26-27 to position 48.

The following indications are noted:

- ROD DRIFT alarm seals in

- ROD OVER TRAVEL alarm seals in

- Rod 26-27 full core display red light out Which one of the following identifies:

(1) the indication that would be displayed on the four-rod group display and (2) the required action for the inoperable control rod lAW Technical Specification 3.1.3, Control Rod Operability?

A. (1)48 (2) Fully insert control rod 26-27 and disarm the HCU B. (1)48 (2) Verify 1 2 control rods are withdrawn and implement GP-1 1, Second Operator Rod Sequence Checkoff Sheets C (1) Blank (2) Fully insert control rod 26-27 and disarm the HCU D. (1) Blank (2) Verify 1 2 control rods are withdrawn and implement GP-1 1, Second Operator Rod Sequence Checkoff Sheets

78. The Unit is at 7% power during reactor startup.

The operator withdraws control rod 26-27 to position 48.

The following indications are noted:

- ROD DRIFT alarm seals in

- ROD OVER TRAVEL alarm seals in

- Rod 26-27 full core display red light out Which one of the following identifies:

(1) the indication that would be displayed on the four-rod group display and (2) the required action for the inoperable control rod lAW Technical Specification 3.1.3, Control Rod Operability?

A. (1) 48 (2) Fully insert control rod 26-27 and disarm the HCU B. (1) 48 (2) Verify ~12 control rods are withdrawn and implement GP-11, Second Operator Rod Sequence Checkoff Sheets C!I (1) Blank (2) Fully insert control rod 26-27 and disarm the HCU D. (1) Blank (2) Verify ~12 control rods are withdrawn and implement GP-11, Second Operator Rod Sequence Checkoff Sheets

Feedback K/A: 214000 A2.03 Ability to (a) predict the impacts of the following on the ROD POSITION INFORMATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Overtravel/in-out (CFR: 41.5 / 45.6)

RO/SRO Rating: 3.6/3.9 Objective: CLS-LP-07 Obj 5b Given plant conditions, determine if the following conditions exist: b. Indications of an uncoupled control rod.

Reference:

SD-07 page 27 TS 3.1.3 Cog Level Low Explanation:

If the control rod is in the overtravel out position, the corresponding digital indicator will be blank since the magnet will not be near any of the 00 to 48 reed switches. lAW TS the rod is declared inoperable then inserted to 00 (within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) and disarmed (within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />). TS 3.1.6 if the RWM is inoperable then if

>12 control rods are withdrawn GP-1 1 would be implemented, unless the rod is at 00 and is not intended to be moved.

Distractor Analysis:

Choice A: Plausible because the full in and 00 indications are at the same point or the exam inee may think that the rod may settle to the 48 position.

Choice B: Plausible because the full in and 00 indications are at the same point or the examinee may think that the rod may settle to the 48 position. These are TS actions for an inoperable RWM, not control rod.

Choice C: Correct answer, see explanation.

Choice D: Plausible because this is the correct indication but these are TS actions for an inoperable RWM, not control rod.

SRO Basis: 10 CFR 55.43(b)-2, Facility operating limitations in the technical specifications and their bases. Application of required actions statements.

Feedback KIA: 214000 A2.03 Ability to (a) predict the impacts of the following on the ROD POSITION INFORMATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Overtravel/in-out (CFR: 41.5 /45.6)

RO/SRO Rating: 3.6/3.9 Objective: CLS-LP-07 Obj 5b Given plant conditions, determine if the following conditions exist: b. Indications of an uncoupled control rod.

Reference:

SO-07 page 27 TS 3.1.3 Cog Level Low Explanation:

If the control rod is in the overtravel out position, the corresponding digital indicator will be blank since the magnet will not be near any of the 00 to 48 reed switches. lAW TS the rod is declared inoperable then inserted to 00 (within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) and disarmed (within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />). TS 3.1.6 if the RWM is inoperable then if

.=::12 control rods are withdrawn GP-11 would be implemented, unless the rod is at 00 and is not intended to be moved.

Oistractor Analysis:

Choice A: Plausible because the full in and 00 indications are at the same point or the examinee may think that the rod may settle to the 48 position.

Choice B: Plausible because the full in and 00 indications are at the same point or the examinee may think that the rod may settle to the 48 position. These are TS actions for an inoperable RWM, not control rod.

Choice C: Correct answer, see explanation.

Choice 0: Plausible because this is the correct indication but these are TS actions for an inoperable RWM, not control rod.

SRO Basis: 10 CFR 55.43(b)-2, Facility operating limitations in the technical specifications and their bases. Application of required actions statements.

Notes From SD-07 Coupling integrity of a control rod shall be checked anytime a control rod is fully withdrawn by verifying that the rod does not reach the overtravel position. An uncoupling check can be performed by maintaining the continuous withdraw signal for approximately 3 to 5 seconds when the control rod has reached position 48 and verifying the control rod does not retract beyond position 48. If the rod is uncoupled, then the four rod display indication will go out for the uncoupled rod and the Rod Over Travel Annunciator A-05 4-2 will illuminate.

SD-07 Rev. 6 Page 27 of 57 C.

One or more control rods C. 1 inoperable for reasons other Inoperable control rod may than Condition A or 8.

be bypassed in the RWM or RWM may be bypassed as allowed by LCO 3.3.2.1, if required, to allow insertion of inoperable control rod and continued operation.

Fully insert inoperable 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> control rod.

AND (continued)

C.

(continued)

C.2 Disarm the associated 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> CRD.

From GP-11:

This procedure provides a method for a second licensed operator or qualified member of the plant technical staff to verify control rod movement and compliance with the prescribed BPWS control rod pattern with the rod worth minimizer (RWM) inoperable in conformance with the requirements of Technical Specification 332.1. If the RWM is inoperable due to bypassed control rod(s) that will not be moved during the startup/shutdow, then this procedure is not required.

Categories K/A:

214000 A2.03 Tier / Group:

T2G2 RO Rating:

3.6 SRO Rating:

3.9 LP Obj:

07-5B Source:

NEW Cog Level:

HIGH Category 8:

Y Notes From SD-07 Coupling integrity of a control rod st"lall be checked anytime a control rod is fully withdrawn by verifying that the rod does not reacl) the overtravel position. An uncoupling check can be performed by maintaining the continuous withdraw signal for approximately 3 to 5 seconds when the control rod t"las reached position 48 and verifying the control rod does not retract beyOnd position 48. If the rod is uncoupled, then the four rod display indication will go out for the uncoupled rod and the Rod Over Travel Annunciator A-05 4-2 will illuminate.

I SD-07 Rev.S Page 27 of 571 C.

One or more control rods C.*1 inoperable for reasons other than Condition A or B.

AND C.

(continued)

C.2 From GP-11:

NOTE--------

Inoperable control rod may be bypassed in the RWM or RWM may be bypassed as allowed bv LCO 3.3.2.1: if required, to allow insertion of inoperable control rod and continued operation.

Fully insert inoperable control rod.

Disarm the associated CRD.

3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> (continued) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> This procedure provides a method for a second licensed operator or qualified member of the plant technical staff to verify control rod movement and compliance with the prescribed BPWS control rod pattern with the rod worth minimizer (RWM) inoperable in conformance with the requirements of Technical Specification 3.3.2.1. If the RWM is inoperable due to bypassed control rod{s) that Will not be moved during the startup/shutdown, then this procedure is not required.

Categories KIA:

214000 A2.03 Tier / Group: T2G2 RORating:

3.6 SRORating

3.9 LPObj

07-SB Source:

NEW Cog Level:

HIGH Category 8:

Y

79. Given the following ATWS conditions on Unit Two:

2A CRD Pump Overcurrent trip 2B CRD Pump Shaft uncoupled HPCI System Under Clearance SLC Both squib valves failed to fire RCIC Running with an unisolable steam supply leak Suppression Pool Level

-24 inches Reactor Power 10%

Reactor Water Level 160 inches Which one of the following identifies the action that would be taken concerning the RCIC system based on the conditions above?

The RCIC system would:

A. be isolated to secure the source of the steam leak lAW OAOP-05.0, Radioactive Spills, High Radiation, and Airborne Activity.

B. have the high suppression pool water level transfer defeated and its suction transferred back to the CST lAW SEP-i 0, Circuit Alterations Procedure.

C remain running because it is needed for boron injection lAW LEP-03, Alternate Boron Injection.

D. be terminated and prevented to reduce level to 90 inches lAW LPC.

79. Given the following ATWS conditions on Unit Two:

2A CRD Pump 2B CRD Pump HPCI System SLC RCIC Suppression Pool Level Reactor Power Reactor Water Level Overcurrent trip Shaft uncoupled Under Clearance Both squib valves failed to fire Running with an unisolable steam supply leak

-24 inches 10%

160 inches Which one of the following identifies the action that would be taken concerning the RCIC system based on the conditions above?

The RCIC system would:

A. be isolated to secure the source of the steam leak lAW OAOP-05.0, Radioactive Spills, High Radiation, and Airborne Activity.

B. have the high suppression pool water level transfer defeated and its suction transferred back to the CST lAW SEP-1 0, Circuit Alterations Procedure.

C~ remain running because it is needed for boron injection lAW LEP-03, Alternate Boron Injection.

D. be terminated and prevented to reduce level to 90 inches lAW LPC.

Feedback K/A: 217000 G2.04.08 Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

Reactor Core Isolation Cooling System (RCIC)

(CFR: 41.10/43.5/45.13)

ROISRO Rating: 3.8/4.5 Objective: CLS-LP-300-J Obj 4 Given plant conditions and a copy of the LEPs, determine which method of alternate boron injection is appropriate.

Reference:

AOP-50/SCCP/LEP-03/LPC Cog Level high Explanation: EOP action that supercedes the AOP action is what the question is asking.

AOP-5 does have a step to isolate the system that is leaking, but SCCP overrides that if the system is required by EOPs. With the ATWS the RCIC system is required for alternate boron injection. The Suppression Pool level is high but this will only transfer the HPCI suction valves as RCIC only transfers on CST level. LEP-03 would first want to use CRD then RCIC as long as suction is from CST.

Distractor Analysis:

Choice A: Plausible because AOP-5 does have a step to isolate the system that is leaking, but SCCP overrides that if the system is required by EOPs.

Choice B: Plausible because the high suppression pool level would transfer HPCI and SEP-10 has a section for transferring the suction to the CST from the Suppression Pool.

Choice C: Correct answer, see explanation Choice D: Plausible becasue LPC does have a step for terminating and preventing but this does not address RCIC.

SRO Basis: 10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Feedback KIA: 217000 G2.04.08 Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

Reactor Core Isolation Cooling System (RCIC)

(CFR: 41.10/43.5/45.13)

RO/SRO Rating: 3.8/4.5 Objective: CLS-LP-300-J Obj 4 Given plant conditions and a copy of the LEPs, determine which method of alternate boron injection is appropriate.

Reference:

AOP-50 1 SCCP 1 LEP-03 1 LPC Cog Level high Explanation: EOP action that supercedes the AOP action is what the question is asking.

AOP-5 does have a step to isolate the system that is leaking, but SCCP overrides that if the system is required by EOPs. With the A TWS the RCIC system is required for alternate boron injection. The Suppression Pool level is high but this will only transfer the HPCI suction valves as RCIC only transfers on CST level. LEP-03 would first want to use CRD then RCIC as long as suction is from CST.

Distractor Analysis:

Choice A: Plausible because AOP-5 does have a step to isolate the system that is leaking, but SCCP overrides that if the system is required by EOPs.

Choice B: Plausible because the high suppression pool level would transfer HPCI and SEP-10 has a section for transferring the suction to the CST from the Suppression Pool.

Choice C: Correct answer, see explanation Choice D: Plausible becasue LPC does have a step for terminating and preventing but this does not address RCIC.

SRO Basis: 10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Notes Actions from AOP-05.O 1.

INITIATE a search to locate and isolate the source of any fl coolant or steam leak.

2.

IF radiography is in progress, AND personnel are in D

danger of abnormal exposure, THEN SECURE radiography.

3.

ENSURE all personnel in the area monitor their dosimetry and report unusual exposure to E&RC.

OAOP-05.C Rev. 23 Page 6 of 10 From SCCP has the actions to leave the system running:

ISOLATE ALL SYSTEMS DISCHARGING INTO THE AREA EXCEPT SYSTEMS REQUIRED:

  • TO BE OPERATED BYAN EMERGENCY OPERATING PROCEDURE
  • FOR DAMAGE CONTROL I

SCCP-14 From LEP-03 A

NOTE:

HPCI/RCIC should be used only if suction is from the CST A

From LPC, RCIC is not on list to Terminate and prevent (HPCI is):

LOWER REATQR WATER LEVEL J!UEsrECTIvE OF ANY REACTOR POWER OR REACTOR WATER LEVELOSCILLATIONS ry TERMflIAJ1NO AND P4flTMflN tNJFCTIOU FRCM TIlE FOLLO%1NG SYSTE1S UNLESS THE SYSTEM IS SEINO USED IO NJEC1 CORON

  • GQNDENSATEFEECWATER
  • NPCI
  • ALTERNATE COCL?.NT NJTCtIflN YTPM!

RC)L-17 Categories KJA:

217000 G2.04.08 Tier! Group:

T2G1 RO Rating:

3.8 SRO Rating:

4.5 LP Obj:

300J-4 Source:

NEW Cog Level:

HIGH Category 8:

Y Notes Actions from AOP-05.0

1.
2.
3.

INITIATE a search to locate and isolate the source of any coolant or steam leak.

IF radiography is in progress, AND personnel are in danger of abnormal exposure, THEN SECURE radiography.

ENSURE all personnel in the area monitor their dosimetry and report unusual exposure to E&RC.

o o

o IOAOP-05.0 Rev. 23 Page 6 of"10 I From SCCP has the actions to leave the system running:

ISOLATE ALL SYSTEMS DISCHARGING INTO THE AREA EXCEPT SYSTEMS REQUIRED:

  • TO BE OPERATED BY AN EMERGENCY OPERATING PROCEDURE
  • FOR DAMAGE CONTROL From LEP-03 NOTE:

HPGIIRCIC should be used only it suction is from the CST.

From LPC, RCIC is not on list to Terminate and prevent (HPCI is):

LOWEll; ReACTOR WATER LEVEL IRlIE$PEC11VE OF ANY REACTOR POWER OR REACTOR

\\VATER I.EVELOSCILLAllONS BY TI"~MIIIAliNG AND PRlilfliNTING IN.I~cnOIJ FRQM THE FOLlOVlIHG SYSTEMS UHLESSTHE

~ YSTEM IS BEING USElJ TO INJECT IIOItOH;

  • CO.'lDENSATelFESOWATI!R
  • IIPCI
  • ' IIHR
  • CORESPRAY Categories KIA:

217000 G2.04.08 RORating:

3.8 LPObj

300J-4 Cog Level:

HIGH Tier / Group:

SRORating:

Source:

Category 8:

T2G1 4.5 NEW Y

80. Unit Two was operating at rated power with the following conditions:

- A dual Unit Loss Of Offsite Power (LOOP)

- Spent Fuel Pool level is lowering rapidly due to a dropped test weight

- RRCP has been entered due to high rad conditions on the refuel floor Which one of the following is the first makeup source to be used for filling the fuel pool and identifies the procedure to perform the action?

A. Emergency Diesel Makeup Pump via hoses lAW OAOP-38.0, Loss of Fuel Pool Cooling B RHR B Loop via Fuel Pool Cooling System lAW OAOP-38.0, Loss of Fuel Pool Cooling C. Emergency Diesel Makeup Pump via hoses lAW OEDMG-002, Spent Fuel Pool Makeup/Spray and Refuel Floor Enhanced Ventilation under Conditions of Extreme Damage D. RHR B Loop via Fuel Pool Cooling System lAW OEDMG-002, Spent Fuel Pool Makeup/Spray and Refuel Floor Enhanced Ventilation under Conditions of Extreme Damage

80. Unit Two was operating at rated power with the following conditions:

- A dual Unit Loss Of Offsite Power (LOOP)

- Spent Fuel Pool level is lowering rapidly due to a dropped test weight

- RRCP has been entered due to high rad conditions on the refuel floor Which one of the following is the first makeup source to be used for filling the fuel pool and identifies the procedure to perform the action?

A. Emergency Diesel Makeup Pump via hoses lAW OAOP-38.0, Loss of Fuel Pool Cooling B!'" RHR B Loop via Fuel Pool Cooling System lAW OAOP-38.0, Loss of Fuel Pool Cooling C. Emergency Diesel Makeup Pump via hoses lAW OEDMG-002, Spent Fuel Pool Makeup/Spray and Refuel Floor Enhanced Ventilation under Conditions of Extreme Damage D. RHR B Loop via Fuel Pool Cooling System lAW OEDMG-002, Spent Fuel Pool Makeup/Spray and Refuel Floor Enhanced Ventilation under Conditions of Extreme Damage

Feedback K/A: 233000 G2.04.06 Knowledge of EOP mitigation strategies.

Fuel Pool Cooling and Clean-up (CFR: 41.10/43.5/45.13)

There are no direct EOP actions associated with FPC, a loss of level in the fuel pool will cause entty into RRCP which is an EOP. So these actions are mitigation strategies to RRCP.

RO/SRO Rating: 3.7/4.7 Objective:

CLS-LP-1 3, Obj. 11. State the sources of makeup water for the Fuel Pool in order of preference.

Reference:

OAOP-38.0 Loss of Fuel Pool Cooling Cog Level High Explanation:

The order of the makeup sources is from the normal fill, Demin water hose stations, Fire protection hose stations, demin water through RHR keepfill, and then other sources that are not service water. For a high capacity water source and the gates installed RHR Loop B would be used via the FPC system. With a LOOP the demin pumps have no power. If no other sources are availble then the proceudre has injection from the EDMP.

Distractor Analysis:

Choice A: Plausible because although this is a makeup source it is not the preferred source (last resort per the procedure) and is performed per the EDMG procedures. Although upon entering the AOP there is a step to start lining this system up for injection because of the time required to get all of the hoses run in the procedure up to the fuel pool.

Choice B: Correct answer see explanation Choice C: Plausible because although this is a makeup source it is not the preferred source (last resort per the procedure). Although upon entering the AOP there is a step to start lining this system up for injection because of the time required to get all of the hoses run in the procedure up to the fuel pool.

Choice D: Plausible because RHR is the high capacity source that will need to be used, but the EDMG procedure does not provide this guidance.

SRO Basis: 10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Feedback KIA: 233000 G2.04.06 Knowledge of EOP mitigation strategies.

Fuel Pool Cooling and Clean-up (CFR: 41.10/43.5/45.13)

There are no direct EOP actions associated with FPC, a loss of level in the fuel pool will cause entry into RRCP which is an EOP. So these actions are mitigation strategies to RRCP.

RO/SRO Rating: 3.7/4.7 Objective:

CLS-LP-13, Obj. 11. State the sources of makeup water for the Fuel Pool in order of preference.

Reference:

OAOP-38.0 Loss of Fuel Pool Cooling Cog Level High Explanation:

The order of the makeup sources is from the normal fill, Demin water hose stations, Fire protection hose stations, demin water through RHR keepfill, and then other sources that are not service water. For a high capacity water source and the gates installed RHR Loop B would be used via the FPC system. With a LOOP the demin pumps have no power. If no other sources are availble then the proceudre has injection from the EDMP.

Distractor Analysis:

Choice A: Plausible because although this is a makeup source it is not the preferred source (last resort per the procedure) and is performed per the EDMG procedures. Although upon entering the AOP there is a step to start lining this system up for injection because of the time required to get all of the hoses run in the procedure up to the fuel pool.

Choice B: Correct answer see explanation Choice C: Plausible because although this is a makeup source it is not the preferred source (last resort per the procedure). Although upon entering the AOP there is a step to start lining this system up for injection because of the time required to get all of the hoses run in the procedure up to the fuel pool.

Choice 0: Plausible because RHR is the high capacity source that will need to be used, but the EDMG procedure does not provide this guidance.

SRO Basis: 10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Notes 2.

IMMEDIATELY ENTER OEDMG-002, Spent Fuel Pool Makeup/Spray and Enhanced Refuel Floor Ventilation Under Conditions of Extreme Damage, AND make preparations to makeup to the fuel pool using the EDMP.

3.2.12 IF a high capacity makeup source of water through the RI-IR System is required to maintain fuel pool level AND the fuel pool gates are installed, THEN PERFORM the following:

CONFIRM one of the following ow paths available for use with the Fuel Pool Cooling System:

RHR Loop B only (RHR Loop B Shutdown Cooling must be secured)

RHR Loop A through RHR Loop Cross-Tie to the RHR Loop B discharge. (Both RHR Loop A and Loop B Shutdown Cooling must be secured).

QAOP-38.O Rev. 22 Page 11 of 35 Actions for Emergency Diesel Makeup Pump:

3.2.19 IF no actions have been successful, THEN ENTER LI OEDMG-OD2, Spent Fuel Pool Makeup/Spray and Refuel Floor Enhanced Ventilation under Conditions of Extreme Damage.

From EMG-002:

3.3 Normal fuel pool makeup methods and the B.5.b requirement for a diverse internal strategy (using installed plant equipment) are contained in OAOP-38.O, Loss of Fuel Pool Cooling. OEDMG-002 is entered when the methods contained in OAOP-38.O have proven to be inadequate or cannot be performed.

Categories K/A:

233000 G2.04.06 Tier / Group:

T2G2 RO Rating:

3.7 SRO Rating:

4.7 LPObj

13-11 Source:

NEW Cog Level:

HIGH Category 8:

LI Li Notes

2.

3.2:12

1.

IMMEDIATELY ENTER OEDMG-002, Spent Fuel Pool Makeup/Spray and Enhanced Refuel Floor Ventilation Under Conditions of Extreme Damage, AND make preparations to makeup to the fuel pool using the EDfvIP.

IF a high capacity makeup source of water through the RHR System is required to maintain fuel pool level AND tt1e fuel pool gates are installed, THEN PERFORM tile following:

CONFIRM one of the following flow paths available for use with the Fuel Pool Cooling System:

D RHR Loop B only (RHR Loop 6 Shutdown Cooling D

must be secured)

RHR Loop A through RHR Loop Cross-Tie to the D

RHR Loop 6 discharge. (60th RHR Loop A and Loop 6 Shutdown Cooling must be secured).

IOAOP-38.0 Rev. 22 Page 1-1 of 351 Actions for Emergency Diesel Makeup Pump:

3.2:19 From EMG-002:

IF no actions have been successful, THEN ENTER OEDMG-002, Spent Fuel Pool Makeup/Spray and Refuel Floor Enhanced Ventilation under Conditions of Extreme Damage.

D 3.3 Normal fuel pool makeup methods and the 6.S.b requirement for a diverse internal strategy (using installed plant eqUipment) are contained in OAOP-38.0, Loss of Fuel Pool Cooling. OEDMG-002 is entered when the methods contained in OAOP-38.0 have proven to be inadequate or cannot be performed.

Categories KIA:

233000 G2.04.06 Tier / Group: T2G2 RORating:

3.7 SRORating

4.7 LPObj:

13-11 Source:

NEW Cog Level:

HIGH Category 8:

81. With Unit Two at rated power, which one of the following identifie (1) the required number of operable SRVs for safety function lAW Technical Specification 3.4.3, Safety/Relief Valves and (2) the bases for this number of operable SRVs?

A. (1)9 (2) prevent overpressurization associated with an ATWS event B (1) 10 (2) prevent overpressurization associated with an ATWS event C. (1)9 (2) prevent overpressurization associated with an MSIV closure D. (1)10 (2) prevent overpressurization associated with an MSIV closure Feedback K/A: 239002 G2.02.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

Safety Relief Valves (CFR: 41.5 / 41.7/43.2)

ROISRO Rating: 3.2/4.2 Objective: CLS-LP-25, Obj. 10 Given plant conditions and TS, including the Bases, TRM, ODCM, and COLR determine the required actions to be taken in accordance with TS associated with the Reactor Recirculation System. (SRO only)

Reference:

TS 3.4.3 and bases document Cog Level Low Explanation:

TS 3.4.3 states 10 must be operational for the safety function, the bases states the reason, ATWS.

Distractor Analysis:

Choice A: Plausible because the bases states that 9 are required for the MSIV closure.

Choice B: Correct answer, see explanation Choice C: Plausible because the bases states that 9 are required for the MSIV closure and the MSIV closure is not the binding failure mode.

Choice D: Plausible because 10 are required for the ATWS and the MSIV closure is not the binding failure mode.

SRO Basis: 10 CFR 55.43(b)-2, Facility operating limitations in the technical specifications and their bases. This is knowledge of tech spec bases to determine the reason 10 are required.

81. With Unit Two at rated power, which one of the following identifies:

(1) the reql:lired number of operable SRVs for safety function lAW Technical Specification 3.4.3, Safety/Relief Valves and (2) the bases for this number of operable SRVs?

A. (1) 9 (2) prevent overpressurization associated with an ATWS event B!'" (1) 10 (2) prevent overpressurization associated with an ATWS event C. (1) 9 (2) prevent overpressurization associated with an MSIV closure D. (1) 10 (2) prevent overpressurization associated with an MSIV closure Feedback KIA: 239002 G2.02.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

Safety Relief Valves (CFR: 41.5/41.7/43.2)

RO/SRO Rating: 3.2/4.2 Objective: CLS-LP-25, Obj. 10 Given plant conditions and TS, including the Bases, TRM, ODCM, and COLR determine the required actions to be taken in accordance with TS associated with the Reactor Recirculation System. (SRO only)

Reference:

TS 3.4.3 and bases document Cog Level Low Explanation:

TS 3.4.3 states 10 must be operational for the safety function, the bases states the reason, ATWS.

Distractor Analysis:

Choice A: Plausible because the bases states that 9 are required for the MSIV closure.

Choice B: Correct answer, see explanation Choice C: Plausible because the bases states that 9 are required for the MSIV closure and the MSIV closure is not the binding failure mode.

Choice D: Plausible because 10 are required for the ATWS and the MSIV closure is not the binding failure mode.

SRO Basis: 10 CFR 55.43(b)-2, Facility operating limitations in the technical speCifications and their bases. This is knowledge of tech spec bases to determine the reason 10 are required.

Notes 3.4.3 Safetv?Rehef Valves (SRVs)

LCO 3.4.3 The safety function of 10 SRVs shall be OPERASLE.

APPUCABILITY:

MODES 1. 2. and 3.

From the Bases document:

APPLICABLE The overpressure protection system must accommodate the most SAFETY ANALYSES severe pressunzattan transient EvaIatioi aac1eterrnined that the rnoLseere4ransient icioit1flar&fres (MP7 fIod by readorscrt n

1f1hr{ie fiIure of the direct scrani i+/-tdith MSIV position) (Ref. 1 For the purpose of the analyses, 9 SRVs are assumed to operate in the safety mode. The analysis results demonstrate that the design SRV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure (110% x 1250 psig 1375 psig). This LCO helps to ensure that the acceptance limit of 1375 psig is met during the Design Basis Event.

(continued APPLICABLE For overpressurization associated with an ATWS event, 10 SRVs are SAFETY ANALYSES assumed to operate in the safety mode. The analysis (Ref. 2)

(continued) results demonstrate that the design capacity is capable of maintaining reactor pressure below the ASME Section Ill Code Serice Level C limits (1500 psig).

From an overpressure standpoint, the design basis events are bounded by the overpressurization associated with the ATWS event described above. Reference 3 discusses additional events that are expected to actuate the SRVs.

SRVs satisfy Criteiion 3 of 10 CER 50.36(cX2)(ii) (Ref. 4).

Categories KJA:

239002 G2.02.25 Tier / Group:

T2G1 RO Rating:

3.2 SRO Rating:

4.2 LP Obj:

25-10 Source:

NEW Cog Level:

LOW Category 8:

Y Notes 3A.3 Safety/Relief Valves (SRVs)

LCO 3.4.3 The safety function of '10 SRVs shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

From the Bases document:

APPLICABLE The overpressure protection system must accommodate the most SAFETY ANAL YSES severe pressurization transient.

of the direct purpose of the analyses;* 9SR\\lsare assumed,tooperate.in,the safety mode. The analysis results demonstrate that the design SRV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure P'l 0% x '1250 psig = 1375 psig). This LCO helps to ensure that the acceptance limit of '1375 psig is met during the Design Basis Event.

(continued)

APPLICABLE For overpressurization associated with an A nNS event. 10 SRVs are SAFETY ANALYSES assumed to operate in the safety mode. The analysis (Ref. 2)

(continued}

results demonstrate that the design capacity is capable of maintaining reactor pressure below the ASME Section III Code Service Level C limits (1500 psig).

Categories KIA:

RORating:

LP Obj:

Cog Level:

From an overpressure standpoint, the design basis events are bounded by the overpressurization associated with the AnNS event described above. Reference 3 discusses additional events that are expected to actuate the SRVs.

SRVs satisfy Criterion 3 of '10 CFR 50.36(c){2)(ii) (Ref. 4).

239002 G2.02.25 Tier / Group: T2G1

3.2 SRORating

4.2 25-10 Source:

NEW LOW Category 8:

Y

82. Unit One is operating at full power when the Main Stack Rad Monitor lost its norm power supply.

Which one of the following identifies the procedure that contains the steps to transfer the Main Stack Rad Monitor to its alternate power supply?

A. IOP-52, 120 Volt AC UPS, Emergency, and Conventional Electrical Systems Operating Procedure B 20P-52, 120 Volt AC UPS, Emergency, and Conventional Electrical Systems Operating Procedure C. 1APP UA-03 6-3, PROCESS SMPL OG VENT PIPE DNSC/INOP D. 2APP UA-03 6-3, PROCESS SMPL OG VENT PIPE DNSC/INOP Feedback K/A: 262002 G2.0l.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Uninterruptable Power Supply (A.C.ID.C.)

(CFR: 41.10/43.5145.2/45.6)

ROISRO Rating: 4.3/4.4 Objective: CLS-LP-11.0, 15a Given plant conditions and a trip or failure of one of the following Radiation Monitors, determine appropriate plant response and use procedures to determine the actions required to control and/or mitigate the consequences of the event:

a. Main Stack.

Reference:

20P-52, Section 8.7, Stack Radiation Monitor UPS Power Supply Transfer Cog Level High Explanation:

The normal power supply for the Main Stack Rad Monitor is from Unit Two. On a loss of power the from the normal power supply the operators will need to transfer to the alternate power supply. This direction is only in the U2 procedure. There is no directions to perform this in the Ui procedure or the APPs for either Unit.

Distractor Analysis:

Choice A: Plausible because the stem states this is Ui but the actions are in the U2 procedure.

Choice B: Correct answer, see explanation.

Choice C: Plausible because the downscale imp annunciator will be actuated on a loss of power but the APP5 do not address transfer of power to backup supply.

Choice D: Plausible because the downscale / mop annunciator will be actuated on a loss of power but the APPs do not address transfer of power to backup supply. U2 is the normal power supply to the rad monitor.

SRO Basis: 10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

82. Unit One is operating at full power when the Main Stack Rad Monitor lost its normal power supply.

Which one of the following identifies the procedure that contains the steps to transfer the Main Stack Rad Monitor to its alternate power supply?

A. 10P-52, 120 Volt AC UPS, Emergency, and Conventional Electrical Systems Operating Procedure B~ 20P-52, 120 Volt AC UPS, Emergency, and Conventional Electrical Systems Operating Procedure C. 1APP UA-03 6-3, PROCESS SMPL OG VENT PIPE DNSCIINOP D. 2APP UA-03 6-3, PROCESS SMPL OG VENT PIPE DNSCIINOP Feedback KIA: 262002 G2.01.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Uninterruptable Power Supply (A.C'/D.C.)

(CFR: 41.10 143.5 1 45.2 145.6)

RO/SRO Rating: 4.3/4.4 Objective: CLS-LP-11.0, 15a Given plant conditions and a trip or failure of one of the following Radiation Monitors, determine appropriate plant response and use procedures to determine the actions required to control and/or mitigate the consequences of the event:

a. Main Stack.

Reference:

20P-52, Section 8.7, Stack Radiation Monitor UPS Power Supply Transfer Cog Level High Explanation:

The normal power supply for the Main Stack Rad Monitor is from Unit Two. On a loss of power the from the normal power supply the operators will need to transfer to the alternate power supply. This direction is only in the U2 procedure. There is no directions to perform this in the U1 procedure or the APPs for either Unit.

Distractor Analysis:

Choice A: Plausible because the stem states this is U1 but the actions are in the U2 procedure.

Choice B: Correct answer, see explanation.

Choice C: Plausible because the downscale 1 inop annunciator will be actuated on a loss of power but the APPs do not address transfer of power to backup supply.

Choice D: Plausible because the downscale 1 inop annunciator will be actuated on a loss of power but the APPs do not address transfer of power to backup supply. U2 is the normal power supply to the rad monitor.

SRO Basis: 10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Notes 8.0 INFREQUENT OPERATIONS.

32 8.1 Transferring UPS Loads From Alternate Source to Primary UPS Unit 2A 32 8.2 Transferring UPS Loads From Alternate Source to Standby UPS Unit 28 35 8.3 Transferring UPS Loads From Standby UPS Unit 2B to Alternate Source 39 8.4 Transferring UPS Loads From Primary UPS Unit 2A to Alternate Source 41 8.5 Alignment of Standby UPS Unit 2B After a Loss of Alternate Source Power 43 8.6 Returning Standby UPS Unit 2B to Normal Operating Condition Upon Regaining Alternate Source Power Supply 45 8.7 Stack Radiation Monitor UPS Power Supply Transfer 47 20P-52 Rev. 53 Page2of78 8.0 INFREQUENT OPERATIONS 34 8.1 Transferring UPS Loads From Alternate Source to Primary UPS Unit 1A 34 8.2 Transferring UPS Loads From Alternate Source to Standby UPS Unit 18 38 8.3 Transferring UPS Loads From Standby UPS Unit I B to Alternate Source 42 8.4 Transfening UPS Loads From Primary UPS Unit IA to Alternate Source 44 8.5 Alignment of Standby UPS Unit lB After a Loss of Alternate Source Power 46 8.6 Returning Standby UPS Unit 1 B to Normal Operating Condition Upon Regaining Alternate Source Power Supply 48 I OP-52 Rev. 35 Page 3 of 74 Unit2 APP UA-03 6-3 Page 1 of 1 PROCESS SMPL OG VENT PIPE DNSCIINOP (Process Sample Off-Gas Pipe Down-Inoperable)

AUTO ACTIONS NONE CAUSE 1.

Off-gas vent pipe (stack) radiation monitor downscale or out of service.

2.

Circuit malfunction.

3.

Change in background counts, possibly from unit power reduction.

Categories K/A:

262002 G2.0 1.23 Tier / Group:

T2G1 RO Rating:

4.3 SRO Rating:

4.2 LP Obj:

I 1-15A Source:

NEW Cog Level:

HIGH Category 8:

Y Notes 8.0 INFREQUENT OPERATIONS.................................................................................. 32 8.1 Transferring UPS Loads From Alternate Source to PrimalY UPS Unit 2A..... 32 8.2 Transferring UPS Loads From Alternate Source to Standby UPS Unit 28.... 35 8.3 Transferring UPS Loads From Standby UPS Unit 2B to Alternate Source.... 39 8.4 Transferring UPS Loads From Primary UPS Unit 2A to Alternate Source..... 4'1 8.5 Alignment of Standby UPS Unit 2B After a Loss of Alternate Source Power. 43 8.6 Returning Standby UPS Unit 2B to Normal Operating Condition Upon Regaining Altemate Source Power Supply..................................................... 45 8.7 Stack Radiation Monitor UPS Power Supply Transfer.................................... 47 120P-52 Rev. 53 Page 2 of 78 I 8.0 INFREQUENT OPERATIONS............................................................................... 34 8.1 Transferring UPS Loads From Alternate Source to PrimalY UPS Unit 1A..... 34 8.2 Transferring UPS Loads From Alternate Source to Standby UPS Unit 18.... 38 8.3 Transferring UPS Loads From Standby UPS Unit '1 B to Alternate Source.... 42 8.4 Transferring UPS Loads From Primary UPS Unit 'lA to Alternate Source..... 44 8.5 Alignment of Standby UPS Unit 1 B After a Loss of Alternate Source Power................................................................................................................. 46 8.6 Returning Standby UPS Unit 1 B to Normal Operating Condition Upon Regaining Alternate Source Power Supply..................................................... 48 l'lOP-52 Rev. 35 Page 3 of 74 I PROCESS SMPL OG VENT PIPE DNSCllNOP (Process Sample Off-Gas Pipe Down-Inoperable)

AUTO ACTIONS NONE CAUSE Unit 2 APP UA-03 6-3 Page 1 of'l

'I.

Off-gas vent pipe (stack) radiation monitor downscale or out of service.

2.

Circuit malfunction.

3.

Change in background counts, possibly from unit power reduction.

Categories KIA:

262002 G2.0 1.23 Tier / Group: T2Gl RORating:

4.3 SRORating

4.2 LPObj

11-15A Source:

NEW Cog Level:

HIGH Category 8:

Y

83. The following conditions exist on Unit Two following a spurious Main Turbine trip at rated power:

SDV HI-HI WTR LVL TRIP BYPASS OTBD NSSS VALVES MTR OVERLOAD Reactor level Reactor Pressure All Control Rods Scram RWCU System In alarm In alarm 185 inches and steady 900 psig with BPVs controlling Fully inserted Being reset lAW LEP-02 Isolated by 2-G31 -FOOl The 2-G31-F004 (RWCU Outboard lsol VIv) failed to automatically close on a valid isolation signal due to motor overload.

Which one of the following identifies the Technical Specification requirements when the RSP is exited?

The RSP can be exited to OGP-05, Unit Shutdown, provided an active LCO is implemented for Technical Specification (1)

The start time of the LCO action completion time is when the (2)

A.

(1) 3.3.1.1, Reactor Protection System (RPS) Instrumentation (2) condition occurred B.

(1) 3.3.1.1, Reactor Protection System (RPS) Instrumentation (2) RSP is exited C. (1) 3.6.1.3, Primary Containment Isolation Valves (PCIV5)

(2) condition occurred D (1) 3.6.1.3, Primary Containment Isolation Valves (PCIV5)

(2) RSP is exited

83. The following conditions exist on Unit Two following a spurious Main Turbine trip at rated power:

SDV HI-HI WTR LVL TRIP BYPASS OTBD NSSS VALVES MTR OVERLOAD Reactor level Reactor Pressure All Control Rods Scram RWCU System In alarm In alarm 185 inches and steady 900 psig with BPVs controlling Fully inserted Being reset lAW LEP-02 Isolated by 2-G31-F001 The 2-G31-F004 (RWCU Outboard Isol Vlv) failed to automatically close on a valid isolation signal due to motor overload.

Which one of the following identifies the Technical Specification requirements when the RSP is exited?

The RSP can be exited to OGP-05, Unit Shutdown, provided an active LCO is implemented for Technical Specification (1)

The start time of the LCO action completion time is when the (2)

A. (1) 3.3.1.1, Reactor Protection System (RPS) Instrumentation (2) condition occurred B. (1) 3.3.1.1, Reactor Protection System (RPS) Instrumentation (2) RSP is exited C. (1) 3.6.1.3, Primary Containment Isolation Valves (PCIVs)

(2) condition occurred D!'" (1) 3.6.1.3, Primary Containment Isolation Valves (PCIVs)

(2) RSP is exited

Feedback K/A: S295006G 2.02.22 Knowledge of limiting conditions for operations and safety limits.

SCRAM (CFR: 41.5/43.2/45.2)

RO/SRO Rating: 4.0/4.7 Objective: CLSLP300C*1 I

11. Given plant conditions, the Unit Shutdown Procedure (GP-05), and the Reactor Scram Procedure, determine if conditions allow exiting the Reactor Scram Procedure.

Reference:

1 OCFR5O.36 OEOP-01-UG, Revision 55, Page 31, Section 3.5 Cog Level: High Explanation:

The EOPs authorize actions outside of technical specifications to mitigate the consequences of an emergency condition. The EOPs also provide for returning the system or component to service, If the system or component is not returned to its standby or operable condition prior to exiting the EOP5, then the appropriate limiting condition of operation shall be implemented in accordance with Technical Specifications The starting time for the limiting condition of operation is the time that the EOPs were exited.

In order to exit EOP, compatibilty with GP-05 along with active LCOs need to be implemented. PCIV G31-F004 is inoperable, TS 3.6.1.3 Condition A (A1)requires isolating the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve witf witj pursQ() jerify the affected penetration flow path is isolated Once per 31 days for isolation devices outside primary containment AND Prior to entering MODE 2 or 3 from MODE 4, if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside primary containment.

Distractor Analysis:

Choice A: Plausible because OGP-01 would be entered in order to restart the reactor and TS 3.3.1.1 would be correct if the unit was in Mode I or 2

- SDV Hi level is not required in Mode 3.

Choice B: Plausible because OGP-05 is correct and TS 3.3.1.1 would be correct if the unit was in Mode 1 or 2

- SDV Hi level is not required in Mode 3.

Choice C: Plausible because OGP-01 would be entered in order to restart the reactor and TS 3.6.1.3 is correct.

Choice D: Correct Answer SRO Only Basis: Requires assessment of plant conditions (RPS SDV Hi Level Bypass and Failed open PCIV) and prescribing a procedure with which to proceed (OGP-05).

Notes Feedback KIA: S295006G 2.02.22 Knowledge of limiting conditions for operations and safety limits.

SCRAM (CFR: 41.5 1 43.2 1 45.2)

RO/SRO Rating: 4.0/4.7 Objective: CLS-LP-300-C*11

11. Given plant conditions, the Unit Shutdown Procedure (GP-05), and the Reactor Scram Procedure, determine if conditions allow exiting the Reactor Scram Procedure.

Reference:

10CFR50.36 OEOP-01-UG, Revision 55, Page 31, Section 3.5 Cog Level: High Explanation:

The EOPs authorize actions outside of technical specifications to mitigate the consequences of an emergency condition. The EOPs also provide for returning the system or component to service. If the system or component is not returned to its standby or operable condition prior to exiting the EOPs, then the appropriate limiting condition of operation shall be implemented in accordance with Technical Specifications. The starting time for the limiting condition of operation is the time that the EOPs were exited.

In order to exit EOP, compatibilty with GP-05 along with active LCOs need to be implemented. PCIV G31-F004 is inoperable, TS 3.6.1.3 Condition A (A1)requires isolating the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve wi1~q9Ll~l~J~q~~ilbiml&~l:Ii~(.~ify the affected penetration flow path is isolated Once per 31 days for isolation devices outside primary containment AND Prior to entering MODE 2 or 3 from MODE 4, if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside primary containment.

Distractor Analysis:

Choice A: Plausible because OGP-01 would be entered in order to restart the reactor and TS 3.3.1.1 would be correct if the unit was in Mode 1 or 2 - SDV Hi level is not required in Mode 3.

Choice B: Plausible because OGP-05 is correct and TS 3.3.1.1 would be correct if the unit was in Mode 1 or 2 - SDV Hi level is not required in Mode 3.

Choice C: Plausible because OGP-01 would be entered in order to restart the reactor and TS 3.6.1.3 is correct.

Choice D: Correct Answer SRO Only Basis: Requires assessment of plant conditions (RPS SDV Hi Level Bypass and Failed open PCIV) and prescribing a procedure with which to proceed (OGP-05).

Notes

3.5 Technical Specifications The EOPs authorize actions outside of technical specifications to mitigate the consequences of an emergency condition. The EOPs also provide for returning the system or component to service.

If the system or component is not returned to its standby or operable condition prior to exiting the EOPs, then the appropriate limiting condition of operation shall be implemented in accordance with Technical Specifications. The starting time for the limiting condition of operation is the time that the EOPs were exited.

OEOP-01-IJG Rev. 55 Page 31 of 151 Completion Times 1.3 1.0 USE AND APPLICATION 1.3 Completion Times PURPOSE The purpose of this section is to establish the Completion Time convention and to provide guidance for its use.

BACKGROUND Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe operation of the unit. The ACTIONS associated with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Times(s).

DESCRIPTION The Completion Time is the amount of time allowed for completing a Required Action. it is referenced to the time of discovery of a situation (e.g, inoperable equipment or variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the unit is in a MODE or specified condition stated in the Applicability of the LCO. Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the unit is not within the LCO Applicability.

3.5 Technical Specifications The EOPs authorize actions outside of technical specifications to mitigate the consequences of an emergency condition. The EOPs also provide for returning the system or component to service. If the system or component is not returned to its standby or operable condition prior to exiting the EOPs, then the appropriate limiting condition of operation shall be implemented in accordance with Technical Specifications. The starting time for the limiting condition of operation is the time that the EOPs were exited.

IOEOP-01-UG Rev. 55

'1.0 USE AND APPLICATION Page 31 of 'IS" I Completion Times 1.3

'1.3 Completion Times PURPOSE BACKGROUND DESCRIPTION The purpose of this section is to establish the Completion Time convention and to provide guidance for its use.

Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe operation of the unit. The ACTIONS associated *with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Times(s).

The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the time of discovery of a situation (e.g., inoperable equipment or variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the unit is in a MODE or specified condition stated in the Applicability of the LCO. Required Actions must be completed prior to the expiration ofthe specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the unit is not within the LCO Applicability.

WHITE 55 OTBD NSSS VALVES MTR OVERLOAD Page 1 of 2 1.0 OPERATOR ACTIONS:

1.1 OBSERVE Automatic Functions:

1.1.1 IF one of the affected valves was being operated. THEN:

1. Valve motion will stop
2. Valve will NOT respond to control signals
3. Valve position will still be indicated 1.2 PERFORM Corrective Actions:

NOTE: Resetting valve motor overload devices or manual operaon of tripped motor-operated valves should only be attempted in emergency situations as directed by the Unit SCO.

CAUTION During manual oPeration of motor-operated valves, personnel should stand clear of the valve vhiIe either:

1. Resetting the thermal 3verload device or 2.

Operating the valve remo:ely.

1.2.1 IF the affected valve is required for operation, THEN PERFORM the following steps:

1. RESET the thermal overload device at the affected valve breaker compartment AND OPERATE the valve again.

2.

IF the themal overload device actuates again. THEN MANUALLY OPERATE the valve.

3. WHEN the valve is broken off its closed or open seat, THEN RESET the themial overload device at the affected valve breaker compartment AND OPERATE the valve.

1.2.2 REFERt0TS. 3.6.1.3 and TRMApp DTable 3.6.1.3-2.

24,PP-A-02 Rev. 32 Page 57 of 5-5 OTBD NSSS VALVES MTR OVERLOAD Page 1 of2 1.0 OPERATOR ACTIONS:

1.1 OBSERVE Automatic Functions:

1.1.1 IF one of the affected valves was being operated, THEN:

1. Valve motion will stop
2. Valve will NOT respond to control signals
3. Valve position will still be indicated 1.2 PERfORM Corrective Actions:

NOTE: Resetting valve motor overload devices or manual operation of tripped motor-operated valves sl10uld only be attempted in emergency Situations as directed by the Unit SeQ.

CAUTION During manual operation of motor-operated valves, personnel sh.ould stand clear of the valve while either:

1. Resetting the thermal overload device Of
2. Operating the valve remotely.

1.2.1 IF the affected valve is required for operation, THEN PERFORM the following steps:

1. RESET the thermal overload device at the affected valve breaker compartment AND OPERATE the valve again.
2. IF the themlal overload device actuates again, THEN MANUALLY OPERATE the valve.
3. WHEN the valve is broken off its closed or open seat, THEN RESET the themlal overload device at the affected valve breaker compartment AND OPERATE the valve.

1.2.2 REFER to T.S. 3.6.1.3 and TRM App. 0 Table 3.6.1.3-2.

12APp-A-02 Page 57 of 791

PCIVs 3.6.1.3 3.6 CONTAINMENT SYSTEMS 3.6.1.3 Primary Containment Isolation Valves (PCIVs)

LCO 16.1.3 Each PC1V, except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3, When associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, Primary Containment Isolation Instrumentation.

CONDITION REQUIRED ACTION COMPLETION TIME A.

NOTE-------

A.l Isolate the affected B hours Only applicable to penetration flow path by penetration flow paths with use of at least one closed two PCIVs.

and de-activated automatic

valve, closed manual valve, blind flange, or check valve One or more penetration with flow through the valve flow paths with one PCIV secured.

inoperable except for MSIV leakage not within limit.

3.6 CONTAINMENT SYSTEMS 3.6.'1.3 Primary Containment Isolation Valves (PCIVs)

LCO 3.6. '1.3 Each PCIV, except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3, VV'hen associated instrumentation is required to be OPERABLE per LCO 3.3.6:1, "Primary Containment Isolation Instrumentation."

CONDITION A.

NOTE-----------

A:I Only applicable to penetration flow paths 'Nith two PCIVs.

One or more penetration flo.... ' paths with one PCIV inoperable except for MSIV leakage not within limit.

AND REQUIRED ACTION COMPLETION TIME Isolate the affected 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.

3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.l.1I shall be OPERABLE.

APPLICABILITY:

According to Table 3.3.1.1-i.

ACTIONS

NOTE---

Separate Condition entry is allowed for each channel.

RPS Instrumentation 3.3.1.1 Tt 3l.t 3 c R: Pn:r Oy: i:Irc

,%PPLICABLE OC:NOION MOD O REOAED REHCE OTHER CHANNEL2 FO 2PECWE TA QLLAE 2URVELLArCE ALLOWABLE FL4O2N OCNDI:ON 3OTE3 AOT2r D.

REOLJIREVENTO VALUE 7.

Ccr CEcharc Vr 1,2 2

OR WLHh OR OR OR CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more required Al Place channel in trip.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> channels inoperable.

OR A.2 NOTE 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Not applicable for Functions 2.a, 2.b, 2.c, 24, or 2.f.

Place associated trip system in trip.

Categories KJA:

S2950006G 2.02.22 Tier! Group:

T1GI RO Rating:

4.0 SRO Rating:

4.7 LP Obj:

CLSLP300C*11 Source:

NEW Cog Level:

HIGH Category 8:

3.3.1 :1 3.3 INSTRUMENTATION 3.3.*'.'1 Reactor Protection System (RPS) Instrumentation LCO 3.3.-l.'I The RPS instrumentation for each Function in Table 3.3.'U-*1 shall be OPERABLE APPLICABILITY:

According to Table 3.3.1.. 1-*1.

ACTIONS

NOT E ---------------------------------------------------------

Separate Condition entry is allowed for each channel.

FUNCttON

7.

ScramOtscharoeVO\\\\m""e W3!er Le'o!et-HIgn CONDITION A.

One or more required channels inoperable.

Categories KIA:

S2950006G 2.02.22 RORating:

4.0 LPObj

CLS-LP-300-C* 11 Cog Level:

HIGH RPS Instrumentation 3.3.1.*'

Tatfe :t 3~ 1.1'"~ (~3J~.:. o~ 3t R~3-:'>>r Pf\\:t~,:.~or. Sj':~m. !n:.~".m~r:~l!:':f:

.'V'PLICASLE

!.. ~ooszc,:(

C*"iHER.

(lPECIFIE::l CC'NOI7'ICNS 1,;:

CO~OI7fONS RSQIJt:tED

,:t5FERENCEO CHANNeLS AAOM i=E?t r,:w=

3'V!l"iE\\1 FtEQIJ1,"iEO aVR\\'EII.LAJIoICE.

ACtiON 0.1 RSQUIREAlEma

~

OR ~.l.4.1";

a.=t

.3
.~.4.:S a,;:( 3.3."L~LH
a,=\\ 3.3:.il.L1':'

"'1.L!:JW *.r..sLE VAL,VE REQUIRED ACTION COMPLETION TIME A.'I Place channel in trip.

  • 12 hours OR A.2

NOTE-------------

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f.

Place associated trip system in trip.

Tier / Group: TIG!

SRORating: 4.7 Source:

NEW Category 8:

84. Following a scram on Unit Two, which one of the following correctly identifies:

(1) the initial response of reactor water level if an SRV is opened and (2) the procedure that contains the guidance to close the MSIVs due to water level?

A. (1) Shrink (2) Reactor Scram Procedure B. (1) Shrink (2) 2APP-A-07, REACTOR WATER LEVEL HIGH/LOW C (1) Swell (2) Reactor Scram Procedure D. (1) Swell (2) 2APP-A-07, REACTOR WATER LEVEL HIGH/LOW Feedback K/A: 295008 A2.05 Ability to determine andlor interpret the following as they apply to HIGH REACTOR WATER LEVEL:

Swell (CFR: 41.10/43.5/45.13)

ROISRO Rating: 2.9/3.1 Objective: CLS-LP-300-C, 10 Given plant conditions and the RSP, determine the required operator actions.

Reference:

RSP I 001-37.3 2APP-A-07, page 12 Cog Level: High Explanation:

Opening of the SRV will cause the reactor water level to swell up due to the reduction in pressure in the vessel and if level reaches the value in figure 1 on the RSP then closure of the MSIVs is directed. The MSIVs will close automatically but only on low level 3, not high level.

Distractor Analysis:

Choice A: Plausible if the examinee thinks that opening the SRV would reduce the water volume in the RPV, The RSP does contain the actions to close the MSIVs.

Choice B: Plausible if the examinee thinks that opening the SRV would reduce the water volume in the RPV, the examinee may think that the closure is an auto action, which are contained in the APP.

Choice C: Correct see explanation Choice D: Plausible because reactor water level will swell, and the examinee may think that the closure is an auto action, which are contained in the APP.

SRO Basis: 10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Notes

84. Following a scram on Unit Two, which one of the following correctly identifies:

(1) the initial response of reactor water level if an SRV is opened and (2) the procedure that contains the guidance to close the MSIVs due to water level?

A. (1) Shrink (2) Reactor Scram Procedure B. (1) Shrink (2) 2APP-A-07, REACTOR WATER LEVEL HIGH/LOW c~ (1) Swell (2) Reactor Scram Procedure D. (1) Swell (2) 2APP-A-07, REACTOR WA TER LEVEL HIGH/LOW Feedback KIA: 295008 A2.05 Ability to determine and/or interpret the following as they apply to HIGH REACTOR WATER LEVEL:

Swell (CFR: 41.10/43.5/45.13)

RO/SRO Rating: 2.9/3.1 Objective: CLS-LP-300-C, 10 Given plant conditions and the RSP, determine the required operator actions.

Reference:

RSP 1001-37.3 2APP-A-07, page 12 Cog Level: High Explanation:

Opening of the SRV will cause the reactor water level to swell up due to the reduction in pressure in the vessel and if level reaches the value in figure 1 on the RSP then closure of the MSIVs is directed. The MSIVs will close automatically but only on low level 3, not high level.

Distractor Analysis:

Choice A: Plausible if the examinee thinks that opening the SRV would reduce the water volume in the RPV, The RSP does contain the actions to close the MSIVs.

Choice B: Plausible if the examinee thinks that opening the SRV would reduce the water volume in the RPV, the examinee may think that the closure is an auto action, which are contained in the APP.

Choice C: Correct see explanation Choice 0: Plausible because reactor water level will swell, and the examinee may think that the closure is an auto action, which are contained in the APP.

SRO Basis: 10 CFR 55.43(b)-5, Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Notes

REACTOR WATER LEVEL I-IIGHFLOW PaOe 1 012 1.0 OPERATOR ACTIONS:

1.1 CONFIRM by multiple indications actual high or low reactor water level:

1.1.1 Reactor water level indication on RTGB Panel P603 may be used ror verification of water level:

1, Reactor Water Level A, C32-Ll-R606A.

2. Reactor Water Level B, C32-Ll-R6066.
3. Reactor Water Level C, C32-Ll-R606C.
4. Reactor Level/Pressure Recorder, C32-R608.

1.2 OBSERVE Automatic Functions:

1.2.1 IF reactor level decreases to 136 inches, THEN a reactor Scram results.

1.2.2 IF reactor level increases to 206 inches. THEN the Main Turbine, RFPTs, RCIC and HPCI turiines will trip.

1.2.3 IF either of the RFPs have tripped AND reactor water level is less than 182 inches. TI-lEN a Recirculation Pump runback will occur.

2APP-A-07 Rev. 32 Page 12 of 45 From the Reactor Scram Procedure:

flWAN J_Yt PL.CLALLMIV lTCHE1OCLO o1 LOWER REAtOR WATER WH RWCU REACTOR WATER LEVEL HIGHlLOW Page 1 of 2 1.0 OPERATOR ACTIONS:

1.1 CONFIRM by multiple indications actual high or low reactor water level:

1.1.1 Reactor water level indication on RTGB Panel P603 may I)e used for verification of water level:

1. Reactor Water Level.4" C32-Ll-R606.'\\.
2. Reactor Water Level B, C32-Ll-R606B.
3. Reactor Water Level C, C32-U-R606C.
4. Reactor Level/Pressure Recorder, C32-R608.

1.2 OBSERVE Automatic Functions:

1.2.1 IF reactor level decreases to '166 inches, THEN a reactor Scram results.

1.2.2 IF reactor level increases to 206 inclles, THEN the Main Turbine. RFPTs, RCIC and HPCI tUri)ines '.vill trip.

1.2.3 IFeitller of tile RFPs have tripped AND reactor water level is less than 182 inches, THEN a Recirculation Pump runback will occur.

12APP-A-07 Page 12 of 451 From the Reactor Scram Procedure:

Cl)

LUI (U)

-J LU>

LU

-j a

z ATTACHMENT 6 Page 19 of 19 FIGURE 21 Reactor Water Level at MSL (Main Steam Line Flood Level>

WHEN REACTOR PRESSURE IS LESS THAN 60 PSIG, USE INDICATED LEVEL.

MSL IS+250 INCHES.

DEOP-Ol-UG I

Rev. 55 Page 106 of 151 300 250 REF LEG TEMP ABOVE OR EQUAL TO 200F REF LEG TEMP BELOW 2COF 200 11111 IIIIII lIIlllII!1lllIJIjI1,15o 100 300 500 700 900 13100 60 200 400 600 800 1,000 REACTOR PRESSURE (PSIG)

(J)

W J:

()

Z --

..J W >

W

..J C

W

~

o -c z

250 200 OEOP-01-UG ATTACHMENT 6 Page 19 of 19 FIGURE 21 Reactor Water Level at MSL (Main Steam Line Flood Level)

MSL 00 300 500 60 200 400 600 800 1,000 REACTOR PRESSURE (PSIG)

WHEN REACTOR PRESSURE IS LESS THAN 60 PSIG, USE INDICATED LEVEL.

MSL IS +250 INCHES.

Rev. 55 REF lEG TEMP ABQVEOR EQUAL TO 200-F

.r REF LEG

,. TEMP BELOW 200"F Page 106 of 151

Categories KJA:

295008 A2.05 Tier / Group:

T1G2 RU Rating:

2.9 SRO Rating:

3.1 LP Ubj:

300-C, 10 Source:

NEW Cog Level:

HIGH Category 8:

Y Categories KIA:

295008 A2.05 Tier / Group: TlG2 RORating:

2.9 SRORating

3.1 LP Obj:

300-C,1O Source:

NEW Cog Level:

HIGH Category 8:

Y

85. Unit Two is operating at 74% power when the FW-V120,FW HTRS 4 & 5 BYP VLV, is inadvertantly opened by mechanics. The valve is bound and can not be reclosed.

Initial Final Feedwater Temperature was 404°F.

Conditions are now stable with reactor power at 81% and Final Feedwater Temperature at 314°F.

(Reference provided)

Which one of the following identifies the required action based on the information above?

Continued operation:

Av is not allowed and reactor shutdown is required lAW OGP-05, Unit Shutdown.

B. is not allowed and a manual reactor scram is required lAW 001-01.01, BNP Conduct of Operations Supplement.

C. is allowed provided the FW Heaters 4 & 5 are isolated lAW 20P-32, Condensate and Feedwater Operating Procedure.

D. is allowed provided reduced thermal limits are established within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> as required by Technical Specifications.

85. Unit Two is operating at 74% power when the FW-V120, "FW HTRS 4 & 5 BYP VLV, is inadvertantly opened by mechanics. The valve is bound and can not be reclosed.

Initial Final Feedwater Temperature was 404°F.

Conditions are now stable with reactor power at 81 % and Final Feedwater Temperature at 314°F.

(Reference provided)

Which one of the following identifies the required action based on the information above?

Continued operation:

A'! is not allowed and reactor shutdown is required lAW OGP-05, Unit Shutdown.

B. is not allowed and a manual reactor scram is required lAW 001-01.01, BNP Conduct of Operations Supplement.

C. is allowed provided the FW Heaters 4 & 5 are isolated lAW 20P-32, Condensate and Feedwater Operating Procedure.

D. is allowed provided reduced thermal limits are established within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> as required by Technical Specifications.

Feedback K/A: 295014 G2.01.25 Ability to interpret reference materials, such as graphs, curves, tables, etc.

Inadvertent Reactivity Addition (CFR: 41.10 / 43.5 / 45.12)

ROISRO Rating: 3.9/4.2 Objective: CLS-LP-34, Obj. 1 Ic Given plant conditions, describe the effect a loss/malfunction of the feedwater heaters will have on:

c. Feedwater Temperature

Reference:

20P-32, Attachment 4 (provided)

Cog Level HI Explanation:

From Attachment 4 of 20P-32 operation is outside of the allowable range (<1 10.3°F) this wil require a Unit shutdown lAW GP-05.

Distractor Analysis:

Choice A: Correct see explanation Choice B: Plausible because the 01 has a table with the Selected Out-of-Service Equipment Contingencies. In this case the FW heater meets the definition of the heater OOS and operation is permitted.

Choice C: Plausible because operation is allowed if the FW heaters are isolated but not with the final temperature greater than the 110.3 limit.

Choice D: Plausible because if turbine bypass is inoperable with a FW Heater OOS then TS 3.7.6 requires this action.

SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations (43(b)(5) a. Assessement of the plant conditions and then prescribing the shutdown lAW the GP.

Notes Feedback KIA: 295014 G2.01.25 Ability to interpret reference materials, such as graphs, curves, tables, etc.

Inadvertent Reactivity Addition (CFR: 41.10/43.5/45.12)

RO/SRO Rating: 3.9/4.2 Objective: CLS-LP-34, Obj. 11c Given plant conditions, describe the effect a loss/malfunction of the feedwater heaters wi" have on:

c. Feedwater Temperature

Reference:

20P-32, Attachment 4 (provided)

Cog Level HI Explanation:

From Attachment 4 of 20P-32 operation is outside of the a"owable range <<11 0.3°F) this wil require a Unit shutdown lAW GP-05.

Distractor Analysis:

Choice A: Correct see explanation Choice B: Plausible because the 01 has a table with the Selected Out-of-Service Equipment Contingencies. In this case the FW heater meets the definition of the heater OOS and operation is permitted.

Choice C: Plausible because operation is allowed if the FW heaters are isolated but not with the final temperature greater than the 110.3 limit.

Choice 0: Plausible because if turbine bypass is inoperable with a FW Heater OOS then TS 3.7.6 requires this action.

SRO Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations (43(b)(5) a. Assessement of the plant conditions and then prescribing the shutdown lAW the GP.

Notes

U 0

C 1)

C C C4 CJ cN C c C C4 C C c4 - - - - - -

C C) C C) C) C C C C C CrJ j cl)c

LL

  • r*

j iZ D

C CJ c c CJ CJ 0

C)

çL)

t 1 t 1

  • t i t t J f f C) C C (V) C C3 Cr)

Cr) cc Lfl in in in in cc cc cc cc r r-r cc cc cc in in ci - cc L -r c cJ

cc in

a c

cc

COCCCQCO rzrr U

x 0)

CO I-W

[C) t C) CI,

00) 0) I-. W

[C)

) C1.

00)

CO I-.

CO

[C)

C)

Ck

0) 0) 0) 0)0) 0) 0) 0) 0) 0)

CO CO CO CO CO CO CO CO CO CO -

I 1 I - I I

ci) 0 Co>

I-ci)0 Ea)

I ci)a)

U (5

LL Ca)

E

-c C)

(5 (4

a0 E

2 LL From OP-32, Attachment 4, Final Feedwater Temperature Vs Power RX Nominal Nominal

  • 1*1O.3°F PWR FVV Temp RlV Temp Reduced F'vV Reduced Temp

'10°F 100 429.0 4'19.0 328.7 99 427.6 4'17.6 327.7 98 426.5 4'16.. 6 327.0 97 425.5 4'15.6 326.3 96 424.. 4 4']4.6 325.7 95 423.4 4'13.6 325.0 94 422.4 4'12.6 324.3 93 42'1.4 4'I'L7 323.7 9,2 420.4 4'10..7 323.0 91 419.5 409.8 322.4 90 4-18.5 408.8 321.7 89 4'17.5 407.9 32'1.'1 88 416.5 406.. 9 320.4 87 415.6 406.0 319.8 86 4'14.6 405.0 319:1 85 413.6 404.1 318.5 84 4-12.6 403.. 1 317.8 83 41'1.7 402.2 317.2 82 4-10.7 40'1.2 316.5 81 409.7 400.3 315.8 80 408.7 399.3 315.2 79 407.6 398.3 314.5 78 406.6 397.3 313.8 77 405.6 396.3 313:1 76 404.5 395.3 312.4 75 403.5 394.2 31'1.7 74 402.4 393.2 311.0 73 40'1.3 392.1 310.3

CAUTION Unit operation outside the bounds of the Loss of Feedwater Heating analysis is prohibited.

9.

IF Step 8.7.2.8c criteria is NOT met. THEN PERFORM the following:

a.

IMMEDIATELY NOTIFY the Unit SCO b

RESTORE unit operation within the bounds of the cycle Loss of Feedwater Heating analysis OR c

COMMENCE unit shutdown in accordance with OGP-05.

20P-32 Rev. 165 Page 117 of 300 Permitted Condition Operation Comment 005 SingI (See NOTES)

FWHOOS Yes Defined as a ID F or greater reduction n nominal feedwater temperature.

FWR FFrR Yes Defined as a 10 F or greater reduction n feecwater temperature.

Defined as a cycle extension strategy.

MSIVOOS Yes-base MSIVOOS permits I MSIV to be inoperable.

IF MSIVOOS. THEN thermal power shall be limited to 70% of rated.

TBPCCS Yes TSPOOS assumes a! turbine bypass valves (ThV) are incoerable.

SLO Yes Permitted with a thermal limi: penalty.

005 Combination (See NOTES)

TBPOOS & WH003 Yes Combined 003 condition is permtted with a thermal limit penalty.

TBPOOS Yes Combined 003 condtion s permitted with a thermal imit penatty.

& FWTR (FFTR)

Operefinq Power/Flow Map ICF Yes-base Permitted operations with thermal limits defined by 003 condition.

Power Coas:down Yes-base Permitted operations with thermal limits detned by OCS condition.

Turbine Control Mode Yes-base Partial arc cpera:ion is supported by safety analysis for all 005 conditions RCR Pump Per Source Yes-base Power source can be protiided by the UAT or SAT or all ODS conditions.

Yes:

Operations are permitted with restrictive thermal limits.

Yes-base:

Operations are permitted with base thermal limits. fo thermal flnit changes are required.

001-01.01 Rev. 29 Page 121 of 177 Categories K/A:

295014 G2.01.25 Tier/Group:

T1G2 RO Rating:

3.9 SRO Rating:

4.2 LPObj

34-11C Source:

NEW Cog Level:

HIGH Category 8:

CAUTION Unit operation outside tile bounds of the Loss of Feedwater Heating analysis is prohibited.

9.

IF step 8.7.2.8.c criteria is NOT met, THEN PERFORM the following:

120P-32 Condition

a.

IMMEDIATELY NOTIFY the Unit seo.

b.

RESTORE unit operation within the bounds of the cycle Loss of Feedvvater Heating analysis OR

c.

COMMENCE unit shutdown in accordance 'ljVith OGP-05.

Rev. 165 Page 1-17 of 300 1 Permitted Operation Comment DOS Single (See NOTES)

FWHOOS Yes Defined as a 10°F or greater reduction in nom inal feedwater temperature.

FW,R I,FFTR)

Yes Delined as a 10 'F or greater reduction in feedwater temperature.

Defined as a cycle extension strategy.

MSIIfOOS Yes-base

IF MSIVOOS_ THEN thermal power shall be limited to 70% of rated.

TepOOS Yes TSPOOS assumes all turbine bypass valves (I BV) are inoperable.

SLO Yes Permitted with a ihemlallimii penalty.

ODS Combination (See NOTES)

TSPOOS & FlNHOOS Yes Combined OOS condition is permitted with a thermal limit penalty.

TBPOOS Yes Combined 005 condition is permitted with a thermal limit penalty.

& FltHR (FFTR)

Operating Power/Flow Map'" ICF Yes-base

  • Pem,iited operations with thermal limits defined by OOS condition.

Power Coastdown Yes-base

  • Pemlitted operations with thermal limits defined by OOS condition.

Turbine Control Mode Yes-base

  • Partial aro operation is supported by saier{ analysis for all OOS conditions.

RCR Pump Pwr Source Yes-base

  • Power source can be provided by the UA T or SAT for all OOS conditions.

Yes:

Operations are permitted with restrictive thermal limits.

Ye.s-base:

Operations are permitted with base thermal limits. No thermallirnit changes are required.

1001-01.01 Categories KIA:

RORating:

LPObj:

Cog Level:

295014 G2.01.25 3.9 34-11C HIGH Rev. 29 Tier / Group: T1 G2 SRO Rating:

4.2 Source

NEW Category 8:

Page 12-1 of '1771

8.7.2 Procedural Steps Initials CAUTION Opening the Feedwater Heater tube side vents will release hot discharges under pressure to the drain trough.

z.

PERFORM the following to vent the tube side of the 4A(B) feed water heater:

OPEN FEEDWA TER HEA TER 4A (B)

CHANNEL INBOARD VENT VALVE, MVD-V69(V76).

CRACK OPEN FEED WA TER HEA TER 4A(B) OUTBOARD CHANNEL VENT VALVE, MVD-V70(V75), to establish a vent path.

aa.

PERFORM the following to vent the tube side of the 5A(B) feed water heater:

OPEN FEED WA TER HEA TER 5A(B)

CHANNEL INBOARD VENT VALVE, MVD-V8 I (V88).

CRACK OPEN FEED WA TER HEA TER 5A(B) CHANNEL OUTBOARD VENT VALVE, MVD-V82(V87), to establish a vent path.

NOTE:

Step 8.7.2.8 ensures unit operation with reduced feedwater temperature is bounded by the cycle Loss of Feedwater Heating analysis.

8.

EVALUATE reduction in final feedwater temperature for compliance with Loss of Feedwater Heating analysis as follows:

a.

RECORD current final feedwater temperature from PPC Display 825.

20P-32 Rev. 166 Page 116 of 301 8.7.2 Procedural Steps CAUTION Opening the Feedwater Heater tube side vents will release hot discharges under pressure to the drain trough.

Z.

PERFORM the following to vent the tube side of the 4A(B) feed water heater:

OPEN FEEDWATER HEATER 4A(B)

CHANNEL INBOARD VENT VAL VE, MVD-V69(V76).

CRACK OPEN FEEDWATER HEATER 4A(B) OUTBOARD CHANNEL VENT VALVE, MVD-V70(V75), to establish a vent path.

aa.

PERFORM the following to vent the tube side of the 5A(B) feed water heater:

OPEN FEEDWATER HEATER 5A(B)

CHANNEL INBOARD VENT VALVE, MVD-V81 (V88).

CRACK OPEN FEEDWATER HEATER 5A(B) CHANNEL OUTBOARD VENT VALVE, MVD-V82(V87), to establish a vent path.

Initials NOTE:

Step 8.7.2.8 ensures unit operation with reduced feedwater temperature is bounded by the cycle Loss of Feedwater Heating analysis.

120P-32

8.

EVALUATE reduction in final feedwater temperature for compliance with Loss of Feedwater Heating analysis as follows:

a.

RECORD current final feedwater temperature from PPC Display 825.

Sllj of.

Rev. 166 Page 116 of 301 I

8.7.2 Procedural Steps Initials b.

RECORD 110.3°F Reduced FFWT value for current reactor power from Attachment 4.

c.

CONFIRM reduction in final feedwater temperature is less than 110.3°F by comparing the following:

8.7.2.8.a 8.7.2.8.b CAUTION Unit operation outside the bounds of the Loss of Feedwater Heating analysis is prohibited.

9.

IF Step 8.7.2.8.c criteria is NOT met, THEN PERFORM the following:

a.

IMMEDIATELY NOTIFY the Unit CRS.

b.

RESTORE unit operation within the bounds of the cycle Loss of Feedwater Heating analysis qj c.

COMMENCE unit shutdown in accordance with OGP-05.

10.

IF feedwater temperature is more than 10°F below nominal (refer to Attachment 4), AND reactor power is greater than or equal to 30% of rated thermal power, THEN PERFORM the following:

a.

ENSURE reactor operation in accordance with applicable FWTR Power to Flow Map.

b.

REFER to 201-03.2 for required actions.

20P-32 Rev. 166 Page 117 of 301 8.7.2 Procedural Steps

b.

RECORD 110.3°F Reduced FFWT value for current reactor power from Attachment 4.

c.

OF CONFIRM reduction in final feedwater temperature is less than 11 0.3°F by comparing the following:

3/)(

OF B.7.2.B.a B.7.2.B.b CAUTION Unit operation outside the bounds of the Loss of Feedwater Heating analysis is prohibited.

9.

IF Step B.7.2.B.c criteria is NOT met, THEN PERFORM the following:

a.

IMMEDIATELY NOTIFY the Unit CRS.

b.

RESTORE unit operation within the bounds of the cycle Loss of Feedwater Heating analysis

c.

COMMENCE unit shutdown in accordance with OGP-05.

10.

IF feedwater temperature is more than 10°F below nominal (refer to Attachment 4), AND reactor power is greater than or equal to 30% of rated thermal power, THEN PERFORM the following:

a.

ENSURE reactor operation in accordance with applicable FWTR Power to Flow Map.

b.

REFER to 201-03.2 for required actions.

Initials 120P-32 Rev. 166 Page 117 of 301 I

8.7.2 Procedural Steps Initials 11.

CONFIRM feedwater flow temperature compensation is accurate by performing the following:

NOTE:

Feedwater Line A temperature can be obtained from any of the following:

PPC Point U2CP_B050 PPC Point U2CP_B051 Feedwater Lines Temperature Recorder, B21-TR-5515 (20 el. Reactor Building) a.

DETERMINE Feedwater Line A temperature AND RECORD temperature and instrument used below:

FW Line A temp Instrument NOTE:

Feedwater Line B temperature can be obtained from any of the following:

PPC Point U2CP_B052 PPC Point U2CP_B053 Feedwater Lines Temperature Recorder, B21TR-5515, (20 el. Reactor Building) b.

DETERMINE Feedwater Line B temperature AND RECORD temperature and instrument used below:

FW Line B temp Instrument c.

OBTAIN Feedwater Line A feedwater flow compensation value using Feedwater Line A temperature recorded in Step 8.7.2.11.a and AND RECORD on Attachment 10, column 1.

20P-32 Rev. 166 Page 1 18 of 301 8.7.2 Procedural Steps Initials NOTE:

NOTE:

120P-32

11.

CONFIRM feedwater flow temperature compensation is accurate by performing the following:

Feedwater Line A temperature can be obtained from any of the following:

PPC Point U2CP B050 PPC Point U2CP B051 Feedwater Lines Temperature Recorder, 821-TR-5515 (20' el. Reactor Building)

a.

DETERMINE Feedwater Line A temperature AND RECORD temperature and instrument used below:

OF FW Line A temp Instrument Feedwater Line B temperature can be obtained from any of the following:

PPC Point U2CP B052 PPC Point U2CP B053 Feedwater Lines Temperature Recorder, 821-TR-5515, (20' el. Reactor Building)

b.

DETERMINE Feedwater Line B temperature AND RECORD temperature and instrument used below:

OF FW Line B temp Instrument

c.

OBTAIN Feedwater Line A feedwater flow compensation value using Feedwater Line A temperature recorded in Step 8.7.2.11.a and AND RECORD on Attachment 10, column 1.

Rev. 166 Page 118 of 301 I

8.7.2 Procedural Steps Initials d.

OBTAIN Feedwater Line B feedwater flow compensation value using Feedwater Line B temperature recorded in Step 8.7.2.11.b and AND RECORD on Attachment 10, column 1.

NOTE:

Process Computer compensation values are located on the second page of the C32B022/B023 screen under HANDLING PARAMETERS, CORRECTION TYPE FACTOR 0.

e.

OBTAIN CORRECTION TYPE FACTOR 0 compensation value for PPC Point U2C32B022 (Feedwater Line A) AND RECORD on 0, column 2.

f.

OBTAIN CORRECTION TYPE FACTOR 0 compensation value for PPC Point U2C32B023 (Feedwater Line B) AND RECORD in 0, column 2.

NOTE:

IF the values compared in the following step are within 0.002, THEN feedwater flow temperature compensation is accurate.

g.

VERIFY the values on Attachment 10, columns 1 and 2 for Feedwater Line A are within 0.002 AND DOCUMENT on Attachment 10.

h.

VERIFY the values on Attachment 10, columns I and 2 for Feedwater Line B are within 0.002 AND DOCUMENT on Attachment 10.

i.

IF feedwater flow temperature compensation is NOT accurate, THEN IMMEDIATELY NOTIFY the duty Reactor Engineer.

2OP-32 Rev. 166 Page 119 of 301 8.7.2 Procedural Steps

d.

OBTAIN Feedwater Line B feedwater flow compensation value using Feedwater Line B temperature recorded in Step 8.7.2.11.b and AND RECORD on Attachment 10, column 1.

Initials NOTE:

Process Computer compensation values are located on the second page of the C32B022/B023 screen under HANDLING PARAMETERS, CORRECTION TYPE FACTOR O.

e.

OBTAIN CORRECTION TYPE FACTOR 0 compensation value for PPC Point U2C32B022 (Feedwater Line A) AND RECORD on 0, column 2.

f.

OBTAIN CORRECTION TYPE FACTOR 0 compensation value for PPC Point U2C32B023 (Feedwater Line B) AND RECORD in 0, column 2.

NOTE:

IF the values compared in the following step are within 0.002, THEN feedwater flow temperature compensation is accurate.

\\20P-32

g.

VERIFY the values on Attachment 10, columns 1 and 2 for Feedwater Line A are within 0.002 AND DOCUMENT on Attachment 10.

h.

VERIFY the values on Attachment 10, columns 1 and 2 for Feedwater Line B are within 0.002 AND DOCUMENT on Attachment 10.

i.

IF feedwater flow temperature compensation is NOT accurate, THEN IMMEDIATELY NOTIFY the duty Reactor Engineer.

Rev. 166 Page 119 of 301 \\

86. While in Mode 3 with Shutdown Cooling (SDC) in service on Unit One, a complete Loss of Off-site Power (LOOP) occurs.

The 1-E11-F009, RHR Shutdown Cooling Inboard Isolation Valve, mechanically binds in a mid-position and cannot be fully opened.

Which one of the following is the minimum level required to support natural circulation and identifies the procedural method for Decay Heat removal that is available?

The minimum Reactor Water Level to support Natural Circulation is (1) inches.

The available method of decay heat removal is (2)

A.

(1) 200 (2) Alternate Decay Heat Removal Using Natural Circulation and FPCCS and SSFPC lAW lOP-I 7, Residual Heat Removal System Operating Procedure B(I) 200 (2) Alternate Shutdown Cooling lAW OAOP-l 5.0, Loss of Shutdown Cooling C. (1) 254 (2) Alternate Decay Heat Removal Using Natural Circulation and FPCCS and SSFPC lAW lOP-i 7, Residual Heat Removal System Operating Procedure D. (1) 254 (2) Alternate Shutdown Cooling lAW OAOP-15.0, Loss of Shutdown Cooling

86. While in Mode 3 with Shutdown Cooling (SDC) in service on Unit One, a complete Loss of Off-site Power (LOOP) occurs.

The 1-E11-F009, RHR Shutdown Cooling Inboard Isolation Valve, mechanically binds in a mid-position and cannot be fully opened.

Which one of the following is the minimum level required to support natural circulation and identifies the procedural method for Decay Heat removal that is available?

The minimum Reactor Water Level to support Natural Circulation is (1) inches.

The available method of decay heat removal is (2)

A. (1) 200 (2) Alternate Decay Heat Removal Using Natural Circulation and FPCCS and SSFPC lAW 1 OP-17, Residual Heat Removal System Operating Procedure B~ (1) 200 (2) Alternate Shutdown Cooling lAW OAOP-15.0, Loss of Shutdown Cooling C. (1) 254 (2) Alternate Decay Heat Removal Using Natural Circulation and FPCCS and SSFPC lAW 10P-17, Residual Heat Removal System Operating Procedure D. (1) 254 (2) Alternate Shutdown Cooling lAW OAOP-15.0, Loss of Shutdown Cooling

Feedback KIA: S295021 A2.03 Ability to determine andlor interpret the following as they apply to LOSS OF SHUTDOWN COOLING:

Reactor water level (CFR: 41.10/43.5/45.13)

ROISRO Rating: 3.5/3.5 Objective: CLS-LP-1 20*06

6. Describe how to determine when natural circulation exists within the Reactor Vessel.

Reference:

OAOP-15, Revision 23, Page 11, Section 3.2.14 Cog Level: High Explanation:

During conditions in which there is no circulation, the reactor vessel water level, as read on B21-LI-R605A(B), should be maintained between 200 and 220, or as directed by the Shift Superintendent based on plant conditions, until forced circulation is restored. With a LOOP present and no actions taken to restore Off-site power (not provided in the question), the only available means of decay heat removal is alternate shutdown cooling utilizing SRVs.

Distractor Analysis:

Choice A: Plausible because 200 inches is correct and OP-17 contains actions for using FPC and SSFPC, but these would only be available if the reactor head is removed and fuel pool gates removed.

Choice B: Correct Answer Choice C: Plausible because 254 inches is the level of the MSLs and could be confused with Natural Circulation level due to the requirement to be at this level during alternate SDC and OP-17 contains actions for using FPC and SSFPC, but these would only be available if the reactor head is removed and fuel pool gates removed.

Choice D: Plausible because Plausible because 254 inches is the level of the MSLs and could be confused with Natural Circulation level due to the requirement to be at this level during alternate SDC SRO Only Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations (43(b)(5)a. Requires assessing plant conditions (LOOP, Mode 3, power availability, impact of power losses) and prescribing correct section of a procedure to provide DHR.

Notes Feedback KIA: S295021 A2.03 Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING:

Reactor water level (CFR: 41.10/43.5/45.13)

RO/SRO Rating: 3.5/3.5 Objective: CLS-LP-120*06

6. Describe how to determine when natural circulation exists within the Reactor Vessel.

Reference:

OAOP-15, Revision 23, Page 11, Section 3.2.14 Cog Level: High Explanation:

During conditions in which there is no circulation, the reactor vessel water level, as read on 821-Ll-R605A(8), should be maintained between 200" and 220", or as directed by the Shift Superintendent based on plant conditions, until forced circulation is restored. With a LOOP present and no actions taken to restore Off-site power (not provided in the question), the only available means of decay heat removal is alternate shutdown cooling utilizing SRVs.

Distractor Analysis:

Choice A: Plausible because 200 inches is correct and OP-17 contains actions for using FPC and SSFPC, but these would only be available if the reactor head is removed and fuel pool gates removed.

Choice B: Correct Answer Choice C: Plausible because 254 inches is the level of the MSLs and could be confused with Natural Circulation level due to the requirement to be at this level during alternate SDC and OP-17 contains actions for using FPC and SSFPC, but these would only be available if the reactor head is removed and fuel pool gates removed.

Choice D: Plausible because Plausible because 254 inches is the level of the MSLs and could be confused with Natural Circulation level due to the requirement to be at this level during alternate SDC SRO Only Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations (43(b)(5)a. Requires assessing plant conditions (LOOP, Mode 3, power availability, impact of power losses) and prescribing correct section of a procedure to provide DHR.

Notes

2.0 AUTOMATIC ACTIONS

Loop A(B) INBOARD INJECTION V4LVE E11-FO15.4iB, will close (Low Level One Only)

The RHR Pump in service for Shutdown Cooling will trip on a loss of suction path.

3.0 OPERATOR ACTIONS 3,1 Immediate Actions None 3.2 Supplementary Actions ii 3.2.1 IF Shutdown Cooling has been lost due to a tripped RHR

[]

Pump, THEN START an RHR Pump in the loop being used for Shutdown Cooling.

NOTE:

During conditions in which there is no circulation, the reactor vessel water level, as read on 82f-LI-R6OA8, should be maintained between 200 and 220, or as directed by the Shift Superintendent based on plant conditions.

until forced circulation is restored.

3.2.2 IF forced circulation has been lost, AND natural circulation has NOT been established, THEN RESTORE AND MAINTAIN reactor vessel water level.

OAOP-15.0 Rev. 23 Page 3 of 21 CAUTION If reactor coolant ten,pera:ire is grea:er than 212rF and reac:or water level has l:ieen raised to greater than 212 inches or 10 minLtes or more, a false RPV low level signal could result when the reference leg condensing pot N12A(B: n3zzie is uncovered as level is subsequentl lowered below 21S inches.

2.0 AUTOMATIC ACTIONS Loop /\\(8) INBOARD INJECTION 1I.4L VE, 0

E11-F015A(B), will close (Low Level One Only)

The RHR Pump in service for Shutdown Cooling will trip 0

on a loss of suction path.

3.0 OPERATOR ACTIONS 3.1 Immediate Actions None 3.2 Supplementary Actions CAUT.lON If reactor coolant temperature is greater tllan 21.2°F and reactor water level has been raised to greater than 218 inches for 10 minutes or more, a false RPV low level signal could result when the reference leg condensing pot N12A(B) nozzle is uncovered as level is subsequently lowered below 218 inches.

3.2.1 IF Shutdown Cooling has been lost due to a tripped RHR Pump, THEN START an RHR Pump in the loop being used for Shutdown Cooling.

o NOTE:

During conditions in which there is no circulation, the reactor vessel water level, as read on B21-Ll-R605A(B), should be maintained between 200" and 220~, or as directed by the Shift Superintendent based on plant conditions, unol rorced circulation is restored.

3.2.2 IF forced circulation has been lost, AND natural 0

circulation has NOT been established, THEN RESTORE AND MAINTAIN reactor vessel water level.

IOAOP-15.0 Rev. 23 Page 3 of21 I

10 OPERATOR ACTIONS j,

IF the reactor coolant temperature is less than 212°F, THEN ENSURE the following valves are open:

INBOARD RX HE4D VENT VLV. 821-F003 OUTBOARD PXHEAD VENT VLY. B21-F004.

k.

MAINTAIN RHR in Shutdown Cooling in accordance with 1(2)OP-17.

IF RHR has NOT been restored in accordance with Step 3.2.11.5, THEN PLACE the RHR loop that was operating in Shutdown Cooling back in service in accordance with I (2)OP-1 7 as soon as conditions permit.

U U

U 3.2.12 IF necessary to minimize reactor coolant temperature rise, THEN PERFORM one of the following feed and I

bleed combinations:

FEED BLEED CONDIFW in accordance with RWCU Reject in accordance 1(2)OP-32 with 1(2)OP-14 CRD in accordance with Reactor Water Level Control 1(2)OP-08 using Main Steam Lines in accordance with 1 (2)OP-32.

Core Spray in accordance Maintaining RPV Level Using with 1(2)OP-18 the Main Steam Line Drains LPCI in accordance with with 1(2)OP-25.

1(2)OP-1 7 IF NEITHER RHR loop can be placed in Shutdown Cooling, THEN PLACE the Condensate System in Condenser Cooling in accordance with 1(2)OP-32.

OAOP-15.0 I

Rev.23 Page 10 of 21 Not AvaD (LOOP) jvail (LOOP)

I 3-23 U

Not Avail (RPS not reset)

U 3.0 OPERATOR ACTIONS

j.

IF the reactor coolant temperature is less than 212°F, THEN ENSURE the following valves are open:

iNBOARD RX HE.4D VENT VLV, B21-F003 0

OUTBOARD RX HEAD VENT VLV B21-F004.

0

k.

MAINTAIN RHR in Shutdown Cooling in accordance with 1 (2)OP-17.

IF RHR has NOT been restored in accordance with Step 3.2.11.5, THEN PLACE the RHR loop that was operating in Shutdown Cooling back in service in accordance with 1 (2)OP-17 as soon as conditions permit.

o o

3.2.12 IF necessary to minimize reactor coolant temperature rise THEN PERFORM one of the following feed and o

P) I bleed combinations:

I~ot Avail (RPS notJ reset)

FEED BLEED CONDJFW in accordance with RWCU Reject in accordance 1 (2)OP-32 with 1(2)OP-14 CRD in accordance with Reactor Water level Control 1 (2)OP-08 using Main Steam Lines in accordance with 1 (2)OP-32.

Core Spray in accordance Maintaining RPV Level Using I with 1(2}OP-18 the Main Steam Line Drains LPCI in accordance with with 1 (2)OP-25.

1(2)OP-17 IF NEITHER RHR loop can lJe placed in Shutdown 0

Cooling, THEN PLACE the Condensate System in Condenser Cooling in accordance "vith 1(2}OP-32.

IOAOP-15.0 Rev. 23 Page 10 of21 I

3.0 OPERATOR ACTIONS 3.2.14 IF ALL of the above methods can NOT maintain reactor vessel coolant temperature below 212°F, THEN INITIATE alternate Shutdown Cooling with the SRVs as follows:

1.

ENSURE ALL control rods are fully inserteth 2.

CONFIRM reactor vessel head is installed and tensioned.

3.

IF the Reactor Recirculation Pumps are running, THEN PERFORM the following:

a.

RAISE AND MAINTAIN reactor water level between 200 and 220 as read on 321-LI-R6O548J. or as directed by Shift Superintendent based on plant conditions.

b.

STOP the running Reactor Recirculation Pumps in accordance with 1(2)OP-02.

4.

SHUT DOWN the RI-IR loop that was operating in U

Shutdown Cooling in accordance with 1(2)OP-17.

5.

PLACE one RHR loop in the Suppression Pool Cooling U

mode in accordance with 1(2)OP-17.

6.

IF Suppression Pool temperature rises above 95°F.

U THEN GO TO OEOP-02-PCCP, Primary Containment Control Procedure AND PERFORM CONCURRENTLY with this procedure.

J OAOP-15.O Rev. 23 Page 11 of 21 Categories K/A:

S295021 A2.03 Tier/Group:

T1G1 RO Rating:

3.5 SRO Rating:

3.5 LP Obj:

CLSLP120*06 Source:

NEW Cog Level:

NIGH Category 8:

3.0 OPERATOR ACTIONS 3.2.14 IF ALL of the allove methods can NOT maintain reactor vessel coolant temperature below 212°F, THEN INITIATE alternate Shutdown Cooling with. the SRVs as follows:

1.

ENSURE ALL control rods are funy inserted.

0

2.

CONFIRM reactor vessel head is installed and 0

tensioned.

3.

IF the Reactor Recirculation Pumps are running: THEN PERFORM the following:

a.

RAISE AND MAINTAIN reactor water level 0

between 200" and 220" as read on B21-U-R605A(8), or as directed by Shift Superintendent I)ased on plant conditions.

b.

STOP the running Reactor Recirculation Pumps in 0

accordance with 1 (2)OP-02.

4.

SHUT DOWN the RHR loop that was operating in Shutdown Cooltng in accordance with 1(2)OP-17.

5.

PLACE one RHR loop in the Suppression Pool Cooling mode in accordance with 1 (2)OP-17.

6.

IF Suppression Pool temperature rises above 95QF, THEN GO TO OEOP-02-PCCP, Primary Containment Control Procedure AND PERFORM CONCURRENTLY with this procedure.

IOAOP-15.0 Rev. 23 Categories KIA:

S295021 A2.03 Tier / Group: T1 G 1 SRO Rating:

3.5 RO Rating:

3.5 LP Obj:

CLS-LP-120*06 Source:

~vv Cog Level:

HIGH Category 8:

0 0

0 Page 11 of 21 I

87. While performing refueling activities on Unit Two, a spent fuel bundle was dropped and the following alarms were received:

AREA RAD REFUEL FLOOR HIGH PROCESS FV( BLDG VENT RAD HIGH Which one of the following identifies:

(1) the immediate operator action that is required to be performed and (2) the bases for the performance of this action?

A. (1) Standby Gas Treatment (SBGT)

(2) Ensures control room operators will receive 2 Rem TEDE B.

(1) Standby Gas Treatment (SBGT)

(2) Ensures control room operators will receive. 5 Rem TEDE C. (1) Control Room Emergency Ventilation (CREV)

(2) Ensures control room operators will receive <2 Rem TEDE D (1) Control Room Emergency Ventilation (CREV)

(2) Ensures control room operators will receive 5 Rem TEDE

87. While performing refueling activities on Unit Two, a spent fuel bundle was dropped and the following alarms were received:

AREA RAD REFUEL FLOOR HIGH PROCESS RX BLDG VENT RAD HIGH Which one of the following identifies:

(1) the immediate operator action that is required to be performed and (2) the bases for the performance of this action?

A. (1) Standby Gas Treatment (SBGT)

(2) Ensures control room operators will receive.:5. 2 Rem TEDE B. (1) Standby Gas Treatment (SBGT)

(2) Ensures control room operators will receive.:5. 5 Rem TEDE C. (1) Control Room Emergency Ventilation (CREV)

(2) Ensures control room operators will receive.:5. 2 Rem TEDE D~ (1) Control Room Emergency Ventilation (CREV)

(2) Ensures control room operators will receive.:5. 5 Rem TEDE

Feedback K/A: S295023G 2.04.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

Refueling Accidents (CFR: 41.10 /43.2/45.6)

ROISRO Rating: 4.6/4.4 Objective: CLSLP302..J*02

2. Given plant conditions with spent fuel damage and a high airborne activity problem in progress, determine if the appropriate automatic actions have occurred in accordance with OAOP-5.0, Radioactive Spills, High Radiation, and Airborne Activity.

Reference:

OAOP-05, Revision 23, Page 2, Section 3.1 Cog Level: High Explanation:

OAOP-05 immediate action for a dropped or damaged fuel assembly is to ENSURE CREVS is in operation.

The dose consequence calculation for the fuel handling accident does not credit the secondary containment or automatic CREVS start, however, it does assume that CREVS is manually initiated within 20 minutes of a dropped/damaged fuel assembly. Based on this analysis, Technical Specifications do not require secondary containment or CREVS automatic initiation instrumentation except during Modes 1, 2, or 3 or during operations with the potential to drain the Reactor vessel. The CREV System is designed to maintain a habitable environment in the CRE for a 30 day continuous occupancy after a DBA without exceeding 5 rem total effective dose equivalent (TEDE).

Knowledge of DBA analysis initial conditions.

Distractor Analysis:

Choice A: Plausible because PROCESS RX BLDG VENT RAD HIGH annunciator is easily confused with the auto start for SBGT verifing Auto actions can be confused with Immediate Actions. SBGT start is a supplemental action which will reduce control room dose and 2 Rem TEDE is a site administrative dose limit and can be confused with the actual Dose Analysis from FHA of 2.69 rem TEDE.

Choice B: Plausible because PROCESS RX BLDG VENT RAD HIGH annunciator is easily confused with the auto start for SBGT verifing Auto actions can be confused with Immediate Actions. SBGT start is a supplemental action which will reduce control room dose and 5 Rem TEDE is correct.

Choice C: Plausible because CREV is correct and 2 Rem TEDE is a site administrative dose limit and can be confused with the actual Dose Analysis from FHA of 2.69 rem TEDE.

Choice D: Correct Answer.

SRO Only Basis: Conditions and limitations in the facility license (43(b)(1)

Notes Feedback KIA: S29S023G 2.04.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

Refueling Accidents (CFR: 41.10 / 43.2 /4S.6)

RO/SRO Rating: 4.6/4.4 Objective: CLS-LP-302-J*02

2. Given plant conditions with spent fuel damage and a high airborne activity problem in progress, determine if the appropriate automatic actions have occurred in accordance with OAOP-S.O, Radioactive Spills, High Radiation, and Airborne Activity.

Reference:

OAOP-OS, Revision 23, Page 2, Section 3.1 Cog Level: High Explanation:

OAOP-OS immediate action for a dropped or damaged fuel assembly is to ENSURE CREVS is in operation.

The dose consequence calculation for the fuel handling accident does not credit the secondary containment or automatic CREVS start, however, it does assume that CREVS is manually initiated within 20 minutes of a dropped/damaged fuel assembly. Based on this analysis, Technical Specifications do not require secondary containment or CREVS automatic initiation instrumentation except during Modes 1, 2, or 3 or during operations with the potential to drain the Reactor vessel. The CREV System is designed to maintain a habitable environment in the CRE for a 30 day continuous occupancy after a DBA without exceeding S rem total effective dose equivalent (TEDE).

Knowledge of DBA analysis initial conditions.

Distractor Analysis:

Choice A: Plausible because PROCESS RX BLDG VENT RAD HIGH annunciator is easily confused with the auto start for SBGT verifing Auto actions can be confused with Immediate Actions. SBGT start is a supplemental action which will reduce control room dose and 2 Rem TEDE is a site administrative dose limit and can be confused with the actual Dose Analysis from FHA of 2.69 rem TEDE.

Choice B: Plausible because PROCESS RX BLDG VENT RAD HIGH annunciator is easily confused with the auto start for SBGT verifing Auto actions can be confused with Immediate Actions. SBGT start is a supplemental action which will reduce control room dose and S Rem TEDE is correct.

Choice C: Plausible because CREV is correct and 2 Rem TEDE is a site administrative dose limit and can be confused with the actual Dose Analysis from FHA of 2.69 rem TEDE.

Choice D: Correct Answer.

SRO Only Basis: Conditions and limitations in the facility license (43(b)(1)

Notes

Unit 2 APP UA-03 3-7 Page 1 of 1 AREA RAD REFUEL FLOOR HIGH AUTO ACTIONS NONE CAUSE 1.

High radiation level in the cask wash area.

2.

Circuit malfunction.

3.

Refueling cavity water seal failure.

OBSERVATIONS 1.

ARM indicator and ip unit Upscale light illuminated on Panel H 12-P600.

ACTIONS 1.

Refer to EOP-03-SCCP, Table 3; enter EOP-03-SCCP as appropriate.

2.

Refer to AOP-05.0, Radioacte Spills, High Radiation, and Airborne Activity.

3.

Suspend refueling operation if due to fuel pool low level from refueling cavity water seal leakage.

4.

If a circuit malfunction is suspected, ensure that a Trouble Tag is prepared.

DEVICEISETPOINTS ARM Channel 29 1<2 40 rnRPhr POSSIBLE PLANT EFFECTS 1.

Suspension of refuel floor activities.

REFERENCES 1.

LL-9353-39 2.

AOP-05.0 3.

EOP-03-SCCP 2APP-LIA-03 Rev. 46 Page 34 of AREA RAD REfUEL FLOOR HIGH AUTO ACTIONS NONE

'1.

High radiation level in the cask wash area.

2.

Circuit malfunction.

3.

Refueling cavity water seal failure.

OBSERV.A.TIONS Unit 2 APP U.A.-03 3-7 Page 1 of 1

'1.

ARM indicator and trip unit Upscale light illuminated on Panel H 12-P600.

ACTIONS

1.

Refer to EOP-03-SCCP, Table 3; enter EOP-03-SCCP as appropriate.

2.

Refer to AOP-OS.O, Radioactive Spills, High Radiation, and Airborne Activity.

3.

Suspend refueling operation if due to fuel pool low !evel from refueling cavity water seal leakage.

4.

If a circuit malfunction is suspected, ensure that a Trouble Tag is prepared.

DE\\iICE/SETPOINTS ARM Channel 29 K2 40 mRlhr POSSIBLE PLANT EFFECTS

1.

Suspension of refuel floor activities.

REFERENCES

1.

LL-9353 - 39

2.

AOP-05.0

3.

EOP-03-SCCP 12APP-uA-03 Rev. 46 Page 34 of 631

Unit 2 APP UA-03 4-S Page 1 of 1 PROCESS RX BLDG VENT RAD HIGH AUTO ACTIONS NONE CAUSE 1.

High airborne activity in Reactor Suiding ventilation exhaust p!enum.

2.

Circuit malfunction.

OBSERVATIONS 1.

Reactor Building Vent Rad Recorder D12-RR-R605 ChannelA or B indicates high radiation level.

2.

Reactor Building Exhaust Plenum Rad Monitor Channel A or B indicates greater than 3 rnRlhr on Panel H12-P6G6.

ACTIONS 1.

Enter EOP-03.SCCP. Secondary Containment Conti-o.

2.

Refer to AOP-05.O, Radioacte SpiIs, High Radiation, and Airborne Activity.

3.

If a circuit malfunction is suspected, ensure that a Troube Tag is prepared.

DEVICEISETPOINTS D12-RR-R605 red or black pen 3 mRihr POSSIBLE PLANT EFFECTS 1.

Possible release to environs.

2.

If airborne activity increases to 4 niRihr. Reactor Building HVAC isolation, a Group 6 isolation, drjwell purge isolation, and initiation of the Standby Gas Treatment System ifl1 occur.

REFERENC ES 1.

LL-9353 - 35 2.

AOP-O5.O 3.

EOP-03-SCCP 4.

Plant Modification 85-081 2APP-UA-03 Rev. 46 Page 41 of 63 PROCESS RX BLDG VENT RAD HIGH AUTO ACTIONS NONE Unit 2 APP U.A.-03 4-5 Page'l of 1

1.

High airborne activity in Reactor Building ventilation exhaust plenum.

2.

Circuit malfunction.

OBSERV.A.TiONS

'1.

Reactor Building Vent Rad Recorder D12-RR-R605 Channel A or B indicates high radiation level.

2.

Reactor Building Exhaust Plenum Rad Monitor Channel A or B indicates greater than 3 mRlhr on Panel H12-P606.

ACTIONS

1.

Enter EOP-03-SCCP, Secondary Containment Control.

2.

Refer to AOP-OS.O, Radioactive Spills, High Radiation, and Airborne Activit'!.

3.

If a circuit malfunction is suspected, ensure that a Trouble Tag is prepared.

DEVICE/SETPOINTS D'12-RR-R605 red or black pen 3 mRlhr POSSIBLE PLANT EFFECTS

1.

Possible release to environs.

2.

If airborne activity increases to 4 mRlhr, Reactor Building HVAC isolation, a Group 6 isolation, dr'Jwell purge isolation, and initiation of the Standby Gas Treatment System \\'Iill occur.

REFERENCES

1.

LL-9353 - 35

2.

AOP-05.0

3.

EOP-03-SCCP

4.

Plant Modification 85-081 12APP-uA-03 Page 41 of 631

1.0 SYMPTOMS 1.1 AREA RAD REFUEL FLOOR HIGH (UA-03 3-7) is in alarm.

1.2 AREA RAD NEW FUEL STORAGE HIGH (UA-03 4-7) is in alarm.

1.3 PROCESS RX BLDG VENT RAD i-il (UA-03 4-5) is in alarm.

1.4 TVRB BLDG VENT RAD HIGH (U.4-03 3-3 is in alarm.

1.5 Area Radiation Monitor (ARM) is in alarm.

1.6 Continuous Air Monitor (CAM) is in alarm.

1.7 Turbine Building once-through effluent monitor indicates elevated (higher than expected or an unanticipated increase) activity.

1.8 Routine surveys indicate high radiation, contamination andlor airborne activity.

1.9 Report of spill. leak. or potential damage to new or spent fuel.

2.0 AUTOMATIC ACTIONS 2.1 IF PROCESS RX BLDG VENT RAD HI-HI (UA-03 3-5) is in alam, THEN the following actions occur:

Reactor Building Ventilation isolation

SBGTSautostart U

Group 6 Isolation.

3.0 OPERATOR ACTIONS 3.1 Immediate Actions

]

3.1.1 IF a fuel assembly was dropped or damaged, THEN U

ENSURE the Control Room Emergency Ventilation System (CREVS) is in operation.

OAOP-05.O Rev. 24 Page 2 of 10 1.0 SYMPTOMS 1.1 AREA RAD REFUEL FLOOR HIGH (UA-03 3-7) is in alarm.

1.2 AREA RAD NEW FUEL STORAGE HIGH (UA-03 4-7) is in alarm.

1.3 PROCESS RX BLDG VENT RAD HI (UA-03 4-5) is in alarm.

1.4 TURB BLDG VENT RAD H.IGH (U.4-03 3-3) is in alarm.

1.5 Area Radiation Monitor (ARM) is in alaml.

1.6 Continuous,",ir Monitor (CAM) is in alarm.

1.7 Turbine Building once-through effluent monitor indicates elevated (higher than expected or an unanticipated increase) activity.

1.8 Routine surveys indicate high radiation, contamination and/or airi)ome activity.

1.9 Report of spill, leak, or potential damage to ne'.v or spent fuel.

2.0 AUTOMATIC ACTIONS 2.1 IF PROCESS RX BLDG VENT RAD HI-HI (UA-03 3-5) is in alam1, THEN the following actions occur:

Reactor Building Ventilation isolation 0

SBGTS auto start 0

Group 6 Isolation.

0 3,0 OPERATOR ACTIONS 3.1 Immediate Actions

]

3.1.1 IF a fuel assembly was dropped or damaged, THEN 0

ENSURE the Control Room Emergency Ventilation System (CREVS) is in operation.

IOAOP-OS.O Rev. 24 Page 2 of 10 I

UPDATED FSAR evision:

21 r&L ENGInEERED SAFETYFruREs Chapter:

6

Pmm, Page:

108 o 121 6.44.12 Fuel Handling Accident

- Control Room Dose Section 15.7.1 discusses the release of activity and its transpoi to the environment following a postulated fuel handling accident (FHA).

The design inpiAs utilized to evalua:e tne ntake & ths acv,ty into the control room and to assess the resultant dose to the control room operators are tabulated in Table 6-22. A sensitivity study of unuttered outsde air inleakage nto the control room was performed evaluating inleakage rates of 10,030 ofm (bouxidirig case). 3000 cfrn {confrol room design), and 0 cfm, Acodent XIQ values are developed as discussed in Section 15..2. Section 1.9.S descr bes the parameters utilized in conjunction with the AQTRAO computer code Reterence 8-35} to convert the Alternative Source Temi activity drawn into the control room during the postulatec accident into a total effective dose equivalent TEDE) dose.

Th 30-day FHA dose to the control room operator from the internal cloud associated with the FHA is calculated to be 2.69 rem TEQE.

The onsi:e control room operator dose criterion established by Reference 8-36 for this accident is that the total control room operator dose should be less than the 10 CFR 50.67 guidelines: i.e., that the total dose should be less than S rem TEDE.

C l UPDA TED FSAR ENGINEERED SAFETY FEA TURES 6.4.4.1.2 Fuel Handling Accident - Control Room Dose Revision:

21 Chapter:

6 Page:

108 oi 121 Section 15.7.1 discusses the release of aotivity and its transport to the environment following a postulated fuel handling aoodent (FH..o,).

The design inputs utilized to ellaluate the intake of this acjjvjt~* into the conlrol room and to assess the resultant dose to the control room operators are tabulated in Table 6-28. A sensitivity study of unfiltered outside air inleakage into the oontrol room 'lias performed ellaluating in leakage rates of 10.000 ofm (bounding case). 3000 cfrn {control rcom design). and 0 cfm. Accident XiQ values are developed as discussed in Secticn 15.9.2. Section '15.9.3 describes the parameters utilized in conjunction with the RADTRAD computer code {Reference 8-35} to con~'ert the Alternative Source Tern, activity drawn into the contrcl room during the postulated accident into a total effeclive dose equivalent (TEDE) dose.

The 30-day FHA dose to the conirol room operator from the internal cloud associated with the FHA is c,3lculated to be 2.69 rem TEOE.

The onsile control room operator dose criterion established by Reference 8-36 for this accident is that the total contro!! room cperator dose should be less than the to CFR 50.67 guidelines; i.e.* that the total dose should be less than 5 rem TEDE.

3.0 OPERATOR ACTIONS 3.2.3 IF new or spent fuel damage is suspected, THEN PERFORM the following:

1.

PLACE any fuel that is being moved in a safe condition.

2.

SECURE further fuel movement.

3.

EVACUATE personnel from the following areas:

Refueling Floor 0

Drywell, if occupied Reactor Building, -17 Elev., if Shutdown Cooling in service.

ECCS Pipe Tunnel U

Any area determined to have the potential for high radiation.

4.

ISOLATE Secondary Containment.

5.

START Standby Gas Trains.

U 3.2.4 NOTIFY E&RC to perform the following as necessary:

Area radiation survey U

Air sampling U

Smear survey U

Posttheaffectedareaasnecessary U

Control access to reduce exposure and U

contamination.

DADP-05.0 Rev 23 Page 4 of 10 3.0 OPERATOR ACTIONS 3.2.3 IF new or spent fuel damage is suspected, THEN PERFORM the following:

1.

PLACE any fuel that is being moved in a safe condition.

D

2.

SECURE further fuel movement.

D

3.

EVACUATE personnel trom the following areas:

Refueling Floor D

Drywell, if occupied D

Reactor Building, -17' Elev., if Shutdown Cooling in D

service.

ECCS Pipe Tunnel D

Any area determined to have the potential for high D

radiation.

4.

ISOLATE Secondary Containment D

5.

START Standby Gas Trains.

D 3.2.4 NOTIFY E&RC to perform the following as necessary:

Area radiation survey D

Air sampling D

Smear survey D

Post the affected area as necessary D

Control access to reduce exposure and D

contamination.

IOAOP-05.0 Rev. 23 Page 4 of 10 I

4.0 GENERAL DISCUSSION Liquid radioactive spills may be caused by valve packing leaks, leaky fittings, system leaks, or system draining evolutions. Liquids spills should be covered with an absorbent material to minimize the spread of contamination. Solid spills may be caused by leaks from the containers or process streams which handle radioactive material or by an accident during the transport of new or spent fuel, radioactive sources, or other solid radioactive materials. Solid spills should be covered by a damp material to minimize the spread of airborne contamination. A spill of highly radioactive solid materials such as spent resin, filter sludge, neutron sources, or irradiated reactor internal components may create a serious personnel exposure problem and should be handled with extreme caution. In addition, high radiation and high airborne activity may accompany a spill.

High airborne activity may occur from reactor coolant leaks, coolant spills, radwaste leaks, sampling, grinding, draining, and other maintenance. High airborne activity in the turbine buildings may require ventilation shutdown or realignment to the recirculation lineup if the ventilation systems are operating in the once-through lineup.

High radiation levels may be caused by radiation streaming, loss of or degraded shielding, fuel element damage, high airborne activity, coolant spills, or radiography.

New or spent fuel damage may occur within the plant during fuel handling operations. Fuel may be damaged if it is inadvertently dropped or allowed to collide with objects. Damage may also be sustained if heavy objects (shipping casks, reactor vessel head, drywell head, etc.) are allowed to fall on the fuel. These accidents may release a substantial amount of radioactive noble gases, halogens, and other fission products into the secondary containment. The secondary containment will be automatically isolated due to high radiation at its ventilation exhaust plenum. Although Standby Gas Treatment (SBGT) System will reduce the activity released to the environs, there is a chance that technical specification limits may be exceeded.

The dose consequence calculation for the fuel handling accident does not credit the secondary containment or automatic CREVS start, however, it does assume that CREVS is manually initiated within 20 minutes of a dropped(damaged fuel assembly. Based on this analysis, Technical Specifications do not require secondary containment or CREVS automatic initiation instrumentation except during Modes 1, 2, or 3 or during operations with the potential to drain the Reactor vessel.

OAOP-05.0 Rev. 23 Page 8 of 10 4.0 GENERAL DISCUSSION Liquid radioactive spills may be caused by valve packing leaks, leaky fittings, system leaks, or system draining evolutions. Liquids spills should be covered with an absorbent material to minimize the spread of contamination. Solid spills may I)e caused by leaks from the containers or process streams which handle radioactive material or by an accident during the transport of new or spent fuel, radioactive sources, or other solid radioactive materials. Solid spills should be covered by a damp material to minimize the spread of airborne contamination. A spill of highly radioactive solid materials such as spent resin, filter sludge, neutron sources, or irradiated reactor internal components may create a serious personnel exposure problem and should be handled with extreme caution. In addition, high radiation and higl1 airborne acUvity may accompany a spill.

High airborne activity may occur from reactor coolant leaks, coolant spills, radwaste leal,s, sampling, grinding, draining, and other maintenance. High airborne activity in the turbine buildings may require ventilation shutdown or realignment to the recirculation lineup if the ventilation systems are operating in the once-through lineup.

High radiation [evels may be caused by radiation "streaming," loss of or degraded shielding, fuel element damage, high airi)orne activity, coolant spillS, or radiography.

New or spent fuel damage may occur within the plant during fuel handling operations. Fuel may be damaged if it is inadvertently dropped oral[owed to collide with objects. Damage may also be sustained if heavy objects (shipping casks, reactor vessel head, dlY'Nell head, etc.) are allowed to fall on the fuel. These accidents may release a substantial amount of radioactive noble gases, halogens, and other fission products into the secondary containment The secondary containment will be automatically isolated due to high radiation at its ventilation exhaust plenum. Although Standby Gas Treatment (SBGT) System will reduce the activir; released to the environs, there is a chance that technical speCification limits may be exceeded.

The dose consequence calculation for the fuel handling accident does not credit the secondary containment or automatic CREVS start, however, it does assume that CREVS is manually initiated within 20 minutes of a droppedfdamaged fuel assembly. Based on this analysis, Technical Specifications do not require secondary containment or CREVS automatic initiation instrumentation except during Modes 1, 2, or 3 or during operations with the potential to drain the Reactor vessel.

IOAOP-05.0 Rev. 23 Page B of 10 I

UPDATED FSAR Revion:

21 CP&L EzIGIIIEERED SAFETY FEA TURES Chapter:

3 CHAPTER 6 TABLES 1 of 1 TABLE 6-28 Control Room Design Inputs Design Basis Accidents Control Room 1.

Control room habitabi!ity vo ume

2g8,650 f 2.

Assumed unfiltered inIeakage 10.000 crn Control Room Ventilation 1.

Normal mode operation outside air intake 2,1C0 cm 2.

Normal mode roughing filter, aerosol removal 0%

3.

Normal mode roughing filter, elemental iod:ne removal 0%

4.

Normal mode roughing filter, organic iodine removal 0%

5.

Time of manual switchover from normal to radiation mode 20 minutes 0.

Radiation mode operation outsde sir ntake I.500 cfm 7.

Radiation mode HEPA ft ter. aerosol removal Radiation mode charcoal filter, elemental iodine removal Radiation mode charcoal fiter, organic iodine removal 10.

Radiation train charcoal depth 2 inches 11.

Radiation mode filtered recrculated airflow 40 cfm 12.

Radiation mode aerosol iodine removal 13.

Radiation mode elemental iodine removal 14.

Radiation mode organic iodne removal NOTES

Sensitivity cases using 3.00 cfm and 0 cThi unfiltered outside air inleakage into the control room were also evaluated. The 10,000 cfm unfiltered inleakage case is bounding for the LOCA, the FHA, and the CRDA events.

For the MSLB event, 0 cfm unfiltered outside air nleakage represents the bounding vaiue.

For the MSLB event, a 5ensitlvity study was performed, isolating the control room at various times between 5.0 seconds and 30 days.

TABLE 6-28 UPDA TED FSAR ENGmEERED SAFETY FEA TURES CHAPTER 6 TABLES Control Room Design Inputs - Design Basis Accidents e - roughing filter. aerosol removal

- aerosol iodine removal

- elemental iodine remollal NOTES Revision:

Chapter:

Page:

2gS.650 fl 10.000 cim' 2.100 cfm 0%

0%

0%

20 minutes'"

1.500 cfm 95%

90%

90%

2inehes 400 cfm 95%

90%

90%

Sensitivity cases using 3.000 cfm a,nd 0 dm unfiliered outside air inleakage into the control room were also evaluated. The 10.000 cim unfiltered inleakage case is boundi,ng for the lOCA. the FHA. and the eRDA ellents.

For the MSLB ellent. 0 cfm unfiltered outside air inleakage represents the bounding value.

For the MSLB ellent. a sensitivity study was performed. isolating the contrel room at various times between 5.5 seconds and 30 days.

21 6

1 of

CREV System B 37.3 BASES BACKGROUND The CREV System is designed to maintain a habitable environment in the (connued}

CRE for a 30 day connuous occupancy after a DBA without exceeding 5 rem total effective dose equivalent (TEDE), A single CREV subsystem operating at a flow rate of 2200 cfm will slightly pressurize the CRE relative to outside atmosphere to minimize infiltration of air from surrounding areas adjacent to the CRE bounday. CREV System operation in maintaining CRE habitability is discussed in the UFSAR, Sections 6.4 and 9.4, (Refs. 1 and 2. respectively).

APPLICABLE The ability of the CREV System to maintain the habitability of the CRE SAFETY ANALYSES is an explicit assumption for the design basis accident presented in the UFSAR (Ref. 3). The radiation!smoke protection. mode of the CREV System is assumed (explicitly or implicitly) to operate following a DBA.

The radiological doses to the CRE occupants as a result of a DBA are summarized in Reference 3. Postulated single active failures that may cause the loss of outside or recirculated air from the CRE are bounded by BNP radiological dose calculations for CRE occupants.

Brunswick Unit 2 S 3.7,3-2 Revision No. 61 Categories K/A:

S295023G2.04.49 Tier/Group:

T1G1 RO Rating:

4.6 SRO Rating:

4.4 LP Obj:

CLSLP.3O2J*O2 Source:

NEW Cog Level:

HIGH Category 8:

BASES BACKGROUND (continued}

CREV System B 3.7.3 The CREV System is designed to maintain a habitable environment in the CRE for a 30 day continuous occupancy after a DBA without exceeding 5 rem total effecti.... e dose equivalent {TED E). A single CREV subsystem operating at a f10Vl rate of s: 2200 cfm ',viii sligh!ly pressurize the CRE relative to outside atmosphere to minimize infiltration of air from surrounding areas adjacent to the CRE boundary. CREV System operation in maintaining CRE habitability is discussed in the UFSAR, Sections 6.4 am! 9.4, (Refs. *1 and 2, respectively).

APPLICABLE The ability of the CREV System to maintain the habitability of the CRE SAFETY.ANAL YSES is an explicit assumption ror the design basis accident presented in the UfSAR (Ref. 3}. The radiation/smoke protection mode ofthe CREV System is assumed (explicitly or implicitly) to operate fol!owing a DBA.

The radiological doses to the CRE occupants as a result of a DBA are summarized in Reference 3. Postulated single active failures that may cause the loss of outside or recirculated air from the CRE are bounded by BNP radiological dose calculations for CRE occupants.

Brunswick Unit 2 B 3.7.3-2 Re.... ision No. 61 Categories KIA:

S295023G 2.04.49 Tier / Group: TIGl RORating:

4.6 SRORating

4.4 LP Obj:

CLS-LP-302-J*02 Source:

NEW Cog Level:

HIGH Category 8:

88. An event on Unit One has resulted in the following plant conditions:

Reactor pressure 1000 psig Reactor Water Level 120 inches Control Rod Positions All unknown APRMs Downscale Drywell pressure 3 psig Supp. Pool pressure 2 psig Supp. Pool water temp 150° F Supp. Pool water level

-4 feet (Reference provided)

Which one of the following identifies the status of the Heat Capacity Temperature Limit (HCTL) and the required procedure for reactor pressure control?

HCTL Pressure Control Leg of Procedure A. has been exceeded RVCP B has been exceeded LPC C. has NOT been exceeded RVCP D. has NOT been exceeded LPC

88. An event on Unit One has resulted in the following plant conditions:

Reactor pressure Reactor Water Level Control Rod Positions APRMs Drywell pressure Supp. Pool pressure Supp. Pool water temp Supp. Pool water level (Reference provided) 1000 psig 120 inches All unknown Downscale 3 psig 2 psig 150 0 F

-4 feet Which one of the following identifies the status of the Heat Capacity Temperature Limit (HCTL) and the required procedure for reactor pressure control?

HCTL Pressure Control Leg of Procedure A. has been exceeded RVCP By has been exceeded LPC C. has NOT been exceeded RVCP D. has NOT been exceeded LPC

Feedback K/A: S295026 A2.03 Ability to determine andlor interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:

Reactor pressure (CFR: 41.10/43.5/45.13)

RO/SRO Rating: 3.9/4.0 Objective: CLSLP300L*05a

05. Given the PCCP, determine the appropriate actions if any of the following limits are approached or exceeded:
a. Heat Capacity Temperature Limit.

Reference:

Heat Capacity Temperature Graph only is given to examinee PCCP.

Cog Level: High Explanation:

HCTL has been exceeded. With rods unknown the operator would be in LPC.

Distractor Analysis:

Choice A: Plausible because rods are unknown, would be in LPC.

Choice B: Correct Answer Choice C: Plausible because HCTL has been exceeded. rods are unknown, would be in LPC Choice D: Plausible because HCTL has been exceeded.

SRO Only Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations (43(b)(5)

Notes Feedback KIA: S295026 A2.03 Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:

Reactor pressure (CFR: 41.10 143.5 145.13)

RO/SRO Rating: 3.9/4.0 Objective: ClS-lP-300-l *05a

05. Given the PCCP, determine the appropriate actions if any of the following limits are approached or exceeded:
a. Heat Capacity Temperature Limit.

Reference:

Heat Capacity Temperature Graph only is given to examinee PCCP.

Cog level: High Explanation:

HCTl has been exceeded. With rods unknown the operator would be in lPC.

Distractor Analysis:

Choice A: Plausible because rods are unknown, would be in lPC.

Choice B: Correct Answer Choice C: Plausible because HCTl has been exceeded. rods are unknown, would be in lPC Choice D: Plausible because HCTl has been exceeded.

SRO Only Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations (43(b)(5)

Notes

I I,,C CJ I

j) ti mm c

a 8 8 SUPPRESSION POOL WATER TEMPERATURE (°F)

C)

ç)ç)QQ C)

-4

0 rnti Z

0

--9 0

Z 0

rn rn cn r>

rn C) rn-n 0

0 om m

mm UI I

T1

-o 00 COO m

r-

.rngm CO rn C

0 It rn

-Ui Cl) o

11[

m.,..

D IE[

r rn U

0 TEl W

liii I

I I 1 z <

i-rn tit-r tW mm z

titt llttVtt rmi m

F

o ID 0

C Cl C

LII Cl 0 Cl UI CD fl 1flfll1

-I

---4 g

ATTACHMENT 5 Page '18 of 27 FIGURE 3 Heat Capacity Temperature Limit L1.

~ 220 W

$ 210

~

=f::l=Il

!;;c

=~I n:: 200 UNSAFE ABOVE ~

SELECTED LINE ~

~ 190 5j 180 l-n:: 170 W

~ 160 s:

..J 150 140 o o D..

Z o

=f:: f::

SAFE BELOW 130 =~~ SELECTED LINE (I) 120 (I)

~ 110 D..

D.. 100

J (I)

(-) 0.25 FT

(-) 1.25 FT

(-) 2.50 FT

~~~ (-) 3.25 FT

S:i~

=t=~ (-) 4.25 FT

=I=~

St=~

,-p.,~

=I::~ (-) 5.50 FT
=~~
=~I=

1=1

=I=~
=I=~

,-I-f-I

-1,150 100 300 500 700 900 1,100 o

200 400 600 800 1,000 REACTOR PRESSURE (PSIG)

SUPPRESSION POOL WATER TEr.... 1PERATURE IS DETERMINED BY:

CAC-TR-4426-*1A, POINT WTR AVG OR CAC-TR-4426-2A, POINTWTRAVG OR COrvlPUTER POINT G050 OR COMPUTER POINT G051 OR CAC-TY -4426-1 OR CAC-TY -4426-2 SELECT GRAPH LINE IMMEDIATELY BELOW SUPPRESSION POOL WATER LEVEL AS THE LIMIT.

I OEOP-01-UG Rev. 55 Page 78 of 151 I

YES BNP VOLV1 IEOP-Ot-LPC REldlStONNO 9 NO MAiNTAIN IIEACTORPRE5S BELOW TIlE HEAT CAPACfY TEMP UM1 1RRESPECTPJE OF THE RESULTING tOOLDOWN RATE RCWI1 UNIT i ONLY MAINTAIN REACTOR PRESS BELOW THE HEAT CAPACITY TEMP Ur.tIT IRRESPECTIVE OFTHERESULTING COOLOOWN RATE

INITIATEA REACTOR SCRAM AND ENTER EOP-9i SPIT-99 REDUCE REACTOR PRE PER THE RCIP SECTION OF EOP-91 AS NECESSARY TO REMAIN IN SAFE REGION OF HEAT CAPACITf TEMP UMIT SPIT-b

/

CONSIDERANTICIRATIONOF EMERGENCY DEPRESSURIZA11ON

\\

PERRCIPSECTIONOF

\\

REACTOR VESSEL CONTROL

\\ PROCEDURE{EOP.01. RVCP)

HCTL (MERGENCY DEPRESSURI THE REACTOR PER THE RCIP SECTION OF EOP-91 SPIT-IS BNPVOL-VI OEOP-02-PCCP REVISION NO 10 Categories K/A:

S295026 A2.03 Tier / Group:

T1G1 RO Rating:

3.9 SRO Rating:

4.0 LP Obj:

CLSLP300L*05A Source:

PREV Cog Level:

HIGH Category 8:

Y SPIT-12 INITIATEA REACTOR SCRAM AND ENTER EOP-01 SP/T-09

~

______ -L ______

~

REDUCE REACTOR PRESS I PER THE RCIP SECTION OF EOP- 01 AS NECESSARY TO REMAIN IN SAFE REGION OF HEAT CAPACITY TEMP UMIT SPIT-10

/~~~~'~~'~~1iiii£"'~~~~~"\\

I CONSIDER ANTICIPATION OF \\

( EMERGENCY DEPRESSURlZAll0N \\

\\

PER RCIP SECTION OF

/

\\

~REACTOR VESSEL CONTROL/

\\~~~,:~~~~~:,,~~~~

.1.

SPIT-11 HCTL

~/".r(;AN~,

~//

THE HEAT

</"

CAPACITY TEMP UMIT

'-.,.-!Ii~L.".,

-.,...,. BE MAINTAINED IN THE /"-~

-.,...,., SAFEREGION /""'"

SPIT-12 NO

---'::e~

f

';EN;-~EP;;~~IZE THE REACTOR PER THE RCIP SECTION OF EOP.01

~~~~~~~~.

SPIT-13 BNP VOL-VI OEOP PCCP Categories KIA:

RORating:

LP Obj:

Cog Level:

REVISION NO: 10 S295026 A2.03 3.9 CLS-LP-300-L *05A HIGH Tier / Group: TIG!

SRORating: 4.0 Source:

PREY Category 8:

Y

89. Unit Two is operating at rated power when half of the Drywell (DW) Coolers are lost.

Which one of the following correctly completes the statements below?

(Assume initial DW and Suppression Pool pressures are equal)

As DW temperature rises, Suppression Pool pressure will rise at (1)

DW pressure.

If DW Air Temperature is not restored to within the LCO limit in (2) hours, the Unit is required to be in Mode 3 within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per TS 3.6.1.4 (Drywell Air Temperature).

A. (1) the same rate as (2) 8 B. (1) the same rate as (2) 12 C (1) a slower rate than (2) 8 D. (1) a slower rate than (2) 12

89. Unit Two is operating at rated power when half of the Orywell (OW) Coolers are lost.

Which one of the following correctly completes the statements below?

(Assume initial OW and Suppression Pool pressures are equal)

As OW temperature rises, Suppression Pool pressure will rise at (1)

OW pressure.

If OW Air Temperature is not restored to within the LCO limit in (2) hours, the Unit is required to be in Mode 3 within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per TS 3.6.1.4 (Orywell Air Temperature).

A. (1 ) the same rate as (2) 8 B. (1 ) the same rate as (2) 12 Cy (1) a slower rate than (2) 8 O. (1 ) a slower rate than (2) 12

Feedback K/A: S295028 A2.05 Ability to determine andlor interpret the following as they apply to HIGH DRYWELL TEMPERATURE:

Torus/suppression chamber pressure: Plant-Specific (CFR: 41.10 /43.5 /45.13)

RO/SRO Rating: 3.6/3.8 Objective: CLSLP004A*1 5a

15. Given plant conditions, determine the effects that the following will have on the Primary Containment, Primary Containment Ventilation and Primary Containment Monitoring:
a. Loss of Drywell cooling.

Reference:

SD-04, Revision 5, Page 25 TS Cog Level: High Explanation:

Reduced DW cooling or rising DW temperature results in DW pressure increases whose severity is dependent upon plant conditions. OAOP-14.0, Abnormal Primary Containment Conditions provides guidance on indications to be monitored and actions to be taken which include verification of cooling system lineups and reductions in power to maintain average temperature below 150°F. Failure to accomplish this may require entry into the OEOP-02-PCCP Primary Containment Control.

Elevated DW temperature causes DW pressure to rise. As DW pressure rises, SP water level rises causing a rise in SP pressure. Due to the downcomers extending 3 feet below the surface of the SP water level a differential pressure will always exist. Temperature response is different from LOCA response due to steam AND non-condensibles being forced into the SP

- steam condensing and non-condensibles collecting in SP air space.

TS 3.6.1.4 (DW Air Temperature) limit of < 150°F, CONDITION A - Drywell average air temperature not within limit, REQUIRED ACTION A.1 Restore drywell average air temperature to within limit has a COMPLETION TIME of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

If temperature is not restored to.

150°F, CONDITION B, REQUIRED ACTION 8.1 Be in MODE 3 has a COMPLETION TIME of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Distractor Analysis:

Choice A: Plausible because SP pressure is changed by the change in SP level only vs pressure, steam, and non-condensibles during a LOCA. The SP air space temperature is in equilibrium with SP water temperature (95°Fmax during normal operations) rising DW pressure would have a direct impact on SP level. However during temperature only (no steam), the DW pressure increase is cushioned by SP water, small changes in SP water level provides small change in SP pressure.

8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to restore temperature is correct.

Choice B: Plausible because SP pressure is changed by the change in SP level only vs pressure, steam, and non-condensibles during a LOCA. The SP air space temperature is in equilibrium with SP water temperature (95°Fmax during normal operations) rising DW pressure would have a direct impact on SP level. However during temperature only (no steam), the DW pressure increase is cushioned by SP water, small changes in SP water level provides small change in SP pressure.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is the time required to get to MODE 3 if not restored within the required Completion Time.

Choice C: Correct Answer Choice D: Plausible because rising at a slower rate is correct and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is the time required to get to MODE 3 if not restored within the required Completion Time.

SRO Only Basis: Application of required actions (Section 3) and surveillance requirements (Section 4) in accordance with rules of application requirements (Section 1). (43(b)(2)

Feedback KIA: S295028 A2.05 Ability to determine and/or interpret the following as they apply to HIGH ORYWELL TEMPERATURE:

Torus/suppression chamber pressure: Plant-Specific (CFR: 41.10/43.5/45.13)

RO/SRO Rating: 3.6/3.8 Objective: CLS-LP-004-A *15a

15. Given plant conditions, determine the effects that the following will have on the Primary Containment, Primary Containment Ventilation and Primary Containment Monitoring:
a. Loss of Drywell cooling.

Reference:

SD-04, Revision 5, Page 25 TS Cog Level: High Explanation:

Reduced OW cooling or rising DW temperature results in DW pressure increases whose severity is dependent upon plant conditions. OAOP-14.0, Abnormal Primary Containment Conditions provides guidance on indications to be monitored and actions to be taken which include verification of cooling system lineups and reductions in power to maintain average temperature below 150°F. Failure to accomplish this may require entry into the OEOP-02-PCCP Primary Containment Control.

Elevated DW temperature causes DW pressure to rise. As DW pressure rises, SP water level rises causing a rise in SP pressure. Due to the downcomers extending 3 feet below the surface of the SP water level a differential pressure will always exist. Temperature response is different from LOCA response due to steam AND non-condensibles being forced into the SP - steam condensing and non-condensibles collecting in SP air space.

TS 3.6.1.4 (DW Air Temperature) limit of ~ 150°F, CONDITION A - Drywell average air temperature not within limit, REQUIRED ACTION A.1 Restore drywell average air temperature to within limit has a COMPLETION TIME of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. If temperature is not restored to ~

150°F, CONDITION B, REQUIRED ACTION 8.1 Be in MODE 3 has a COMPLETION TIME of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Distractor Analysis:

Choice A: Plausible because SP pressure is changed by the change in SP level only vs pressure, steam, and non-condensibles during a LOCA. The SP air space temperature is in equilibrium with SP water temperature (95°Fmax during normal operations) rising DW pressure would have a direct impact on SP level. However during temperature only (no steam), the DW pressure increase is cushioned by SP water, small changes in SP water level provides small change in SP pressure.

8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to restore temperature is correct.

Choice B: Plausible because SP pressure is changed by the change in SP level only vs pressure, steam, and non-condensibles during a LOCA. The SP air space temperature is in equilibrium with SP water temperature (95°Fmax during normal operations) rising DW pressure would have a direct impact on SP level. However during temperature only (no steam), the DW pressure increase is cushioned by SP water, small changes in SP water level provides small change in SP pressure.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is the time required to get to MODE 3 if not restored within the required Completion Time.

Choice C: Correct Answer Choice D: Plausible because rising at a slower rate is correct and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is the time required to get to MODE 3 if not restored within the required Completion Time.

SRO Only Basis: Application of required actions (Section 3) and surveillance requirements (Section 4) in accordance with rules of application requirements (Section 1). (43(b)(2)

Notes 3.6 CONTAINMENT SYSTEMS 3.6.1.4 Drvwetl Air Temperature Drvwell Air Temperature 3.6.1.4 LCO 3.6.1.4 Drywell average air temperature shall be 150°F.

APPLICABIUTY:

MODES 1,2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Drywell average air A. 1 Restore drywell average air 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> temperature not within limit, temperature to within limit.

B.

Required Action and B.l Be in MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.

B.2 Be in MODE 4.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> 4.

Drywell Temperature A loss of RBCCW to the drywell results in drywell temperature and pressure increases whose severity is dependent upon plant conditions. OAOP-14.0, Abnormal Primary Containment Conditions provides guidance on indications to be monitored and actions to be taken which include verification of cooling system lineups and reductions in power to maintain average temperature below 150°F.

Failure to accomplish this may require entry into the OEOP-02-PCCP Primary Containment Control.

Loss of RBCCW to the Drywell due to all RBCCW pumps tripping SD-21 Rev. 5 Page 25 of 42 Categories KJA:

RO Rating:

LP Obj:

Cog Level:

S295028 A2.05 CLSLPOO4A* I 5A HIGH Tier / Group:

SRO Rating:

Source:

Category 8:

T1G1 3.8 NEW YF 3.6 Notes Drywell Air Temperature 3.6.1.4 3.6 CONTAINMENT SYSTEMS 3.6.'1.4 Drywell Air Temperature LCO 3.6. '1.4 Drywell average air temperature shall be s '150°F.

APPLICABILITY:

MODES '1,2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A

DrY'Nell average air A.-I Restore drywell average air 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> temperature not within limit.

temperature to within limit B.

Required Action and B.-I Be in MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met AND 1 SO-21 Categories KIA:

RORating:

LP Obj:

Cog Level:

B.2 Be in MODE 4.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />

4.

Orywell Temperature A loss of RBCCW to the drywell results in drywell temperature and pressure increases whose severity is dependent upon plant conditions. OAOP-14.0, Abnormal Primary Containment Conditions provides guidance on indications to be monitored and actions to be taken which include verification of cooling system lineups and reductions in power to maintain average temperature below '150°F.

Failure to accomplish this may require entry into the OEOP-02-PCCP Primary Containment Control.

Loss of RBCCW to the Orywell due to all RBCCW pumps tripping S295028 A2.05 3.6 CLS-LP-004-A*15A mGH Rev. 5 Tier / Group: T1 G 1 SRO Rating:

3.8 Source

NEW Category 8:

YF Page 25 of 421

90. The following plant conditions exist on Unit Two:

- An ATWS with a spurious Group I Isolation has occurred

- HPCI is injecting to the RPV to maintain RPV level

- SUPPRESSION CHAMBER LVL HI-HI is in alarm Which one of the following identifies the action required for long term HPCI system operation and the reason for this action?

When suppression pool temperature reaches 140°F, (1) to prevent (2)

A. (1) lower HPCI flow to less than 2000 gpm lAW LPC (2) pump bearing damage B. (1) lower HPCI flow to less than 2000 gpm lAW LPC (2) a loss of NPSH C (1) defeat the automatic suction transfer logic and transfer HPCI suction to the CST lAW SEP-10 (2) pump bearing damage D. (1) defeat the automatic suction transfer logic and transfer HPCI suction to the CST lAW SEP-10 (2) a loss of NPSH

90. The following plant conditions exist on Unit Two:

- An ATWS with a spurious Group I Isolation has occurred

- HPCI is injecting to the RPV to maintain RPV level

- SUPPRESSION CHAMBER L VL HI-HI is in alarm Which one of the following identifies the action required for long term HPCI system operation and the reason for this action?

When suppression pool temperature reaches 140°F, (1) to prevent (2)

A. (1) lower HPCI flow to less than 2000 gpm lAW LPC (2) pump bearing damage B. (1) lower HPCI flow to less than 2000 gpm lAW LPC (2) a loss of NPSH C~ (1) defeat the automatic suction transfer logic and transfer HPCI suction to the CST lAW SEP-10 (2) pump bearing damage D. (1) defeat the automatic suction transfer logic and transfer HPCI suction to the CST lAW SEP-10 (2) a loss of NPSH

Feedback K/A: 295029 G2.01.07 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

High Suppression Pool Water Level (CFR: 41.5/43.5/45.12/45.13)

ROISRO Rating: 4.4/4.7 Objective:

LOl-CLS-LP-0l 9-A, 26g: Given plant conditions and one of the following events, use plant procedures to determine the actions required to control and/or mitigate the consequences of the event:

High Suppression Pool water level.

Reference:

001-37.5 SUPPRESSION CHAMBER LVL HI-HI APP Cog Level

- High Explanation: HPCI system is normally aligned to the CST, with the torus high water level this transfers to the torus. this meets the KA by having to evaluate the suction path has transferred to the torus and the operational implications of the high torus tempeature on continued operation of the HPCI system. this requires the suction to be transferred back to the CST lAW SEP-I 0.

From : The lube oil and control oil for both HPCI and RCIC are cooled by the water being pumped. Very high lube oil temperatures can result in loss of lubricating qualities in the oil and thus cause damage to the bearings. Suction for HPCI and RCIC is aligned to the Condensate Storage Tank (CST) if it is available. The HPCI automatic suction transfer logic can be defeated to allow this lineup if necessary provided suppression pool temperature is approaching 140°F.

Distractor Analysis:

Choice A: Plausible because reducing flow would be a correct action if HPCI NPSH was the concern. At high temperatures with low level in the torus this could be a correct action. Pump bearing damage is a correct statement.

Choice B: Plausible because reducing flow would be a correct action if HPCI NPSH was the concern. At high temperatures with low level in the torus this could be a correct action. a loss of NPSH would be correct for the reason to reduce flow.

Choice C: Correct answer, see explanation Choice D: Plausible because transferring the suction is correct but the concern is for pump bearing damage.

SRO Basis: 10 CFR 55.43(b)-S Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Feedback KIA: 295029 G2.01.07 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

High Suppression Pool Water Level (CFR: 41.5/43.5/45.12/45.13)

RO/SRO Rating: 4.4/4.7 Objective:

LOI-CLS-LP-019-A, 26g: Given plant conditions and one of the following events, use plant procedures to determine the actions required to control and/or mitigate the consequences of the event:

High Suppression Pool water level.

Reference:

001-37.5 SUPPRESSION CHAMBER LVL HI-HI APP Cog Level - High Explanation: HPCI system is normally aligned to the CST, with the torus high water level this transfers to the torus. this meets the KA by having to evaluate the suction path has transferred to the torus and the operational implications of the high torus tempeature on continued operation of the HPCI system. this requires the suction to be transferred back to the CST lAW SEP-10.

From: The lube oil and control oil for both HPCI and RCIC are cooled by the water being pumped. Very high lube oil temperatures can result in loss of lubricating qualities in the oil and thus cause damage to the bearings. Suction for HPCI and RCIC is aligned to the Condensate Sto(age Tank (CST) if it is available. The HPCI automatic suction transfer logic can be defeated to allow this lineup if necessary provided suppression pool temperature is approaching 140°F.

Distractor Analysis:

Choice A: Plausible because reducing flow would be a correct action if HPCI NPSH was the concern. At high temperatures with low level in the torus this could be a correct action. Pump bearing damage is a correct statement.

Choice B: Plausible because reducing flow would be a correct action if HPCI NPSH was the concern. At high temperatures with low level in the torus this could be a correct action. a loss of NPSH would be correct for the reason to reduce flow.

Choice C: Correct answer, see explanation Choice 0: Plausible because transferring the suction is correct but the concern is for pump bearing damage.

SRO Basis: 10 CFR 55.43(b )-5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Notes MAXIMUM INJECTION

.,VSTEM PRESSURE (P5I)

CONDENSATE1FEEDWATER 1250 CRt FLOW NAY OE 1490 MAXIMIZED PER EOP0i. SEP-00 RCIC WITH SUCTION FROM CST IF AVAILABLE. DEFEAT LOW REACTOR 1190 PRESS AND HIGH AREA TEMPERATURE ISOLATION LOGIC IF NECESSARY PER CIRCUW ALTERATION PROC EDURE c,.oISEP-io HPCI WITH SUCTION FROM 1250 OST IF AVAILAbLE, OFIAT 1IPCI NI SUPPRESSION POOL I.EVEL SUCflO1 TRANSFER AND I-IGt-I AREA TEMPERATURE ISOLATION LOGIC IF NECESSARY PER CIRCUIt ALTERATION PROCEDURE EOP.0I-SEP-10)

IPUI-ESTALI5H RHR SERVICE 200 WATER FLOW AS SOON AS POSSIBLE CAUTION OPERATION OF HPGI OR ROIG WITH SUCTION TEMPERATURES ABOVE 140? MAY RESULT IN EQUIPMENT DAMAGE RC1L-23 Distractor plausibility:

CAUTION I

HPCI FLOW ABOVE 2000 GPM WITH SUCTION FROM CST AND GST LEVEL BELOW 5 FEET MAY RESULT IN VORTEXNG AND EQUIPMENT DAMAGE I

RCFL Categories K/A:

295029 G2.01.07 Tier/Group:

T1G2 RO Rating:

4.4 SRO Rating:

4.7 LP Obj:

19-A 26G Source:

BANK Cog Level:

HIGH Category 8:

TABLE I M1MUM SYSTEM INJECTION PRESSURES Notes 1

MAXIMUM SYSTEM INJECTlD.'I PRESSURES SYSTEM CONDENSATEIFEEDWA.TER CRD FI.OW NAY DIl:

MAXIMIZED PER eOP.Of. SEP. 09 RCiC WITH SUCTION FROM CST IF AVAILABLE. DEFEAT LOW REA.CTOR PRESS AND HIGH AREA TEMPERATURE ISOLATION LOGIC IF NECESSARY PER *CIRCUIT ALTERATION PROCEDURE" leop* Ot* sep* 10)

HPCI WITH SUCTION FROM CST II' AVAILABLe. OeFeAT HP(;I HI SUI"PRI!$$ION POOt. I.I!Yf!1.

SUCTION TAANSFSR ANI) HIGH ARIOA TIOMPERATURS ISOLATION lOGIC If: NeCeSSARY PER *CIRCUIT ALTeRATION PROCEDURe"

~EOP* 01* SEP* 10) lPCI* ESTABLISH RHR SERVICE WATER FI.OW AS 500H AS POSSIBLE CAUTION ERATION OF HPCI OR RCI H SUCTION TEMPERATURE OVE 140*' MAY RESULT IN EQUIPMENT DAMAGE Distractor plausibility:

CAUTION HPCI FLOW ABOVE 2000 GPM WITH SUCTION FROM CST AND CST LEVEL BELOW 5 FEeT MAY RESULT IN VORTEX<<NG AND EQUIPMENT DAMAGE Categories KIA:

RORating:

LPObj:

Cog Level:

295029 G2.01.07 4.4 19-A 26G HIGH MAXIM LIM INJECTION PRESSURE (PSIG) 1250 141)0 1190 12110 200 Tier / Group: T1 G2 SRO Rating:

4.7 Source

BANK Category 8:

91. Which one of the following identifies the controlling document and the required action to be taken if SJAE Offgas Radiation monitor readings increase 50% during steady state rated power operation?

Notify E&RC to perform the Surveillance I Test Requirement (SRITR) required by (1)

, which confirms the SJAE release rate is within limits within (2) following the monitor reading increase.

A. (1) ODCM 7.3.2, Radioactive Gaseous Effluent Monitoring Instrumentation (2) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> B. (1) ODCM 7.3.2, Radioactive Gaseous Effluent Monitoring Instrumentation (2) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> C (1) T.S. 3.7.5, Main Condenser Offgas (2) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> D. (1) T.S. 3.7.5, Main Condenser Offgas (2) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

91. Which one of the following identifies the controlling document and the required action to be taken if SJAE Offgas Radiation monitor readings increase 50% during steady state rated power operation?

Notify E&RC to perform the Surveillance I Test Requirement (SRITR) required by (1)

,which confirms the SJAE release rate is within limits within (2) following the monitor reading increase.

A. (1) ODCM 7.3.2, Radioactive Gaseous Effluent Monitoring Instrumentation (2) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> B. (1) ODCM 7.3.2, Radioactive Gaseous Effluent Monitoring Instrumentation (2) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> C~ (1) T.S. 3.7.5, Main Condenser Offgas (2) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> D. (1) T.S. 3.7.5, Main Condenser Offgas (2) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

Feedback K/A: S295038G 2.02.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

High Off-Site Release Rate (CFR: 41.7/41.10 / 43.2/43.3/45.3)

RO/SRO Rating: 3.9/4.6 Objective: CLSLP30*08

08. Given plant conditions and Technical Specifications, including the Bases, TRM, ODCM, and COLR, determine whether given plant conditions meet minimum Technical Specifications, TRM, or ODCM requirements associated with the Condenser Air Removal/Augmented Offgas System.

Reference:

101-03.1, Revision 10, Page 44, Item #57 (CODSR)

Cog Level: High Explanation:

NOTIFY E&RC to confirm release rate is within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following a monitor reading increase of greater than or equal to 50% without an accompanying increase in thermal power. SR 3.7.5.1 Distractor Analysis:

Choice A: Plausible because the SJAE Rad Monitor operability is required by ODCM 7.3.2 and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is correct.

Choice B: Plausible because the SJAE Rad Monitor operability is required by 00CM 7.3.2 and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is a timeframe for another Required Action in this spec.

Choice C: Correct Answer Choice D: Plausible because TS 3.5.7 is correct and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is a timeframe for another Required Action in this spec.

SRO Only Basis: Application of Surveillance Requirements and timeframe greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Notes Feedback KIA: S29503BG 2.02.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

High Off-Site Release Rate (CFR: 41.7/41.10/43.2/43.3/45.3)

RO/SRO Rating: 3.9/4.6 Objective: CLS-LP-30*OB OB. Given plant conditions and Technical Specifications, including the Bases, TRM, ODCM, and COLR, determine whether given plant conditions meet minimum Technical Specifications, TRM, or ODCM requirements associated with the Condenser Air Removal/Augmented Offgas System.

Reference:

101-03.1, Revision 10, Page 44, Item #57 (CODSR)

Cog Level: High Explanation:

NOTIFY E&RC to confirm release rate is within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following a monitor reading increase of greater than or equal to 50% without an accompanying increase in thermal power. SR 3.7.5.1 Distractor Analysis:

Choice A: Plausible because the SJAE Rad Monitor operability is required by ODCM 7.3.2 and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is correct.

Choice B: Plausible because the SJAE Rad Monitor operability is required by ODCM 7.3.2 and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is a timeframe for another Required Action in this spec.

Choice C: Correct Answer Choice D: Plausible because TS 3.5.7 is correct and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is a timeframe for another Required Action in this spec.

SRO Only Basis: Application of Surveillance Requirements and timeframe greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Notes

ATTACHMENT 1 Page 39 of 67 ITEM SI-ilF CHECK LIST NOTES OPER FREQ TIME TSOPER NO.

MODE LIMITS RECORD SJAE OFF(4S RAD MONiTOR DD

1. 2. 3 b

07-13 012-RM-K60tA. NOTIFY E&RC to confirm release rate is within lmits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> oIIowing a monitor reading increase of greater than or equal to 50% without an l3-1 accompanying increase in thermal pcwer.

SR 3.7.5.1 RECORD S4E OFFGAS R40 MON.TOR CD

1. 2. 3 07-12 D12-RM-K6018. NOTIFY E&RC to confirm reLease rate is within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following a nionitor reading increase of greater than or equal to 50% without an accompanying increase in thermal pcf.ver.

SR 3.7.5.1 PERFORM channel check utilizing the

07-13 Reference calculabon on Table I SJAE OFF-GAS RAE) calcufaton on MONITORS 012-RM-keoIA and B 00CM Table I TR 7.3.2-1 FunctionS, TR 7.3.2.1 PERFORM channel check on SERViCE R

c 07-13 channel WATER EFFLLJENTRAD MONITOR operabe 012-RM-K605. 00CM Tabie 7,3. 1-i Function 3, TR 7.3.1.1 i

PERFORM channel check on RAE)WASTE 8

c 07-13 channel EFFLUENT RAD MONITOR D12-RM-K6C4 operabe on Control Room Panel 2-H12-P804 with recorder D12-ROO1 on XU-2. CCCM Tab!e 7.3.1-i, Item 1. TR7.3.l.I During operation of the main condenser air ejector.

SHIFT Dayshift BRUNSWICK STEAM ELECTRIC PL4NT DAILY SURVE1LLANCE REPORT CONTROL OPERATORS 101-03.1 Rev. 101 ITEM SHIFT CHECK LIST NOlES NO.

57 RECORD SJAE OFFGAS RADMONITOR DD D12-RM-K6Cl"IA. NOTIFY E&RC to confilm rel~ase rate is within limits,... ijhin 4 hrurs follO\\lotng a monitor reading increase o*f greater than or equal to 50% without an accompanying increase in thermal pow~r.

SR3.7.5.1 58 RECORD SJAE OFFGAS RAD MONITOR DD D12-RM-K601B. NOTIFY E&RC to confirm re!~ase rate is within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> follo',\\;ng a monitor reading increase of greater than or equal to 50% without an accompanying increase in thermal pow~r.

SR 3.7.5.1 59 PERFORM ohannel check utilizing the calculation on Table SJAE OFF-GAS RAD MONITORS D'12-RM-K601A and B ODCM TR 7.3.2-1 Function 6, rn 7.3.2.1 00 PERFORM channel cheok on SERVICE R

~\\-:41ER EFFLUENT RAD MONITOR D12-RM-K605, ODCM Table 7.3.1-1, Function 3, TR 7.3.1.1 61 PERFORM ohannel check on P.ADI*lIASTE EFFLUENT RAD MOM TOR D12-RM-K604 en Control Room Pan~12-H12-Pa04 'Qi1h recorder D'12-ROOt on XU-3, o!)CM Table 7.3.1-1, !t~m I, TR 7.3.1.1

'During operation of the main oondenser air ejector.

SHIFT Davshift 1'101-03.1 OPER MODE 1,2' 3' 1,2',3'

'6 6

6 ATTACHMENT 'I Page 39 of6?

FREQ TIME':

b 07-13 13-19 b

07-13 13-19 0

07-13 e

07-13 e

07-13 TSiOPER LIMITS Refer~nce calculation on Table 1 channel operable channel operable BRUNSWICK STEAM ELECTRIC PL4.NT DAilY SURVEIllANCE REPORT CONTROL OPERATORS Rev. 101

Main Condenser Offgas 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Main Condenser Offgas LCO 3.7.5 The gross gamma activity rate of the noble gases measured at the main condenser air ejector shall be 243,600 pCiisecond after decay of 30 minutes.

APPLICABILITY:

MODE 1, MODES 2 and 3 with any main steam line not isolated and steam jet air ejector (SJAE) in operation.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Gross gamma activity rate of A.1 Restore gross gamma 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> the noble gases not within activity rate of the nobie

limit, gases to within limit.

B.

Required Action and B.1 Isolate all main steam lines. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> associated Completion Time no met.

B.2 Isolate SJAE.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> OR B.3.1 Be in MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND B.3.2 Be in MODE 4.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Brunswick Unit 1 3.7-18 Amendment No. 203 3.7 PLANT SYSTEMS 3.7.5 Main Condenser Offgas Main Condenser Offgas 3.7.5 LCO 3.7.5 The gross gamma activity mte of the noble gases measured at the main condenser air ejector shall be ::; 243,600 jJCilsecond after decay of 30 minutes.

APPLICABILITY:

MODEl, MODES 2 and 3 with any main steam line not isolated and steam jet air ejector (SJAE) in operation.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Gross gamma activity rate of.1l.* 1 Restore gross gamma 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> the noble gases not within activity mte of the noble limit.

gases to within limit B.

Required Action and B.1 Isolate all main steam lines. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.

OR B.2 Isolate SJAE.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> OR B.3.1 Be In MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND B.3.2 Be in MODE4.

313 hours0.00362 days <br />0.0869 hours <br />5.175265e-4 weeks <br />1.190965e-4 months <br /> Brunswick Unit 1 3.7-18 Amendment No. 203

Main Condenser Offgas 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1,

NOTE

Not required to be performed until 31 days after any main steam line not isolated and SJAE in operation.

Verify the gross gamma activity rate of the noble 31 days gases is 243600 pCiisecond after decay of 30 minutes.

AND Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> afteraa5O%

increase in the nominal steady state fission gas release after factoring out increases due to changes in THERMAL POWER level 8runswick Unit 1 3.7-19 Amendment No. 203 SUR'.,.'EILLANCE REQUIREMENTS SURVEILLANCE Main Condenser Offgas 3.7.5 FREQUENCY SR 3.7.5.1

NOTE----------------------------

Brunswick Unit 1 Not required to be performed until 31 days after any main steam line not isolated and SJAE in operation.

Verify the gross gamma activity rate of the noble gases is ::; 243,600 ~Cilseoond after decay of 30 minutes.

3.7-19 31 days Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after a :::50%

increase in the nominal steady slate fission gas release after factoring oul increases due to changes in THERMAL POWER level Amendment 1\\10. 2113

Radioactive Gaseous Effluent Monitoring Instrumentation 7.3.2 Table 7.12-aage 2 of 4)

Radactie I3asecus Effluent Mitang reburrentatcn FJNC.

PJCA6_E aL:iEt ccrilyr1cr4 TE&T ftp tLC0S 3ft C%NLS 9EFEfNCO aECuIRE,lETS 3TPCNT OThER ZER ft REJIRD S,CIFIED FJt2TDt CCt.FEriSArOv CCt4DrIC?3 tE4.SLL9E3 Al 2.

Rea,tsor Buildng Ventilatcn Ffonhtorir System (continueo}

e.

£airplerFlcwRaie At at tme; 1

0 TR 7.3.2.1 srernent Device TR 7.12.6 TR 7.3.2. C 3.

Turbine Buildng Venblatcn Monito,ir System a.

NobieOasMtWiy Atalmes 1

3 TR 7.a2.

(b Atnikr TR 7.3.2.3 TR 7.3.15 TR 7.3.2.10 b.

bineSanpIer Ata[tmes 1

C TR 7.3.12 NA Cartroge o.

Parclate Sampler Ata[ bmes 1

C TR 7.3.2.2 NA Ftter d.

System Effluent Fbiw At a tines 1

0 TR 7.3.2i NA Rate Measurement TR 7.3.2.

Device TR 7.3.2.10 e.

Lcv Rar,e Lairpler At al tines 1

0 TR 7.3.21 (Cl F Rate TR 7.3.2.5 easarernent Desice TR 7.3.2.10 f.

Md:High Rante (rn:

1 0

TR 7.3.2.10 NA Sampler F1cv Rate Measurenrent 0eioe 4.

ManCondenserOtf-Gas (ei 1

B TR 7.32.1 Treatment System Ncble TR 7.3.2.3 Gsa Acbiity Mcnhtor TR 7.12.6 (0onseam of AOG R 7.3.2.10 Treatment System)

(continued)

(a)

Specif a inatn.anentatii dentlicaban nuntere are provided in Appendix 3.

b)

Alarnscrp setpcints ahat be detemiine in accocdaie teth 00CM methodacgy and set ia ensure the limta of 000MS 7.3.7, Dose RateGaseous Effluents, are nt eteeeded.

Ic)

Alarmr seoints shall be deterntined in aconntiartce eith asscctsed desi9n specitcatiDn(5) and setto ensure the ltnit.s at COCMS 7.3.7, Dose RateGaseous Effluents, are na exceeded.

d)

Provides alarm.

ie)

Doring Main Condenser 0ff-Gas Tresurent System operatcn irn)

During MWHig RareSymaacn Brunswidc Units 1 and 2 7.3.2-18 Rev. 32 I Categories K/A:

S295038G2.02.42 Tier/Group:

T1G1 RO Rating:

3.9 SRO Rating:

4.6 LP Obj:

CLSLP3O*O8 Source:

NEW Cog Level:

HIGH Category 8:

Y Radioactive Gaseous Effluent Monitoring Instrumentatlon T<ible 7.:?.2-1 ipage 2 off 4)

Radioacti ** -e C-,aSECUS Effluent ~,'itocn!! 1nslrurr,entat'crI FUNCTlml '"

,.P?LlCA8:'E REaLiI.~EO CCNDlTICNS TEST fb:::O:SOR CH>.NNELS REfERIENCEO REQUIREMENTS OThER

?ER fROM REQUIRED S"=:CIFIEO F;Jr..;:::TiO:-I CCI/PEN!l"'TO.RY CCN:)liIC/,"S MEASU~ES"

  • . l
2.

Re:actor Build\\!1g Ventilation Monitoring Systsm (rontinuec}

e.

Sarrpler Row Rate At all fmas D

TR 7.3.2.1 Measurerr,ent Device; TR 7.3.2.6 TR 7.3.2.10

3.

TLiltine Buildillg Venlilaton Monitorir" Systam

3.

Notle Gas Acti'.i.ly

.<\\t all tim;;s TR 7.32.;

M::fIitcr TR 7.3.2.3 TR 7.3.2.5 TR 7.3.2.10

b.

looine Sarrpler At 'II! limss C

TR 7.3.22 Cmr:dge

c.

PartiCIIlate S<1mj:ler AtaN lim;;s C

TR 7.3.2.2 Fiiter

d.

System Effiuem Flem At all lim;;s 0

TR 7.3.2.*

Rate Measurement TR 7.3.2.0 De-~ice TR 7.3.2.10

e.

low Range Sarrpler At aU lim;;s 0

TR 7.3.2.1 FbwRate TR 7.3.2.0 Measurement Device; TR 7.3.2.10

f.

MeiHigh Rang;;

(m) 0 TR 7.. 3.2.10 Sarrpler ReI'.' Rate ME-<1SlJrement Device;

4.

Main Condenser Off-Gas (e)

B TR 1.3.2.1 Treatment Systam Noble TR 7.3..2.3 Gas Activity Monitor '"

TR 7.3..2.6 (Dovmstream of AOG TR 7.3.2.10 T reafment Systam)

(a)

Speci/a instrumentatioo idE!1~licalion numbers <ire provided in AppEIldix E.

(b)

AlarJl1l~rip setpoinfs shall be determined in acoocdance \\";th ODCM melhodo:cgy and set 10 ensure the lim(s oi ODCMS 7.3.7. 'Dose Rate-Gaseous Effluents; are no! elQOeeaed.

7.3.2 Al.ARWiRiP S:=TPO!NT V,.LUE (0;1 (b)

NA NA NA (e)

NA (bJ (CQntinued)

(c}

AJarll1/tr,p setpoints shall be detemlined in oY..o::roance \\\\;th assccia:;ed desig\\!1specifoation{s) and set to anS!!re the limits of COCMS 7.:'..7. 'Dose Rate---Gaseous Effluents; are not exceec'ed.

(d)

Provides alaml.

(e)

During Main Coodenser Oft-Gas Treatment System operation (m)

During MdiHigll Rang;; System opeaIion Brunswick Units 1 and 2 Categories KIA:

RORating:

LPObj:

Cog Level:

S295038G 2.02.42 3.9 CLS-LP-30*08 HIGH 7.3.2-10 Rev. 32 1 Tier / Group: TIGl SRORating:

4.6 Source

NEW Category 8:

Y

92. The following plant conditions exist on Unit Two due to a malfunction of the Air Dryer:

- SERVICE AIR PRESS-LOW is in alarm

- RB INSTR AIR RECEIVER 2A PRESS LOWis in alarm

- RB INSTR AIR RECEIVER 2B PRESS LOW is in alarm

- Instrument Air pressure is 93 psig and recovering Based on the above indications, which one of the following correctly identifies:

(1) the status of the Service Air Dryer Bypass Valve, SA-PV-5067, and (2) the procedure that contains the steps to close the Reactor Building Inboard and Outboard Isolation Valves (BFIVs)?

A (1) open (2) OAOP-20.O, Pneumatic (Air/Nitrogen) System Failures B. (1) open (2) 2APP-UA-O1, Service Air Press-Low C. (1) closed (2) OAOP-20.O, Pneumatic (Air/Nitrogen) System Failures D. (1) closed (2) 2APP-UA-O1, Service Air Press-Low

92. The following plant conditions exist on Unit Two due to a malfunction of the Air Dryer:

- SERVICE AIR PRESS-LOW is in alarm

- RB INSTR AIR RECEIVER 2A PRESS LOW is in alarm

- RB INSTR AIR RECEIVER 2B PRESS LOW is in alarm

- Instrument Air pressure is 93 psig and recovering Based on the above indications, which one of the following correctly identifies:

(1) the status of the Service Air Dryer Bypass Valve, SA-PV-5067, and (2) the procedure that contains the steps to close the Reactor Building Inboard and Outboard Isolation Valves (BFIVs)?

A'I (1) open (2) OAOP-20.0, Pneumatic (Air/Nitrogen) System Failures B. (1) open (2) 2APP-UA-01, Service Air Press-Low C. (1) closed (2) OAOP-20.0, Pneumatic (Air/Nitrogen) System Failures D. (1) closed (2) 2APP-UA-01, Service Air Press-Low

Feedback K/A: 300000 A2.01 Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

Air dryer and filter malfunctions (CFR: 41.5/45.6)

RO/SRO Rating: 2.9/2.8 Objective:

CLS-LP-46, 07i: Given plant conditions, determine if the following automatic actions should occur: Air Dryer is bypassed.

CLS-LP-037.1, 8b: State how the RBHVAC is affected by the following: Loss of Instrument Air.

Reference:

RB INSTR AIR RECEIVER 28 PRESS LOW (UA-01 1-2)

SERVICE AIR PRESS LOW (UA-01 5-4)

OAOP-20, Pneumatic (Air/Nitrogen) System Failures Cog Level: High Explanation:

The air dryer malfunction has caused air pressure to lower. The Service Air low pressure alarms comes in at 107 psig. At 105# decreasing the Service Air system isolates, thus the 0 psig indication on Service Air.

The alarms for the receivers low pressure come in at 95# and are located in the Reactor Building. With these alarms in the operators are required to close the BFIVs while there is still sufficient air pressure remaining to make the secondary containment isolation valves close in accordance with the AOP supplemental actions.

Distractor Analysis:

Choice A: Correct answer; The air dryer bypass valve opens at 98# and dropping and the steps are in the AOP for closing the BFIVs.

Choice B: Plausible because the air dryer bypass valve is open, but the guidance for closure of the BFIVs is contained in the AOP or RB INSTRAIR RECEIVER 2A(B) PRESS LOWAPP.

Choice C: Plausible because the AOP is the correct procedure for closure of the BFIVs, but the air dryer bypass valve would be open (requires system knowledge to know the setpoint for the bypass opening).

Choice D: Plausible because the student may not know the setpoint of the bypass valve opening and the guidance for closure of the BFIVs is contained in the AOP or RB INSTR AIR RECEIVER 2A(B) PRESS LOWAPP.

SRO Basis: 10 CFR 55.43(b)-5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

The first part of the question is RO knowledge (setpoint for the auto opening of the air dryer bypass valve the second part is Assessing plant conditions (normal, abnormal, or emergency) and then prescribing a procedure to mitigate, recover, or with which to proceed.

Notes Feedback KiA: 300000 A2.01 Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

Air dryer and filter malfunctions (CFR: 41.5 /45.6)

RO/SRO Rating: 2.9/2.8 Objective:

CLS-LP-46, 07i: Given plant conditions, determine if the following automatic actions should occur: Air Dryer is bypassed.

CLS-LP-037.1, 8b: State how the RBHVAC is affected by the following: Loss of Instrument Air.

Reference:

RB INSTR AIR RECEIVER 2B PRESS LOW (UA-01 1-2)

SERVICE AIR PRESS LOW (UA-01 5-4)

OAOP-20, Pneumatic (Air/Nitrogen) System Failures Cog Level: High Explanation:

The air dryer malfunction has caused air pressure to lower. The Service Air low pressure alarms comes in at 107 psig. At 105# decreasing the Service Air system isolates, thus the 0 psig indication on Service Air.

The alarms for the receivers low pressure come in at 95# and are located in the Reactor Building. With these alarms in the operators are required to close the BFIVs while there is still sufficient air pressure remaining to make the secondary containment isolation valves close in accordance with the AOP supplemental actions.

Distractor Analysis:

Choice A: Correct answer; The air dryer bypass valve opens at 98# and dropping and the steps are in the AOP for closing the BFIVs.

Choice B: Plausible because the air dryer bypass valve is open, but the guidance for closure of the BFIVs is contained in the AOP or RB INSTR AIR RECEIVER 2A(B) PRESS LOW APP.

Choice C: Plausible because the AOP is the correct procedure for closure of the BFIVs, but the air dryer bypass valve would be open (requires system knowledge to know the setpoint for the bypass opening).

Choice D: Plausible because the student may not know the setpoint of the bypass valve opening and the guidance for closure of the BFIVs is contained in the AOP or RB INSTR AIR RECEIVER 2A(B) PRESS LOWAPP.

SRO Basis: 10 CFR 55.43(b)-5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

The first part of the question is RO knowledge (setpoint for the auto opening of the air dryer bypass valve the second part is Assessing plant conditions (normal, abnormal, or emergency) and then prescribing a procedure to mitigate, recover, or with which to proceed.

Notes

4.

IF RB INSTR AiR RECEIVER IA(2A) PRESS LOW (UA-O1 1-1) OR RB IWSTR AIR RECEiVER IB(2B,)

PRESS LOW (UA-Ol 1-2) alarm is received, THEN PERFORM the following:

NOTE:

Isolation of the Reactor Building supply and exhaust dampers will render the building ventilation system inoperable. Consideration should be given for starting the Standby Gas Treatment System to ensure the Reactor Building differential pressure remains negative.

a.

IF necessary. THEN START the Standby Gas 0

Treatment System.

NOTE:

Local Tee Handles may be used to close the Reactor Building Isolation Dampers if insufficient control air is available. 1 (2)OP-37.1 provide L

instructions for manual operation of Reactor Building Isolation Valves.

b.

CLOSE the following dampers:

RB VENT 1NBO VALVES, 1A2A)-BFiV-RB and iCi2C)-BF1V-RB RB VENT OUTBD VALVES, 1B(2B)-BF1V-RB and ID(2D)-BF1 V-RB

) OAOP-20.O Rev. 35 Page 5 of 18 Unit 2 APP UA-O1 5-3 Page 1 of 2 AIR DRYER 2A TROUBLE AUTO ACTIONS 1.

Service air dryer bypass valve SA-PV-5067 will begin to open if service air header pressure decreases to 98 psig.

2.

If control power is lost or interrupted the dryer will fail safe, providing continued air flow through one tower.

3.

If a dryer tower moisture sensing probe related fault or malfunction occurs, the dryer control system will default to a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> drying cycle.

4.

IF RB INSTR AIR RECEIVER 1A(2A) PRESS LOW (UA-01 1-1) OR RB INSTR AIR RECElVER 1B(2B)

PRESS LOW (UA-01 1-2) alarm is received, THEN PERFORM the following:

NOTE:

Isolation or the Reactor Building supply and exhaust dampers will render the building ventilation system inoperable. Consideration should be given ror starting the Standby Gas Treatment System to ensure me Reactor Building differential pressure remains negative.

a.

IF necessary, THEN START the Standby Gas Treatment System.

o NOTE:

Local "Tee Handles" may be used to close the Reactor Building Isolation Dampers if insufficient control air is available. 1 (2)OP-37.1 provide instructions for manual operation of Reactor Building Isolation Valves.

b.

CLOSE the following dampers:

IOAOP-20.0 AIR DRYER.2A TROUBLE AUTO ACTIONS

,~B VENTJNBD VALVES, 0

1A(2A)-BFlV-RB and tC(2C)-BFIV-RB

.~B VENTOUTBD VALVES, 0

1B(2B)-BFlV-RB and 1D(2D)-BFIV-RB Rev. 35 Page50f18 I Unit 2 APP UA-01 5-3 Page*lof2

  • 1.

Service air dryer bypass valve SA-PV-5067 will begin to open if service air header pressure decreases to 98 psig.

2.

If control power is lost or interrupted the dryer will fail safe, providing continued air flow through one tower.

3.

If a dryer tower moisture sensing probe related fault or malfunction occurs, the dryer control system will default to a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> drying cycle.

RB INSTR AIR RECEIVER 2B PRESS LOW AUTO ACTIONS 1.

Standby Instrument Air Compressor 2B starts and loads.

2.

High Pressure Bottle Rack Isolation Valve, RNA-SV-5481 opens, supplying SRVs and AC-V17 with a pneumatic source.

CAUSE 1.

Low air pressure (95 psig) in instrument air receiver 2B.

2.

Loss of plant air compressors.

3.

Instrument air pipe rupture or air leak.

4.

Circuit malfunction.

OBSERVATIONS 1.

Standby compressor starts automatically and loads (it will unload at 105 psig).

2.

Service air header may have isolated.

3.

Pressure Indicator 2-RNA-Pl-5268 (XU-51) indicates approximately 100 psig.

ACTIONS 1.

Check that standby compressor is running.

2.

Check to see if instrument air pressure is maintaining or increasing above 95 psig.

3.

Check plant compressors.

4.

Check for instrument air ruptures.

5.

Isolate any instrument air piping leaks or ruptures.

6.

Isolate nonessential air supplies in order to maintain more than 95 psig on instrument heade[

7.

Ensure the High Pressure Bottle Rack Isolation Valve, RNA-SV-5481 (XU-51) opens.

8.

If a circuit malfunction is suspected, ensure that a WRIJO is prepared.

9.

If secondary containment isolation is required, close secondary containment isolation valves 2B-BFIV-RB and 2D-BFIV-RB prior to accumulator air pressure bleedoff.

2APP-UA-0i Rev. 64 Page 6 of 102 Categories K/A:

300000 A2.01 Tier/Group:

T2G1 RO Rating:

2.9 SRO Rating:

2.8 LP Obj:

46-71 Source:

NEW Cog Level:

HIGH Category 8:

Y RB INSTR AIR RECEIVER 2B PRESS LOW AUTO ACTIONS

'I.

standby Instrument Air Compressor 2B starts and loads.

2.

High Pressure BoUIe Rack Isolation Valve, RNA-SV-548'1 opens, supplying SRV's and CAC-V17 with a pneumatic source.

CAUSE

'I.

Low air pressure (95 psig) in instrument air receiver 2B.

2.

Loss of plant air compressors.

3.

Instrument air pipe rupture or air leak.

4.

Circuit malfunction.

OBSERVATIONS

'1.

Standby compressor starts automatically and loads (it will unload at '105 psig).

2.

Service air header may have isolated.

3.

Pressure Indicator 2-RNA-PI-5268 (XU-51) indicates approximately 100 psig.

ACTIONS

'1.

Check tllat standby compressor is running.

2.

Check to see if instrument air pressure is maintaining or increasing above 95 psig.

3.

Check plant compressors.

4.

Cl1eck for instrument air ruptures.

5.

Isolate any instrument air piping leaks or ruptures.

6.

Isolate nonessential air supplies in order to maintain more than 95 psig on instrument header.

7.

Ensure the High Pressure Bottle Rack Isolation Valve, RNA-SV-548'1 (XU-51) opens.

8.

If a circuit malfunction is suspected, ensure that a WRfJO is prepared.

9.

If secondary containment isolation is required, close secondary containment isolation valves 2B-BFIV-RB and 20-BFIV-RB prior to accumulator air pressure bleedoff.

!2APP-UA-O'l Categories KIA:

300000 A2.0 1 RO Rating:

2.9 LP Obj:

46-71 Cog Level:

HIGH Rev. 64 Tier / Group: T2G 1 SRO Rating:

2.8 Source

~vv Category 8:

Y Page 6 of 1021

93. Unit Two is operating at power with Reactor Recirculation Loop A isolated due to abnormal seal leakage. A fire in the reactor building occurs and the Site Incident Commander has requested that MCC 2XA-2 be de-energized for fire suppression.

Which one of the following identifies the impact that deenergizing MCC 2XA-2 has on RHR Loop A availability and the procedure which provides this guidance under the above plant conditions?

Deenergizing MCC 2XA-2 will make RHR Loop A Inoperable but Available provided that the 2-El l-FO15A, Inboard Injection Vlv, (1) to support LPCI lAW (2)

A.

(1) is maintained (de-energized) opened (2) OAP-025, BNP Integrated Scheduling B. (I) is maintained (de-energized) opened (2) 001-01.08, Control of Equipment and System Status C (1) has a Dedicated Operator is assigned for manual operation (2) OAP-025, BNP Integrated Scheduling D. (1) has a Dedicated Operator is assigned for manual operation (2) 001-01.08, Control of Equipment and System Status

93. Unit Two is operating at power with Reactor Recirculation Loop A isolated due to abnormal seal leakage. A fire in the reactor building occurs and the Site Incident Commander has requested that MCC 2XA-2 be de-energized for fire suppression.

Which one of the following identifies the impact that deenergizing MCC 2XA-2 has on RHR Loop A availability and the procedure which provides this guidance under the above plant conditions?

Deenergizing MCC 2XA-2 will make RHR Loop A Inoperable but Available provided that the 2-E11-F015A, Inboard Injection Vlv, (1) to support LPCIIAW (2)

A. (1) is maintained (de-energized) opened (2) OAP-025, BNP Integrated Scheduling B. (1) is maintained (de-energized) opened (2) 001-01.08, Control of Equipment and System Status C~ (1) has a Dedicated Operator is assigned for manual operation (2) OAP-025, BNP Integrated Scheduling D. (1) has a Dedicated Operator is assigned for manual operation (2) 001-01.08, Control of Equipment and System Status

Feedback K/A: S600000G 2.02.37 Ability to determine operability andlor availability of safety related equipment.

Plant Fire On Site (CFR: 41.7 /43.5 / 45.12)

ROISRO Rating: 3.6/4.6 Objective: CLS-LP

Reference:

OAP-025, Revision 39, Page 9, Section 3.1 Cog Level: High Explanation:

Requires knowledge of equipment powered from MCC 2XA-2 (opposite Unit power E5). With the RR Loop A isolated (RR Discharge and Disch Bypass valves will be close

- required for LPCI) and FO15A (located in ECCS Pipe Tunnel - RB 20) can be manually opened by a dedicated operator. 001-01.08 has recently been revised to support implementation of OPS-NGGC-1 000, Fleet Conduct of Operations. Risk assessment and equipment removal from service guidance has been removed from 01-01.08.

Distractor Analysis:

Choice A: Plausible because if the valve is de-energized in its intended state (such as a PCIV in the closed dierection) this could be considered correct although for this case the open direction for this normally closed valve would have to have interlocks defeated to have both the F015 and F017 both open (potential to pressurize low pressure piping). OAP-025 is correct.

Choice B: Plausible because if the valve is de-energized in its intended state (such as a PCIV in the closed dierection) this could be considered correct although for this case the open direction for this normally closed valve would have to have interlocks defeated to have both the F015 and F017 both open (potential to pressurize low pressure piping). 001-01.08 no longer provides guidance for evaluating MR/PSA system availability.

Choice C: Correct Answer Choice D: Plausible because RHR Loop is available and 001-01.08 no longer provides guidance for evaluating MRIPSA system availability.

SRO Only Basis: Knowledge of administrative procedures that specify implementation, and/or coordination of plant normal procedures.

Notes Feedback KIA: S600000G 2.02.37 Ability to determine operability and/or availability of safety related equipment.

Plant Fire On Site (CFR: 41.7/43.5/45.12)

RO/SRO Rating: 3.6/4.6 Objective: CLS-LP-

Reference:

OAP-025, Revision 39, Page 9, Section 3.1 Cog Level: High Explanation:

Requires knowledge of equipment powered from MCC 2XA-2 (opposite Unit power E5). With the RR Loop A isolated (RR Discharge and Disch Bypass valves will be close - required for LPCI) and F015A (located in ECCS Pipe Tunnel - RB 20') can be manually opened by a dedicated operator. 001-01.08 has recently been revised to support implementation of OPS-NGGC-1000, Fleet Conduct of Operations. Risk assessment and equipment removal from service guidance has been removed from 01-01.08.

Distractor Analysis:

Choice A: Plausible because if the valve is de-energized in its intended state (such as a PC IV in the closed dierection) this could be considered correct although for this case the open direction for this normally closed valve would have to have interlocks defeated to have both the F015 and F017 both open (potential to pressurize low pressure piping). OAP-025 is correct.

Choice B: Plausible because if the valve is de-energized in its intended state (such as a PCIV in the closed dierection) this could be considered correct although for this case the open direction for this normally closed valve would have to have interlocks defeated to have both the F015 and F017 both open (potential to pressurize low pressure piping). 001-01.08 no longer provides guidance for evaluating MRIPSA system availability.

Choice C: Correct Answer Choice D: Plausible because RHR Loop is available and 001-01.08 no longer provides guidance for evaluating MRIPSA system availability.

SRO Only Basis: Knowledge of administrative procedures that specify implementation, and/or coordination of plant normal procedures.

Notes

3.0 DEFINITIONS 3.1 Available (Availability)

The status of a system, structure or component (SSC) that is OPERABLE, in service or can be placed in a FUNCTIONAL state within a reasonably short period of time consistent with its intended need. The SSC must be capable of meeting all of its most limiting requirements for the plant mode under consideration Using a manual means for placing an SSC in service requires a dedicated operator assigned to be cognizant of the SSC along with a written procedure for its restoration. A dedicated operator for the purpose of this definition is one who is specifically assigned the task and 2vailable, as necessary, to perform the required actions.

3.2 Backbone Schedule A preliminary schedule consisting of work items that are either required to he performed or have been designated by management as high priority items.

The following items would nomally comprise the backbone schedule:

Implementing Supervisor recommendations KeyI(a)(1 Equipment priority action items Required SurveillancesiPMs System Outages Committed Items Priority 1 & 2 CAPRs, CORRs. and regulatory committed items Modification ECs determined a priority by Engineering representative or Scheduler (must be ready to work with work orders in ready or approved status)

Engineering recommendations Reactivity Management flagged Work Orders 13 Compensatory Actions Measures that are used to niitigate the impact and minimize the duration of an ELEVATED RISK activity. These measures may include CONTINGENCY PLANS or procedural controls.

3.4 Contingency Planning A look ahead process whereby potential problems are systematically identified, assessed, and addressed by adding plans or mitigating actions.

The necessity for a contingency plan is based on the potential consequences as well as the probability of a problem occurring.

DAP-025 Rev 39 I

Page 9 of 121 3.0 DEFINITIONS 3.1 Available (Availability)

The status of a system, structure or component (SSC) that is OPERABLE, in service or can be placed in a FUNCTIONAL state within a reasonably short period of time consistent with its intended need. The SSC must be capable of meeting all of its most limiting requirements for the plant mode under consideration. Using a manual means for placing an SSC in service requires a dedicated operator assigned to be cognizant of the SSC along with a written procedure tor its restoration. A "dedicated" operator for the purpose of this definition is one who is specifically assigned the task and available, as necessary, to perform the required actions.

3.1 Backbone Schedule A preliminary schedule consisting of work items that are either required to be performed or have been deSignated by management as high priority items.

The following items would nom1ally comprise the backbone schedule:

Implementing Supervisor recommendations Key/(a)(1) Equipment priority action items Required Surveillances/PMs System Outages Committed Items - Priority 1 & 2 C.4.PR's, CORR's, and regulatory committed items Modification EC's determined a priority by Engineering repre.sentative or Scheduler (must be ready to work with work orders in ready or approved status)

Engineering recommendations Reactivity Management flagged Work Orders 3.3 Compensatory Actions Measures that are used to mitigate the impact and minimize the duration of an ELEVATED RISK activity. These measures may include CONTINGENCY PLANS or procedural controls.

3.4 Contingency Planning IOAP-025 A look ahead process whereby potential problems are systematically identified, assessed, and addressed by adding plans or mitigating actions.

The necessity for a contingency plan is based on the potential consequences as well as the probabllity of a problem occurring.

Rev. 39 Page 9 of 121 I

ATTACHMENT 3 Page 15 of 19 480V Substation E5?MCCIPaneI Load Summary Load: 480V Motor Control Center 22XA-2 Locadc9: Ur t Reaotor 3T. ldin ZY NE Draw ng

Reference:

-2 3c-9 Upstream Power Source: dBOV Substation E5 COMPT LOAD DESCRIPTION EFFECTS ON LOSS OF POWER DF5 F?lR Cuibaard lniection Valve Loss of load 2-El1-F3l7ATS 3.5.1, 36.1.3, 3.5.2, 3.3.2.1:

D3 RHR Inboard lnection Valve 2-El -FO15A Loss of load ITS 3.51, 3.8.1.3, 3.5.2,3.3.3.1)

DGO HR Tor,s Spray Valve 2-El i-F028A Loss of load tTS3.8.1,3.8.1.3.3.&2.3. 3.3.3.1) 0D7 Rx ecrculation Pump 2A Discharge Valve Losa of load 2-B22-FO31ATS 3.4.1, 3.5.1) 008 Rx Recrculation Pump 2A Discharge Loss of load Sypass Valve 2-232-F032A (TS 3.4.1 3.5.11 001-50.1 Rev. 42 Page 24 of 55 ATTACHMENT 3 Page 15 of 19 4BOV Substation E5/MCC/Panei Load Summary Load: 480V Motor Control Center 2-2XA-2 Location: Unit 2 Reactor Building 20' NE Drawing

Reference:

F-03D49 Upstream Power Source: 4S0V SUbstation E5 COMPT LOAD DESCRIPTION EFfECTS ON LOSS OF POWER DFS RHR Outboard Injection Valve Less of load 2-E11-F017A(TS 3.5.1. 3.6.1.3, 3.5.2, 3.3.3.1}

DF3 RHR Inboard Injecticn Val'.e 2-E11-FOt5A Loss of load

{IS 3.5.1. 3.8.1.3, 3.5.2, 3.3.3.1}

DGO R HR Torus Spray Valve 2-E 11-1"028.11, Loss of load (IS 3.5.1. 3.8.1.3, 3.6.2.3, 3.3.3. t) 007 Rx Recirculation Pump 2A Discharge Valve Loss of load 2-B;?2-F031A iTS 3.4.1. 3.S.1}

008 Rx Recirculation Pump 2A Discharge Loss of load Bypass Valve 2-B32-F032A (is 3.4.1, 3.5.1)

\\ 001-50.1 Rev. 42 Page 24 of 55\\

ATTACHMENT 2A Page 2 of 3 Residual Hear Removal System Loop A Panel Lineup Number Description Positicn Checked Venfied Indication Loop A Control Room Panel H12 P801 El lFcOeC Pump C Shuown Cooling CLOSED Suction V El l-FCOeA Pump A Shutdovn Cooling CLOSED Suction V El l-V32 Check Vae Bypass VIv CLOSED El l-FO17A Ouoard lrec:ion Vlv OPEN El l-F0le4 Drywell Spray Otbd Isol Vlv CLOSED Ell-F1O4A HX 2A Inboard Vent Vlv CLOSED Ell-FOI5A Inboard InjeciionVlv CLOSED El l-FO2IA Orywell Spray Inbd led Vlv LOSEO Ell-FIO3A HX 2A Outboard Vent /1v CLOSES El 1-F024A Torus Cooling Isol Vhi CLOSED Ell-FC4A HX24Bypass1v OPEN El l-F027A Torus Spray led Vi CLOSES El 1-FOl IA HX 2 Drain To Torus Vl CLOSES El 1-FcO4C Pump C Torus Suction Viv OPEN Eil-F028A Torus Discharge IsolVlv CLOSED El I-F026A

-IX 2A Drain To RCIC VIv CLOSES El I-FcO4A Pump A Torus Sucbon VIv OPEN E1l-FcO2A HX 2A Cutlev OPEN El I-FC7A Mn Flow Sypass V CLOSED El 1-FO2GA Pump A&C Torus Suction Vlv CPEN El i-Fc47A HX Z InletV OPEN El l-FE8OA Manual lnection ilv OPEN El l-PDV-Ffl38A HX Z SW Disch VIv CLOSED CS-517A Containment Spray Valve Control OFF Think Swilch 20P-17 Rev. 155 Page 244 of 297 Categories K/A:

S600000G2.02.37 Tier/Group:

T1G1 RO Rating:

3.6 SRO Rating:

4.6 LP Obj:

Source:

NEW Cog Level:

HIGH Category 8:

Numb:r El1-FOOeC Ell-FOOM Ell-V32 Ell-F017.4 El1-F01M Ell-Fl04A Ell-FOl5A El1-F021A El1-Fl03A El1-F024A Ell-F048A Ell-F027A Ell-FOllA Ell-FOO4C Ell-F028A E11-F02M El1-FOO4A E1l-FOO3.4 E11-FOO7A Ell-F020A E11-F047A E11-FOOOA E11-PDV-F068A CS-S17A ATTACHMENT 2.4-Page 2 of 3 Residual Heat Removal System Loop A Panel Lineup Description Positionl Checked Indication Loop A Control Room - Panel H12-P801 Pump C Shulx:lc'IIn Cooling CLOSED SuctionV'N Pump A Shutdown Cooling CLOSED SuclionV'N Check Va'Ne 8}'Jlass VIII CLOSED Outboard Injection VIII OPEN Drywell Spray Otbd Isol VIII CLOSED

!-IX 2A Inboard Vent V'N CLOSED Inboard Injection V'N CLOSED Drywell Spray lnbd (501 Iflv CLOSED

!-IX 2.4 Outboard Vent Vlv CLOSED Torus Cooling 1501 VIII CLOSED

!-IX 2A Bypass 'v1v OPEN Torus Spray lsol V'N CLOSED

!-IX 2A Drain To Torus Vlv CLOSED Pump C Torus Suction V'N OPEN Torus Discharge lsol V'N CLOSED

!-IX 2A Drain To RCIC VI" CLOSED Pump A Torus Suction VIII OPEN

!-IX 2A Outlet Vlv OPEN Min Flow Bypass IJI~'

CLOSED Pump A&C Torus Suction Ifl...

OPEN HX 2A Inlet V'N OPEN M.anuallnieciion Vlv OPEN

!-IX 24 SW Disch VI" CLOSED Containment Spray Va'Ne Control OFF Think Switch Verified 1 20P-17 Rev. 155 Page 244 of 2971 Categories KIA:

RORating:

LP Obj:

Cog Level:

S600000G 2.02.37 3.6 HIGH Tier / Group: T1 G 1 SRO Rating:

4.6 Source

NEW Category 8:

94. What action is required to be taken if Alternate Safe Shutdown (ASSD) Staffing drops below minimum complement due to an emergent on-shift AC illness and what procedure provides the guidance for this action?

The guidance for establishing an (1) if ASSD staffing composition is less than the minimum required is provided by (2)

A (1) ASSD Impairment (2) OASSD-00, User Guide B.

(1) Active LCO for T.S. 5.2.2, Facility Staff Organization, (2) OASSD-00, User Guide C. (1) ASSD Impairment (2) 001-01.01, BNP Conduct of Operations Supplement D. (1) Active LCO forT.S. 5.2.2, Facility Staff Organization, (2) 001-01.01, BNP Conduct of Operations Supplement

94. What action is required to be taken if Alternate Safe Shutdown (ASSD) Staffing drops below minimum complement due to an emergent on-shift AO illness and what procedure provides the guidance for this action?

The guidance for establishing an (1) if ASSD staffing composition is less than the minimum required is provided by (2)

A'I (1) ASSD Impairment (2) OASSD-OO, User Guide B. (1) Active LCO for T.S. 5.2.2, Facility Staff Organization, (2) OASSD-OO, User Guide C. (1) ASSD Impairment (2) 001-01.01, BNP Conduct of Operations Supplement D. (1) Active LCO for T.S. 5.2.2, Facility Staff Organization, (2) 001-01.01, BNP Conduct of Operations Supplement

Feedback K/A: SG2.O1.05 Conduct of Operations Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

(CFR: 41.10 / 43.5/45.12)

ROISRO Rating: 2.9/3.9 Objective: CLSLP304M*1 3m

13. Given ASSD procedures and plant conditions that require use of ASSD procedures, determine the following:
m. The manpower required to support the ASSD actions.

Reference:

OASSD-00, Revision 37, Page 30, Section 5.3.3 Cog Level: High Explanation:

The ASSD staffing composition may be less than the minimum requirements for a period of time not to exceed two hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore requirements to within the minimum requirements of the shift ASSD staffing.

If the ASSD staffing composition is less than the minimum required, establish an Alternative Safe Shutdown Impairment in accordance with OPLP-1.5, Alternative Shutdown Capability Controls, and OFPP-020, Impairment Notification.

Distractor Analysis:

Choice A: Correct Answer Choice B: Plausible because an impairment is the same as LCO (001-01.01) but impairments are not established against TS 5.2.2 and ASSD User Guide is correct.

Choice C: Plausible because ASSD impairment is correct and 001-01.01 provides staffing requirements for TS 5.2.2 but directs use of OASSD-00 procedure use for required staffing.

Choice D: Plausible because Plausible because an impairment is the same as LCO (001-01.01) but impairments are not established against TS 5.2.2 and 001-01.01 provides staffing requirements for TS 5.2.2 but directs use of OASSD-00 procedure use for required staffing.

SRO Only Basis: Requires knowledge of TS 5.2.2 Facility Staff Organization

- and prescribes the procedure required for guidance during periods of ASSD minimum complement not maintained.

Notes Feedback KIA: SG2.01.05 Conduct of Operations Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

(CFR: 41.10/43.5/45.12)

RO/SRO Rating: 2.9/3.9 Objective: CLS-LP-304-M*13m

13. Given ASSD procedures and plant conditions that require use of ASSD procedures, determine the following:
m. The manpower required to support the ASSD actions.

Reference:

OASSD-OO, Revision 37, Page 30, Section 5.3.3 Cog Level: High Explanation:

The ASSD staffing composition may be less than the minimum requirements for a period of time not to exceed two hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore requirements to within the minimum requirements of the shift ASSD staffing. If the ASSD staffing composition is less than the minimum required, establish an Alternative Safe Shutdown Impairment in accordance with OPLP-1.5, Alternative Shutdown Capability Controls, and OFPP-020, Impairment Notification.

Distractor Analysis:

Choice A: Correct Answer Choice B: Plausible because an impairment is the same as LCO (001-01.01) but impairments are not established against TS 5.2.2 and ASSD User Guide is correct.

Choice C: Plausible because ASSD impairment is correct and 001-01.01 provides staffing requirements for TS 5.2.2 but directs use of OASSD-OO procedure use for required staffing.

Choice D: Plausible because Plausible because an impairment is the same as LCO (001-01.01) but impairments are not established against TS 5.2.2 and 001-01.01 provides staffing requirements for TS 5.2.2 but directs use of OASSD-OO procedure use for required staffing.

SRO Only Basis: Requires knowledge of TS 5.2.2 Facility Staff Organization - and prescribes the procedure required for guidance during periods of ASSD minimum complement not maintained.

Notes

5.0 INSTRUCTIONS 5.3 General Guidelines for ASSD Staff 6.31 All ASSD Staffing Roster members must be capable of prompt response when events are in progress that may require entry into ASSO procedures.

5.12 All ASSD members shall obtain a designated radio at the beginning of shift and ensure that it is charged.

rNol-E:

Planned reduction of ASSD personnel below the minimum number required is NOT permitted.

5.3.3 The ASSD staffing composition may be less than the minimum requirements for a period of time not to exceed two hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore requirements to within the minimum requirements of the shift ASSD staffing.

5.3.4 If the ASSD staffing composition is less than the minimum required.

establish an Alternative Safe Shutdown Impairment in accordance with OPLP-1.5, Alternative Shutdown Capability Controls, and OFPP-020, Impairment Notification.

5.3.5 If an impairment exceeds two hours, initiate a Condition Report.

5.3.6 With both units in Mode 4 or 5, ASSD staffing is not required.

DASSO-QO Rev. 37 Page 30 of 53 5.0 INSTRUCTIONS 5.3 General GUidelines for ASSD Staff 5.3.1 All ASSD Staffing Roster members must be capable of prompt response when events are in progress that may require entrl into ASSD procedures.

5.3.2 All ASSD members shall obtain a deSignated radio at the beginning of shift and ensure that it is charged.

NOTE:

Planned reduction of ASSn personnel below the minimum number required is NOT permitted.

5.3.3 The ASSD staffing composition may be less than the minimum requirements for a period of time not to exceed two hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore requirements to within the minimum requirements of the shift ASSn staffing.

5.3.4 If the ASSn staffing composition is less than the minimum required, establish an Alternative Safe Shutdown Impairment in accordance '.'1ith OPLP-1.5, Alternative Shutdown Capability Controls, and OFPP-020, Impairment Notification.

5.3.5 If an impairment exceeds two hours, initiate a Condition Report.

5.3.6 With both units in Mode 4 or 5, ASSD staffing is not required.

10ASSD-OO Rev. 37 Page 30 of 531

5.0 INSTRUCTIONS 5.4 Minimum ASSD Nuclear Shift Staffing/Assignments 5.4.1 Senior Reactor Operators:

1 Unit 1 SCO: Unit 1 Remote Shutdown Panel 1

Unit 2 SCO:

Unit 2 Remote Shutdown Panel 5.4.2 Auxiliary Operators:

1 Unit 1 Reactor BuildinglMCC Operator or as directed by the Unit SCO 1

Unit 2 Reactor BuildinglMCC Operator or as directed by the Unit SCO 1

Diesel Generator Operator or as directed by the Unit SCO 1

Emergency Switchgear Operator or as directed by the Unit SCO 1

Service Water Building Operator or as directed by the Unit SCO QASSD-00 Rev. 37 Page 31 ofj 5.0 INSTRUCTIONS 5.4 Minimum ASSD Nuclear Shift Staffing/Assignments 5.4.*1 Senior Reactor Operators:

Unit 1 seo: Unit 1 Remote Shutdown Panel 1

Unit 2 seo: Unit 2 Remote Shutdown Panel 5.4.2 Auxiliary Operators:

\\ OASSD-OO Unit 1 Reactor Building/MCe Operator or as directed by the Unit seo 1

Unit 2 Reactor BuildingJMCe Operator or as directed by the Unit seo 1

Diesel Generator Operator or as directed by the Unit seo 1

Emergency Switchgear Operator or as directed by the Unit seo 1

Service Water Building Operator or as directed by the UnitSeO Rev. 37 Page 31 of 53\\

9.4 Operations Leadership Role in Station Activities (continued) 5.

Operators work closely with station support personnel to establish appropriate priorities for resoMng plant equipment and station program deficiencies. Being aware of the integrated effect of equipment out of service and establishing priorities for equipment return-to-service consistent with plant impact are key components of this philosophy.

6.

Operations pursues the root cause(s) of problems; provides direction to implement corrective actions and hold department and station personnel accountable for achieving expected levels of perfomance.

9.5.

Operations Shift StaflinO Standards Operations ensures that the Control Room is adequately staffed for plant operations with appropriately qualified individuals. Additionally, Operations ensures staffing is adequate to meet regulatory and programmatic requirements.

Expectations 1.

General a.

The CRS and Shift Manager are responsible for ensuring that only qualified watchstanders hold required positions. Personnel should verify they are qualified for the position to be held prior to assuming the watch.

b.

Individual qualifications for specifc positions can be found in REG-NGGC-0012, Confirmation of Personnel Qualifications Associated with Commitments to Regulatory Guide 1.8.

c.

The shift complement may be one less than the minimum requirement for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of onduty shift members provided immediate action is taken to restore the shift complement to within the minimum requirements. This provision does not permit any shift member position to be unmanned upon shift change due to an oncoming shift member being late or absent.

d.

Shift staffing shall meet the requirements of the indMdual plant license/Tech Specs and other regulatory and programmatic required positions at all times. Required staff numbers and positions can be found in Attachment I Shift Staffing.

OPS-NGGC-1000 Rev.2 I

Page46of 1491 9.4 Operations Leadership Role in Station Activities (continued)

5.

Operators work closely with station support personnel to establish appropriate priorities for resolving plant equipment and station program deficiencies. Being aware of the integrated effect of equipment out of service and establishing priorities for equipment return-to-service consistent with plant impact are key components of this philosophy.

6.

OperatiOns pursues the root cause(s} of problems; provides direction to implement corrective actions and hold department and station personnel accountable for achieving expected levels of perfornlance.

9.5.

Operations Shift Slaffing Standards Operations ensures that the Control Room is adequately staffed for plant operations with appropriately qualified individuals. Additionally, Operations ensures staffing is adequate to meet regulatory and programmatic requirements.

Expectations

1.

General

a.

The CRS and Shift Manager are responsible for ensuring that only qualified watchstanders hold required positions. Personnel should verify they are qualified for the position to be held prior to assuming the watch.

b.

Individual qualifications for specific positions can be found in REG-NGGC-0012, Confirmation of Personnel Qualifications Associated with Commitments to Regulatory Guide 1.8.

c.

The shift complement may be one less than the minimum requirement for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift meml)ers provided immediate action is taken to restore the shift complement to within the minimum requirements. This provision does not permit any shift member position to be unmanned upon shift change due to an oncoming shift meml)er I)eing late or absent.

d.

Shift staffing shall meet the requirements of the individual plant licenseffech Specs and other regulatory and programmatic required positions at all times. Required staff numbers and positions can be found in Attachment 1 "Shift Staffing".

I OPS-NGGC-1000 Rev. 2 Page 46 of 1491

Attachment I

- Shift Staffing Shift Manning BNP Position Minimum Note staffing SM I

CRS 2

SRO/STA 1

RO 3

AO 9

9.5 Operations Shift Staffing In addition to the requirements of OPS-NGGC-1000. the following requirements apply:

9.5.1 General The following table outlines the administrative guideline for the normal Operations shift complement Any deviation from the normal shift complement must remain in accordance with Section 5.2.2 of Technical Specifications, and apphcable sections of IJASSD-OO.

OFPP-031. and Attachment 13. (Attachment 13 contains a listing of required ERO Watch Stations and qualifications for each and ASSD position& This attachment may be used as a tool to support determining shift staffing requirements.)

BNP Watchstations BNP Shift Complement License Shift Manager (SM) 1 Shift Manager SRO Control Room Supervisor (CRS) 2 CRSs (1 for each unit)

SRO Reactor Operator (RO) 4 Reactor Operators rtypically. 2 ROISRO for each unit)

Auxiliarj Operator (AO) 9 (includes 2 in Radwaste)

N/A Operations Center/Field SRO 1 Operations Center/Field SRO SRO STA 1 STA STA Qualified

The STA may stand watch as a CRS or Reactor Operator provided the following requirements are met:

At least 4 SROs are available on shift (this includes the STA hut does NOT include the Fire Brigade Advisor which may be filled by an RO licensed individual).

Another Licensed Operator is designated to relieve the STA as Unit CRS or RO.

(Relief as Reactor Operator is required if only one operator is assigned to a unit Relief as CR5 shall he filled from the CR5 position on the shift staffing roster.)

The designated relief must NOT be assigned as the Fire Brigade Advisor.

The designated relief has taken turnover on the affected unit.

The designated relief must he able to relieve the STA within 10 minutes.

001-Ui.131 Rev.

Page 14 of 177 Sheet 1 of 2 Shift Manning BNP Position SM CRS SRO/STA RO AO - Shift Staffing Sheet 1 of2 Minimum Note staffing 1

2 1

3 9

9.5 Operations Shift Staffing In addition to the requirements of OPS-NGGC-1000, the following requirements apply:

9.5.1 General The following table outlines the administrative guideline for the normal Operations shift complement A.flY deviation from the normal shift complement must remain in accordance with Section 5.2.2 of Technical Specifications, and applicable sections of OASSO-OO, OFPP-031, and Attachment 13. (Attachment 13 contains a listing of required ERO Watch Stations and qualifications for each and ASSD positions. This attachment may be used as a tool to support determining shift staffing requirements.)

BNP Watchstations BNP Shift Com~lement License Shift Manager (SM) 1 Shift Manager SRO Control Room Supervisor (CRS) 2 CRSs (1 for each unit)

SRO Reactor Operator (RO) 4 Reactor Operators (typically, 2 RO/SRO for each unit)

Auxiliary Operator (AO) 9 (includes 2 in Radwaste)

N/A Operations Center/Field SRO 1 Operations Center/Freid SRO SRO STA' 1 STA ST A Qualified

'The STA may stand watch as a CRS or Reactor Operator provided the following requirements are met:

At least 4 SROs are available on shift (thiS includes the STA but does.NOT include the Fire Brigade Advisor which may be filled by an RO licensed individual).

Another Licensed Operator is designated to relieve the STA as Unit CRS or RO.

(Relief as Reactor Operator is required if only one operator is assigned to a unit Relief as CRS shall be filled from the CRS position on the shift staffing roster.)

The deSignated relief must NOT be assigned as the Fire Brigade Advisor.

The deSignated relief has taken turnover on the affected unit.

The deSignated relief must be able to relieve the STA within 10 minutes.

1001-01.01 Rev. 29 Page 14 of 1771

ATTACHMENT 13 Page 1 of 2 Operations Staffing Roster Date:__________________

Shift:________________

STA P613 STA SRO PB11 Unit 1 CRS/U-1 RSD Panel SRO PB1I Unit 2 CRS1U-2 RSD Panel ARO PB12 Unit 1 RB MCC Operator ARC P812 Unit 2 RB MCC Operator ARC PB12

  • FB Advisor CREC PBI7 CREC AC PB14 SW Operator AC P614 DC Operator AC P814 Emergency Switchgear Operator FB (SIC) FBO2FBD3 FB (SIC)

FB FBO2 F8 F8 FBO2 FB FB FGO2 FB FB F802 FB Security Contact E&RC Contact Maintenance Contact A

May hold an RO OR SRO license.

Security Key Accountability iiiui; ii*;

Unit I RB AC Unit2 RBAO Outside AC Unit 1 CRS Un1t2CRS Shift Manager 001-01.01 Rev 29 Page 97 of 177 Date: ______

ATTACHMENT 13 Page 1 of 2 Operations Staffing Roster Shift ______ _

May hold an RO OR SRO license.

Security Key Accountability 1001-01.01 Rev. 29 Page 97 of 1771

5.2.2 Facility Staff The facility staff organization shall include the following:

a.

A total of three non-licensed operators shall be assigned for Brunswick Units I and 2 at all times.

Organization 5.2 5.2 Organization 5.2.2 Facility Staff continued) b.

At east one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, when either unit is in MODE 1, 2, or 3. at least one licensed Senior Reactor Operator (SRO) shall be present in the control room. With one unit in MODE 1, 2, or 3 and the other unit defueled, the minimum shift crew shall include a total of two SROs and two ROs.

c Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and Specifications 5.2.2.a and 5.2.2.g for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

d.

An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

e.

Deleted.

f.

The operations manager or assistant operations manager shall hold an.

SRO license.

g.

The shift technical advisor shall serve in an advisory capacity to the shift superintendent on matters pertaining to the engineering aspects assuring safe operation of the unit when either unit is in MODE 1.2, or 3.

Eiunswick Unit 1 5.0-3 Amendment No. 253 I

5.2.2 Facility Staff The facility staff organization shall include the fol[owing:

a.

A total of three non-licensed operators shall be assigned for Brunswick Units 1 and 2 at all times.

5.2 Organization 5.2.2 Facility Staff (continued)

( continued)

Organization 5.2

b.

At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, when either unit is in MODEl, 2, or 3. at least one licensed Senior Reactor Operator (SRO) shall be present in the control room. V'I'ith one unit in MODEl, 2, or 3 and the other unit defueled, the minimum shift crew shall include a total of two SROs and two ROs.

c.

Shift crew composition may be lee.s than the minimum requirement of 10 CFR SO.54(m){2)(i) and Specifications 5.2.2.a and 5.2.2.g for a peliod of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

d.

An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence: provided immediate action is taken to fill the required pOSition.

e.

Deleted.

f.

The operations manager or assistant operations manager shall hold an SRO license.

g.

The shift technical advisor shall serve in an advisory capacity to the shift superintendent on matters pertaining to the engineering aspects assuring safe operation of the unit when either unit is in MODE 1, 2, or 3.

Brunswick Unit '1 5.0-3 Amendment No. 253 I

Categories K/A:

SG2.0L05 Tier/Group:

T3 RO Rating:

2.9 SRO Rating:

3.9 LP Obj:

CLSLP3O4M*13M Source:

NEW Cog Level:

HIGH Category 8:

Y Categories KIA:

SG2.01.05 Tier / Group: T3 RORating:

2.9 SRORating

3.9 LP Obj:

CLS-LP-304-M* 13M Source:

NEW Cog Level:

HIGH Category 8:

Y

95. OFH-1 1, Refueling, prohibits control rod withdrawal during the core load sequence until a neutronic bridge is established.

Which one of the following core loading sequences establishes a neutronic bridge as described in OFH-1 1?

Four fuel bundles are loaded around (1)

,then fuel is loaded in all fuel cells in a line between SRMs (2)

A. (1) SRMsAand Donly (2) AandD B.

(1) SRMs B and D only (2) B and D C. (1) each of the four SRMs (2) A and D D (1) each of the four SRMs (2) B and D

95. OFH-11, Refueling, prohibits control rod withdrawal during the core load sequence until a neutronic bridge is established.

Which one of the following core loading sequences establishes a neutronic bridge as described in OFH-11?

Four fuel bundles are loaded around (1)

,then fuel is loaded in all fuel cells in a line between SRMs (2)

A. (1) SRMs A and D only (2) A and D B. (1) SRMs Band D only (2) Band D

c. (1) each of the four SRMs (2) A and D D~ (1) each of the four SRMs (2) Band D

Feedback K/A: SG2.01.42 Conduct of Operations Knowledge of new and spent fuel movement procedures.

(CFR: 41.10/43.7/45.13)

RO/SRO Rating: 2.5/3.4 Objective: CLSLP305C*

Reference:

OFH-1, Revision 93, Page 9, Section 4.37 Cog Level: High Explanation:

Provide ENP-24-12, Figure 1 as a reference From FH-11, 4.37 To help ensure that an unmonitored criticality will not occur, control rod withdrawal is not allowed during the core reload sequence until a neutronic bridge is established. The neutronic bridge ensures that two SRMs are neutronically coupled, thus monitoring the loaded area of the core. The reload sequence has three basic steps. Four fuel bundles are loaded around each of the four SRMs, the neutronic bridge is established and a spiral reload of the other fuel bundles completes the sequence. The neutronic bridge is established by loading fuel in all fuel cells in a line between two SRMs. These SRMs must be on opposite sides of the core and the line of loaded fuel cells must intersect the center of the core (A to D would not intersect the center, B to D would).

Distractor Analysis:

Choice A: Plausible because loading fuel around 2 SRMs and a line between them would establish a neutron bridge (between those 2 SRM5) but not PAW OFH-1 1 and A&D are adjacent.

Choice B: Plausible because loading fuel around 2 SRMs and a line between them would establish a neutron bridge (between those 2 SRM5) but not JAW OFH-1 I and A&D are adjacent.

Choice C: Plausible because loading fuel around all SRMs is correct but A&D are adjacent.

Choice D: Correct Answer SRO Only Basis: IOCFR55.43.6 Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity.

IOCFR55.43.7 Fuel handling facilities and procedures.

Notes Feedback KIA: SG2.01.42 Conduct of Operations Knowledge of new and spent fuel movement procedures.

(CFR: 41.10/43.7/45.13)

RO/SRO Rating: 2.5/3.4 Objective: CLS-LP-305-C*

Reference:

OFH-1, Revision 93, Page 9, Section 4.37 Cog Level: High Explanation:

Provide ENP-24-12, Figure 1 as a reference From FH-11, 4.37 To help ensure that an unmonitored criticality will not occur, control rod withdrawal is not allowed during the core reload sequence until a neutronic bridge is established. The neutronic bridge ensures that two SRMs are neutronically coupled, thus monitoring the loaded area of the core. The reload sequence has three basic steps. Four fuel bundles are loaded around each of the four SRMs, the neutronic bridge is established and a spiral reload of the other fuel bundles completes the sequence. The neutronic bridge is established by loading fuel in all fuel cells in a line between two SRMs. These SRMs must be on opposite sides of the core and the line of loaded fuel cells must intersect the center of the core (A to D would not intersect the center, B to D would).

Distractor Analysis:

Choice A: Plausible because loading fuel around 2 SRMs and a line between them would establish a neutron bridge (between those 2 SRMs) but not lAW OFH-11 and A&D are adjacent.

Choice B: Plausible because loading fuel around 2 SRMs and a line between them would establish a neutron bridge (between those 2 SRMs) but not lAW OFH-11 and A&D are adjacent.

Choice C: Plausible because loading fuel around all SRMs is correct but A&D are adjacent.

Choice D: Correct Answer SRO Only Basis: 10CFR55.43.6 Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity.

10CFR55.43.7 Fuel handling facilities and procedures.

Notes

4.0 PRECAUTIONS AND LIMITATIONS 4.34 RPS shorting links SHALL be removed for control rod withdrawal (except for control rods removed in accordance with Technical Specifications) in the refuel mode when core verification AND subsequent strongest rod out verification have NOT been performed. Control rods may be withdrawn with the shorting links installed, provided core verification (QENP-24.13),

subsequent strongest rod out verification (single control rod suhcriticality test in accordance with OFH-1 1) have been performed, and the one-rod-out refuel interlocks have been demonstrated to be operable.

4.35 An SRO with no other concurrent duties shall directly supervise all core alterations.

4.36 Members of fuel handling crew, scheduled for consecutive daily duty, should NOT normally work more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> out of each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.37 To help ensure that an unmonitored criticality will NOT occur, control rod withdrawal is NOT allowed during the core reload sequence until a neutronic bridge is established. The neutronic bridge ensures that two SRM5 are neutronically coupled, thus monitoring the loaded area of the core. The reload sequence has three basic steps. Four fuel bundles are loaded around each of the four SRMs, the neutronic bridge is established and a spiral reload of the other fuel bundles completes the sequence. The neutronic bridge is established by loading fuel in all fuel cells in a line between two SRM& These SRMs must be on opposite sides of the core and the line of loaded fuel cells must intersect the center of the core.

4.38 With fuel removed, if a control rod is withdrawn without blade guides installed, the insertion capability shall be removed for the control rod.

4.39 The Bridge Operator should immediately push the STOP button if the bridge fails to respond to Operator commands, such as speed changes or jogs. The STOP button will prevent all bridge movement 4.40 If attaching tools, such as ajet pump grapple or control blade latching tool, to either the monorail or frame mounted hoist, verify proper thread engagement/size by ensuring there is no play in the connection prior to thread engagement of three (3) full tum& The correct tool and coupling thread size is 7116-14 UNC. Additionally, a 112-13 UNC bolt will not fit into a proper size tool (7/16-14 U NC); thus, this check may be performed if practical. Failure to detect mis-matched thread sizes will significantly reduce the load capacity of the tool/hoist.

4.41 Indication of criticality observed on the SRM indicators during functional, subcritical, or shutdown margin rod checks shall be reason to teminate fuel loading until a complete evaluation of the cause of the criticality indication is determined.

OFH-1 1 Rev. 93 Page 9 of 55 4.0 PRECAUTIONS AND LIMITATIONS 4.34 RPS shorting links SHALL be removed for control rod withdrawal (except for control rods removed in accordance with Technical Specifications) in the refuel mode when core verification AND subsequent strongest rod out verification have NOT been performed. Control rods may be withdrawn with the shorting links installed, provided core verification (OENP-24.13),

subsequent strongest rod out verification (single control rod sub criticality test in accordance with OFH-11) have IJeen performed, and the one-rod-out refuel interlocks have been demonstrated to IJe operable.

4.35 An SRO with no other concurrent duties shall directly super"ise all core alterations.

4.36 Members of fuel handling crew, scheduled for consecutive dail~' duty, should NOT nom13l1y work more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> out of each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.37 To help ensure that an unmonitored criticality will NOT occur, control rod withdrawal is NOT allowed during the core reload sequence until a neutronic bridge is established. The neutronic bridge ensures that two SRMs are neutronically coupled, thus monitoring the loaded area of the core. The reload seQuence has three basic steps. Four fuel bundles are loaded around each ofthe four SRMs, the neutronic bridge is established and a spiral reload of the other fuel bundles completes the sequence. The neutronic bridge is established by loading fuel in all fuel cells in a line between two SRMs. These SRMs must be on OPPOsite sides of the core and the line of loaded fuel cells must intersect the center of the core.

4.38 With fuel removed, if a control rod is withdrawn without blade guides installed, the insertion capabiliPf shall be removed for the control rod.

4.39 The Bridge Operator should immediately push the STOP button if the bridge failS to respond to Operator commands, such as speed changes or jogs. The STOP button will prevent an bridge movement.

4.40 If attaching tools, such as a jet pump grapple or controll)lade latching tOOl, to either the monorail or frame mounted hoist, verify proper thread engagement/size by ensuring there is no play in the connection prior to thread engagement of three (3) full turns. The correct tool and coupling thread size is 7116-14 UNC. Additionally, a 112-13 UNC IlOlt wiI[ not fit into a proper size tool {7/16-14 UNC); thus, this check may be performed if practical. Failure to detect mis-matched thread sizes will significantly reduce the load callacity of the tOOl/hoist.

4.41 Indication of criticality observed on the SRM indicators during functional, subcritical, or shutdown margin rod checks shall be reason to temlinate fuel loading until a complete evaluation of the cause oftM criticality indication is determined.

IOFH-11 Rev. 93 Page90f55 I

FIGURE 09.1-2 IN-Core Instrurnenation Location Diagram 00 LI IRM DETECTOR LOCATION DETECTOR LOCATION PLANT NORTH CORE RM LOCATION A

12.41 B

36-41 C

20-33 0

283.3 CORE IRM LOCATION E

28.25 F

20-26 G

36-09 H

12-09 SD-091 Rev. 6 Page 49 of 61 Categories K/A:

SG2.O1.42 RORating:

2.5 LP Obj:

Tier/Group:

T3 SRO Rating:

3.4 Source

BANK CORE SRM LOCATION A

12.33 B

2641 C

36-26 0

20.17 Cog Level:

HIGH Category 8:

Y FIGURE 09.1-2 IN-Core Instrumentation Location Diagram r

r r!,

LA..J r

(, ~

~ ~

L L

rHl I I

L 12 CORE SRM LOCATION A

12.33 a

28-41 C

36-25 0

20*11 I SD-09.1 Categories KIA:

SG2.01.42 RO Rating:

2.5 LPObj

Cog Level:

HIGH PLANT ~

NORTH,..".-

n h

/'" ~

l/

B,-l

--41

~

ffr

-33 4:f. ~

t:l.f; ~

'( -:;

-25

~ ~

~ :J w

-11 L.~

w

--09 W

20 28 D IRM DETECTOR LOCATION ZSIO SRM DETECTOR LOCATION CORE CORE IRM LOCATION IRM LOCATION A

12*41 E

28*25 B

36-41 F

20*25 C

20*33 G

36*09 0

28.33 H

12*09 Rev. 6 Page490f61 I Tier / Group: T3 SRORating:

3.4 Source

BANK Category 8:

Y

96. With Unit Two operating at power, Annunciator RCIC TURBINE STM LINE DRN POT LEVEL HI alarms and the RO observes the E51-F054, F025, & F026 indicate closed on Panel P601.

Which one of the following identifies the cause of the above indications and the operability status of RCIC?

(Reference provided)

These valves are closed due to loss of (1) and after taking the appropriate actions in the annunciator procedure the system would be declared Inoperable and(but)

(2)

A.

(1) pneumatics (2) Unavailable B (1) pneumatics (2) Available C. (1) DC Power (2) Unavailable D. (1) DC Power (2) Available

96. With Unit Two operating at power, Annunciator RCIC TURBINE STM LINE DRN POT LEVEL HI alarms and the RO observes the E51-F054, F025, & F026 indicate closed on Panel P601.

Which one of the following identifies the cause of the above indications and the operability status of RCIC?

(Reference provided)

These valves are closed due to loss of (1) and after taking the appropriate actions in the annunciator procedure the system would be declared Inoperable and (but)

(2)

A. (1) pneumatics (2) Unavailable B:-" (1) pneumatics (2) Available C. (1) DC Power (2) Unavailable D. (1) DC Power (2) Available

Feedback K/A: SG2.02.15 EQUIPMENT CONTROL Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc.

(CFR: 41.10 143.3/45.13)

ROISRO Rating: 3.9/4.3 Objective: CLSLP016*15e

15. Given plant conditions, predict the RCIC System response to the following conditions:
s. Loss of instrument air.
e. DC power failure.

Reference:

2APP A-03 3-5, Revision 49, Page 44 Cog Level: High Explanation:

Valves fail closed on loss of DC power or Pneumatics, however with a loss of power, position indication on P601 will also be lost. Per APP A-03, 3-5

- If either E51-F025 or E51-F026 has been failed closed for more than 5 minutes, perform the following:

a. Close Turbine Trip and Throttle Valve, E51 -V8, to prevent water hammer damage from a RCIC auto start.
b. If RCIC must be started, proceed to OP-16.

this would still make RCIC available for use per the procedure but it is inoperable because it will not auto start as required.

This will make the RCIC system inoperable but available to be restarted per the procedure.

Distractor Analysis:

Choice A: Plausible because loss of pneumatics only is correct and the system will not start in auto when required, but could be manually started.

Choice B: Correct Answer Choice C: Plausible because a loss of power will cause valves to fail closed, but with loss of power position indication will be lost and the system will not start in auto when required.

Choice D: Plausible because pneumatics and power will cause valves to fail closed, but with loss of power position indication will be lost and it is available to start per the procedure which makes it available.

SRO Only Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, emergency conditions.

Notes Feedback KIA: SG2.02.15 EQUIPMENT CONTROL Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc.

(CFR: 41.10/43.3/45.13)

RO/SRO Rating: 3.9/4.3 Objective: CLS-LP-016*15e

15. Given plant conditions, predict the RCIC System response to the following conditions:
s. Loss of instrument air.
e. DC power failure.

Reference:

2APP A-03 3-5, Revision 49, Page 44 Cog Level: High Explanation:

Valves fail closed on loss of DC power or Pneumatics, however with a loss of power, position indication on P601 will also be lost. Per APP A-03, 3-5 -If either E51-F025 or E51-F026 has been failed closed for more than 5 minutes, perform the following:

a. Close Turbine Trip and Throttle Valve, E51-V8, to prevent water hammer damage from a RCIC auto start.
b. If RCIC must be started, proceed to OP-16.

this would still make RCIC available for use per the procedure but it is inoperable because it will not auto start as required.

This will make the RCIC system inoperable but available to be restarted per the procedure.

Distractor Analysis:

Choice A: Plausible because loss of pneumatics only is correct and the system will not start in auto when required, but could be manually started.

Choice B: Correct Answer Choice C: Plausible because a loss of power will cause valves to fail closed, but with loss of power position indication will be lost and the system will not start in auto when required.

Choice D: Plausible because pneumatics and power will cause valves to fail closed, but with loss of power position indication will be lost and it is available to start per the procedure which makes it available.

SRO Only Basis: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, emergency conditions.

Notes

RCIC STEAM POT Partial P&ID RCIC STEAM POT Partial P&ID

~

QRAIN!

I

.f.£lI.

F04S

Unit 2

APP A-03 3-S Page 2 of 2

ACTIONS (Cont a)

CAUTION If Main Steam Line Drain Vlv, NVD-F021, fails to close, then the Main Steam Line Drain Inboard and Outboard Isolation valves must be closed.

6. If reguii-ed, then close Main Steam Line Drain Inbd lad
Vlv, B21-F0l6, and Main Steam Line Drain Otbd Isol Vlv, B2l-FOl.
7. If alarm fails to clear within five minutes after completion of actions 1 2,3, 5, or 6, then dispatch an Auxiliary Operator to the Drwell access roof to determine if the HPCI/RCIC Condensate Drain Line Back Pressure Orifice is plugged or the drain line isolated.

NOTE Greater than 500 psig on HPCI/RCIC Back Pressure Orifice Inlet Pressure Gauge, 2-MVD-PI-7146 would be an indication of a plugged orifice.

8.

IF back pressure orifice is plugged, a,

Open HPCI/RCIC (ond Drn Line Back Press Orifice Bypass Valve, 2-MVD-VSOO2.

b.

Close BPCI/RCIC Cond Drn Line Back Press Orifice Inlet Isol

Valve, 2-MVD-VS000.

c.

Close HPCI/RCIC Cond Drn Line Back Press Orifice Outlet Isol

Valve, 2-MVD-VSOOl.

d.

Place valves under proper administrative control.

9.

If NPCI/RCIC Cond Drain Line is isolateth a.

Open HPCI/RIC (ond Drn Line Back Press Orifice Inlet Isol

Valve, 2-MVD-VS000.

b.

Open HPCI/RCIC Cond Drn Line Back Press Orifice Outlet Isol

Valve, 2-MVD-VSOOl.

10.

IF a circuit malfunction is suspected, ensure that a WE/JO is prepared.

DEVICE/SETPOINTS Level Switch ES1-LSH-NO1O-l Instrument failure in the Switch Point tl dry condition/1980 my.

Level Switch ESl-LSH-NOl0-l Also detects instrument failure in the Switch Point #2/0 +/- 2 wet condition.

Incox-porates water.

100 sec time delay in annunciator circuitry.

POSSIBLE PLANT EFFECTS Damage to the RCIC turbine due to high moisture carryover on the steam.

REFERENCES 1.

LL-9364 50 2.

OP-iS, RCIC System Operating Procedure 2APP-A-03 Rev. 49 Page 45 of 102 Unit 2 APP A-03 3-5 Page 2 of 2 ACTIONS (Cont'd)

CAUTION If r~in Steam Line Drain v~v, MVD-F021, fails to close, then the Main Steam Line Drain Inboard and Outboard Isolation valves must be closed.

6. If required, then close [*lain Stearn Line Drain Inbd Iso1 V1v, B21-F016, and Main Steam Line Drain Otbd Iso1 V1v, B21-F019.
7. If alarm fails to clear within five minutes after completion of actions 1 2,3, 5, or 6, then dispatch an Auxiliary Operator to the Drywe11 access roof to determine if the HPCI/RCIC Condensate Drain Line Back Pressure Orifice is plugged or the drain line isolated.

NOTB:

Greater than 500 psig on HPCI/RCIC Back Pressure Orifice Inlet Pressure Gauge, 2-1*lYD-PI-7146 would be an indication of a plugged orifice.

8.

IF back pressure orifice is plugged,

a.

Open HPCI/RCIC Cond Drn Line Back Press Orifice Bx~ss Valve, 2-['lVD-V5002.

b.

Close HPCI/RCIC Cond Drn Line Back Press Orifice Inlet Iso1 Valve, 2-t-1VD-V5000.

c.

Close HPCI/RCIC Cond Drn Line Back Press Orifice Outlet Isol Valve, 2-t-r.rD-V5001.

d.

Place val,,'es under proper adrninistrat ive control.

9.

If HPCI/RCIC Cond Drain Line is isolated:

a.

Open HPCI/RCIC Cond Drn Line Back Press Orifice Inlet Isol

Valve, 2-M'~~-V5000.
b.

Open HPCI/RCIC Cond Drn Line Back Press Orifice Outlet Isol Valve, 2-M'.rD-V5001.

10. IF a circuit malfm1ction is suspected, ensure that a WR/JO is prepared.

D~vlCE/SETPOINTS Level Switch E51-LSH-NOIO-l Switch Pc,int #1 Level Switch E51-LSH-NOIO-l the Switch Point #2/0" +/- 2" water.

POSSIBLE PLANT EFFECTS Instrument failure in the dry condition/1980 mV.

.~lso detects instrument fa.ilure ill wet condition.

Inc01~orates 100 sec time delay in ~1nunciator circuitry.

Damage to the RCIC turbine due to high moisture carryover on the steam.

REFERENCES

1.

LL-9364 - 50

2.

OP-16, RCIC System Opera.t ing Procedure I2APP-A-03 Rev. 49 Page 45 of '1021

Unit 2

APP A-03 3-S Page 1 of 2

RCIC TIJREINE STh LINE DRN POT LEVEL HI RCIC Turbine Steam Line Water Drain Pot High Level)

AUTO ACTIONS 1.

Supply Drain Pot Drain Bypass Valve, ES1-F054, opens.

CATJSE 1.

Heavy condensate load dul-ing steam line warmup.

2.

Normal orifice clogged.

3, HPCIJRCIC ond.

Drain Line Back Pressure Orifice is plugged.

4.

Drain line isolation valves to main condenser closed.

5.

Drain pot level instrument failure or loss of instrument, power.

6.

Circuit malfunction.

OBSERVATIONS 1.

RCIC Supply Drain Pot Drain Byp Valve, ESl-F054, opened.

NOTE:

If alarm occurs and the ES1-F054 valve does not automatically open, the most probable cause is instrument failure or loss of instrument power (Panel 2B-Rx PH1O1T CKT 14),

NOTE:

Additional LED indications are available inside the level element control box device HSE (RB 20 elevation) as follows:

Normal status No annunciator No LEDs illuminated High Water level Green LED on Instrument failure Red LED on ACTIONS 1.

Ensure Supply Drain Pot Drain Byp Vlv, ESl-F054, is open.

2.

Ensure RCIC Supp Pot Inbd Isolation Valve, ESi-F025, is open.

3.

Ensure RCIC Supp Pot Outbd Isolation Valve, ES1-F026, is open.

NOTE:

Valves E51-F025 and E51-F026 will close on loss of instrument air and will also close if E5l-F045 is not fully closed.

Valves ES1-F025 and E51-F026 cannot be opened in either of these conditions.

4.

If either ESl-Ft25 or ESl-F026 ha been failed closed for more than 5 minutes, perform the following:

a.

Close Turbine Trip and Throttle Valve, ESl-V8, to prevent water hammer damage frcm a RCIC auto start.

b.

If RCIC must he started, proceed to OP-16.

S.

Ensure Main Steam Drain Lme Vlv, MVD-F02l, is cload.

2APP-A-03 Rev. 49 Page 44 of 102 Unit 2 APP "~.-03 3-S Page 1 of 2 RCIC TURBINE STI-l LINE DRN POT LEVEL HI iRCIC Turbine Steam Line Water Drain Pot High L=vel)

"~UTO ACTIONS C.... USE

1.

Supply Drain Pot Drain Bypass Valve, ESI-F054, opens.

1.

Heavy condensate load during steam line warmup.

2.

Normal orifioe clogged.

3.

HPCI!RCIC Cond. Drain Line Baok Pressure Orifioe is plugged.

4.

Drain line isolation valves to main oondenser olosed.

S.

Drain pot level instrument failure or loss of instrument po;;er.

6.

Cirouit malfunction.

OBSERVATIONS

1.

RCIC Supply Drain Pot Drain Byp Valve, ES1-F054, opened.

NOTE:

If alarm ooours and the E51-FOS4 valve does not automatioally open, the most probable oause is instrument failure or loss of instrument power (Panel 2B-Rx "HIO" CKT 14/.

NOTE:

Additional LED indioations are available inside the level element control box device H5E (P£ 20' elevation) as follo;;s:

Normal status No annunciator No LEDs illuminated Green LED on Red LED 011.

ACTIONS High Water level Instrument failure

1.

Ensure Supply Drain Pot Drain Byp "lIlv, ESI-FOS4, is open.

2.

Ensure RCIC Supp Pot IOOd Isolation Valve, ESI-F025, is open.

3.

Thlsure RCIC Supp Pot Outbd Isolation Valve, ESI-F026, is open.

NOTE:

Valves E51-F025 and ESI-F026 will close on loss of instrument air and will also close if ESI-F045 is not fully olosed.

Valves ESI-F025 and E51-F026 oannot be opened in either of these conditions.

4.

If either ESI-F025 or ES1-F026 has been failed olosed for more than 5 minutes, perform the following:

a.

Close Turbine Trip and Throttle Valve, ESI-V8, to prevent water hammer damage from a RCIC auto start.

b.

If RCIC must be started, proceed to OP-16.

S.

Ensure Main Steam Drain Line Vlv, MVD-F02l, is olosed.

I2APp-A-03 Rev. 49 Page 44 of '1021

I IF rapid isolation is NOT desired, THEN PERFORM the following to isolate and de-pressurize the RCIC steam supply line:

a.

CLOSE STEAM SUPPLY INBOARD 1SOL VLV, E51-F007.

b.

OPEN HPCLRCIC COND DRN LINE SACK PRESS ORiFiCE BYPASS VALVE MVD-V5002.

c.

OPEN TURBiNE STEAM SUPPLY VL E1-FO45, AND MONITOR turbine response.

d.

CLOSE SUPPLYDRAIN POT1N8D DRAIN VLV, E1-FO25.

e.

CLOSE SUPPLY DRAiN POT OTBD DRAIN VLV, E51-FO26.

20P-16 Rev. 107 Page 34 oi1 R

R3r9rr Uae 8.4 Isolating the RCIC System Steam Supply 8.4.1 Initial Conditions 1.

All applicable prerequisites listed in Section 4.0 are met.

8.4.2 Procedural Steps 1.

IF rapid isolation of RCIC steam line is desired, THEN PERFORM the following:

a.

CLOSE STEAM SUPPLYIWBOARD ISOL VLV, E1-FOO7.

b.

CLOSE STEAMSUPPLYOUTBOARDISOL VLV, E1-FOO8.

8.4 Isolating the RCIC System Stearn Supply 8.4.1 Initial Conditions

1.

All applicable prerequisites listed in Section 4.0 are met.

8.4.2 Procedural Steps

1.

IF rapid isolation of RCIC steam line is desired, THEN PERFORM the following:

a.
b.

CLOSE STEAM SUPPL Y INBOARD /SOL VL V.

E51-F007.

CLOSE STEAM SUPPL Y OUTBOARD ISOL VL 11, E5'1-FOOB.

CAUTION o

o o

Opening the TURBINE STEAM SUPPL Y VL V. E51-F045, to de-pressurize the RCIC steam line will roU the RCIC turbrne,

2.

IF rapid isolation is NOT desired, THEN PERFORM the following to isolate and de-pressurize the RCIC steam supply line:

a.

CLOSE STEAM SUPPL Y INBOARD ISOL VL V.

0 E51-FOO7,

b.

OPEN HPCllf?CIC COND DRN LINE BACK 0

PRESS ORIFICE BYPASS VALVE, MVD-V5002.

c.

OPEN TURBINE STEAM SUPPL Y VL V, 0

E51-F045, AND MONITOR turbine response.

d.

CLOSE SUPPLY DRAIN POT lNBD DRAIN VLV; 0

E51-F025.

e.

CLOSE SUPPLY DRAIN POT OTaD DRAIN VL1I, 0

E51-F026.

R Rijlefijr Use 12oP-16 Rev. 107 Page 34 of 891

8.12 Controlled Manual Start of the RCIC System With Turbine Steam Line R

Drain Pot High Level or RCIC Pump Low Discharge Pressure Refrence Indicated 8.12.1 Initial Conditions 1.

IF RCIC is being operated for a planned evolution (non emergency operation), THEN Health Physics (HPs) shall be notified to attend the pre-job briefing AND a log entry made to identify the individual contacted.

2.

One of the following conditions exist:

a.

The RCIC turbine has been shutdown or tripped and annunciator RC!C TURBINE STM LINE DRN POT LEVEL HI (A-03 3-5) sealed in.

b.

The RCIC turbine has been shutdown or tripped and the RCIC PUMP DISH PRESS LOW annunciator (A-02, 1-6) is sealed in.

3.

A controlled manual start of RCIC is desired.

8.12.2 Procedural Steps CAUTION The RCIC turbine has the potential for failures that could cause personnel injuries. The potential is most significant when the system is initially started after control system maintenance, or after an extended period of being idle. Announcing turbine starts and clearing of all personnel from the RCIC area are required during this period of risk.

Permission to access this area during initial RCIC roll requires the approval of the Unit sco.

1.

EVACUATE all personnel f 0111 he RCIC turbine area.

20P-16 Rev. 107 Page 52 of 89 8.12 Controlled Manual Start of the RCIC System With Turbine Steam Line Drain Pot High Level or RCIC Pump Low Discharge Pressure Indicated 8.12.1 Initial Conditions IF RCIC is being operated for a planned evolution (non-emergency operation), THEN Health Physics (HPs) shall be notified to attend the pre~ob briefing AND a log entry made to identify the individual contacted.

o

2.

One of the follol;1,1ng conditions exist

3.

8.12.2

a.
b.

The RCIC turbine has been shutdown or tripped and annunciator RCIC TURBINE STM LINE DRN POT LEVEL HI (A-03 3-5) sealed in.

The RCIC turbine has been shutdown or tripped and the RCIC PUMP DISCH PRESS LOW annunciator (A-02, 1-6) is sealed in.

A controlled manual start of RCIC is desired.

Procedural Steps CAUTION o

o o

The RCIC turbine has the potential for failures that could cause personnel injuries. The potential is most significant when the system is initially started after control system maintenance, or after an extended period of being idle. Announcing turbine starts and clearing of all personnel from the RCIC area are required during this period of risk.

Permission to access this area during initial RCIC roll requires the approval of the Unit SCO.

EVACUATE all personnel from the RCIC turbine area.

0 R

Reference Use 120p--.16 Rev. '107 Page 52 of 891

ATTACHMENT 2A Page5of 30 PANEL 4A LOCATION NORMAL SUPPLY Reference Drawing LL-3024-6 Control Building 49 ft East Switchboard 2A CIRCUIT LOAD EFFECT Rx.Annunciator Logic, 2-H12-P630 1.

Auto transfers to alternatesource, PaneI4B circuit 1.

Panels 80t/503 2.

Receive annunciator A5-&.

2 HPCI Flow controler 1.

Controller fails downscale.

E41-FIC-K800 (24 VDC;.

2.

Loss of f;ow ndicatcn.

3.

Receive annunciator A 1-2-5.

4.

Loss of HPCI 5.

Loss of ASSD function.

HPC1 Supervisory Lights 1.

Loss of E41-VS and E4l-V indicaticn 2.

Loss HPCI oil tank level HilLO alarm.

HPCI Vertical Board meters 1.

Loss of pressure transmitters/meters R6OI R32, R503, R80(

(52.5 VDC)

HPCI Turbine Speed Control 1.

Loss of speed control, 2GM and soeed sensor.

2.

Loss of speed indicaon on veitcal board.

E41-F053, E41-FOE4, E41-F026 1.

Fail closed.

2.

E41-F054 and 241-F026 loss of indication.

251-F006,E51-F025 1.

Failclosed.

2.

Loss of indication, 001-50 Rev. 45 I

Reference Drawin CIRCUIT lOAD 1

R>:. Annunciator logic, 2-H12-P630

i.

Panels 60 !fe03

2.

2 HPCI Flow controller

[1 E41-FIC-K600 (24 \\lDC)

2.
3.
4.
5.

HPCI Super~'isory Lights

1.
2.

HPCI Vertical Board meters

[1.

(52.5 VDC)

HPCl Turbine Speed Control

i.
2.

E41-F053, E41-F054. E41-F026

1.
2.

E51-F005, E51-F025

1.
2.

1001-50 ATTACHMENT 2A Page 5 of 30 NORMAL SUPPLY Switchboard 21\\

EFFECT Auto transiers to alternate source, Panel4B circuit 1 Receive annunciator Ae-5-S.

Controller fails downsca.ie.

loss of flo'l/ indication.

Reoeive annunciator A 1-2-5.

loss of HPCr loss of ASSD function.

loss of E41-1/S and E41-V9 indication loss HPCI oil lank le... el HillO alam *.

loss of pressure transmitters/meters Re01, R602. Re~3, R60!

loss of speed control, EGM and speed sensor.

loss of speed indication on vertioal board.

Faif closed.

E41-F054 and E41-F026 loss ofindication.

Fail clcsed.

loss of indication.

Rev. 45

ATTACHMENT 25 Page 90132 PANEL: 4B LOCATION:

NORMAL SUPPLY:

Reference Drawing: LL-3024-7 Control Building 49 ft South Switchboard 2B Cku:,

LOAD EFFECT E

Recrc Pump 3 Auxiliary EqtLpment 1.

Loss of aJternate control power to:

Aternate Control Power Recirc B Gen. Fielo Breaker, control, frp and indicaior.

Recirc B Scoop Thbe Power Failure Look & Reset Recirc Lube Oil Pumps S-i ano 8-2, control and ndication.

Reciro B Lock out lTrp Logic ATWS Trip Logic S 2.

Normal power is from Panel 1OA, ckt 2.

Backup Scram valve. 2-C12-1 lflS 1.

Backup Scram valve fails closed; Div I Backup Scram valve can still fun Div II Backup Scram Logic 1.

Sc-ram Discharge Volume Vent and 2ran Vaves wil no receive a close valves will sti I function with Div I.

2.

DFWLCS wD rot receive auto set down from Div II. Digital eedwater w 2.

Ten-second tme delay por to scram reset, will not function for B RPS 7

Spare Spare RCIC Flow controller 1.

Controller sails cownecale.

E51 -FlC-KDO (24 VOC) 2.

Loss of flow indication.

3.

Receive annunciator A3-6-5.

RCIC Supervisory Lights 1.

Loss of E51-V8 and 251-VQ ndic-ation.

RCIC Vertical Board meters 1.

Loss of pressure :ransmittersmeters ROOl, RO2, RD3, RCIJ4 on the P i2.5 VDC1 E51-F02& E51-FCD4. E51-F054 1.

Fail closed.

2.

Loss of indicaon.

HPCI E41-FC25 1.

Fail closed.

2.

Loss of indicaon.

RCIC 2GM 1.

Loss of speed control.

2.

Loss of speed indication on RIGS RCIC Initiation and Control Logic 1.

RCIC will not auto initiate. Cannot be manually operated.

2.

Receive annunciator A3-1-4.

2.

Mm flow valve will not auto open.

4.

Barometric condenser vacuum tank auto level control nop.

001-50 I

Rev. 45 r

PIPING HY QLlo PUi tC 1i4 13 tLJRB.VNDQR h1 REv)SL PLR C 6t164 55 R-Vt5fD PIR rc 64l2 A

PROGRESS ENERGY li-(MQRMATIOtI ON HS oNAw:1o CDMPI IFS WllI Ctk %l

tJtT 2 5i S/A 2-FP46 (l PO 729E48B SH 2)

REACIOR BL.i1DING RLACJOR CORE ISOLATION NOTE RLVISONS TO THIS ORAWING MUST ALSO 1E COOLiNG SYSII M P4CORPORATO ON lE CORRESPONDING ORAW14S:

PIPING I)IAG AM C) 1476, C) 0427, 0 04219, I) 04220 &

1) 04221 Dø229 PANEl:4B Reference Drawing: lL..,3024-7 Ckt#

LOAD 5

Recirc Pump 8 AuxillaI'I Equipment Alternate Control PeweI t)

Backup Scram valve, 2-C12-i=11 08 Div II Backup Scram Logic 7

Spare 8

RCIC Flo'll contrcller E5l-FIC-K800 (24 VDq RCtC Supervisory lights RCIC Vertical Board meters (52.5VDCi E51-F026. E51-F004. E51-F054 HPCr E41-F025 It RC1CEGM RCIC Initiation and Contrel Logic 1001-50 ATTACHMENT 29 Page 9 of 32 LOCATION:

J Contra! Building 49 ftSouth NOBMAt SUt>,P:L Y:

.. Switchboard 2B EFFECT

1. Loss of alternate control power to:

Recilc B Gen. Field Breaker, oontre!, trip and indication.

Recilc B Scoop Tube Power Failure Lock & Reset Recife Lube Oil Pumps 8-1 and B-2. control and indicatien.

Recifc B lock out ITrip logic A TWS Trip logic 8

2. Normal power is irom Panel lOA, ckt 3.
1. Backup Scram val'Je fails dosed; Div I Backup Scram valve can still fun,
1. Sc*ram Discharge Volume Vent and Drain Va.lves will nOl receive a close lIallies wilf still function wilh Dill I.
2. DFWLCS w~1 not receive auto, set down from Dill rI. Digital Feedwater w
3. Ten-second time delav prior to scram reset, will not function for B RPS !

Spare

1. Conlroller fails downscale.
2. Loss of flow indication.
3. Receive annunciatcr />'3-6-5.
1. Loss of E51-V8 and E51-V9 indicatien.
1. Loss of pressure transmitters/meters ReOl, R602, R60S. R804 on the R
1. Fai.!closed.
2. Loss of indication.
1. Fail closed.
2. Loss of indication.
1. Loss of speed control.
2. Loss of speed indication on RTG8
1. RCIC 'liill noi auto iniliate. Cannot be manuati'l operated.
2. Receive annunciator />'3-14.

2,_

Min flc'II valve will not auto open.

4. Barometric condenser lIacuum tank auto level oontrol inop.

Rev. 45 56 55

~~~'-L-------------~---~~~----------------;A C3 PROGRESS ENERGY

Q 1 R NIT S:

I QUPPJ

  • NS RUM1S &

PIPING AR REIXII IY JNII &

SSIM LMiS 2 th NLSS OTHWS N0ED UEFNCE LRWIIN(S SEC D Ø1q.

. I 511 IMEiAT lfl\\

- N RA1 ICNS Af*

MJI I; L,IHFS IR)JH QN S Iy

4. A...L INSRUAENT R.CKS AFE PREIXED 2H2l.

S. Ki *FR O OG1C [NrR1QC<,

&A.L ANUNCLA0 ALAQS R[ PRE-IXED 2 W2 Pfl XX, 7.

-. OENDTS VALVE LEAKOFT WHICH WILL E NORM&L OPEN WILL 3E PIPED TO CRW, UNDER CASS 6e.

8. VENDOR FURNIS ED.

9 < )oio f,S MASIR EQUIPMN LISt NUAt*R

. SC-CS24 IS USt) fl Sl Ci filHfR IC RkI O FtC 3S? AI[)

FO NSR OWR JIY SOUCf 10 1

NiL.

U, x=:s: CLASS

=Qc)..LTY Cj.,S EE 0- ø2i9 FQF ADDI IIQNA NOTES t2 HIGH POThT UTNT CAE S NORMALY MQVEO AND ORAIN tOSF NSTAI D FOR iS1FM VENTING ii SE ECHNICAL

  • PoRr PI52 FOR APPUCADLE ASPE SCrION XI RFQUIRMN7S.

LF Categories KJA:

SG2.02.15 Tier! Group:

T3 RO Rating:

3.9 SRO Rating:

4.3 LP Obj:

CLSLPO16*15E Source:

NEW Cog Level:

HIGH Category 8:

YF UNlf &:

.), I NSmUI,lENtAl ION PENElRt'\\nONS ARE MULTI-LINES THROUGH ONl:

4. ALL INSTRUMENT RACKS M~E PREnXED H21"
5.

REFER INTERLOCK.

6. ALL ANNUNCIATOR ALARMS N~E PI~EFlXED HI2**P60,**xX".
7.

DENOTES W"LVE LEAKOFF 'NHICl-I VaLL BE NORMAllY OPEN WIll BE PIPED TO C.R.W. UNDER CLASS 160",

B. VENDOR fURN1StIED.

MA$rEI~ EQU!PMENI LIsr NUMBER, (1,2.3.M.-)

AOantONAl NOTES.

12. HtGH POiNT VENT CAP !S NORMAlt,Y R(MOVEO ANO DRAIN INSTALLED r.OR VENrtNG.

R(PORf F"OR APPLICABLE XI Categories KIA:

SG2.02.15 Tier / Group:

RORating:

3.9 SRORating

LPObj:

CLS-LP-016*15E Source:

Cog Level:

HIGH Category 8:

T3 4.3 NEW YF

97. The following conditions exist on Unit One after a transient:

Jet Pump Flow Loop A 22 Mlbs/hr Jet Pump Flow Loop B 33 Mlbs/hr Recirc Pump A Percent Speed 47%

Recirc Pump B Percent Speed 66%

Total Core Flow (UICPWTCF) 55 MIbs/hr Which one of the following identifies the Required Action lAW T.S. 3.4.1, Recirculation Loops Operating, and the bases for this action?

Recirculation (1) mismatch is exceeded requiring Recirculation Loop A to be considered out of service (2)

A (1) Loop Flow (2) to ensure that assumptions of the LOCA analysis are satisfied B. (1) Loop Flow (2) due to the inability to detect significant degradation in jet pump performance C. (1) Pump Speed (2) to ensure that assumptions of the LOCA analysis are satisfied D. (1) Pump Speed (2) due to the inability to detect significant degradation in jet pump performance

97. The following conditions exist on Unit One after a transient:

Jet Pump Flow Loop A Jet Pump Flow Loop B Recirc Pump A Percent Speed Recirc Pump B Percent Speed Total Core Flow (U1CPWTCF) 22 Mlbs/hr 33 Mlbs/hr 47%

66%

55 Mlbs/hr Which one of the following identifies the Required Action lAW T.S. 3.4.1, Recirculation Loops Operating, and the bases for this action?

Recirculation (1) mismatch is exceeded requiring Recirculation Loop A to be considered out of service (2)

A'! (1) Loop Flow (2) to ensure that assumptions of the LOCA analysis are satisfied B. (1) Loop Flow (2) due to the inability to detect significant degradation in jet pump performance C. (1) Pump Speed (2) to ensure that assumptions of the LOCA analysis are satisfied D. (1) Pump Speed (2) due to the inability to detect significant degradation in jet pump performance

Feedback K/A: SG2.02.22 Equipment Control Knowledge of limiting conditions for operations and safety limits.

(CFR: 41.5 / 43.2 I 45.2)

RO/SRO Rating: 4.0/4.7 Objective: CLSLP002*34

27. Explain why there is a limit for mismatch between total Jet Pump Loop flows
34. Given plant conditions and Technical Specifications, including the Bases, TRM, ODCM, and COLR determine the required action(s) to be taken in accordance with Technical Specifications associated with the Reactor Recirculation System. (SROISTA only)

Reference:

Unit 1 Technical Specification 3.4.1 and BASES Cog Level: High Explanation:

Two recirculation loops are normally required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied.

Jet pump loop flow mismatch should be maintained within the following limits:

- jet pump loop flows within 10% (maximum indicated difference 7.5 xl 0 lbslhr) with total core flow less than 58 x10 lbs/hr

- jet pump loop flows within 5% (maximum indicated difference 3.5 xl 06 lbslhr) with total core flow greater than or equal to 58 xl 06 lbs/hr Distractor Analysis:

Choice A: Correct Answer Choice B: Plausible because Loop flow mismatch is correct and vibrations would be a result of low or reverse flow.

Choice C: Plausible because Pump Speed used to be the indication utilized and LOCA analysis is correct.

Choice D: Plausible because Pump Speed used to be the indication utilized and vibrations would be a result of low or reverse flow.

SRO Only Basis: Application of Required Actions and Knowledge of TS Bases.

Notes Feedback KIA: SG2.02.22 Equipment Control Knowledge of limiting conditions for operations and safety limits.

(CFR: 41.5 143.2 1 45.2)

RO/SRO Rating: 4.0/4.7 Objective: CLS-LP-002*34

27. Explain why there is a limit for mismatch between total Jet Pump Loop flows
34. Given plant conditions and Technical Specifications, including the Bases, TRM, ODCM, and COLR determine the required action(s) to be taken in accordance with Technical Specifications associated with the Reactor Recirculation System. (SRO/STA only)

Reference:

Unit 1 Technical Specification 3.4.1 and BASES Cog Level: High Explanation:

Two recirculation loops are normally required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied.

Jet pump loop flow mismatch should be maintained within the following limits:

- jet pump loop flows within 10% (maximum indicated difference 7.5 x1 06 Ibs/hr) with total core flow less than 58 X102 Ibs/hr

- jet pump loop flows within 5% (maximum indicated difference 3.5 x106 Ibs/hr) with total core flow greater than or equal to 58 x106 Ibs/hr Distractor Analysis:

Choice A: Correct Answer Choice B: Plausible because Loop flow mismatch is correct and vibrations would be a result of low or reverse flow.

Choice C: Plausible because Pump Speed used to be the indication utilized and LOCA analysis is correct.

Choice D: Plausible because Pump Speed used to be the indication utilized and vibrations would be a result of low or reverse flow.

SRO Only Basis: Application of Required Actions and Knowledge of TS Bases.

Notes

Recirculation Loops Operating 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCD 3.4.1 Two recirculation loops with matched flows shall be in operaon, OR One recirculalion oop may be in operation provided the following limits are applied when the associated LCD is applicable:

a.

LCO 3.2.1.

AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR), single loop operation limits specified in the COLR; b.

LCO 3.2.2, MINIMUM CRITICAL POWER RATIO (MCPR), single loop operation limits specified in the COLR; c.

LCO 3.2.3. LINEAR HEAT GENERATION RATE (LHGR) single loop operation limits specified in the COLR: and d.

LCO 3.3.1.1, Reactor Protection System RPS) Insfrumentation, Function 2.b (Average Power Range Monitors Simulated Thermal PowerHigh), Allowable Value of Tabe 3.3.1.1-1 is reset for single loop operation.

APPLICABILITY:

MODES I and 2.

ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A.

Requirements of the LCD A.1 Satist the requirements of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> not met, the LCD.

(continued)

Siunswick Unit 1 3.4-1 Amendment No. 246 Recirculation Loops Operating 3.4."1 3.4 RE.A.CTOR COOL.A.NT SYSTEM (RCS) 3.4.. 1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched lIows shall be in operation.

One recirculation loop may be in operation provided the following limits are applied when the associated LCO is applicable:

a.

LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR), single loop operation limits specified in the COLR;

b.

LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO {MCPR}, single loop operation limits specified in the COLR;

c.

LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," single loop operation limits specified in the COLR; and

d.

LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation, Function 2.b (Average Power Range Monitors Simulated Thermal Power-High), Allowable Value of Table 3.3. '1.1-1 is resetror single loop operation.

APPLICABILITY:

MODES 1 and 2.

ACTIONS CONDITION A.

Requirements of the LCO nolmet Brunswick Unit 1 A.1 REQUIRED ACTION COMPLETION TIME SatisPj the requirements of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> the LCO.

( continued) 3.4-1 Amendment No. 246 I

Recirculation Loops Operating ACTIONS continuedi 3.4.1 COMPLETION CONDITION REQUIRED ACTION TIME 5.

Required Action and 8.1 Se n MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not met.

OR No recirculation loops in operation.

SURVEILLANCE_REQUIREMENTS SURVEILLANCE FREQUENCY SR 34.i.1

NOTE----

Not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after both recrcuation toops are in operation.

Verify recirculation loop jet pump flow mismatch with 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> both recirculation loops in operation:

10% of rated core flow when operating at

< 75% of rated core flow; and b.

5% of rated core flow when operating at 75% of rated core flow.

Snjnswjck Unit 1 3.4-2 Amendment No. 244 Recirculation Loops Operating 3A:l ACTIONS (continued)

CONDITION REQUIRED.A.CTlON B.

Required Action and B.1 Be in MODE 3.

associated Completion Time of Condition.A. not met.

No recirculation loops in operation.

SURVEILLANCE REQUIREMENTS SR 3.4:1.1 Brunswick Unit 1 SUR\\lEILLANCE

NOTE--------------------------

Not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after both recirculation lOops are in operation.

Verify recirculation loop jet pump flow mismatch with both recirculation loops in operation:

a.

5: -t 0% of rated core flow when operating at

<: 75% of rated core flow; and

b.

5: 5% of rated core flO'A' when operating at

75% of rated core flow.

3.4-2 COMPLETION TIME 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> FREQUENCY 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Amendment No. 244

Recirculation Loops Operating B 3.4.1 BASES APPLICABLE For AREVA ftiel, the COLR presents single loop operation APLHGR limits SAFETY ANALYSES in the form of a multiplier that is applied to the two loop operation (continued)

APLHGR limits.

The transient analyses of Chapter 15 of the UFSAR have also been evaluated for single recirculation loop operation. The evaluation concludes that results of the transient analyses are not significantly affected by the single recirculation loop operation. There is, however, an impact on the fuel cladding integritj SL since some of the uncertainties for the parameters used in the critical power determination are higher in single loop operation. The net result is an increase in the MCPR operating limit.

During single recircutation loop operation, modification to the Reactor Pratection System (RPS) average power range monitor APRM)

Simulated Themial PowerHigh Allowable Value is required to account for the different analyzed limits between two-recirculation drive now loop operation and operation with only one loop. The APRM channel subtracts the W value from the measured recirculation drive flow to effectivej shift the limits and uses the adjusted recirculation drive flow value to determine the APRM Simulated Themial PowerHigh Function trip setpoint.

Recirculation loops operating satisfies Criterion 2 of 10 CFR S0.36(cX2)(ii) (Ref. 4).

LCO Two recirculation loops are normally required to be in operation with their flows matched within the limits specified in SR 3.4.1,1 to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. Alternately, with only one recirculation loop in operation, modifications to the required APLHGR limits LCO 3.2.1, AVERAGE PLANAR L1NEAR HEAT GENERATION RATE (APLHGRy), MCPR limits (LCO 3.2.2.

MINIMUM CRITICAL POWER RATIO MCPRfl, LHGR limits (LCO 3.2.3, LINEAR HEAT GENERATION RATE (LHGR)), and APRM Simulated Thermal Power High Allowable Value (LCO 3.3.1.1), as applicable, must be applied to allow continued operation. The COLR defines adjustments or modifications required for the APLHGR, MCPR, and LHGR limits for the current operating cycle.

continued)

Brunswick Unit 1 6 3.4.1-3 Revision No.58 B.A.SES Recirculation Loops Operating B 3.4.1 APPLICABLE For AREVA fuel, the COLR presents single loop operation APLHGR limits SAFETY AN.A.L YSES in the form of a multiplier that is applied to the two loop operation (continued}

APLHGR limits.

lCO Brunswick Unit 1 The transient analyses of Chapter 15 of the UFSAR have also been evaluated for single recirculation loop operation. The evaluation concludes that results of the transient analyses are not significantly affected by the single recirculation loop operation. There is, however, an impact on the fuel cladding integrity SL since some of the uncertainties for the parameters used in the critical power determination are higher in single loop operation. The net result is an increase in the MCPR operating limit.

During single recirculation loop operation, modification to the Reactor Protection System (RPS) average power range monitor (APRM)

Simulated Themlal Power-High Allowable 'Value is required to account for the different analyzed limits bet'A'een two-recirculation drive tlow loop operation and operation with only one loop. The APRM channel subtracts the l!.W value from the measured recirculation drive flow to effective~1 shift the limits and uses the adjusted recirculation drive flow value to determine the APRM Simulated Themlal Power-High Function trip setpoint.

Recirculation loops operating satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii) (Ref. 4}.

Two recirculation loops are normally required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure !hat dunng a LOCA caused by a break of the piping of one recirculation loop

!he assumptions of the LOCA analysis are satisfied. Alternately, with only one recirculation loop in operation, modifications to the required APLHGR limits (LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), MCPR limits (LeO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"), LHGR limits (LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)"), and APRM Simulated Thermal Power-High Allowable Value (LCO 3.3.1.1), as applicable, must be applied to allo\\'/ continued operation. The COLR defines adjustments or modifications required for the APLHGR, MCPR, and LHGR limits for the current operating cycle.

( continued)

B 3.4.1-3 Revision No. 58

Recirculation Loops Operating 3.4.1 APPLICABILITY In MODES I and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the liniiting design basis transients and accidents are assumed to occur.

In MODES 3,4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recircuation loops are not important.

ACTIONS Al With the requirements of the LCO not met, the recirculation loops must be restored to operation with matched flows within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. A recirculation loop is considered not in operation when the pump in that loop is idle or when the mismatch between total jet pump flows of the two loops is greater than the required limits. The loop with the lower flow must be considered not in operation. Should a LOCA occur with one recirculation loop not in operation, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses. Therefore, only a limited time is allowed to restore the inoperable loop to operating status.

Alternatively, if the single loop requirements of the LCO are applied to operating limits and RPS setpoints, as applicable, operation with only one recirculation loop would satisfy the requirements of the LCO and the initial conditions of the accident sequence.

The 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable time to complete the Required Action (i.e.. reset the applicable limits or setpoints for single recirculation loop operation), and on frequent care monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected.

This Required Action does not require tripping the recirculation pump in the lowest flow loop when the mismatch between the total jet pump flaws of the two loops is greater than the required limits. However, in cases where large flow mismatches occur, low flow or reverse flow can occur in the low flaw loop jet pumps. causing vibration of the jet pumps.

If zero or reverse flaw is detected, the condition should be alleviated by changing pump speeds to re-establish forward flow.

I continued)

BASES Brunswick Unit 1 B 3.4.1-4 Revision No. SB BASES APPLICABILITY ACTIONS Brunswick Unit 1 Recirculation Loops Operating B 3.4.1 In MODES 1 and 2, requirements ior operation ofthe Reactor Coolant Recirculation System are necessary' since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur.

In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important.

With the requirements of the LCO not met, the recirculation loops must be restored to operation with matched flows within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. A recirculation loop is considered not in operation when the pump in that loop is idle or when the mismatch between total jet pump flows of the two loops is greater than the required limits. The loop with the lower flow must be considered not in operation. Should a LOCA occur with one recirculation loop no! in operation, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses. Therefore, only a limited time is allowed to restore the inoperable loop to operating status.

Altematively, if the single loop requirements of the lCO are applied to operating limits and RPS setpoints, as applicable, operation with only one recirculation loop would satisfl/ the requirements of the LCO and the initial conditions of the accident sequence.

The 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Completion Time is bae.ed on the low probability of an accident occurring during this time period, on a reae.onable time to complete the Required Action (I.e., reset the applicable limits or setpoints for single recirculation loop operation), and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickiy detected.

This Required Action does not require tripping the recirculation pump in the lowest flow loop when the mismatch between the total jet pump flows of the two loops is greater than the required limits. However, in cases where large flow mismatches occur, low flow or reverse flow can occur in the low flow loop jet pumps, causingllibralion of the jet pumps. If zero or reverse flow is detected, the condition should be alleviated by changing pump speeds to re-establish forward flow.

(continued)

B 3.4.1-4 Revision No. 58

Recirculation Loops Operating B 3.4.1 ACTIONS (continued)

With no recirculation loops in operation or the Required Action and associated Completion Time oi Condition A not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the pant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. In this condition, the recirculation loops are not required to be operating because of the reduced severity of DBAs and minimal dependence on the recirculation loop coastdown characteristics. The allowed Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This SR ensures the recirculation loops are within the allowable limits for mismatch. At low core flow (i.e.. < 76% of rated core flow), the MCPR requirements provide larger margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of early boiling transition during a LOCA is reduced. A larger flow mismatch can, therefore, be allowed when core flow is < 75% of rated core flow. The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop.

The mismatch is measured in terms of the percent of rated core flow.

If the flow mismatch exceeds the specified limits, the loop with the lower flow is considered not in operation. The SR is not required when both loops are not in operation since the mismatch limits are meaningless during single loop or natural circulation operation. The Surveillance must be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after both loops are in operation. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is consistent with the Surveillance Frequency for jet pump OPERABILITY verification and has been shown by operating experience to be adequate to detect off normal jet pump loop flows in a timely manner.

REFERENCES 1.

UFSAR. Section 5.4.1.3.

2.

UFSAR, Chapter 15.

3.

NEDC-31776P, Brunswick Steam Electric Plant Units I and 2 Single Loop Operation, February 1990.

4.

10 CFR E0.36(c)t2)(ii).

Brunswick Unit 1 B 3.4.1-6 Revision No.58 Categories KJA:

SG2.02.22 Tier / Group:

T3 RO Rating:

4.0 SRO Rating:

4.7 LP Obj:

CLSLP0O2*34 Source:

NEW Cog Level:

HIGH Category 8:

Y BASES BASES Recirculation Loops Operating B 3.4:1 ACTIONS

.!U.

(continued)

With no recirculation loops in operation or the Required Action and associated Completion Time of Condition A not met, the plant must be brought to a MODE in which the LCO does not apply_ To achieve this status, th.e plant must be brought to MODE 3 within 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s_ In this condition, the recirculation loops are not required to be operating because of the reduced severity of DBAs and minimal dependence on the recirculation loop coastdown characteristics. The allowed Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and *... 'ithout challenging piant systems.

SURVEILLANCE SR 3.4.U REQUIREMENTS This SR ensures the recirculation loops are within the allowable limits for mismatch_ At low core t/ow (Le., -< 75% of rated core flow), the MCPR requirements provide larger margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of early boiling transition during a LOCA is reduced_ A larger flow mismatch can, therefore, be aI/owed when core tlow is < 75% of rated core flow. The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop_

The mismatch is measured in terms of the percent of rated core flow. If the flow mismatch exceeds the specified limits, the loop wiltl the lower flow is considered not in operation. The SR is not required when both loops are not in operation since the mismatch limits are meaningless during single loop or natural circulation operation. The Surveillance must be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after both loops are in operation_ The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> frequency is consistent with the Surveillance Frequency for jet pump OPERA.BILITY verification and has been shown by operating experience to be adequate to detect off normal jet pump loop flows in a timely manner.

REFERENCES

1.
2.
3.

UFSAR, Section SA 1.3.

UFSAR, Chapter 15.

4.

NEDC-31776P, Brunswick Steam Electric Plan! Units 1 and 2 Single Loop Operation, February 1990.

10 CFR SD_36(c)(2)(ii)_

Brunswick Unit 1 63A1-5 Revision No. 58 Categories KIA:

RORating:

LP Obj:

Cog Level:

SG2.02.22 4_0 CLS-LP-002*34 mGH Tier / Group: T3 SRORating: 4.7 Source:

NEW Category 8:

Y

98. Which oneQf the following identifies the procedure required to control drywell pressure within PCPL-A lAW PCCP and the release rate restrictions, if any, in effect during the venting?

A SEP-01 Section 1 Venting Primary Containment irrespective of Off Site Release rate B. SEP-01, Section 2 Venting Primary Containment via the Suppression Chamber within Site Release Rate Limit C. SEP-01, Section 3, Venting Primary Containment via the Drywell within Site Release Rate Limit D. OLDMG-003, Containment Venting Under Conditions of Extreme Damage irrespective of Off Site Release Rates

98. Which one,Qf the following identifies the procedure required to control drywell pressure withinPCPL-A:IAW PCCP and the release rate restrictions, if any, in effect during the venting?

A~ SEP-01 Section 1 'j Venting Primary Containment irrespective of Off Site Release rate B. SEP-01, Section 2.,. Venting Primary Containment via the Suppression Chamber within Site Release Rate Limit C. SEP-01, Section 3, Venting Primary Containment via the Drywell within Site Release Rate Limit D. OEDMG-003, Containment Venting Under Conditions of Extreme Damage irrespective of Off Site Release Rates

Feedback K/A: SG2.03.11 Radiation Control Ability to control radiation releases.

(CFR: 41.11 /43.4/45.10)

ROISRO Rating: 3.8/4.3 Objective: CLSLP300L*08d

8. Given the Primary Containment Control Procedure and plant conditions, determine if the following actions are required:
c. Venting the primary containment while staying within radioactivity release rate limits
d. Venting the primary containment IRRESPECTIVE of radioactivity release rate limits

Reference:

001-37.8, Revision 4, Page 33, Step PC/P-18 Cog Level: High Explanation:

Action to vent the primary containment is taken before drywell pressure rises to Primary Containment Pressure Limit A to assure that the integrity of the primary containment is maintained and to prevent core damage that might be caused by the inability to vent the reactor, as necessary, to permit injection of water to cool the core. Venting of the primary containment is performed irrespective of the off-site radioactivity release rate that will occur, and defeating isolation interlocks if necessary, because the consequences of not doing so may be either severe core damage or loss of primary containment integrity and uncontrolled radioactive release much greater than might otherwise occur. Note that primary containment venting is performed only, as necessary, to restore and then maintain pressure below the limit.

Distractor Analysis:

Choice A: Correct Answer.

Choice B: Plausible because within ODCM limits is utilized during SEP-01 section 1 when venting the torus due to containment Hydrogen/Oxygen concentration concerns and before exceeding PCPL-A Choice C: Plausible because within ODCM limits is utilized during SEP-01 section 1 when venting the drywell due to containment Hydrogen/Oxygen concentration concerns and before exceeding PCPL-A.

Choice D: Plausible because irrespective is correct and after exceeding PCPL-A is wrong SRO Only Basis: Detailed knowledge of diagnostic steps and decision points in the EOPs that involve transitions to emergency contingency procedures.

Notes PRIMARY CONTAINMENT PRESSURE LIMIT-A The lesser of the pressure capability of the primary containment, pressure at which containment vent valves sized to reject all decay heat from the containment can be opened and closed, or pressure at which SRVs can be opened and will remain open (Figure 2).

DE0P-01-UG Rev. 55 Page 71 of 151 Feedback KIA: SG2.03.11 Radiation Control Ability to control radiation releases.

(CFR: 41.11/43.4/45.10)

RO/SRO Rating: 3.8/4.3 Objective: CLS-LP-300-L *08d

8. Given the Primary Containment Control Procedure and plant conditions, determine if the following actions are required:
c. Venting the primary containment while staying within radioactivity release rate limits
d. Venting the primary containment IRRESPECTIVE of radioactivity release rate limits

Reference:

001-37.8, Revision 4, Page 33, Step PC/P-18 Cog Level: High Explanation:

Action to vent the primary containment is taken before drywell pressure rises to Primary Containment Pressure Limit A to assure that the integrity of the primary containment is maintained and to prevent core damage that might be caused by the inability to vent the reactor, as necessary, to permit injection of water to cool the core. Venting of the primary containment is performed irrespective of the off-site radioactivity release rate that will occur, and defeating isolation interlocks if necessary, because the consequences of not doing so may be either severe core damage or loss of primary containment integrity and uncontrolled radioactive release much greater than might otherwise occur. Note that primary containment venting is performed only, as necessary, to restore and then maintain pressure below the limit.

Distractor Analysis:

Choice A: Correct Answer.

Choice B: Plausible because within ODCM limits is utilized during SEP-01 section 1 when venting the torus due to containment Hydrogen/Oxygen concentration concerns and before exceeding PCPL-A Choice C: Plausible because within ODCM limits is utilized during SEP-01 section 1 when venting the drywell due to containment Hydrogen/Oxygen concentration concerns and before exceeding PCPL-A.

Choice D: Plausible because irrespective is correct and after exceeding PCPL-A is wrong SRO Only Basis: Detailed knowledge of diagnostic steps and decision points in the EOPs that involve transitions to emergency contingency procedures.

Notes PRIMARY CONTAINMENT PRESSURE LlMIT-A The lesser of the pressure capability of the primary containment, pressure at which containment vent valves sized to reject all decay heat from the containment can be opened and closed, or pressure at which SRVs can be opened and will remain open (Figure 2).

IOEOP-01-UG Rev. 55 Page 71 of 151 I

-rT C

Cl) z H

C) 0

-1

C

-4 m

Inl cci

-o Cl)cl) 0 o

mmo

-w

HH r

ma m

t;x

)OO

-tJ-U-o

.- C.,

DRYWELL PRESSURE (PSIG)

N)

W ai 0)

V H

Hi ii.

iii III.11 H

I m0

-o C)

C,,

0,

-v 1

-b 01

-s

= i.

G)CDQ D C -.x 4

.mgm k) N) Z

-H 01 I

(!)

(j)

Il. -

W a:

I CJ)

U) w t:t:

a.

...I

..J W

~

il2 C

100 1'10 80 70 60 50 40 30 20 10 0

ATTACHMENT 5 Page 17 of 27 FIGURE 2 Primary Containment Pressure Limit-A

  • 10 0

10 20 30 40 50 GO 70 8Q PRIMARY CONTAINMENT WATER LEVEL (FEET)

IF USING THE FOLLOWING INSTRUMENT:

CAC-PI-1230 CAC-PI-4176 CAC-PR-1257-1 PCPl-A IS:

70 PSIG USE THE GRAPH USE THE GR.f\\PH IOEOP-01-UG Rev. 55 Page 77 of 151 I

PCPL-A j DRYWELL PHESS PCPL.A VENT ThE PRMARV CTT RRSPCTIV& OF OFcsIrE NZUASE RAXF

$ETtON I OF pcjp. i STEP BASES:

Action to vent the primary containment is taken before drywell pressure rises to Primary Containment Pressure Limit A, defined to be the lesser of either:

a.

The pressure capability of the containment, or b.

The maximum containment pressure at which vent valves sized to reject all decay heat from the containment can be opened and closed, or C.

The maximum containment pressure at which SRVs can be opened and will remain open, or d.

The maximum containment pressure at which reactor vent valves can be opened and closed.

This action is taken to assure that the integrity of the primary containment is maintained and to prevent core damage that might be caused by the inability to vent the reactor, as necessary, to permit injection of water to cool the core.

The directions Lo vent before drywell pressure reaches PCPL-A allows, but does not require, venting at significantly lower pressures Early or extended venting can permit primary containment pressure reductions before significant fuel damage occurs, thereby increasing the capacity of the containment to retain lission products and reducing the radioactivity released to the environment, If the primary containment has failed, venting may also reduce the offsite dose by directing fission products through an elevated release point.

001-37.8 Rev.

I Page 33 of 58 PCPL-A STEP BASES:

rmVWllll. PRESS R~AClfeS PCPL*A VENT THE PRIMARY elMT IRRESPECTIVE OF OFFSITE "IELEASe RAtE PER SECTION 1 OF 1001"*01* SEP. 01 "CiP*HI Action to vent the primary containment is taken before drywell pressure rises to Primary Containment Pressure Limit A,defined to be tile lesser of either:

a.

The pressure capability of the containment, or

b.

The maximum containment pressure at whicll vent valves sized to reject all decay heat from tile containment can be opened and closed, or

c.

The maximum containment pressure at which SRVs can be opened and will remain open, or

d.

The maximum containment pressure at which reactor vent valves can be opened and closed.

This action is taken to assure that the integrity of the primary containment is maintained and to prevent core damage that might be caused by tile inability to vent the reactor, as necessary, to permit injection of water to cool tile core.

The directions to vent "before drywell pressure reaches pePl-A" allows, but does not require, venting at significantly lower pressures. Early or extended venting can permit primary containment pressure reductions before significant fuel damage occllrs, thereby increasing the capacity of the containment to retain fission products and reducing the radioactivity released to the environment. If the primary containment has failed, venting may also reduce the offsile dose by directing fission prodllcts throllgh an elevated release point.

1001-37.8 Rev. 4 Page 33 of 581

STEP PCIP.18 (continued)

Venting of the primary containment is performed irrespective of the off-site radioactivity release rate that will occur, and defeating isolation interlocks if necessary, because the consequences of not doing so may be either severe core damage or loss of primary containment integrity and uncontrolled radioactive release much greater than might otherwise occur. Note that primary containment venting is performed only, as necessary, to restore and then maintain pressure below the limiL Primary containment venting is performed using Primary Containment Venting, EDP-O1 -SEP-01.

001-37.8 Rev. 4 Pane 34 of 58 STEP PC/P*18 (continued) venting of the primary containment is petiormed irrespective of the off-site radioactivity release rate that will occur, and defeating isolation interlocks if necessary, because the consequences of not doing so may be either severe core damage or loss of primary containment integrity and uncontrolled radioactive release much greater than might otherwise occur. Note that primary containment venting is petiormed only, as necessary, to restore and then maintain pressure below the limit Primary containment venting is performed using Primary Containment Venting, EOP-01-SEP-0'I.

1001-37.8 Rev. 4 Page 34 of 581

PRIMARY CONTAINMENT VENTING 1.0 ENTRY CONDITIONS As dhrected by the PC/P section of Primary Contaniment Control Procedure, EOP-O2-PCCP OR As directed by the PC/H section of Primary Containment Control Procedure, EOP-02-PCCP 2.0 OPERATOR ACTIONS CO:

2.1 IF while executing this procedure, it is recognized the actions can NOT be performed, OR will NOT be effective, THEN GO TO Containment Venting Under Conditions of Extreme Damage, OEDMG-003, if directed by the Unit SCO.

CO:

2.2 IF venting for pressure control, THEN PERFORM Section 1, on page 3.

CO:

2.3 IF venting for H2/02 control, THEN PERFORM section of procedure directed by SCO.

I OEOP-Oi-SEP-Oi Rev. 24 Page 2 of 22 Categories KJA:

SG2.03.I 1 Tier / Group:

T3 RU Rating:

3.8 SRU Rating:

4.3 LP Ubj:

CLSLP3OOL*O8D Source:

NEW Cog Level:

HIGH Category 8:

PRIMARY CONTAINMENT VENTING 1.0 ENTRY CONDITIONS As directed by the PC/P section of Primary Containment Control Procedure, EOP-02-PCCP OR As directed by the PC/H section of Primary Containment Control Procedure, EOP-02-PCCP 2.0 OPERA TOR ACTIONS CO:

2.1 IF while executing this procedure, it is recognized the actions can NOT be performed, OR will NOT be effective, THEN GO TO Containment Venting Under Conditions of Extreme Damage, OEDMG-003, if dlrected by the Unit SCO.

CO:

2.2 IF venting for pressure control, THEN PERFORM Section oJ, on page 3.

CO:

2.3 IF venting for H2/02 control.. THEN PERFORM section of procedure directed by SCO.

IOEOP-ool-SEP-O'l Rev. 24 Categories KIA:

SG2.03.11 RO Rating:

3.8 LP Obj:

CLS-LP-300-L*08D Cog Level:

HIGH Tier / Group: T3 SRO Rating:

4.3 Source

~vv Category 8:

0 0

0 Page 2 of 22 1

99. During non-ATWS emergency conditions on Unit Two, Emergency Depressurization is required with reactor pressure at 1100 psig.

Which one of the following identifies the bases for the Minimum Number of SRV5 Required for Emergency Depressurization and the required procedure utilized if this number of SRVs open cannot be achieved?

The Minimum Number of SRVs Required for Emergency Depressurization is based on the low pressure ECCS system with the lowest head being capable of making up the SRV steam flow at the Minimum (1)

(2)

Procedure is required if the minimum number of SRVs cannot be opened.

A.

(1) Reactor Flooding Pressure (2) Primary Containment Flooding B. (1) Reactor Flooding Pressure (2) Alternate Emergency Depressurization C. (1) Alternate Reactor Flooding Pressure (2) Primary Containment Flooding D (1) Alternate Reactor Flooding Pressure (2) Alternate Emergency Depressurization

99. During non-ATWS emergency conditions on Unit Two, Emergency Depressurization is required with reactor pressure at 1100 psig.

Which one of the following identifies the bases for the Minimum Number of SRVs Required for Emergency Depressurization and the required procedure utilized if this number of SRVs open cannot be achieved?

The Minimum Number of SRVs Required for Emergency Depressurization is based on the low pressure ECCS system with the lowest head being capable of making up the SRV steam flow at the Minimum (1)

(2)

Procedure is required if the minimum number of SRVs cannot be opened.

A. (1) Reactor Flooding Pressure (2) Primary Containment Flooding B. (1) Reactor Flooding Pressure (2) Alternate Emergency Depressurization C. (1) Alternate Reactor Flooding Pressure (2) Primary Containment Flooding D:' (1) Alternate Reactor Flooding Pressure (2) Alternate Emergency Depressurization

Feedback KJA: SG2.04.17 Emergency Procedures I Plan Knowledge of EOP terms and definitions.

(CFR: 41.10 /45.13)

ROISRO Rating: 3.9/4.3 Objective: CLSLP300H*002

2. Given plant conditions and the Emergency Operating Procedures, determine if execution of the Alternate Emergency Depressurization Procedure is required.

Reference:

OEOP-01-UG, Revision 55, Page 70, Attachment 5 (Definitions)

RVCP Cog Level: High Explanation:

The Minimum Number of SRVs Required for Emergency Depressurization (5) is defined to be the least number of SRVs which correspond to a Minimum Alternate Reactor Flooding Pressure sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow at the corresponding Minimum Alternate Reactor Flooding Pressure.

If the number of SRVs specified cannot be opened, the reactor must be depressurized by other means. A list of alternate systems that can be used for depressurizing the reactor is included in the Alternate Emergency Depressurization Procedure, EOP-01 -AEDP.

Distractor Analysis:

Choice A: Plausible because Minimum Reactor Flooding Pressure is easily confused with Minimum Alternate Reactor Flooding Pressure and Primary Containment Flooding requires exiting all EOPs which is wrong for the given conditions.

Choice B: Plausible because Minimum Reactor Flooding Pressure is easily confused with Minimum Alternate Reactor Flooding Pressure and AEDP is correct.

Choice C: Plausible because Minimum Alternate Reactor Flooding Pressure is correct and Primary Containment Flooding requires exiting all EOPs which is wrong for the given conditions.

Choice D: Correct Answer SRO Only Basis: Detailed knowledge of diagnostic steps and decision points in the EOPs that involve transitions to emergency contingency procedures.

Notes MINIMUM ALTERNATE FLOODING PRESSURE The lowest reactor pressure at which steam 110w through open SRVs is sufficient to preclude any clad temperature from exceeding 1500F even if the reactor core is not completely covered OEOP-Oi-UG Rev. 55 Page 69 of 151 Feedback KJA: SG2.04.17 Emergency Procedures I Plan Knowledge of EOP terms and definitions.

(CFR: 41.10/45.13)

RO/SRO Rating: 3.9/4.3 Objective: CLS-LP-300-H*002

2. Given plant conditions and the Emergency Operating Procedures, determine if execution of the Alternate Emergency Depressurization Procedure is required.

Reference:

OEOP-01-UG, Revision 55, Page 70, Attachment 5 (Definitions)

RVCP Cog Level: High Explanation:

The Minimum Number of SRVs Required for Emergency Depressurization (5) is defined to be the least number of SRVs which correspond to a Minimum Alternate Reactor Flooding Pressure sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow at the corresponding Minimum Alternate Reactor Flooding Pressure. If the number of SRVs specified cannot be opened, the reactor must be depressurized by other means. A list of alternate systems that can be used for depressurizing the reactor is included in the Alternate Emergency Depressurization Procedure, EOP-01-AEDP.

Distractor Analysis:

Choice A: Plausible because Minimum Reactor Flooding Pressure is easily confused with Minimum Alternate Reactor Flooding Pressure and Primary Containment Flooding requires exiting all EOPs which is wrong for the given conditions.

Choice B: Plausible because Minimum Reactor Flooding Pressure is easily confused with Minimum Alternate Reactor Flooding Pressure and AEDP is correct.

Choice C: Plausible because Minimum Alternate Reactor Flooding Pressure is correct and Primary Containment Flooding requires exiting all EOPs which is wrong for the given conditions.

Choice D: Correct Answer SRO Only Basis: Detailed knowledge of diagnostic steps and decision points in the EOPs that involve transitions to emergency contingency procedures.

Notes MINIMUM ALTERNATE FLOODING PRESSURE The lowest reactor pressure at which steam flow through open SRVs is sufficient to preclude any clad temperature from exceeding 1500°F even if the reactor core is not completely covered I

OEOP-O*!-UG Rev. 55 Page 69 of *15"1 I

ATTACHMENT 5 Page 10 of 27 Definitions MINI MUM CORE FLOODING INTERVAL The greatest amount of time required to flood the reactor to the top of the active Tuel with reactor pressure at the minimum reactor flooding pressure and at least the minimum number of SRVs required for emergency depressurization open.

MINIMUM INDICATED LEVEL The highest reactor water level instrument indication which results from off-calibration instrument run temperature conditions when reactor water level is actually at the elevation of the instrument variable leg tap.

M1NIMUM NUMBER OF SRVS REQUIRED FOR EMERGENCY DEPRESSURIZATION The least number of SRVs which correspond to a minimum alternate reactor flooding pressure sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow at the corresponding minimum alternate reactor flooding pressure.

MINIMUM REACTOR FLOODING PRESSURE The minimum SRV reopening pressure; 50 psid with 5 SRVs open. This pressure is utilized to assure sufficient liquid injection into the reactor to maintain SRVs open and to flood the reactor to the elevation of the main steam lines during the flooding evolution when the reactor is shutdown.

MINIMUM SRV REOPENING PRESSURE The lowest reactor pressure at which an SRV will lufly open and remain fully opened when its control switch is placed in the OPEN position.

MINIMUM STEAM COOLING REACTOR WATER LEVEL The lowest reactor water level at which the covered portion of the reactor core will generate sufficient steam to preclude any clad temperature in the uncovered portion of the core from exceeding 150DF. This limit is used during an ATWS event to prevent fuel damage when level is lowered below TAF (Unit 1 only:

Figure 18; Unit 2 only: Figure 18A).

OEOP-01-UG Rev. 55 Page 70 of 151 ATTACHMENT 5 Page 10 of 27 Definitions MINIMUM CORE FLOODING INTERVAL The greatest amount of time required to flood the reactor to the top of the active fuel with reactor pressure at the minimum reactor flooding pressure and at least the minimum number of SRVs required for emergency depressurization open.

MINIMUM INDICATED LEVEL The highest reactor water level instrument indication which results from off-calibration instrument run temperature conditions when reactor water level is actually at the elevation of the instrument variable leg tap.

MINIMUM NUMBER OF SRVS REQUIRED FOR EMERGENCY DEPRESSURIZATION The least number of SRVs which correspond to a minimum altemate reactor flooding pressure sufficiently low that tile ECCS witll the 101,vest head will be capable of making up the SRV steam flow at the corresponding minimum alternate reactor flooding pressure.

MINIMUM REACTOR FLOODING PRESSURE The minimum SRV reopening pressure; 50 psid witl) 5 SRVs open. This pressure is utilized to assure sufficient liquid injection into the reactor to maintain SRVs open and to flood the reactor to the elevation of the main steam lines during the flooding evolution when the reactor is shutdown.

MINIMUM SRV REOPENING PRESSURE The lowest reactor pressure at which an SRV will fully open and remain fully opened when its control switch is placed in the OPEN position.

MINIMUM STEAM COOLING REACTOR WATER LEVEL The lowest reactor water level at which the covered portion of the reactor core will generate sufficient steam to preclude any clad temperature in the uncovered portion of the core from exceeding 1500°F. This limit is used during an ATWS event to prevent fuel damage when level is lowered below T AF (Unit 'I only:

Figure 18; Unit 2 only: Figure'18A).

I OEOP-O'l-UG Rev. 55 Page 70 of '15'1 I

STEPS RCIP-23 through RC/P-25 The Minimum Number of SRVs Required for Emergency Depressurizatiori (5) is defined to be the least number of SRVs which correspond to a Minimum Alternate Reactor Flooding Pressure sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow at the corresponding Minimum Alternate Reactor Flooding Pressure.

The Minimum SRV Re-opening Pressure is the lowest reactor pressure at which an SRV will remain fully open with its control switch in the open position. The accuracy of the re-opening pressure and the indication available to determine reactor pressure result in conditions such that the SRVs are not always open when the pressure indicated is 50 psig. One hundred psig has been selected as a value which can be used to determine the SRV5 have failed to function. When reactor pressure is below this value, depressurization is considered complete and reactor pressure reduction need not be augmented by use of additional systems even if less than the minimum number of SRVs are open.

If the number of SRVs specified cannot be opened, the reactor must be depressurizeci by other means. A list of alternate systems that can be used for depressurizing the reactor is included in the Alternate Emergency Depressurization Procedure, EOP-O1-AEDP. However, since event independence must be maintained and specific plant conditions cannot be presumed, no priority regarding system use is indicated. This approach provides an operator the flexibility of being able to use whatever system(s) may be most appropriate under current plant conditions.

001-37.4 Rev. 8 Page 59 of 78 PERFORM ALTERNATE EFi.ERGE.NCY DEPRESSURI2.ATION PROCEDURE (EOP. 01. AEDP IRRSPECTNE OI OFFSITE RADIOACTIVITY RELEASE RATE 1

STEP BASES:

STEPS RC/P*23 through RC/P*25 PERFORM "ALTERNATE EMERGENCY DEPRESSURIZATION PROCEDURE" (EOP. 01. AEDPI IRRESPECTIVE OF OPPSIlE RADIOACTIVITY REll:.... SE RAT!:

RCIP-ZS STEP BASES:

The Minimum Number of SRVs Required for Emergency Depressurization (5) is defined to be the least number of SRVs which correspond to a Minimum Alternate Reactor Flooding Pressure sufficiently low that the ECCS with the lowest !lead will be capable of making up the SRV steam flow at the corresponding Minimum Alternate Reactor Flooding Pressure.

The Minimum SRV Re-opening Pressure is the lowest reactor pressure at which an SRV will remain fully open with its control switch in the open position. The accuracy of the re-opening pressure and the indication available to detelmine reactor pressure result in conditions such that the SRVs are not always open when the pressure indicated is 50 psig. One hundred psig has been selected as a value which can be used to determine the SRVs have failed to function. When reactor pressure is below this value, depressurization is considered complete and reactor pressure reduction need not be augmented by use of additional systems even if less than the minimum number of SRVs are open. If the number of SRVs specified cannot be opened, the reactor must be depressurized by other means. A list of alternate systems that can be used for depressurizing the reactor is included in the Alternate Emergency Depressurization Procedure, EOP-01-AEDP. However, since event independence must be maintained and specific plant conditions cannot be presumed, no priority regarding system use is indicated. This approach provides an operator the flexibility of being able to use whatever system(s) may be most appropriate under current plant conditions.

1001-37.4 Rev. 8 Page 59 of 781

ALTERNATE EMERGENCY DEPRESSURIZATION PROCEDURE 1.0 ENTRY CONDITIONS As directed by the RCIP section of Reactor Vessel Control Procedure, EOP-Ol-RVCP OR As directed by the RC/P section of Level/Power Control, EOP-O1-LPC OR As directed by SAMG Primary Containment Flooding, SAMG-Ol 2.0 OPERATOR ACTIONS NOTE:

Manpower:

I Control Operator I Auxiliary Operator I Independent Verifier Special equipment:

4 jumpers (32, 33, 34, and 35:

1 Ilathead screwdriver I locking screwdriver tape NOTE:

Performance of this procedure will affect any main steam line leakage control pathways established by EOP-Oi -SEP-i t 2.1 EVACUATE the Unit 1 and 2 Turbine Buildings using the following actions:

CO:

2.11 SOUND the Unit 1 and Unit 2 Turbine Building evacuation alarms AND ANNOUNCE the evacuation.

CO:

2.1.2 REQUEST the SCO to notify the TSC that the Turbine Building is being evacuated due to potential high radiation conditions during the alternate emergency clepressurization.

CO:

2.2 IF either Unit 1 or Unit 2 Turbine Building ventilation is in service in the once-through lineup, THEN SECURE that units turbine building ventilation (OP-37.3).

OEOP-Oi-AEDP Rev. 18 Page 2 of 16 Categories K/A:

SG2.04.17 Tier/Group:

T3 RO Rating:

3.9 SRO Rating:

4.3 LP Obj:

CLSLP3OOH*OO2 Source:

NEW Cog Level:

HIGH Category 8:

Y ALTERNATE EMERGENCY DEPRESSURIZATION PROCEDURE 1.0 ENTRY CONDITIONS As directed by the RC/P section of Reactor Vessel Control Procedure, EOP-01-RVCP OR As directed by the RC/P section of Level/Power Control, EOP-01-LPC OR As directed by SAMG Primary Containment Flooding, SAMG-O'l 2.0 OPERATOR ACTIONS NOTE:

Manpower:

Special eqUipment:

'1 Control Operator

'1 Auxiliary Operator

'I Independent Verifier 4 jumpers (32, 33, 34, and 35}

1 flathead screwdriver 1 locking screwdriver tape NOTE:

Performance of this procedure will affect any main steam line leakage control pathways established by EOP-O'l-SEP-'!'L 2,1 EVACUATE tile Unit 1 and 2 Turbine Buildings using the following actions:

CO:

CO:

CO:

2.2 2.1:1 2.1.2 SOUND the Unit 1 and Unit 2 Turbine Building evacuation alarms AND ANNOUNCE the evacuation.

REQUEST the SCO to notify the TSC that the Turbine Building is being evacuated due to potential high radiation conditions during the alternate emergency depressurization.

IF either Unit 1 or Unit 2 Turbine Building ventilation is in service in the once-througlliineup, THEN SECURE that units' turbine building ventilation (OP-37.3).

I OEOP-O'I-AEDP Rev. '18 Categories KIA:

SG2.04.17 RO Rating:

3.9 LP Obj:

CLS-LP-300-H*002 Cog Level:

mGH Tier / Group: T3 SRO Rating:

4.3 Source

~VV Category 8:

Y o

o o

Page 2 of 161

100. An ATWS has occurred on Unit Two:

ARI has been actuated.

No blue lights are lit on the Full Core Display.

Suppression Pool Temperature is 112° F.

The 2A SLC pump has a red light indication.

The 2B SLC pump has a green light indication The SLC A Squib Valve Continuity white light is lit The SLC B Squib Valve Continuity white light is extinguished.

Which one of the following identifies the procedure that an AC would be directed to perform based on the above conditions and the resultant effect of those actions?

A Perform LEP-02, Section 2 to insert control rods in order to shutdown the reactor by venting the Scram Air Header.

B. Perform LEP-02, Section 6 to insert control rods in order to shutdown the reactor by venting the overpiston area of the control rods.

C. Perform LEP-03, Section 2 to inject boron to shutdown the reactor using RCIC.

D. Perform LEP03, Section 3 to inject boron to shutdown the reactor using RWCU via the SLC tank.

100. An ATWS has occurred on Unit Two:

ARI has been actuated.

No blue lights are lit on the Full Core Display.

Suppression Pool Temperature is 112 0 F.

The 2A SLC pump has a red light indication.

The 2B SLC pump has a green light indication The SLC A Squib Valve Continuity white light is lit The SLC B Squib Valve Continuity white light is extinguished.

Which one of the following identifies the procedure that an AO would be directed to perform based on the above conditions and the resultant effect of those actions?

A'! Perform LEP-02, Section 2 to insert control rods in order to shutdown the reactor by venting the Scram Air Header.

B. Perform LEP-02, Section 6 to insert control rods in order to shutdown the reactor by venting the overpiston area of the control rods.

C. Perform LEP-03, Section 2 to inject boron to shutdown the reactor using RCIC.

D. Perform LEP~03, Section 3 to inject boron to shutdown the reactor using RWCU via the SLC tank.

Feedback K/A: SG2.04.35 Emergency Procedures I Plan Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects (CFR: 41.10 /43.5/45.13)

RO/SRO Rating: 3.8/4.0 Objective: CLSLP300J*005 5.

Given plant conditions and the Local Emergency Procedures, determine which sections of the Alternate Control Rod Insertion Procedure should be utilized for Control Rod Insertion (EOP-01 -LEP-02).

4. Given plant conditions and the Local Emergency Procedures, determine which method of the Alternate Boron Injection is appropriate (EOP-01-LEP-03)

Reference:

OEOP-01 -LEP-02 Cog Level: High Explanation:

Based on the conditions given, determines that scram valves have not opened (no blue lights on full core display) and that Boron is injecting with A pump running (red light on) and B squib valve opened (white light extinguished) so LEP-03 is not required. The pumps discharge into a c9rflmoIheader before going to the squib valves. Requires assessment of alternate control rod insertion seOtions anddetermines venting the scram air header is appropriate.

Distractor Analysis:

Choice A: Correct Answer Choice B: Plausible because venting of the over piston area will insert the control rods but would be the inappropriate decision for rod insertion given the conditins. The operational effect is reactor shutdown with control rod insertion.

Choice C: Plausible because suppression pool temperature is greater than 1100 F and boron injection is required. With A pump running but the A squib valve not open and no B pump a common misconception is that SLC flow will not occur to the Reactor. this would be correct under different conditions in the stem. The operational effect is reactor shutdown with boron injection.

Choice D: Plausible because suppression pool temperature is greater than 1100 F and boron injection is required. With A pump running but the A squib valve not open and no B pump a common misconception is that SLC flow will not occur to the Reactor. this would be correct under different conditions in the stem. The operational effect is reactor shutdown with boron injection.

SRO Only Basis: Assessing plant conditions and prescribing a section of a procedure with which to proceed.

Notes Feedback KIA: SG2.04.35 Emergency Procedures I Plan Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.

(CFR: 41.10/43.5/45.13)

RO/SRO Rating: 3.8/4.0 Objective: CLS-LP-300-J*005

5. Given plant conditions and the Local Emergency Procedures, determine which sections of the Alternate Control Rod Insertion Procedure should be utilized for Control Rod Insertion (EOP-01-LEP-02).
4. Given plant conditions and the Local Emergency Procedures, determine which method of the Alternate Boron Injection is appropriate (EOP-01-LEP-03)

Reference:

OEOP-01-LEP-02 Cog Level: High Explanation:

Based on the conditions given, determines that scram valves have not opened (no blue lights on full core display) and that Boron is injecting with A pump running (red light on) and B squib valve opened (white light extinguished) so LEP-03 is not required. The pumps discharge into a c9ffi..rnQ!lJl~~der before going to the squib valves. Requires assessment of alternate control rod insertion seCtions and"determines venting the scram air header is appropriate.

Distractor Analysis:

Choice A: Correct Answer Choice B: Plausible because venting of the over piston area will insert the control rods but would be the inappropriate decision for rod insertion given the conditins. The operational effect is reactor shutdown with control rod insertion.

Choice C: Plausible because suppression pool temperature is greater than 110° F and boron injection is required. With A pump running but the A squib valve not open and no B pump a common misconception is that SLC flow will not occur to the Reactor. this would be correct under different conditions in the stem. The operational effect is reactor shutdown with boron injection.

Choice D: Plausible because suppression pool temperature is greater than 110° F and boron injection is required. With A pump running but the A squib valve not open and no B pump a common misconception is that SLC flow will not occur to the Reactor. this would be correct under different conditions in the stem. The operational effect is reactor shutdown with boron injection.

SRO Only Basis: Assessing plant conditions and prescribing a section of a procedure with which to proceed.

Notes

21 INSERT control rods by one or more of the following methods:

2.71 DE-ENERGIZE the scram pilot valve solenoids AND VENT the scram air header Section 2 on Page 9 2.7.2 RESET RPS AND INITIATE a manual scram. Section 3 on Page 14.

2.7.3 SCRAM indMdual rods with the scram test switches, Section 4 on Page 17.

2.7.4 INSERT control rods with the Reactor Manual Control LI System, Section 5 on Page 21.

2.7.5 VENT the over piston area of control rods, Section 6 on Page 22.

OEOP-O1-LEP-02 Rev. 26 Page 3 of 29 2.2 INJECT boron with one or more of the following systems:

ENOTE:

System(s) should be selected in order listed and based upon system II availability and accessibility.

CO:

CRD, Section 1 on page 3 LI NOTE:

HPCI/RCIC should be used only if suction is from the CST.

CO:

HPCl/RCICSection2onpage14 LI CO:

RWCU via SLC tank, Section 3 on page 21 LI CO:

RWCU with borax, Section 4 on page 31 LI OEOR-Di-LEP-03 Rev. 27 Page 2 of 41 2.7 INSERT control rods by one or more of the following methods:

2.7.1 2.7.2 2.7.3 2.7.4 2.7.5 DE-ENERGIZE the scram pilot valve solenoids AND VENT the scram air header, Section 2 on Page 9.

RESET RPS AND INITIATE a manual scram, Section 3 on Page '14.

SCRAM individual rods with the scram test switches, Section 4 on Page H.

INSERT control rods with the Reactor Manual Control System, Section 5 on Page 21.

VENT the over piston area of control rods, Section 6 on Page 22.

I OEOP-01-LEP-02 Rev. 26 2.2 INJECT boron with one or more of the following systems:

o o

o o

o Page 3 of 29 I NOTE:

System(s) should be selected in order listed and based upon system availability and accessibility.

CO:

CRD, Section 1 on page 3 NOTE:

HPCIIRCIC should be used only jf suction is from the CST.

CO:

HPCIIRCIC, Section 2 on page 14 CO:

RWCU via SLC tank, Section 3 on page 21 CO:

RWCU with borax, Section 4 on page 31 I OEOP-01-LEP-03 Rev. 27 o

o o

o Page 2 of 4-1 I

K/A:

SG2.04.35 RORating:

3.8 LP Obj:

CLSLP3OOJ*OO5 Cog Level:

NIGH Tier / Group:

T3 SRO Rating:

4.0 Source

NEW Category 8:

Categories

~026 Categories KIA:

RORating:

LPObj:

Cog Level:

SQUIB F004A SQUIB F004B SG2.04.35 3.8 CLS-LP-300-J*005 HIGH Tier / Group: T3 SRORating:

4.0 Source

NEW Category 8:

f002A F002B PUMP COO1A PUMP C001B

Retrieved from "https://kanterella.com/w/index.php?title=ML101590078&oldid=5297989"