ML19283D002
| ML19283D002 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 10/03/2019 |
| From: | Thomas Setzer Operations Branch I |
| To: | Exelon Generation Co |
| Shared Package | |
| ML18347A730 | List: |
| References | |
| EPID L-2019-OLL-0039 | |
| Download: ML19283D002 (22) | |
Text
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
R.E. Ginna Date of Examination:
07/24/19 Examination Level: RO
~ SRO D Operating Test Number:
N2019-301 R Administrative Topic (see Note)
Type Describe activity to be performed Code*
Conduct of Operations M,S Ability to make accurate, clear, and concise KIA-2.1.18 logs, records, status boards, and reports.
3.6 Perform a Daily Surveillance Log Conduct of Operations M,R Knowledge of procedures, guidelines, or KIA-2.1.37 limitations associated with reactivity 4.3 management.
Calculate SDM for an Operating Reactor with a Misaligned Control Rod Equipment Control D,R Ability to recognize system parameters that are KIA-2.2.42 entry-level conditions for Technical 3.9 Specifications.
HCO Review of STP-0-36QC Radiation Control D,R Ability to control radiation releases.
KIA-2.3.11 Determine Maximum Reactor Vessel Venting 3.8 Time Emergency Plan NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).
- Type Codes and Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (S 3 for ROs; s 4 for SROs and RO retakes)
(N)ew or (M)odified from bank (~ 1)
(P)revious 2 exams (S 1, randomly selected)
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
RE. Ginna Date of Examination:
07/24/19 Examination Level: RO D SRO
[Zl Operating Test Number:
N2019-301 R Administrative Topic (see Note)
Type Describe activity to be performed Code*
Conduct of Operations D,R Ability to interpret reference materials, such as KIA-2.1.25 graphs, curves, tables, etc.
4.2 Perform a Critical Rod Position Calculation in accordance with 0-1.2.2 Conduct of Operations N,R Ability to explain and apply system limits and K/A-2.1.32 precautions.
4.0 Determine Operating Limits for Station 13A Transmission in accordance with 0-6.9 Equipment Control M,R Ability to apply Technical Specifications for a K/A-2.2.40 system.
4.7 Determine limitations in accordance with A-52.12, Nonfunctional Equipment Important to Safety Radiation Control N,R Ability to approve release permits.
K/A-2.3.6 Review and Approve Gas Decay Tank Release 3.8 Permit Emergency Plan M,R Knowledge of the emergency action level 2.4.41 thresholds and classifications.
4.6 Determine Protective Action Recommendations in accordance with EP-CE-111 NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).
- Type Codes and Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (s 3 for ROs; S 4 for SROs and RO retakes)
(N)ew or (M)odified from bank (~ 1)
(P)revious 2 exams (S 1, randomly selected)
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:
RE. Ginna Date of Examination:
07/24/19 Exam Level: RO
~ SRO-I D SRO-U D Operating Test Number:
N2019-301 R Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*
Safety Function
- a. 006 Emergency Core Cooling System (ECCS) [006 A4.01 M,S,A,EN 2
(4.1/3.9)] Establish RCS Injection in AP-RCS.4 with Cl Valve Failures
- b. 062 AC Electrical Distribution System [062 A4.01 (3.3/3.1 )]
M,S,A 6
Transfer 4160V Auxiliary Loads and Take Actions for Loss of Bus
- c. 012 Reactor Protection System (RPS) [012 A4.04 (3.3*/3.3)]
D,S 7
Defeat Failed RCS Temperature Channel
- d. 010 Pressurizer Pressure Control System (PZR PCS) [010 A4.03 D,S,L 3
(4.0/3.8)] Placing L TOP in Service
- e. EPE W/E06 Degraded Core Cooling (EPE W/E06 EA2.2 (3.5/4.1)]
M,S,A,EN 4P Vent RCS for Accumulator/RHR Injection
- f. 045 Main Turbine Generator (MT/G) System [045 A4.01 (3.1/2.9)]
M,S,A 4S Perform Intercept and Reheat Stop Valve Test with Low EH System Pressure
- g. 026 Containment Spray System (CSS) [026 A2.08 (3.2/3.7)]
D,S,EN 5
Secure Containment Spray in E-1
- h. APE 026 Loss of Component Cooling Water (CCW) [APE 026 N,S,A 8
AA1.02 (3.2/3.3)] Respond to Complete Loss of CCW Flow In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U
- i. 039 Main and Reheat Steam System (MRSS) [039 A3.02(3.1/3.5*)]
D, E 4S Locally Close MSIVs
- j. 033 Spent Fuel Pool Cooling System (SFPCS) [033 G2.1.29 D,R 8
(4.1/4.0)] Alternate SFP Cooling Systems (A to B)
Isolate CI/CVI Valves All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.
ES 301 C
t IR on ro
/I Pl t S t
oom n-an
- ys ems 0 tr u me F orm ES 301 2
- Type Codes Criteria for R /SR0-1/SRO-U (A)lternate path 4-6/4-6 /2-3 (C)ontrol room (D)irect from bank
- S 9/S 8/S 4 (E)mergency or abnormal in-plant
~ 1/<?: 1/<?: 1 (EN)gineered safety feature
~ 1/<?: 1/<?: 1 (control room system)
(L )ow-Power/Shutdown
~1/<?:1/<?:1 (N)ew or (M}odified from bank including 1 (A)
~ 2/'2 2/'2 1 (P)revious 2 exams
- S 3/S 3/S 2 (randomly selected)
(R)CA
~ 1/'2 1/'2 1 (S)imulator
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:
RE. Ginna Date of Examination:
07/24/19 Exam Level: RO D SRO-I
~ SRO-U D Operating Test Number:
N2019-301 R Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*
Safety Function
- a. 006 Emergency Core Cooling System (ECCS) [006 A4.01 M,S,A,EN 2
(4.1/3.9)) Establish RCS Injection in AP-RCS.4 with Cl Valve Failures
- b. 062 AC Electrical Distribution System [062 A4.01 (3.3/3.1 )]
M,S,A 6
Transfer 4160V Auxiliary Loads and Take Actions for Loss of Bus
- c. N/A
- d. 010 Pressurizer Pressure Control System (PZR PCS) [010 A4.03 D,S, L 3
(4.0/3.8)] Placing L TOP in Service
- e. EPE W/E06 Degraded Core Cooling (EPE W/E06 EA2.2 (3.5/4.1)]
M,S,A,EN 4P Vent RCS for Accumulator/RHR Injection
- f. 045 Main Turbine Generator (MT/G) System [045 A4.01 (3.1/2.9)]
M,S,A 4S Perform Intercept and Reheat Stop Valve Test with Low EH System Pressure
- g. 026 Containment Spray System (CSS) [026 A2.08 (3.2/3.7)]
D,S,EN 5
Secure Containment Spray in E-1
- h. APE 026 Loss of Component Cooling Water (CCW) [APE 026 N,S,A 8
AA1.02 (3.2/3.3)] Respond to Complete Loss of CCW Flow In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U
- i. 039 Main and Reheat Steam System (MRSS) [039 A3.02(3.1/3.5*)]
D, E 4S Locally Close MSIVs
- j. 033 Spent Fuel Pool Cooling System (SFPCS) [033 G2.1.29 D,R 8
(4.1/4.0)] Alternate SFP Cooling Systems (A to B)
Isolate CI/CVI Valves All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2
- Type Codes Criteria for R / SRO-I / SRO-U (A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank
- 59/SS/::54 (E)mergency or abnormal in-plant
~1,~1,~1 (EN)gineered safety feature
~ 1 / ~ 1 / ~ 1 (control room system)
(L)ow-Power/Shutdown
~1,~1,~1 (N)ew or (M)odified from bank including 1 (A)
~2,~2,~1 (P)revious 2 exams
- s; 3 / S 3 I ::s; 2 (randomly selected)
(R)CA
~1,~1,~1 (S)imulator
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:
RE. Ginna Date of Examination:
07/24/19 Exam Level: RO D SRO-I D SRO-U
~
Operating Test Number:
N2019-301 R Control Room Systems:* 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*
Safety Function
- a. 006 Emergency Core Cooling System (ECCS) [006 A4.01 M,S,A,EN 2
(4.1/3.9)] Establish RCS Injection in AP-RCS.4 with Cl Valve Failures
- b. N/A
- c. N/A
- d. 010 Pressurizer Pressure Control System (PZR PCS) [01 O A4.03 D,S,L 3
(4.0/3.8)] Placing L TOP in Service
- e. N/A
- f. N/A
- g. N/A
- h. APE 026 Loss of Component Cooling Water (CCW) [APE 026 N,S,A 8
AA1.02 (3.2/3.3)] Respond to Complete Loss of CCW Flow In-Plant Systems:* 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U
- i. 039 Main and Reheat Steam System (MRSS) [039 A3.02(3.1/3.5*)]
D,E 4S Locally Close MSIVs
- j. N/A
Isolate CI/CVI Valves All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2
- Type Codes Criteria for R / SRO-I / SRO-U (A)lternate path 4-6 I 4-6 / 2-3 (C)ontrol room (D)irect from bank
- 59/:::8/S4 (E)mergency or abnormal in-plant
.::1/.::1/i!:1 (EN)gineered safety feature
.:: 1 /.:: 1 / i!: 1 (control room system)
(L )ow-Power/Shutdown
.::1/.::1/i!:1 (N)ew or (M)odified from bank including 1 (A)
.::2/.::2/i!:1 (P)revious 2 exams s; 3 /::: 3 / s 2 (randomly selected)
(R)CA
.::1/.::1/i!:1 (S)imulator
Appendix D Scenario Outline Form ES-D-1 Facility:
Ginna Scenario No.:
1 Op-Test No.: N2019-301R Examiners:
Operators:
Initial Conditions:
The plant is at 70% power due to Grid issues.
Turnover:
Plant has been at 70% power for 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />. Off-site power circuit 767 is OOS Critical Tasks: CT #1: Take MANUAL control of Pressurizer pressure to prevent an automatic Reactor trip from occurring.
CT #2: ECA-2.1-A Control the AFW flowrate to 50 gpm per SG in order to minimize the RCS Cooldown rate before a severe challenge (Orange Path} develops to the integritt CSF (EOP-Based}
Event Malf.
Event Event No.
No.
Type*
Description 1
N(BOP)
Raise Turbine Load to 100% in accordance with 0-5.2, Load R(ATC)
Ascension N(US) 2 PZR02D l(ALL)
PT-449, Pressurizer Pressure, fails HIGH TS(US) 3 OVR-C(ALL)
Loss of 4160V Bus 128 EDS44D TS(US) 4 STM05A M(ALL)
Both Steam Generators faulted downstream of MSIVs (MSIVs fail STM05B to close)
STM03 5
SIS02A C(ATC)
Safety Injection fails to Auto Actuate (manual successful)
SIS02B C(US) 6 RPS07E C(ATC)
RHR Pump 'A' fails to Auto Start after SI initiation (manual successful 7
Entry into ECA-2.1, Uncontrolled Depressurization of Both Steam Generators (N}ormal, (R)eactivity, (l}nstrument, (C}omponent, (M)ajor
Appendix D Scenario Outline Form ES-D-1 Ginna July 2019 NRC Simulator Exam #1 The plant is at 70% power following a plant power reduction to address Electrical Grid issues associated with storm damage that also resulted in a loss of Offsite Power Circuit 767. The plant has been at 70% power for 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />. Energy Operations has notified the station that grid stability has been restored and requests the station return to full power. Additionally, Offsite Power Circuit 767 will be available in approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
The following equipment is Out-of-Service: Offsite Power Circuit 767. A-52.12 submitted for TRM TR 3.8.1, 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Action.
Shortly after taking the watch, Operators will commence raising power at 10%/HR in accordance with 0-5.2, Load Ascension.
Approximately 2 minutes after commencing load ascension, PT-449, Pressurizer Pressure, fails high causing the Pressurizer Spray valves to OPEN. The Operator will respond in accordance with AR-F-2, PRESSURIZER HI PRESS 2310 PSI, and enter AP-PRZR.1, Abnormal Pressurizer Pressure. AP-PRZR.1 will refer the Operator to ER-INST.1, Reactor Protection Bistable Defeat After Instrumentation Loop Failure, for the defeat of PT-449. The Operator will address Technical Specification LCO 3.3.1, Reactor Trip System (RTS) Instrumentation; LCO 3.3.3, Post Accident Monitoring (PAM) Instrumentation, and Technical Requirements Manual TR 3.4.3, Anticipated Transient Without Scram (ATWS) Mitigation. If RCS pressure lowers<
2175 psig during the transient, then the Operator will also address Technical Specification LCO 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits.
Approximately 12 minutes into the scenario, 4160V Bus 12B is lost. Operators will respond in accordance with AP-ELEC.1, Loss of 12A and/or 12B Susses. 'B' Emergency Diesel Generator will automatically start and energize the 480 VAC safeguards Buses 16 and 17. The Operator will address Technical Specification LCO 3.8.1, AC Sources - MODES 1, 2, 3, and 4, and Technical Requirements Manual TR 3.8.1, Offsite Power Sources.
Approximately 30 minutes into the scenario, a large Steam line break occurs downstream of the MSIVs. MSIVs will NOT close. Safety Injection fails to automatically actuate requiring the Operators to manually initiate Safety Injection. RHR Pump 'A' fails to automatically start on SI initiation, Operators will manually start 'A' RHR Pump.
The crew will enter E-0, Reactor Trip or Safety Injection, and transition to E-2, Faulted Steam Generator Isolation. The crew will have to transition to ECA-2.1, Uncontrolled Depressurization of Both Steam Generators.
The scenario will terminate at Step 16 of ECA-2.1, after the crew has determined whether SI Termination criteria have been met and either terminates SI at Step 17 or returns to Step 2; or transitions FR-P.1, Response to Imminent.Pressurized Thermal Shock Condition, due to an ORANGE path on Integrity CSFST.
Appendix D Scenario Outline Form ES-D-1 Critical Tasks:
Take MANUAL control of Pressurizer pressure to prevent an automatic Reactor trip from occurring.
Safety Significance: Failure to manually control Pressurizer pressure will result in degrading Over-Temperature Delta-T conditions ultimately causing an automatic Reactor trip. In this case, Pressurizer pressure can be manually controlled from the control room. Therefore, failure to manually control Pressurizer pressure also represents a "demonstrated inability by the crew to take an action or combination of actions that would prevent a challenge to plant safety.
Additionally, under the postulated plant conditions, failure to manually control Pressurizer pressure (when it is possible to do so) results in a "significant reduction of safety margin beyond that irreparably introduced by the scenario."
Reduce AFW flow to both SGs to 50 gpm each per ECA-2.1, in order to minimize the RCS cooldown rate before a severe (orange-path) challenge develops to the integrity CSF (EOP-Based).
Safety Significance: Failure to control the AFW flow rate to the SGs leads to an unnecessary and avoidable severe challenge to the integrity CSF. Also, failure to perform the critical task increases the challenges to the subcriticality and the containment CSFs beyond those irreparably introduced by the postulated plant conditions.
Thus, failure to perform the critical task constitutes "demonstrated inability by the crew to take an action or combination of actions that would prevent a challenge to plant safety." It also causes a "significant reduction of safety margin beyond that irreparably introduced by the scenario."
Appendix D Scenario Outline Form ES-D-1 Facility:
Ginna Scenario No.:
2 Op-Test No.: N2019-301R Examiners:
Operators:
Initial Conditions:
Plant is at 48% Power EOL.
Turnover: Circuit 7T is OOS. MDAFW Pum12 'B' is OOS for bearing re12lacement.
Critical Tasks: CT #1 - E-3 -A: Isolate feedwater flow into and steam flow from the ru12tured SG before a transition to ECA-3.1 occurs.
CT #2 - E B: Establish/maintain an RCS tem12erature so that transition from E-3 does not occur because the tem12erature is either too high to maintain reguired subcooling or too low causing a challenge to the subcriticalit~ or integrit~ CSF.
nt Malf.
Event Event I
No.
No.
Type*
Description 1
N(BOP)
Load Reduction per 0-2.1, Normal Shutdown to Hot Shutdown N(US)
R(ATC) 2 ROD07 l(ALL)
T REF Fails Low 3
CVC07A C(ATC)
PCV-135 Fails Closed C(US) 4 EDS04D C(BOP)
Loss of Bus 18 RPS07R C(US)
'D' Service Water Pump fails to automatically start (manual TS(US) successful) 5 SGN04B C(ALL)
Steam Generator Tube Leak TS(US) 6 SGN04B M(ALL)
Steam Generator Tube Rupture 7
RPS07A C(ATC)
Safety Injection Pumps 'A' and 'B' fail to Auto Start (manual RPS07B C(US) successful) 8 RPSO?J C(ATC)
CNMT Recirc Fan 'D' fails to Auto Start (manual successful)
(N)ormal, (R)eactivity,
{l)nstrument, (C)omponent, (M)ajor
Appendix D Scenario Outline Form ES-D-1 Ginna July 2019 NRC Simulator Exam #2 The Plant is at 48% Power, EOL conditions. Station Management has decided to shutdown the unit due to the extended Circuit 7T outage.
The following equipment is Out-of-Service: Off-Site Power Circuit 7T. A-52.12 submitted for TRM TR 3.8.1, 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Action; Motor Driven Auxiliary Feedwater Pump 'B' is Out-of-Service for bearing replacement. A-52.4 submitted for ITS LCO 3.7.5, 7 day Action.
Operators will commence Shutdown in accordance with 0-2.1, Normal Shutdown to Hot Shutdown, at 10% / HR.
Approximately 2 minutes after commencing the shutdown, T REF input to Rod Control fails LOW causing inward rod motion. The Operators should determine Rod Motion is not called for and the HCO places RODS in manual per AP-RCC.1, Continuous Control Rod Withdrawal/Insertion.
Approximately 10 minutes into the scenario, PCV-135 fails closed causing a loss of letdown flow. The HCO should recognize the failure of PCV-135 and take manual control to restore Letdown flow per AR-A-11, LETDOWN LINE HI PRESS 400 PSI.
Approximately 17 minutes into the scenario, a fault on 480V Bus 18 will occur, resulting in Bus 18 de-energizing. The Operator will respond in accordance with AR-L-23, BUS 18 UNDER VOLTAGE SAFEGUARDS, and/or AR-L-5, SAFEGUARD BUS MAIN BREAKER OVERCURRENT TRIP, and enter AP-ELEC.17/18, Loss of Safeguards Bus 17/18. Operators will start Service Water Pump 'D'. The Operator will address Technical Specification LCO 3.8.1, AC Sources - Modes 1, 2, 3, and 4; and LCO 3.8.9, Distribution Systems - Modes 1, 2, 3, and
- 4.
Approximately 27 minutes into the scenario, a 5 gpm Steam Generator Tube Leak (SGTL} will develop on the 'B' Steam Generator. The Operator will respond in accordance with AR-PPCS-R47 AR, SGTL INDICATED, and enter AP-SG.1, Steam Generator Tube Leak, and commence a load reduction. The Operator will address Technical Specification LCO 3.4.13, RCS Operational Leakage, and LCO 3.4.17, Steam Generator (SG) Tube Integrity.
Approximately 32 minutes into the scenario, the Steam Generator Tube Leak will rise to 375 gpm. The Operator will recognize that the Charging System will not maintain Pressurizer Level thereby requiring a Reactor Trip and Safety Injection actuations and transition to E-0, Reactor Trip or Safety Injection.
Safety Injection Pumps 'A' and 'B' and Containment Recirc Fan 'D' fail to automatically start on Safety Injection signal. Manual Start is successful.
The crew will transition to E-3, Steam Generator Tube Rupture.
The scenario will terminate at Step 22 of E-3 after the crew has completed RCS depressurization and secured SI and RHR Pumps.
Appendix D Scenario Outline Form ES-D-1 Critical Tasks:
Isolate feedwater flow into and steam flow from the ruptured SG (B) so that minimum AP between the B SG and A SG is not less than 250 psid once target temperature is reached (Entry into ECA-3.1 at Step 16 RNO). (EOP-Based)
Safety Significance: Failure to isolate the ruptured SG causes a loss of ~p between the ruptured SG and the intact SG. Upon a loss of ~P. the crew must transition to a contingency procedure that constitutes an incorrect performance that "necessitates the crew taking compensating action which complicates the event mitigation strategy." If the crew fails to isolate steam from the SG, or feed flow into the SG, the ruptured SG pressure will tend to decrease to the same pressures as the intact SG, requiring a transition to a contingency procedure, and delaying the stopping of RCS leakage into the SG.
While in EOP-E-3, establish/maintain an RCS temperature so that transition from E-3 does not occur because the RCS temperature is in either (1) Too high to maintain 20°F of RCS Subcooling OR (2) below 284°F (RCS Integrity Red Path Limit) (EOP-Based)
Safety Significance: Failure to establish and maintain the correct RCS temperature during a SGTR leads to a transition from E-3 to a contingency procedure. This failure constitutes an incorrect performance that necessitates the operator taking compensating action that would unnecessarily complicate the event mitigation strategy.
ES-401 PWR Examination Outline Form ES-401-2 Facility: Ginna Date of Exam:
RO K/A Category Points SRO-Only Points Tier Group K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
Total A2 G*
Total
- 1.
1 3
3 3
3 3
3 18 3
3 6
Emergency and Abnormal Plant 2
1 2
2 N/A 1
2 N/A 1
9 2
2 4
Evolutions Tier Totals 4
5 5
4 5
4 27 5
5 10 1
3 3
3 2
2 3
2 3
3 2
2 28 3
2 5
- 2.
Plant 2
1 1
1 1
1 1
0 1
1 1
1 10 0
2 1
3 Systems Tier Totals 4
4 4
3 3
4 2
4 4
3 3
38 5
3 8
- 3. Generic Knowledge and Abilities 1
2 3
4 10 1
2 3
4 7
Categories 2
2 3
3 2
2 1
2 Note: 1.
Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the "Tier Totals" in each K/A category shall not be less than two). (One Tier 3 radiation control K/A is allowed if it is replaced by a K/A from another Tier 3 category.)
- 2.
The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.
- 3.
Systems/evolutions within each group are identified on the outline. Systems or evolutions that do not apply at the facility should be deleted with justification. Operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.
- 4.
Select topics from as many systems and evolutions as possible. Sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5.
Absent a plant-specific priority, only those K/As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6.
Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
- 7.
The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/As.
- 8.
On the following pages, enter the K/A numbers, a brief description of each topic, the topics' IRs for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel-handling equipment is sampled in a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2. (Note 1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9.
For Tier 3, select topics from Section 2 of the K/A catalog and enter the K/A numbers, descriptions, I Rs, and point totals(#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
G* Generic K/As These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.
These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.
ES-401 2
Form ES-401-2 IIES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions-Tier 1/Group 1 (RO/SRO)
E/APE # / Name I Safetv Function K1 K2 K3 A1 A2 G*
KIA Topic(s)
IR 000008 (APE 8) Pressurizer Vapor Space 8
PRT level pressure and temperature 3.8 1
Accident/ 3 000009 (EPE 9\\ Small Break LOCA / 3 12 Letdown isolation 3.4 2
000011 (EPE 11) Large Break LOCA / 3 3
Consequences of managing LOCA with loss of 3.7 3
ccw 000022 (APE 22) Loss of Reactor Coolant 3
Relationship between charging flow and PZR 3.0 4
Makeup/ 2 level 000025 (APE 25) Loss of Residual Heat 3
LPI pumps 3.4 5
Removal Svstem / 4 000027 (APE 27) Pressurizer Pressure 3
Controllers and positioners 2.6 6
Control Svstem Malfunction / 3 000029 (EPE 29) Anticipated Transient 6
Breakers, relays, and disconnects.
2.9 7
Without Scram / 1 000038 (EPE 38) Steam Generator Tube 4
Reflux boiling 3.1 8
Rupture/ 3 000040 (APE 40; BW E05; CE E05; W E12) 2.2.40 Ability to apply technical specifications for a 3.4 9
Steam Line Rupture-Excessive Heat system.
Transfer I 4 000054 (APE 54; CE E06) Loss of Main 4
Actions contained in EOPs for loss of MFW 4.4 10 Feedwater /4 000055 (EPE 55\\ Station Blackout/ 6 3
Actions necessary to restore power 3.9 11 000056 (APE 56) Loss of Offsite Power I 6 77 Auxiliary feed pump (running) 4.1 12 000057 (APE 57) Loss of Vital AC 2.1.7 Ability to evaluate plant performance and make 4.4 13 operational judgments based on operating Instrument Bus/ 6 characteristics, reactor behavior and instrument interpretation.
000058 {APE 58) Loss of DC Power I 6 1
Use of de control power by D/Gs 3.4 14 000077 (APE 77) Generator Voltage and 3
Under-excitation 3.3 15 Electric Grid Disturbances / 6 (W E04) LOCA Outside Containment I 3 3
Desired operating results during abnormal and 3.8 16 emergency situations.
(W E 11) Loss of Emergency Coolant 2.2.37 Ability to determine operability and/or 3.6 17 Recirculation / 4 availability of safety related equipment (BW E04; W E05) Inadequate Heat 2
Facility's heat removal systems, including 3.9 18 primary coolant, emergency coolant, the decay Transfer-Loss of Secondary Heat Sink/ 4 heat removal systems and relations between the proper operation of these systems to the operation of the facilitv.
000015 (APE 15) Reactor Coolant Pump 10 When to secure RCPs on loss of cooling or seal 3.7 76 Malfunctions / 4 iniection 000022 (APE 22) Loss of Reactor Coolant 2
Charging pump problems 3.7 77 Makeup/ 2 000029 (EPE 29) Anticipated Transient 2.4.2 Knowledge of system set points, interlocks and 4.6 78 Without Scram / 1 automatic actions associated with EOP entry conditions.
000058 (APE 58) Loss of DC Power I 6 2.4.34 Knowledge of RO tasks performed outside the
- 4. 1 79 main control room during an emergency and
-- ----- ____________ ------ ---------~. resu!~ant operational effects---------------
ES-401 3
Form ES-401-2 000077 (APE 77) Generator Voltage and 4
VARs outside the capability curve 3.6 80 Electric Grid Disturbances / 6 (W E 11) Loss of Emergency Coolant 2.4.3 Ability to identify post-accident instrumentation.
3.9 81 Recirculation / 4 KIA Category Totals:
3 3
3 3
3/3 3/3 Groua Point Total:
18/6
ES-401 4
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emeraencv and Abnormal Plant Evolutions-Tier 1/Grouo 2 RO/SRO)
E/APE #/Name I Safety Function K1 K2 K3 A1 A2 G*
KIA Topic(s)
IR 000005 (APE 5) Inoperable/Stuck Control Rod / 1 1
Stuck or inoperable rod from 3.3 19 in-core and ex-core NIS, in-core or loop temperature measurements 000028 (APE 28) Pressurizer (PZR) Level Control 3
Controllers and positioners 2.6 20 Malfunction / 2 000032 (APE 32) Loss of Source Range Nuclear 1
Startup termination on source-3.2 21 Instrumentation/ 7 ranae loss 000033 (APE 33) Loss of Intermediate Range Nuclear 2
Level trip bypass 3.0 22 Instrumentation/ 7 000060 (APE 60) Accidental Gaseous Radwaste Release/ 9 1
A radiation-level alarm, as to 3.1 23 whether the cause was due to a gradual (in time) signal increase or due to a sudden increase (a "spike"), including the use of strip-chart recorders, meter and alarm observations 000061 (APE 61) Area Radiation Monitoring System Alarms 1
Detectors at each ARM 2.5 24 17 svstem location 000067 (APE 67) Plant Fire On Site / 8 2.1.7 Ability to evaluate plant 4.4 25 performance and make operational judgments based on operating characteristics, reactor behavior and instrument interoretation.
000076 (APE 76) High Reactor Coolant Activity/ 9 6
Actions contained in EOP for 3.2 26 hiah reactor coolant activitv (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 2
Normal, abnormal and 3.3 27 emergency operating procedures associated with (Natural Circulation Ooerations).
000005 (APE 5) Inoperable/Stuck Control Rod / 1 2.2.22 Knowledge of limiting 4.7 82 conditions for operations and safety limits.
000024 (APE 24) Emergency Boration / 1 6
When boron dilution is taking 3.7 83
!olace 000033 (APE 33) Loss of Intermediate Range Nuclear 8
Intermediate range channel 3.4 84 Instrumentation / 7 operability (W E16) High Containment Radiation /9 2.4.4 Ability to recognize abnormal 4.7 85 indications for system operating parameters which are entry-level conditions for emergency and abnormal operatino procedures.
KIA Category Point Totals:
1 2
2 1
2/2 1/2 Group Point Total:
9/4
ES-401 5
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant S stems-Tier 2/Grou p 1 (RO/SRO)
Svstem # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
KIA Tooic(s)
IR 003 (SF4P RCP) Reactor Coolant 2
RCP seals and seal water supply 2.7 28 Pump 003 (SF4P RCP) Reactor Coolant 14 Starting requirements 2.6 29 Pumo 004 (SF1; SF2 CVCS) Chemical and 3
Boundary isolation valve leak 3.6 30 Volume Control 005 (SF4P RHR) Residual Heat 2
RHR flow rate 3.3 31 Removal 005 (SF4P RHR) Residual Heat 4
RHR valve malfunction 2.9 32 Removal 006 (SF2; SF3 ECCS) Emergency 10 Theory of thermal stress 2.5 33 Core Coolina 007 (SF5 PRTS) Pressurizer 2.2.38 Knowledge of conditions and limitations 3.6 34 Relief/Quench Tank in the facility license.
008 (SF8 CCW) Component Cooling 1
Setpoints on instrument signal levels for 3.2 35 normal operations, warnings and trips Water that are armlicable to the CCWS 008 (SF8 CCW) Component Cooling 3
All flow rate indications and the ability to 3.0 36 evaluate the performance of this closed-Water cycle coolinq system.
010 (SF3 PZR PCS) Pressurizer 2
PZR 3.2 37 Pressure Control 012 (SF? RPS) Reactor Protection 1
120V vital/instrument power system 3.4 38 012 (SF? RPS) Reactor Protection 1
RPS channels, components and 3.3 39 interconnections 013 (SF2 ESFAS) Engineered 1
ESFAS-initiated equipment which fails to 4.5 40 Safetv Features Actuation actuate 022 (SF5 CCS) Containment Coolinq 1
Coolinq of containment penetrations 2.5 41 026 (SF5 CSS) Containment Spray 1
Containment pressure 3.9 42 039 (SF4S MSS) Main and Reheat 8
Effect of steam removal on reactivity 3.6 43 Steam 059 (SF4S MFW) Main Feedwater 3
Feedwater oumo suction flow pressure 2.5 44 061 (SF4S AFW) 2 AFW electric drive pumps 3.7 45 Auxiliary/Emergency Feedwater 062 (SF6 ED AC) AC Electrical 5
Methods for energizing a dead bus 2.9 46 Distribution 062 (SF6 ED AC) AC Electrical 1
Major system loads 3.5 47 Distribution 063 (SF6 ED DC) DC Electrical 2
Components using DC control power 3.5 48 Distribution 064 (SF6 EDG) Emergency Diesel 3
ED/G (manual loads) 3.6 49 Generator 064 (SF6 EDG) Emergency Diesel 4
Overload ratings 3.1 50 Generator 073 (SF? PRM) Process Radiation 1
Those systems served by PRMs 3.6 51 Monitoring
ES-401 6
Form ES-401-2 076 (SF4S SW) Service Water 2
SWS valves 2.6 52 078 (SF8 IAS) Instrument Air 2.4.35 Knowledge of local auxiliary operator 3.8 53 tasks during an emergency and the resultant operational effects.
078 (SF8 IAS) Instrument Air 1
Instrument air compressor 2.7 54 103 (SF5 CNT) Containment 2
Containment isolation/containment 3.9 55 intearitv 004 (SF1; SF2 CVCS) Chemical and 2.2.12 Knowledge of surveillance procedures.
4.1 86 Volume Control 013 (SF2 ESFAS) Engineered 4
Loss of instrument bus 4.2 87 Safety Features Actuation 059 (SF4S MFW) Main Feedwater 3
OverfeedinQ event 3.1 88 061 (SF4S AFW) 2.4.34 Knowledge of RO tasks performed 4.1 89 outside the main control room during an Auxiliary/Emergency Feedwater emergency and the resultant operational effects 078 (SF8 IAS) Instrument Air 1
Air drver and filter malfunctions 2.9 90 KIA Categorv Point Totals:
3 3
3 2
2 3
2 3/3 3
2 2/2 Group Point Total:
28/5
ES-401 7
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant S stems-Tier 2/Grouo 2 /RO/SRO)
Svstem # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
KIA Tooic(s)
IR 002 (SF2; SF4P RCS) Reactor 12 Code Safety valves 3.0 56 Coolant 011 (SF2 PZR LCS) Pressurizer 2
PZR heaters 3.1 57 Level Control 014 (SF1 RPI) Rod Position 1
Loss of offsite power 2.8 58 Indication 017 (SF? ITM) In-Core Temperature 3
Indication of superheating 3.7 59 Monitor 029 (SF8 CPS) Containment Purge 2.4.49 Ability to perform without reference to procedures those actions that require 4.6 60 immediate operation of system components and controls.
033 (SF8 SFPCS) Spent Fuel Pool 2
Spent fuel leak or rupture 2.9 61 Cooling 034 (SF8 FHS) Fuel-Handling 1
RCS 2.5 62 Eauioment 035 (SF 4P SG) Steam Generator 1
RCS 4.4 63 068 /SF9 LRS) Liauid Radwaste 4
Automatic isolation 3.8 64 079 (SF8 SAS**) Station Air 1
Cross-connect with IAS 2.9 65 017 (SF? ITM) In-Core Temperature 2
Core damage 4.1 91 Monitor 029 (SF8 CPS) Containment Purge 1
Maintenance or other activity taking place 3.6 92 inside containment 075 (SF8 CW) Circulating Water 2.1.25 Ability to interpret reference materials such as graphs, monographs and tables 4.2 93 which contain oerformance data.
KIA Cateaorv Point Totals:
1 1
1 1
1 1
0 1/2 1 1
1/1 Grouo Point Total:
10/3
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 I Facilit:t::
Date of Exam:
I Category KIA#
Topic RO SRO-only IR IR 2.1.36 Knowledge of procedures and limitations involved in core 3.0 66 alterations 2.1.45 Ability to identify and interpret diverse indications to validate 4.3 67 the response of another indication
- 1. Conduct of Operations 2.1.35 Knowledge of the fuel-handling responsibilities of SROs.
3.9 94 2.1.39 Knowledge of conservative decision making practices 4.3 95 Subtotal 2
2 2.2.6 Knowledge of the process for making changes toprocedures 3.0 68 2.2.7 Knowledge of the process for conducting special or infrequent 2.9 69 tests
- 2. Equipment Control 2.2.13 Knowledge of tagging and clearance procedures.
4.3 96 2.2.18 Knowledge of the process for managing maintenance activities 3.9 97 durina shutdown ooerations.
Subtotal 2
2 2.3.13 Knowledge of radiological safety procedures pertaining to 3.4 70 licensed ooerator duties 2.3.14 Knowledge of radiation or contamination hazards that may 3.4 71 arise during normal, abnormal, or emergency conditions or activities
- 3. Radiation 2.3.4 Knowledge of radiation exposure limits under normal and 3.2 72 Control emeraencv conditions 2.3.12 Knowledge of radiological safety principles pertaining to 3.7 98 licensed ooerator duties Subtotal 3
1 2.4.29 Knowledge of the emergency plan.
3.1 73 2.4.43 Knowledge of emergency communications systems and 3.2 74 techniaues.
2.4.8 Knowledge of how abnormal operating procedures are used in 3.8 75 conjunction with EOPs.
- 4. Emergency Procedures/Plan 2.4.12 Knowledge of general operating crew responsibilities during 4.3 99 emeraencv ooerations.
2.4.37 Knowledge of the lines of authority during implamentation of an 4.1 100 emeraencv olan.
Subtotal 3
2 Tier 3 Point Total 10 7