ML20094E098
| ML20094E098 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 09/06/1995 |
| From: | Sellman M ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| References | |
| NUDOCS 9511070061 | |
| Download: ML20094E098 (21) | |
Text
b (k
Ent;;rgy Oper-ti:nn, ins.
R$er Bend Station 5485 U S. H.ghway 61
-==~ ENTERGY 7e"L's ;
2 (504) 381 4374 FAX (504) 3814872 JOHN R. McGAHA, JR.
Vce Pres 4 dent Operations September 6,1995 U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011
Subject:
River Bend Station Submittal of Formal NRC Examination Comments River Bend Station - Unit i Docket No. 50-458 File No.:
G9.5, G1.41.26 RBF1-95-214 RBG-41935 l
RBEXEC-95-137 Gentlemen:
In accordance with NUREG-1021, Revision 7, Supplement 1(ES-402, Attachment 3),
enclosed are four questions submitted for your review and consideration. The enclosed questions (and suppoding documentation) are a result of the NRC administered licensing examination conducted at the River Bend Station the week of August 18,1995. The chief examiner during the examination was Mr. Howard Bundy.
If you have any questions regarding the attached, please contact Mr. L. Grant Lewis at (504) 381-4752.
Sincerely, j }(( L S~
_Q,.
T e M c C A JRM/JJF enclosure 660033 9511070061 950906 PDR ADOCK 05000458 y
PDR k
I
}
1 (c
River Bend Station Submittal of Formal NRC Examination Comments September 6,1995 l
RBF1-95-0214 j
RBG-41935 RBEXEC-95-137 Page 2 of 2
]
l cc:
U. S. Nuclear Regulatory Comnussion Attention: Document Control Desk Mail Stop PI-37 Washington, DC 20555 NRC Sr. Resident Inspector P. O. Box 1051 St. Francisville, LA 70775
)
David Wigginton NRR Project Manager U. S. Nuclear Regulatory Commission NRR Mail Stop 13-H-3, One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. Howard Bundy U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington,TX 76011 i
l
- 1) Question #24 RO/#27 SRO l
If total feedwater flow drops below the RRS interlock level, the Reactor Recirculation l
pumps will downshift to slow speed.
l l
What is the PRIMARY reason for this interlock?
)
- a. pump cavitation.
- b. flow control valve cavitation.
- c. excessive axial thrust on the pump.
- d. inaccurate wide range level indication.
Answer: b
Reference:
LOTM-7-5, page 12 of 38, section V.A.1 Recirculation System Enabling Objective 3.3 K/A 202002K108, (3.1/3.2)
Comment:
Both (a) and (b) are correct. LOTM-7-5 page 12 of 38, section V.A.1 does give the key's answer but does not mention anything about being the PRIMARY reason for the interlock. There is some conflicting information that supports answer (a). LOTM 34-6 (Feedwater Level Control System), page 5 of 16, B.3 specifically states that the low total feedwater flow limit is based on preventing cavitation of the Reactor Recirculation pumps due to inadequate subcooling. HLO-005-05 (lesson plan), page 10 of 27,4.1.4 does not give a PRIMARY component. Also HLO-060-5 (lesson plan), page 9 of 26 " NOTE" reinforces answer (a) to the students.
Recommendation:
Accept both answer (a) and (b).
Reactor Water Cleanup (Chapter 14),
Floor and Equipment Drains (Chapter 49), and Nuclear Boiler Instrumentation (Chapter 3).
V.
SYSTEM OPERATION A.
Nomal Operation 1.
Cavitation Interlocks a.
Cavitation is prevented by inhibiting high speed operation with total feed water flow less than 25%.
If pumps are operating at high speed and feed flow drops below 25%, pumps auto i
transfer to low speed after a 15-second time delay.
Startirc transferring to high speed is inhiExted.
White light illuminates above the cavitation reset pushbutton on P680 panel, indicating a " seal-in" condition.
This interlock downshifts the pumps because feed flow of less than 25
)
percent will not provide adequate NPSH to prevent cavitation of the FCVs during fast speed recirc pump operation.
This interlock may be bypassed by taking SW-127A(B), Low Total Flow Interlock Bypass switch, located on panel H22-B33-P001A(B) to BYPASS.
b.
If reactor level falls to level 3 (9.7"),
auto transfer to slow speed is initiated.
This also allows for more accurate wide range level reading.
The above interlocks will downshift both pumps to slow speed.
c.
If both loops receive a main steamline temperature /recirc pump suction differential temperature signal of <8oF, auto transfer to slow speed will be initiated after a 15-second time delay.
LOIM-7-5 Page 12 of 38
i 4
4 differential pressure transmitter (1C33-FT-N003A,B,C,D) is sent to the Main Control Room where it passes through an SRU to be converted from 4
current to voltage.
The following relationship exists between AP and flow:
Flow = constant x VAP.
Since the resultant flow signal is not linear, the signal then passes through a square root extractor (Figure 1).
The square root extractor linearizes the signal such that flow indications are linear and the control system responds in a
linear fashion.
1 When approximately 40% rated steam flow is sensed in all 4 steam lines, a close signal is sent to t
B21-F033 (steam line inboard drain) and B21-F069 (steam line outboard drain).
Each of the four flow signals is then sent to a summing circuit whose output is a total steam flow 4
signal.
In addition, the four steam line flows are displayed on meters on P680 (pounds /hr x 106).
3.
Feedwater Flow (Figure 4)
Feedwater flow is measured by a venturi type flow element (1C33-FEN 001A,B) located in each of the two feedwater lines to the reactor vessel.
Each venturi differential pressure signal is measured by a transmitter (1C33-FT-N002A,B).
Similar to steam flow, the feedwater flow signals pass through an SRU and a square root extractor (Figure 1).
The output signal from the square root extractor is sent to a summer where the two feedwater flow signals are summed to give the total feedwater flow.
The two feedwater flows are also displayed on meters on P680 (figure 10).
Low total feedwater flow
($25%
rated for 15 seconds) provides a signal to transfer the Reactor Recirculation pumps from fast to slow speed during downpower.
This limit is based on preventing
)
cavitation of the Reactor Recirculation pumps due to inadequate subcooling.
It also prevents upshift from slow to fast speed during power ascension if this permissive is not met.
C.
Component Description 1.
Feed Flow / Steam Flow Summer (Figure 1 - K602)
The total feed water flow and steam flow signals are input into the feed flow / steam flow summer.
LOTM-34-6 Page 5 of 16
4.0 Description of controls and interlocks OBJ #4a.
4.1 Pump Cavitation Interlocks CB NOTE:
List the cavitation interlocks on the board as each is discussed.
4.1.1 Cavitation is prevented by inhibiting high speed operation with j
total feed water flow less than 25%.
If pumps are operating at high speed and feed flow drops e
below 25%, pumps auto transfer to low speed after a 15-second time delay.
Starting or transferring to high speed is inhibited.
White light illuminates above the cavitation reset pushbutton e
on P680 panel, indicating a " scal-in" condition.
This interlock downshifts the pumps because feed flow of less y
than 25 percent will not provide adequate NPSH in the downcomer for fast speed rectre pump operation.
This interlock may be bypassed by taking SW-127A(B).
located on panel H22-833-P001 A (B) to BYPASS.
l 4.1.2 If main steamline temperature /recim pump suction differential temperature falls to 8'F, auto transfer to slow speed will bc initiated after a 15-second time delay.
This interlock is provided because the low pressure area at the recire pump suction might fall below saturation pressure of the coolant under such a low differential temperature condition This interlock may be bypassed by taking SW-125A(B.
located on panel H22-B33-P001 A(B) to BYPASS.
4.1.3 If reactor level falls to level 3 (9.7") auto transfer to slow spece is initiated.
4.1.4 All of the above interlocks are specifically designed to prevent or limit cavitation at the:
Jet pumps
)
Recirculation pumps e
Flow Control Valves.
HLO405-45 PAGE 10 OF 27
(
linearize the signal.
TP-06 3.0.4 Two meters on P680-3B indicate feedwater flow is the two lines with a scale of pounds /hr x 105 3.0.5 The two feedwater flow signals are input to a total feed flow summer to provide a total feedwater flow signal.
The output of this summer is sent to:
Total feedwater flow indication (steam flow / feed flow recorder P680-3B).
Feed flow / steam flow summer Bailey alarm cards K618A/B send a low total feedwater flow signal to the Recirculation System. This signal will
])r transfer recirculation pumps to LFMG during a downpower.
It also prevents upshift from slow to fast speed during power ascension (if this permissive is not met).
NOTE:
Ask students what the purpose of the
's-downshift is and what the setpoint is?
/
ANS: Purpose is to prevent pump cavitation due to inadequate subcooling
($25% rated flow for 15 seconds).
4.0 Feedwater Level Control System Components OBJ#3.1 4.1 Feed Flow / Steam Flow Summer TP-04 4.1.1 The total feedwater flow and steam flow signals are input into the feed flow / steam flow summer.
4.1.2 Its purpose is to provide a method for the Feedwater Level Control System to anticipate Rx level changes and take appropriate corrective action.
4.1.3 The principle is simple.
If feed flow matches steam flow, level should be relatively constant.
If a match does not exist, a change in level can be expected.
4.1.4 The feedwater flow input is negative HLO-060-5 Page 9 of 26
- 2) Question #47 RO/#46 SRO Which of the following will prevent RCIC discharge to the CST through the test line return valves (F022, F029)?
- a. The CST suction valve is open (F010)
- b. RCIC minimum flow valve open (F019)
- c. CST level is less than 6.5 inches.
- d. Reactor vessellevelis 55 inches.
i Answer: d
Reference:
LOTM-20-4, RCIC, page 3, II.A.2 and table #1 RCIC objective 4 and 5.j K/A 217000A301, (3.5/3.5)
Comment:
Answer (c) could also be correct. Operators know that RCIC pump suction is normally lined up to the CST as long as there is at least about 3 and a half feet in the CST. This lineup automatically shifts the suction path from the CST to the suppression pool when there i:!ee: t.% 3'5" in the CST which correlates to a trip instmment "0" on the back panels. When the swap of the suction valves takes place, the test return valves are interlocked with the suppression pool suction valve to go shut and will prevent RCIC discharge to the CST. Since 6.5" is less than 3'5" this would be a correct answer. The confusion is that the response choice is not clear if this is actual tank level or a trip instrument level. To get actual CST level, the operator calls the auxiliary control room which gives actual CST tank level in feet. The operator can also call up a computer point which is also in units of feet (actual tank level). See LOTM-20-4, page 3, II.A.2. Also see table 2 (LOTM-20-4, page 15 of 24). Also see Tech. Spec. table 3.3.3-2 note ** page 3/4 3-39 for tank level correlation. Also see Tech. Spec. table 3.3.5-2, page 3/4 3-57 for trip setpoint.
Recommendation:
Accept both answer (c) and (d)
i 1
1 C.
Genera _l D,escription f-The R5C system is started automatically upon receipt of a low reactor water level signal (level 2) or manually by the operator. Water from the CST or suppression pool is l
pumped into the core by a turbine-driven pump powered by reactor steam.
[
D.
Basic System Flow Path I
The RCIC pump suction is normally lined up to the Condensate Storage Tank (CST).
This provides an adequate supply of high purity water for system operation. Steam to operate the turbine is supplied via piping from Main Steam Line (MSL) A upstream of the inboard Main Steam Isolation Valve (MSIV). The RCIC pump discharges to the reactor vessel upper head spray nozzle. A backup source of water for the RCIC pump is available from the suppression pool. Shifting to this source of water under normal condi-tions requires deliberate operation of valves by the operator.
IL SYSTEM DETAILS A.
Detailed Flow Path (Figure 1) 1.
Steam Flow l,
Steam for operation of the RCIC turbine is provided from MSL A inside the drywell upstream of the inboard MSIV (B21-F022A). The steam piping size is 8" up to the point that taps off to the RHR system, and then reduces to 4" to supply the RCIC turbine.
The steam piping to the turbine is kept hot to allow rapid starting of the turbine, therefore allowing rated RCIC system flow to be attained in less than 30 seconds when needed. This is accomplished by a normal valve lineup and piping and steam trap arrangement which maintains the steam piping at near normal operating tem-perature.
The turbine is designed for immediate starting with no warmup prior to operation at rated speed. Turbine exhaust steam is directed to the suppression pool for conden-sation via 12" piping.
A gland seal air system is provided to prevent steam leakage from the turbine glands, govemor and throttle valves. Air is provided by a compressor to counteract steam leakage from the above points.
2.
Water Flow l
The RCIC pump suction is normally lined up to the CST. Suction automatically shifts to the suppression pool when:
a low level exists in the CST (0"), or e
a high level exists in the Suppression Pool (+6.5") with a RCIC isolation sig-l nal not present.
NQII: Tech Specs require RCIC suction to shift at $6.5", or 20'-3.5" actual Suppression Pool level. Current serpoint is 19'-10.9", which satisfies this requirement.
l LOTM-20-4 PAGE3 OF 24
7 i
~
Table 2 (continued) 1H13*P601/21A ANNUNCIATORS DIV II RCIC ISOL STM 60 psig (3 see TD)
RCIC System Auto Isolation:
SPLY PRESS LOW e
RCIC and RHR steam supply inboard i
isol. valve (1E51*F063) closes.
RCIC steam line warmup isol. valve (1E51*F076) closes.
RCIC pump min flow to suppression pool valve (1E51*F019) closes.
RCIC injection isol. valve (F013) e closes.
RCIC WARMUP LINE Valve not closed None.
ISO VLV E51-F076 NOT FULLY CLOSED RCIC TURBINE STEAM 0" Increasing RCIC steam supply drain trap bypass valve SPLY WATER DRAIN (mid-range on trap) (1E51*F054) opens.
TRAP LVL HI RCIC TURB TRIP PMP 20" Hg vac Dect.
SUCT PRESS LOW (0.5 second TD)
RCIC SUCT XFER 0" on meter RCIC pump suppression pool suction CST LEVEL LOW valve (1E51*F031) opens.
RCIC pump CST suction valve (1E51*F010) closes.
RCIC test bypass valve to CST (1E51*F022) closes.
RCIC test return valve to CST (1E51*F059) closes.
RCIC ISOLATION Ambient 182*F RCIC System Auto Isolation:
RCIC RM HI AMB OR Vent Diff 96 F AT RCIC turbine trips.
VENT DITF TEMP RCIC steam supply valves (1E51*F063 and IE51*F064) close.
j RCIC inject isol valve (1E51*F013)
Closes.
RCIC pump suppression pool suction valve (1E51*F031) closes.
RCIC min flow valve to suppression pool (1E51*F019) closes.
RCIC warmup line shutoff valve (1E51*F076) closes.
LOTM-20-4 PAGE 15 OF 24
$g.
A o4
/
, ~.
2.,
/, -
TAett 3.3.3-2 (Continued) 9 O
s 7
n EfERGEIGCY CORE COOLING SYSTEM ACIDATION INSTRLSENTATION SETPOINIS ALLOWABLE TRIP FLRitilOlt,'
TRIP SETPOINT VALUE D.
Loss OF POWER (continued) 2.
Divisten III a.
4.16 kw Stan 6y Bus ifnderveltage a.
4.16 kw Basis -
(Sustained landerveltage) 3045 1 153. welts 3045 1 214 velts b.
3 1 0.3 sec. time 3 1 0.33 sec. time delay delay b.
4.16 kw Staney Bus IInderveltage a.
4.16 kw Basis -
(Degraded Voltage) 3777 1 30 welts 3777 1 75 volts b.
60 1 6 sec. Line 60 1 6.6 sec. time delay l
delay (w/o LOCA) c.
3 1 0.3 sec. time 3 1 0.33 sec. time delay delay (w/LOCA) l 85ee Bases Figure S 3/4 3-1.
CC(8ettee of CST is at EL 95'1".)
The levels are measured from the lastrument zero level of EL 90'6".
- (Bettee of seppression pesi is at EL 70'.)
The levels are measured from the lastrument zere level of EL 89'9".
- These are inverse time delay voltage relays er instantaneous voltage relays with a time delay.
The weltages shown are the enzimum that will not result in a trt.
Lower weltage conditlens will result P
in decreased trip times.
TAliLE 3.3.5-2 REACTOR CORE I50LAIION COOLING SYSIEN ACTUATION INSTRUNENTATION SETPOINTS
~
g N
ALLOWABLE g
FUNCTIONAL UNITS TRIP SE1 VALUE
[
1.
Reactor Vessel Water Level - Low Low level 2 3 -43 inches
- 3 -47 inches xZ 2.
Reactor Vessel Water Level - High Level 8
$ 51 inches *
$ '52 inches Condensate Storage 5nk Levei l ow~
> 0 inches )
-4.5 friches 4.
Suppression Pool Water Level - High 5 6.5 inches 1 8 inches 5.
Manual Initiation NA NA
- See Bases Figure B 3/4 3-1.
T 20
7
- 3) Question #52 RO/#49 SRO While in a refueling outage, which of the following requires Primary Containment Integrity to be established?
a.
Both trains of Standby Gas Treatment becomes inoperable.
- b. One LPRM detector will be replaced with a new one.
- c. All source range detectors are discovered to be inoperable.
- d. Irradiated fuelis to be moved in the fuel pool.
Answer: b
Reference:
TS 3.6.1.2, page 6-2 amendment 35 and definition 1.7 Core Alterations Primary containment objective 9A and 9B K/A 223001GO11, (3.3/4.2)
Comment:
No correct answer. Answer (b) assumes that replacing an LPRM is a core alteration. For our plant (BWR/6), the LPRM resides in a dry tube and if removed during a refuel outage, would have only a negligible (if any) effect on core reactivity. Replacing an LPRM is not a core alteration. See attached NRC safety evaluation of T.S. amendment No. 29, dated October 12,1988.
Recommendation:
Throw out the question since there is no correct answer.
u.g\\
6C - 31(g p. -
Attached to-
/
UNITE 3 8TATES NUCLEAR REGULATORY COMMISSION
{
wAaMasseTose. D. c. 20ess SAFETY EVALUATION:BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 29 TO FACILITY OPERATING LICENSE N0._NPF-47 GULF STATES _ UTILITIES _ COMPANY RIV_ER BEND STATION, UNIT 1 DOCKET NO. 50-458
1.0 INTRODUCTION
By letter dated August 5, 1988 Gulf States Utilities Company (GSU)
(the licensee) requested an amendment to Facility Operating License No.
NPF-47 for the River Bend Station, Unit 1.
The proposed amendment would modify the Technical Specifications (TSs) to revise the definition of Y
core alteration to exclude the. normal moveme.nt (including. replacement) of local power range monitors. (LPRMs) from this definition.,
7.0 EVALUATION Technical Specification Definition 1.7, CORE ALTERATION, currently does not consider normal movement of the source range monitors, intermediate range monitors, traversing in-core probes, or special moveable detectors to be considered a core alteration. This change request would provide 7
the same exclusion for LPRMs.
River Bend Station is a BWR/6 boiling water reactor which incorporates certain design changes compared to earlier boiling water reactors. One of these changes is the introduction of a dry tube that houses the LPRM strings. The dry tubes extend from the bottom of the reactor pressure vessel vertically to the top of the core. Thus, removal and installation of the LPRMs from underneath the reactor pressure vessel can be accomplished without the removal of the reactor vessel head and fuel does not need to be moved from around the dry tube for maintenance or replacement of LPRMs. The LPRM strings are only removed from the core when they are being replaced and they have no normal drive mechanisms.
Based on the above discussion, the staff concludas that the exclusion of the LPRMs in the definition of core alteration is acceptable.
~'
With the modification of the definition of cure alteration discussed above, the footnote excepting replacement of LPRM strings applicable to Action 3 and Action 9 of Table 3.3.1-1 is no longer necessary. The staff concludes that deletion of the footnote is acceptable.
3.0 ENVIRONMENTAL CONSIDERATION
The amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The staff has determined that the amendment involves no significant station vocw.u,t m OCT 211988
~ - - - - - -
increase in the' amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposures.
The Cosmission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public connent on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR Section 51.?2(c)(9).
. Pursuant to 10 CFR 51.22(b), no environmental impact statement or environ-mental assessment need be prepared in connection with the issuance of the amendment.
4.0 CONCLUSI0g The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations.
and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date: October 12, 1988 Principal Contributor:
W. Paulson y
=
4
- 4) Question #75/#71 SRO With the reactor initially at 100% power, a loss ofinstrument air to which of the following will NOT eventually result in an automatic REACTOR PROTECTION SYSTEM SCRAM?
a.
Condensate and heater drain pump recirc valves.
- b. Feedwater regulating valves,
- c. Turbine steam seals and SJAE.
I
- d. Scram inlet and outlet valves.
Answer: d
Reference:
AOP-0008, R7A, section 4.0 AOP-0008 objective 3.a -
K/A 295019K201, (3.8/3.9)
Comment:
The words "will not eventually" leaves this question with no correct answer. Answer (a) does not send a direct RPS trip but will eventually cause a feedpump trip and eventually a RPV level 3 scram (see AOP-0008, Rev.7A page 3 of 16, section 3.2). Answer (b) will cause the feedwater regulating valves to lock up and fail"as is". Changes in power, pressure, temperature, and other non-constant external and internal forces (even fuel depletion) will cause a steam flow to feed flow mismatch and you'll get either a high or low RPV level scram. Also the air pressure that is locked up, keeping the feedwater regulating valves in a steady, constant position, will eventually bleed slowly. Answer (c) will result in a loss of condenser vacuum which does not send a direct RPS trip but will eventually cause a turbine trip and a resultant Rx scram (see AOP-0008, Rev. 7A page 3 of 16, section 3.5). Answer (d) will cause the rods to go in, and will cause either the scram discharge volume to fill faster than it can drain (assuming that the scram discharge volume vent & drain valves did not fail closed since the air line is common to the scram inlet and outlet valves which lost air) and cause a RPS scram or because the turbine is on the line and Rx power level is going down due to rods going in will cause Rx pressure to go below 849 psig (eventually because of other steam loads) and cause the MSIV's to go closed and cause a RPS scram (see LOTM-5 Fig.12, Tech. Spec table 2.2.1-1(6)&(9),
f pages 2-4&5, and Tech. Spec. table 3.3.2-2(2.c), page 3/4 3-19).
Recommendation:
4 Throw out the question since there is no correct answer.
i i
T I.0 P.URPOSE/ DISCUSSION il l
1.1 The purpose of this procedure is to provide guidance to the operators in the event l'
that Instrument Air System air pressure.is lowering or lost.
l 1.2 A total loss of instrument air pressure may be caused by a break in the instrument air header, or by a loss of all air compressors. A multitude of actions occur as a result oflow instrument air supply pressure. These actions occur at various times
. depending upon the rate of the instrument air pressure drop. The actions listed in 3.0, Automatic Actions, are listed in decreasing order of significance to the Nuclear Steam Supply System.
}
2.0 SYMPTOMS l
2.1 Lowering Instmment Air Header Pressure.
l 2.2 Amber indicating lights for compressor llAS-CI A, CIB and/or CIC.
f 2.3 Various AOV's will fail (see Attachment 1) in a random manner.
1 j
3.0 AUTOMATIC ACTIONS l
3.1 Control rods individually scram as the scram valves fail open. The Scram Discharge
+
Volume vent and drain valves fail closed. CRD flow control valves fail closed.
(
__ y 3.2 Condensate and heater drain pumps recire valves will open and " starve" the reactor feed pumps; causing them to trip
-._p.
3.3 The Feedwater Reg Valves will lock up and fail "as is" on low air pressure (85 psig).
3.4 All normal HVAC will fail due to closure of AOD's.
_._y 3.5 Loss of steam seals and SJAE will result in a loss of condenser vacuum.
Drywell and containment drains will isolate.
3.6 4.0 IMMEDIATE OPERATOR ACTIONS 4.1 Make a plant wide Qaitronics announcement to cease non essential use of air.
4.2 If any of the following occurs, Scram the Reactor and enter AOP-0001 REACTOR SCRAM:
4.2.1 When individual rod movement is observed.
4.2.2 When the instrument air header pressure decreases to 65 psig (IIAS-PIl05) on 1H13*P870.
5.0 SUBSEOUENT OPERATOR ACTIONS 5.1 Monitor air header pressure and if it lowers to 50 psig, verify closed or close the MSIV's.
A O P-0008 REV.7A PAGE 3 OF 16
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iS TABLE 2.2.1-1 i
<g REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS l
E FUNCTIONAL UNIT ALLOWABLE o
TRIP SETPOINT VALUES
[
1.
Intermediate Range Monitor, Neutron Flux-High 5 120/125 divisions 5 122/125 divisionf
=
of full scale of full scale j
2.
Average Power Range Monitor:
a.
Neutron Flux-High, Setdown
< 15% of RATED
< 20% of RATED
~ THERMAL POWER THERMAL POWER b.
Flow Blased $1mulated Thermal Power-High
- 1) Two Recirculation Loop Operation i
i a) Flow Blased 1 0.66 W+485, with 1 0.66 W+51%, with a maximum of a maximum of i
b) High Flow Claged 5 111.0% of RATED 1 113.0% of RATED l THERMAL POWER THERMAL POWER i
- 2). Single Recirculation Loop Operation i
a) Flow Blased
< 0.66 W+42.7%, with
<0.66W+45.7%,wik a maximum of a maximum of i
b) High Flow Clamped
< 111.0% of RATED
< 113.0% of RATED gg THERMAL POWER-THERMAL POWER l
m# o c.
Neutron Flux-High
-< 118% of RATED
< 120% of RATED O $"
THERMAL POWER THERMAL POWER O G5 m
d.
Inoperative NA NA
- C 3.
Reactor Vessel Steam Dome Pressure - High 5 1064.7 psig i 1079.7 psig ar 4.
Reactor Vessel Water Level - Low, Level 3
> 9.7 inches above
>8.7inchesabovel 2,
instrument zero*
instrument zero :
5.
Reactor Vessel Water Level-High, Level 8 5 51.0 inches above
$ 52.1 inches above; s
instrument zero" Instrument zero y
Main Steam Line Isolation Valve - Closure
$ 85 closed i 12% closed 6.
"See Bases Figure 8 3/4 3-1.
l f
TABLE 2.2.1-1 (Continued) l m2g REACTOR PROTECTION SYSTEN INSTRUNENTATION SETPOINTS
=
9 ALLOWABLE o
FUNCTIONAL UNIT TRIP SETPOINT VALUES E
7.
Main Steam Line Radiation -
High
< 3.0 x full power
< 3.6 x full power '
q background background
~
8.
Drywell Pressure - High 1 1.68 psig
< 1.88 psig
)
9.
Scram Discharge Volume Water Level - High a.
Level Transmitter - LiSN6n1A and B *
< 49"
< 53" LISN601C and D
?49" 3 51.7" i
b.
Float Switche:;
- LSN013A and 8
< 48.76"
< 53.50"
-7 46.88" 7 49.00" LSN013C and 0
{
- 10. Turbine Stop Valve - Closure 1 5% closed i 7% closed
- 11. Turbine Control Valve Fast Closure, Trip 011 Pressure - Low 1 530 psig 2 465 psig 2 :o o rn 12.
Reactor Mode Switch Shutdown Position NA NA (n
a O $$
- 13. Manual Scram.
NA NA o m
@ o N
3n E
M
- See Bases Figure B 3/4 3-1.
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JUN 0 81923 y
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5 RIVER BEND - UNIT 1 3/4 3-19 AMENOMENT NO,I3. N
-