ML20006E277

Rev 0 to Decrease of RHR Pump Rated Flow by 10% in LPCI Mode, Safety Evaluation

ML20006E277
Person / Time
Site:	FitzPatrick
Issue date:	02/05/1990
From:	Stoll C, Walz V, Warbis W POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20006E269	List: ML20006E268 ML20006E275 ML20006E276 ML20006E277
References
JAF-SE-90-024, JAF-SE-90-024-R00, JAF-SE-90-24, JAF-SE-90-24-R, NUDOCS 9002220496
Download: ML20006E277 (9)

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NO.-'JAF-SE-90-024 -REV. O h g ggy QA CAT. 1

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-EXPERIMENT X OTHER

, Change of RHR- .;

NUMBER Pump-P.ated Flow- {

to 8910 GPM in-LPCI Mcde TITLE:: Decrease of'RHR Pump Rated Flow by 10% in LPCI Mode ,

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The proposed modification, test, or experiment:

~ .1 - . '( ) Does -

Increase the probability of occurrence or conse-(X) Does Not quences of an accident or malfunction of structures,-

systems, or components important to safety previously evaluated in the FSAR.

1

-2. (-) ;Does -

Create the possibility of an accident or malfunction

-(X) Does Not of a different type than any evaluated previously '

-in the FSAR.

3. () Does -

Reduce the margin of sefety as defined in the basis-(X) Does Not for cny Technical Specification.

'4. () Does. -

Involve an unreviewed safety question based on 1, 2,- f

+.

. '(X)- Does Not and 3 above. 1 g , 5.- (X) Does -

Involve a change in the Technical Specifications- ,j Q () Does Not (Section(s) 4'.5.A.3 ).. j 6.. () Does -

Require pre-implementation review by the NRC.

(X) Does Not n

7. ( ): Does -

Degrade the Security Plan, Quality Assurance Program. ')

(X) Does'Not or the Fire Protection System. ,

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8. - .( ) Does -

Affect the environmental impact of the plant or (X). Does Not involve an unreviewed environmental question. i

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Preparer: C. H. Stoll (General Electric Co.)/ 2/5/90 Engineer Date  !

Distribution PORC Chairman. Reviewer: W. A. Warbis (General Electric Co.)/ 2/5/90

.Q.A~. Superintendent Engineer Date r

SRC Chairman (WPO) .

,j, Licensing (WPO)

Director, PES (WPO)

Approved: V. M. Walz M ./rd / 2/4/10 Technical Services Supt'. O Date i

l. Mod. Coordinator G.'Rorke~, WPO Reviewed: 90-011' / 2/6/90

-D. Burch, WPO. PORC Meeting No. Dct'e r

9002220496 900209 M I FORM.MCM14, 4.1 (MAY 1988) "

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NUCLEAR SAFETY-EVALUATION-

~ A '. - - . SCOPE OF EVALUATION F This evaluation demonstrates that the Residual Heat Removal. l k System (RHR) (including the Low Pressure Coolant Injection (LPCI)

U ' mode of operation): for the James A. FitzPatrick Nuclear Power i

~ Plant - (JAFNPP) is capable of performing its intended function-and i that there is no ' impact on the JAFNPP = Emergency Core Cooling '

System (ECCS) licensing basis for the following' condition:

The rated flow for RHR pump (s) ' operating ita the IPCI mode decreased from 9900 GPM to - 8910 GPM (a 10% reduction in rated LPCI flowrate) which results in a smaller (if any) y  : reduction in the other modes of RHR operation. 4 The functions - of the RHR system that could potentially be af- r fected are:-

a. To provide core cooling in the event of a Loss-of-Coolant  !

' Accident (LOCA) via the LPCI mode of operation,

b. To provide inventory makeup in the LPCI mode during 1 postu-lated events in compliance with 10CFR50 Appendix R,

.c. To provide torus water (suppression pool) cooling when L operating in the pool-cooling mode of operation, '

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. .;. . To - remove decay heat from the reactor vessel at low reactor-vessel' pressures in order to achieve and maintain cold shut-down of the reactor, and

e. To remove 1 heat from the ' drywell and wetwell in si'.aations 4 where it.is beneficial to do so.

Each function >is-assessed, herein, assuming RHR pump rated _ flow

, of 8910 gpm in the cLPCI mode'and a rated flow ' consistent' with this same pump performance when operating in the other modes ~of

~

l RHR operation. The assessment of the LPCI system performance R during LOCAs - also determines the impact of this change on the JAFNPP~ECCS licensing. basis.

B. REASON-FOR EVALUATION The~results of recent surveillance tests conducted by NYPA indi- I y cate - that the performance of RHR pumps A and C are near the L current technical specification limit which is based on required

LPCI- flowrate. The current technical specification requirements ,'

for the'LPCI flowrate are based on calculated results from older licensing evaluation n.odels which are overly conservative. The

current LOCA licensing basis for JAFNPP util&zes the SAFER /GESTR ECCS-LOCA methodology (Reference 1) and regulrements such as LPCI flowrate may be relaxed without having a significant impact on plant safety or ECCS limits. Therefore, the LPCI technical Page .

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specification.'flowrate requirements are more stringent than necessary when utilizing the newerLtechnology.

Sensitivity analyses performed for JAFNPP with the ' SAFER /GESTR

.model' (Reference 2) indicate a-maximum-increase in ghe limiting-licensing fuel peak cladding temperature (PCT) of 88 F for a - 10%

reduction in rated LPgI flow rate. Theo current limiting licensing.

PCT is more than 600 F below the 2200 F allowable limit. There-fore, JAFNPP would still meet all requirements of 10CFR50.46 and Appendix K of 10CFR50 with significant - margin even' with a 10%

lower LPCI flow rate.

An evaluation of a decrease in the LPCI system rated flow, which subsuntiates that no significant safety hazard would result, could reduce the potential for forced shutdowns during the operating cycle if indicated LPCI flow should decrease.

The purpose of this safety evaluation is to justU'y JAFNPP con- '

tinued power operation until the next refueling outage (currently scheduled for March- 31, 1990) with reduced LPCI flow (as low as 8910 gpm from one'RHR pump in the LPCI mode).

.C.- SAFETY EVALUATION 2

C.1 LPCI System Performance During LOCAs 4

The RHR pumps are aligned to the LPCI system and are dedicated to supplying emergency inventory makeup flow to the reactor vessel i upon occurrence of a LOCA signal. -l The LPCI system is an integral part of the ECCS that replenishes reactor vessel inventory during LOCAs that rapidly depressurize the vessel. A sensitivity study (Reference 2) was performed that i Jvaried LPCI system and other ECCS systen performance require.ments  !

for ~ the' JAFNPP. The- sensitivity study: demonstrated that a 10%

reductior-in L P C I - s y s t e m -- r a t e d flow. would. result in a maximgm increase ~in the licensing peak cladding temperature-(PCT) of 88 F and an insignificant increase in metal water reaction for the limiting large break accidents- with no change in fuel MAPLHGR 11 limits. For small break accidents, the requirements for the LPCI  !

' flow rate are less stringent than for large breaks because the -

loss'of coolant inventory is less significant and the fuel clad-  ;

ding heat transfer is higher- throughout the transient due to I

+1 steam cooling. '

Therefore, a decrease of the LPCI. system rated flow from 9900 gpm-to 8910 gpm has no impact on the JAFNPP licensing basis (the fuel MAPLHGR limits are unchanged) and the LPCI remains capable of performing its intended function during postulated LOCAs.

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JAMES A.J FITZPATRICK NUCLEAR PCE.3 PLANT l NUCLEAR SAFETY EVALUATION ]

.C.2 LPCI System Performance During Appendix R Events -

q The LPCI system is also relied upon to supply reactor inventory -

. makeup ' during postulated ' Appendix R events.' . These are not pipe j break events but' - postulated fire events- that can threaten the

' ability of the plant. to maintain reactor vessel inventory u depleted by decay hoat and sensible heat bolloff. '

Reference 5. documents an analysis of.a worst case Appendix . R i event which assumes an RHR pump is utilized in the LPCI mode to replenish inventory. The calculated guel peak cladding tempera-ture-(PCT) .during this-event was 1013 F. The Appendix R require-F ment occur ' for is toPCT prevent belowfuel f:F.1 adding damage which is not axpected to 1500-The PCT- for this event 5 analysis results) is estimated to in(based crease on' lessthe thanRefe5ence 60 F assuming a 10%

redugtion in LPCI flowrate which maintains a large margin to e 1500 F. Therefore, the ability of the LPCI system to perform this function in L compliance with Appendix R is not compromised by a 10% reduction.in rated LPCI flow.

C.3 RHR Pool. Cooling Performance During LOCA Events_

Another .'.nportant function of the RHR system is to remove heat from the suppression pool.: For this mode of operation the RHR.  ;

pumps are aligned to circulate pool water through a heat ex-  ;

changer an'd back to the . suppression pool. The most stringent.

requirements for heat removal from the pool occur during postu--

lated LOCA' events. A flowrate of 8000 gpm was assumed - for this-mode ' of: operation in pool temperature analyses (Reference 4). ..'

Since-the design:of the RHR pump is based on the higher flowrate  ;

requirements of LPCI,-the pump flowrate is normally throttled in '

the pool cooling mode to excessive flow. In the pool cooling -mode 'of operation, preventis there no i difference in ' elevation '

head and. pressure head between the system suction and discharge, unlike:the situation in the LPCI mode. Therefore, the' ability-to  :

. perform this function is-not compromised by a 10% reduction in rated.LPCI flow.

C.4 RHR Shutdown Cooling Performance i The RHR pumps may also be aligned to circulate reactor water through a heat exchanger for decay heat removal. The design

.flowrate-in this mode of operation.is 7700 gpm. In this mode of operation the. system is. capable of cgoling down the reactor

- vessel to cold shutgwn conditions (212 F) within approximately

- 20' hours and to 125 F within approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> additional time with a flowrate of 7700 gpm. Therefore, the ability to perform this function is unaffected by a 10% reduction in rated LPCI flow (from 9900 gpm to 8910 gpm per pump).

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-C.5 Containment Spray Performance  :-

Since the design of the'RHR pump is based on the higher flowrate

' requirements of LPCI,. the pump- flowrate in the- containment J

q 'drywell;and wetwell spray mode is less than in the LPCI mode.

l The! design buis LOCA containment response .,nalyses for .JAFNPP which consi(c the containment (drywell or wetwell) spray-systems are.the: analysis to. determine the allowable bypass leakage be- i tween the drywell and - wetwell airspace described in paragraph 5.2.4.4 of- the FSAR and the design basis accident analysis -

described in: paragraph 14 .'6.1. 3 - o f the - FSAR. The analyses ofL containment bypass leakage are performed to' show that~ the con-tainment response for design basis loss-of-coolant- accidents remains within containment pressure design . limits considering.' u wetwell sprays. -These bypass leakage analyses are primarily  ;

dependent on the time delay for operator initiation of the sprays '

and are-not strongly sensitive to the wetwell opray flowrate.

The -design basis accident containment response analysis in-4 paragragh 14.6.1.3 has-cases with and without containment sprays.

Case = D without containment sprays gives the highost (limiting) values for long term drywell temperature and containment pres-sure. It is judged that a 10% reduction in spray flow rate would i not result in the cases which consider sprays .to become more

' limiting than' Case D.'

, Therefore, based on engineering judgement, a 10% reduction in the

. containment spray flow rate should have' .no impact on the!

Fitzpatrick containment response analysis including the - bypass leakage analysis (paragragh 5.2.4.4) or the design basis accident  :

analysis (paragragh 14.6.1.3.3).

Some plants have considered use of drywell sprays to obtain a )

reduction in long term drywell temperature envelope for equipment qualification (EQ).. Also, the Emergency Operating Procedures' (EOPs)) based on BWROG Emergency Procedure Guidelines . (Revision 4) call for the operator to use drywell or wetwell sprays te control.

-the containment pressure or tsmperature !if these parameters approach design limits. .The drywell and wetwell spray systems typically have significantly more flow capability than that required for controlling the containment-pressure or temperature so that EQ envelopes which consider containment sprays and operator actions to mitigate pressure or temperature per EOPs

• q would-be expected to be unaffected by a 10% reduction in spray flowrate. .

C.6 Evaluation of the Effect on the FSAR b.-

C.6.1 Residual Heat Removal System (FSAR Chapter 4.8_)

y As described in section C.1, the LPCI system is capable of f  ;

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JAMES A. FITZPATRICK NUCLEAR = POWER' PLANT '

%" ~ NUCLEAR, SAFETY-EVALUATION p '

q I4 x . performing L its intended' ECCS function with a rated flow l , decrease from- 9900 gpm to 8910 gpm. . The LPCI- rated ' flow is- -!'

. addressed' in Table . 4.8-1 of Chapter 4.8 of the JAFNPP FSAR Update. Tf . the rated flow of a RHR pump were to ' decrease from 9900 gpm to 8910<gpm, the'flowrate.would be below the amount 4 specified' in this table. - As described in sections- "

C.2,.C.3,'C.4, and C.5 the . functions of -inventory makeup.

e s during Appendix R events, pool cooling and shutdown cooling.

with RHR, and the containment spray cooling with RHR are all  ;

unaffected with an RHR pump flowrate of 8910 gpm. Therefore, >

the rest of.this Chapter of the-FSAR-is unaffected by this- .,

condition, a B~ '!

C.6.'2 ECCS -(FSAR Section 6_)

This section of the FSAR discusses the intended function of p LPCI but references the SAFER /GESTR-LOCA report for calcu- -

l: lated performance results. As described in section C.1, the

l. LPCI system is capable of performing its intended ECCS function with- a rated flow decrease from 9900 gpm to'8910 i o

gpm. Therefore section 6 of the FSAR is unaffected by'this

. condition.

C.7 - Impact on Plant-Technical Specifications

,. The LPCI system' rated flow is referenced in'the JAFNPP Technical O! Specifications section. 4. 5. A. 3. Section ' C.1 above demonstrated that the ECCS licensing decrease -(1.e. ,. over 500}F margin to the 2200 F regulatory limit-asis'is unaffected-remains).. The fuel MAPLHGR limits in the JAFNPP Technical Specifications section 3.5.H are not restricted by LOCA analysis

but by fuel- design limits (14.4 kW/ft) and remain unchanged.

Therefore the margin to ' thermal- limits delineated within the JAFNPP. Technical Specifications would not be affected. '

If the LPCI flowrate were to decrease to 8910 gpm per pump, it

. would be below the LPCI flowrate specified in 4.5.A.3. However, this evaluation substantiatesethe fact that no safety hazard or significant degradation of safety margins would' occur.

10.8- Impact on Reload Evaluation TheLreload licensing document (Reference 6) provides the thermal limits for the respective cycle based on the licensed performance of'JAFNPP systems and equipment. The ECCS thermal limits reported 4 in J the reload analysis are those determined from the limiting-L. LOCA events. The sensitivity studies performed in Reference 2 demonstrate that the thermal limits during. limiting LOCA events

.(i.e., fuel MAPLHGR limits) are unaffected by a 10% decrease ~ in the- LPCI rated flowrate. Therefore, this condition would have no impact on the current and future reload licensing analyses.

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. JAMES As FlTZPATRICK NUCLEAR POWER PLANT i i NUCLEAR. SAFETY EVALUATION C;9 Summary of Safety Evaluation This Lsafety evaluction was performed ' to support continued power operation '.of until-the nextJAFNPP with*aoutage refueling 10% reduction of RHR (currently scheduled pump for flowrates March

. 31, 1990). This postulated condition has been -demonstrated to ~!

have.no impact on the capability of the RHR system to perform its intended functions. Additionally, it was demonstrated that this condition- would have no impact on the JAFNPP licensing ' basis .

documented in References 1 and 7 (i.e., no change in fuel MAPLHGR.

limits) . ; There - is no effect on RHR system and component safety bases. .as' defined in the FSAR.- A review of plant Technical Specifications.. to assess the effects on ' applicable Limiting ,

,, -Conditions.of Operation, Limiting Safety System Settings,: Safety I

Limits, and- roactor thermal parameters concludes that a 10% -1 decrease in rated flowrate does not significantly reduce the l margin of safety as defined in the bases for the- Technical

-Specifications. J C.10 Evaluation Summary

' Based on the'ab'ove' evaluation, it is determined that a decrease of 10% in RHR system = rated flowrate does.not constitute a sig- .

nificant- hazard- as defined in 10CFR50.92 for the following

. reasons:.

a. It .does not increase the probability of occurrence or the o consequences of an accident evaluated previously in the' safety analysis report A decrease in the rated flowrate is a performance condition that . is in response to accident condi-L tions. Therefore, this change has no impact on'the conditions i

, that would initiate an accident. An LPCI system flow reduc-h -tion was shown to have no significant impact on the ' JAFNPP

ECCS licensing basis and therefore does- not increase the

, consequences-of any accident analyzed in the safety analysis report.

l- b. It does not create a possibility for an 6ccident of a dif-o ferent type than any evaluated previously in the safety 1: analysis report. This is because the condition would be a l; change in performance for the response of the LPCI system to jA abnormal or accident conditions within the JAFNPP. Conse-quently, the conditions leading to such events are unaf-facted.

lU c. ,The margin of safety as defined in the basis for the Techni-

\ ' cal Specifications would not be significantly reduced. The

/ fuel MAPLHGR limits in JAFNPP Technical Specifications sec-D tion 3.5.H would remain unchanged. The margin of safety is P reflected in the operating limits and Limiting Safety System Settings of the Technical Specifications. The postulated LPCI Page 1

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- system . rated : flow . decrease . would' not change any of these - i

,' . lim i ts ; ; The ..,nsequences of transients; or accident events- l g, ,

have' been assesaed and- the. appropriate safety - -limits ' or .j i;u ' regulatory. requirements would-not be affected.

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'^D. -REFERENCES

-1. ~"JAFNPP- SAFER /GESTR - LOCA- Analysis", General ~ Electric Company, NEDC-31317P, October 1986.

2. "Sensitivityfof the JAFNPP Safety Systems Performance to Fundamental System Parameters", General Electric Company, b-- MDE-83-0786,-July 1986.

, 4.- "JAFNPP_ Suppression Pool Temperature Response",

General Electric Company, NEDC-24361-P, August 1981.

, 5.- " Analysis to' Extend Operator Action Time For Alternate Shutdown Panels In Support Of Fitzpatrick Compliance To.

Appendix R", General Electric Nuclear Energy, MDE-137-0585,

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November 1985.

6.. " Supplemental Reload Licensing Report for JAFNPP Reload 8 (Cycle 9)", General Electric Nuclear Energy, 23A5898, June 1988.

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