ML20055D002

Discusses Util 900514 Application for Amend to License NPF-47,increasing Suppression Pool Temp from 95 F to 100 F. Requests That Util Demonstrate Applicability of Items Addressed in Encl Hatch Safety Evaluation to Facility

ML20055D002
Person / Time
Site:	Hatch, River Bend
Issue date:	06/11/1990
From:	Abbate C Office of Nuclear Reactor Regulation
To:	Deddens J GULF STATES UTILITIES CO.
References
NUDOCS 9007030062
Download: ML20055D002 (2)

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June 11c 1990

.Docke't No. 50-458 Gulf States Utilities ATTN: Mr. James C. Deddens Senior Vice President (RBNG)

Post Office Box 220 St. Francisville, Louisiana 70775

Dear Mr. Deddens:

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SUBJECT:

APPLICATION FOR AMENDMENT TO LICENSE TO INCREASE SUPPRESSION POOL TEMPERATURE FROM 95 F TO 100 F Oy letter dated May 14, 1990, Gulf Stat (s Utilities (GSU) filed an application for a license amendment to increase the eximum temperature for the suppression -

pool from 95 F to 100*F. The Edwin I. Hatch Nuclear Plant (Hatch) also requested a similar amendment in 1986 and received staff approval in 1989. Because the increase in suppres;lon pool temperaturn changes the loads identified in the USAR, t the staff requests that GSU oemonstrate applicability of the items addressed in _

the Hatch Safety Evaluation (SE) (Enclosure 1) to the River Bend Station or provide additional inforuetion to address items identified in the Hatch SE. The _

information will facilitiate the staff's review. If you nave any questions, please call me at (301) 492-1322. -

Sincerely, Or.'g!nal VM %,

Claudia M. Abbate, Project Engineer Project Directorate IV-2 Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation J-

Enclosure:

DISTRIBUTION:

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, Mark Wetterhahn, Esq. Ms. H.' Anne Plettinger y Bishop,' Cook, Purcell & Reynolds - 3456 Villa Rose Drive r 1401 L-Street,.N.W.

Baton Rouge, Louisiana .70B06 - t Washington , D.C.. 20005 1

'Mr. Les' England Director - Nuclear Licensing Gulf States Utilities Company P. O. Box 220

%. St. Francisv111e, Louisiana 70775 i h- Mr. Philip G. Harris V Cajun Electric Power Coop. Inc.

10719 Airline Highway P. O. Box 15540 Baton Rouge, Louisiana 70895

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Senior Resident Inspector P. D. Box 1051 St. Francisville, Louisiana 70775 '

President of West Feliciana

. Police. Jury 1

P. O. Box 1921

• St. Francisv111e, Louisiana 70775- 1 l3 .

Regional Administrator, Region IV '

f U.S. Nuclear Regulatory Commission 611 Ryan Plaza' Drive, Suite 1000 Arlington, Texas 76011 e

, Mr. J. E. Booker s

Manager-Nuclear Industry Relations .!

7 P. 0. Box 2951 Beaumont, Texas 77704 '

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, Mr. Willian H. Spell, Administrator Nuclear Energy Division Office of. Environmental Affairs

,%<t JP, 0.. Box 14690 4 -3 Baton Rouge, Louisiana 70898 '

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1-t Mr. "J. David McNeill, III William G. Davis, Esq.

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i NUCLE AR REGULATORY COMMISSION usmNGTON O C 20555 5AFETY EVALUAT10N BY THE CrFICE OF hlCLEAR REACTOR REGULAi!Ct; SUPPORTING AMENDMEhT N05.165 AND 102 TO F AtillTY CPERATINC LICENSES DPR-E7 AND NPF-5 GE0kCIA POWEP COMPANY OGLEThCRPE POWER CCRFORAilCh Ml't.'.ClTX ELECTRIC AUTH0 TIT Y OF GEORG1 A CITY OF LALTON, GEORGI A EDklh 1. HATCH hUCLEAR PLANT, UhlTS 1 AND 2 D,0C KET 1105. 50-321 AND 50-366 1.C !h'RCDUCTION By letter catec September 6,1906, Georgia Power Company, the litersee for the Ecw ; n I. Hatch Nuclear Plant, Units 1 and 2 requested changes te Technical Specifications (TS) 3.7 ano associated Bases 3.7. A.1 for Hatch Unit 1, and 15 3.6.2.1 anc t. .C .2.1 ar.d ossociated Bases 3/4.t.2 for hatch- Uni t 2. These specifications deal with the limiting conditiors of operation (LCO) of tN suppression pool (SP) during normal plant operation at cor.ditions 1, 2 and 3 for tcth the units and the associated surveillance requirement for Unit 2.

Specifically, the croposed u1ange would raise the suppression pool temperature lircit during norir.11 operation f rom 95' F to 10C' f. The 105" F limit on allev.atale pool ter.perature during safety system testing, which adds heat to the cuppre.ssion pool, will not be changed. Alsc, the suppression poci terrperature limits (SPTL) requiring immediate plant shutdown (110 F) and vessel cepressuruction (120' F), will retrain unchanged.

In recent years, high sumertime teraperatures have caused the temperature of the Altamaha River, which serves as the ultirkate heat sink fcr the plant sersice water anc residual heat removal (RHR) systems, to rise to the point where an insufficierit differential temperature is available to naintain the suppession poc1 temperature beim 95* F. R equ e s t for emergency relief f ror' the TS LCO has been iminent on a number of occasions, and processins of an energency TS change to increase the 95' f limit was in progress (trir.g August 1987 when the LC0 was cleared.

To avoid the necessity of submitting energercy TS changes regarcir.g the 95' F i hnit , the licensee proposes to raise the limit from 95* F to 100 F du ri ng normal o ration. In support of this increase in the suppression pool ten-perature limit during normal operation, tre licensee pros ided the General f E'trt ric (GE) Ccnpiny's safety evaluet icn (EAS-E-0388, dated March EEC' c f f the suppression pool ternperature limit for l'ork I containent and its appii-p . catiiity to Fatch Units. The 60 report discussed the iraputt o' the proposed L. in crea se it. the pool's operational temperature limi t on (1) cor.ta' nwrt response l (' '

(2) safety-relie f s alve (SRV) operation, (3,' emrgency core coo'.ir<g sy stw. I FCCS$

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perf arnar.ce , ( d) tiPSF fcr saf ety !.y stem purrps, ( S) Hf.ch emergeng cp ra ting procedures (EOFs), and (f!) anticipated transm t withou* scram MTW5, l

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l 2.0 EVALUATION .

The events which involve \the suppression pool can be divided into 'two general'

.h categories: safety relier valve (Sav> eischarge to the po61 via the Say y discharge lines and T-Quenthers, and discharges to the pool via the drywell to wetwell vent pipes during design basis loss-of-coolant accidents (LOCA). These W are evaluated in sections 2.1 and 2.2 below, j;

f' 2.1 LOCA-RELATED C0hTAINMENT LOADS The GE safety evaluation of the suppression pool (SI) temperature limit for Hatch Uniti. I and 2. discussed the ranges for operational temperature limits for SP! water under LOCA conditions to ensure that containment pressures and temper-j atores and hydrodynamic loads under such condit40ns do not exceed the design d values. The GE evaluation concludes that a normal operating suppression pool 3 tenperature up to 100' F for the Hatch uniti, will not affect the design loads.

The following paragraphs (a) through (d) sumarize thef.e evaluations and discuss their erplication to the Hatch units..

1. (a) Containthent Pressure and Temperature E'esign Limits The GE report compared the pressure and temperature design limits for several Mark I plants (including Hatch) to the predicted maximum p, containmentipressure and temperatures during a LOCA. The report noted j

- that because the design limits are very high for such containments, there

.. is a large margin between the predicted values under LOCA conditions and the design values that would support a large increase in the normal operational pool tenperature. Specificaib , the report pointed out that T based on design pressure and temper.ature consideration alone, an

. operational pool temperature in the range of 133' F to 1616 F should be

l. acceptable, t (b) Steam Condensation

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With rbgard to' the ability of the suppression pool to ensure complete

! steari tendensation following a LOCA, the report stated that based on an analyris of test data-for the Mark I full scairc test facility (FSTF), GE determinui that a norrcal operational pool tempcrature:in the range of 118' F to 133' F would ensure complete steam condensation because it would y correspond to the tested maximur. pool temperatures f or which complete steam condenution was confirmed.

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p, c) Condensation 0:cillatiot+ Loads 1// ' 1 gq-

~j, The report pointed out that condensatlon oscillation'(CO) loads are f' prirrarily af fected by two hydrodynamic ph rannters, i.e. , pool terrperature f and the enthalpy fin throJ9h the downComer vents. Using the

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3 GE-developed correlation between these two parameters and the CO ' loads under transient conditions, the CO loads for the expected LOCA conditions and the conditions sim lated during.the FSTF test were determined and '

, corrpared with plant-specific predictions to determine the margin between the expected and the design C0 loads and, subsequently, the associated margin in the pool temperature. The licensee stated that consideration of Hatch plant-sprcific bounding hydrodynamic parareters would result in a CO load that is less than thet assumed in the containment loads evaluation even w:th a normal operational pool temperature.of.110* F (the shutdown limit).

(d) Chuggins Loads The GE report stated that a review of chugging data obtained during the Hark 1 FSTF tests (NEDE 24539-F) indicated that chugging occurs only with .

sr..all-break LOCAS and relatively low pool temperatures (less than 135' F).

The report concluded that the _proposeo increase in the normal operational pool. tmperature limit will have no impact on chugging loads. <

On the basis of the GE information, the staff concludes that the j LOCA-related containr.ient loads resulting from the proposed increase in. -l normal operational pool teraperature limit will be within the containment- '

design 1 cads.

2.2 SPV 0FERAT10NAL LOADS e

The SRV operational loads can be divided into two categories. The SRV air clearing load and SRV condensation loads..

< - (a) SRV Air Clearing Loads l The SRV air clearing . loads result from the expulsion of air out of the l SRV discharge line into the. suppression pool. The expansion and contraction. A cf the air bubble creates an oscillatory load on the containnent wall- and submerged structures. The SRV air clearing load will increase with a Ugher initial pool temperature. However, the staff notes that the US Marr containment program requires that the limiting SRV air clearing loat: te be considered in containment structural evaluations be determinec

• on cne basis of the first actuation of an SRV at-the maximum pool teroperature

permitted by the Mark I plant TS (120' F) with the reactor at operating ,

pressure. The Hatch units also have the same TS limit for suppression pool that would require the reactor to be depressurized. Therefore,-the

' i' stuff agrees'with the licensee that the SRV air clearing load resulting y . f rom the proposed increase of normal operational pool temperature from I/ 9E* F to 100* F will be bounded by the liniting SRY air clearing load for the Hatch units.

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'(b) SRV Condensation Loads The licensee referred to GE Topical Report NEDO-30832, " Elimination of Lirdt on BWR Suppression Poci Temperature for SRV Discharge with Quenchers" submitted to the NRC by the BWR Owners' Group in March 198E.

This report had concluded that the local pool temperature limits for the suppression pool to ensure stebu condensation under stable conditions during SRY steam discharge into the pool specified in NUREG-0783,

" Suppression Pool Temperature Limits for BWR Containments" dated

. November 19P.1, could be eliminated for BWRs that utilized T or X-quencFer

-devices. GE concluded the above, baseo on their f 4ndings (tabulated in the NE00-30832 repert) that the SRV condensation loads with the above-devices were low in comparison with other loads (e.g. , SRV air clearing loads) considered in containment structural evaluation. The staff has not yet completed its evaluation of the above : report. Therefore, for this safety evaluation, the staff has used the criterion for local pool teoterature limit during SRV steam discharge into tFe pool that is identified in hUREG-0783 to assess whether the peak. local pool terperature resulting from the proposed initial pool temperature of 100' F will meet the criteria given in the NUREG. In January 1983, and in February 1983, the licensee provided plant-unique analysis reports for Hatch, Units 1 anc E long term centainment programs, in-these reports, using an initial pool temperature of 95* F and other hydrodynamic paraneters, the licensee calculated a bounding local pool temperature of 199* F for the Hatch units during transients involving SRV actuations.

The licensee concluded that the Hatch units, therefore, complied with-the NUREG-0783 limit for local pool temperature during SRV steam discharge inte the pool (200 F). Based on the review of these reports, the staff' concluced (SER, dt,ted January 25,'1984) that the licensee employed a conservative methodology to analyze pool temperature transients-involving SRV actuations to demonstrate the plant's compliance with NUREG-0783. The staff, therefore, found the calculated temperatures acceptable.

Ey providing credit for quencher submergence as allowed by the NUREG, the staff has reevaluated the local poc1 temperature limit for the Hatch units, and concluded that a limit of 204' F is appropriate (Hatch units have about 8 feet quencher submergency; steam flux through quencher perforations is less than 42 lbs m/f te-sec, when the peak local pcol temperature is reached). The staff has determined that the proposed

-increase of operational pool temperature by 5* F will not result in a peak pool local temperature higher than the estimated allowable limit of 204' F. Thtrefore, the staff concludes that there is reasonable assurance that the proposed normal operational pool tenperature limit of 100* F will not compromise the ability of the suppression pool to condense steam under stable conditions during SRV discharge of steam into the pool and, therefore, meett the criteria of NUREG-0783. Furthernore, the staff notes that the

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proposed TS changes will not alter the existing reouirements for (1) poci .l cooling whenever the pool temperature exceeds 100* F, (2) scramming the 3 reactor whenever the pool temperature exceeds 110' F, and (3) depressurizing the- reactor whenever the pool temperature exceeds 120' F.

T.? E CCS PERF0D%NCE The core cooling capability of the ECCS pumps is determined by the ability to keep the peak clad temperature of the fuel to less than 2200* F for all postulated  ;

loss of coolant accident (LOCA) events, considering an arbitrary single failure. '

For -the Hatch units, the inost limiting LOCA event is a large break in the L discharge line of. the recirculatior loop coupled with a single failure of the i low pressure coolant injection (LPL1) valve on the other loop. For this postulated event, the two core spray pumps are the only effective means for L core cooling.

l The GE report (EAS-19-0388) presented the results at an LCCS analysis using l 110' F as the initial pool temperature instead of the 95* F used in the original ECCS calculations. The results indicate that there is no significant l- impact on the LOCA analysis. Thus, the proposed TS change would not adversely affect ECCS performance.

On the basis of the GE information, the staff concludes that ECCS performance i will remain within the limits set by 10 CFR 50, Appendix K, and thus is i accep tab le .

F.4 NPSH FOR SAFETY SYSTEF PUMPS In accordance with Regulatory Guide 1.1, it is required that the RHR and core i spray pumps have adequate net positive suction head (NPSH) without dependence i on positive containment pressure during the worst case LOCA with a single l failure. '

~ l The initial NPSH calculations for the Hatch units were performed using ari .

l initial suppression pool water temperature of 95' F and assuming that all the energy in the reactor pressure vessel was absorbed by the suppression pool water following'a LOCA. Using these and other assumptions, the peak ,

suppression pool temperature was calculated to occur at 6.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> following a I LOCA. At that tirre, the NPSH margins for both the RHR pumps and the core l spray pumps were determined to be adequate (3.94 f t. and 1.34 ft. , respectively). l The GE report (EAS-19-0388) presents the results of a re-analysis using all of the assumptions of the initial analysis except that the initial pool temperature was assumed to be.110 F and realistic energy source terms were used. The 4 eriergy it:put to the suppression pool was taken to be the blowdown energy from l the LOCA plus decay heat calculated using the May-Witt decay heat correlation, I l

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-E-which includes.a 101 factor for conservatism. The energy input also was calculated using the 1979 ANS decay heat correlation which represents the best estimate decay heat correlation', and results in a calculated peak pool temperature of about 190' f.

Using the revised assumptions and the May-Witt decay heat correlction, GE calculated that the maxirrum suppressior, pool temperature would be approximately 2120 f which would still result in adequate NPSH for the RHR pumps. At this temperature, the core sprey pumps may operate with some cavitation since- the i required NPSH is about 0.05 feet higher than the available NPSH. i however, the GE report points out that the time at which the peak pool terrperature occurs is more than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> into the accident, by which tire only about 10% of the core spray rated flow would be required to remove-the decay heat.- The >

recuired NPSH at such reduced flow is significantly less than the NPSH required at full fitv. The report also notes that Revision 4 to the Emergency Procedure Cuidelines (EPGs) instructs the plant operators to reduce the ECCS pump flow and to turn off unneeded pumps when adequate core cooling is assured. The GE report concludes that, based on the actual NPSH requirements for the core spray '

pumps at high water temperatures and'the required mode of pump operation, the .

increase .in initial pool temperoture will still result in adequate NPSH for the '

core spray purrps. - -

Eased on the GE report, and noting the conservatism built into the Hay-Witt currelation plus the fact that the calculation was run using 110' F rather -

thcn the proposec 100' f as the initial pool temperature, the staff concludes that the CHR ano core spray pumps will have adequate NPSH. The NPSH evaluation is limited to the PHR and core spray pumps because neither the NPCI or'the RCIC pumps would be operated beyond 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> into a LOCA event. The peak pool t teroperoture and the resultant minimum NPSH availability do not occur until after 6 bcurs into the event.

Ttt staff therefore concludes thet the increase in suppression pcol temperature-requested by the licensee would not have an adverse impact upon' the operation of the safety system pumps.

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i 2.5 EMEPCENCY OPERATING PROCEDURES (EOPs)

TFe GE report points out, ccrrectly, that the proposed change in the suppression poc1 temperature limit would result in some needed changes to the E0Ps. .However, the staff is not now reviewing the adequacy of E0Ps prior to implementation.

Thus, this SER does not address charces to the E0Fs. As a matter of interest ,

however, the licensee now is revising the Hatch E0Ps to be in accordance with Revision 4 to the Emergency Procedure Guidelines (EPGs). The staff expects that any changes to the E0Ps required as a result of this proposed change will be incorporatec as a part of the ongoing E0P rcvision, which will be subject to later staff inspection for adequacy.

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' 4,1 2.0 ' ATWS EVALUATION The TS for each of the Hatch units now require that the. reactor be scramed by placing the mode switch in the Shutdown position wher.ever the suppression pool temperature exceeds 110' F. This TS requirement is not changed as a result of the requested TS amendment. Therefore, we conclude that the proposed change has no irnpact on the Aik5 evaluation.

2.7

SUMMARY

In sumary, the staff has examined the impacts of the proposed TS changes on (1) LOCA-related containna:nt loads, (2) safety-relief valve (SRV) operational-(4) NPSH for safety system punps, loacs, (5) Emergency (3) ECCS performance Operating calculations (6) ATWS evaluation, Procedures, and and has con that the proposed changes are acceptable.

3.0 ENVIRCimENTAL (01.SIDERAT10N These anendr4r.ts involve changes to the installation or use of f acility-components located within the restrictecLarea as defined in 10 CFR Part 20.

The staff has determined that the amendments involve no significant increase l' in the amounts, and no significant char.ge in the types, cf an? ef fluents that '

I may be released offsite, and that there is r.o significar.t increase in individual or cuculative occupational radiation exposure. The Cormission has previously issued a propmed finding that the amendmerts involve no L sigtificant hazards consideration and there has been no public comment on such E f i nd i r.g . Accordingly, the amendments taect the eli l

categcrical exclusion set forth in 10 CFR 31.22(c)(giaility

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criteria 9). Pursuant to 10 for CFR l

51.2E(b), no environmental 1 npact statement or environmental assesstcnt- need be prepared in connection with the issuance of these amendnents.

4.0 CONCLUSION

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. The Cone.ission made a proposed determination that1 these amendments involve no sigr.ificant hazards considt. ration which was published in the Federal Register on Novenber 2,1988 (E3 FR 44251), and consulted with the state of Georgia. No L public con. rents were received, and the state of Georgia did not have ar.y

ccntrents.

! We'have concluded, based on the considerations' discussed above, that: (1) trere is reasor.6ble assurance that th( 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 Comndssion's regulations, and the issuance of the amendn.ents will r.ct be inimical to'the corinon defense and security or to the health crd safety of the public.

Principal Contributors: Raj K. Anand, SPLB, DEST, i;r.P George Thomas, SRXE, CEST, HRR Lawrence P. Crocker, PD 11-3, DRP I/11, IMR I

Cated: July 18,1989 .

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