regulat(. f information distribution system … · 69-10430, 08/08/94 ~ idap ts3.id2 page 4 of 4...

REGULAT(. f INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9608)00'112 DOC.DATE: 96/08/13 NOTARIZED: NOFACIL:5()-27~Diablp Canyon Nuclear Power Plant, Unit 1, Pacific Ga

5(q323 Diabfp Canyon Nuclear Power Plant, Unit 2, Pacific GaAUTH.NW$k AUTHOR AFFILIATION

RUEGER,G.M. Pacific Gas a Electric Co.RECIP.NAME RECIPIENT AFFILIATION

Document Control Branch (Document Control Desk)

SUBJECT: Forwards completed Licensing Basis Impact Evaluation ofFSAR Update change which contains SE performed IAW10CFR50.59 re reanalysis of inadvertent ECCS actuationaccident.

DISTRIBUTION CODE: A001D COPIES RECEIVED:LTR ENCL SIZE'.TITLE: OR Submittal: General Distribution

DOCKET I0500027505000323

NOTES:

RECIPIENTID CODE/NAME

PD4-2 LABLOOM,S

COPIESLTTR ENCL

1 11 1

RECIPIENTID CODE/NAME

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INTERNAL: ACRSNRR/DE/EMCBNRR/DSSA/SPLBNUDOCS-ABSTRACT

EXTERNAL: NOAC

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NOTE TO ALL "RIDS" RECIPIENTS:PLEASE HELP US TO REDUCE WASTEl CONTACT THE DOCUMENT CONTROL DESK,ROOM OWFN 5D-5(EXT. 415-2083) TO ELIMINATE YOUR NAME FROM

DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 13 ENCL 12

1

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'I

Pacific Gas and Electric Company

August 13, 1996

PGKE Letter DCL-96-177

245 Market Street, Room 937-N9B

San Francisco, CA 94105Blaiiingcirlcln'ss

Mail Code N9BP.O. Box 770000San Francisco, CA 94177415/973-4684 Fax 415/973-2313

Gregory M. RuegerSenior Vice President andGeneral ManagerNuclear Power Generation

U.S. Nuclear Regulatory CommissionATIN: Document Control DeskWashington, D.C. 20555

Docket No. 50-275, OL-DPR-80Docket No. 50-323, OL-DPR-82Diablo Canyon Units 1 and 2Regnal sis of Inadvertent Emer enc Core Coolin S stem Actuation Accident

Dear Commissioners and Staff:

On May 8 and 9, 1996, PG8 E informed the NRC via telephone conversationsthat we were reanalyzing the inadvertent emergency core cooling systemactuation accident described in Chapter 15 of the Diablo Canyon Final SafetyAnalysis Report (FSAR) Update. As discussed in those conversations, the newanalysis predicts a limited amount of water relief from the pressurizer safetyvalves (PSVs). The Diablo Canyon PSVs have been demonstrated to beoperable under these conditions by the Westinghouse Corporation based ontesting performed by the Electric Power Research Institute.

PGKE has completed a Licensing Basis Impact Evaluation (LBIE) of the FSARUpdate change. The LBIE contains a safety evaluation performed in accordancewith 10 CFR 50.59. The results of the safety evaluation concluded that thischange does not involve an unreviewed safety question.

We are enclosing the safety evaluation for your information.

Sincerely,

V.Gregory M. Rueger

cc: Steven D. BloomL. J. CallanKenneth E. PerkinsMichael D. TschiltzDiablo Distribution

I[a (

Enclosure H

RLJ/371PDR ADQCK 050002759608200ii2 960813 '

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PGKE Letter DCL-96-177

ENCLOSURE

LICENSING BASIS IMPACT EVALUATION

69-10430 08/08/94 ~IDAP TS3. ID2

Page 1 of 4

TITLE: LICENSING BASIS IMPACT EVALUATION (LBIE) SCREEN

REFERENCE DOCUMENT No. 5. Doc. Rev. No.10(i.e., indicate the Procedure Number, DCP Number, or other reference document forwhich the Screen is done, including the document revision number or date).

Reference Document Title

Sponsoring Organization: Sponsor: 1

(Print)

DESCRIPTION

Summarize the proposed activity or CTE and how it differs from the presentlyapproved condition. The reason for the proposed activity or CTE should also be

described. Cite applicable drawings and other documents as necessary todescribe the current condition. , Briefly describe how the issue may interfacewith the licensing basis (documents}.

The change being addressed is in the FSAR Chapter 15 Accident Analysis forSpurious Safety Injection Actuation. The previous analysis does not addresspressurizer overfill. The neo analysis demonstrates that overfi I I does occurprior to SI termination. but the pressurizer safety valves sill operate reliablyfor the fluid conditions resulting during the eater release period.

The new Spur ious Safety Injection System (SIS) analysis was performed in responseto IYestinghouse Nuclear Safety Advisory Letter (NSAL) 93-13. This NSAL

identified to A'estinghouse customers that pressurizer overfill was of greaterconcern during a Spurious SIS event than previously believed. Earlier0'estinghouse rework had failed to account for decay heat in a conservative manner.This applies to EPRI NP-2296, which +as authored by IY'estinghouse and isreferenced in our Spurious SI SER. The contribution of POP floe to the analysiswas also non-conservatively ignored, as documented in NSAL 93-13. Supplement 1 ofOctober 1994. The concern associated /4ith overfilling the Pressurizer is thatoriginally the Pressurizer Safety Valves (PSV) vere not qualified, for eaterrelease. Their failure mould r'esult in a small break loss of coolant accident(SBLOCA). The PSVs have subsequently been qualified for eater release for thebrief period and high RCS fluid temperatures predicted for this scenario.

Higher temperatures are advantageous for PSV eater release operation, becausethey promote fluid flashing as opposed to eater slug floe. A large degree ofsubcooling leads to valve chatter and instability that can damage valve seats andresult in incomplete reseating. The extent of'alve damage after subcooled I4/aterdischarge is difficult to quantify. If the leakage rate is less than the normalmake-up flot4/, it is not technically a Small Break LOCA (per ANS 51.1/N18.2-1973).

The current analysis demonstrates that eater release begins at 12 minutes 5.5seconds after the SI is initiated. The fluid temperature at that time is 6'16'.8

F. At 16'.5 minutes, the fluid temperature has reduced to 603.2 F. Operators canbe assumed to terminated SI by 16 minutes. Five operator crees, timed on theOCPP simulator, demonstrated times between 10 minutes. Z6'econds and 13 minutes,25 seconds to terminate 1 CCP floe. An additional minute was needed to establishnormal charging controls. The average of the five cree times pIus 1 minute forcharging controls leads to an estimate of 13.5 minutes to terminate the event.Vse of 16 minutes conservatively bounds this response time. An additional 0.5

«II ~

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minutes'is assumed to conservative bound the period betheen SI termination andthe end of water release.

OCPP PSVs are Crosby 6M6 valves of the same type qualified for hot water releasein O'CAP 11677. OCAP 11677, "Pressurizer Safety Relief Valve Operation for A'aterDischarge During a Feedwater Line Break," was performed For the benefit of 35units that had accident analyses demonstrating pressurizer overfill following afeedline break. OCPP units 1 and Z were not among those 35 units, butIr/'estinghouse has verified that the A'CAP is applicable to OCPP's valves. Thetemperatures and duration predicted for water relief during the OCPP Spurious SIevent are bounded by those predicted for water relief at some of the &CAP unitswith Crosby 6M6 valves, specifically. Seabrook 1 & 2 estimated relief fluidtemperature varies from 605 F down to 603 F, and is assumed to occur for 17minutes.

k'estinghouse has reviewed the O'CAP and OCPP data and concludes that OCPP's PSVswill be operable for water relief during the worst case Spurious SI pressurizeroverfill transient. JYVREG 0737, item II.O.I, states that "licensees andapplicants shall conduct testing to qualify the reactor coolant system relief andsafety valves under expected operating conditions for design-basis transients andaccidents." Nestinghouse letter PGE-96-565 verifies that the EPRI testingprogram summarized in O'CAP 11677 satisfies this requirement for PG&E, as it hassatisfied the same requirement for many other utilities.

It is also noted that the single failure criterion does not apply to the PSV.&'estinghouse states that within their IYRC-approved methodology, spring operatedrelief valve failures are considered passive failures. Passive Failures do notneed to be assumed for short term accident responses per AP$ -5B.9-19BI.

The licensing impact will be, first, that FSAR Update section 15.Z.14 will nowreference the new Spurious SI analysis. Second. operator action will be creditedfor SI termination within 16 minutes. Other documents, such as training materialand the Accident Analysis Profile, must be updated.

SCREENING FOR DETERMINING THE NEED FOR PRIOR REGULATORY AGENCY APPROVAL

Does this activity or CTE involve a change to the Facility OperatingLicense (OL), including OL Attachments (Technical Specifications,Environmental Protection Plan and Antitrust Conditions)?* If "Yes". submit an LAR to the NRC and continue this Screen subject

to the appr oval of the contents of the LAR. LARg Do notrelease the Reference Document above for use. construction. etc.,until the LA is received. The originator of the Reference Documentshould provide a reconciliation between the LA and LAR to the PSRCto justify release for use, construction, etc.

Is the Reference Document a procedure?( If "No", skip the next question.)

Yes No

( )~ (X)

( ) (X)

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Yes No

Does the Procedure Commitment Database (PCD) contain any commitment ( )** ( )

to a Regulatory Agency that must be changed and which would either:

a) Require notification to that agency, orb) Require prior approval from that agency?

** Follow the requirements of IDAP XI4. ID2, Commitment ChangeProcess. Continue this Screen subject to the contents of therequest for prior regulatory approval. .Requesting document Jf

If no prior approval is required. continue the Screen.

SCREENING FOR DETERMINING THE NEED FOR A SPECIFIC EVALUATION

For the activity or CTE under consideration answer the following questions. Any"Yes" response (except for the, answers to items 3.a and 4.a below) requires theappropriate sections of Form 69-10431 (LBIE) to be completed.

Yes No

'SECTION 1. 0 F a 3

a)

b)

'c)

Does it involve a change to the facility design, function ormethod of performing the function, as described in the SAR.

including text. tables and figures+ and including the FireProtection Program ( FSAR Update, Section 9 ') and QualityAssurance Program (FSAR Update. Chapter 17)? (+See Appendix7.5 of TS3. ID2)

Does it involve a change to procedures, system operation. oradministrative control over plant activities as described inthe SAR, including procedures related to the Fire ProtectionProgram (FSAR Update. Section 9.5) and the Quality AssuranceProgram (FSAR Update, Chapter 17)?

Does it resul.t in a test, experiment, condition, orconfiguration that might affect safe operation of the plantbut was not anticipated, described, or evaluated in the SAR?

(X) ( )

( ) (X)

( ) (X)

SECTION 2. ntal Pr

a)

b)

Does it involve changes to or new effluents discharged to theair, fresh water, sea water. or land?

Does it involve a change to the quantity* or use or storage ofmaterials classified as hazardous (including oils) or thegeneration of hazardous wastes? (*See Paragraph 5.4.2 ofTS3. ID2)

( ) (X)

( ) (X)

c) Does it result in a disturbance of previously undisturbedland?

d) Does it alter surface water runoff patterns or amounts?

e) Does it involve work within the SLO-2 archeological site

( ) (X)

( ) (X)

"( ) . (X)

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boundary?

SECTION 3. m r enc

Yes No

a)

b)

Does the Emergency Plan (EP) require review on the basis ofAppendix 7.1? If "No", skip the next question and signature.

If "Yes", does the activity or CTE result in a change to theEP?

( ) (X)

() ()

Emergency Plan Reviewer Signature / Date

SECTION 4. S c iYes No

a) Do any of the security plans (PSP. SCP,, STOP) require review ( ) (X)on the basis of Appendix 7.2? If "No", skip the nextquestion and signature.

b) If "Yes", does the activity or CTE result in a change to a

Security Plan?

If so, which plan(s)?

() ()

Security Plan Reviewer Signature / Date

REMARKS: For each Screen Section above having all "No" answers, provide the logicfor the "No" answers if clarification is required.

The proposed change is to reference an additional analysis in FSAR Update Chapter15. It does not affect operating procedures or the physical condition of theplant: therefore, it does not impact the security plan, emergency plan, orenvironmental protection. The analysis does credit a specific time for Operatorsto respond to an inadvertent SIS, and it credits PSVs to pass eater for up to 4.4minutes and reseat. This can be interpreted to mean a change in the method ofperforming a function as described in the SAR.

REFERENCES/ATTACHMENTS:

FSAR Update Section 15.2.14 mark up0'estinghouse Project Letter PGE-96-565, Nay 31, 1996NSAL 93-13NSAL 93-13. Supplement 1

Based on the above criteria, I have determined that an LBIE is ~ is notrequired.

Preparer Sig a ure ate

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Independent Technical Reviewer Signature Oate

Based on my.indep ndent te hnical review, I concur with the above conclusion.'4 . ~ t nl"~c

t69-10431. 08/08/94 . ~PACIFIC GAS'AND ELECTRIC COMPANY

NUCLEAR POWER GENERATIONIDAP TS3. ID2

Page 1 of 5

TITLE: LICENSING BASIS IMPACT EVALUATION (LBIE)L

REFERENCE DOCUMENT No.(i .e., indi cate the Procedure Number . DCP Number. or other reference document forwhich the Evaluation is done, including the document revision number or date).

Reference Document Title:

Sponsoring Organization: Sponsor:(Print)

As a result ofLBIE have beenevaluation.

the LBIE Screen ( Form 69-10430), indicate which sections of thiscompleted and are attached. Refer to TS3. ID2 to complete each

[Xj SECTION 1

and[ 3 SECTION 2

[ j SECTION 3

[ 3 SECTION 4

10 CFR 50.59 Safety Evaluation (including 10 CFR 50 '4(a)(3)OL Condition 2.C.(5)b./2.C.(4)b. Evaluations)Environmental Protection EvaluationEmergency Plan Evaluation - 10 CFR 50.54(q)Security Plans'valuation - 10 CFR 50.54(p)

Explain why this LBIE is being performed (i.e., Why were Screen questions answered"Yes"?)

The proposed change is to reference a new analysis for Spurious Operation of thesafety injection system (SIS) at poser, FSAR Update 15.Z.14. This is an ANS

Condition II event, an event of moderate 'frequency, that should at worst result ina shutdown with the capability of restarting the unit. The new analysis creditsoperator termination of the event at 16'inutes and I4/ater relief through the PSVsfor up to 4.4 minutes. The PSVs must reseat to avoid an unisolable small breakLOCA. which is an ANS Condition III event. The neo analysis is a result of a

Westinghouse Nuclear Safety Advisory Letter MENSAL 93-13) that indicated previousspurious SI overfill calculations vere non-conservative.

Because the neh analysis credits specific operator timing and PSV eater release,it is considered to be a change in the method of performing a plant safetyfunction as desciibed in the SAR.

APPROVED (PLANT MANAGER)

Yes No

PSRC REVIEW: MEETING NO. V0 t)7/ DATE 7 2G <6 RECOMMEND APPROVAL (X) ( )

DATE

69-10431 08/08/94 ~IDAP TS3. ID2

Page 2 oi'

TITLE: LICENSING BASIS IMPACT EVALUATION (LBIE)

SECTION l. af t v tio

For the issue under consideration, provide an explanation justifying each of theYes/No answers. The detail provided shall be commensurate with the nuclear safetysignificance of the activity or CTE.

1. May the probability of occurrence of an accidentpreviously evaluated in the SAR be increased?

Yes No

(X)

Justification:

The change relates to the mitigation of a specific accident anddoes not impact accident initiation during normal operation.The results of the analysis demonstrate successful eventtermination without initiating of an additional accident.Therefore, the probability of an accident previously evaluatedin the SAR is not increased.

2. May the consequences of an accident previously evaluatedin the SAR be increased?

(X)

Justification:Although no physical changes are being implemented, the revisedanalysis represents a change in the predicted accidentresponse. Currently, pressurizer overfill is not an analyzedconsequence discussed in the DCPP FSAR Vpdate. Instead, theFSAR Vpdate analysis is concerned with Departure from NucleateBoiling (DNB) and input variables like decay heat are biased ina manner that is conservative to minimizing DNB margin. The

neo analysis biases input variables in a manner to promoteoverfill. The conclusion of the net analysis is that overfillrt ill occur, but the PSVs are qualified to reseat after SItermination. The consequences of RCS fluid floe through thePSVs to the Pressurizer Relief Tank (and subsequently throughthe PRT rupture disk) is bounded by the evaluation documente'din NSAL 93-13, Supplement 1. This supplement states "theradiological releases (offsite doses) resulting from breakingthe PRT rupture disk are rt/ithin acceptable limits."

Hence the Condition 2 success criteria continues to bemaintained -- pressure in the RCS stays below 110'f design.fuel is not damaged, and the Condition 2 Spurious SI does notinitiate a Condition 3 event. Therefore, there is no increasein consequences of an accident previously evaluated in the SAR.

3. May the probability of occurrence of a malfunction of (X)

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equipment important to safety previously evaluated in theSAR be increased?

Yes No

Justification:The analysis methodology changes do not result in any physicalchanges or operational changes at OCPP. The net analysiscredits PSV rater operation for this scenario ahereas pastscenarios only credited PSV steam operation. The ElectricPoser Research Institute (EPRI) tested different models of PSVsincluding the OCPP model, Crosby 6M6. 0'estinghouse publishedthe results in &CAP 11671. This &CAP documents that Crosby 6M6valves vere adequate for eater relief dohn to 603 F for up to17 minutes as predicted for Seabrook Vnits 1 8 Z. 0'estinghousehas since reviewed the A'CAP and the OCPP specific Spurious SItransient. The OCPP transient predicts rater temperatures aslow as 603.2 F for up to 4.4 minutes. The conclusion is thatthe OCPP PSVs vill operate reliably using the same technicalbasis as for Seabrook and the other units identified in h/CAP

12677.

There are no other changes in the predicted equipment response.Therefore there is no increase in the probability of occurrenceof a malfunction of equipment important to safety previouslyevaluated in the SAR.

4. Nay the consequences of a malfunction of equipmentimportant to safety previously evaluated in the SAR beincreased?

(X)

Justification:As noted, the subject analysis credits PSV I4/ater reliefoperation; bort/ever, the consequences of PSV failure are nodifferent than if the failure occurred as described in section15.2.12 or 25.3.2. No additional equipment is credited.Therefore, the consequences of a malfunction of equipmentimportant to safety previously evaluated in the SAR sill not beincreased.

5. Nay the possibility of an accident of a different typethan previously evaluated in the SAR be created?

(X)

Justification:T'h e change relates to the mitigation of' specific accident anddoes not impact normal operation. The results of the analysisdemonstrate successful event termination r4/ithout initiating ofan additional accident. Therefore. the possibility of an

69-10431 08/08/94I DAP TS3. I02

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accident of a different type than previously evaluated in theSAR is not created.

Yes No

Hay the possibility of a malfunction of equipmentimportant to safety of a different type than anypreviously evaluated in the SAR be created?

(X)

Justification:The change relates to the documentation of the mitigation of a

specific accident and does not impact normal operation, henceit has no impact on potential malfunctions during normaloperation.

The results of the analysis demonstrate acceptable mitigationof the Spurious SI event crediting only the safety-relatedsystems that here previously evaluated in the SAR. The PSVs

sere previously credit'ed for steam release. Therefore, thischange does not create the possibility of a malfunction ofequipment important to safety of a different type than anypreviously evaluated in the SAR.

Is there a reduction in the margin of safety 'as definedin the basi s for any Techni cal Speci fi cati on?

(X)

Justification:The analytical work discussed here is consistent with alltechnical specifications. This analytical work does notcontradict the statement that the PSVs provide the safetyfunction of RCS pressure protection, nor does it affect thesafety function of any other equipment since the licensingbasis scenario still relies on PSV operation.

The analysis does contradict statements made in SER 4. SER 4concludes that overfill sill not occur 14ithin ZO minutes of a

Spurious SI. The SER conclusion is based on the pre-NSAL 93-13methodology and fails to account For RCS expansion due to decayheat. This situation is acceptable because DCPP continues tomeet the criteria established for Spurious SI mitigation andPSV operation, although the mechanism for demonstrating thiscompliance has changed. Specifically, NUREG 0737, item II.0.1,ahich states that "licensees and applicants shall conduct

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testing to qualify the reactor coolant system relief and safetyvalves under expected operating conditions for design-basistransients and accidents, " was previously addressed by anaccident analysis that demonstrated no eater relief. Thechange being implemente'd with this 50. 59 review means that norItem II.0.2 is satisfied by the EPRI testing program, that issummarized in A'CAP 2167T. 0estinghouse letter PGE-96-565verifies that the testing program applies to the PGSEPressurizer Safety Valves.

Yes No

Therefore, there is no reduction in the margin of safety asdefined in the basis for any Technical Specification.

Is there a change to the Fire Protection Program ( FPP)(FSAR Update, Section 9.5, including tables, figures, andappendices)?

(X)

9. Is there a change to the Quality Assurance (QA) Program "( )t

(FSAR Update, Chapter 17)?

Complete and attach the next form sheet to this 10 CFR

50.59 Safety Evaluation.

(X)

Based upon the above criteria and justification. I have determined that anunreviewed safety question is" is not~ involved. A change to the DCPP

Technical Specifications is* is not ~ involved. Further. anyresulting changes to the FPP or QA Program are documented as being within thelicensing is.

7 tePreparer S'ture

s

REVIEWED: Based on my independent technical review,

g.'..j ~. i.Qi,~,Independent Technical Reviewer Signature

Date

I, concur with the above

. l! 'i.Date

*If an unreviewed safety question. change to the DCPP Technical Specifications orother license amendment is involved. NRC approval is required prior toimplementing the activity or CTE.

0

t I

„V

DCPP UHITS I 5 2 FSAR UPDATE

Following the actuation signal, the suction of the coolant charging pumps is diverted from the volumeI A

control tank to the RWST. The charging injection valves between the charging pumps and the injection header

open automatically. The charging pumps then pump RMST water through the header and injection line and into

the cold legs of each loop. The safety injection pumps also start automatically but provide no flow when

the RCS is at normal pressure. The passive injection system and the low-head system also provide no flow at

normal RCS pressure.

m DNSERTAHERE]An SIS signal normally results in a reactor trip followed by a turbine trip. However, it cannot be assumed

that any single fault that actuates the SIS will also produce a reactor trip. Therefore, two different

courses of events are considered.

Case A: Trip occurs at the same time spurious injection starts

Case B: The reactor protection system produces a trip later in the transient.

For Case A, the operator should determine if the spurious signal was transient or steady state in nature,

i.e., an occasional occurrence or a definite fault. The operator will determine this by following approved

procedures. In the transient case, the operator would stop the safety injection and bring the plant to the

hot shutdown condition. If the SIS must be disabled for repair, boratlon should continue and the plant

brought to cold shutdown.

For Case B, the reactor protection system does not produce an irwediate trip and the reactor experiences a

negative reactivity excursion causing a decrease in the reactor power. The power unbalance causes a drop in

T and consequent coolant shrinkage, and pressurizer pressure and level drop. Load will decrease due toavg

the effect of reduced steam pressure on load if the electrohydraulic governor fully opens the turbine

throttle valve. If automatic rod control is used, these effects will be lessened until the rods have moved

out of the core. The transient is eventually terminated by the reactor protection system low-pressure tripor by manual trip.

15.2-38

Hay 1995 Revision 10

I

T

I1

OCPP UNITS I /h 2 FSAR UPDATE

15.2.15 References

1. W. C. Gangloff, An Evaluation of Antici ated 0 erational Transients in Mestin house

Pressurized Mater Reactors, WCAP-7486, Hay )971.

B. N. Ritter, Jr. and R. F, Barry, TiilNKLE-ANulti-Bieenaional Neutron Kinetioe ~Cpm uterCode, WCAP-7979-P-A (Proprietary) and 'WCAP-8028-A (Non-Proprietary), January 1975.

C. Hunin, FACTRAN A Fortran IV Code for Thermal Transients in UO Fuel Rod, WCAP-7908, I

June 1972.

T. W. T. Burnett, et al., LOFTRAN Code Descri tton, 'WCAP-7907, June 1972.

H. Chelemer, et al., Im roved Thermal Desi n Procedure, WCAP-8567 (Proprietary) and

MCAP-8568 (Non-Proprietary), July 1975.

Technical S ecifications, Diablo Canyon Power Plant Units I and 2, Appendix A to LicenseNos. DPR-80 and DPR-82, as amended 'to the date of the most recent FSAR Update Revision.

H. Chelemer, et al., Subchannel Thermal Anal sis of Rod Bundle Cores, MCAP-7015, Rev. I,January 1969.

H. A. Hangan, Over ressure Protection for Mestin house Pressurized Water Reactor, MCAP-

7769, October 1971.

J. S. Shefcheck, A lication of the THINC Pro ram to PWR Desi n, MCAP-7359-L, August1969 (Proprietary), and WCAP-7838, January 1972.i

10. T. Horita, et al., Dro ed Rod Nethodolo for Ne ative Flux Rate Tri Plants, 'MCAP-

10297-P-A (Proprietary) and MCAP-10298-A (Non-Proprietary), June 1983 .

m DNsERT B HER+

15.2-41September 1992 Revision 8

INSERT A

The potential for DNB, loss of fuel integrity, or excessive cooldown isevaluated in the analyses presented in this section. Reference 11 addressesthe potential for pressurizer overfill and concludes that overfill could occur forup to 4.4 minutes prior to termination by the operator. Reference 12demonstrates that pressurizer safety valves will successfully operate duringthis transient without adverse consequences.

INSERT B

11. Westinghouse letter PGE-96-584, Diablo Can on Units 1 and 2S urious Safet In'ection Calculation Note, June 1996.

12. Westinghouse letter PGE-96-565, Diablo Can on Units 1 and 2S urious Safet ln'ection/Pressurizer Safet Valve Water Relief FinalResults, May 31, 1996.

I

WestinghouseElectric Corporation

Energy Systems Nuclear Technoroi.y Oivision

Box 355Pittsburgh Pennsylvania 15230 0355

Mr. M. J. AngusPacific Gas & Electric Company245 Market StreetSan Francisco, CA 94105

PGE-96-565

NSD-NT-OPL-96-250

May 31, 1996

Attention: Mr. Ralph Berger

PACIFIC GAS AND ELECTRIC COMPANYNUCLEAR PLANT, DIABLOCANYON UNITS 1 & 2

ri us afe In'i n Pr ri r afe Valve Wat r Relief Fin Re its

Dear Mr. Angus:

Westinghouse has completed the analyses of the DCPP FSAR Chapter 15 Spurious SI event and the

subsequent evaluation of the pressurizer safety valves (PSVs) and water relief. The conclusions of the

Westinghouse evaluation are attached. Note that a revision to DCPP FSAR Section 15.2.14 will be

provided following upcoming discussions with PG&E personnel. In conclusion, DCPP can be expected

to mitigate the Spurious SI transient without any adverse concern for PSV operability.

Ifyou have any questions, please contact the undersigned.

Very truly yours,

JCBJohn C. Hoebel, Project ManagerPacific Gas & Electric Project

Attachment

cc: D. B. Miklush, 1LM. J. Angus, 1LW. G. Crockett, 1LJ. E. Tomkins,'LJ. A. Reichanadter QV Sales), 1L, 1AT. Lee, 1L

v .R. Berger, 1L, 1A

RECEIVED

JUN 13. 1996

CMMsvppoRT seRvIccslIMs

AiTACHMENTTO PGF 96-565

, 230372

TRANSIENT ANALYSIS

Westinghouse has performed an evaluation of the DCPP pressurizer safety valves (PSVs) for water reliefduring an inadvertent ECCS actuation transient. The first portion of the overall evaluation consisted of a

new analysis of the FSAR Chapter 15.2.14 Spurious Operation of the Safety Injection System event. The

transient analysis for this event is characterized as follows:

High-head Safety Injection (Si) Actuation / Reactor Trip at time zero.

SI injection and reactor coolant system (RCS) heatup due to decay heat cause pressurizer level to

increase. Eventually pressure increases towards the PSV setpoint as spray becomes ineffective incontrolling pressure; at about 553 seconds one PSV opens on steam (pressurization rate about

1.7 psi/sec) at a pressure of 2475 psia (nominal setpoint -1% tolerance). Steam flow capacity of a

single PSV is assumed to be a constant 420,000 lb/hr (116.67 lb/sec).

PSV reseats at 2375 psia (nominal setpoint -5% blowdown), pressurization to opening setpointoccurs again. The LOFTRAN model predicts six PSV opening cycles on steam prior to pressurizer

fili'ressurizer fills with water at 725.5 seconds, and PSV partially opens to relieve water. Water reliefcalculated using a homogeneous equilibrium saturated model.

The transient analysis assumes SI termination via operator action at 960 seconds (16 minutes).Water relief ceases shortly following SI termination.

Water relief through the PSVs begins at an initial water temperature of 616.8'F and terminates 4.4minutes later at a final water temperature of 603.2'F

VALVEANALYSIS

The transient information for postulated Spurious SI event for DCPP has been evaluated. An analysis has

been performed to document that the Pressurizer Safety Valves at DCPP willbe operable for the initialsteam pops followed by slightly subcooled water relief that lasted from 725 seconds to 990 seconds of the

transient.

The analysis used the methodology in WCAP-11677, "Pressurizer Safety Valve Relief Operation ForWater Discharge During A Feedwater Line Break", as the basis for the justification. The WCAP, which

was developed from data extracted from the EPRI-CE PWR Safety Valve Test Report (EPRINP-2770-LD Vol. 6), showed that'the Crosby Safety Valves could be cycled in slightly subcooled water

following steam discharges without deleterious effects. The DCPP ECCS evaluation was centered on

determining whether the amount of subcooling permitted stable valve lift, and ifso, how many lifts can

be expected during the water relief phase of the transient. Based on the evaluation, a stable performance

is expected during the postulated water relief phase of the transient with a water temperature of 603'F.The number of actual valve opening cycles during water relief was determined to be two.

In conclusion, DCPP can be expected to mitigate the transient without any adverse concern foroperability.

westinghouseEnergySystemsBusinessUnit

NUCLEAR SAFETY ADVISORY LETTER

THIS IS A NOTIFICATIONOF A RECENTLY IDENTIFIED POTENTIAL SAFETY ISSUE PERTAINING TO BASICCr)MPANENTS SUPPLIED BY WESTINGHOUSE. THIS INFORMATION IS BEING PROVIDED TO YOU SO THATAREVIEW OF THIS ISSUE CAN BE CONDUCTED BY YOU TO DETERMINE IF ANY ACTION IS REQUIRED.

P.O. Box 355. Pittsburgh. PA 1523O-0355

Subject: Inadv'ertent ECCS Actuation at Power

Basic Component: Transient Accident Analysis

Plants: See Page 2, Table 1

Substantial Safety Hazard or Failure to Comply Pursuant to 10 CFR 21.21(a)Transfer of Information Pursuant to 10 CFR 21.21(b)Advisory Information Pursuant to 10 CFR 21.21(c)(2)

Reference:

Number: NSAL-93-013

Date: June 30, 1993

Yes 0 No ClYes EYes 0

SM Ih IARY

Westinghouse has discovered that potentially non-conservauve assumptions were used in the licensinganalysis of the Inadvertent Operation of the ECCS at Power accident. Based on preliminary sensitivityanalyses, use of revised assumptions could cause a water solid condition in less than the 10 minutesassumed for operator action time. If the PORVs were blocked, the PSRVs would relieve water andpotentially cause the accident to degrade from a Condition II incident to Condition III incident withoutother incidents occurring independently. Per ANS-051.1/N18.2-1973, a Condition II event cannotgenerate a more serious event of the Condition III or IV type without other incidents occurringindependently.

Westinghouse is unable to determine whether a defect causing a substantial safety hazard or a failure tocomply resulting in a substantial safety hazard exists because sufficient plant specific information is notavailable. Under 10 CFR 21.21(b), ifWesunghouse determines that there is insufficient informationavailable to provide the capability to perform an evaluation, then Westinghouse must inform affectedlicensees of this determination.

Additional information. ifrequired. may be obtained from the originator. Telephone 4 I2-374A302.

OriginatorG. G. ent H. A. Sepp, er

Strategic Licensing Issues

Vavrek

TABLE I PLANT APPLICABILITYLIST

Byron I & 2Braidwood I & 2Zion I & 2V. C. SummerD. C. Cook I & 2Shearon HarrisW. B. McGuire I & 2Catawba I & 2Beaver Valley I & 2J. M. Farley I & 2

Vogtle I & 2SeahrookMillstone 3North Anna I & 2Surry I & 2Salem I & 2Diablo Canyon I & 2Wolf CreekCallawaySequoyah I & 2Watts Bar I & 2Haddam Neck (note I)

Almaraz I & 2Doel 1,2 & 4VandellosAsco I &2KrskoBeznau I & 2Ringhals 2, 3 & 4Tihange I & 3ZoritaC. N. des ArdennesC. N. BR3Kori 3 &4Yonggwong I & 2Maansham I &,2Mihama 2Ohi I & 2 (note I)Takahama I (note I)

Notes: I. Westinghouse is not cognizant of the current ECCS design for these plants.

NShL.034l3 Sheet 2 of 5

TECHNICAL DESCRIPTION

ISSUE DESCRIPTIONThe Inadvertent Actuation of the Emergency Core Cooling System (ECCS) accident {also referred to as

the Spurious SI event) is a Condition II incident as defined by ANS-051.1/N18.2-1973, "Nuclear SafetyCriteria for the Design of Stationary Pressurized Water Reactor Plants." A Condition II incident isdefined as a fault of moderate frequency, which, at worst, should result in a reactor shutdown with theplant being capable of returning to operation. A Condition II event cannot generate a more serious eventof the Condiuon III or IV type without other incidents occumng independently.

Standard Review Plan NUREG-0800, Rev. 1, Section 155.1, "Inadvertent Operauon of ECCS thatIncreases Reactor Coolant Inventory," states that to meet the requirements of GDC 10, 15, and 26 forincidents of moderate frequency an incident of moderate frequency should not generate a more seriousplant condiuon without other faults occurring independently. To address this, Westinghouse adopted thefollowing criterion:

The pressurizer shall not become water solid as a result of this Condition II transient within theminimum time required for the operator to identify the event and terminate the source of fiuidincreasing the RCS inventory. Typically, a 10 ptinute operator action time has been assumed.

The basis for demonstrating that the pressurizer will not become water solid is to preclude the possibilityof discharging primary coolant through the Power Operated Relief Valves (PORYs) and/or the PressurizerSafety Relief Valves (PSRVs), causing the incident to progress from one of moderate frequency to aninfrequent small break LOCA incident. A small break LOCA condition could result from failure of thePSRVs to close after discharging water since the PSRVs were typically not designed for water relief.

'Based on a review of the analysis methods used to evaluate this accident, it was discovered that thesemethods were developed with the primary emphasis on criteria for maintaining RCS pressure below-thedesign value and ensuring that fuel cladding integrity is maintained. These methods did not emphasizethe criterion for preventing the pressurizer from becoming water solid within the allowable operator actiontime. Sensitivity analyses performed for this accident have shown that some analysis assumptions arenon-conservauve with respect to maximizing the potential for pressurizer filling. Revised analysisassumptions that conservatively consider the potential for pressurizer filling for the Inadvertent Operationof the ECCS at Power accident have been found to have a significant effect on the rate at which thepressurizer water volume increases.

TECHNICALEVALUATION%'esunghouse has performed preliminary sensitivity analyses that indicates for some plant specificapplications using revised assumptions the pressurizer can become water solid in less than 10 minutes.To conclude that Standard Review Plan NUREG-0800 is met, it must be demonstrated that the pressurizerdoes not become water-solid in the minimum allowable operator action time, that the PSRVs do not open,or that the PSRVs are capable of successfully closing following water relief. IfECCS fiow is notterminated before water is discharged through the PSRVs, it cannot be demonstrated without plantspecific PSRV operability assessments that this accident does not lead to a more serious plant condition.Water relief through the PORVs is not.a concern, because the PORV block valves can be used to iso ate

the PORVs if they fail to close. If=ECCS flow is not terminated before the pressurizer becomes watersolid and water is discharged through the PSRVs, it can not be demonstrated that this accident does notlead to a more serious Condition 111 LOCA event.

'BASAL 934 I3 Sheet 3 of 5

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TECHNICALEVALUATION(con't)The analysis for licensing basis assumed maximum ECCS fiow which typically includes an additional 5to 10 percent margin on discharge pressure above the vendor's specified pump performance. Thelicensing basis analysis assumed that the PORVs, the pressurizer water level control system, the steamdump system and the steam generator PORVs were not available to help mitigate this accident since theyare considered to be control grade functions. Also, no credit was taken for letdown since it is isolatedfollowing a safety injection signal for those plants which use charging pumps for high head saferyinjection pumps.

ASSESSMENT OF SAFETY SIGNIFICANCEAnalyses of the Inadvertent ECCS Actuation at Power accident using revised analysis assumptions withthe primary emphasis on conservatively demonstrating acceptability with respect to pressurizer fillinghave been performed. These analyses show a potential for reaching a water solid condition before theallowable operator acuon time. Without the appropriate operator action to terminate the ECCS flowprior to reaching a water-solid pressurizer condition, the accident may progress from a Condition II to amore severe Condition III LOCA event as a result of failure of the PSRVs due to water relief through thevalves.

Although Westinghouse previously adopted the conservative criterion of preventing the pressurizer frombecoming water solid, the acceptability of water leakage from the RCS for Inadvertent Operation ofECCS Condition II events is supported by NUREG-0800 and ANS-051.1. To meet the applicableCondition II criteria, the magnitude of any water relief inust not exceed that of the normal makeupsystems (which it will not by definition since this is the cause of the water relief) and the ability toorderly shutdov n the reactor must be maintained. The latter implies that the RCS must ultimately beisolated. Hence. the PSRVs must either not open or must be capable of closing after release of subcooledwater.

NRC AWARENESS/REPORTING CONSIDERATIONSWestinghouse is unable to 'determine if this issue would cause a substantial safety hazard or a failure tocomply resulting in a substantial safety hazard because sufficient plant specific information is notavailable. This information is being transferred to the applicable plants pursuant to 10 CFR 21.21(b).The NRC has not been notified of this issue.

RECOMMENDED ACTIONS

Licensees should first determine if their current licensing basis requires them to analyze theInadvertent Operation of the ECCS at Power accident. Ifthis accident is not included within theircurrent licensing basis, no additional action is required.

Licensees should determine if their Pressurizer Safety Relief Valves are capable of closingfollowing discharge of subcooled water. If the PSRVs were designed or qualified to relievesubcooled water, the Inadvertent ECCS Actuation at Power accident will not degrade into a moreserious Condition IIIevent, since these valves will close once ECCS flow has been terminated. Itshould be noted that the licensees may have qualified these valves in compliance to NUREG-0737, Item II.D.1.

3. If the PSRVs are not designed or qualified for subcooled water relief, the licensees should re-

evaluate the Inadvertent ECCS Actuation at Power accident using one or a combination of thefollowing options.

!4SAi..934 li Sheet 4 of 5

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RECOMMENDED ACTIONS (con't)

O tion I; Reduce the maximum ECCS flow used in the safety analysis. Preliminary sensitivityanalyses have shown that using less conservative flow mav sufficiently delay ftlflng thepressurizer such that the operator action to terminate the accident can be successfully credited.

~Otton ll: 'Use a less restricdve operator response time. Per ANSVANS-Sg.g-1992, "Timeresponse design criteria for safety-related operator actions," credit can be taken in the analysis forthe operator to stop one pump at 7 minutes, a second pump at 8 minutes, and depending on theplant specific design, the third at 9 minutes. Preliminary sensitivity analyses have shown thatusing these less restrictive operator action times mav sufficiently delay or prevent filling thepressurizer.

v

~Otton fft: Credit the use of one or more PORVs to help mitigate the accident. Preliminarysensitivity analyses have shown that ifa water solid pressurizer condition is reached, one PORYshould be sufficient to maintain pressure below the PSRV setpoints and prevent discharge ofwater through the pressurizer safety relief valves. To credit this option, the licensee would haveto ensure that at least one PORY is always available (PORV block'alve is opened). This optioncould also be credited if the PORVs are blocked by ensuring that the Emergency OperatingProcedures (EOPS) instruct the operators to open at least one PORV block valve before thePORV setpoint is reached. Use of this option may require a change to the plant EOPS and/or theplant technical specifications to ensure that at least one PORV is available since most technicalspecifications currently allow the PORVs to be isolated during power operation.

.'BASAL 934 I3 Sheet 5 of 5

anal NUCLPAR SAFEIY AD Y I3HTERSystemsBusinessUnit

THIS IS h NOIIFICATIONOF h RECENT-Y IDENTIFIEDPOTEYHhL SAFETY ISSUE PERTAININGTO BASIC

COMPONERIS SUPPLlED BY WESTINGHOUSE. THIS INFORMATIONIS BEING PROVIDED TO YOU SO THAThREVIEW OF THIS ISSUE ChN BE CONDUCTED BY YOU TO DETERMINE IF hNY hCIMN IS REQUIRED.

P.O. Box 355, Piaaburgh, Ph I52304355

Subject: Inadvertent ECCS Actuation at Power

Bisic Component: Transient Accident Analysis.

Plants: See Page 2, Table 1

Number: NSAL-93413,Supplement 1

Date: Oct. 28, 1994

Substantial Safety Hazard or Failure to Comply Pursuant to 10 CFR 21.21(a) Yes D No ilTransfer of Information Pursuant to 10 CFR 21.21(b) Yes DAdvisory Information Pursuant to 10 CFR 21.21(c)(2) Yes D

Reference:

As previously described in Nuclear Safety Advisory Letter {NSAL) 93413, dated June 30, 1993, Westinghousediscovered that potentially non~nservative assumptions were used in the licensing analysis of the Inadvertent Operationof. the ECCS at Power accident.

Ia addition to thc information provided in the origiaal NSAL, this supplement provides additional information relatedto this issue and specifically notes that Positive Displacement Pump (PDP) operation during normal operating conditionswilltend to aggravate the event by reducing the time to reach a pressurizer water solid condition. Ifwater relief fromthc pressurizer does occur, the piping downstream of the PSRVs/PORVs must be qualified for subcooled water relief.Normal operation of a PDP, concurrent with initiation of an Inadvertent ECCS Actuatioa event, willserve to increasethc injection flow by approximately 100 gpm and, without operator action, shorten thc time to reach a pressurizer watersolid condition. Though not all plants listed in Table 1 necessarily operate a PDP during normal plant operations, thissupplement to NSAL 93<13 willbe transmitted to those plants identified in the original NSAL to ensure continunity.

'Hds supplemental information dces not pose a substaatial safety hazard or failure to comply per the definitionsprovidcdin 10 CFR Part 21.21(a).

hddidooai information, ifrequired, may be obtained from the originator. Telephone 412-374-5036.

Originator(s):J. S. Galembush H. A. Sep, ger

Strategic Licensing Issues

NSAIAOOlS.~ 1 Sheet 1 of 7

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Byron 1 &2Braidwood 1 &2

'ion 1 &2V. C. SummerD. C. Cook 1 & 2Shearon HarrisW. B.,McGuire 1 & 2Catawba 1 &2Beaver Valley 1 &2J. M. Parley 1 &2Vogtle1 & 2SeabrookMillstone 3

North Anna 1 & 2Surry 1 &2Salem 1 & 2Diablo Canyon 1 & 2Wolf CreekCallawaySequoyah 1 & 2Watts Bar 1 & 2Haddam Neck (note 1)

Almaraz 1 & 2Doel 1,2& 4VandellosAsco 1 &, 2KrskoBeznau 1 &2Ringhals 2, 3 & 4Tihange 1 & 3ZoritaC. N. des ArdennesC. N. BR3Kori 3 & 4Yonggwong 1 & 2Maansham 1 & 2Mihama 2Ohi 1 & 2 (note 1)Takahama 1 (note 1)

Notes: l. Westinghouse is not cognizant of the current ECCS design for these plants.

NSALr%41), Oeppicmcaa 1 Sheet 2 of 7

TECHNICAL DESCRIFI'ION

In the original issue of NSAL 93413, the Inadvertent Actuation of the Emergency Core Cooling System

(ECCS) accident is noted as a Condition II incident as defined by ANS451.1/N18.2-1973, "Nuclear Safety

Criteria for the Design of Stationary Pressurized Water Reactor Plants." A Condition 11 incident is defined

as a fault of modeiate frequency, which, at worst, should result in a reactor shutdown with the plant being

capable of returning to operation.' Condition II event cannot generate a more serious event of the

Condition IIIor IV type without other incidents occurring independently.

As described in NSAL 93413, the historical analysis methodology for the "Inadvertent Operation of the

ECCS at Power" event used assumptions to conservatively demonstrate that the DNBR safety analysis and

RCS pressure limits are met and that these assumptions may not be conservative with respect to maximizing

the RCS inventory increase.

Standard Review Plan NUREG4800, Rev. 1, Section 15.5.1, "Inadvertent Operation ofECCS that Increases

Reactor Coolant Inventory," states that "specific criteria to meet the requirements of GDC 10, 15, and 26

for incidents of moderate frequency are:

a. Pressure in the reactor coolant and main steam systems should be maintained below 110% of the

design values,

b. Fuel cladding integrity shall be maintained by ensuring that the minimum DNBR remains above the

95/95 DNBR limit for PWRs, and,

c. An incident of moderate frequency should not generate a more serious plant condition without other

faults occurring independently."

To address criterion (c), Westinghouse has historically applied the following more restrictive criterion forease in interpreting the transient results.

The pressurizer shall not become water solid as a result of this Condition II transient withinthe minimum time required for the operator to identify the event and terminate the source

of fluid increasing the RCS inve'ntory. Typically a 10 minute operator action time has been

assumed.

It is easy to conclude that criterion (c) is met if it can be demonstrated that the pressurizer does not become

water-solid in the minimum allowable operator action time. However, ifECCS flow is not terminated before

the pressurizer becomes water solid, it is more difficultto demonstrate that this Condition II event does not

lead to a more serious plant condition. Note that no credit for automatic actuation of RCS coolant letdown

(pressurizer level control), pressurizer pressure control (PORVs), steam generator PORVs, or steam dumpis taken since these are considered control grade systems. Without these systems available, it is anticipated

that an Inadvertent ECCS Actuation at Power event could potentially lead to a water-solid pressurizercondition and result in a Condition GI LOCA event ifcorrective action is not taken in a timely manner. Anincrease in the assumed injected flow due to the potential for concurrent operation of a positive displacement

pump at the time of event initiation would further reduce the time to reach a water solid pressurizer

condition, and hence, reduce the time available for appropriate operator actions.

NSALAOD,~ i Sheet 3 of 7

V

The historic."Westinghouse" internal acceptance criterion of preventing the pressurizer from reaching a

water-solid condition during Condition IIevents clearly eliminates any concerns of escalating a Condition 11

event to a Condition IG or IV event. However, this criterion is overly conservative and due to changes inanalysis modeling assumptions made to conservatively analyze this event for proper consideration of .

pressurizer water volume, this criterion is now being challenged within the minimum allowable operatoraction time of 10 minutes typically assumed.

*. However, merely reaching a water-solid pressurizer condition does not imply that the event willescalate into

that of a Condition IQ or IV event. ANS 51.1/N18.2-1973, lists Example (15) of a Condition II event as

a "minor reactor coolant system leak which would not prevent orderly reactor shutdown and cooldownassuming makeup is provided by normal makeup systems only." Here, "normal makeup systems" is definedas those systems normally used to maintain reactor coolant inventory under respective conditions of startup,hot standby, power operation, or cooldown, using on-site power.

Since the cause of the water relief is the ECCS fiow, the magnitude of the leak will be less than orequivalent to that of the ECCS (i.e., operation of the ECCS maintains RCS inventory during the postulatedevent and establishes the magnitude of the subject leak)., Therefore, the above example of a Condition 11

event is met provided "orderly reactor shutdown"'is also met.

To ensure "orderly reactor shutdown" can occur, the RCS pressure boundary must ultimately be isolatable.once the source of the ECCS fiow is terminated. To ensure the RCS pressure boundary can be isolated, thePressurizer Safety Relief Valves (PSRVs) must function as designed and the powermperated relief and/or

~ block valves must be available to the operator (after the minimum allowable operator action time) to provideisolation functions.

For continued conservatism in the safety analysis methodology, it is assumed that PSRVs must not pass waterin order to ensure their integrity and continued availability. Therefore, the Westinghouse internal eventcriterion for this Condition II event is revised such that subcooled water discharge through the PressurizerSafety Relief Valves shall be precluded for a Condition II transient.

Hence, a water-solid pressurizer condition should be precluded when the pressurizer is at or above the setpressure of the PSRUs. An exception to this criterion can be made ifthe utilitycan support a position thattheir PSRVs are. designed and qualified to relieve subcooled water.

The plant technical specifications generally require the PORVs and block valves to be operable. Theiroperability is determined, in part, on the basis of their capability to manually control reactor coolantpressure. With one or more PORVs available, the PSRV setpoint will not be reached during this event.Any water discharge from the RCS would be through the PORV(s). Furthermore, isolation of the RCSfollowing operator action to terminate ECCS flow is obtainable by available block valve closure.

For the potential condition of the plant operating with all the PORVs blocked, RCS pressure would exceedthe PORV settings and continue to increase towards the PSRV setpoint. To preclude water relief throughthe PSRVs, either action to terminate the ECCS flow to avert a water-solid condition or to confirm that atleast one PORV is unblocked and available for water relief, prior to reaching water-solid condition, mustbe taken within the minimum operator action time.

NSAIA3OD. Oeppkcocct i Sheet 4 of 7

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"~"',: -'„The acceptabBity of water leakage from the RCS for the Inadvertent Operation of ECCS at Power'tzn Ilevent is further supported by statements contamed in NUREMN00, Section 15 5.1 - 15.5 2.

Specifically, Section IH Review Procedures indicate (first paragraph on page 15.5.14):Jg )i....

,"The results of the applicant's analysis are reviewed and compared to the acceptance criteria presented";"'--~.-: ' '"h subsection H regarding maximum pressure in the reactor coolant and main steam systems and the

minimum critical heat flux ratio (MCHFR) or departure f'mm nucleate boBing ratio (DNBR). Thevariations with time during the transient of the neutron power, ......;.. pressure relief valve flow rate,

." . "; «nd flow rate from the reactor coolantsystem to the containment system (ifapplicable) are reviewed."

..;,,= 'Iherefore, based on the aforementioned information, it is interpreted that Condition H criteria can be met

..: - with some water relief from the RCS. To meet the applicable Condition H criteria, the magnitude of any'-.; -,"„" ="';, - water relief must not exceed that of the normal makeup systems (which it wBl not by definition since the

ECCS flow is the cause of the ~ater relief) and as long as orderly shutdown of the reactor can occur. 'Ihe~ - 'atter implies that the RCS must ultimately be isolatable, Hence, PSRVs must not be exposed to discharge

of liquid as a result of the pressurizer reaching a water solid condition,

Option H of the original NSAI. 93413, which references ANSVANS-58.8-1984, "Time Response DesignCriteria for Nuclear Safety-Related Operator Actions," recommends the use of a less restrictive operatorresponse time. Per ANSVANS-58.8-1984, the operator action times for event indication are based on specifictime tests. Time test 1 requires 5 minutes and time test 2 requires (1 + N*1) minutes where "N" signifiesthe number of discrete manipulations required. PORVs should be expected to be available unless they wereblocked due to a leakmg PORV condition. Operator action associated with assuring PORV availabilitygenerally consists of manually opening a block valve to allow it to actuate on demand. An acceptableminimum time to assume initial operator action would therefore be 7 minutes.

Two additional concerns must also be addressed in conjuction with potential water relief through either thePORVs or PSRVs (ifqualified for such). The definition of a Condition H incident states that the event at~orst "should result in a reactor shutdown with the plant being capable of returning to operation. In orderto meet this condition, the piping downsteam of the PSRVs and/or PORUs must be qualified for water relief.Secondly, water relief may result in overpressurizing the Pressurizer Relief Tank (PRQ, breaking the rupturedisk, and spilling contaminated fluid into containment. Therefore, the radiological consequences of thisoccurrence must also be evaluated.

To conclude that Standard Review Plan NUREG4800 is met, it must be demonstrated that 1) the pressurizerdoes not become water-solid within the minimum allowable operator action time, 2) the PSRVs do notrelieve water or that the PSRVs are capable of successfully closing following subcooled water relief, 3) thedownstream piping is capable ofhandling the water discharge flow, and 4) that the radiological consequencesof breaking the PRT rupture disk do not violate the applicable offsite dose limits. Water relief through thePORVs is not a concern because the PORV block valves would be available to isolate the PORVs shouldthey fail to close.

l4$ALhOOD, Sepal~ i Sheet 5 of7

~"i ~ i

'Ihe assessments provided in the original NSAL remain valid. Analyses of the Inadvertent ECCS Actuation

. at Power accident using revised analysis assumptions with the primary emphasis on conservatively.demonstrating acceptability with respect to pressurizer fillinghave been pmformed. These analyses show

a potential for reaching a watermlid condition before the ten (10) minute allowable operator action timetypically assumed. Without the appropriate operator action to teaninate the ECCS flow prior to reaching

a water~lid pressurizer condition, the accident may progress from a Condition Il to a more severe

Condition IIILOCA event as a result of the potential failure of the PSRVs to close after water 'relief.\

Although Westinghouse previously adopted the conservative criterion of preventing the pressurizer frombecoming water solid, the acceptability of water leakage from the RCS for Inadvertent Operation of ECCSat Power Condition II events is supported by NUREG4800 and ANS-51.1/N18.2-1973. To meet the.applicable Condition IIcriteria, the magnitude ofany water reliefmust not exceed that of the normal makeupsystems (e.g., operation of the ECCS) and the ability to orderly shutdown the reactor must be maintained.The latter implies that the RCS must ultimately be isolated. Hence, the PSRVs must either not relieve wateror must be capable of closing after release of subcooled water.

Without appropriate operator action to terminate safety injection flow prior to reaching a water-solidpressurizer condition, the Inadvertent ECCS Actuation at Power event may progress from a Condition II toa more severe Condition III LOCA event as described above. While this occurrence may result in aviolation of one of the applicable licensing basis criteria for a Condition II event it is not considered asignificant safety concern. As a LOCA event, discharge of coolant out of the PSRVs and PORVs due toECCS flow is not significantly adverse relative to other Condition IIILOCA events currently analyzed. Thisis because the pressurizer is located on the hot leg (a hot leg LOCA being less severe than a cold leg LOCA)and because the Inadvertent ECCS Actuation at Power event typically models maximum ECCS flow (tomaximize the effects of the initiating event) which is a benefit for LOCA. As such, the Inadvertent ECCSActuation at Power induced LOCA is bounded by the existing small break LOCA analyses.

Relative to the qualification of the PSRV/PORV downstream piping, it has been demonstrated that thethermal hydraulic loads downstream of these valves, generated for water solid discharge, are bounded bythe steam-slug discharge case which was used for the design of the pressure safety and relief system.Therefore, the downstream piping is qualified under the existing design criteria for the water solid dischargeevent. An evaluation of the radiological consequences has been performed which bounds the Table 1 plantsfor which the required analysis information is available (i.e., U.S. plants). The radiological releases (offsitedoses) resulting from breaking the PRT rupture disk are within acceptable limits.

AW NE IDE

Westinghouse has determined that this supplemental information does not represent a substantial safety hazardor a failure to comply resulting in a substantial safety hazard. The NRC has not been notified of this issue.

NDED

The recommendations provided in the original issue of NSAL 93413 remain valid. The purpose of thissupplement is to provide additional information related to this issue and specifically note that PDP operationduring normal operating conditions will tend to accelerate the event by reducing the time to reaching apressurizer water solid condition. Ifwater relief from the pressurizer is predicted, the PSRUs and the piping

NSAlr9$0l$, Ssppl~cct i

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Sheet 6 of 7

down!eeam of the PSRVs and PORVs must be qualified for subcooled water relief. Normal operation ofa PDP, concurrent with initiation of an inadvertent ECCS Actuation event, will serve to increase theinjection Bow by approximately 100 gpm and, without operator action, shorten the time to reaching apressurizer water solid condition.

MAIA501S, ~keaea 1 Sheet 7 of 7