ML17257A503

Forwards Draft Evaluation of SEP Topic XV-19,LOCA.Requests Info Re Whether as-built Facility Differs from Licensing Basis Assumed in Assessment within 30 Days of Receipt of Ltr

ML17257A503
Person / Time
Site:	Ginna
Issue date:	07/27/1981
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Maier J ROCHESTER GAS & ELECTRIC CORP.
References
TASK-15-19, TASK-RR LSO5-81-07-086, LSO5-81-7-86, NUDOCS 8107300043
Download: ML17257A503 (14)

Text

pl July 27, 1981 Docket No. 50<<244 ISO5-81-07-086 Mr. John E. trier Vice President Electric and Steam Production Rochester Gas 5 Electric Corporation 89 East Avenue Rochester, New York 14649

SUBJECT:

R.E.

GINNA >> SEP TOPIC XV-19 (SYSTEMS),

LOSS OF COOLANT ACCIDENT Enclosed is our draft evaluation of SEP Topic XV-19 (systems),

Evaluation of the radiological consequences will be issued separately.

This evaluation compares your facility with the criteria currently used by the regulatory staff for licensing new facilit)es.

Please inform us if your as-built facil-ity differs from the licensing basis assumed in our assessment within 30 days of receipt of this letter.

This evaluathpn will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect as-built condii tions at your facility.

This topic assessment may be revised in the future if your facility design is changed or if NRC criteria relating to these topics are modified before the integrated assessment is completed.

Sincerely,

Enclosure:

As stated cc w/enclosure:

See next page Bjo73OOO4s 810727 PDR ADOCK 05000244 PDR Dennis H. Crutchfield, Chief Operating Reactors Branch No.

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GINHA TOPIC, XV-19:

LOSS OF COOLANT ACCIDENTS RESULTIHG FROM SPECTRUM OF POSTULATED PIPING BREAKS WITHIN THE REACTOR COOLANT PRESSURE BOUNDARY I.

INTRODUCTION The capability of the R.E.

Ginna Emergency Core Cooling System to miti- !

'ate the consequences of a spectrum of Loss of Coolant Accidents (LOCAs) is evaluated to assure that pipe breaks in the reactor coolant system (RCS) do not result in a loss of core cooling capability.

Detailed ac-ceptance criteria for Emergency Core Cooling System (ECCS) performance are contained in 10 CFR 50.46 and in Standard Review Plan Sections 15.6.5, 6.3 and supporting appendices.

The five main criteria for ac-ceptance are:

1.

Peak clad temperature less than 2200'F 2.

Maximum cladding oxidation less than 175 3.

Total hydrogen generation less than 1/ of'otal zirconium in the active fuel region 4.. Maintenance of eoolable geometry 5.

Long term coolability A spectum of break sizes up to and including a double-ended break of the largest pipe at various break locations is examined using an approved evaluation model which conforms to the requirements of Appendix K to 10 CFR 50 to verify that the acceptance criteria are met for a variety of postulated loss of coolant accidents.

II.

E VALUATION The Gi nna power plant ECCS provides emergency core cooling water at three delivery pressures.

The high pressure safety injection (HPI) system delivers borated water at up to 1500 psi.

This is different than'urrent PWR's which use safety grade charging pumps which are capable of inject-ion at operating pressure, about 2235 psi.

Intermediate pressure passive injection is provided by the accumulators which are held at 700 psi by nitrogen gas overpressure.

The HPI system and the accumulators discharge into lines to each cold leg.

Low pressure cooling water from the refuel-ing water storage tank'is deli'vered by the residual heat removal (RHR) system which becomes available at 300 psi.

The low pressure injection flow is pumped directly into the reactor vessel through separate nozzles.

More compl,ete descriptions of these systems are provided in the Ginna safe shutdown report, (Ref. 1),

and in the Final Safety Analysis Report (Ref. 7).

Switchover from injection to recirculation mode is.covered'n SEP'Topic YI-7.B.

Analysis for the large pipe breaks was performed for the Cycle 8 fuel reload in Reference 2, with the staff evaluation presented in Reference 3.

The small break analysis was performed generically for Westinghouse two-loop PWRs,during the initial Appendix K reviews.

Since the small breaks were clearly demonstrated to be non-limiting, later reloads re-evaluated only the large break spectrum.

In response to the NRC's Bulletins and Orders Task Force, additional small break analyses were performed on a generic basis.

The Westinghouse calculations of Reference 4 were reviewed by the staff in Reference 5.

Lar e Break Anal sis The cycle 8 fuel reload safety analysis (Reference

2) examined six dif-ferent pipe breaks in the cold leg.

Hot leg br eaks were not examined.

Three double area guillotine breaks with discharge coefficients (Cn) of 1.0, 0.6,'.and 0.4 were analyzed.

The other three breaks considered were split breaks with discharge coefficients of 1.0, 0.6, and 0.4.

The sel'ection of breaks for this analyses was justified, based on previous

. evaluations, which clearly identified the cold leg split and guillotine breaks as the mos't limi'tsng.

The a'ssumptions'nd computer codes used in the LOCA analyses are cover-ed in Refer ence 2.

Some of the more important assumptions include:

l.

Initial power at 1021 2.

'Reactor trip is neglected for large breaks 3.

All accumulator water bypasses core until termination of bypass 4.

Linear Heat Generation Rate of 13.76 kw/ft 5.

Total peaking factor is 2.32 6.

Fuel at beginning of life (BOL) conditions These and the other assumptions used for these analyses were in accord-ance with 10 CFR 50.46 and Appendix K, 10 CFR Part 50 and have-.been shown (Ref. 2) to result in conservatively high peak cladding temperatures.

Results - Lar e Break The limiting case of the six breaks examined in the ECCS analysis for Ginna was the double-ended cold leg guillotine break with Cn=0.4.

The peak clad temperature predicted was 1922'F, considerably beTow the 10 CFR 50;46 l,imit of 2200'F.

Clad oxidation (peak and total) was also well

~ within limits. It should be noted that guillotine breaks with a dis-charge coefficient smaller than 0.4 are not required in accordance with Ref erence 6.

h

\\

Small Break Anal sis As discussed above, plant-specific small break analyses were not performed because it was felt that the small breaks would not be the limiting case.

since Westinghouse two-loop plant designs have historically been shown to be large'break limited.

Small Break Anal ses - Post TMI Generic analyses of small break LOCAs were submitted by Westinghouse in response to NRC Bulletins and Orders Task Force requirements.

The staff has accepted these analyses as a basis for providing information on plant response and as an aide to developing guidelines for operator action.

The staff considers these generic analyses to be representative of the response for Ginna to a postulated small break LOCA, however, the Ginna power plant would respond somewhat differently to small breaks than more recent PWR's because the high pressure injection pump has a 1500 psia shutoff head.

Most PWR's (and all the latest) utilize a safety grade char ging system for HPI which has a shutoff head above the operating pres-

. sure of 2250 psia.

The effect of the lower Ginna HPI shutoff head is that small breaks will result in depressurization to 1500 psi, and full repressurization of the RCS is not possible using the safety grade HPI.

Since 1500 psi corresponds to a saturation temperature of 596'F, and the hot-leg temperature is over 600'F, a depressurization to 1500 psi could result in some voiding unless the reactor is cooled.

This added steam formation may be significant depending on when the reactor is trip-ped and the size of the break.

It is unlikely that core cooling would be affected during a small LOCA of this size, but the time to reach sub-cooled conditions may be extended in comparison with recent PWR's.

Results - Small Break - Post TMI As a result of the review of these analyses, the staff expressed concern about the applicability of current evaluation models and their applica-tion to the expanded scope of small break LOCA analyses now being consid-ered.

As part of the TMI Task Action Plan, which is beyond the scope of the SEP r eview, Westinghouse is to revise and resubmit the small break analysis methods for staff approval.

Plant specific calculations, using these revised methods will then be required to show compliance with 10 CFR 50.46.

These analyses should place special emphasis on accidents which actuate the HPI.

I I II.

CONCLUS ION The loss of coolant accidents analyzed for the Ginna nuclear power plant meet the acceptance criteria.

New small break LOCA analyses using revised evaluation models will be conducted as part of the TMI Task Action Plan and will not be included as part of the SEP review.

REFERENCES l.

Safe Shutdown Systems for R.

E. Ginna Nuclear Power Plant, SEP Topic VII-3, May 13, 1981.

2.

"ECCS Analysis for R.

E. Ginna Reactor with ENC WREN-2 PWR Evaluation Model."

Exxon Nuclear Corporation Report, XN-NF-77-58, December 1977.

3.

"R.

E. Ginna Nuclear Power Station Cycle 8 Reload Safety Evaluation Report."

May 1, 1978.

4.

"Report on Small Break Accidents for Westinghouse NSSS System" Westinghouse Nuclear Energy Systems Report, WCAP-9600, June 1979.

5.

"Generic Evaluation of Feedwater Transients and Small Break Loss of Coolant Accidents in Westinghouse Designed Operating Plant", NUREG-0611, January 1980.

6.

Status.Report by the Directorate of Licensing in the Matter of Westinghouse Electric Company ECCS Evaluation Model Conformance to 10 CFR Part 50, Appendix K, October 15, 1974 and the Supplement, dated November 13, 1974.

7.

"R. E. Ginna Nuclear Power Plant Final Safety Analysis Report" September 1969.

REGULATOR NFORMATION DISTRIBUTION S TEM (RIDS)

ACCESSION NBR ~ 8107300043 DOC ~ DATE! 81/07/27 NOTARIZED NO DOCKK~Tt 0 FAGIL<:50 244 Rober t Emmet. Ginna* Nucl ear Plant'i Unit ir Rochester G

05000244' AUTH',NAME'UTHOR AFFILIATION C RUTCHF I EL D t D'>>

Operating Reactors" Branch 5-RECIP ~ NAME~

RECIPIENT AFF IL'IATION MAIER'rJ">> E ">>

Rochester:

Gas 8 Electric Corp'UBJECTS Forwards dr aft~ evaluation of SEP Topic XV 19~ LOCA>>Re'quests info re whether 'as, built facility differs. from. licensing.

basis assumed in assessment within 30 days of receipt of

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Docket No. 50-244 LS05 07-086 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 July 27, 1981 Mr. John E. Maier Vice President Electric and Steam Production Rochester Gas 8 Electric Corporation 89 East Avenue Rochester, New York 14649

Dear Mr. Maier:

JuL 2>>~

jgaoQW Qa'4I~~~ ~

lorp 6'UBJECT:

R.E.'INNA - SEPTOPIC XV-19 (SYSTEYjS),

LOSS OF COOLANT ACCIDENT Enclosed is our draft evaluation of SEP Topic XY-19 (systems).

Evaluation of the radiological consequences

'will be issued separately.

This evaluation compares your facility with the criteria currently used by'the regulatory staff for licensing new facilities.

Please inform us if your as-built facil-ity differs from the licensing basis assumed in our assessment within 30 days of receipt of this letter.

This evaluation will be'a basic input to the integrated safety assessment for your facility unless you identify changes needed to re'fleet as-built condi-tions'at your facility.

This topic assessment may be r'evised in the future if your facility design is changed or if NRC criteria relating to these topics are modified before the integrated as'sessment is completed.

Sincerely,

Enclosure:

Asstated'ji/

jp(;irreg(( '~i<4i 4<~,cl-Dennis M. Crutchfield, Chef Operating Reactors Branch No.

5 Division of Licensing cc w/enclosure:

See next page

'~O I

li p"

e J

R. E.

GINNA TOPIC.XV-19:

LOSS OF COOLANT ACCIDENTS RESULTING FROM SPECTRUM OF POSTULATED PIPING BREAKS WITHIN THE REACTOR COOLANT PRESSURE BOUNDARY I.

INTRODUCTION The capability of the R.E.

Ginna Emergency Core Cooling System-to miti- <

gate the consequences of a spectrum of Loss of Coolant Accidents (L'OCAs) is evaluated to assure that pipe breaks in the reactor coolant s'stem (RCS) do not result in a loss of core cooling capability.

Detailed ac-ceptance criteria for Emergency Core Cooling System (ECCS) performance are contained in 10 CFR 50.46 and in Standard Review Plan Sections 15.6.5, 6.3 and supporting appendices.

The five main criteria for ac-ceptance are:

1.

Peak clad temperature less than 2200'F 2.

Maximum cladding oxidation less than 17%

3.

Total hydrogen generation less than 1% of total zirconium in the'ctive fuel region 4.

Maintenance of eoolable geometry 5.

Long term coolability A spectum of break sizes up to and including a double-ende'd break of the largest pipe at various break locations is examined using an approved evaluation model which conforms to the requirements of Appendix K to 10 CFR 50 to verify that the acceptance criteria are met for a varie'ty of postulated loss of coolant accidents.

II.

EVALUATION The Ginna power plant ECCS provides emergency core cooling water at three delivery pressures.

The high pressure safety injection (HPI) system delivers borated water at up to 1500 psi.

This is different than'urrent PWR's which use safety grade charging pumps which are capable of inject-ion at operating pressure, about 2235 psi.

Intermediate pressure passive injection is provided by the accumulators which are held at 700 psi by nitrogen gas 'ov'erpressure.

The BPI sys'em and the accumulators discharge into lines to each cold leg.

Low pressure cooling water from the refuel-ing"water storage taiik is deli'vered by the residual heat removal (RHR) system which becomes available at 300 psi.

The low pressure injection flow is pumped directly into the reactor vessel through separate nozzles.

More complete descriptions of these systems are provided in the Ginna safe shutdown report, (Ref. 1),

and in the Final Safety Analysis Report (Ref. 7).

Switchover from injection to recirculation mode is covered in SEP "Topic VI-7.B.

Analysis for the large pipe breaks was performed for the Cycle 8 fuel reload in Reference 2, with the staff evaluation presented in Reference 3.

The small break analysis was performed generically for Westinghouse two-loop PWRs.during the initial Appendix K reviews.

Since the small breaks were clearly demonstrated to be non-limiting, later reloads re-evaluated only the large break spectrum.

In response to the NRC's Bulletins and Orders Task Force, additional small break analyses were performed on a'generic basis.

The Westinghouse calculations of Reference 4 were reviewed by the staff in Reference 5.

Lar e Break Anal sis The cycle 8 fuel reload safety analysis (Reference

2) examined six dif-ferent pipe breaks in the cold leg..Hot leg breaks were not examined.

Three double area guillotine breaks with discharge coefficients (Cn) of 1.0, 0.6,::and 0.4 were analyzed.

The other three breaks considere6 were split breaks with discharge coefficients of 1.0, 0.6, and 0.4.

The sel'ection of breaks for this analyses was justified, based on previous

. evaluations, which clearly identified the cold leg split and guillotine brea s as e mos smiting.

The a'ssumptions and computer codes used in the LOCA analyses are cover-ed in Reference 2.

Some of the more important assumptions include:

l. Initial power at 1025 2.

Reactor trip is neglected for large breaks 3.

All accumulator water bypasses core until termination of bypass 4.

Linear Heat Generation Rate of 13.76 kw/ft 5.

Total peaking factor is 2.32 6.

Fuel at beginning of life (BOL) conditions These and the other assumptions used for these analyses were in accord-ance with 10 CFR 50.46 and Appendix K, 10 CFR Part 50 and have;been shown (Ref.

2) to result in conservatively high peak cladding temperatures.

Results

- Lar e Break The limiting case of the six breaks examined in the ECCS analysis for Ginna was the double-ended cold leg guillotine break with Cn=0.4.

The peak clad temperature predicted was 1922'F, considerably beTow the 10 CFR 50;46 l.imit of 2200'F.

Clad oxidation (peak and total) was also well within limits. It should be noted that guillotine br eaks with a dis-charge coefficient smaller than 0.4 are not required in accordance with Reference 6.

Small Break Anal sis As discussed above, plant-specific small break analyses were not performed because it was felt that the small breaks would not be the limiting case since Westinghouse two-loop plant designs have historically been shown to be large break 1 imited.

Small Break Anal ses - Post TNI Generic analyses of small break LOCAs were submitted by Westinghouse in response to NRC Bulletins and Orders Task Force requirements.

The staff has accepted these analyses as a basis for providing information on plant response and as an aide to developing guidelines for operator action.

The staff considers these generic analyses to be representative of the response for Ginna to a postulated small break LOCA, however, the Ginna power plant would respond somewhat differently.to small breaks than more recent PWR's because the high pressure injection pump has a 1500 psia shutoff head.

Host PWR's (and all the latest) utilize a safety grade charging system for HPI which has a shutoff head above the operating pres-sure of 2250 psia.

The effect of the lower Ginna HPI shutoff head is that small breaks will result in depressurization to 1500 psi, and full repressurization of the RCS is not possible using the safety grade HPI.

Since 1500 psi corresponds to a saturation temperature of 596'F, and the hot leg temperature is over 600'F, a depressurization to 1500 psi could result in some voiding unless the reactor is cooled.

This added steam formation may be significant depending on when the reactor is trip-ped and the size of the break.

It is unlikely that core cooling would be affected during a small LOCA of this size, but the time to reach sub-cooled conditions may be extended in comparison with recent PWR's.

Results - Small Break - Post Tl1I As a result of the review of these analyses, the staff expressed concern about the applicability of current evaluation models and their applica-tion to the expanded scope of small break LOCA analyses now being consid-ered.

As part of the THI Task Action Plan, which is beyond the scope of the SEP review, Westinghouse is to revise and resubmit the small break analysis methods for staff approval.

Plant specific calculations, using these revised methods will then be required to show compliance with 10 CFR 50.46.

These analyses should place special emphasis on accidents which actuate the HPI.

III.

CONCLUSION The loss of coolant accidents analyzed for the Ginna nuclear power plant meet the acceptance criteria.

Hew small break LOCA analyses using revised evaluation models will be conducted as part of the THI Task Action Plan and will notbe included as part of the SEP review.

REFERENCES 1.

Safe Shutdown Systems f'r R.

E. Ginna Nuclear Power Plant, SEP Topic VII-3, May 13, 1981.

2.

"ECCS Analysis for R.

E. Ginna Reactor with EHC WREM-2 PWR Evaluation Model."

Exxon Nuclear Corporation Report, XH-NF-77-58, December 1977.

3.

"R.

E. Ginna Nuclear Power Station Cycle 8 Reload Safety Evaluation Report."

May 1, 1978.

4.

"Report on Small Break Accidents for 'Westinghouse NSSS System" Westinghouse Nuclear Energy Systems Report, WCAP-9600, June 1979.

5.

"Generic Evaluation of Feedwater Transients and Small Break Loss of Coolant Accidents in Westinghouse Designed Operating Plant", NUREG-0611, January 1980.

6.

Status Report by the Directorate of Licensing in the Matter of Westinghouse Electric Company ECCS Evaluation Model Conformance to 10 CFR Part 50, Appendix K, October 15, 1974 and the Supplement, dated November 13, 1974.

7.

"RE E. Ginna Nuclear Power Plant Final Safety Analysis Report" September 1969.