ML20129A491

Requests Confirmation That New Westinghouse Small Break LOCA Model,Notrump,Will Be Used in Analysis,Per TMI Action Items II.K.3.30 & II.K.3.31.Plans & Schedule for Completing Items Also Requested.Safety Evaluation Encl

ML20129A491
Person / Time
Site:	Farley
Issue date:	05/23/1985
From:	Varga S Office of Nuclear Reactor Regulation
To:	Mcdonald R ALABAMA POWER CO.
References
TASK-2.K.3.30, TASK-2.K.3.31, TASK-TM NUDOCS 8506040626
Download: ML20129A491 (10)

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'May 23, 1985 i

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- Docket Nos.'50-348 Distribution sDockeMfile~

and 50-364 NRC PDR L'PDR' ~~

ORB #1 RDG Gray file HThompson -0 ELD EJordan BGrimes Mr.-R. P.~ Mcdonald JPartlow ACRS (10)'

- Senior Vice President EReeves.$) CParrish Alabama Power. Company-

- Post Office Box 2641 Binningham, Alabama 35291

SUBJECT:

. COMPLETION OF REVIEW 0F ITEM II.K.3.30, REVISED SMALL BREAK LOCA METHODS, NUREG-0737 FOR JOSEPH M. FARLEY NUCLEAR MANT, UNITS 1 AND 2 OnMay[211985, the NRC approved the new Westinghouse small break LOCA model, NOTRUMP, for use in satisfying the TMI Action Item II.K.3.30. The ical Reports, WCAP-10079

-Westinghouse and.WCAP-10054. model was documented The Westinghouse in the two Top (WOG) references NOTRUMP Owners Group their new licensing small break LOCA model to satisfy the requirements of

- TMI Action Item-II.K.3.30. Our Safety Evaluation of II.K.3.30 for the

~ members of WOG is enclosed.

. It is our understanding that you are a member of the WOG and that NOTRUMP is to be used in the small break LOCA analysis for the Farley Nuclear Plant,

. Units 1 and 2. If this is corre'ct, this completes the TMI Action Item II.K.3.30 for your plant. In accordance with the TMI Action Item II.K.3.31, your plant specific analysis is due within one year of receipt of this ,

letter. Please advise this office.within 60 days if this is not correct and provide your plans and schedule for completing II.K.3.30 and II.K.3.31.

On November 2,1983 in Generic Letter No. 83-35,' the NRC provided clarification and proposed a generic resolution of TMI Action Item II.K.3.31.

That .is, _ resolution of II.K.3.31 may be accomplished by generic analysis to demonstrate that the previous analyses performed with WFLASH were conservative.

Future plant specific analysis performed for y.our plant by Westinghouse for

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reloads or Technical Specification amendments (those beyond 90 days of the _.

date of this letter) should be calculated with the new code, NOTRUMP.

Sincerely,

/s/SAVarga Steven A. Varga, Chief-Operating Reactors Branch #1 Division of Licensing i

Enclosure:

- As: stated 8506040626 850523 i- PDR ADOCK 05000348

- Cc:

D. Wigginton ,

ORB #1:DL .0RB -

L CParrishfp ERe ves/ts SValp 05 85

- 05$85 Q . 05/g85 '

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UNITED STATES

-8 o NUCLEAR REGULATORY COMMISSION J j wAsHNTON, D. C. 20555

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May 25,.-1985 Docket Nos. 50-348 and 50-364

'Mr. R. P. Mcdonald Senior Vice President Alabama Power Company Post Office Box 2641 Birmingham, Alabama 35291 .

SUBJECT:

COMPLETION OF REVIEW 0F. ITEM II.K 3.30, REVISED SMALL BREAK LOCA METHODS, NUREG-0737 FOR JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2-On.May 21 1985, the NRC approved the new Westinghouse small break LOCA-model,. NOTRUMP, for use in satisfying the TMI Action Item II.K.3.30. The ical Reports, WCAP-10079 Westinghouse and WCAP-10054.:model was documented The Westinghouse in the two Top (WOG) references NOTRUM Owners Group their new licensing small break LOCA model to satisfy the requirements of TMI-Action' Item II.K.3.30. Our Safety Evaluation of II.K.3.30 for the-members of WOG is enclosed..

It is our understanding that you are a member of the WOG and that NOTRUMP is

~ to be used in the small break LOCA analysis for the Farley Nuclear Plant,

' Units 1 and 2. If this'is correct, this completes the TMI Action. Item II.K.3.30.for you'r plant. In accordance with the TMI Action Item II.K.3.31,

your ' plant specific analysis is due within one year of receipt of this letter. Please advise this office within 60 days if this-is not correct and

. . provide your plans 'and schedule for completing II.K.'3.30 and II.K.3.31.- ,

On NovemLer 2, 1983 in Generic Letter No. 83-35, the NRC provided clarification and oroposed a generic resolution of TMI Action Item II.K.3.31.

That is, resolution of II.K.3.31 may be accomplished by generic analysis to demonstrate that-the previous analyses performed with WFLASH were conservative..

- Future' plant specific analysis performed for your plant by Westinghouse for reloads or Technical Specification amendments (those beyond 90 days of the-date of this letter) should be calculated with the new code, NOTRUMP.

l. Sincerely, e a/g , C t j Operating Reactors nch #1

, Division of Licensing

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Enclosure:

As stated CC:

D. Wigginton

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Mr. R. P. Mcdonald - Joseph M. Farley Nuclear Plant Alabama Power Company cc: -Mr.:W. O. Whitt .

D. Biard MacGuineas Esquire Executive Vice President- Volpe, Boskey and Lyons Alabama Power Company 918 16th Street, N.W.

Post Office Box 2641

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Washington, DC 20006 Birmingham, Alabama.- 35291~

. Charles R. Lowman Mr.-Louis B. Long, General Manager Alabama Electric Corporation Southern Company Services, Inc. Post Office Box 550 Post Office Box 2625 Andalusia, Alabama 36420 Birmingham,' Alabama 35202 Dr. J. Nelson Grace Chairman Regional Administrator - Region II E Houston County Commission U.S. Nuclear Regulatory Commission

. Dothan, Alabama 36301 101 Marietta Street, Suite- 2900

-Atlanta, Georgia 30303 George F.' Trowbridge, Esquire

. Shaw, Pittman, Potts and Trowbridge Ira L. Myers, M.D.

-1800 M Street, N.W. ' State Health Officer

-Washington, DC 20036_ State Department of Public Health State Office Building Robert A.. Buettner, Esquire . - Montgomery, Alabama 36130

-Balch, Bingham, Baker, Hawthorne, Williams and Ward Post Office Box 306 Birmingha~m, Alabama 35201 Resident Inspector .

'U.S. Nuclear ~ Regulatory Commission

-Post Office Box 24 - Route 2 ^

Columb'ia, Alabama .36319 State Department of Public Health 1 ATTN: State Health Officer > '

State Office Building Montgomery, Alabama 36104 '

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SAFETY EVALUATION TMI ACTION ITEM II.K.3.30 FOR WESTINGHOUSE PLANTS 2

NUREG-0737 is a report transmitted by a letter from D. G. Eisenh'ut, Director i of the Division of Licensing, NRR, to licensees of operating power reactors and applicants for operating reactor licenses forwarding TMI Action Plan requirements which have been approved'by the Commission for implementa-tion.'.Section II.K.3.30 of Enclosure 3 to NUREG-0737 outlines the Commission requirements for the industry to demonstrate its small break loss of coolant accident (SBLOCA) methods continue to comply with the requirements of Appendix K to 10 CFR Part 50.

The technical issues to be addressed were outlined in NUREG-0611, " Generic

- Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in

' Westinghouse-Designed Operating Plants." In' addition to the cuncerns listed in

= NUREG-0611, the staff requested licensees with U-tube steam generators to -

assess their computer codes with the Semiscale S-UT-08 experimental results.

Th.is request was made to validate the code's ability to calculate the core coolant level depression as influenced by the steam generators prior to loop seal clearing.

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' ' In response to TMI Action Item II.K.3.30, the Westinghouse Owners' Group (WOG) has elected to reference the Westinghoue NOTRUMP code as their new

, licensing small break LOCA model. . Referencing the new computer code did not imply deficiencies in WFLASH to meet the Appendix K requirements. ,The decision was based on desires of the industry to perform licensing evaluatio'ns with a ,

computer program specifically designed to' calculate small break LOCAs with ,

y, - greater phenomenological accuracy than capable by WFLASH. ,

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l The following documents our evaluation of the WOG response to TMI Action Item II.K.3.30 confirmatory items.

II.

SUMMARY

OF REQUIREMENTS NUREG-0611 required licensees and applicants with Westinghouse NSSS designs to address the following concerns:

A. Provide confirmatory validation of the small break LOCA model to

. adequately calculate the core heat transfer'and two phase coolant level during core uncovery conditions.

B. Validate the adequacy of modeling the primary side of the steam generators as a homogeneous mixture.

C. Validate the condensation heat transfer model and affects of non-condensible gases.

D. Demonstrate, through noding. studies, the adequacy of the SBLOCA model to calculate flashing during system depressurization.

5 E. Validate the polytropic expansion coefficient applied in the accumu- [

lator model, and

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F. Validate the SBLOCA model with. LOFT tests L3-1 and L3-7. In addition, I validate the model with the Semiscale S-UT-08 experimental data.

Detailed responses to the above items are documented in WCAP-10054, l

" Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code."

B III. EVALUATION The following s the staff's evaluation of the TMI Action Ite$ require-ments outlined above.

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l A. Core Heat Transfer Models i

The Westinghouse Owners Group (WOG) referenced the NOTRUMP computer code as their new computer program for small break loss of coolant accident (SBLOCA) evaluation. NOTRUMP was benchmarked against core uncovery experiments conducted at the Oak Ridge Natiogal , Laboratory (ORNL). These tests were performed under NRC sponsorship. The good agreement between the calculations and the data confirmed the adequacy of the drift flux model used for core hydraulics as well as the core heat transfer models of clad temperature predictions- .

The staff finds the core thermal-hydraulic models in NOTRUMP accept-able. This item is resolved.

B. Steam Generator Mixture Level Model NUREG-0611 requested licensees and applicants with Westinghouse designed NSSSs to justify the adequacy of modeling the primary system of the steam generators as a homogeneous mixture. This question was directed to the WFLASH code. NOTRUMP, the new SBLOCA licensing code models phase separation and incorporates flow regime maps within the steam generator tubes. The adequacy of this model was demonstrated through benchmark analyses with integral experiments, in particular with Semiscale test S-UT-08.

The staff finds the steam generator model in NOTRUMP acceptable.

, This item is resolved.

C. Noncondensible Affects On Condensation Heat Transfer  ?

NUREG-0611 requested validation of the condensation heat transfer 8 correlations in the Westinghouse SBLOCA model and an assessment of 3

m, the consequences of noncondensible gases in the primary coolant.

The condensation heat transfer model used in NOTRUMP is based on steam experiments performed by Westinghouse on a 16-tube PWR steam generator model. For two phase conditions, an empirical correlation developed by Shah is applied.

The staff finds the condensation heat transfer correlation in NOTRUMP acceptable.

The influences of noncondensible gases on the condensation heat transfer was demonstrated by degrading the heat transfer coefficient in the steam generators. The heat transfer degradation was calculated using a boundary layer approach. For this calculation, the noncon '

densible gases generated within the primary coolant system were col-lected and deposited on the surface of the steam generator tubes.

The sources of noncondensibles considered were:

(1) Air dissolved in the RWST.

(ii) Hydrogen dissolved in the primary system.

(iii) Hydrogen in the pressurizer vapor space.

(iv) Radiolytic decomposition of water.

With a degradation factor on the heat transfer coefficient, the limiting 58LOCA was reanalyzed for a typical PWR. The WOG, thereby, concluded that formation of noncondensible gases in quantities that 1 may reasonably be expected for a 4-inch cold leg break LOCA presents no serious detriment on the PWR system response in terms of core l uncovery or system pressure. What perturbation was observed was l minor in nature. I The staff finds acceptable the Westinghouse submittal on the influences of noncondensible gases on design bases SBLOCA events. Our conclusion is based on the limited amount of noncondensible gases a(ailable dur-ing a design, basis SBLOCA event, as well as results obtained from Semi- ,

scale experiments which reached similar conclusions while injecting noncondensible gases in excess amount expected during a SBLOCA design basis event. This item is resolved.

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,7 D. Nodalization Studies For Flashing During Depressurization P

As a consequence of the staff's experience with modeling 58LOCA events with NRC developed computer codes (in particular the THI-2 accident), the staff questioned the adequacy of the nodalization in the _ licensing model to calculate the depressurization of the primary system.

The staff therefore requested validation of the Westinghouse a Evaluation Model to properly calculate the. depressurization expected during~a 58LOCA event.

Through nodalization studies and validation of the NOTRUMP licensing model with integral experiments (e.g., LOFT and Semiscale), Westing-4 house demonstrated the acceptability of the nodalization and nonequi-librium models.

I The staff finds the Westinghouse model acceptable for calculating depressurization during S8LOCA events. This item is resolved.

E. Accumulator Model

• WFLASH, the previous Westinghouse small break loss of coolant accident

" (S8LOCA) analysis code, applied a polytropic gas expansion coefficient of 1.4 to the nitrogen in the accumulators. The WOG was requested to .

validate.thisaccumulatormodeiinlightofdataobtainedthroughthe  !

. LOFT experimental programs for SBLOCAs. Westinghouse reviewed the applicable LOFT data and determined the need to perform full scale accumulator tests. Based upon these tests Westinghouse modified the I' polytropic expansion coefficient to a more realistic value. Of inter-est is Westinghouse's conclusion that the selection of either a high j or low expansion coefficient had negligible effect on the calculated peak clad temperature (PCT). This insensitivity is only appropriate -

to NOTRUMP, with its nonequilibrium assumptions. '

The staff finds acceptable the polytropic expansion coefficient in '

the NOTRUMP code. This item-is resolved.

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F. Code Validation

, Following the Taree Mile Island event of 1979, staff analyses with

, NRC developed computer codes led to concerns -that detailed nodali-zation was required to simulate realistic systems responses to postu-lated SBLOCAs. As a consequence, licensees and applicants with Westing-1 house plants were requested to validate their licensing tools with integral experiments. In specific, the NRC requested that the computer codes be. validated with the LOFT L3-1 and L3-7 experimental data. In addition, the staff also requested that the code be benchmarked with

the Semiscale S-UT 08 experimental data.

Westinghouse performed the above benchmark analyses. For the LOFT tests, Westinghouse showed good agreement between the NOTRUMP calcu-

lations and the experimental data. For the S-UT-08 test, Westinghouse demonstrated that NOTRUMP did a reasonable job calculating the experi-mental data. However, this required a more' detailed nodalization of the. steam generators then used in the licensing model. With the less

detailed licensing nodalization, the pre-loop-seal-clearing core level

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i depression phenomenon, as observed in the S-UT-08 data, was not con-L servatively calculated for very small breaks. However, the calculated peak clad temperature was demonstrated to be higher (more conservative)'

with the coarse : .dalization. The staff, therefore, finds acceptable the NOTRUMP computer code and the associated nodalization for SBLOCA l design basis evaluation..

This item is resolved.

IV. CONCLUSION i

The Westinghouse Owners Group (WOG), by referencing WCAP-10079jand WCAP-10054, have identified NOTRUMP as their new thermal-hydraulic iomputer program for calculating small break loss of coolant accidents ($8LOCAs). .The staff finds acceptable the use of NOTRUMP as the new Westinghouse licensing l tool for calculating S8LOCAs for Westinghouse NSSS designs.

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t The responses to NUREG-0611 concerns, as evaluated within this SER, have also been fou.7d acceptable.

This SER comple,tes the requirements of TMI Action Item II.K.3.30 for licensees and applicants with Westinghouse NSSS designs who were members of the WOG and referenced WCAP-10079 and WCAP-10054 as their response to this item.

Within one year of receiving this SER, the licensees and applicants with Westinghouse NSSS designs are required to submit plant specific analyses with

, NOTRUMP, as required by TMI Action Item II.K.3.31. Per generic letter 83-35, compliance with Action Item II.K.3.31 may be submitted generically. We require that the generic submittal include validation that the limiting break location a not shifted away from the cold legs to the hot or pump suction legs.

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