ML20129F260

Forwards Safety Evaluation of TMI Action Item II.K.3.30 Re Use of Westinghouse Small Break LOCA Model NOTRUMP.Plant- Specific Analyses Due within 1 Yr of Receipt of Ltr,Per TMI Item II.K.3.31

ML20129F260
Person / Time
Site:	Mcguire, Catawba, McGuire, 05000000
Issue date:	06/26/1985
From:	Novak T Office of Nuclear Reactor Regulation
To:	Tucker H DUKE POWER CO.
References
TASK-2.K.3.30, TASK-2.K.3.31, TASK-TM GL-83-35, TAC-58041, TAC-58042, NUDOCS 8507170293
Download: ML20129F260 (11)

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f(pME%qI UNITED STATES y ) 3m (gg NUCLEAR REGULATORY COMMISSION 5 /1 [ WASHINGTON, D. C. 20555

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a / June 26, 1985 cDodit[ hob 501365,5hA3E0[ 1"

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Mr. H. B. Tucker, Vice President Nuclear Production Department Duke Power Company -

422 South Church Street Charlotte, North Carolina 28242

SUBJECT:

TMI Action Item II.K.3.30, Catawba and McGuire Nuclear Stations 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 Westinghouse model was documented in the two Topical Reports, WCAP-10079 and WCAP-10054. The Westinghouse Owners Group (WOG1 references NOTRUMP as 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.

Since you are a member of the WOG and use NOTRUMP in the small break LOCA analysis for the Catawba and McGuire Nuclear Stations, this completes the TMI Action Item II.K.3.30 for these plants. In accordance with the TMI Action Item II.K.3.31, yoter plant-specific analyses are due within one year of receipt of this letter.

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, res-olution 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 analyses performed for your plants 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.

Sincerely, Thomas M. Novak, Assistant Director for licensing Division of Licensing

Enclosure:

As stated DESIG ATED ORIGINAL Certtrieg gy 0 (y y

8507170293 85062639 PDR ADOCK 050 P

1 CATAWBA Mr. H. B. Tucker, Vice President Nuclear Production Department Duke Power Company 422 South Church Street Charlotte, North Carolina 28242 cc: William L. Porter, Esq. North Carolina Electric Membership Duke Power Company Corp.

P.O. Box 33189 3333 North Boulevard Charlotte, North Carolina 28242 P.O. Box.27306 Raleigh, North Carolina 27611 J. Michael McGarry, III, Esq.

Bishop, Liberman, Cook, Purcell Saluda River Electric Cooperative, and Reynolds Inc.

1200 Seventeenth Street, N.W. P.O. Box 929 Washington, D. C. 20036 Laurens, South Carolina 29360 North Carolina MPA-1 Senior Resident Inspector Suite 600 Route 2, Box 179N 3100 Smoketree Ct. York, South Carolina 29745 P.O. Box 29513 Raleigh, North Carolina 27626-0513 Regional Administrator U.S. Nuclear Regulatory Commission, Mr. C. D. Markham Region II Power Systems Division 101 Marietta Street, N.W., Suite 2900 Westinghouse Electric Corp. Atlanta, Georgia 30323 P.O. Box 355 Pittsburgh, Pennsylvania 15230 Robert Guild, Esq.

P.O. Box 12097

. NUS Corporation Charleston, South Carolina 29412 2536 Countryside Boulevard Clearwater, Florida 33515 Palmetto Alliance 2135 i Devine Street Mr. Jesse L. Riley, President Columbia, South Carolina 29205 Carolina Environmental Study Group 854 Henley Place Karen E. Long Charlotte, North Carolina 28208 Assistant Attorney General N.C. Department of Justice Richard P. Wilson, Esq. P.O. Box 629 Assistant Attorney General Raleigh, North Carolina 27602 S.C. Attorney General's Office P.O. Box 11549 Columbia, South Carolina 29211 Piedmont Municipal Power Agency 100 Memorial Drive Greer, South Carolina 29551

i CATAWBA

• cc: . Spence Perry, Esquire Associate General Counsel Federal Emergency Management Agency Room 840 500 C Street, S.W.

Washington, D. C. 20472 Mark S. Calvert, Esq.

Bishop, Liberman, Cook, Purcell & Reynolds 1200 17th Street, N.W.

Washington, D. C. 20036 Mr. Michael Hirsch Federal Emergency Management Agency Office of the General Counsel Room 840 500 C Street, S.W.

Washington, DC 20472 Brian P. Caisidy, Regional Counsel Federal Emergency Management Agency, Region I .

J. W. McCormach P0CH Boston, Massachusetts 02109 4

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Mr. H. B. Tucker McGuire Nuclear Station Duke Power Company CC*

Mr. A. Carr Dr. John M. Barry Duke Power Company Department of Environmental Health P. O. Box 33189 Mecklenburg County 422 South Church Street 1200 Blythe Boulevard Charlotte, North Carolina 28242 Charlotte, North Carolina 28203 Mr. F. J. Twogood County Manager of Mecklenburg County Power Systems Division 720 East Fourth Street Westinghouse Electric Corp. Charlotte, North Carolina 28202 P. O. Box 355 Pittsburgh, Pennsylvania 15230 Chairman, North Carolina Utilities Commission Mr. Robert Gill Dobbs Building Duke Power Company 430 North Salisbury Street Nuclear Production Department Raleigh, North Carolina 27602 P. O. Box 33189 Charlotte, North Carolina ?8242 Mr. Dayne H. Brown, Chief Radiation Protection Branch J. Michael McGarry, III, Esq. Division of Facility Services Bishop, Liberman, Cook, Purcell Department of Human Resources and Reynolds P.O. Box 12200 1200 Seventeenth Street, N.W. Raleigh, North Carolina 27605 Washington, D. C. 20036 Mr. Wm. Orders Senior Resident Inspector c/o U.S. Nuclear. Regulatory Commission Route 4 Box 529 Hunterville, North Carolina 28078 Regional Administrator U.S. Nuclear Regulatory Commission, Region II 101 Marietta Street, N.W., Suite 2900 Atlanta, Georgia 30323 R. S. Howard Operating Plants Projects Regional Manager Westinghouse Electric Corporation - R&D 701 P. O. Box 2728 Pittsburgh, Pennsylvania 15230

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SAFETY EVALUATION TMI ACTION ITEff II.K.3.30 FOR CATAWBA AND MCGUIRE NUCLEAR STATIONS NUREG-0737 is a report transmitted by a letter from D. G. Eisenhut, Director 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 concerns 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.

This 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.

~ 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 greater phenomenological accuracy than capable by WFLASH.

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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.

E. Validate the polytropic expansion coefficient applied in the accumu-lator model, and F. Validate the SBLOCA model with LOFT tests L3-1 and L3-7. In addition, validate the model with the Semiscale S-UT-08 experimental data.

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

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

III. EVALUATION '

The following is the staff's evaluation of the TMI Action Iten require-ments outlined above. l l'

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A. Core Heat Transfer Models 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 National 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.

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

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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 i

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 k

in the steam generators. The heat transfer degradation was calculated

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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:

(i) Air dissolved in the RWST.

(ii) Hydrogen dissolved in the primary system.

(iii) Hydrogen in the pressurizer vapor space.

(iv) Radioly' tic 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 may reasonably be expected fcr a 4-inch cold leg break LOCA presents

. 30 serious detriment on the PWR system response in terms of core ui overy or system pressure. What perturbation was observed was mir.ir in nature.

2 The staff finds acceptable the Westinghouse submittal on the influences of nonc.ndensible gases on design bases 58LOCA events. Sur conclusion is based on the limited amount of noncondensible gases svailable dur-

] ing a des \gn basis SBLOCA event, as well as results obtained from Semi-l scale expe~iments which reached similar c.:aclusions while injecting noncondensiale gases in excess amount expected during a SBLOCA design 1

basis event. This item is resolved.

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D. Nodalization Studies For Flashing During Depressurization As a consequence of the staff's experience with modeling SBLOCA events with NRC developed computer codes (in particular the TMI-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 Evaluation Model to properly calculate the depressurization expected during a SBLOCA event.

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

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

E. Accumulator Model WFLASH, the previous Westinghouse small break loss of coolant accident (SBLOCA) analysis code, applied a polytropic gas expansion coefficient of 1.4 to the nitrogen in the accumulators. The WOG was requested to validate this accumulator model in light of data obtained through .ae l

LOFT experimental programs for SBLOCAs. Westinghouse reviewed the l applicable LOFT data and determined the need to perform full scale accumulator tests. Based upon these tests, Westinghouse modified the l

polytropic expansion coefficient to a more realistic value. Of inter-est is Westinghouse's conclusion that the selection of either a high 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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m Code Validation F.

Following the Taree Mile Island event of 1979, staff analyses with NRC developed computer codes le.1 to concerns that detailed nodali-zation was required to simulate reali,stic systems responses to postu-lated SBLOCAs. As a consequence, licensees and applicants with Westing-house plants were requested'to validate their licensing tools with.

. integral experiments. In specific, the NRC requested that the computer codesbevalidated[withtheLOFTL3-1andL3-7 experimental. data. In addition, the staff also requested that the code be benchmarked with

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y the Semiscale S-UT-08 experimental data.

Westingnouse perfornad 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 reasonableu jo' calculating the experi-mental data. However, this required a more detailed nodalization of thy l' team generators then used in jhe licensing model. With the less

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

. depression phenomenon, as observed in the S-UT-08' data, was not con-1 servatively calculated for very small breaks. However, the calculated

,, peak clad temperature was demonstrated to be higher (more conservative)

,) with the coarse nodalization. The staff, therefore, finds acceptable

. the NOTRUMP computer code and the associated nodalization for SBLOCA

)_; \ design W sis evaluation.

This item is resolved.

IV. CONCLUSION

, TheWestingtfoaseOwnersGroup(WOG),byreferencingWCAP-10079and j WCAP-10054, nave identified NOTRUMP as their new thermal-hydraulic I:omputer program for calculating sma!1 break loss of coolant accidents (SBLOCAs). The staff. finds acceptable the use of NOTRUMP as the new Westinghouse licensing l tool for calculating SBLOCAs for Westinghouse NSSS designs, s

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t The responses to NUREG-0611 concerns, as evaluated within'this SER, have also been found acceptable.

This SER completes 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 ith -

NOTRUMP, as required by THI 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 has not shifted away from the cold legs to the hot or pump suction legs.

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