ML20126F560

Forwards Safety Evaluation on Item II.K.3.30 Re Small Break LOCA Analysis.Confirmation of Participation in Westinghouse Owner Group,Use of Notrump & Schedule for Completing Items II.K.3.30 & II.K.3.31 Requested within 30 Days

ML20126F560
Person / Time
Site:	Comanche Peak
Issue date:	06/06/1985
From:	Noonan V Office of Nuclear Reactor Regulation
To:	Spence M TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
References
TASK-2.K.3.30, TASK-2.K.3.31, TASK-TM GL-83-35, NUDOCS 8506170533
Download: ML20126F560 (12)

Text

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,4'p uag(Cg UNITED STATES y 7, NUCLEAR REGULATORY COMMISSION

• ;* ;j WASHINGTON, D. C. 20555

%, ...../ JUN 0 61985 Docket Nos.: 50-445 and 50-446 Mr. M. D. Spence President Texas Utilities Generating Company 400 N. Olive Street L . B . 81 Dallas, Texas 75201

Subject:

Resolution of TMI Action Items II.K.3.30 and II.K.3.31 Related to the Small Break LOCA Analysis for Comanche Peak 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 (WOG) 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.

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 Comanche Peak Steam Electric Station, Units 1 and 2. If this is correct, you should amend your FSAR to state that you are a member of the WOG, and you should reference WCAP-10079 and WCAP-10054 in stating that NOTRUMP is to be used for your small break LOCA analysis. This documentation will complete 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 30 days if our understanding of your participation in the WOG and use of NOTRUMP is not correct and provide your plans and schedule for completing II.K.3.30 and II.K.3.31.

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8506170533 850606

{DR ADOCM 05000445 PDR l

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. Mr. M. D. Spence .

On %verter 2, ISE3, in Ger.eric Letter No. 63-35, the FC crevided clarificatico and pre;csed a generic resolution cf TMI Actica Iter II.K.3.31.

That is, resciutien of II.C.3.31 ray be acemclished by gereric analysis te r decenstrate that the previous ar.alyses perfcmed with WFLASH were conservacive.

Future plant s;ecific analysis perfcmed fer y:cr plant by Westinghouse fer reloads or Technical Specificatice amer.*ents (those teyced 90 days of t6e i

date of this letter) should be calculated with tne r.ew code, .43TRtPP.

l Sincerely,  ;

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. .r. ,, recter for Cerarde eak .iect Divisico of Li ensing Office of helear 7eacter Regulatica

Enclosure:

As stated cc: See next page f

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C JUN 061985

. Mr. M. D. Spence I

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

. Future plant specific analysis perfonned for your plant by Westinghouse for I reloads or Technical Specification amendments (those beyond 90 days of the {

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

Sincerely, N

Vincent S. Noonan, Director for Comanche Peak Project Division of Licensing Office of Nuclear Reactor Regulation

Enclosure:

As stated cc: See next page DISTRIBUTION:

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COMANCHE PEAK 'jUk 06 70 Mr. M. D. Spence President Texas Utilities Generating Company 400 N. Olive St., L.B. 81 Dallas, Texas 75201 cc: Nicholas S. Reynolds, Esq. Mr. H. Shannon Phillips Bishop, Liberman, Cook, Resident Inspector / Comanche Peak Purcell & Reynolds Nuclear Power Station 1200 Seventeenth Street, N. W. c/o U. S. Nuclear Regulatory Washington, D. C. 20036 Commission '

P. O. Box 38 Robert A. Wooldridge, Esq. Glen Rose, Texas 76043 Worsham, Forsythe, Sampels &

Wooldridge Regional Administrator 2001 Bryan Tower, Suite 2500 U. S. NRC, Region IV Dallas, Texas 75201 611 Ryan Plaza Drive Suite 1000 Mr. Homer C. Schmidt Arlington, Texas 76011 Manager - Nuclear Services Texas Utilities Generating Company Lanny A. Sinkin, Executive Director Skyway Tower Nuclear Information and 400 North Olive Street Resource Service L. B. 81 1346 Connecticut Ave., N.W. 4th Floor Dallas, Texas 75201 Washington, D. C. 20036 Mr. Robert E. Ballard, Jr. B. R. Clements Director of Projects Vice President Nuclear Gibbs and Hill, Inc. Texas Utilities Generating Company 11 Penn Plaza Skyway Tower New York, New York 10001 400 North Olive Street, LB#81 Dallas, Texas 75201 {

Mr. A. T. Parker (

Westinghouse Electric Corporation Ms. Billie Pirner Garde l P. O. Box 355 Citizens Clinic Director Pittsburgh, Pennsylvania 15230 Government Accountability Project 1901 Que Street, N. W.

Renea Hicks, Esq. Washington, D. C. 20009 Assistant Attorney General Environmental Protection Division David R. Pigott, Esq.

P. O. Box 12548, Capitol Station Orrick Herrington & Sutcliffe

! Austin, Texas 78711 600 Montgomery Street San Francisco, California 94111 Mrs. Juanita Ellis, President Citizens Association for Sound Anthony Z. Roisman, Esq.

Energy Trial Lawyers for Public Justice 1426 South Polk 2000 P. Street, N. W.

tilas, Texas 75224 Suite 611 Ms. Nancy H. Williams CYGNA 101 California Street San Francisco, California 94111

COMANCHE PEAK cc: Mr. Dennis Kelley Resident Inspector - Comanche Peak c/o U. S. NRC P. O. Box 1029 Granbury, Texas 76048 Mr. John W. Beck Manager - Licensing Texas Utilities Electric Company Skyway Tower 400 N. Olive Street L. B. 81 Dallas, Texas 75201 Mr. Jack Redding Licensing Texas Utilities Generating Company 4901 Fairmont Avenue Bethesda, Maryland 20814 William A. Burchette, Esq.

Heron, Burchette, Ruckert & Rothwell Suite 700 1025 Thomas Jefferson St., N. W.

Washington, D. C. 20007 Mr. James McGauhy Southern Engineering Company of Georgia 1800 Peachtree Street, N. W.

Atlanta, Georgia 30367-8301 ,

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,' . Enclosure l

SAFETY EVALUATION IMI ACTION ITEti II.K.3.33 FOR WESTI!GHOUSE PLANTS NUREG-0737 is a report transmitted by a letter from D. G. Eisenhut, Director of the Division of Licensing, NRR, to licensees of operatirg 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 Com..ission requirements for the industry to demonstrate its small break loss of coolant i

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-0511, " 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.

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In response to TMI Action Item II.K.3.30, the Westinghouse Owners Group j (WOG) has elected to reference the Westinghoue NOTRUMP code as their new I 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 evaluations 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 docuter.ts our evaluation of the WOG response to TMI Action .

Ite II.K.3.30 confirmatory items.

II.

SUMMARY

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

A. Pree4de confirmatory' val'idation of the small break LOCA model tc 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." l III. EVALUATION The following is the staff's evaluation of the TMI Action Ite'm require-cents outlined above.

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A. Core Heat Transfer Models Tne Westinghouse Owners Group (WOG) referenced the NOTRUMP computer code as their new computer program for small break loss of coolant acc.ident (SBLOCA) evaluatid . NOTRUMP was bencnmarked against core unceve y experimer.ts conducted at the Oak Ricgs Nacional Labcratcry (ORNL). These tests were performed under NRC spcnsorship. The good agreement between the calculations and the data confirmed the adequacy cf the d*if t flux rede.1 used for core hyd a.lics s wel' as t.'.e ccre 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 adecuacy 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.

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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 correlations in the Westinghouse SBLOCA model and an assessment of 3

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ne 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:

(i) 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 SBLOCA was reanalyzed for a typical PWR. The WOG, thereby, l concluded that formation of noncondensible gases in quantities that i 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 uncovery or system pressure. What perturbation was observed was minor in nature.

The staff finds acceptable the Westinghouse submittal on the influences of noncondensible gases on design bases SBLOCA events. Our conclusion l is based on the limited amount of noncondensible gases available 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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D. Nocalization 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 questiened 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 the 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 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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F. Ccce Valication Folicwing the T.iree Mile Island event of 1979, staff analyses with NRC cevelcped computer codes led to cencerns that cetailec ncdali-Zatien was required to si=uiate realistic systers respcases to peste-lated 55LOCAs. As a consequence, licensees and applicants with Westing-house plants were requested to validate.their licensing tcols with ints;rai ex;sricents. -In specific, tne NRC repuested tr.a: tne cc:p.:er ceces be validated with the LOFT LS-1 and LS-7 experimental cata. In acditien, the staff also requested that the code be benchmarked with the Semiscale 5-UT-0E experimental cata.

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-05 test, Westinghouse de onstrated that NOTRUMP did a reasenable iot calculating the experi-mental data. However, this required a more detailed nodalization of the steam generators then used in the licensing codel. With the less detailed licensing nodalization, the pre-loop-seal-clearing core level depression phenomenon, as observed in the 5-UT-08 data, was not con-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 cc puter code and the associated nodalization for SSLOCA design basis evaluation.

This item is resolved.

IV. CONCLUSION The Westinghouse Owners Group (WOG), by referencing WCAP-10079.and WCAP-10054, have identified NOTRUMP as their new thermal-hydraulic'co:puter program for calculating small break loss of coolant accidents (SELOCAs). The  !

staff finds acceptable the use of NOTRUMP as the new Westingicuse licensing tool for calculating SSLOCAs for Westinghouse NSSS designs.

The responses to NUREG-0511 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

:e sees and a;plicants with Westingnouse NSSS designs who .ere me-bers cf :..e W3G and referenced WCAP-10079 and WCAP-10054 as their response to this item.

Within ene year cf receiving this SER, the licensees and applicants with Westinghouse NSSS designs are required to submit plant specific analyses with NOTRUMP, as required by THI Action Item II.K.3.31. Per generic letter 83-35, co ;iiance witn 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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