ML20011D867

Forwards 10CFR50.46 Annual ECCS Model Changes Rept.Rept Covers Effect of ECCS Evaluation Model Mods on Peak Cladding Temp Results Reported in Chapter 15,Section 6 of FSAR

ML20011D867
Person / Time
Site:	Vogtle
Issue date:	12/22/1989
From:	Hairston W GEORGIA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
ELV-01184, ELV-1184, NUDOCS 9001020202
Download: ML20011D867 (10)

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December 22, 1989 t w wc ~

ELV-01184 Docket Nos. 50-424 ,

50-425 ,

U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Gentlemen:

Vogtle Electric Generating Plant 10CFR50.46 Annual ECCS Model Chanaes Report The October 17, 1988, revision to 10CFR50.46 required applicants and 4 holders of operating licenses or ' construction permits to notify the Nuclear  !

Regulatory Commission (NRC) of errors and changes in the ECCS Evaluation Models,~which-are not significant, on an annual basis. Enclosed is Georgia' Power Company's report in compliance with this requirement for the Vogtle Electric Generating Plant Units 1 and 2.

Attachment A provides information regarding the effect of the ECCS Evaluation Model modifications on the peak cladding temperature (PCT) results reported in Chapter 15, Section 6 of the Vogtle Electric Generating Plant Units 1-and 2 Final Safety Analysis Report (FSAR). ' Attachment B provides a sunnary of the plant change safety evaluations performed under the provisions of 10CFR50.59 that impact PCT. Please note that the facility change safety evaluations included in ' Attachment B reflect only those which result in non-zero PCT impact assessments. This information package constitutes Georgia- Power Company's report to the NRC as part of annual reporting required by 10CFR50.46(a)(3)(ii).  !

It has been determined'that compliance with the requirements of 10CFR50.46- l continues to be maintained when the effects of plant design changes, performed under 10CFR50.59, which.could affect the large break LOCA and small break LOCA analyses results are combined.with the effects of the ECCS Evaluation Model modifications applicable to Vogtle Units 1 and 2.

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U. S. Nucl' ear Regulatory Commission Page 2 i If you have-any questions regarding this report, please contact this office.

Sincerely, (A).h. ) bed W W. G. Hairston, III  !

WGH,III/ gps Attachment cc: Georaia Power Comoany Mr. C. K. McCoy Mr. G. Bockhold, Jr.

Mr. P. D. Rushton Mr. R. M. Odom j NORMS U.S. Nuclear Reaulatory Commission Mr. S. D. Ebneter, Regional Administrator  :

Mr. J. B. Hopkins, Licensing Project Manager, NRR Mr. J. F. Rogge, Senior Resident Inspector, Vogtle ,

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ATTACHNENT A EFFECT OF WESTINGH0USE ECCS EVALUATION MODEL' N0DIFICATIONS ON THE LOCA ANALYSIS RESULTS FOUND IN CHAPTER 15. SECTION 6 0F THE V0GTLE UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT BACKGROUND The October.17, 1988, revision to 10CFR50.46 required applicants and-holders of _ operating licenses. or construction permits to notify the Nuclear Regulatory Commission (NRC) of errors and changes in the Emergency Core Cooling System (ECCS) Evaluation Models on an annual basis, when the errors and changes are not significant. Reference 1 defines a significant error or change as one which results in a calculated fuel->eak cladding temperature (PCT) different by more than 500F from tie temperature calculated for the limiting transient using the last acceptable model, or as a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 500F.  ;

In References 2 and 3,-information regarding modifications to the.  !

Westinghouse large break and small break Loss-of-Coolant-(LOCA) ECCS Evaluation Models was submitted to the NRC. The following presents an assessment of the effect of the modifications to the Westinghouse ECCS Evaluation Models on the LOCA analysis results found in Chapter 15, Section 6 of the Vogtle Units 1 and 2 Final Safety Analysis Report (FSAR).

LARGE BREAK LOCA ECCS EVALUATION MODEL The large break LOCA analysis for Vogtle Units l'and-2 was examined.to assess the effect of the applicable modifications to the Westinghouse'large' break LOCA ECCS Evaluation Model on PCT results reported-in Chapter 15,

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Section 6 of the FSAR. The large break LOCA analysis results' were calculated using the 1981 version of the Westinghouse large break LOCA ECCS Evaluation Model in July 1988 (Reference 4). The limiting break analysis assumed the following information important to the large break LOCA analyses:

o 17x17 Standard Fuel Assembly 1 o Core-Power - 1.02

• 3411 MWT '

o Vessel Average Temperature - 589.60F -

0 Steam Generator Plugging Level - 5%

o Fg - 2.32  ;

o F-delta-H - 1.55 l

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ATTACHMENT A Page 2 For Vogtle Units 1 and 2, the limiting break resulted from the double-ended guillotine rupture of the cold leg piping with a discharge coefficient of CD = 0.6 for the maximum safeguards condition. The calculated PCT was 1995.80F.

The following modification to the Westinghouse ECCS Evaluation Models would affect the large break LOCA analysis results found in Chapter 15, Section 6 of the Vogtle Units 1 and 2 FSAR:

DOWNCOMER OVERFILLING DELAY 1981 ECCS Evaluation Model:

In the 1981 ECCS Evaluation Model, a modification as discussed in Reference 2 was made to delay downcomer overfilling. The delay corresponds to backfilling of the intact cold legs. Data from tests simulating cold leg injection during the post-large break LOCA reflood phase which have adequate safety injection flow.to condense all of the available steam flow show a significant amount of subcooled liquid to be present in the cold leg pipe test _ section. This situation corresponds to the so-called maximum safety injection scenario of ECCS '

Evaluation Model analyses. j For maimum safety injection scenarios, the reflooding model in the 4 Westinghouse 1981 ECCS Evaluation Model uses a WREFLOOD code version F which predicts the downcomer to overfill. Flow through the vessel = side of the break-is computed based upon the available head of water in the 4 downcomer in WREFLOOD using an incompressible flow in- an open channel i' method. -A modification to the WREFLOOD computer code was made to consider the cold leg inventory which would be present in conjunction -!

with the enhanced downcomer level in the non-faulted loops.

WREFLOOD code logic was altered to consider the filling'of the cold legs together with downcomer overfilling. Under this coding update, when the downcomer level exceeds its maximum value. as input to i WREFLOOD, liquid flow into the intact cold leg, asLwell as spillago out  :

'the break, is-considered. This logic modification stabilizes the j overfilling of the vessel downcomer.as it approaches its equilibrium i level. In some cases, this change could delay the downcomer _  !

overfilling' process, which could result in a PCT penalty. The -

l magnitude of the possible PCT penalty was assessed by reanalyzing a -

i plant which is maximum safeguards limited (Co - 0.6 Double-Ended Cold a Leg Guillotine case), and which'is most sensitive to the changes in the WREFLOOD code. The PCT penalty of 160F, which resulted for. this case, represents the maximum PCT penalty which could be exhibited for any plant due to the WREFLOOD logic change.

This change re) resents a model enhancement in terms of the consistency of the approac1 in the WREFLOOD code and the actual response of the

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ATTACHMENT A Page 3 downcomer level. Since Appendix K to 10CFR50 does not require the explicit treatment of the mass storage feature, this modification represents an enhancement rather than an error. However, to assess the margin available for accommodating potential plant changes, a 160F penalty in the peak cladding temperature will be tracked for this code modification.

RESULTANT LARGE BREAK LOCA PCT As discussed above, modifications to the Westinghouse large break LOCA ECCS Evaluation Model could affect the large break LOCA analysis results by altering the PCT as follows:

A. Analysis calculated result 1995.80F B. Modifications to Westinghouse ECCS Evaluation Model + 16.00F C. ECCS Evaluation Model Modifications Resultant PCT -2011.80F

[@CLUSION An evaluation of the effect of modifications to the Westinghouse large break 1981 ECCS Evaluation Model, as reported in Reference 2, was performed for the large break LGCA analysis results found in Chapter 15. Section 6 of the Vogtle Units 1 and 2 FSAR. When the effects of the ECCS model changes were combined with the current plant analysis results, it was determined that compliance with the requirements of 10CFR50.46 would be maintained.

SMALL BREAK LOCA ECCS EVALUATION MODEL The small break LOCA analysis for Vegtle Units 1 and _2 was also examined to assess the effect of the applicable modifications to the Westinghouse ECCS Evaluation Models on PCT results reported in Chapter 15, Section 6 of the FSAR. The small break LOCA analysis results were calculated using tha October 1975 version of the Wettinghouse small break LOCA ECCS EvaNation Model incorporating the WFLASH computer code. For Vogtle Units 1 and 2, the limiting size small break resulted from a four-inch equivalent diameter break in the cold leg. The calculated PCT was 15370F, The analysis assumed the following information important to the small break LOCA analyses:

o 17x17 Standard Fuel Assembly o Core Power = 1.02

• 3411 MWT

ATTACHNENT A Page 4 o Vessel Average Temperature = 589.60F o Steam Generator Plugging Level = 5%

o Fq = 2.20 at 10 ft o F-delta H ='l.55-As discussed below, the modifications to the Westinghouse ECCS Evaluation Models discussed in References 2 and 3 do not affect the WFLASH small break LOCA analysis results found in Chapter 15, Section 6 of the Vogtle Units 1 and 2 FSAR.

WFLASH ECCS EVALUATION MODEL Following the accident at Three Mile Island Unit 2, additional attention was focused on the small break LOCA, and Westinghouse submitted 4 report, WCAP-9600 (Reference 5), to the NRC detailing the performance of the WestNghouse small break LOCA Evaluation Model which utilized the WFLASH computer code. In NUREG 0611 (Reference 6), the NRC staff questioned the validity of certain models in the WFLASH computer code and required licensees to justify continued acceptance of the model.Section II.K.3.30 of NUREG-0737 (Reference 7) clarified the NRC post-TMI requirements regarding small break LOCA modeling and required licensee's to revise their small break LOCA ECCS models along the guidelines specified in NUREG-0611.

Following the issuance of NUREG-0737, Westinghouse and the Westinghouse Owners Group decided to develop the NOTRUMP (Reference 8) computer code for use in a new small break LOCA ECCS Evaluation Model (Reference 9). The NRC approved the use of NOTRUMP fx small break LOCA ECCS analyses in May 1985.

Since approval of the N01 RUMP small break LOCA ECCS Evaluation Model in '

1985, the WFLASH computer code has not been maintained as part of the Westinghouse ECCS Evaluation Model computer codes.

In Section II.K.3.31 of NUREG-0737, the NRC required that each licensee submit a new small break LOCA analysis using an NRC-approved small break l

LOCA Evaluation Model which satisfied the requirements of NUREG-0737 l section II.K.3.30. NRC Generic Letter 83-35 (Reference 10) relaxed the l requirements of item II.K.3.31, by allowing a more generic response and l providing a basis for retention of the existing small break LOCA analyses.

l Provided that the previously existing model results were demonstrated to be conservative with respect to the new small break LOCA model approved under the requirements of NUREG-0737 section II.K.3.30 (NOTRUMP), plant-specific analyses using the new small break LOCA Evaluation Model would not be '

required. In WCAP-ll145 (Reference 11), Westinghouse and the Westinghouse Owners Group demonstrated that the results obtained from calculations with l WFLASH were conservative relative to those obtained with NOTRUMP.

Compliance with item II.K.3.31 of NUREG-0737 has been completed by .

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i ATTACHMENT A Page 5 referencing WCAP-11145 as documented in Supplement 3 to the Vogtle Safety  :

Evaluation Report (Reference 12).

Westinghouse, therefore, has not been modifying, investigating, or  !

evaluating proposed changes to the WFLASH portion of the small break LOCA l ECCS Evaluation Model. There are no modifications to report.

SBLOCTA-IV COMPUTER CODE l

Modifications were made to the small break LOCTA-IV computer code used in i the small break LOCA ECCS Evaluation Model. Since the small break LOCTA-IV code modifications could, at most, result in a very small benefit, the i effect of modification to the small break LOCTA IV code modifications do ,

not need to be assessed or tracked. l RESULTANT SMALL BREAK LOCA PCT As discussed above, modifications to the Westinghouse small break LOCA ECCS Evaluation Model do not affect the small break LOCA analysis results and 40 i not alter the resultant PCT.

A. Analysis calculated result 1Dl0F l B. Modifications to Westinghouse ECCS Evaluation Model + 00F i j C. ECCS Evaluation Model Modifications Resultant PCT =1Dl0F CONCLUSION  ;

Aa evaluation of the effect of modifications to the Westinghouse small break October 1975 ECCS Evaluation Model using WFLASH was performed for the small break LOCA analysis results found in Chapter 15, Section 6 of the '

Vogtle Units 1 and 2 FSAR. When the effects of the small break ECCS model changes were combined with the current plant analysis results it was  ;

determined that compliance with the requirements of 10CFR50.46 would be  ;

maintained.

REFERENCES ,

1. " Emergency Core Cooling Systems; Revisions to Acceptance Criteria," i Federal Register, Vol. 53, No.180, pp. ' 35996-36005, dated September 16, 1988.

2. NS-NRC-89-3464, "10CFR50.46 Annual Notification for 1989 of Modifications in the Westinghouse ECCS Evaluation Models," Letter from W. J. Johnson (Westinghouse) to T. E. Murley (NRC), dated October 5, 1989.  ;

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l ATTACHMENT A Page 6 J

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3. NS NRC-89 3463, " Correction of Errors and Modifications to the NOTRUMP  !

Code in the Westinghouse Small Break LOCA ECCS Evaluation Model Which  !

Are Potentially Significant," Letter from W. J. Johnson (Westinghouse) {

to T. E. Murley (NRC), dated October 5,1989.

4. WCAP-9220 P A, Revision 1 (Proprietary), WCAP-9221-A, Revision 1 )

(Non Proprietary), " Westinghouse ECCS Evaluation Model - 1981 Version," i 1981 Eiche1dinger. C. j

5. " Report on Small Break Accidents for Westinghouse Nuclear Steam Supply System," WCAP-9601 (Non-Proprietary), June 1979, WCAP-9600  !

(Proprietary), June 1979.  ;

I

6. " Generic Evaluation of Feedwater Transients and Small Break  !

Loss of Coolant Accidents in We:tinghouse Designed Operating Plants,"  !

NUREG-0611, January 1980.  ;

7. " Clarification of TMI Action Plan Requirements," NUREG-0737, November 1980. l

8. "NOTRUMP - A Nodal Transient Small Break and General Network Code," ,

WCAP-10079 P A (Proprietary), WCAP 10080-A (Non Proprietary),  ;

Meyer, P. E., et al., August 1985.

9. " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP i Code," WCAP-10054-P A (Proprietary), WCAP-10081 A (Non-Proprietary),  !

Lee, N., et al., August 1985. j.

10. " Clarification of TM1 Action Plan Item II.K.3.31," NRC Generic Letter i 83-35 from D. G. Eisenhut, November 2, 1983.

11. ' Westinghouse Small Break ECCS Evaluation Model Generic Study with the NOTRUMP Code," WCAP-lll45-P A (Proprietary), WCAP-ll372-A,  !

(Non Proprietary), Rupprecht, S. D., et al., October 1986.

12. " Safety Evaluation Report Related to the Operation of Vogtle Electric .j Generating Plant, Units 1 and 2," NUREG-ll37, Supplement 3, dated i August 1986. )

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ATTACHNENT B EFFECT OF SAFETY EVALUATIONS PERFORNED i DN THE LOCA ANALY$!$ RESULTS FOUND IN CHAPTER 15, SECTION 6 0F THE  !

i V0GTLE UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT  !

l LARGE BREAK LOCA I

DESCRIPTION OF PLANT MODIFICATIONS j The large break Loss-of-Coolant (LOCA) analysis results have been j supplemented by safety evaluations of changes which could affect the PCT as i follows:

l. A safety evaluation to determine the effect for a change of the charging flow rates used in the FSAR Chapter 15, Section 6 large break  !

LOCA analysis due to increased runout flow of the charging pumps was performed for Vogtle Units 1 and 2. This evaluation determined that

' the large break LOCA analysis PCT results could be affected by a 20F increase.

2. A safety evaluation to determine the effect of a change in safety t injection flow was performed for the Vogtle Units I and 2 FSt.R Chapter 15, Section 6 large break LOCA analysis. This evaluation determined  ;

that the large breik LOCA analysis PCT results could be affected by a  :

30F increase.

3. A safety evaluation to determine the effect of containment purging during a LOCA was performed for the Vogtle Units 1 and 2 Chapter 15, .,

Section 6 large break LOCA analysis. This evaluation determined that the large break LOCA analysis PCT results could be affected by a 100F increase.

RESULTANT LARGE BREAK LOCA PCT  ;

As discussed above, plant modifications could affect the resultant PCT as follows: ,

Resultant PCT from ECCS Evaluation Model Modifications Reprted in Attachment A 20ll,80F

1. Safety Evaluation. for Charging Pump Increased Runout + 2_.00F

2. Safety Evaluation for Safety Injection Flow Changes + 3.00F

3. Safety Evaluation for Containment Purging + 10.00F Total Resultant PCT -2026.80F CONCLUSIONS It was determined that compliance with the requirements of 10CFR50.46 would be maintained when safety evaluations for changes which affected the large f

• ATTACHNENT B Page 2 break LOCA analysis results were combined with the effect of the large

' break ECCS Evaluation Model modifications applicable to Vogtle Units 1 and

2. )

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SMALL BREAK LOCA y

DESCRIPTION OF PLANT MODIFICATIONS The small break LOCA analysis results have been supplemented by a safety evaluation which could affect the PCT as-follows: )

1. A safety evaluation to determine the effect of changing instrumentation l uncertainties due to Veritrak transmitters was performed for the Vogtle  ;

Units 1 and 2 FSAR Chapter 15, Section 6 small break LOCA analysis. ]

This evaluation determined that the small break LOCA analysis PCT results could be affected by a 3.70F increase.

RESULTANT SMALL BREAK LOCA PCT As discussed above, plant modifications could affect the resultant PCT as follows- t l

Resultant PCT from ECCS Evaluation Model Modifications Reported in Attachment A 1537.00F

1. Safety Evaluation for Veritrak Transmitters + 3.70F l Total Resultant PCT -1540.70F CONCLUSIQ!iS It was determined that compliance with the requirements of 10CFR50.46 would be maintained when safety evaluations for changes which affected the small break LOCA analysis results were combined with the effect of the small break ECCS Evaluation Model modifications applicable to Vogtle Units 1 and

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