ML20059J557

Forwards 10CFR50.46 ECCS Model Significant Change Rept Based on WCAP-13451 & in Compliance W/Reporting Requirements of 10CFR50.46(a)(3)(ii)

ML20059J557
Person / Time
Site:	Vogtle
Issue date:	11/05/1993
From:	Mccoy C GEORGIA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LCV-0199, LCV-199, NUDOCS 9311120262
Download: ML20059J557 (9)

Text

?*

• Ger:ga Power Company 40 hverness Cemer Parkway Pod O' ice Got 1295 6;rTrogTiam aim >di!a 35201 IER#cre 205 877-7122 '

m

c. t uccoy Georgia Power r t.uuaem November 5, 1993 LCV-0199 ,

n Docket Nos. 50-424 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Gentlemen:

V0GTLE ELECTRIC GENERATING PLANT 10 CFR 50.46 ECCS MODEL SIGNIFICANT CHANGE REPORT Attached is Georgia Power Company's 10 CFR 50.46 emergency core cooling system (ECCS) model significant change report based on WCAP-13451 and in compliance with the reporting requirements of 10 CFR 50.46(a)(3)(ii). It is based on information provided by Westinghouse on October 6, 1993, concerning errors and changes assessed against the Vogtle ECCS models.

The assessed sum of the absolute magnitude of the errors / changes is greater than 500F for the Vogtle small break loss of coolant accident (SBLOCA);

however, the net effect of the errors / changes is a reduction in the SBLOCA peak cladding temperature (PCT) for Vogtle. The SBLOCA PCT results are affected by three issues: safety injection in the broken loop, the. improved condensation model, and the drift flux flow regime. The large break LOCA PCT results are not affected by these issues. The attached report also provides a summary of the plant change evaluations performed under the provisions of 10 CFR 50.59 that were previously assessed. The effects on the LOCA results will be incorporated in a future FSAR update.

Because the NOTRUMP small break ECCS model errors / changes affect many plants, the Westinghouse Owners Group is reviewing this issue and possible  :

development of a generic program for resolution. A schedule for resolution and reanalysis will be provided following the completion of the Westinghouse Owners Group review.

If you have any questions regarding this report, please contact this office.

Sincerely, C. K. McCoy h8 cc: (See next page}10003 4 'f i i

9311120262 9f'1105 "i PDR ADOCK 35000424 R PDR

Geoigia Powerk~ n .

J. S. Nuclear Regulatory Commission Page'2 W

CKM/BCA/HWM: gps Attachment '

cc: Georaia Power Company '-

Mr. J. B. Beasley, Jr.

Mr. M. Sheibani NORMS U.S. Nuclear Reaulatory Commission Mr. S. D. Ebneter, Regional Administrator ,

Mr. D. S. Hood, Licensing Project Manager, NRR Mr. B. R. Bonser, Senior Resident Inspector, Vogtle _

LCV-0199 Mcn1 i

ATTACHMENT V0GTLE ELECTRIC GENERATING PLANT 10 CFR 50.46 ECCS MODEL SIGNIFICANT CHANGE REPORT ,

BACKGROUND Provisions in 10 CFR 50.46 require 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, and within 30 days of discovery when the errors and

~

changes are significant. Reference 1 defines a significant error or change as one which results ;n a calculated fuel peak cladding temperature (PCT) different by more than 500F from the 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.

The following presents an assessment of the effects of the significant errors and changes to the Westinghouse ECCS evaluation models on loss of coolant accident (LOCA) analyses p6rformed for Vogtle Electric Generating Plant (VEGP) Units 1 and 2. The current LOCA analysis results were reported in the 1992 annual report (reference 1) based t the format  ;

presented in WCAP-13451 (reference 2). The LOCA analysis, assessments, and safety evaluation results reported herein will be included in a future VEGP Final Safety Analysis Report (FSAR) update. ,

i LARGE BREAK LOCA ECCS EVALUATION MODEL The large break LOCA (LBLOCA) analysis PCT for VEGP Units 1 and 2 remains unchanged since the 1992 annual report. The LBLOCA analysis results are based on the Westinghouse BASH large break ECCS evaluation model (reference

3) as approved by the NRC for VEGP-specific application (reference 4). The

• limiting size break analysis assumed the following information important to the LBLOCA analyses:

o 17x17 VANTAGE-5 Fuel Assembly o Core Power - 1.02

• 3565 Mwt o Vessel Average Temperature - 571.90F c Steam Generator Plugging Level - 10%

o FQ = 2.50 o F-delta-H = 1.65

1 t%

ATTACHMENT '

Page 2 for VEGP Units 1 and 2, the limiting size break continues to be the double-ended guillotine rupture of the cold leg piping with a discharge coefficient of CD - 0.6. The LBLOCA analysis calculated PCT remains 20250F .

(Reference 1).

The transition core penalty, containment purge, and Tavg uncertainty items continue to be listed separately per the format of WCAP-13451. These items are listed separately because they are not explicitly modeled. The PCT assessment values for these items are 10, 11, and 500F, respectively. ,

The steam generator flow area application and structural metal heat modeling assessments continue to be listed separately per the forrrt of WCAP-13451. These items are prior BASH large break ECCS model asses 'nts, the absolute sum of which do not exceed 500F. The PCT assessment vi 4 for these items are +10 and -250F, respectively.

NEW BASH LARGE BREAK ECCS LOCA MODEL ASSESSMENTS There are no new BASH large break ECCS model assessments since the 1992 annual report. l 10 CFR 50.59 EVALUATION ASSESSMENTS e Two plant changes pursuant to 10 CFR 50.59 which affect the LBLOCA analysis results have been evaluated. The evaluations were for fuel reconstitution and the addition of a permanent radiation shield. ,

Because the VEGP Unit 1 Cycle 5 core design contains one fuel assembly that has been reconstituted with one stainless steel filler rod, an evaluation of the LBLOCA PCT has been performed. The evaluation concluded that a 20F penalty must be assessed to the VEGP Unit 1 PCT results.

A permanent reactor vessel head radiation shield was installed on VEGP Unit  !

2. The previous radiation shielding was temporary, installed during the refueling outage and removed following the outage. The shield will now become an integral part of the reactor vessel head. Due to the additional heat sink surface area and mass, the vessel flooding rate is reduced, which results in a 10F increase in PCT. Thus, the evaluation concluded that a i 10F penalty must be assessed to the VEGP Unit 2 PCT results. At the next i refueling outage for VEGP Unit 1, the permanent shielding will also be installed, and the same IDF penalty will be assessed to VEGP Unit 1.

l 1

LICENSING BASIS LBLOCA PCT Based on the above discussions concerning the VEGP-specific application of the Westinghouse BASH large break ECCS Evaluation Model, the licensing I basis LBLOCA PCT is as follows:-  !

I

1 ATTACHMENT Page 3

.A. 1992 Annual Report LBLOCA Analysis of Record (Reference 1)

1. BASH Large Break ECCS Model Analysis Result 2025.00F

2. Evaluation for Containment Purging + 10.00F

3. Evaluation for +/- 60F Uncertainty Bard + 11.00F '

4. Evaluation for Transition Cycle Penalty + 50.00F B. BASH Large Break ECCS Model Assessmants Prior to 1993

1. Steam Generator Flow Area Application + 10.00F

2. Structural Metal Heat Modeling - 25.00F C. 10 CFR 50.59 Evaluations

1. Fuel Reconstitution (VEGP Unit 1) + 2.00F

2. Permanent Radiation Shield (VEGP Unit 2) + 1.00F D. 199310 CFR 50.46 BASH Large Break LLCS Model /.ssessments None + 0.00F LICENSING BASIS LBLOCA PCT (Unit 1) -

ZD31J0F (Unit 2) - 2082.00F CONCLUSION An evaluation of the effect of errors and changes to the Westinghouse BASH 1arge break ECCS evaluation model was performed on the LBLOCA analysis results. When the effects of the BASH ECCS model errors / changes and safety evaluations were combined with the VEGP LBLOCA analysis results, it was determined that VEGP Units 1 and 2 are in compliance with the requirements of 10 CFR 50.46(b).

SMALL BREAK LOCA ECCS EVALUATION MODEL Significant errors / changes were assessed against the small break LOCA (SBLOCA) analysis PCT for VEGP Units 1 and 2 since the 1992 annual report.

The SBLOCA analysis results are based on the Westinghouse NOTRUMP small break ECCS evaluation model (reference 5) as approved by the NRC for VEGP-specific application (reference 4). The limiting size break analysis assumed the following information important to the SBLOCA analyses:

o 17x17 VANTAGE-5 Fuel Assembly e Core Power - 1.02

• 3565 MWT o Vessel Average Temperature - 571.90F o Steam Generator Plugging Level - 10%

a,

i

.; w ,

ATTACHMENT ,

Page 4 o FQ = 2.48 at 9.5 ft o F-delta-H - 1.70 For VEGP Units 1 and 2, the limiting size small break continues to ha a 3-inch equivalent diameter break in the cold leg. The current SBLOCA analysis calculated PCT is 18090F (reference 1).

The steam ger 'rator lower level tap raiocation and Tava uncertainty items '

continue to bi listed separately per the format of WCAP-13451. These items are listed se;arately because they are not explicitly modeled. The PCT assessment va ues on these items are 15 and 40F, respectively.

The Bessel function error was previously assessed as a 250F penalty (reference 1) to the NOTRUMP small break ECCS model and is' listed separately per the format of WCAP-13451.

NEW NOTRUMP SMALL BREAK ECCS MODEL ASSESSMENIS The following errors and changes to the NOTRUMP small break ECCS evaluation model would affect the SBLOCA PCT analysis results:

Safety In.iection in the Broken Loon Westinghouse has completed an evaluation of a potential issue concerning the modeling of safety injection (SI) flow into the broken reactor coolant ,

system (RCS) loop for the SBLOCA. In previous analyses, Westinghouse assumed that modeling SI flow into the broken RCS loop would result in a lower calculated PCT because additional SI flow would be expected to provide additional core cooling. Therefore, in previous analyses, SI flow into the broken RCS loop was modeled as spilling directly into the containment sump to provide conservative PCT results. Recent evaluations indicate for SBLOCA events using NOTRUMP, that modeling SI flow into the ,

broken RCS loop will actually result in a significant increase in PCT. The increase in PCT occurs as a result of competition between the steam venting Nt the break and the SI to the broken loop both exiting through the break.

Ine competition between the steam and the SI results in higher RCS pressures, therefore, lower delivered SI flow rates to the intact RCS loops and an increase in PCT. The effect of this issue is a 1500F increase in PCT. Therefore, a 1500F penalty has been assessed against the VEGP SBLOCA PCT results.

Imoroved Steam Condensation Model lhe SI in the broken loop issue described above is significant with regard to the effect on calculated SBLOCA PCT results. An offsetting steam condensation benefit has been identified which more than offsets the above 1500F PCT penalty. Improved condensation of the loop steam in the intact loops results in lower RCS pressures and increased SI' flow rates. The

n. .

I i ATTACHMENT ,

Page 5 '

i increased SI flow rates result in a lower calculated PCT. Thus, the- -

negative effects of SI in the broken loop can'be fully offset by an '

improved SI steam condensation model in the intact RCS loops. Hence, e y 1500F benefit has been assessed against the VEGP SBLOCA PCT results to .

offset the PCT penalty for SI in the broken loop.

The Westinghouse Owners Group is reviewing both issues, SI in the broken loop and the improved SI steam condensation model, for possible development of a generic program for resolution that is acceptable to the NRC. '

Westinghouse is keeping the NRC (Reactor Systems Branch) informed of-the resolution of these two issues.

Drift Flux Flow Reaime Errors Errors were discovered by Westinghouse in both WCAP-100B1-A and related ,

coding in NOTRUMP where the improved TRAC-P1 vertical flow regime map is evaluated. This flow regime map is only used during counter-current flow >

conditions in vertical flow links. The error occurs as a result of an i unbounded parameter which may lead to a discontinuity in the flow regime i map under certain circumstances. The effect of this issue is a 130F decrease in PCT. The error has been correct?d in NOTRUMP. A benefit of 130F has been assessed against the VEGP SBLOCA PCT results.

li0 CFR 50.59 EVALUATION ASSESSMENTS Two plant changes pursuant to 10 CFR 50.59 which affect the SBLOCA analysis results have been evaluated. The evaluations were for fuel reconstitution i

1 and a potential loose part.

Because the VEGP Unit 1 Cycle 5 core design contains one fuel assembly that '

has been reconstituted with one stainless steel filler rod, an' evaluation of the SBLOCA PCT results has been performed. The evaluation concluded  ;

that a 10F penalty must be assessed to the VEGP Unit 1 SBLOCA PCT results.

An evaluation was performed to determine the effect of a loose part 'in the 'l RCS. The loose part has been identified as the ring from the quick click '

pin on the fuel handling tool. The ring was lost in the spent fuel pool-and may have been carried to the RCS by a fuel assembly. The evaluation concluded that a 20F PCT penalty must be assessed against the VEGP Unit 1 i SBLOCA PCT results. )

1 LICENSING BASIS SBLOCA PCT'  !

Based on the above discussions concerning the VEGP-specific application of the Westinghou:e NOTRUMP small break ECCS evaluation model, the licensing basis SBLOCA PCT is as_ follows:

l c: . ., 1 1

ATTACHMENT Page 6 ,

A. 1992 Annual Report SBLOCA Analysis of Record (Reference 1)

1. NOTRUMP Small Break ECCS Model Analysis Result 1809.00F i

2. Evaluation for Steam Generator Lower Level Tap Relocation + 15.00F

3. Evaluation for +/- 60F Uncertainty Band -

4.00F B. NOTRUMP Small Break ECCS Model Assessments Prior to 195 Bessel function error + 25.00F C. 10 CFR 50.59 Evaluations

1. Fuel Reconstitution (VEGP Unit 1) + 1.00F

2. Loose Parts (VEGP Unit 1) + 2.00F .

D. 199310 CFR 50.46 NOTRUMP Small Break ECCS Model Assessments '

l. SI in Broken Loop + 150.00F

2. Improved Condensation Model - 150.00F

3. Drift Flux Flow Regime -

13.00F LICENSING BASIS SBLOCA PCT (Unit 1) - 1843.00F (Unit 2) = 1840.00F CONCLtL510E An evaluation of the effect of errors and changes to the Westinghouse NOTRUMP small break ECCS Evaluation Model was performed on the SBLOCA 4 analysis results. When the effects of the NOTRUMP' ECCS model errors / changes and safety evaluations were combined with the VEGP SBLOCA ,

analysis results, it was determined that the sum of the absolute magnitude J of the errors / changes is greater than 500F. Therefore, in compliance with 10 CFR 50.46(a)(3)(ii), a 30-day report to the NRC is required. However, the net effect of the errors / changes is a reduction in SBLOCA PCT results.

Because the NOTRUMP sndll break ECCS errors / changes affect many plants, the '

Westinghouse Owners Group is reviewing this issue and possible development .

of a generic program for resolution. A schedule for resolution and '

reanalysis will be provided following the completion of the Westinghouse i Owners Group review.

REFERENCES

1. ELV-05255, "Vogtle Electric Generating Plant,10 CFR 50.46 ECCS Model 1992 Report," letter from C. K. McCoy-(GPC) to USNRC, dated February 23, 1993.  :)

1

2. WCAP-13451, " Westinghouse Methooology for Implementation of 10 CFR 50.46 Reporting," dated October 1992.

1

3. "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," WCAP-11524-A (Non-Proprietary), March 1987.

I l

. J

i ATTACHMENT-Page 7

4. ELV-02166, "Vogtle Electric Generating Plant, Request for Technical Specifications Changes VANTAGE-5 Fuel Design," letter from W. G. Hairston, III, to USNRC, dated November 29, 1990.

5. " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP ,

Code," WCAP-10081-A (Non-Proprietary).

,