ML18153D110

Forwards Rept of ECCS Evaluation Model Changes,Per 10CFR50.46.Calculated Peak Clad Temp for Small & Large Break LOCA Analyses Listed.No Generic Changes to ECCS Evaluation Models Implemented During June 1991 Through May 1992

ML18153D110
Person / Time
Site:	Surry, North Anna
Issue date:	08/31/1992
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
92-560, NUDOCS 9209090153
Download: ML18153D110 (11)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 August 31, 1992 United States Nuclear Regulatory Commission Serial No.92-560 Attention: Document Control Desk NA&f/GLD-CGL R1 Washington, D. C. 20555 Docket Nos. 50-280 50-281 50-338 50-339 License Nos. DPR-32 DPR-37 NPf-4 NPf-7 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 NORTH ANNA POWER STATION UNITS 1 AND 2 REPORT OF ECCS EVALUATION MODEL CHANGES PURSUANT TO REQUIREMENTS OF 10CFR50.46 Pursuant to 1OCfR50.46(a)(3)(ii), Virginia Electric and Power Company is providing information concerning changes to the ECCS evaluation models used in existing licensing analyses. Information is also provided concerning the effect of these changes upon the existing analyses for Surry and North Anna Power Stations.

In a July 31, 1992 letter, Westinghouse advised us that no generic changes to the ECCS evaluation models had been implemented for the current reporting period (June 1991 through May 1992).

Although no generic changes were implemented, there were plant specific changes associated with application of the evaluation models for Surry and North Anna.

Attachment 1 provides a report describing these plant specific evaluation model changes. Attachments 2 and 3 provide information regarding the effect of the ECCS evaluation model changes on the LOCA analysis results in the currently approved Surry and North Anna analyses, respectively. To summarize the information in Attachments 2 and 3, the calculated peak clad temperature (PCT) for the small and large break LOCA analyses for Surry and North Anna are given below. None of these results constitute a significant change from the previously reported PCTs.

Surry Units 1 and 2 - Small break: 1852°f Surry Units 1 and 2 - Large break: 2173°f North Anna Units 1 and 2 - Small break: 1873°f North Anna Unit 1 - Large break: 2141 °f North Anna Unit 2 - Large break: 2131 °f

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• e Since none of the calculated temperatures exceed 2200°F, no further action is required.

If you have questions or require additional information, please contact us.

Very truly yours, nr.cWcJ~

\;t. L. Stewart Senior Vice President - Nuclear Attachments:

1. Report of Changes in Application of ECCS Evaluation Models

2. Effect of ECCS Evaluation Model Modifications - Surry Units 1 and 2

3. Effect of ECCS Evaluation Model Modifications - North Anna Units 1 and 2 cc: U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W.

Suite 2900 Atlanta, Georgia 30323 Mr. M. W. Branch NRC Senior Resident Inspector Surry Power Station Mr. M. S. Lesser NRC Senior Resident Inspector North Anna Power Station

ATTACHMENT 1 REPORT OF CHANGES IN APPLICATION OF ECCS EVALUATION MODELS

1.0* Background This report provides a summary of the changes to the evaluation model and LOCA analysis results from those last reported for North Anna Units 1 and 2 and Surry Units 1 and 2 (References 1 and 2). These changes are described in Section 2.0 below. It has been concluded that none of these changes are significant, as defined in 10CFR50.46(a)(3)(i).

2.0 Evaluation Model Changes 2.1 Revised Small Break LOCA Analysis - NOTRUMP (Surry Units 1 and 2)

Our previous 10CFRS0.46 report (Reference 1) included results for a revised small break LOCA analysis employing the NOTRUMP evaluation model. This analysis, documented in Reference 3, has since received NRC approval and is reported as the analysis of record.

- 2.2 Evaluation of Low Head Safety Injection Measured Flow (Surry Units 1 and 2)

In March 1992, flow measurements taken on the Unit 1 low head safety injection pumps indicated that delivered flow during a large break LOCA may be less than that assumed in the existing analysis of record. Measurement data were evaluated and corrected for conditions present during the measurement process. It was concluded that reducing the previously assumed low head safety injection flow would appropriately bound the expected flow rate for both units. A similar test and data evaluation are scheduled to be performed during the next Unit 2 refueling outage.

The impact of this flow reduction upon the existing analysis of record (Reference 4)

. was quantified by employing a conservative estimate of the clad heatup rate associated with the limiting case large break analysis result. A PCT penalty of 23°F was assessed and is reported in Attachment 2.

2.3 15x15 Assembly Increased Grid Pressure Drop (Surry Units 1 and 2)

In Reference 5, Westinghouse advised us of results from assessing a potential safety issued affecting Surry Units 1 and 2. Hydraulic tests on the 15x15 Optimized Fuel Assembly (OFA) indicated that design values for the grid loss coefficients had been underestimated. This resulted in the core pressure drop assumed in the safety and design analyses being underestimated by approximately 10%.

The large break LOCA analysis of record (Reference 4) employs assumptions for Surry Improved Fuel (SIF), which contains the Zircaloy grid design from the OFA fuel.

The effect of the increased grid pressure drop results in a 26°F increase in peak clad temperature (PCT) for the limiting case large break analysis. This penalty is presented in Attachment 2. The effects of the grid issue were quantified by using the present 1992 BART model with appropriate input changes to reflect the increased grid pressure drop. The penalty presented in Attachment 2, therefore, contains effects of

the "BART evaluation model changes (from the Reference 4 analysis to the present model) and the increased grid pressure drop.

2.4 Revised Large Break LOCA Analysis - Increased SGTP (North Anna Unit 1)

Since our previous 10CFRS0.46 report (Reference 2), a reanalysis of the large break LOCA event has been performed for operation of North Anna Unit 1 with extended steam generator tube plugging (SGTP). This analysis (Reference 6), which assumes 35% uniform SGTP and 95% rated thermal power, has since received NRC approval and is reported as the analysis of record.

3.0 References

1. Letter from W. L. Stewart (Virginia Electric and Power Co.) to NRC, "Surry Power Station Units 1 and 2, North Anna Power Station Units 1 and 2 - Report of ECCS Evaluation Model Changes Per Requirements of 10CFRS0.46," Serial No.91-428, August 23, 1991.

2. Letter from W. L. Stewart (Virginia Electric and Power Co.) to NRC, "North Anna Power Station Units 1 and 2 - Report of Errors/Changes in Application of ECCS Evaluation Models Per 10CFRS0.46," Serial No.92-091, February 10, 1992.

3. Letter from W. L. Stewart (Virginia Electric and Power Co.) to NRC, "Surry Power Station Units 1 and 2 - Proposed Technical Specifications Change - Fi:\h Increase/Statistical DNBR Methodology," Serial No 91-374, July 8, 1991.

4. Letter from W. L. Stewart (Virginia Electric and Power Co.) to NRC, "Surry Power Station Units 1 and 2 - Proposed Technical Specifications Change - Surry Improved Fuel Assembly," Serial No.87-188, May 26, 1987.

5. Letter from H. A. Sepp (Westinghouse - Strategic Licensing Issues), "Nuclear Safety Advisory Letter - 15x15 OFA Core Pressure Drop Increase," NSAL-92-002E, July 29, 1992.

6. Letter from W. L. Stewart (Virginia Electric and Power Co.) to NRC, "North Anna Power Station Units 1 and 2 - Proposed Technical Specifications Change -

Reduction in Maximum Reactor Power Level Due to Increased Steam Generator Tube Plugging Level," Serial No.92-042, January 28, 1992.

ATTACHMENT 2 EFFECT OF ECCS EVALUATION MODEL MODIFICATIONS -

SURRY UNITS 1 AND 2 r

Effect of Westinghouse ECCS Evaluation Model Modifications - Surry The information provided herein is applicable to Surry Power Station Units 1 and 2. It is based upon reports from Westinghouse Electric Corporation for issues involving the ECCS evaluation models and plant specific application of the models in the existing analyses. Peak clad temperature (PCT) values and margin allocations represent issues for which permanent resolutions have been implemented. Section A presents the detailed assessment for small break LOCA. The large break LOCA details are given in Section B.

SECTION A - SMALL BREAK LOCA MARGIN UJIUZAJION - SURRY UNITS l AND 2 A. PCT for Analysis of Record (AOR) 1852 Of (Reference 1)

B. Prior Evaluation Model PCT Assessments Identified Issues Not Applicable C. Current Evaluation Model PCT Assessments Identified Issues Not Applicable SBLOCA Licensing Basis PCT 1852 Of (AOR PCT + PCT Assessments)

SECTION B - LARGE BREAK LOCA MARGIN UTILIZATION - SURRY UNITS 1 AND 2 A. PCT for Analysis of Record (AOR) 1969 Of (Reference 2)

B. Prior Evaluation Model PCT Assessments

1. 1991 Report (Reference 3) + 155 Of C. Current Evaluation Model PCT Assessments

1. Evaluation of LHSI Flow Measurement + 23 Of

2. 15x15 Grid Increased Pressure Drop * + 26 Of LBLOCA Licensing Basis PCT 2173 Of (AOR PCT+ PCT Assessments)

References on following page.

Effect of Westinghouse ECCS Evaluation Model Modifications - Surry References

1. Letter from W. L. Stewart (Virginia Electric and Power Co.) to NRC, "Surry Power Station Units 1 and 2 - Proposed Technical Specifications Changes - Ft\h Increase/Statistical DNBR Methodology," Serial No.91-374, July 8, 1991.

2. Letter from W. L. Stewart (Virginia Electric and Power Co.) to NRC, "Surry Power Station Units 1 and 2 - Proposed Technical Specifications Changes - Surry Improved Fuel Assembly," Serial No.87-188, May 26, 1987.

3. Letter from W. L. Stewart (Virginia Electric and Power Co.) to NRC, "Surry Power Station Units 1 and 2, North Anna Power Station Units 1 and 2 .- Report of ECCS Evaluation Model Changes Per Requirements of 10CFR50.46," Serial No.91-428, August 23, 1991.

ATTACHMENT 3 EFFECT OF ECCS EVALUATION MODEL MODIFICATIONS -

NORTH ANNA UNITS 1 AND 2

L Effect of Westinghouse ECCS Evaluation Model Modifications - North Anna The information provided herein is applicable to North Anna Power Station Units 1 and 2. It is based upon reports from Westinghouse Electric Corporation for issues involving the ECCS evaluation models and plant specific application of the models in the existing analyses. Peak cladding temperature (PCT) values and margin allocations represent issues for which permanentresolutions have been implemented.

Section A presents the detailed assessment for small break LOCA. The large break LOCA details are given in Section B.

SECTION A - SMALL BREAK LOCA MARGIN UTILIZATION - NORTH ANNA UNITS l AND 2 A. PCT for Analysis of Record (AOR) 1873 Of (Reference 1)

B. Prior Evaluation Model PCT Assessments Identified Issues Not Applicable C. Current Evaluation Model PCT Assessments Identified Issues Not Applicable SBLOCA Licensing Basis PCT 1873 Of (AOR PCT+ PCT Assessments)

SECTION B - LARGE BREAK LOCA MARGIN UTILIZATION - NORTH ANNA UNITS 1 AND 2 Unit 1 Unit 2 A. PCT for Analysis of Record (AOR) . 2141 Of 2116 Of (Reference 1 for Unit 1, Reference 2 for Unit 2)

B. Prior Evaluation Model PCT Assessments

1. . 1991 Report (Reference 4) + 0 Of {1} + 25 Of C. Current Evaluation Model PCT Assessments Identified Issue Not Applicable LBLOCA Licensing Basis PCT 2141 Of 2131 Of (AOR PCT+ PCT Assessments)

Notes and References are on the following page.

Effect of Westinghouse ECCS Evaluation Model Modifications - North Anna Notes

{1} The existing analysis was performed with an evaluation model version that included corrections and/or input changes to address the applicable issues.

References

1. "North Anna Power Station Units 1 and 2 - Implementation of Extended SGTP Small Break LOCA Analysis," 10CFRS0.59 Safety Evaluation 92-SE-OT-005, January 21, 1992.

2. Letter from W. L. Stewart (Virginia Electric and Power Co.) to NRC, "North Anna Power Station Unit 1 - Proposed Technical Specification Change - Reduction in Maximum Reactor Power Level Due to Increased Steam Generator Tube Plugging Level," Serial No.92-042, January 28, 1992.

3. "North Anna Power Station Unit 2 Reload Safety Evaluation - North Anna 2 Cycle 9, Pattern ET," 10CFR50.59 Safety Evaluation 92-SE-OT-042, April 13, 1992.

4. Letter from W. L. Stewart (Virginia Electric and Power Co.) to NRC, "Surry Power Station Units 1 and 2, North Anna Power Station Units 1 and 2 - Report of ECCS Evaluation Model Changes Per Requirements of 10CFRS0.46," Serial No.91-428, August 23, 1991.