ML18153A648

Provides 30-day Rept Re Changes to ECCS Evaluation Models & Application in Existing Licensing Analyses

ML18153A648
Person / Time
Site:	Surry
Issue date:	01/09/1997
From:	Ohanlon J VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
96-635, NUDOCS 9701140074
Download: ML18153A648 (9)

Text

e e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 January 9, 1997 United States Nuclear Regulatory Commission Serial No.96-635 Attention: Document Control Desk NL&OS/GDM RO Washington, D.C. 20555 Docket Nos. 50-280 50-281 License Nos. DPR-32 DPR-37 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 30-DAY REPORT OF ECCS EVALUATION MODEL CHANGES PURSUANT TO REQUIREMENTS OF 10CFR50.46 Pursuant to 10CFR50.46(a)(3)(ii), Virginia Electric and Power Company is providing information regarding changes to the ECCS Evaluation Models and their application in existing licensing analyses. Information is also provided which quantifies the effect of these changes upon reported results for Surry Power Station, and demonstrates continued compliance with the acceptance criteria of 10CFR50.46. In our March 14, 1996 letter (Serial No.96-111 ), we committed to perform a reanalysis of the small break loss of coolant accident (LOCA). The results of the reanalysis are included in this report as the revised analysis of record.

Attachment 1 provides a report describing plant-specific evaluation model changes associated with the application of the small break LOCA evaluation models for the Surry Units. Information regarding the effect of the ECCS Evaluation Model changes upon the reported LOCA analysis of record results is provided in Attachment 2. To summarize the information in Attachment 2, the calculated PCT for the small break LOCA analysis is 1717° F. This result, although a reduction, represents a significant change, as defined in 10CFR50.46(a)(3)(i).

We have evaluated these issues and the associated changes in the applicable licensing basis PCT results. These results demonstrate compliance with the requirements of 10CFR50.46(b). Although the Surry small break LOCA changes described in Attachment 1 are significant, the licensing basis PCT results provide increased margin to

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the limit. No further action is required to demonstrate compliance with 10CFRS0.46 requirements.

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

Very truly yours, James P. O'Hanlon Senior Vice President - Nuclear Attachments:

1) Report of Changes in Application of ECCS Evaluation Models - Surry Units 1 and 2

2) Effect of ECCS Evaluation Model Changes - Surry Units 1 and 2 cc: Regional Administrator U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W.

Suite 2900 Atlanta, Georgia 30323 Mr. R. A. Musser NRC Senior Resident Inspector Surry Power Station

e ATTACHMENT 1 REPORT OF CHANGES IN APPLICATION OF ECCS EVALUATION MODELS SURRY UNITS 1 AND 2

Revised Small Break LOCA Analysis Surry Units 1 and 2 1.0 Background This report provides a summary of changes in small break LOCA analysis results from those last reported for Surry Units 1 and 2 (1 ). These changes are described in Section 2.0 below. It has been concluded that these changes are significant as defined in 10 CFR 50.46(a)(3)(i).

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

Since our previous 10CFR50.46 report (1 ), a revised analysis of the small break LOCA transient has been performed for Surry Units 1 and 2. This revised analysis has been implemented as the analysis of record for both Units via a station 10CFR50.59 evaluation (2), consistent with the provisions of Surry Technical Specification 6.2.C, relating to the Core Operating Limits Report (COLR).

The key analysis input changes required to provide acceptable results for the current small break LOCA analysis are listed below and discussed further in the following paragraphs (2):

- Assumption of 15% uniform steam generator tube plugging Peak Heat Flux Hot Channel Factor, F(Q}, of 2.50 (changed from 2.32)

Peak value for Enthalpy Hot Channel Factor, F~h of 1.70 (changed from 1.65)

- A minimum delivered HHSI flow rate calculated for LOCA analysis

- A full core of Surry Improved Fuel (SIF) with ZIRLO' cladding and PERFORMANCE+ features (bounds operation with 15x15 Standard and SIF mixed cores)

Safety Injection in all loops (from safety injection occurring in intact loop only)

COSI Condensation Model (from homogeneous equilibrium condensation model)

This analysis was performed using the 1985 Small Break LOCA Evaluation Model with NOTRUMP (3). Technical Specification 6.2.C lists this as an acceptable reference methodology for determination of relevant power distribution limits in the Core Operating Limits Report.

e In the previous NOTRUMP evaluation model, safety injection is delivered only in the intact loop, and the least resistant safety injection line is assumed to spill on the containment floor. This modeling was assumed to be conservative since the additional safety injection was considered to be a benefit. This assumption was based on older models which employed a homogenous equilibrium assumption for the mixing of different phases.

Sensitivity studies with the previous evaluation model determined that safety injection in the broken loop, in conjunction with the existing condensation model, resulted in a PCT penalty. Reference (4) has documented this change to the NOTRUMP evaluation model.

This modeling is used in this reanalysis.

To offset the penalties associated with the revised safety injection assumption, Westinghouse has incorporated a new condensation model in the NOTRUMP evaluation model. This model, referred to as the COSI model, is based on tests which modeled the configuration of the SI piping to the RCS cold leg. Use of this more realistic model for condensation of steam by pumped SI is demonstrated to provide a benefit larger than any penalty associated with injecting into the broken loop (4).

The analysis assumed a peak Heat Flux Hot Channel Factor, FQ(z), value of 2.50 and a peak Nuclear Enthalpy Hot Channel Factor, F~h, value of 1.70. As required by Technical Specification 6.2.C, the Core Operating Limits Report (COLR) documents the applicable limit values of key core-related parameters for each reload core. These analytical values bound the limits in the current cycle specific COLR's.

A new revised hot rod axial power shape was used in the analysis. This power shape was chosen from a generic database of potential shapes achievable during power operation by assessing the characteristics which yield limiting small break LOCA results.

The selected shape has been identified as the most limiting within the bounds of the proposed K(z) curve.

The flow rates for the HHSI are provided by an engineering model of the HHSI subsystem that is based on the system configuration and measured data from the plant. This model includes allowances for imbalance between the separate injection lines, HHSI pump degradation, and instrument accuracy. The HHSI pump curves used in the model are based on the actual measured plant data for the installed HHSI pumps in each unit. For the calculated HHSI flows, it is assumed that the HHSI flow recirculation line is open above RCS pressures of 1000 psig and that it is closed below that RCS pressure. This is consistent with previous assumptions used to calculate HHSI flow rates versus RCS pressure for small break LOCAs. Other assumptions regarding HHSI system configuration, such as water levels and back pressures, are set to provide limiting conditions for the specified test condition.

The analysis assumes a full core of Surry Improved Fuel (SIF) with ZIRLO' cladding and PERFORMANCE+ design features. This modeling is applicable to full or mixed cores of either SIF or LOPAR fuel product. The only mechanism available to cause a transition core to have a greater calculated small break LOCA PCT than a full core of either fuel

product is the possibility of flow redistribution due to fuel assembly hydraulic resistance mismatch. The small break evaluation model assumes only one core channel. This assumption is acceptable, since the flowrate during a small break LOCA is low, providing enough time to maintain flow equilibrium and eliminate crossflow effects. As stated in Reference (5), mixed core hydraulic resistance mismatches are not a significant factor for small break analyses and it is not necessary to apply a LOCA analysis transition core penalty.

For this analysis of the small break LOCA, cases were run assuming 2 inch, 3 inch, and 4 inch effective diameter cold leg breaks. The 6 inch break case was not run, but previous evaluations have demonstrated its peak clad temperature (PCT) to be less limiting than the 2, 3, and 4 inch breaks. The 6 inch break produces a more rapid depressurization and accumulator actuation, which results in primary core recovery sooner than for the smaller break cases. The PCT occurs during this initial, deep uncovery. Analyses of 6 inch break.cases also typically exhibit a second, more shallow core uncovery, but fuel rod heatup is limited during this period by three factors: 1) greater accumulator and safety injection flows limit the uncovery to the top portion (approximately 2 feet) of the core, 2)

• the larger break size allows more energy removal from the core and 3) the duration of the second uncovery is ultimately limited by significant additional flow from the low head safety injection pumps, which provide for full core recovery. The 3 inch cold leg break was found to be the most limiting break size for the small break LOCA.

The revised analysis of record PCT is 1717° F. Since this result is more than 50° F different from the existing analysis of record, implementation of this analysis represents a significant change, as defined in 10CFR50.46(a)(3)(i). The resulting licensing basis PCT demonstrates that operation at the rated thermal power of 2546 MWt will comply with all of the acceptance criteria specified in 10CFR50.46. Attachment 2 provides the PCT result for the revised analysis of record.

3.0 References (1) Letter from J. P. O'Hanlon (Va. Electric & Power Co.) to USN RC, "Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Surry Power Station Units 1 and 2, Report of ECCS Evaluation Model Changes Pursuant to Requirements of 10CFR50.46" Serial No.96-390, August 1, 1996.

(2) Surry Power Station 10CFR50.59 Safety Evaluation,96-161, "Surry Power Station Units 1 and 2 - Safety Evaluation for Revised Small Break Loss of Coolant Analysis (SBLOCA)", December 16, 1996.

(3) WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using The NOTRUMP Code," August 1985.

(4) WCAP-10054, Addendum 2, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model," August 1994.

(5) WCAP-10444, Addendum 2, "Vantage 5H Fuel Assembly," April 1988.

ATTACHMENT 2 EFFECT OF ECCS EVALUATION MODEL CHANGES SURRY UNITS 1 AND 2

Effect of ECCS Evaluation Model Changes Surry Units 1 and 2 e

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 cladding temperature (PCT) values and margin allocations represent issues for which permanent resolutions have been implemented.

Section A - Small Break LOCA Margin Utilization - Surry Units 1 and 2 A. PCT for Analysis of Record (AOR) {1} 1717°F (1)

8. Prior PCT Assessments Allocated to AOR 0°F SBLOCA Augmented PCT for AOR 1717°F C. PCT Assessments for 10CFR50.46(a)(3)(i) Accumulation 0°F SBLOCA Licensing Basis PCT (AOR PCT+ PCT Assessments) 1717°F Notes:

{1} The previous SBLOCA AOR reported in the previous 30-day report (Reference 2) was 1852°F. The associated Licensing Basis PCT reported was 1813°F.

References:

(1) Surry Power Station 10CFRS0.59 Safety Evaluation,96-161, "Surry Power Station Units 1 and 2 - Safety Evaluation for Revised Small Break Loss of Coolant Analysis (SBLOCA)", December 16, 1996.

(2) Letter from J. P. O'Hanlon (VEPCO) to Document Control Desk (USNRC), "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 Per Requirements of 10CFRS0.46," Serial No.96-390, August 1, 1996.