ML18153A509

Provides Info Re Changes to ECCS Evaluation Models & Application in Existing Licensing Analyses,Per 10CFR50.46(a)(3)(ii)

ML18153A509
Person / Time
Site:	Surry, North Anna
Issue date:	03/27/1997
From:	Ohanlon J VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9704070301
Download: ML18153A509 (14)

Text

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e e VIRGINIA ELECTRIC AND PowER CoMPANY RICHMOND, VIRGINIA 23261 March 27, 1997 United States Nuclear Regulatory Commission Serial No.97-174 Attention: Document Control Desk NL&OS/GDM RO Washington, D.C. 20555 Docket Nos. 50-280/281 50-338/339 License Nos. DPR-32/37 NPF-4/7 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY AND NORTH ANNA POWER STATIONS UNITS 1 AND 2 REPORT OF EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODEL CHANGES PURSUANT TO THE REQUIREMENTS OF 10CFRS0.46 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 North Anna and Surry Power Stations, and demonstrates continued compliance with the acceptance criteria of 10CFR50.46. contains excerpted portions of Westinghouse reports describing the changes to the Westinghouse ECCS Evaluation Models which are applicable to North Anna and Surry and have been implemented during calendar year 1996.

Information regarding the effect of the ECCS Evaluation Model changes upon the reported LOCA analysis of record (AOR) results is provided for the North Anna and Surry Power Stations in Attachments 2 and 3, respectively. To summarize the information in 0

Attachments 2 and 3, the calculated peak cladding temperatures (PCT) for the small and large break LOCA analyses for North Anna and Surry are given below. These results do not include significant changes based on the criterion in 10CFR50.46(a}(3}(i}.

North Anna Units 1 and 2 - Small break: 1675° F North Anna Units 1 and 2 - Large break: 2068°F Surry Units 1 and 2 - Small break: 1717° F

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SurryUnits 1 and 2 - Large break: 2113°F We have evaluated these issues and the associated changes in the applicable licensing basis PCT results. These results demonstrate compliance with the requirements of

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• 1 OCFR50.46(b). No further action is required to demonstrate compliance with 10CFR50.46 requirements.

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

Very truly yours,

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Senior Vice President - Nuclear Attachments:

1) Westinghouse Report of ECCS Evaluation Model Changes - North Anna Units 1 and 2 and Surry Units 1 and 2

2) Effect of ECCS Evaluation Model Changes - North Anna Units 1 and 2

3) Effect of ECCS Evaluation Model Changes - Surry Units 1 and 2 Commitments contained in this letter: None 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 NRC Senior Resident Inspector North Anna Power Station

ATIACHMENT 1 WESTINGHOUSE REPORT OF ECCS EVALUATION MODEL CHANGES NORTH ANNA POWER STATION UNITS 1 AND 2 SURRY POWER STATION UNITS 1 AND 2

TRA!'lSLATION OF FLUID CONDITIONS-FROM SATAN TO LOCTA

Background

An error was discovered in the coding related to the translation of fluid conditions between the SATAN blowdown hydraulics code and the LOCTA code used for subchannel analysis of the fuel rods. In p~rfonning axial interpolatjons to translate the SATAN fluid conditions onto the mesh nodalization used by the LOCTA code, the length of the lower core channel fluid connection to the *lower plenum node was* incorrectly calculated.

• This change is considered to be a Non-Discretionary Change as described in WCAP-13451.

Affected Evaluation Models 1981 Westinghouse Large Break LOCA Evaluation Model 1981 Westinghouse Large Break LOCA Evaluation Model with BART 1981 Westinghouse Large Break LOCA Evaluation Model with BASH Estimated Effects Representative plant calculations with the corrected model demonstrated that this correction resulted in approximately a +/-15 °F effect on large break LOCA evaluation models. Therefore a 15 °F penalty will be assessed against existing analyses.

SMALL BREAK LOOP SEAL VOLUME ISSUE Backe:round

. An inconsistency between the approved methodology for calculation of variable area fluid node inputs in NOTRUMP and the raw data upon which it is based was discovered. This inconsistency led to certain fluid nodes being modelled as normal "T-nQdes" rather than the correct modelling as "inverted T-nodes". A. new, consistent methodology was develop.ed and will be implemented in all future analyses. ~

This change is considered to be a Non-Discretionary Change as described in WCAP-1345 l.

Affected Evaluation Model 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP

• Estimated Effect A representative NOTRUMP calculation has led to a ge.neric conservative assessment of 0°F for this issue.

LOCBART FUEL ROD INTERNAL PRESSURE MODEL Back2round

_An *error ~as discovered in the LOCBART code related to improper calculation of fuel rod crack and dish void volumes. This volume dis~bution is calculated in the LOCBART code*

for the purpose of calculating rod internal pressure. The fact that there is an extra *unpowered rod length below the fuel stack included in the LOCBART nodalization was not properly accounted for in these calculations. .

This change is considered to be a Non-Discretionary Change as described in WCAP-13451.

Affected Evaluation Model 1981 Westinghouse Large Break LOCA Evaluation Model with BASH Estimated Effects Calculations with the corrected model demonstrated that this correction had negligible effect on predicted results. Therefore no penalty or benefit need be assigned to existing analyses.

LOCBART CLAD CREEP AND BURST Back2round An error was discovered in the LOCBART code related to improper modeling of fuel rod*

cladding creep and burst. The high temperature creep model did not properly account for*

pellet/clad contact which can often occur during crcepdown in the first several seconds of a large break LOCA when there is still a compressive stress exerted on the clad. The incorrect clad *strain which was calculated during* this time period then contributed some residual cumulative* effect on the stress and strain _calculations of clad ballooning and burst during the later phase of the transient. For the burst model, logic previously used in the code only examined the highest temperature axial node at any given timestep to determine if th~

temp*erature and hoop stress met burst conditions. However. since for the Zirlo clad model there is an additional requirement that a prcburst strain of 10% must have occurred in order to burst. it is possible on occasion that a node which is not the ~ighest temperature one for a given timestep simultaneously' meets all of the burst criteria first. Due to the small magnitude of effects and the* interaction between these two items. they are being evaluated ~ a single, closely related effect.

This change is considered *to be a Non-Discretionary Change as described in WCAP-1345 l.

Affected Evaluation Models 1981 Westinghouse Large Break LOCA Evaluation Model 1981 Westinghouse Large Break LOCA Evaluation Model with BASH Estimated Effects Representative plant calculations with the corrected model demonstrated that this revision produced small reductions in calculated peak clad temperature over a limited range. Because the effect varies between clad material and specific plant transient conditions. the small benefit will be conservatively assessed as a O °F change for evaluating existing analyses of record.

FISSION GAS COMPosmoN IN NOTRUMP

Background

Previously. the accepted methodology for treaonent of the fission gas component of the fuel rod gap heat transfer coefficient in NOTRUMP assumed that the fission gas was composed

-. .entirely of Xenon. This was inconsistent with the assumed *composition in the rod* heat up code (SBLOCTA) where the fission gas is assumed to be composed of 85% Xenon and 15'k Krypton. The NOTRUMP methodology was chang~d to make it c.onsistent with the assumption in SBLOCTA.

This change is considered to be a Discretionary Cbange as described in Section 4.1.1 of WCAP-1345_1. .

Affected Evaluation Model 1985 SBLOCA Evaluation Model with NOTRUMP Estimated Effect The nature of this change has led to an estimate that the effect on analyses is negligible. Its Discretionary nature therefore allows this change to be implemented on a forward fit basis with no impact assessed on existing analyses.

• ATTACHMENT 2 EFFECT OF ECCS EVALUATION MODEL CHANGES NORTH ANNA UNITS 1 AND 2

Attachment 2 Effect of ECCS Evaluation Model Changes - 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 permanent resolutions have been implemented. The assessments for small break and large break LOCA are presented in Sections A and 8, respectively.

Section A - Small Break LOCA Margin Utilization - North Anna Units 1 and 2 A. PCT for Analysis of Record 1704°F (1)

8. Prior PCT Assessments Allocated to AOR -29°F

1. NOTRUMP Specific Enthalpy Error +20°F (2)

2. SALIBRARY Double Precision Errors -15°F (2)

3. Fuel Rod Initialization Error +1D°F (3)

4. Loop Seal Elevation Error -44°F (3)

SBLOCA Augmented PCT for AOR 1675°F C. PCT Assessments for 10CFR50.46(a}(3)(i) Accumulation {1} 0°F SBLOCA Licensing Basis PCT (AOR PCT+ PCT Assessments) 1675°F Section 8 - Large Break LOCA Margin Utilization - North Anna Units 1 and 2 A. PCT for Analysis of Record 2013°F (1)

8. Prior PCT Assessments Allocated to AOR 40°F

1. LBLOCA/Seismic SG Tube Collapse +30°F (1)

2. BASH Accumulator Empty Flag +10°F (1)

LBLOCA Augmented PCT for AOR 2053°F C. PCT Assessments for 10CFR50.46(a}(3)(i) Accumulation {1} 15°F

1. Translation of Fluid Conditions from SATAN to LOCTA {2} {3} +15°F LBLOCA Licensing Basis PCT (AOR PCT+ PCT Assessments) 2068°F Notes { } and References ( ) on the following page

Effect of ECCS Evaluation Model Changes - North Anna Notes:

{1} The accumulation of changes (sum of absolute magnitudes) is less than 50° F and is not significant as defined in 10CFR50.46(a)(3)(i).

{2} The current report is the initial quantification of effects for this issue.

{3} Refer to the Westinghouse Report of ECCS Evaluation Model Changes provided in Attachment 1. Westinghouse has determined that a +15°F adjustment is applicable to North Anna.

References:

(1) Letter from J. P. O'Hanlon (VEPCO) to Document Control Desk (USNRC), "Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, 30-Day Report of ECCS Evaluation Model Changes Per Requirements of 10CFR50.46,"

Serial No.95-608, November 29, 1995.

(2) Letter from J. P. O'Hanlon (Va. Electric & Power Co.) to USNRC, "Virginia Electric and Power Company, North Anna and Surry Power Station Units 1 and 2, Report of ECCS Evaluation Model Changes and 30-Day Report of ECCS Evaluation Model Changes Per Requirements of 10CFR50.46," Serial No.96-111, March 14, 1996.

(3) Letter from J. P. O'Hanlon (Va. Electric & Power Co.) to USNRC, "Virginia Electric and Power Company, North Anna Power and Surry Power Station Units 1 and 2, Report of ECCS Evaluation Model Changes and 30-Day Report of ECCS Evaluation Model Changes Per Requirements of 10CFR50.46" Serial No.96-390, August 1, 1996.

AITACHMENT 3 EFFECT OF ECCS EVALUATION MODEL CHANGES SURRY UNITS 1 AND 2

Attachment 3 Effect of Westinghouse ECCS Evaluation Model Changes ** 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 cladding temperature (PCT) values and margin allocations represent issues for which permanent resolutions have been implemented. The assessments for small break and large break LOCA are presented in Sections A and B, respectively.

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

B. 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 {1} 0°F SBLOCA Licensing Basis PCT (AOR PCT+ PCT Assessments) 1717°F Section B - Large Break LOCA Margin Utilization - Surry Units 1 and 2 A. PCT for Analysis of Record (AOR) 2120°F (2)

B. Prior PCT Assessments Allocated to AOR -16°F

1. ZIRLO' Cladding -16°F LBLOCA Augmented PCT for AOR 2104°F C. PCT Assessments for 10CFR50.46(a)(3)(i) Accumulation {1} 21°F

1. Vessel & SG Calculation Errors in LUCIFER -6°F (2)

2. LBLOCA Rod Internal Pressure Issues 0°F (2)

3. Translation of Fluid Conditions from SATAN to LOCTA {2} {3} +15°F LBLOCA Licensing Basis PCT (AOR PCT+ PCT Assessments) 2113°F Notes { } and References ( ) on the following page

. _J

Effect of Westinghouse ECCS Evaluation Model Changes - Surry Notes:

{1} The accumulation of changes (sum of absolute magnitudes) is less than 50°F and is not significant, as defined in 10CFR50.46(a)(3)(i).

{2} The current report is the initial quantification of effects for this issue.

{3} Refer to the Westinghouse Report of ECCS Evaluation Model Changes provided in Attachment 1. Westinghouse has determined that a +15° F adjustment is applicable to Surry.

References:

(1) Letter from J. P. O'Hanlon (Va. Electric & Power Co.) to USNRC, "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" Serial No.96-635, January 9, 1997.

(2) Letter from W. L. Stewart (VEPCO) to Document Control Desk (USNRC}, "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, 30-Day Report of ECCS Evaluation Model Changes Per Requirements of 10CFR50.46," Serial No.94-254, April 27, 1994.