ML18153A935
| ML18153A935 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 04/27/1994 |
| From: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 94-254, NUDOCS 9405050369 | |
| Download: ML18153A935 (23) | |
Text
e e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 April 27, 1994 United States Nuclear Regulatory Commission Serial No.94-254 Attention: Document Control Desk NA&F/GLD-CGL RO' Washington, D.C. 20555 Docket Nos. 50-280 50-281 License Nos. DPR-32 DPR-37 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY REPORT OF ECCS EVALUATION MODEL CHANGES PURSUANT TO REQUIREMENTS OF 10CFRS0.46 SURRY POWER STATION UNITS 1 AND 2 Pursuant to 10CFR50.46(a)(3)(ii), Virginia Electric and Power Company is providing information concerning 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 addition, we advised you by an October 20, 1993 letter (Serial No.93-642) that reanalysis of the large break LOCA was planned to be completed by the end of the first quarter of 1994.
The results of this reanalysis are included in this report.
Attachment 1 contains excerpted portions of the Westinghouse report describing the changes to the Westinghouse ECCS Evaluation Models which are applicable to Surry and have been implemented during calendar year 1993. In addition to these generic changes, there were plant-specific changes associated with application of the large break LOCA evaluation model for the Surry units. Attachment 2 provides a report describing these plant-specific evaluation model changes.
Attachment 3 provides information regarding the effect of the ECCS Evaluation Model changes upon the reported LOCA results for the Surry Power Station analysis of record. To summarize the information in Attachment 3, the calculated PCT for the small and large break LOCA analyses for Surry are given below. None of these results include significant changes, as defined in 10CFR50.46(~
Surry Units 1 and 2 - Small break:..,,-1*83.8°F Surry Units 1 and 2 - Large break: 2114°F
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- e 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). No further action is required to demonstrate compliance with 10CFRS0.46 requirements.
If you have questions or require additional information, please contact us.
Very truly yours, Cv* W. L. Stewart
,""' Senior Vice President - Nuclear Attachments:
- 1. Westinghouse Report of ECCS Evaluation Model Changes for 1993 - Surry Units 1 and 2
- 2. Report of Changes in Application of ECCS Evaluation Models - Surry Units 1 and 2
- 3. Effect of ECCS Evaluation Model Changes - Surry 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
e ATTACHMENT 1 WESTINGHOUSE REPORT OF ECCS EVALUATION MODEL CHANGES FOR 1993 SURRY UNITS 1 AND 2
VESSEL AND STEAM GENERATOR CALCULATION ERRORS IN LUCIFER
Background
The LUCIFER code is used to generate the component databases, from raw input data, to be used in the small and large break LOCA analyses. Errors were found in the VESCAL subroutine of the LUCIFER code. These errors were in the geometric and mass calculations of the vessel and steam generator portions of the needed data. All LOCA analyses using the LUCIFER code outputs are affected by these error corrections. The errors were corrected in a manner to maintain the consistency of the LUCIFER code.
The errors were determined to be a Non-Discretionary Change as described in Section 4.1.2 of WCAP-13451 and. were corrected in accordance with Section 4.1.3 of WCAP-13451.
Affected Evaluation Models 1985 SBLOCA Evaluation Model 1981 ECCS Evaluation Model 1981 ECCS Evaluation Model with BART 1981 ECCS Evaluation Model with BASH Estimated Effect
~epresentative plant calculations indicate a net PCT effect of -16°F for small break LOCA and a -6°F for large break LOCA.
1-1
ISIDI DRIFT FLUX ERROR
Background
An error was discovered both in WCAP-10079-P-A and the relevant coding in NOTRUMP SUBROUTINE ISHIIA which led to an incorrect calculation of the drift flux in NOTRUMP when a laminar film annular flow was predicted. The affected equation in WCAP-10079-P-A is Equation G-74 wherein a factor of 'g', the gravitational constant, was inadvertently omitted from both the documentation and the equivalent coding. The correction of this error returned NOTRUMP to consistency with the ultimate reference for the affected correlation.
This was determined to be a Non-discretionary Change as described in Section 4.1.2 of WCAP-13451 and was corrected in accordance with Section 4.1.3 of WCAP-13451. .
Affected Evaluation Models 1985 Small Break LOCA Evaluation Model Estimated Effect
. Representative plant analyses were used to estimate a generic PCT effect of 0°F.
1-2
NOTRUMP POINT KINETICS ERROR Backwund An error was discovered in the coding used in the NOTRUMP User External SUBROUTINE VOLHEAT. The coding did not correctly perform the calculation described by Equation 3-12-28 of WCAP-10054-P-A. This calculation is only used ~uring the time when the Point Kinetics option is used to determine the core power before reactor trip. Therefore, any analysis which used the more conservative assumption of constant core power until *reactor trip time is not affected by this error.
The correction of this erro~ returned NOTRUMP to consistency ~ith WCAP-10054-P-A.
This was determined to be a Non-discretionary Change as described in Section 4.1.2 of WCAP-13451 and was corrected in accordance with Section 4.1.3 of WCAP-13451.
Affected Evaluation Models 1985 Small Break LOCA Evaluation Model Estimated Effect Representative plant analyses were used to estimate a generic PCT effect of 0°F.
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e NOTRUMP DRIFT FLUX FLOW REGIME MAP ERRORS
Background
Errors were discovered in both WCAP-10079-P-A and related coding in NOTRUMP SUBROUTINE DFCORRS where the improved TRAC-Pl vertical flow regime map is evaluated. In Evaluation Model applications, this model is only used during counter-current flow conditions in vertical flow links. The affected equation in WCAP-10079-P-A is Equation G-65 which previously allowed for unbounded values of the parameter C.,. -contrary to the intent of the original source of this equation.
This allowed a discontinuity to exist in the flow regime map under some circumstances. This was corrected by placing an upper limit of 1.3926 on the parameter C... as reasoned from the discussion in the original source. As stated, this correction returned NOTRUMP to consistency with the original source for the affected equation.
Further investigation of the DFCORRS uncovered an additional closely related logic error which led to discontinuities under certain other circumstances. This error was also corrected and returned the coding to consistency with WCAP-10079-P-A.
This was determined to be a Non-discretionary Change as described in Section 4.1.2 of WCAP-13451 and was corrected in accordance with Section 4.1.3 of WCAP-13451.
Affected Evaluation Models 1985 Small Break LOCA Evaluation Model Estimated Effect Representative plant calculations indicated PCT effects ranging from -13°F to -55°F. For the purposes of tracking PCT, an estimated effect of-13°F will be assigned to this change.
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CORE NODE INITIALIZATION ERROR Backm,und An error was discovered in. how the properties of CORE NODE components were initialized for non-existent regions in the adjoining FLUID NODE. In particularly this led to artificially high core temperatures during the timestep when the core mixture level crossed a node boundary, conservatively causing slightly more core mixture level depression than appropriate during this timestep. Correction of this error allows for a smoother mixture level uncovery transient during node crossings.
This was determined to be a Non~iscretionary Change as described in Section 4.1.2 of WCAP-13451 and was corrected in accordance with Section 4.1.3 of WCAP-13451.
Affected Evaluation Models 1985 Small Break LOCA Evaluation Model Estimated Effect The nature of this error led to an estimated generic PCT effect of 0°F.
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e NOTRUMP BEAT LINK POINTER ERROR
Background
An error was discovered in how NOTRUMP initialized certain HEAT LINK pointer variables at the start of a calculation. Correction of this error returned NOTRUMP to consistency with the original intent of this section of coding.
This was determined to be a Non-discretionary Change as described in Section 4.1.2 of WCAP-13451
.and was corrected in accordance with Section 4.1.3 ofWCAP-13451.
Affected Evaluation Models 1985 Small Break LOCA Evaluation Model Estimated Effect Representative plant analyses were used to estimate a generic PCT effect of 0°F.
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e FUEL ROD MODEL .
ERRORS IN .
Background
A number of minor programming errors were corrected in the fuel rod heat up code used in SBLOCA analyses. These corrections were related to: . ,
- 1. Individual rod plenum temperatures .
- 2. Individual rod stack lengths
- 3. Oad thinning logic *
- 4. Pellet/clad contact logic
- 5. Corrected* gamma redistribution
- 6. Including Zr02 thickness at t=O initialization
- 7. Numerics and convergence criteria of initialization.
These changes were determined to be Non-Discretionary Changes in accordance with Section 4.1.2 of WCAP-13451 and were implemented in accordance with Section 4.1.3 of WCAP-13451 .
.Affected Evaluation Models 1975 SBLOCA Evaluation Model 1985 SBLOCA Evaluation Model Estimated Effect The cumulative effect *of the error corrections and convergence criteria change was found to be less than approximately +4°F. This change is therefore judged to have a negligible effect on PCT and on a generic basis the estimated effect will be reported as 0°F.
1-7
LARGE BREAK LOCA FUEL ROD MODEL ERRORS.
Background
M'mor err~rs in ili~ m
\*od heat up ~e ~ed Large Break LOCA analyses were corrected. These errors concerned conditions which exist during periods of pellet/clad contact and the internal book-keeping logic associated with clad thbmlng. .. . . . . . . . - .- -
These changes were-determined to be Non-DiscretionaryCllanges in accordance with Section 4.1.2 of WCAP-13451 .
.and were implemented in -accordance
~
with Section . -*
4J.3 of WCAP-13451.
Affected Evaluation Modeis , . .
1981 ECCS Evalua.ti~n Model*with BASH Estimated Effect ,
Representative plant calculations have shown that these corrections have a negligible effect on PCT for near Beginning-of-Life (BOL) fuel rod conditions (i.e. < 2000 MWD/MTU). These effects .
become prevalent as burnup increases, but are not expected to be of any significance until pelle~clad contact is predicted for steady-state operating conditions (typically > 8000 MWD/MTU). These
-corrections therefore result in a negligible PCT impact for Large Break LOCA licensing basis PCT' s which are calculated with near BOL conditions. This impact is being reported generically as 0°F.
1-8
mGH TEMPERATURE FUEL ROD BURST MODEL Background .
A model for calculating the prediction of zircaloy cladding burst behavior above the previous limit of 1742°F was implemented.. This :model was-described
. ~ . . - to the NRC in:
Letter ET-NRC-92-3746, N. J. Liparulo to R. C. Jones (NRC), "Extension of NUREG-0630 Fuel Rod Burst Strain and Assembly Blockage Models to High Fuel Rod Burst Temperatures", September 16, 1992.
- This was determined to be a Non-discretionary Change as described in Section 4.1.2 of WCAP-13451 and was corrected in accordance with Section 4.L3*of WCAP-13451.
Affected Evaluation Models 1981 ECCS Evaluation Model with BASH Estimated Effect
. The effect of the extended burst model has been directly incorporated in the Analysis of Record for those plants who are affected.
1-9
HOT ASSEMBLY AVERAGE ROD BURST EFFECTS
Background
The rod heat up ~de'rised in.Smail Br~ LOCA calculatio~* contains a model to calculate the amount of clad strain that accompanies rod burst. However, the methodology which has historically been used is to not apply this burst strain model to the hot assembly average rod. This was done so as to minimize the rod gap and therefore maximize the heat transferred to the fluid channel, which in tum would maximize the hot rod temperature. However, due to mechanisms governing the zirc-water temperature excursion (which is the subject of the SBLOCA Limiting Time-in-Life penalty for the hot rod), modeling of clad burst strain for the hot assembly average .rod can result in a penalty for the hot rod by increasing the channel enthalpy at the time of PCT. Therefore, the methodology has been revised such that burst strain will also be modeled on the hot assembly average rod.
This was determined to be a Non-discretionary Change as described in Section 4.1.2 of WCAP-13451 and was corrected in accordance with Section 4.1.3 of WCAP-13451.
Affected Evaluation Models 1975 SBLOCA Evaluation Model 1985 SBLOCA Evaluation Model Estimated Effect Representative plant calculations have*shown that this change introduces an approximately 10%
increase in the SBLOCA Limiting Time-in-Life penalty on the hot rod. However, this penalty is being offset in affected plants PCT Summary Sheets by the Revised Burst Strain Limit Model described on the following page. These models will be implemented concurrently in the Small Break Evaluation Model rod heat-up code in 1994.
- 1-10
e REVISED BURST STRAIN LIMIT MODEL Background _
A revised burst strain limit model which limits strains is being implemented into the rod heat up codes used in both Large Break and Small Break LOCA. This model, which is identical to that previously approved for use for Appendix K analyses of Upper Plenum Injection plants with WCOBRAfl'RAC, as described in WCAP-10924-P-A, Rev. 1, Vol. 1, Add. 4, "Westinghouse Large Break LOCA Best Estimate Methodology: *Volume l: Model Description and Validation, Addendum 4: Model Revisions," 199L
- This has been determined to be a Non-Discretionary Change as discussed in Section 4.1.2 of WCAP-13451 and is being implemented in accordance with Section 4.1.3 of WCAP-13451.
Affected Evaluation Models _
- 1981 ECCS Evaluation Model
- 1981 ECCS Evaluation Model with BART 1981 ECCS Evaluation Model with BASH Estimated Effect The estimated effect on Large Break LOCA PCT's ranges from negligible to a moderate, unquantified benefit which will be inherent in calculations once this model is implemented. In Small Break LOCA, representative-plant calculations indicate that the magnitude of the benefit is conservatively estimated to be exactly offsetting to the penalty introduced by the Hot Assembly Average Rod Burst issue documented on the previous page. This model will be implemented in both Large Break and Small Break Evaluation Models during 1994.
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e LARGE BREAK LOCA ROD INTERNAL PRESSURE ISSUES Issue Description Westinghouse recently completed an evaluation of a potential issue concerning the impact of increased beginning of life rod internal pressure (RIP) uncertainties on LOCA analyses. Historically, beginning of life fuel pressure and temperature uncertainties, were based upon end of life considerations. These RIP uncertainties were found to be potentially nonconservative. During the evaluation of this issue, a second issue related to the applicability of generic IFBA fuel analyses to updated LOCA Evaluation Models was also identified and combined with this issue since the underlying mechanisms were the same.
Technical Evaluation .
The technical evaluation of this issue concluded that both the RIP uncertainty and the current IFBA designs with 200 psig initial fill pressure fuel typically will result in a maximum +/- 15°F PCT variation. Consequently, RIP manufacturing uncertainties and 200 psig initial fill pressure IFBA fuel do not have significant effects on the large break LOCA analyses. Also, based on these results, it was concluded that only nominal RIP (with an upper bound J)ias) should be used in the LOCA .
analyses for fuel designs with an initial cold fill pressure ~ 200 psig. This is consistent with past
. LOCA analysis.
Specific analyses were performed for all plants with initial fill pressure < 200 psig. It was demonstrated that the acceptance criteria of 10 CFR 50.46 continued to be met for each of these plants.
Assessment of Safety Significance A 10 CFR, Part 21 evaluation concluded that the effects of low initial fill pressure and increased RIP uncertainty will not*represent a defect creating a substantial safety hazard and, more likely than not, will not result in a failure to comply with any applicable regulation relating to a substantial safety hazard. This conclusion was based upon the implementation of an extended burst and blockage correlation for burst temperatures above 1742°F and a more realistic minimum bumup assumption at hot full power conditions. In addition, any new reloads which would utilize low ( < 200 psig) initial fill pressure fuel would be specifically analyzed.
Recommended Actions Resolution of this issue may have resulted in a plant specific PCT change and would be shown on the attached annual 50.46 report PCT _Margin Utilization Sheets. Aside from determining reporting requirements relative to 10 CFR 50.46, no other utility action is necessary.
1-12
e e
-SMALL BREAK LOCA LIMITING TIME IN LIFE - ZIRC/WATER OXIDATION TEMPERATURE EXCURSION .
Issue Description
. Westinghouse recently completed an evaluation of a potential issue with regard to burst/blockage modeling in the Westinghouse small break LOCA evaluation model.* This potential issue involved a number of synergistic effects, all related to the manner in which the small break model accounts for the swelling and burst of fuel rods, modeling of the rod burst strain, and resulting effects on clad temperature and *oxidation from the metal/water reaction models and channel blockage.
Technical Evaluation Fuel rod burst during the course of a small break LOCA analysis was found to potentially result in a significant temperature excursion above the clad temperature transient for a non-burst case. Since the methodology for SBLOCA analyses had been to perform the analyses at a near beginning of life (BOL) condition, where rod internal pressures are relatively low, most analyses did not result in the occurrence of rod burst, and therefore may not have reflected the most limiting time in life PCT. In order to evaluate the effects of this phenomenon, Westinghouse has developed an analytical model which allows the prediction of rod burst PCT effects based upon the existing analysis of record.
Assessment of Safen, Significance A 10 CFR Part 21 evaluation concluded that the effects of the burst/blockage modeling in the Westinghouse small break LOCA evaluation model will not represent a defect creating a substantial safety hazard and, more likely than not, will not result in a failure to comply with any applicable regulation relating to a substantial safety hazard.
Recommended Actions Resolution of this issue may have resulted in a plant specific PCT change. Since evaluation of the issue was in progress in 1992, some PCT effect may have previously been reported as a temporary impact. The evaluation for this issue has been finalized, and remaining PCT effects are now considered a permanent change with respect to evaluating 1993 reporting requirements. Aside from determining reporting requirements relative to 10 CFR 50.46, no other utility action is necessary.
1-13
ATTACHMENT 2 REPORT OF CHANGES IN APPLICATION OF ECCS EVALUATION MODELS SURRY UNITS 1 AND 2
e Revised Large Break LOCA Analysis for Reduced LHSI Flowrate and Uprated Power
1.0 Background
This report provides a summary of changes in LOCA analysis results from those last reported for Surry Units 1 and 2 in November 1993 (1 ). These changes are described in Section 2.0 below. It has been concluded that these changes are not significant, as defined in 10CFR50.46(a)(3)(i).
2.0 Evaluation Model Changes 2.1 Revised Large Break LOCA Analysis (Surry Units 1 and 2)
Since our previous 10CFR50.46 report (1 ), a revised analysis of the large break LOCA transient has been performed for Surry Units 1 and 2. This revised analysis has been implemented as the analysis of record (AOR) by means of a station 10CFR50.59 evaluation (2) consistent with the provisions of Surry Technical Specification 6.2.C (relating to the Core Operating Limits Report). This discussion summarizes the changes incorporated in the analysis. Analysis assumptions conservatively included increased steam generator tube plugging (SGTP) and uprated core power. The key changes in assumptions from the prior analysis are:
- Analysis performed with the 1981 Evaluation Model with BASH (3),
- Uprated core thermal power of 2546 MWt, *
- Assumption of 15% uniform SGTP (supports operation with a peak SGTP of 15% in any steam generator),
Improved spacer grid heat transfer model (4),
- Hot assembly relative power factor of 1.465,
- Containment accumulator water temperature of 105°F,
- Safety Injection, 1 HHSI + 1 LHSI, spilling to 1O psig containment pressure, and
- Assumed fuel temperature and rod internal pressure associated with core average burnup of 500 MWD/MTU and Surry Improved Fuel (SIF) with the PERFORMANCE+ design features.
These changes are discussed further in the following paragraphs.
2-1
This analysis was performed using the Westinghouse 1981 large break LOCA evaluation model with BASH (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. The present analysis is the first application of the BASH evaluation model for Surry Power Station. The previous AOR was performed using the BART evaluation model.
The analysis assumes an uprated core power of 2546 MWt, with the associated primary and secondary system parameters. This additional conservatism is included in anticipation of a future license amendment request for operation at uprated power (from 2441 MWt).
The analysis conservatively assumes that 15% of the tubes in each steam generator is plugged. Since large break LOCA results are sensitive to SGTP, this assumption is used to demonstrate and ensure continued compliance with the 10CFR50.46 ECCS acceptance criteria for potential operation with increased SGTP. The previous AOR assumed 7% SGTP.
The assumed initial accumulator water temperature is 105°F. The water temperature in each accumulator is assumed to equal the temperature in the surrounding containment compartment during full power operation. At Surry Power Station, the accumulators are located on the containment floor. A review of temperature detector data at this location has confirmed that a temperature of 105°F is a representative conservative nominal value for use in large break LOCA analysis. This assumption has been changed from 90°F and validated in accordance with Westinghouse guidance which implements the recommendations of Reference (5).
The Surry analysis uses a corrected version of the LOCBART code, which is part of the BASH Evaluation Model. Westinghouse has corrected and improved the spacer grid heat transfer model used in the BART and BASH ECCS Evaluation Models (4). Since this model change is primarily a correction to the evaluation model, it has been implemented in all versions of the BART and BASH evaluation models without prior NRC review. This process for addressing model changes is documented in WCAP-13451 (6).
The safety injection (high head (HHSI) and low head (LHSI)) is assumed to spill to 1O psig containment pressure instead of the value of O psig used in the previous analysis.
This assumption has been validated by confirming that containment pressure (conservatively calculated using the Westinghouse evaluation .model) does not decrease to less than 1O psig during the transient prior to the time of peak clad temperature. The analysis also assumed degraded LHSI flow performance data which are conservatively less than previous Surry LHSI test results. The performance data used in this analysis was benchmarked to a LHSI flow of 2970 gpm for the three lines delivering to a RCS backpressure of O psig (14.7 psia) with a full refueling water storage tank (RWST).
2-2
e The analysis assumed a reference cosine axial power distribution with a peak Heat Flux Hot Channel Factor, FQ(z), value of 2.32. In addition, the analysis assumed a slightly greater hot assembly relative power factor of 1.465 (versus 1 .442 previously).
This assumption was required to bound the anticipated characteristics of the low-low leakage reload cores in future Surry fuel cycles.
Employing these assumptions in the current version of the 1981 ECCS Evaluation Model with BASH, it has been demonstrated that operation at an assumed core thermal power of 2546 MWt with SGTP up to 15% in any steam generator will comply with all of the acceptance criteria specified in 10CFR50.46. Attachment 3 provides the PCT result for the revised analysis of record, in conjunction with appropriate margin assessments which address BASH evaluation model issues.
3.0 References (1) Letter from W. L. Stewart (Va. Electric & Power Co.) to NRC, "Report of ECCS Evaluation Model Changes and 30-Day Report Per Requirements of 1 OCFR50.46, Surry Power Station Units 1 and 2, North Anna Power Station Units 1 and 2," Serial No. 93-1828, November 9, 1993.
(2) "Surry Power Station Units 1 and 2 - Safety Evaluation for Revised Large Break LOCA Analysis Break LOCA Analysis," 10CFR50.59 Safety Evaluation 94-082, March 28, 1994.
(3) WCAP-10266-P-A, Rev. 2, "The 1981 Version of the Westinghouse ECCS Evaluation Model using the BASH Code," March 1987.
(4) Letter from Nick Liparulo (Westinghouse-Manager, Nuclear Safety & Regulatory Activities) to USNRC, "Notification of Changes to the Westinghouse Large Break LOCA ECCS Evaluation Model," ET-NRC-92-3787, December 22, 1992, transmits WCAP-10484, Addendum 1, "Spacer Grid Heat Transfer Effects During Reflood."
(5) Letter from Nicholas J. Liparulo (Westinghouse-Manager, Nuclear Safety &
Regulatory Activities) to USNRC, "Results of Technical Evaluation of Containment Initial Temperature Assumptions for Large Break Loss of Coolant Accident Analysis," ET-NRC-92-3699, June 1, 1992.
(6) Letter from N. J. Liparulo (Westinghouse-Manager, Nuclear Safety & Regulatory Activities) to USNRC, "Westinghouse Methodology for Implementation of 10CFR50.46 Reporting," ET-NRC-92-3755, October 30, 1992, transmits WCAP-13451, "Westinghouse Reporting Methodology for Implementation of 10 CFR 50.46 Reporting."
2-3
ATTACHMENT 3 EFFECT OF ECCS EVALUATION MODEL CHANGES SURRY UNITS 1 AND 2
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 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 - Surry Units 1 and 2 A. PCT for Analysis of Record (AOR) 1852°F (1)
B. Prior PCT Assessments Allocated to AOR 0°F
- 1. Safety Injection in the Broken Loop 0°F (2)
- 2. NOTRUMP Drift Flux Flow Regime Map Errors {1} - 13°F (2)
SBLOCA Augmented PCT for AOR 1839°F C. PCT Assessments for 10CFR50.46(a)(3)(i) Accumulation
- 1. Vessel & SG Calculation Errors in LUCIFER {1} - 16°F
- 2. Hot Assembly Average Rod Burst Effects {1} 2°F
- 3. Revised Burst Strain Limit Model {1} - 2°F
- 4. SBLOCA Limiting Time in Life-Zirc/Water Oxidation {1} 15°F SBLOCA Licensing Basis PCT (AOR PCT+ PCT Assessments) 1838°F Section B - Large Break LOCA Margin Utilization - Surry Units 1 and 2 A PCT for Analysis of Record (AOR) 2120°F (3)
B. Prior PCT Assessments Allocated to AOR 0°F LBLOCA Augmented PCT for AOR 2120°F C. PCT Assessments for 10CFR50.46(a)(3)(i) Accumulation 0°F
- 1. Vessel & SG Calculation Errors *in LUCIFER {1} - 6°F
- 2. LBLOCA Rod Internal Pressure Issues {1} 0°F LBLOCA Licensing Basis PCT (AOR PCT+ PCT Assessments) 211 4 ° F Notes { } and References ( ) on the following page 3-1
Notes:
{1} Refer to the Report of Westinghouse ECCS Evaluation Model Changes for 1993 provided in Attachment 1.
References:
(1) Letter from W. L. Stewart (Va. Electric & Power Co.) to NRC, "Surry Power Station Units 1 and 2 - Proposed Technical Specifications Changes - F~H Increase/Statistical DNBR Methodology," Serial No.91-374, July 8, 1991.
(2) Letter from W. L. Stewart (Va. Electric & Power Co.) to NRC, "Report of ECCS Evaluation Model Changes and 30-Day Report Per Requirements of 10CFR50.46-Surry Power Station Units 1 and 2, North Anna Power Station Units 1 and 2," Serial No. 93-182B, November 9, 1993.
(3) "Surry Power Station Units 1 and 2 - Safety Evaluation for Revised Large Break LOCA Analysis," 10CFRS0.59 Safety Evaluation 94-082, March 28, 1994.
3-2