ML023500330

Safety Analysis Transition to Westinghouse Meeting Objectives

ML023500330
Person / Time
Site:	Prairie Island
Issue date:	12/13/2002
From:	NRC/NRR/DLPM/LPD4
To:
References
Download: ML023500330 (17)

Text

Prairie Island Safety Analysis Transition To Westinghouse Meeting Objectives

1. Inform NRC of plans and schedules

2. Obtain feedback from the NRC Overview Transferring non-LOCA safety analysis scope from NMC to Westinghouse approved methods for Prairie Island.

Technical Specification reference additions and other changes consistent with Westinghouse methods will be needed.

Generally, the change is with approved methods. A few clarifications will be discussed.

Plan to change methods for both units in September 2003 with the Prairie Island Unit 2 Cycle 22 refueling outage.

Plan to have submittal to NRC in February 2003.

ENCLOSURE 2

- i

Prairie Island Safety Analysis Transition To Westinghouse Meeting Agenda Introduction / Meeting Agenda - Jack Leveille Meeting Objectives - Cliff Bonneau Safety Analysis Transition Program Scope - Lamar Brown Schedule - Jack Leveille

Prairie Island Safety Analysis Transition To Westinghouse Safety Analysis Transition Program Scope Overview FSAR CHAPTERS 3 and 14 Fuel Analysis Core Physics Thermal/Hydraulics Fuel Rod Design FSAR CHAPTER 14 Most FSAR Chapter 14 events will be reanalyzed by Westinghouse Replacement steam generator design characteristics will be incorporated Non-LOCA Analysis LOCA Analysis Radiological Analysis FSAR CHAPTER 14 AND APPENDICES I AND K Containment Integrity Analysis Outside Containment Analysis

Prairie Island Safety Analysis Transition To Westinghouse Safety Analysis Transition Program Scope Fuel Analysis CORE PHYSICS (FSAR Chapters 3 and 14)

Nominal parameters will be calculated: power distributions, reactivity coefficients, rod worths, boron concentrations Core physics data to be used in safety analysis will be calculated: nominal parameters plus event specific parameters (ejected rod worths, steamline break peaking factors, etc.)

Power history data will be calculated for input to fuel rod design Approved Codes/Methodology Topicals

1. PHOENIX-P, WCAP-11596-P-A

2. ANC, WCAP-10965-P-A THERMAL/HYDRAULICS (FSAR Chapters 3 and 14)

The Revised Thermal Design Procedure will be implemented Core limit lines and axial offset limit lines will be calculated assuming implementation of Relaxed Axial Offset Control strategy Confirmation will be provided that DNBR limits are met Approved Codes/Methodology Topicals

1. Revised Thermal Design Procedure, WCAP-11397-P-A

2. VIPRE, WCAP-14565-P-A

Prairie Island Safety Analysis Transition To Westinghouse Safety Analysis Transition Program Scope Fuel Analysis FUEL ROD DESIGN (FSAR Chapter 3)

No change to current Westinghouse methodology as applied to Prairie Island PAD will continue to be used to verify that all fuel rod design criteria are met Fuel rod temperatures will be calculated for input to LOCA and non-LOCA analyses Approved Codes/Methodology Topicals

1. PAD, WCAP-10851-P-A, Rev 1

Prairie Island Safety Analysis Transition To Westinghouse Safety Analysis Transition Program Scope NON-LOCA Analysis EVENT APPROVED CODES/METHODOLOGY RCCA Bank TWINKLE / FACTRAN / VIPRE-01 Withdrawal from Subcritical, FSAR 14.4.1 RCCA Bank RETRAN-02 Withdrawal at Power, FSAR 14.4.2 RTDP RCCA Misalignment, Statically Misaligned RCCA - ANC /

FSAR 14.4.3 VIPRE-01 (Statically Misaligned Dropped RCCA(s) - Generic 2-loop WOG RCCA and Dropped Statepoints (LOFTRAN),

RCCA(s))

WCAP-11394-P-A methodology Boron Dilution, Reactor Critical - RETRAN (reactivity FSAR 14.4.4 insertion bounded by RWAP)

Reactor Subcritical - NMC methods retained Startup of an Inactive Operation with a loop out of service Loop, FSAR 14.4.5 precluded by Tech Specs. FSAR will be revised to reflect this. No analysis.

Feedwater RETRAN-02 Malfunction, FSAR 14.4.6 RTDP

Prairie Island Safety Analysis Transition To Westinghouse Safety Analysis Transition Program Scope NON-LOCA Analysis EVENT APPROVED CODES/METHODOLOGY Excessive Load RETRAN-02, or evaluation vs. core

Increase, thermal limits FSAR 14.4.7 Loss of Flow, RETRAN-02 / VIPRE-01 FSAR 14.4.8.1 RTDP Locked Rotor, RETRAN-02 / VIPRE-01 FSAR 14.4.8.2 RTDP (rods in DNB analysis)

Loss of Load I RETRAN-02 Turbine Trip, RTDP (DNB analysis)

FSAR 14.4.9 Loss of Normal RETRAN-02 Feedwater and Loss of AC Power, FSAR 14.4.10, 14.4.11 Steam Line Break RETRAN-02 / VIPRE-01 / ANC (core response),

FSAR 14.5.5 RCCA Ejection, TWINKLE / FACTRAN FSAR 14.5.6 ATWS, FSAR 14.8 RETRAN-02 (validation of Diverse Scram System)

Prairie Island Safety Analysis Transition To Westinghouse Safety Analysis Transition Program Scope NON-LOCA Analysis Approved Code Topicals

1. TWINKLE

2. FACTRAN

3. RETRAN-02

4. LOFTRAN

5. ANC

6. VIPRE-01 WCAP-7979-P-A, TWINKLE - A Multidimensional Neutron Kinetics Computer Code, January 1975.

WCAP-7908-A, FACTRAN -A FORTRANIV Code for Thermal Transients in a U02 Fuel Rod, December 1989.

WCAP-14882-P-A, RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses, April 1999.

WCAP-7907-P-A, LOFTRAN Code Description, April 1984.

WCAP-10965-P-A, ANC: Westinghouse Advanced Nodal Computer Code, September 1986.

WCAP-14565-P-A, VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis, October 1999.

Prairie Island Safety Analysis Transition To Westinghouse Safety Analysis Transition Program Scope NON-LOCA Analysis Approved Methodology Topicals

1. Reload

2. DNB

3. Overpower AT and Overtemperature AT Trip Setpoints

4. Dropped Rod

5. Rod Ejection

6. Boron Dilution

7. Pressurizer Safety Valve Modeling

8. Steam Line Break WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985.

WCAP-11397-P-A, Revised Thermal Design Procedure, April 1989.

WCAP-8745-P-A, Design Bases for the Thermal Oveipower AT and Thermal Overtemperature AT Trip Functions, September 1986.

WCAP-11394-P-A, Methodology for the Analysis of the Dropped Rod Event, January 1990.

WCAP-7588, Rev. 1-A, An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors, January 1975.

NSPNAD-8102-PA, Revision 7, Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods For Application to PI Units, (Appendix G), July 1999.

WCAP-12910, Rev. 1-A, Pressurizer Safety Valve Set Pressure Shift, Westinghouse Owners Group, May 1993.

WCAP-9226-P-A, Revision 1, Reactor Core Response to Excessive Secondary Steamr Releases, February 1998.

Prairie Island Safety Analysis Transition To Westinghouse Safety Analysis Transition Program Scope LOCA Analysis No change to current Westinghouse methodology as applied to Prairie Island LARGE BREAK ANALYSIS (FSAR Chapter 14.6)

The SECY Upper Plenum Injection methodology using WCOBRA/TRAC will continue to be employed Appendix K and Superbounded cases will be run for the limiting break using replacement steam generator data Approved Codes/Methodology Topicals

1. WCOBRA/TRAC, WCAP-10924-P-A SMALL BREAK ANALYSIS (FSAR Chapter 14.7)

NOTRUMP will continue to be employed A complete break spectrum will be analyzed using replacement steam generator data Approved Codes/Methodology Topicals

1. NOTRUMP, WCAP-10054-P-A

2. Small Break Evaluation Methodology, WCAP-10079-P-A

Prairie Island Safety Analysis Transition To Westinghouse Safety Analysis Transition Program Scope LOCA Analysis POST-LOCA ANALYSES (FSAR Chapter 14.10)

Subcriticality will continue to be verified using current methods Boron buildup will continue to be analyzed using current methods LOCA HYDRAULIC FORCES (FSAR Chapter 3.6)

MULTIFLEX will continue to be employed The LOCA forces determined for Unit 2 for the 25% SGTP program will be re-validated The impact of the replacement steam generator will be evaluated for Unit 1 Approved Codes/Methodology Topicals

1. MULTIFLEX, WCAP-8708-P-A

Prairie Island Safety Analysis Transition To Westinghbouse Safety Analysis Transition Program Scope Radiological Analysis Fauske and Associates (Westinghouse subsidiary) has performed dose analysis using the Alternate Source Term Methodology under a separate program and a report will be submitted separately by NMC The Alternate Source Term Methodology will be implemented when approved.

If the Alternate Source Term Methodology is not approved for use on Prairie Island for Unit 2 Cycle 22 (August 2003), NMC methodology may be retained for Locked Rotor analysis Methodology

1. Alternate Source Term Methodology, Reg Guide 1.183

Prairie Island Safety Analysis Transition To Westinghouse Safety Analysis Transition Program Scope Containment Integrity Analysis LOCA MASS & ENERGY RELEASES (FSAR Appendix K)

This data will be provided by NMC for the replacement steam generator STEAMLINE BREAK MASS & ENERGY RELEASES (FSAR Chapter 14.5.5)

Mass & energy releases will be calculated using LOFTRAN for both Unit 1 (RSG) and Unit 2 (OSG)

Various power levels, break sizes, and single failures will be examined As a contingency, NOTRUMP may be used to support break quality assumptions for the RSG's Approved Codes/Methodology Topicals

1. Methodology for Calculation of SLB M&E's Inside Containment, WCAP-8822

Prairie Island Safety Analysis Transition To Westinghouse Safety Analysis Transition Program Scope Containment Integrity Analysis CONTAINMENT RESPONSE (FSAR Chapter 14.5.5 and Appendix I)

Containment response for LOCA and SLB will be performed using either Gothic v. 7.0 or MAAP5 Codes/Methodology Topicals (Currently under NRC Review)

1. Gothic v. 7.0, NRC-02-082, Kewaunee Nuclear Power Plant Request for Use of Gothic v. 7.0 in Containment Design Basis Accident Analysis, Sept 2, 2002

2. MAAP5, WCAP-15844, Rev 0, March 2002

Prairie Island Safety Analysis Transition To Westinghouse Safety Analysis Transition Program Scope Outside Containment Analysis LONG TERM SLB OUTSIDE CONTAINMENT (FSAR APPENDIX I)

Westinghouse will calculate mass & energy releases using LOFTRAN NMC will perform compartment calculation to determine temperature profiles for equipment qualification Approved Codes/Methodology Topicals

1. SLB M&E Outside Containment Methodology, WCAP 10961, Rev 1 SHORT TERM M&E RELEASES OUTSIDE CONTAINMENT (FSAR APPENDIX I)

Westinghouse will validate M&E releases from steamline and feedline breaks NMC will perform compartment pressurization calculations No NRC approval required Methodology

1. Short term M&E releases outside containment, Appendix E of ANSI/ANS-58.2-1980

Prairie Island Safety Analysis Transition To Westinghouse Safety Analysis Transition Program Scope Implementation Implementation of the Westinghouse methodology will require:

Changes to the Technical Specifications References to methodology topicals Format (RAOC)

FSAR changes The validity of the safety analysis will be confirmed on a cycle by-cycle basis in accordance with the methodology described in the Westinghouse Reload Methodology Topical, WCAP-9272-P-A.

Prairie Island Safety Analysis Transition To Westinghouse NRC Involvement Schedule

Change both units in September 2003 with Prairie Island Unit 2 Cycle 22

> Submittal for Technical Specification Change in February 2003

> Methodology references in TS (adding methodologies rather than replacing)

> Changes for Relaxed Axial Offset Control

> No Setpoint changes

° Approvals needed for other submittals

> MAAP 5, or Gothic 7.0 (not on PI docket)

> Alternate Source Term Methodology (ASTM)

• Contingencies

> Locked rotor accident analysis (may need to maintain current analysis until ASTM is approved)

> 3-D Rod Ejection (may need approval for previously submitted Westinghouse methodology, depending on results of analysis)

> Calculation of Liquid Entrainment for Replacement Steam Generators (may need to use NOTRUMP)