ML17179B126

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Summary of 930923 Meeting W/Util to Discuss BWR Safety Analysis Topical Repts for Vendor Independence Program
ML17179B126
Person / Time
Site: Dresden, Quad Cities, LaSalle  
Issue date: 10/04/1993
From: Siegel B
Office of Nuclear Reactor Regulation
To:
Office of Nuclear Reactor Regulation
References
NUDOCS 9310080301
Download: ML17179B126 (40)
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i UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 October 4, 1993 Docket Nos. 50-237, 50-249 50-254, 50-265 and 50-373, 50-374 LICENSEE:

Commonwealth Edison Company FACILITIES:

Dresden, Units I and 2 Quad Cities, Units I and 2 LaSalle, Units I and 2

SUBJECT:

SUMMARY

OF MEETING ON BWR SAFETY ANALYSIS A meeting was held between the NRC and Commonwealth Edison Company (CECo) at One White Flint North in Rockville, Maryland on September 23, 1993.

The purpose of the meeting was for CECo to discuss its vendor independence program and, in particular, the approach being utilized to obtain staff approval of its BWR safety analyses topical reports. Enclosure I is a list of meeting attendees. is a copy of the CECo handout that contains the specifics of the proposed methodology to obtain vendor independence.

The introduction and

  • conclusion were presented by T. Riech, the methodology overview and reload application methodology were presented by R. Tsai, and the transient analysis and core thermal limit methodology were presented by J. Freeman.

The staff provided comments during CECo's presentation that were primarily in the form of cautions and advice with regard to how CECo should proceed to obtain staff approval and avoid potential pitfalls that could delay the approval process.

The staffs most significant comments related to the presentation are as follows:

I.

CECo should adhere to the proposed submittal schedule, if significant slippage is anticipated, the staff should be given as much advance notice as possible.

2.

CECo submittals should address the treatment of uncertainties in its methodology.

3.

CECo should assure that its submittals comply with staff conditions contained in Safety Evaluations related to source code topicals previously approved.

4.

To ~ssure methodologies used for thermal limit analyses are valid, CECo will have to provide adequate benchmark data.

5.

CECo should address the effects of mixed cores in its core application topicals.

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OFC NAME DATE COPY Commonwealth Edison Company October 4, 1993 CECo also stated that they would meet with the staff sometime in the second quarter of each calendar year to provide an update of its vendor independence program.

Enclosures:

1.

Meeting attendees

2.

CECo handout cc w/enclosures:

See next page DISTRIBUTION:

w/o enclosures:

T. Murley/F. Miraglia J. Partlow J. Roe J. Zwolinski J. Dyer J. Stang C. Patel J. Kennedy C. Moore OGC E. Jordan R. Jones L. Phillips R. Frahm ACRS (10)

A. Gody, Jr. 17G21 PM: POI II-2 BSIEGf{f original signed by Chandu Patel for Byron L. Siegel, Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation w/enclosures:

Docket File (50-237, 249, 254, 265, 373, and 374)

NRC & Local PDRs PDIII-2 r/f B. Siegel B. Clayton RII I D: POI II-2 JDYER :JiM_

  • Mr. D. L. Farrar Commonwealth Edison Company cc:

Michael I. Miller, Esquire Sidley and Austin One First National Plaza Chicago, Illinois 60690 Mr. G. Spedl Plant Manager Dresden Nuclear Power Station 6500 North Dresden Road Morris, Illinois 60450-9765 U. S. Nuclear Regulatory Commission Resident Inspectors Office Dresden Station 6500 North Dresden Road Morris, Illinois 60450-9766 Chairman Board of Supervisors of Grundy County Grundy County Courthouse Morris, Illinois 60450 Regional Administrator Nuclear Regulatory Commission, Region III 799 Roosevelt Road, Bldg. #4 Glen Ellyn, Illinois 60137 Illinois Department of Nuclear Safety Office of Nuclear Facility Safety 1035 Outer Park Drive Springfield, Illinois 62704 Robert Neumann Office of Public Counsel State of Illinois Center 100 W. Randolph Suite 11-300 Chicago, Illinois 60601 Dresden Nuclear Power Station Unit Nos. 2 and 3

L

  • commonwealth Edison Company cc:

Mr. Stephen E. Shelton Vice President Iowa-Illinois Gas and Electric Company P. 0. Box 4350 Davenport, Iowa 52808 Michael I. Miller, Esquire Sidley and Austin One First National Plaza Chicago, Illinois 60690 Mr. Richard Bax Station Manager Quad Cities Nuclear Power Station 22710 206th Avenue North Cordova, Illinois 61242 Resident Inspector U. S. Nuclear Regulatory Commission 22712 206th Avenue North Cordova, Illinois 61242 Chairman Rock Island County Board of Supervisors 1504 3rd Avenue Rock Island County Office Bldg.

Rock Island, Illinois 61201 Illinois Department of Nuclear Safety Office of Nuclear Facility Safety 1035 Outer Park Drive Springfield, Illinois 62704 Regional Administrator, Region III U. S. Nuclear Regulatory Commission 799 Roosevelt Road, Bldg. #4 Glen Ellyn, Illinois 60137 Robert Neumann Office of Public Counsel State of Illinois Center 100 W. Randolph Suite 11-300 Chicago, Illinois 60601 Quad Cities Nuclear Power Station Unit Nos. 1 and 2

  • commonwealth Edison Company cc:

Phillip P. Steptoe, Esquire Sidley and Austin One First flational Plaza Chicago, Illinois 60603 Assistant Attorney General 100 West Randolph Street Suite 12 Chicago, Illinois 60601 Resident Inspector/LaSalle, NPS U. S. Nuclear Regulatory Commission Rural Route No. 1 P. 0. Box 224 Marseilles, Illinois 61341 Chairman LaSalle County Board of Supervisors LaSalle County Courthouse Ottawa, Illinois 61350 Attorney General 500 South 2nd Street Springfield, Illinois 62701 Chairman Illinois Commerce Commission Leland Building 527 East Capitol Avenue Springfield, Illinois 62706 Illinois Department of Nuclear Safety Office of Nuclear Facility Safety 1035 Outer Park Drive Springf1eld, Illinois 62704 Regional Administrator, Region III U. S. Nuclear Regulatory Commission 799 Roosevelt Road, Bldg. #4 Glen Ellyn, Illinois 60137 Robert Neuman Office of Public Counsel State of Illinois Center 100 W. Randolph Suite 11-300 Chicago, Illinois 60601 LaSalle County Station Unit Nos. 1 and 2 Robert Cushing Chief, Public Utilities Division Illinois Attorney General's Office 100 West Randolph Street Chicago, Illinois 60601 Michael I. Miller, Esquire Sidley and Austin One First National Plaza Chicago, Illinois 60690 Mr. G. Diederich LaSalle Station Manager LaSalle County Station Rural Route 1 P. 0. Box 220 Marseilles, Illinois 61341

B. Siegel J. Kennedy R. Jones L. Phillips R. Frahm T. Rieck S. Silady K. Kovar R. Tsai J. Freeman MEETING WITH CECo TO DISCUSS BWR SAFETY ANALYSES TOPICAL REPORTS FOR CECo'S VENDOR INDEPENDENCE PROGRAM SEPTEMBER 23. 1992 Affiliation NRC NRC NRC NRC NRC CECo CECo CE Co CE Co CECo ENCLOSURE 1

~ ENCLOSURE 2 Commonwealth Edison Company BWR Safety Analysis Vendor Independence Program BSANRCA.DOC NRC/CECo Meeting September 23, 1993 USNRC Offices White Flint, Maryland

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo *Meeting - 9/23/93 AGENDA

1. Introduction Review Agenda, Meeting Objectives, CECo/NFS Organization, CECo VIP Prag ram
2. Methodology Overview Approach, Schedule, Computer Codes Used, Planned Topical Reports
3. Transient Analysis Methodology Reactor System Models, 1 *D Kinetics Methodology, Computer Codes Used, Benchmarking Studies, Topical Report
4. Core Thermal Limit Methodology Core Thermal Limit Method Overview, Computer Codes Used, Benchmarking Studies, Topical Report
5. Reload Application Methodology Bases of Reload Analysis Applications
6. Conclusions CECo Summary
7. NRC Discussions NRC Feedback on CECo BWR SA VIP BSANRCB.DOC

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo *Meeting - 9/23/93 AGENDA

1. Introduction Review Agenda, Meeting Objec;:tives, CECo/NFS Organization, CECo VIP Program
2. Methodology Overview Approach, Schedule, Computer Codes Used, Planned Topical Reports
3. Transient Analysis Methodology Reactor System Models, 1-D Kinetics Methodology, Computer Codes Used, Benchmarking Studies, Topical Report
  • '4. Core Thermal Limit Methodology Core Thermal Limit Method Overview, Computer Codes Used, Benchmarking Studies, Topical Report
5. Reload Application Methodology Bases of Reload Analysis Applications
6. Conclusions CECo Summary
7. NRC Discussions NRC Feedback on CECo BWR SA VIP BSANRCB.DOC

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 9/23/93 (Introduction)

MEETING OBJECTIVES To inform NRC of CECo's overall plan, schedule, and technical approach for the CECo BWR Safety Analysis Vendor Independence Program (VIP).

To obtain NRC.feedback on CECo plan on

BSANRCC.DOC

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' r BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 9/23/93 (Introduction)

CECO VIP OBJECTIVES

  • 12 Nuclear Units in CECo System
  • In-House Capability Benefits Operational Safety & Technical Support

- Enhance understanding of plant transient behavior

- Enhance understanding of plants

  • - Enhance regulatory support

- Enhance operational support

- Enhance input & evaluation of plant modifications & other changes Economic

- Timely support to stations

- More efficient designs

- Vendor cost replacement BSANRCCl.DOC

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 9/23/93 (Introduction)

CECO VIP STATUS PWR Nuclear Design

  • W based methodology
  • NRC approved 1983/1991
  • 22 CECo reload design analyses BWR Nuclear Design
  • GE/SPC based methodology
  • NRC approved 1992/1993
  • 9 CECo reload design analyses PWR Safety Analysis
  • EPRI (RETRAN/VIPRE) based methods
  • Zion thermal limit approved 1990
  • B/B SGTR approved 1 992
  • Zion RTDP/OTL\\T-OPL\\T approved 1993
  • B/B thermal limit expected 1993

-. B/B transient analysis to be submitted 1 st Quarter 1994 BWR Safety Analysis Began development in December 1990 BSANRCO.DOC

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 9/23/93 (Introduction)

CECO BWR UNITS & FUEL VENDORS Station

  1. of NSSS Contain-Current Name Units Type ment Fuel Type Vendor Dresden 2

BWR/3 Mark I Siemens Power Corp Quad-Cities 2

BWR/3 Mark I General Electric*

LaSalle 2

BWR/S Mark II General Electric*

  • Transition to SPC fuel in 1995 BSANRCE.DOC

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 9/23/93 I

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Mamgemcrs 10 Stiff MO!rilcrs BSANRCF.DOC (Introduction)

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BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 9/23/93 AGENDA

1. Introduction Review Agenda, Meeting Objectives, CECo/NFS Organization, CECo VIP Program
2. Methodology Overview Approach, Schedule, Computer Codes Used, Planned Topical Reports
3. Transient Analysis Methodology Reactor System Models, 1-D Kinetics Methodology, Computer Codes Used, Benchmarking Studies, Topical Report
4. Core Thermal Limit Methodology Core Thermal Limit Method Overview, Computer Codes Used, Benchmarking Studies, Topical Report

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5. Reload Application Methodology Bases of Reload Analysis Applications
6. Conclusions CECo Summary
7. NRC Discussions N RC Feed back on CECo BWR SA VIP BSANRCB.DOC

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BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 9/23/93 (Methodology Overview)

APPROACH AND SCHEDULE

  • Vendor Nuclear Design Methods
  • CECo/EPRI Safety Analysis Methods
  • Topicals in Three Parts:
1. Transient Analysis Methodology

- Topical submitted to NRC - 12/94

- SER requested - 12/95

2. Core Thermal Limit Methodology

- Topical submitted to NRC - 12/95

- SER requested - 12/96

3. Reload Application Methodology

- Topical submitted to NRC - 12/96

- SER requested - 6/97 BSANRCG.DOC

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'.A BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 9/23/93 (Methodology Overview)

TOPICAL REPORTS

1. Transient Analysis Methodology
  • CECo BWR Plant Models - RETRAN
  • 1 D Kinetics Methodology - PETRA
  • CECo Plant Data Benchmarking
  • PB NRC Test Problem Benchmarking
2. Core Thermal Limit Methodology
  • CECo BWR Core Models - FIBWR2
  • Steady State Data Benchmarking
  • Statistical Methods
3. Reload Application Methodology
  • Bases for Reload Analysis Applications BSANRCJ.DOC

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 9/23/93 (Methodology Overview)

METHODOLOGY OVERVIEW

  • 1 D Kinetics Methodology Vendor nuclear design methods PETRA - generic 1 D kinetics method
  • Transient Analysis Methodology RETRAN, FIBWR2, ESCORE based
  • Core Thermal Limit Methodology Fl BWR2 based Vendor CPR Correlations Statistical Methods for Uncertainty (Consistent with NRC approved methods)

BSANRCH.DOC

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 9/23/93 (Methodology Overview)

COMPUTER CODES Code Use Source Previously

  • Name Approved by NRC CASMO 3-D Lattice SPC YES Physics (2 Group)

MICRO BURN 3-D Core SPC YES Simulator (2 Group)

PETRA 3-D to 1-D ScandPower NO Cross Section Collapse (2 Group)

RETRAN-02, Transient EPRI

~

YES MODS Analysis ESCO RE Transient EPRI YES Analysis FIBWR2 Transient Scientech, Inc.

NO*

Analysis &

Core Thermal Limit

  • FIBWR code has been previously approved, of which FIBWR2 is an enhanced version with transient capability BSANRCl.DOC

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 9/23/93 AGENDA

1. Introduction Review Agenda, Meeting Objectives, CECo/NFS Organization, CECo VIP Program
2. Methodology Overview Approach, Schedule, Computer Codes Used, Planned Topical Reports
3. Transient Analysis Methodology Reactor System Models, 1-D Kinetics Methodology, Computer Codes Used, Benchmarking Studies, Topical Report
4. Core Thermal Limit Methodology Core Thermal Limit Method Overview, Computer Codes Used, Benchmarking Studies, Topical Report
5. Reload Application Methodology Bases of Reload Analysis Applications
6. Conclusions CECo Summary
7. NRC Discussions NRC Feedback on CECo BWR SA VIP BSANRCB.DOC

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 9/23/93 (Transient Analysis Methodology)

REACTOR SYSTEM MODEL

  • Reactor system models: RETRAN02, Mods
  • Steady state core thermal hydraulics models: FIBWR2
  • Gap conductance: ESCORE
  • 1 D kinetics data: PETRA PETRA ESCO RE FIBWR2S.S.

BSANRCP.DOC

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, r RETRAN THERMAL LIMIT METHODOLOGY TRANSIENT INPUTS

.... _.___ CHANGE DECK

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 9/23/93 (Transient Analysis Methodology)

TRANSIENT ANALYSIS COMPUTER CODES

  • RETRAN02, Mods (EPRI), Approved by NR~ with SER on 11/1/91 provides overall system response
  • FIBWR2 (Scientech), FIBWR has been previously approved for other utilities provides RETRAN with steady state core flow distribution and pressure drop data
  • ESCORE (EPRI), Approved by NRC with SER on 5/23/90 provides RETRAN with fuel rod gap conductance data
  • PETRA (Scandpower), provides RETRAN with core 1 D kinetics data
  • Utility codes writte~ under CECo QA
  • procedures provide automation of data transfer between Vendor Codes, RETRAN, FIBWR2, ESCORE, and PETRA BSANRCR.DOC

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 9/23/93 (Transient Analysis Methodology)

CECO BWR RETRAN SYSTEM MODELS Description Dresden Quad-Cities LaSalle Number of Core Nodes 24 24 25 Number of Core Neutronic 24 24 25 Reqions Number of Core Conductors 24 24 25 Number of Core Bypass 24 24 25 Conductors Number of Core Reflectors 2

2 2

Number of Main Steam Line 7

7 7

Nodes Non-Equilibrium Option used?

YES YES YES Alqebraic Slio Ootion used?

YES YES YES Recirculation Flow Control System MG MG Valve Number of Recirculation Loops 2

2 2

Number of Safety Valve Junctions 8

8 0

Number of Relief Valve Junctions 5

5 0

Number of Combination 1

1 18 Safety/Relief Valves BSANRCO.DOC

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 9/23/93 (Transient Analysis Methodology)

RETRAN l D KINETICS FEEDBACK MODEL I.

RETRAN contains a flexible 1-D kinetic model. A separate input file provides the coefficients which describe the diffusion theory constants (f3eff Lal B12 D1 Ls12 KLn YLn 1/Vl Laz B22 Dz ICLfz YLfz 11v2 ) and their thermal feedback relationships.

nl n2 n3

. 1

. 1 k 1 Z(xl,x2,x3) = Li=l Lj=l Lk=l C(i,j,k) [(x1 1- )(x2J- )(x2 - )]

where:

xl = [u(t)-u(O)]/u(O)

II.

The relative moderator density is fit with a quadratic relationship and the fuel temperature is fit with a linear relationship without cross terms:

Z(xl,x2)= C(l,1,1) + C(2,l,l) xl + C(3,l,l) xl2 + C(l,2,1) x2 Ill.

Fuel temperature (Doppler) feedback is only applied to Lal and Ls12*

IV.

The RETRAN calculated relative moderator density values will be used for the coefficient fitting.

BSANRCK.DOC

BSANRCN.DOC BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting* 9/23/93 (Transient Analysis Methodology)

KINETICS DATA FLOW (link with SIEMENS Methodology)

BASE POINT '

MICROBURN MIC PET PETRA RETRAN CORE SUB MODEL 1-D d PETRA

, r CASMO

~

POSTPET

  • EXECUTE FOR EACH CONTROL STATE
  • PRESSURE PERTUBATION CASE
  • CHANGE FORMAT OF CROSS SECTIONS
  • COLLAPSE CROSS SECTIONS
  • CALCULATE 1-DIMENSIONAL DENSITIES ensities (all)
  • COLLAPSE CROSS SECTIONS
  • CALCULATE FIT COEFFICIENTS
  • ADD DOPPLER COEFFICIENTS
  • CALCULATE CROSS SECTION LIMITS RETRAN 1-D CROSS SECTION FILE

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE 1'ROGRAM NRC/CECo Meeting - 9/23/93 (fransient Analysis Methodology) 1 D KINETICS METHODS COMPARISONS Lattice Phvsics 3-D Simulator 1-D Linkage Transient GE:

TGBLA

~

PANACEA

~

CRNC

~.

ODYN Si.emens:

CASM0-3

~ MICROBURN

~ PRECOT2 ~ COTRANSA2 Studsvik:

CASM0-3

~ SIMULATE-3

~

SLICK

~ RETRAN Scandpower:RECORD

~

PRESTO

~

PETRA

~ RAMONA ABB:

PHOENIX

~. POLCA

~

WPOL

~

BISON PP&L, WPPSS, PECO:

CPM-2

~ SIMULATE-E

~ SIMTRAN ~ RETRAN CECO:

CASM0-3

~ MICROBURN ~ PETRA ~ RETRAN BSANRCL.DOC

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 9/23/93 (Transient Analysis Methodology) l D KINETICS CODE: PETRA

  • Developed by Scandpower, Inc.
  • Collapses nodal distribution of cross-section data using appropriate averaging schemes (adjoint flux weighting)
  • Allows input of 1 D density distributions (f ram RETRAN)
  • Allows input of global kinetic parameters
  • Calculate polynomial fit coefficients for cross-section parameters needed by RETRAN 1 D kinetics model
  • Adjust fast absorption cross-section to match the axial power shape and eigenvalue from the 1 D solution to the results from the 30 simulator BSANRCM.DOC

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 9/23/93 (Transient Analysis Methodology)

PLANT DATA BENCHMARKING..

  • CECo BWR Units Start Up Tests

- Pressure Regulator Setpoint Change

- Reactor Water Level Setpoint Change

- Bypass Valve Change

- Feedwater Pump Trip

- Two Recirculation Pump Trip

- MSIV Closure

- Turbine Trip

- Recirculation Flow Performance (LaSalle)

- Recirculation Control System (LaSalle)

- Loss of Offsite Power (LaSalle)

  • Startup Test Data Acceptance Criteria

- Steam Dome Pressure

- Downcomer Level

- Feedwater Flow Rate

- Recirculation Loop Flow Rate

- Reactor Power

  • Benchmarking Discussions

- Sensitivity study summary

  • Peach Bottom NRC Test Problem BSANRCQ.DOC
  • I~.,-

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo-Meeting - 9/23/93 AGENDA

1. Introduction Review Agenda, Meeting Objectives, CECo/NFS Organization, CECo VIP Program
2. Methodology Overview Approach, Schedule, Computer Codes Used, Planned Topical Reports
3. Transient Analysis Methodology Reactor System Models, 1-D Kinetics Methodology, Computer Codes Used, Benchmarking Studies, Topical Report
4. Core Thermal Limit Methodology Core Thermal Limit Method Overview, Computer Codes Used, Benchmarking Studies, Topical Report *
5. Reload Application Methodology Bases of Reload Analysis Applications
6. Conclusions CECo Summary
7. NRC Discussions N RC Feed back on CECo BWR SA VIP BSANRCB.DOC

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 9/23/93 (Core Thermal Limit Methodology)

CORE THERMAL LIMIT METHOD OVERVIEW.

  • Steady State Core Thermal Hydraulic Analysis
  • Hot Channel Model
  • Operating Limit MCPR (Statistical methods)

PLANT DATA FUEL DATA

~

~

RETRAN PLANT MODEL

  • ONE DIMENSIONAL HYDRAULICS
  • ONE DIMENSIONAL KINETICS
  • CORE AVERAGE Hgap (ESCOREl
  • CONSIDERS SYSTEM AS A WHO E CORE BOUNDARY CONDITIONS l,

FUEL DATA ~

CORE MODEL

  • MULTI.CHANNEL HYDRAULICS FIBWR2
  • HOT CHANNEL Hgap (ESCORE) i.e. Loss Coef, Bundle Phys Char ACPR UNADJUSTED
  • PERFORMANCE BOUNDING ANALYSIS ACPR OPERA TING LIMIT MCPR BSANRCS.DOC

J' BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 9/23/93 (Core Thermal Limit Methodology)

THERMAL LIMIT COMPUTER CODE - FIBWR2

.* Developed by Scientech, Inc.

  • Use RETRAN generated boundary conditions including time varying axial power
  • Use power distribution generated by 30 simulators
  • Detailed geometric modeling of a fuel assembly including inlet orifice, fuel support piece, lower tie plate, unheated fuel regions, grid spacers, water tubes, and upper tie plate

~ Predicts flow distribution for a given power distribution for multi-channel BWR core

  • Fuel rod heat conduction model BSANRCT.DOC

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE4'ROGRAM NRC/CECo Meeting - 9/23/93 (Core Thermal Limit Methodology)

FIBWR2 CORE MODEL BENCHMARKING

  • Steady state analysis: vendor codes and plant test data

vendor codes

  • Thermal limit analysis (using vendor CPR 1

correlations): Vendor methodologies and CPR test data BSANRCV.DOC

.t BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo-Meeting - 9/23/93 (Core Thermal Limit Methodology)

STATISTICAL METHOD

  • Applied to parameters with large impact on ACPR
  • Perform RETRAN/FIBWR2 calculations
  • Statistical methods consistent with NRC approved
  • Analysis to determine 95/95 criteria for unadjusted ACPR/ICPR
  • Statistical method consistent with vendor methodology BSANRCW.DOC

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CEco-Meeting - 9/23/93 AGENDA

l. Introduction Review Agenda, Meeting Objectives, CECo/NFS Organization, CECo VIP Program
2. Methodology Overview Approach, Schedule, Computer Codes Used, Planned Topical.

Reports 3*. Transient Analysis Methodology Reactor System Models, 1-D Kinetics Methodology, Computer Codes Used, Benchmarking Studies, Topical Report

4. Core Thermal Limit Methodology Core Thermal Limit Method Overview, Computer Codes Used, Benchmarking Studies, Topical Report
5. Reload Application Methodology Bases of Reload Analysis Applications
6. Conclusions CECo Summary
7. NRC Discussions N RC Feed back on CECo BWR SA VIP BSANRCB.DOC

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BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo-Meeting - 9/23/93 (Reload Application Methodology)

Applications Topical Purpose and Scope

  • Provide a representative transient analysis for one specific unit/cycle
  • Define CECo application of Reactor System Transient Analysis and Thermal Limit Methodology for Limiting Events
  • Define CECo appl'ication of Reactor System Transient Analysis Methodology for ASME Overpressurization Events
  • Define CECo position for Non-Limiting Events
  • Define CECo Methodology for development of Safety Limit for Technical Specifications BSANRCWl.DOC

BWR SAFETY ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo-Meeting - 9/23/93 AGENDA

1. Introduction Review Agenda, Meeting Objectives, CECo/NFS Organization, CECo VIP Program
2. Methodology Overview Approach, Schedule, Computer Codes Used, Planned Topical Reports
3. Transient Analysis Methodology Reactor System Models, 1-D Kinetics Methodology, Computer Codes Used, Benchmarking Studies, Topical Report
4. Core Thermal Limit Methodology Core Thermal Limit Method Overview, Computer Codes Used,.

Benchmarking Studies, Topical Report

5. Reload Application Methodology Bases of Reload Analysis Applications
16. Conclusions CECo Summary
7. NRC Discussions N RC Feed back on CECo BWR SA VIP BSANRCB.DOC

-~*

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BWR SAFE1Y ANALYSIS VENDOR INDEPENDENCE PROGRAM NRC/CECo Meeting - 9/23/93 CONCLUSIONS

  • Ongoing communication to assure acceptable schedule and priority
  • Future NRC/CECo Meeting Plan BSANRCZ.DOC 2nd-quarter, 1994: to present preliminary transient analysis results 2nd-quarter, 1995: to discuss transient analysis topical and to present preliminary thermal limit methods results 2nd-quarter, 1996: to discuss thermal limit methods topical

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BWR SAFETY ANALYSIS VENDOR INDEPENDENCE"PROGRAM NRC/CECo *Meeting - 9/23/93 AGENDA

l. Introduction Review Agenda, Meeting Objectives, CECo/NFS Organization, CECo VIP Program
2. Methodology Overview Approach, Schedule, Computer Codes Used, Planned Topical.

Reports

3. Transient Analysis Methodology Reactor System Models, 1-D Kinetics Methodology, Computer Codes Used, Benchmarking Studies, Topical Report
4. Core Thermal Limit Methodology Core Thermal Limit Method Overview, Computer Codes Used, Benchmarking Studies, Topical Report
5. Reload Application Methodology Bases of Reload Analysis Applications
6. Conclusions CECo Summary
7. NRC Discussions N RC Feed back on CECo BWR SA VIP BSANRCB.DOC

HPCI7°6.r-t=-'1'1-FEE DW ATER 501 2*'-r+--" r-~

LPCI 104 51 41 Core VoluMes Vol400 to 425 00 and Quad-* Ci ties RETRAN Model 101 100 LPCI 91 Bl RELIEF VALVES SUPPRESSION POOL 102 SAFETY VALVES 300 DRY WELL 10 174 INBOARD MSIV 175 OUTBOARD M~IV e 602 TURBINE CONTROL VALVE STOP VALVE

e BWR Reactor System Models Dresden and Quad-Cities Description Dresden Number of Core Nodes 24 Number of Core Neutronic Regions 24 Number of Core Conductors 24 Number of Core Bypass Conductors 24 Number of Core Reflectors 2

Number of Main Steam Line Nodes 7

Non-Equilibrium Option used?

YES Algebraic Slip Option used?

YES Recirculation Flow Control System MG Number of Recirculation Loops 2

Number of Safety Valve Junctions 8

Number of Relief Valve Junctions 5

Number of Combination Safety/Relief 1

l.. -----*-**------ -- --*-

~

LaSalle RETRAN Model RETRAN Model TURBINE CONTROL 11 VALVE Quad-Cities 24 24 24 24 2

7 YES YES MG 2

8 5

1

  • -. __ _f

,.,~ARD LaSalle 25 25 25 25 2

7 YES YES Valve 2

0 0

18

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