ML20213G656
| ML20213G656 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 05/13/1987 |
| From: | Vietticook A NRC OFFICE OF SPECIAL PROJECTS |
| To: | Charemagne Grimes NRC OFFICE OF SPECIAL PROJECTS |
| Shared Package | |
| ML20213G658 | List: |
| References | |
| NUDOCS 8705180388 | |
| Download: ML20213G656 (67) | |
Text
- _ - _ .
t 8 s* M co UNITED STATES
/ 'o NUCLEAR REGULATORY COMMISSION E ' ) e,( [,j W ASHINGTON. D. C. 20655
/ E 1
- W 1 6 1907 -
. .Y. .
- O C2 Y i Docket No(s). 50-445 AQ.WCE g and 50-446 MEMORANDUM FOR:
Christopher I. Grimes, Director Comanche Peak Project Division Office of Special Projects FROM:
Annette Vietti-Cook, Project Manager Comanche Peak Project Division Office of Special Pro.iects
SUBJECT:
FORTHCOMING MEETING WITH TV ELECTRIC DATE & TIME: 6 Ike5 d ay , Jm 2, N7 6 I: 00pm. -
s':oo p m.
LOCATION: USNRC dhhhimismeseseMYWWF91!T!P Eosf-tJef-Wof BIdg. Ecom 3R N E a d - Wof Hr3k wa[ 91 s~o Bethesda, Maryland PURPOSE:
To discuss plant specific analysis for steam generator tube rupture for Cor.anche Peak.
NRC Applicant PARTICIPANTS:
J. Lyons J. Hicks, et al.
H. Schierlino M. Hodges R. .lones
/-
l J.A %.
Annette Vietti-Cook, Pro.iect Manager Comanche Peak Project Division L C1 Achief .'" Office of Special Pro.iects
( O t0tT *. H 30I - 492 5 33 cc: J. Keppler J. Axelrad P. McKee I. Barnes cc: See next page
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W. G. Counsil .
Comanche Peak Steam Electric Station l
' Texas Udilities' Electric Company Units 1 and 2 1 s .
CC' . .
Thomas G. Dignaq, Jr. Resident Inspector / Comanche Peak l Ropes & Gray Steam Electric Station l 225 Franklin Street c/o U.S. Nuclear Regulatory Commission '
Boston, Messachusett's 0?l10 P.O. Box 38 s
s Glen Rose, Texas 76043
- 1) T _q Robert A. Wooldridge, Et<1. / , s Regional Administrator, Region P/ j Worsham, Forsythe, Tampels & 1 U.S. Nuclear Regulatory Commissdon '
Wooldridge , 611 Ryan Plaza Crive, Suite 1000 2001 Bryan Tower, Suites 2500 Arlington, Texas. 76011 a Dallas, Texas 75201 1 M ! ' anny A. Sinkin 1
Mr.\ homer C. Schmidt Christic Institute l Manager .'. Nuclear Services 1394 North Capitol Street )
Texis Utilities Generating Comoany Washington, D.C. 2000?
Skyway Tower . ; 3 400 North Olive Street, Lib. fil Ms. Billie Pirner Garde Dallas, Texas 75201 < Citizens Clinic Director l Government Accountability Project l Mr. Robert E. Ballard, Jr. 1555 Connecticut Avenue, N.W. j Suite 202 1 Director of Projects Washington, D.C. 20009 Gibbs and Fill, Inc.
11 Pen Pihza s
New Yorh-New York 10001 David R. Pigott, Esq.
Orrick, Herrington A Sutcliffe T '\ 4 .600 Montgomery Street !
San Francisco, California Mr. R. 5. Koward' 94111 Westinghouse Electric Corporation P. O. Box 355 Anthony Z. Roisman, Esq.
,~
o PittsbuYgh, Pennsylvenia 1!230 Trial Lawyers for Public Justice
, .,'! 2000 P. Street, NW ,
Renea Hicks, Esc. Suita 611 )
Assistant Attorbey General Washington, D.C. 20036 Environmental M tection Division P. O. Box 12548, Capitol Station Nancy E. Wiegers Austin, Texas 78711 Spiegel'8 McDiermed 1350 New York Avenue, NW I Mrs. Juanita Ellis, President, Washington, D.C. 20005-4798 :
Citizens Association for Sound. Energy 1426 South Polk ^ s Ms. Billie Pirner Garde Dallas, Texas .75??4 Citizens Clinic Director i 4
s Government Accountability Project Ms. Nancy H. Williams '
104 East Wisconsin Avenue GGNA ADpleton, WI. 54915-8605 l'
101 California Street San Francisco, California 94111 de i
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9 3 Texas Utilities Electric Company Comanche Peak Electric Station Units 1 and 7 CC*
Resident inspector - Comanche Peak Joseph F. Fulbricht c/o U.S. Nuclear Regulatory Commission Fulbright & Jaworski P. 0. Box 1029 1301 McKinney Street ;
Granbury, Texas 76048 Houston, Texas 77010 j Mr. John W. Beck Vice President j Texas Utilities Generating Company !
Skyway Teser 400 N. Olive Street, LB#81 Dallas, Texas 1 75201 Mr. Jack Redding i
C/0 Datel Service Corp. ,
Texas Utilities Generating Company 7910 Woodmont Avenue, Ste. 208 ,
i Bethesda, Maryland 20814 William A. Burchette, Esc.
Counsel for Tex-La Electric Cooperative j of Texas Heron, Burchette, Puckert & Rothwell Suite 700 1025 Thomas Jefferson Street, NW Washington, D.C. 20007 GDS Associates, Inc.
PS25 Cumberland Parkway Suite 450 Atlanta, Georgia 30339 Administrative Judge Peter Bloch U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Elizabeth B. Johnson Administrative Judge Oak Ridge National Laboratory P. O. Box X, Building 3500 Oak Ridge, Tennessee 37835 Dr. Kenneth A. McCollom 1107 West Knapp Stillwater, Oklahoma 74075 Dr. Walter H. Jordan 881. Outer Drive Oak Ridge, Tennessee 37830
1 A LIST OF ATTENDEES 4 i
l NRC/TU ELECTRIC MEETING JUNE 2, 1987 i i
J Boatwright TV Electric l C. Grimes NRC/0SP R. Hegar TV Electric i
J. Hicks TU Electric l
A. Hussain TV Electric R. Jones NRC/NRR J. Lee Duke Power Co.
Sui-Sang Lo TU Electric
(
J. Lyons NRC/OSP 1 P. McKee NRC/OSP l
B. Rice TV Electric l H. Schierling NRC/0SP D. Woodlan TV Electric !
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TV ELECTRIC SGTR PRESENTATION JUNE 2, 1987 I. INTRODUCTION JACK HICKS SR. ENGINEER, LICENSING i
II. REACTOR ENGINEERING INTRODUCTION DR. AUSAF HUSAIN i DIRECTOR OF REACTOR 1 ENGINEERING 1
III. TRANSIENT ANALYSIS OVERVIEW BRENT RICE I SUPERVISOR, TRANSIENT ANALYSIS i
IV. TU ELECTRIC RETRAN-02 MODEL DR. sui-SANG Lo SR. ENGINEER, TRANSIENT ANALYSIS V. SGTR PLANT SPECIFIC ANALYSIS JAMES BOATWRIGHT STAFF ENGINEER, TRANSIENT ANALYSIS VI.
SUMMARY
DR. AUSAF NUSAIN r
n1UELECTRIC
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ENCLOSURE 3 TV Electric Viewgraphs i SGTR Analysis l NOTE: Proprietary Inforination Deletea i
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l OVERVIEW OF THE TRANSIENT ANALYSIS I SECTION OF TU ELECTRIC Presented l
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by j i
i Brent L. Rice !
7 l
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1 June 2,1987 i i
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n1UELECTRIC l
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TOPICS OF DISCUSSION
- Transient Analysis Section Responsibilities 1 l
- Personnei ,
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- Education
- Background / Experience q
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- Past RETRAN Activities Electric Power Research Institute (EPRI)
- Comanche Peak Analyses Performed To Date Overview Of Plans For Resolution Of SGTR lssue Schedule Summation
e f
/
TRANSIENT ANALYSIS SECTION RESPONSIBILITIES 4
- System Transient Analysis (Non-LOCA)
\
Reactor Thermal-Hydraulics i
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Transient Fuel Performance Analysis i
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Reactor Protection System Setpoint Methodology i
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-J
Y TRANSIENT ANALYSIS PERSONNEL l
1 Supervisor j
- 7 Engineers - ,
1 Summer Student i
Education 1 Ph.D. Nuclear Engineering ;
I 6 M.S. Nuclear Engineering 6 B.S. Nuclear Engineering l
1 M.S. Electrical Engineering 1 M.S. Physics 2 B.A. Physics and Mathematics l
I 1
4
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BACKGROUND. l 1
- 30 Years of Commercial, Nuclear Power, ' Engineering Experience l
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- 17 Years of RELAP and RETRAN Experience j i
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- Several Steam Generator Tube Rupture (SGTR) Analyses performed by Transient Analysis Personnel in Previous Positions
f PAST RETRAN ACTIVITIES AT TU ELECTRIC 1
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I EPRI ,
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Active Member of RETRAN-02 Users Group since 1982 )l I
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- Numerous Presentations of Analysis Results j l
1 Current Vice-Chairman of Users Group - James Boatwright (TU Electric) ,
Charter Member of RETRAN-03 Working Group (Formed in 1986) l l
Participated in Numerous RETRAN-02 and R ETR AN Workshops
l RETRAN ANALYSIS EXPERIENCE FOR COMANCHE PEAK i
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- Thermal-Hydraulic Analyses for CPSES Emer'gency Drill Scenarios (1983 to Present) l l
. . j
- Multi-event
{
Best Estimate l
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- Main Steamline Break with Faulted MSIV Bypass Valves (1984) ,
1 l
. j
- Licensing Analysis ,
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- Energy, Inc. Review of RETRAN Model (1986) l l
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Thermal-Hydraulic Cooldown Analysis for the Fire Protection Program Review (1986)
Quasi- Best Estimate ,
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1 RETRAN ANALYSIS EXPERIENCE FOR COMANCHE PEAK (continued)
Comanche Peak Simulator Benchmarks .(1986 to Present)
- Best Estimate Comanche Peak FSAR Benchmarks (1987)
- Complete Loss of Flow i
- Loss of Normal Feedwater
- Licensing Analysis
1 l
ANALYSES FOR j EMERGENCY DRILL SCENARIOS Reactor Coolant Pump Trip, Steamline Break, Steam Generator Tube Rupture ,
Feedline Break, Reactor Coolant Pump Trip, Steam Generator Tube Rupture Reactor Trip l
MSIV Closure, S/G Safety Valve Stuck Open, Steam Generator Tube Rupture, Loss of Offsite Power Control Rod Ejection, Reactor Coolant Pump Trip, Loss of )
1 ECCS l
Primary Coolant Leak, Power Ramp Preceding a LOCA 1
Reactor Trip at 48% Power, S/G Atmospheric Relief Valve l
Cooldown, RCS Leak of 75 gpm, One-Half ECCS Available l
l 9 l 1
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COMANCHE PEAK SIMULATOR BENCHMARKS i
1
- Steam Generator Tube Rupture l
Reactor Trip Partial Loss of Flow Complete Loss of Flow 4/4 MSIV Closures .,
l Loss of Normal Feedwater l
- Power Ramp (100% - 75% - 100%) !
Main Steamline Break
- Stuck Open Pressurizer PORV Turbine Trip From 10% Power
- Small Hot Leg Break !
/0 1
- f ,
SUMMARY
OF QUALIFICATIONS l
)
Well Educated Staff 1
Extensive RETRAN Analysis Experience Many T.ansients Simulated
- Best Estimate Analyses
- Licensing Analyses SGTR issue Presents Excellent Opportunity to Apply This Experience
//
i OVERVIEW OF PLANS FOR RESOLUTION OF SGTR ISSUE Member Of SGTR Subgroup Of Westinghouse Owners' Group (WOG) Since its inception NRC Has Accepted WOG Methodology i i
i SER Requires Evaluation Of Applicability Of WOG Methodology To Each Plant 1
TU Electric Has Determined That Comanche Peak - Specific Analysis Required I
i Analysis Will Be Performed in-House Utilizing WOG Methodology and RETRAN l2 j
e i SCHEDULE !
l All Open Licensing issues For CPSES, Unit 1 Are. Scheduled )'
To Be Resolved By March 1,1988.
- Safety Evaluation Report on WOG Submittal Requires that NTOL Plants Resolve the Plant-Specific SGTR issue prior to Receipt of an Operating License.
TU ELECTRIC Topica9 On Steam Generator Tube Rupture Will Be Submitted to the NRC by November 1,1987.
1 l
I Request that NRC Review of Topicals be Completed and SER issued by March 1,1988.
/3
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SUMMATION l i
Meeting Purpose ;
- Introduce Ourselves , ;
Discuss Qualifications !
- Present Progress To Date 1
RECEIVE FEEDBACK TO MINIMlZE IMPACT ON SCHEDULE l i
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i THE RETRAN-02 MODEL !
E-QB COMANCHE PEAK STEAM ELECTRIC STATION. UNIT 1 i i
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Presented 1
by Sui-Sang to June 2,1987 RnlELECTRIC '
/r
1 i
f TOPICS OF DISCUSSION 1
- 1. RETRAN Safety Evaluation Report (SER) i I
1
- 2. Comanche Peak RETRAN Model Design Control
- 3. RETRAN Best Estimate Base Model Description i
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L RETRAN-02 COMPUTER CODE
- Thermal-hydraulic code for analyzing nuclear reactor l system transients ]
1
- Developed under the sponsorship of the Electric Power q Research Institute (EPRI) 1
- Widely used in the nuclear utility indu~stry l
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- 1. 'RETRAN-02 SAFETY EVALUATION REPORT I 1
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1
- SER issued by the NRC on RETRAN-02/ MOD 002 l 1
- Error corrections to MOD 002 contained in MOD 003 .
3
- Error corrections to MOD 003 and updates contained in MOD 004- )
i
- Comanche Peak RETRAN model developed from MOD 004 !
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1 REPRESENTATIVE TRANSIENTS IN STANDARD REVIEW PLAN i
- Steam Generator Tube Rupture -
- Decrease in Feedwater Temperature l
- Steam System Piping Failures '
- Turbine Trip i
- Loss of Normal AC Power l 1
- Loss of Normal Feedwater Flow i
- Feedwater System Pipe Breaks i
- Total and Partial Loss of Forced Reactor Coolant Flow j
- Reactor Pump Shaft Seizure ( !
- Uncontrolled Control Rod Assembly Withdrawal at Power
- Boron Dilution at Power I
/9
I ,.y 4 .
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- 2. COMANCHE PEAK RETRAN MODEL DESIGN CONTROL
. e
- Satisfy the QA requirement outlined in USNRC Regulatory Guide 1.64, ANSI N05.2.11 and 10CF'R50,'ATppendix' B
)
I
- Outline the requirements for the performance and control of safety related decign activities
- The Comanche Peak RETRAN model developed in
, i ' compliance with the Design Control Procedures e ,
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i DESIGN VERIFICATION ,
1 Design Review l
Technical Review performed by Energy incorporated (EI)
Review Comments evaluated and properly implemented by TU Electric 1
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- 3. RETRAN BEST ESTIMATE BASE MODEL DESCRIPTION l
- Specific Features of Best Estimate Base Model '
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- Non-equilibrium Pressurizer Model ,
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- Pressurizer Pressure Control l l
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- Pressurizer Level Control l l
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- Rod Control System l
- Steam Generator Water Level Contol l 1
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- Steam Dump Control 1
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1 RETRAN BASE MODEL FOR COMANCHE PEAK UNIT 1
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SUMMARY
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- The RETRAN-02 Code accepted by the NRC for the non-LOCA transient applications I I
- Comanche Peak RETRAN best estimate base model developed and controlled under an approved QA program
- Technical Review performed by El
- The Steam Generator Tube Rupture licensing model derived from this base model s
I O 4 l
i COMANCHE PEAK STEAM ELECTRIC STATION STEAM GENERATOR TUBE RUPTURE i l
PLANT-SPECIFIC ANALYSIS 1
Presented I
by !
James Boatwright June 2,1987 n1UELECTRIC '
2T
1 TOPICS OF DISCUSSION j
SUMMARY
OF NRC and WOG ACTIVITIES CONCERNING THE STEAM GENERATOR TUBE RUPTURE ISSUE 1
PLANT-SPECIFIC REQUIREMENTS j 1
l TU ELECTRIC PLANS FOR RESOLUTION OF ISSUE q
TU ELECTRIC ACTIVITIES TO DATE l
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PROPRIETARY STATEMENT l
1 This presentation contains information proprietar3 to TU Electric and We s tin g h o u s e Electric Corporation; it is presented in l confidence and is to be used soleiy for the purpose for which it is furnished and returned upon request. This document and such information is not to be reproduced, transmitted, disclosed, or used otherwise in whole or in part without authorization of TU Electric and Westinghouse Electric Corporation.
1 i
SGTR HISTORY :
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Steam Generator Tube Rupture is classified as an i
ANS Condition IV event
]
- Several SGTRs have occurred in last few years !
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Current Design-Basis SGTR analysis is based on the ASSUMPTION that primary - secondary leakage is terminated within 30 minutes 4
SGTR events at Prairie Island & Ginna required significantly longer It was required that utilities re-analyze Design-Basis SGTR event to reconcile the above inconsistencies
i o 8 l
WESTINGHOUSE OWNERS GROUP i
SGTR Subgroup of WOG formed for generic resolution of SGTR issues Selected " Reference" plant with the least Margin to ;
Overfill l i
- Based on existing analyses l 1
Determined conservative times for the completion of the significant operator actions required for event I termination I
i Performed Sensitivity Studies on initial Conditions, Setpoints, Control Systems, etc.
29
WESTINGHOUSE OWNERS GROUP (Cont.)
- Evaluated Radiological Consequences .
Submitted Topical Reports to the NRC for Approval of Methodology
- WCAP 10698 .
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- Supplement 1 to WCAP 10698 I i
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i NRC ACTIONS )
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i NRC issued SER on Supplement 1 on Dec.17,1985 l
- NRC issued SER on WCAP 10698 on March 30,1987 )
l Accepted Generic Methodology I J
)
Identdied Plant-Specific Requirements 4
31
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PLANT- SPECIFIC REQUIREMENTS Evaluate Applicability of WOG Operate Actions Times for Plant-Specific, Design-Basis SGTR
- Evaluate Radiological Consequences using Plant-Specific Mass Releases i
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Determine Structural Adequacy of Water Filled Main Steamlines ,
Provide Equipment / Qualification List of Equipment Credited in ERGS and Design-Basis Scenario ,
- Evaluate Effects on the Margin to Overfiil due to Design Differences from the Reference Plant l
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e d TU ELECTRIC ACTION PLAN OPERATOR ACTION TIMES Bases for WOG Operator Action Times
- WOG Operator Action Times based on Emergency Response Guidelines (ERGS), Revision 1 !
- Action Times validated on SEABROOK and SNUPPS i simulators (Both 4-loop W plants very similar to 1 CPSES) l Comparison between CPSES & W ERGS l 1
- NO Significant impact on Operator Action Times
- CPSES & the CPSES ERGS are very similar to the Bases used to develop the WOG Operator Action Times
- The Conservative Operator Action Times Validated by the WOG are applicable to CPSES 33
TU ELECTRIC ACTION PLAN l l
EVALUATION OF RADIOLOGICAL CONSEQUENCES l 1
1 Conservative lodine Transport Model to be used 1
Conservative initial Concentrations used to Bound l 1
Proposed Tech Specs J Evaluated with Pre-Accident and Accident initiated lodine Spikes 1
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. o TU ELECTRIC ACTION PLAN ;
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STRUCTURAL ADEQUACY OF MAIN STEAMLINES UNDER WATER FILLED CONDITIONS 1
Analysis in Progress by Stone & Webster I
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j TU ELECTRIC ACTION PLAN .
I EQUIPMENT & EQUIPMENT QUALIFICATIONS i i
1
- Comprehensive List of Equipment Available to assist the Reactor Operators in SGTR event termination has been l prepared
- Equipment & Instrumentation Addressed in CPSES ERGS l
l List of Equipment REQUIRED for Termination of the Design-Basis SGTR event has been prepared Qualifications of Required Equipment are currently being Verified 36
l l
- 11) ELECTRIC ACTION PLAN i l
EFFECTS DUE TO DESIGN DIFFERENCES BETWEEN CPSES AND THE REFER?.NCE PLANT 1
CPSES IS 4-LOOP PLANT l
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- Larger RCS Volume {
l More heat storage capacity f
LARGER PRESSURIZER
- Slower depressurization rate CPSES S/G (and Steam) VOLUME GREATER ;
- More Margin to Overfill
)
SMALLER DIAMETER S/G U-TUBES
- Smaller Break Flow Rate GREATER CAPACITY ECCS
- Higher Equilibrium Pressure 37
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TU ELECTRIC ACTION PLAN '
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THERMAL-HYDRAULIC MODEL )
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Use RETRAN02 MOD 004 l
- Quantify effects due to Design Differences '
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- Determine Plant Specific Mass Releases RETRAN02 Applied to SGTR events by INPO & NSAC BEST-ESTIMATE Simulations of Ginna and Prairie Island Events 1
". ..t h e RETRAN code is an excellent thermal hydraulic j code for steam generator tube rupture analysis."
l (NSAC 77) 1 l
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CPSES - SPECIFIC MODEL 1
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l Derived from Design-Reviewed Best-Estimate Model )
" Beneficial" Control Systems inoperable Nodalization similar to that used by NSAC l
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, s CPSES - SPECIFIC ANALYSIS CASES ANALYZED l CONSISTENT WITH WOG METHODOLOGY: l
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NOMINAL BASE CASE i
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CONSERVATIVE BASE CASE l l
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DESIGN - BASIS CASE 4'l
,, r CPSES SPECIFIC MODEL BASE CASE MODELING ASSUMPTIONS i
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1 PROPRIETARY 1 l
1 CPSES SPECIFIC MODEL BASE CASE MODELING ASSUMPTIONS !
(Cont.)
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i CPSES - SGTR - NOMINAL BASE CASE 2.4 messumzen mssuac 2.2 -
1 2- l 1.8 -
1.6 -
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$! 1.2 -
Al PROPRIETARY 8 1-1 1
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1 0.4 -
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1750.0 - -
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l NOMINAL BASE CASE l
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DESIGN - BASIS CASE l l
. n ACTIVE FAILURE STUDIES 1
i l
Analyses in Progress 1
1 I
i To Be Evaluated to Determine Worst Single Active Failure l With Respect To:
- Margin to Overfill Radiological Consequences
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- Most Likely Different Failures i
4 Most Significant Possible Active Failures:
I AFW Flow Controller Fails Fully Open Failure of MSIV on Ruptured Loop to Close ARV on intact Loop Fails to Open
- ARV on Ruptured Loop Fails to Close 55
l i
l e 1 REVISED FSAR DESIGN - BASIS ANALYSIS i
i TWO CASES TO BE PRESENTED i
- Worst Single Active Failure With Respect To Margin To j I
Overfill Worst Single Active Failure With Respect To Radiological Consequences i
t i
'l 56 l
< , a a
SUMMARY
l 1
I RETRAN IS APPROPRIATE FOR SGTR ANALYSES l
j ANALYSIS METHODS CONSISTENT WITH APPROVED METHODOLOGY 1
CPSES - SGTR RESULTS COMPARE FAVORABLY WITH WOG RESULTS i
RESULTS OF ACTIVITIES TO DATE ARE VERY POSITIVE l
THE ACTIVE FAILURE STUDIES AND RADIOLOGICAL CONSEQUENCES CALCULATIONS HAVE YET TO BE COMPLETED CONFIDENT THAT MODELS, METHODS, AND PROGRAM ARE SUFFICIENT TO ASSURE RESOLUTION OF.SGTR ISSUE 57 l
l