ML20247K067

Summary of 890518 Meeting W/Util in Rockville,Md to Discuss Reload Design & Planned Topical Repts Re Rod Exchange Methodology,Core Thermal Hydraulics & Transient Analysis. Agenda & List of Attendees Encl

ML20247K067
Person / Time
Site:	Wolf Creek
Issue date:	05/14/1989
From:	Pickett D Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8906010198
Download: ML20247K067 (28)

Text

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- May 14, 1989 {

Docket No. 50-482 LICENSEE: Wolf Creek Nuclear Operating Corporation FACILITY: Wolf Creek Generating Station

SUMMARY

OF MEETING WITH WOLF CREEK NUCLEAR OPERATING CORPORATION TO DISCUSS RELOAD DESIGN The subject meeting was held on May 18, 1989 in the NRC offices in Rockville, Maryland. A list of attendees is included in Enclosure 1. Enclosure 2 is a copy of the licensee's presentation.

The meeting was held at the licensee's request to discuss their plans for submitting topical reports for review and approval by the staff so that the licensee could perform their own core reload analysis. Current reload analysis work is being performed by outside contractors.

The licensee presented their design goals and objectives of acquiring full reload design capability (with the exception of LOCA analysis) in place for Cycle 7 start-up which is currently planned for June 1993. They discussed their organizational structure, capabilities and planned methodologies.

Topical reports in the areas of Rod Exchange Methodology, Core Thermal Hydraulics, Transient Analysis, Core Design, and Reload Safety Evaluation are scheduled for submittal. The proposed schedule calls for the topical reports to be submitted between June 1990 and December 1991. This will allow approximately one and one-half years from the final topical report submittal to the anticipated beginning of Cycle 7.

The staff discussed the proposed methodologies, computer codes and schedule with the itcensee. In p6rticular the staff discussed areas to be emphasized in the topical reports and noted weaknesses in similar submittals by other licensees. In summary, the staff felt that the proposal schedule was reasonable. The staff also indicated that additional meetings with the l licensee's technical staff may be both desirable and beneficial during the review process.

/s/

Douglas V. Pickett, Project Manager Project Directorate - IV 1 8906010198 890514 Division of Reactor Projects - III, l

PDR ADDCK 05000482 P PNV IV, V and Special Projects Office of Nuclear Reactor Regulation

Enclosures:

As stated l

cc w/ enclosures:

See next page DISTRIBUTION

, Docket file NRC PCR Local PDR J. Snfezek PD4 Reading F. Hebdon D. Pickett OGC-Rockville E. Jordan B. Grimes W. Hod R. Jones ACRS (10) R. Martin (Region IV)ges PD4 Plant File DOCUMENT NAME: WOLF CREEK D PICKETT 5/19 PD4/LA M PD4/PM DPickett[:cs p FHebdbnd PD4/Dd . i PNoonan i ,

05g89 05/p3/89 05/14/89

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[ .. 'May 14, 1989

.- Docket No. 50-482 LICENSEE: Wolf Creek Nuclear Operating Corporation FACILITY: Wolf Creek Generating Station

SUMMARY

OF MEETING WITH WOLF CREEK NUCLEAR OPERATING CDPe0 RATION TO DISCUSS RELOAD DESIGN The subject meeting was held on May 18, 1989 in the NRC offices in Rockville, Ma ryland. A list of atti:adees is included in Enclosure 1. Enclosure 2 is a copy of the licensee's presentation.

The meeting was held at the licensee's request to discuss their plans for submitting topical reports fc review and approval by the staff so that the licensee could perform their own core reload anelysis. Current reload analysis work is being performed by outside contractors.

The licensee presented their design goals and objectives of acquiring full reload design cbpability (with the exception of LOCA enalysis) in place for Cycle 7 start-up which is currently planned for June 1993. They discussed their organizational structure, capabilities and planned methodologies.

Topical reports in the areas of Rod Exchange Methodology, Core Thermal Hydraulics, Transient Analysis, Core Design, and Reload Safety Evaluation are scheduled for submittal. The proposed schedule calls for the topical reports to be submitted between June 1990 and December 1991. This will allow approximately one and one-half years from the final topical report submittal to the anticipated beginning of Cycle 7.

The staff discussed the proposed methodologies, computer codes and schedule with the licensee. In particular the staff discussed areas to be emphar,ized in the topical reports and noter weaknesses in similar submittals by other licensees. In sumary, the staff felt that the proposal schedule was reasonable. The staff also indicated that additions 1 meetings with the licensee's technical staff may be both desirable and beneficial during the  !

review process. .

/s/

Douglas V. Pickett, Project Manager Project Directorate - IV Division of Reactor Projects - III, IV, Y and Special Projects Office of Nuclear Reactor Regulation

Enclosures:

As stated cc w/ enclosures:

See next rage DISTRIBUTION Docket File NRC PDR Local PDR J. Sniezek PD4 Reading F. Hebdon D. Pickett OGC-Rockville E. Jordan B. Grimes W. Hod R. Jones ACRS(10) R. Martin (Region IV)ges PD4 Plant File DOCUMENT NAME: WOLF CREEK D PICKETT 5/19 PD4/LA M PD4/PH PD4/Dd PNoonan I DPickett:cs Q FHebden 05g89 05/$/89 05/;4/89

. f[ Ne,' i UNITED STATES

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""n NUCLEAR REGULATORY COMMISSION

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g WASH 1NGTON, D. C. 20555 May 24, 1989

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l Docket No. 5C-482  !

LICENSEE: Wolf Creek Nuclear Operating Corporation FACILITY: Wolf Creek Generatirg Statior

SUMMARY

OF MEETING WITH WOLF CREEK NUCLEAR OPERATING CORPORATION TO DISCUSS RELOAD DESIGN The subject meeting was held on May 18, 1989 in the NRC offices in Rockville, Ma ryland. A list of attendees is included in Enclosure 1. Enclosure 2 is a copy of the licensee's presentation.

The meeting was held at the licensee's request to discuss their plans for submitting topical reports for review and apprcval by the staff so that the licensee could perforn their own core reload analysis. Current reload analysis work is being performed by outside contractors.

The licensee presented thcir cesign goals and objectives of acquiring full relcad design capebility (with the exception of LOCA analysis) in place for Cycle 7 start-up which is currently plarr.ed for Jur,e 1993. They discussed their organizational structure, capabilities and planned methodologies.

Topical reports in the areas of Rod Exchange Methodology, Core Thermal hydraulics, Transient Analysis, Core Design, and Reload Safety Evaluation are scheduled for submittal. The proposed schedule calls for the topical reports to be submitted between June 1990 and December 1991. This will allow approximately one and one-half years from the final tcpical report submittel to the anticipated beginning of Cycle 7.

The staff discussed the proposed methodologies, computer codes and schedult with the licensee. In particular the staff discussed areas to be emphasized

• in the topical reports and noted weaknesses in similar submittels ty other licensees. In summary, the staff felt that the proposal schedule was .

reasonable. The staff also indicated that additional meetings with the licensee's technical staff may be both desirable and beneficial during the review process.

T4 a &

(;,y Douglas V. Pickett, Project Manager Project Directorate - IV Division of Reactor Projects - III, IV, Y and Special Projects Office of Nuclear Reactor Regulation

Enclosures:

As stated l

cc w/ enclosures:

See next page

- 4 4

Mr. Bart D. Withers

~

Wolf Creek Generating Station Wolf Creek Nuclear Operating Corporation Unit No. I cc:.

Jay Silberg, Esq. Mr. Gerald Allen <

Shaw, Pittman, Potts & Trowbridge Public Health Physicist 1800 N Street, NW Bureau of Air Quality & Radiation Washington, D.C. 20036 Control Division of Environment Chris R. Rogers, P.E. Kansas-Department of Health-Manager,, Electric Department and Environment Public Service Commission Forbes Field Building 321 P. O. Box 360 Topeka,. Kansas 66620

' Jefferson City, Missouri 65102 Mr. Gary Boyer, Plant Manager Regional Administrator, Region III Wolf Creek Nuclear Operating Corp.

U.S. Nuclear Regulatory Comission P. O. Box 411 799 Roosevelt Road Burlington, Kansas 66839 Glen Ellyn, Illinois 60137

, Regional Administrator, Region IV Senior Resident- Inspector / Wolf Creek U S. Nuclear Regulatory Comission c/o U. S. Nuclear Regulatory Comission Office of Executive Director P. O. Box 311 for Operations Burlington, Kansas 66839 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Mr. Robert-Elliot, Chief Engineer Utilities Division Mr. Otto Maynard, Manager Licensing Kansas Corporation Comission Wolf Creek Nuclear Operating Corp.

4th Floor - State Office Building P. O. Box 411 Topeka, Kansas 66612-1571 Burlington, Kansas 66839 Office of the Governor

' State of Kansas Topeka, Kansas 66612 Attorney General 1st Floor - The Statehouse Topeka, Kansas 66612 Chairman, Coffey County Comission

-Coffey County Courthouse Burlington, Kansas 66839 i

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Enclosure 1,

' Meeting Attendees Wolf Creek Nuclear Operating Corporation f

Reload Design Meeting Hay 18, 1989 NRC WCNOC Wayne Hodges- Robert Hagan.

Robert Jones: Terry Garrett'.

' Douglas Pickett Otto Maynard 4

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Enclosure 2 RELOAD DESIGN MEETING BETWEEN NRR AND WCNOC f

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MAY 18,1989 Wi$LF CREEK NUCLEAR OPERATING CORPORATION

NOLF CREE NUCLEAR OPERATING (IRPCFATICN Plans for the Develognant, Review, Appzwal, and Application of Reload Design Methods at Wolf Creek Generating Station p m unted by otto Maynard Manager Regulatory services Dr. Robert Hagan Manager pelaar Services

. Terry Garrett Manager Nuclear Safety Analysis May 18, 1989 i

M Creek thr laar nn.-e-4,v, r. _ .. _: :nn REEMD IEEEEEE REC natn'I'tr.

JGIN A 1.0 Opening Remarks Manager Regulatory Servi s/

otto Maynard o Purpose 1

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2.0 Reload Design Goal Manager Nuclear Services /

Dr. Robert Hagan o @jectives o organization Structure / )

Raspensibilities j

3.0 Overview of N30C Methods W Nuclear Safety Analysis /

W Garrett  ;

1 o Safety Analysis Transient Analysis BNFC Methodology REMN, REAPS Codes IOCA - Westinghouse / WCNOC Interfaces o Reload Safety Core s k rmal Hy4="14r=

BIEC Methods BNCW CHF Correlation VIPRE Code Ralewi Safety Evaluation

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o Core Design 1

4.0 7bpicals l

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6.0 NRC Faar'harir l

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Meetirn Purcose o Describe the wolf creek objectives for Reload Design o Discuss Rod Exchange Metberblev;y o Describe - Nuclear Safety Analysis Secticn Structure,

- Methodology and Code Package to be used o Describe 'Ibpical Report Contant, and

- Schedule for Topical Snhniasion o Cbtain NRC - Assessment of Reisw and Approval Schedule 0

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Reload Design anala and Objectives Chganizational Structure and Respannihilities Dr. Robert Hagan Mam melaar services 1

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, Goals and (biectives l

1 o Present a 'Ibpical Report on Rod Exchange Methodology l

o Obtain NRC Approval of the Rod Exchange Methodology o Present Reload Design Topical % M in.the Areas of Core 'Jhanaal Hydraulics Transient Analysis Core Design Reload Safety Evaluation o Obtain lac hwal of the Reload Design Methodology I

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l-Nblf Creek Plans for Rod Exchance Mlunr -

o Rod Exchange Methodology for Startup Cycle 6.(r+ h r 1991) o '1bpical Snhniasion in July 1990 (1.0 Year Review Time)

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, -c Nblf Creek Plans for Reload Desian

o. Full Reload Design Capability (except IOCA) o In Place for Cycle 7 Startup (June 1993) o Allows for a 1.0 Year Review Period from Time of Last W ie=1 Submission l

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1 Wolf Creek Oreanizaticn / Structure L

l Manager Nuclear Services  !

Dr. Robert Hagan l

I I Manager Manager Nuclear Safsty Analysis Nuclear Fuel Terry Garrett Steven Clark i I Safety Analysis Fuel Ccntracts t

Dr. Jin-Shou Haou, Imad Engineur Jim Samis, Supervisor Steven Sorrell, Engineer III Ron Rather, Enginaar III-Srikant Mehta, Engineer II Open Position, Engineer I Mike Carroll, Engineer II Dao Nguyen, Engineer I Open Position, Engi - III Reload Safety Fuel Perfe m ance Glenn Neises, Engineer II Scotty F&gnmai, Technical William Kennamore, Engineer I Staff Enginaar Matt Morris, Engineer III Core Design con Iong, Isad Eng4m Rod Kliewer, Engi w III Jeff Blair, Engineer III

, R114ntt Jackson, Enciw II Open Position, Eng'_ mar III Risk Assessment Mike Hall, Senior Engineer Vem Inckert, Senior Engineer

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Otmlifications of Mrl=e Safety Analysis Fuocauel -

i o Nuclear Safety Analysis has 16 Professionals (2 openings) 1 o Experience Ccznbines to 123 Cunnlative Years 1 i

o Experience Breakdom 6 2 10 Years of Nuclear Experience 10 2 5 Years of Nuclear W anoe I 14 2 2 Years of Nuclear Experience 4

2 Managers 37 n =ilative Years of Nuclear j W *'De  !

2 Imad Engineers 26 nmilative Years of Nuclear i W *"" i' 11 Staff 60 nmilative Years of Mel=*

W *'De j o Om14fications 1 Ph.D/NE 1 Ph.D/ME 1 Ms/ME 1 NE/NE 11 as/NE  !

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. Reload Design Methodology

% Gazzett Manager a r1 = r safety analysis 8.

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!! Responsibilities

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Nuclear Safety Analysis i

Safety Analysis:

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Transient Analysis (REIRAN, REUS5)  !

Non-IOCA Transients SGIR Analysis

'nx:hnical Specificaticrt Changes 10CFR50.59 Evaluations PRA Support (MAAP Code, )*vhilar Wir%t Analysis Fivmu)

Mrvialing severe Wiriarit Behavior Radiological Cornequence Analysis for Design Basis Arv'iriants Containment Analysis for Design Basis Wiriartts ~

Technical Support to Operations and Other Engineering Groups IOCA Cognizance Reactor Protection System Setpoint thcartainty Analysis Reload Safety:

Core 'Ihemal Hydraulic Analysis (VIPRE code)

Reload Safety Analysis Checklist Reload Safety Evaluations Todmif al Specification Changes 10CER50.59 Evaluations Core Operational / Safety Limit Analysis Transient D e Analysis Evaluaticrt of Reload sensitive transients Core Desian Caze Design Physics Parameters Maneuvering Aralysis Plant Operations Package Static Reactivity Chapter 15 events ROCA M4rden, Inadvertent r,witng and Operation of a

! Fuel AsLambly in an T r ,+r Position Safety Evaluations Technical Specification Changes Fuel Rod nanign

• Rod mirr+umga calilatims Risk Assesam mt:

Pw*=hiliatic Risk Assessilent IPE Generic Istter

• Planned at this time

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. 1 Reload Desian Methodolcxpr I o NRC Approved BGC Methods were Purchased to j Enable a Technology Transfer o HGC Methods will be Used 2 roughout the Reload Process I

o mis will Allow for NRC Review of PIw-Approved Methodology O

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Transient Analysis 1

o Use BREC Methodology o REIRAN-02 Cmputer Code (NSSS 'lhermal Hydraulics) - EPRI

, i o REIAP5 Coquter Code (NSSS 'Ihemal Hydraulics) - BNK: Version 1

o Licensing Analysis and Evaluation itxial of USAR Chapter 15 Analyses Ioss of Flow Iocked Rotor Turbine Trip Rod WitMr=ml at Power W

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Core 'Ihemal Hydraulics o Use INGC Methodology o VIPRE Cmputer Code - EPRI 1

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o Una NRC Appa m!d BHCMV CHF Correlation .i o Use BHEC Statistical Core Design Procedure (SCD)

I o IEEB Transient Analysis 4 l

'o Core Operating Limit Analysis

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Core Desian I o .Use BIGC Nothodology l

o 5G0 Cceputer Codes PEEL NILIF NO3LE FIAM PDQ o Cross-sections Generated by PEEL and NULIF o Reactivity Parameters Calculated Using lomrR o Power Distributions Analysis Perfnmarl Using FIAM o Pin-to-Box Factors and Isotcpics Calculated Using PDQ l

o Peaking Factors Deramined frca FIAM and PDQ Mnrials

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6/90 7/90 Trarisielt 'Ibgxi. cal '

12/90 core Design wir=1 12/91 12/91 i

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Core Ubarmal Hvriran1(cs Wr al i i

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1.0 ' Introduction 2.0 VIPRE Code Description i

3.0 NCDS Description 4.0 Core Nnrialing 5.0 NOGS Core Shermal Hydraulic Analysis 6.0 References I

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Rod Exchance Methodolocv Topicql 1 1.0 Introduc.ticn I

1 2.0 Rod Exchange Calculations

. 3.0 Rod Exchange Measurenant Procedure 4.0 Benchmark i

5.0 Verification of Westinghouse Core Design (Cycle 6) 6.0 Rasults and Conclusions h

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Transient 'Iboical 1.0 Introduction 2.0 Approach 3.0 Sumary and Conclusions 4.0 Transient Selection 5.0 Plant Initial Conditions 6.0 REEAP Plant Model Description 7.0 REIRAN Plant Model Description 8.0 USAR Mmtive Analyses 9.0 Key Safety Parameters for Reload Analysis 10.0 References l

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1.0 Introduction 2.0 Overview of the cale'lational Model 3.0 Model Verification & Reliability 4.0 Model Applications to Reactor Operations 5.0 Model Applications to Safety Evaluations 6.0. References Appendix A Statistical Methods for the Delamination ard M11r= tion w -ici-Appendix B >=414=7 M** Code Sunnary Description I

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Reload Safety Evaluati m M cal 1.0, Introduction 2.0 Reload Safety Evaluation Process 3.0 Core Design 4.0 Core bmnal }fydraulics 5.0 Transient Analysis i

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. n-1 IIDLE enEG BELSAs SESIts Stetsett

1. UCLfA8 SAF(77 ANAL 7$f$

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1989 Ney Jese L' . . July Aegest '

leptester Octeter sevester Decenter 1990 Joseery fetteery North April Ney legiesteg of Cycle $ Jose Core T/N Topical $stettlet to BRC Jely Red tacbesse Topical lebettt:1 to set Aegest Septester October Revoeter

!- Decenter Treasteet Teptcal Satelttet to Att 1991 Joseery Febreary Marc 6 April Ney Jose set Sf 8 Isssed for the Approval of the Core T/s Teptcal July sec $tt Issees for tee Aeproval of the Red lacheste Topical Aegast Septeeter Octeter Revoeter Septesteg of Cycle 6 Decenter Core Destge Testcal le6stttet to 80C R$t Testcal lobotstal to SRC i a NGC $ER !sseed for the Approval of 16e freestent Topical 1992 Jessary .

Fe6reary Aerc6 i

, April Ney Jose Jely Aegest Septoster Octeter Nevester 4 I

Decee6er NRC $lt Isseed for the Approval of the Core Dest $s Topics) ese the ASE Topical i

1993 Jeseary Febreery rc6 i April aey

. Begiesteg of Cycle F Jose i

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