ML20002C100

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Summary of 801203 Meeting W/Util Re Plans in Area of Core Reload Licensing for Coming Yr.Meeting Identified Two Topical Repts Germane to Plans Submitted to NRC But to Which NRC Appeared Oblivious.Viewgraphs Encl
ML20002C100
Person / Time
Issue date: 12/12/1980
From: Berggren J
Office of Nuclear Reactor Regulation
To:
Office of Nuclear Reactor Regulation
References
NUDOCS 8101090116
Download: ML20002C100 (22)
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,3, UNITED STATES NUCLEAR REGULATORY COMMISSION

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DEC 12 $80 VENDOR:

Virginia Electric and Power Company

SUBJECT:

LICENSING TOPICAL REPORTS RELATIVE TO VEPCO CORE RELOAD LICENSING On December 3,1980, the NRC staff met with representatives of VEPC0 to discuss VEPCO's plans in the area of core reload licensing in the coming year.

The VEPC0 representatives presented their plans using viewgraphs of the attached charts (Enclosure 1).

The meeting identified two licensing topical reports germaine to their plans and which had been submitted to NRC but to which NRC appeared oblivious. These are VEP-FRD-19 entitled "PDQ07 Discrete Model" submitted July 25,1976 and VEP-FRD-20 entitled "P0Q07 One Zone Model" submitted January 25, 1977. A copy of VEP-FRD 19 has been traced but no records of VEP-FRD-20 have been found and a request, for an additional copy has been made to VEPCO. These reports will be included in the NRC topical report review program.

Per VEPCO's recuest a follow-up technical meeting has been scheduled for 10:00 a.m. on January 28,198T in room P-ll4 in the Phillips Building, Bethesda, Maryland. The discussions will include: topical reports to be submitted in 1981, basic scope of NRC involvement and initial NRC technical feedback.

A list of the attendees is attached (Enclosure 2).

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._._ John S. Berggren, Project Manager

~ Standardization & Special Projects Branch Division of Licensing

Enclosures:

1.

Agenda 2.

List of Attendees cc: See attached list I

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ENCLOSURE 1 AGENDA I.

Nuclear Fuel Engineering at Yepco...............W. R. Benthall II. Reload Design and Analysis Capability...........M. L. Smith III. Reload Safety Analysis Capabili ty...............R. W. Cross IV. Vepco's Reload Analysis Program.................R. W. Cross V.

Suma ry......................................... W. R. B en tha l l VI. NRC Response - Discussion.

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VEPCO'S RELOAD ANALYSIS PROGRAM NRC MEETING OBJECTIVES Outline Vepco's Nuclear Fuel Engineering Capabilities e

Outline Vepco's Core Reload Analytical and Technical Resources e

Explain Vepco's Program For Complete, In-House Reload Design e

and Safety Analysis for 1981 License Submittals.

Solicit NRC Perspective on Vepco's Reload Licensing Prcgram e

Identify NRC/Vepco Teams and Resources For Review of Vepco's e

Reload Licensing Capability Establish Date For First Working Meeting to Define Details of the e

Scope and Schedule for NRC Involvement.

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VEPCO'S NUCLEAR ORGANIZATION Executive Vice President Power J.11. Ferguson Vice President Fuel Resources W. N. Thomas

  • Quality llanager Manager Assurance fluclear Operations Nuclear Technical Department

& !!aintenance Services Manager Nuclear Fuel L. M. Girvin f

I Director Director Director St tI s Nuclear Fuel Nuclear fuel Nuclear fuel Operations Engineering Procurement

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1 NUCLEAR FUEL ENGINEERING Director Nuclear Fuel 8

Engineering 5

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2 Nuclear Fuel Nuclear Dealgn Group AnkShsih*bYot5p i-Supervisor - M. L. Smith Supervisor - R. W. Cross i

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i NUCLEAR FUEL ENGINEERING RESPONSIBILITY Provide Design, Licensing and Safety Analyses For Nuclear Fuel Cores Provide System Safety Analyses To Support Core Operation 4

STAFF EXPERIENCE 16 Engineers Staff Size 1 Technician 70+ Man Years Experience 9

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I VEPCO RELOAD ANALYSIS PROGRAM OBJECTIVES e Development of Complete In-House Reload Core Design and Safety Analysis Capability.

e Utili:e In-House Reload Analysis Capability to Perform Complete Design and Licensing Scope For Relcad Licensing Submittals Stgrting In 1981.

e Provide In-House Resource For Plant Specific, Core Related Analytical Support of Nuclear Station Operations.

e Provide In-House Resource For Enhancing Training Of j

Station Personnel on Nuclear System Transients and Core Related Effects.

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ADVANTAGES OF UTILITY IN-HOUSE RELOAD ANALYSIS CAPABILITY Improved Communications Between Reload Designer, Safety Analysts e

and Plant Operator Which Will Lead to Improved Safety Evaluations e Better Plant Operational Support As A Result of Plant Specific Evaluations ImprovedResponsivenesstoNRginResolvingLicensingandSafety e

Issues e Provide More Optimum Core Designs For Improved Availability and Fuel Utilization o Provide Flexibility to Better Support Plant Operations and Schedular Changes e Enhance Operator Understanding of System Transients Through Train-ing Emphasizing Core Related Effects 9

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NUCLEAR FUEL CESIGN GRCUP RESPONSIBILITIES e Develop Reload Cycle Designs Determine Number of Fuel Assemblies and Enrichment Based on Reload Cycle Lifetime Requirements Develop Loading Pattern Which Will Meet Peaking Factor and Other Constraints.

e Calculate Reactor Physics Input to Safety Analysis Basic Core Reactivity Parameters Core Peaking Factors,

Transient Specific Parameters Compare Reload Cycle Values to Values Assumed in FSAR or Subsequent Reference Safety Analysis e Provide Operational Support Analysis Calculate Rod Worth, Reactivity and Power Distributions For Sisetup Physics Tests Calculation of Input For Rod Worth Measurement With Rod l

Swap Technique Provide Calculations For Operator Curve Book Fr eparation l

and Core Follow Support Reactor Operations by Evaluating Any Anomalies e

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NUCLEAR FUEL DESIGN GROUP STAFF EXPERIENCE i

1 e Eight Engineers and One Technician l

- Average of over Five Years Related Experience With a Range From One to Ten Years e Reload Cycle Designs Developed

- Surry Unit 1, Cycles 5 and 6 Surry Unit 2, Cycle 5

- North Anna Unit 1, Cycles 2 and 3

- Verified Vendor Designs For Earlier Cycles e

e Development of Rod Swap Analysis Techniques i

- Developed Analy:is Methodology

- Provided Calculations For Both Rod Swap And Standard Technique for North Anna 1 Cycle 2 Comparison

- Provided Rod Swap Predictions For Surry Unit 2, Cycle 5 e Development of Improved Cycle Designs

- Developed 18 Month Reload Cycles

- Developed Low Leakage Core Loading Patterns

- Participating in DOE Sponscred LWR Fuel UtiTization Improvement Program i

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NUCLEAR FUEL ENGINEERING RESOURCES AVAILABLE i

e Cornputer Facilities Access to Two IBM 3033 Cc=puters At Vepco r

Access to Two CDC 7600 Cc=puters At Westinghouse e Consultants I

NAI Energy Incorporated Westinghouse i

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NUCLEAR FUEL DESIGN GRCUP REACTOR PHYSICS MCCELS USED e PDQ7 Discrete Model Two Dimensional, Two Energy Group Model With Thermal-Hydraulic Feedback Represents each fuel pin in Quarter, Half, Full Core Geometry Topical submitted for NRC Review in 1976 e PDQ7 One Zone Model Two Dimensional Two Energy Group Model With Thermal-Hydraulic Feedback Uses 6x6 Mesh For Each Fuel Assembly In Quarter Half or Full Core Geometry Topical Submitted for NRC Review In 1977 s FLAME 3 Nodal Model One Energy Group, Three Dimensional Model with Thermal-Hydraulic Feedback One Node Per Assembly Radially With 32 Axial Nodes in Quarter, Half and Full Core Geometry Topical Submitted For NRC Review in 1978 e STANDARD WESTINGHOUSE CODES Access to the Westinghcuse Codes Used Fce Design And Licensing Is Expected In 1st Quarter,1981 Codes to Be Accessed Include TORTOISE, PALADON, APPOLO, and ARK e

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.__DEVELOPMENLOF RELOAD CYCLE _. DESIGN _!Klu0l>S

..DEVELOPh1ENT OF_EDQ3 O

DiSCliETE MODEL

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FLAME 3 MODEL s

DEVFLOPMEl'lT GF METI-ICDCLOGY FOR

.EVAljJATION OE HEL OAD. - O CYCLE SAFETY AtlALYSIS lilPUT USING vel >CO MODELS

  • DEVELOPMENT.CF.. l-D._ _ Q AXIAL MODEL

. RELOAD _CYCLELORE_ LOADING _ PAT _TElsN.DEVE!.OPMFf1T,

SUPPORT OF STARTUI' TESTING AllD CCliE Ol'LisATIGil N4 75 76 7

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1 NUCLEAR FUEL SAFETY ANALYSIS GROUP RESPONSIBILITIES -

Perform incore and system safety analysis to support core operation training, and licensing of Vepco nuclear units.

Develop / review safety evaluation reports and coordinate the prepara-tion of the safety analysis necessary to support core related j

licensing.

Develop safety analysis models and strategies.

Monitor and perform qualitative engineering and/or safety evalua-tion for all generic fuel rela *ed areas that may impact incore fuel design and performance or core safety.

Review and evaluate the impact pn incore fuel performance and safety resulting from a vendor dispositicn of fuel or fuel comconents which deviate frem the design or manufacturing criteria.

Monitor applicable research and development efforts.

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NUCLEAR FUEL SAFETY ANALYSIS GRCUP STAFF EXPERIENCE

  • Six Engineers, of which 5 have minimum experience level of 4 years.

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  • All personnel have a minimum of one graduate (:2:ter's) degree
  • Prior work experience includes vendor (Bettis, KAPL) and consultant (BNWL)
  • Three Engineers have been heavily involved in the utility working group effort to benchmark tne RETRAN code and one Engineer has participated in the development of the COBRA code.
  • Group members have presented 8 capers at professional society meetings and authored or coauthored 8 additional published reports.

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NUCLEAR FUEL SAFETY ANALYSIS GROUP REACTOR SAFETY MODELS RETRAN

- Generalized, variable geometry thermal-hyoraulics code.

- Uses one-dimensional, homogeneous, equilibrium form of the conservation equations.

- Includes point kinetics model with reactivity feedback, two surface heat transfer, nca-equilibrium pressuri:er model, and general control system models.

- Topical submittal for NRC review expected in early 1981 COSRA-IIIC/MIT

- Detailed thermal / hydraulic analysis code using a sub-channel forumulation

- Perform DNB analysis in PWR core using single-pass approach.

- Topical submitted for NRC Review in 1979 STANDARD WESTINGHOUSE CODES

- Access to the Westinghouse Codes used for design and licensing is planned for 1st Quarter 1981.

- Codes to be accessed include LOFTRAN, THINC-III, FACTRAN, TWINKLE, WIT-6.

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PROPOSED 1981 SCHEDULE

,RELDAD SAFEIY #{ALYSIS INPUT TOPICAL 00TDRIE R9' SD11TIAL -

_, SYSTEM TRNISifNT #4ALYSIS TOPICAL SI)RRY 2, CYCE 6 l

,PBIGIMRK REIRNi 10 CAPABILITY 4.

j DD/Q.0P #6 iU10t%RK 1-D AXIAL PilVSICS MODEL i

VEPC0/flRC KETINGS

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NRC/VEPC0 AS REWIRW TEGlilCAL W EiltE

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1981 - PROPOSED ACTION ptAN 1.

Initial Meeting Review Vecco reload capab'lity development plan Develop general guidelines for interaction Meet parties involved and appoint point of contact for respective organizations 2.

Technical Meeting - Week of January 26, 1981 Discuss content of topical reports to be submitted in 1981 Determine basic scope of hRC in~volvement required Review initial NRC technical feedback 3.

Submit topical reports - On or about March 1,1981 Non-LOCA System Transient Analysis (RETRAN) Topical Reload Safety Analysis (RSAC) Topical 4.

Technical Review Meeting - Week of April 13, 1981 NRC provide initial informal ccmments Clarify NRC questions 5.

Schedule further meetings as required to resolve informally as many comments as possible.

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e VEPCO'S RELCAO ANALYSIS PROGRAM XEY POINTS e Process For Resolving Most NRC Questions Would Be Expedited Through Technical Review Meetings vs. Formal NRC

' Questions.

The Codas Vepco Will Be Using Will Also Be Utilized By Other e

Utilities (RETRAN, CCERA, PDQ)

Vepco Can Utilize Westinghousar Codes For a Conveniant Benchmarking e

Tool TMI Review Cor:mittees Findings Encourage Utility Development of 4

Safety Analysis Capability Objectives and Benefits of In-House, Utility Reload Capability Are e

Significant - Utility and NRC Both Benefit NRC Should Encourage Utility Programs By Establishing Priority and e

Resources to Enable Timely Review-

ENCLOSURE 2 o

LIST OF ATTENDEES MEETING WITH VEPCO/NRC DECEMBER 3.1980 ORGANIZATION _

NAME ORB #3/DL/NRR L. B. Engle ORB #1/DL/NRR D. Neighbors SSPB/DL/NRR J. S. Berggren CPB/DSI/NRR L. E. Phillips CPB/DSI/NRR (Fuels Section) )

R. O. Meyer CPB/DSI/NRR (Physics Section W. L. Brooks CPB/DSI/NRR S. C. Gupta

?.SB/DSI/NRR F. Orr VEPCO Marvin L. Smith VEPCO Roger W. Cross VEPCO W. R. Benthall VEPCO R. M. Berryman e

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MEETING MINUTES 7ISTRIBUTION STANDARDIZATION AND SPECIAL PROJECTS BRANCH DIVISION OF LICENSING v c.

L. B. Engle D. Neighbors J. S. Berggren L. E. Phfilips 7

R. O. Meyer W. L. Brooks S. C. Gupta F. Orr Marvin L. Smith

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Roger W. Cross W. R. Benthall 3

2 R. M. Berryman

'4. Johnston T. Speis A. Dromerick D. Fieno B. Sheron G. Mazetis l

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