ML20058B743

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Summary of 901010 Meeting W/Util & Consultants in Rockville, MD Re Licensee Plan to Pursue Reload Technology.List of Attendees Also Encl
ML20058B743
Person / Time
Site: Palo Verde  
Issue date: 10/19/1990
From: Peterson S
Office of Nuclear Reactor Regulation
To:
Office of Nuclear Reactor Regulation
References
NUDOCS 9010300410
Download: ML20058B743 (40)
v • d • e


Text

,

o October 19, 1990 Docket Nos. 50 528, 50 529 DISTRIBUTION ' Docket File and 50 530 TRirag1Ta---

JPartlow NRC & LPDRs MVirgilio DCrutchfield JDyer LICENSEE: Arizona Public Service Company SPeterson CTrammell OGC EJordan FACILITY:

Palo Verde Nuclear Generating Station ACRS(10)

MSlosson, 17G21 MCarusco LKopp PD5 rf

SUBJECT:

MEETING

SUMMARY

. RELOAD TECHNOLOGY TRANSFER On October 10, 1990 the staff met in Rockville, Maryland with representatives ofandconsultantsforArizonaPublicServiceCompany(APS)todiscussthe licensee's plan to pursue licensing of reload engineering capability for Palo Verde Nuclear Generating Station. A list of attendees is provided in.

APS had previousl 1990, to discuss the Combustion Engineering (CE)y met with the staff on May 31, load technology designed program to provide re to APS.

APS focused their discussion on the schedule and scope of their program to obtain reload technology from CE.

The purpose of the meeting was to obtain staff feedback and discuss future actions. APS provided an overview of their proposed topical report which will describe their reload analysis

)rocess and the independent results of CE and APS analyses.

APS stated that t1ey currently plan on submitting a request to license their reload capability by the second quarter of 1991. APS informed the staff that it will be seeking NRC approval of their reload design capability by the end of the first quarter of 1992 and that CE methodology would be used. The presentation handouts from APS are provided as Enclosure 2.

The staff indicated that APS's general approach to obtain reload technology was appropriate. The staff em)hasized that APS should continue to receive timely revisions to the CE met 1odology and computer codes used in their reload analysis. The staff stated that clear cross references to previous submittals from CE, describing the methodology APS will use, should be included in APS's submittal.

0@alsiente Dy:

Sheri R. Peterson, Project Manager Project Directorate V 9010300410 901019 Division of Reactor Projects. Ill, PDR ADOCK 05000520 IV, Y and Special Projects P

PU Office of Nuclear Reactor Regulation

Enclosures:

1.

List of Attendees 2.

Handouts I

cc w/ enclosure:

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0 OU ICIAL REGUKU COPY Document lame:

PT MI6 5UMMARY

Mr. William F. Conway Palo Verde Arizona Public Service Company CC:

Arthur C. Gehr Esq.

Mr. Jack R. Newman Snell & Wilmer Newman & Holtzinger P.C.

3100 Valley Center 1615 L Street, NW, Suite 1000 Phoenix, Arizona 85073 Washington, DC 20036 James A. Boeletto, Esq.

Southern California Edison Company P. O. Box 800 Rosemead, California 91770 Senior Resident Inspector U.S. Nuclear Regulatory Commission HC-03 Box 293-NR Buckeye, Arizona 85326 Regional Administrator, Region V U. S. Nuclear Regulatory Commission 1450 Maria Lane Suite 210 Walnut Creek, California 94596 Mr. Charles B. Brinkman Washington Nuclear Operations Combustion Engineering, Inc.

i 12300 Twinbrook Parkway, Suite 330 Rockville, Maryland 20852 Mr. Charles Tedford, Director Arizona Radiation Regulatory Agency 4814 South 40 Street Phoenix, Arizona 85040 Chairman Maricopa County Board of Supervisors 111 South Third Avenue Phoenix, Arizona 85003 L

l (8) s I

ENCLOSURE 1 List of Attendees MEETING WITH ARIIDRK PUBLIC SERVICE COMPANY (APS)

RELOKD TECHNOLOGi TRAN5FER OCT5EtR 10.1990 NRC APS Sheri Peterson William F. Mcdonald Mark Caruso Lun-Chih(Larry)Hwang Larry Kopp Peri Shankar NilsBreckenridge(CE)

Matt Reid B

Richard A. Bernier I

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1

ENCLOSURE 2 ARIZONA PUBLIC SERVICE Palo Verde Nuclear Generating Station Presentation to NRC AELOAD 5

TRANSFER T E CH t4 0L OGY

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x PVN05 UCLEA FUEL MA AGEMENT October 10,1990 p 3 m

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ti PRESENTATION INTRODUCTION BACKGROUND l

REVIEW PROPOSED TOPICAL REPORT OUTLINE

SUMMARY

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NRC FEEDBACK / FUTURE ACTIONS i

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INTRODUCTION i

l APS PREVIOUSLY MET WITH NRC TO DISCUSS PLANS TO PURSUE i

LICENSING OF RELOAD ENGINEERING CAPABILITY l-APS PROPOSED FUTURE WORKING LEVEL MEETINGS TO REVIEW PROGRESS OF LICENSING REPORT WITH NRC l

l APS WILL PRESENT AN OUTLINE OF THE PROPOSED LICENSE l

REPORT OR ' TOPICAL REPORT ~

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APS REQUESTS NRC FEEDBACK ON THE PROPOSED FORMAT AND I

CONTENT OF TOPICAL REPORT i

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'N BACKGROUND REVIEW APS PROGRAM TO OBTAIN RELOAD TECHNOLOGY

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REVIEW APS SCHEDULE FOR APPLYING RELOAD

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TECHNOLOGY i

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APS PLAN TO OBTAIN RELOAD ENGINEERING CAPABILITY i

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OBTAIN QUALIFIED STAFF i

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TRAIN STAFF IN PREVIOUSLY LICENSED TECHNOLOGY i

DEMONSTRATE AND LICENSE CAPABILITY

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ARIZONA PUBLIC SERVICE 4

Palo Verde Nuclear Generating Station NUCLEAR FUEL MANAGEMENT ORGANIZATION 1

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DEVELOPMENT OF CAPABILITY i

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OBJECTIVES:

ACQUISITION OF QUALIFIED COMPUTER CODES f

ACQUISITION OF QUALIFIED METHODS FOR SPECIFIC EVENTS ANALYSIS i

QUALIFICATION OF STAFF CAPABILITY j

HOW:

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CE/APS RELOAD TECHNOLOGY TRANSFER i

PROGRAM l

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___.__i

s Reload Technology Transfer i

Program Overview i

Option in i986 APS-CE Fuel Fabrication Contract Scope All Reload Engineering Technology Except O

LOCA O

Fuel Mechanical Design O

Fuel Fabrication t

Program i

Three Phases 1

/ Classroom Lecture

/ On the Job Training (OJT)

= Independent Analysis s

Goal 1

1 l

NRC approval Ist quarter 1992 I

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APS-CE Reload Technology Transfer Program Status 4

I 1

Phase 1

- Classroom Training 10/87 through 3/90 6000 Engineer-hours Phase 2 - On the Job Training 1/88 through 12/90 17000 Engineer-hours i

Phase 3 - Independent Analysis i

4/90 through 6/91 12000 Engineer-hours APS - Unit 3 cycle 3 (late cycle 2 shutdown)

CE

- Unit 3 cycle 3 (early cycle 2 shutdown) j i

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SG. '

i Schedule Name ! APS Reloed Technology Transfer Program 4

Responsible A2-of Date

1-Oct-90.7:00mm Schedule File : RTT T

I 87 88 89 90 91 92 93 Start End Jan 4

Task Name Date Date 2

4 3

2 2

2 4

Reloed Technology Tranefer 1-Oct-87 28-Jun-91

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j Phase 1 - Classroom Training 1-Oct-87 28-Feb-90 l

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~~~~~~~----------

i Phase 2 - On 1he Jc6 Training 4-Jan-88 28-Dec-90 Phase 3 - I;4ep.As.t Analysis 2-Apr-90 28-Jun-91 APS Prepare Topical Report 1-Apr-91 28-Jun-91 l

N WRC Review Topical Report 1-Jul-91 5-Mar-92 l

Earliest Possible APS Reload 5-Mar-92 12-May-93 l

Reload Analysis 5-Mar-92 23-Mar-93 l

Startup PVNGS Unit 2 Cycle 5 12-May-93 12-May-93 l

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" Detail Task

===== Suunnery Task

          • Baseline

- " (Progress)

- (Progress)

Conflict

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(Slack)

- (Slack)

Resource delay l

Progress shows Percent Achieved on Actuel a Milestone scale: 4 treeks per character i

TIME LINE cantt Chart Report, Strip 1 s

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APS RELOAD TOPICAL REPORT PROPOSED OUTLINE i

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I ABSTRACT a

II INTRODUCTION k

III OVERVIEW OF RELOAD ANALYSIS PROCESS IV COMPARISON OF APS INDEPENDENT ANALYSIS TO CE ANALYSIS V

SUMMARY

AND CONCLUSIONS t

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N APS RELOAD TOPICAL REPORT SECTION 3

. PURPOSE:

t PROVIDE AN OVERVIEW OF THE RELOAD PROCESS r

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ADDITIONALLY:

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1.

PROVIDE REFERENCE TO PREVIOUSLY LICENSED METHODOLOGY 2.-

IDENTIFY APS AND VENDOR SCOPE j

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APS RELOAD TOPICAL REPORT SECTION 3 (cont.)

i PURPOSE - PROVIDE OVERVIEW OF RELOAD PROCESS l

DESCRIBE RESPONSIBILITIES WITHIN APS SCOPE l

1.

NUCLEAR ANAIXSIS i

2.

CORE THERMAL HYDRAULIC ANALYSIS l

3.

FUEL PERFORMANCE ANALYSIS 4.

NON-LOCA TRANSIENT ANALYSIS l

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COLSS/CPCS ANALYSIS 5

4 i

L I

APS RELOAD TOPICAL REPORT t

E SECTION 4 I

i 4

I i

j PURPOSE:

t I

L PROVIDE COMPARISON OF PRINCIPAL RESULTS OF APS

' INDEPENDENT' RELOAD DESIGN TO CE DESIGN

  • APS UNIT 3 CYCLE 3' (LATE CYCLE 2 SHUTDOWN)
  • CE UNIT 3 CYCLE 3 (EARLY CYCLE 2 SHUTDOWN)

PROVIDE RESULTS OF CE REVIEW OF APS RELOAD i

CALCULATIONS 4

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SECTION 3 OVERVIEW -

NUCLEAR ANALYSIS I.

INPUTS TO NUCLEAR ANALYSIS i

A.

GROUNDRULES B.

FUEL MANAGEMENT SCOPING STUDIES l

C.

PREVIOUS CYCLE ASBUILT ANALYSIS i

D.

PREVIOUS CYCLE SHUTDOWN BURNUP i

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SECTION 3 i

OVERVIEW - NUCLEAR ANALYSIS II.

' NUCLEAR RELOAD ANALYSES A.

MODELS AND DEPLETIONS I

B.

PHYSICS DATA FOR NON-LOCA TRANSIENT ANALYSES C.

PHYSICS DATA FOR FUEL PERFORMANCE ANALYSIS l

D.

PHYSICS DATA FOR LOCA ANALYSIS 1

E.

PHYSICS DATA FOR THERMAL HYDRAULIC ANALYSIS I

I l

F.

PHYSICS DATA FOR FUEL MECFANICAL DESIGN ANALYSIS t

I G.

PHYSICS DATA FOR COLSS/CPCS ANALYSIS i

H.

ASBUILT ANALYSIS i

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SECTION 3 OVERVIEW - NUCLEAR ANALYSIS III.

NUCLEAR ANALYSIS OUTPUTS A.

PHYSICS DATA FOR SAFETY ANALYSIS B.

PHYSICS DATA FOR PLANT OPERATION i

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SECTION 4 NUCLEAR ANALYSIS RESULTS COMPARISONS DESIGN PARAMETER REF CE APS CRITICAL BORON CONCENTRATION BORON WORTH MTC 4

FTC KINETICS PARAMETERS

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SECTION 4 NUCLEAR ANALYSIS RESULTS COMPARISONS i

j DESIGN PARAMETER REF CE APS s

SCRAM REACTIVITY BALANCE - HOT FULL POWER STEAM LINE BREAK ANALYSIS l

WORTH OF ALL CEAS IN CORE i

i STUCK CEA ALLOWANCE k

WORTH OF ALL CEAS I

HIGHEST WORTH CEA STUCK OUT i

ANALYSIS SCRAM WORTH i

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SECTION 3 OVERVIEW - THERMAL HYDRAULIC ANALYSIS i

I.

INPUT TO THERMAL-HYDRAULIC ANALYSIS i

A.

PHYSICS POWER DISTRIBUTION DATA B.

FUEL PERFORMANCE ANALYSIS INPUTS C.

FUEL MECHANICAL DESIGN INPUTS D.

OTHER INPUTS b

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SECTION 3 OVERVIEW - THERMAL HYDRAULIC ANALYSIS II.

THERMAL HYDRAULICS RELOAD ANALYSES A.

' DEVELOP DESIGN THERMAL HYDRAULIC MODEL B.

STATISTICAL COMBINATION OF UNCERLpar rtES ANALYSIS (IE. VERIFY DNBR LIMIT)

C.

DNBR DATA FOR NON-LOCA TRANSIENT ANALYSIS 1

-IN SECTION 3 OVERVIEW - THERMAL HYDRAULIC ANALYSIS 9

III.

THERMAL HYDRAULIC RELOAD ANALYSES OUTPUTS f

i A.

THERMAL HYDRAULIC MODEL FOR DOWNSTREAM ANALYSES

- B.

DNBR LIMIT AND DNBR PROBABILITY DISTRIBUTION

. FUNCTION (PDF)

C.

'DNBR VS INTEGRATED RADIAL PEAKING FACTOR DATA FOR NON-LOCA TRANSIENT ANALYSIS 6

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SECTION 4 THERMAL HYDRAULIC ANALYSIS RESULTS COMPARISONS t

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DESCRIPTION. OF-DIFFERENCES IN T/H MODELS

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COMPARISON. OF DNBR I.IMIT VERIFICATION ANALYSES

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r SECTION 3

~ OVERVIEW -- FUEL PERFORMANCE ANALYSIS

'I.

INPUTS TO FUEL PERFORMANCE ANALYSIS A.

- PHYSICS POWER-DISTRIBUTION DATA l

B.

FUEL MECHANICAL ANALYSIS INPUTS 1.

FUEL ROD GEOMETRY 2.

FILL GAS PRESSURE 'AND COMPOSITION C.

THERMAL HYDRAULIC ANALYSIS INPUTS IE.

COOLANT CONDITIONS D.

LOCA ANALYSIS INPUTS IE.

LOCA LINEAR HEAT RATE LIMIT l

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SECTION 3

-OVERVIEW - FUEL PERFORMANCE ANALYSIS II.

FUEL PERFORMANCE RELOAD ANALYSES

- i IE.

STEADY STATE THERMAL AND MECHANICAL BEHAVIOR OF FUEL ROD I

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III.

FUEL PERFORMANCE RELOAD ANALYSES OUTPUTS A.

LOCA DATA 1.

FATES FILES FOR INITIALIZATION 2.

FUEL ROD INTERNAL PRESSURE f

i B.

NON-LOCA TRANSIENT ANALYSIS DATA i

l.-

FATES FILES FOR INITIALIZATION 2.

CORE AVERAGE GAP CONDUCTANCES t

C.

COLSS/CPCS ANALYSIS DATA 1.

FUEL CENTEliLINE MELT LIMIT l

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SECTION 4 FUEL PERFDRMANCE ANALYSIS RESULTS COMPARISONS e

DESIGN PARAMETER REF CE APS i

POWER TO CENTERLINE MELT I

MAXIMUM INTERNAL ROD PRESSURE t

MAXIMUM FUEL CENTERLINE TEMP.

l MAXIMUM FUEL AVERAGE TEMP.

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SECTION 3 OVERVIEW - NON LOCA TRANSIENT ANALYSIS I.

- INPUTS TO TRANSIENT. ANALYSES A.

PHYSICS ANALYSIS INPUTS 1

1.

REACTIVITY COEFFICIENTS 2.

ROD WORTHS 3.

DELKIED NEUTRON FRACTION (ETC.)

B.

THERMAL HYDRAULIC ANALYSIS INPUTS IE.

THERMAL HYDRAULIC MODELS C.

FUEL PERFORMANCE ANALYSIS INPUTS i

1.

FATES FILES t

2.

GAP CONDUCTANCES i

D.

OTHER INPUTS IE. GROUNDRULES i

rs SECTION 3 i

OVERVIEW - NON LOCA TRANSIENT ANALYSIS II.

SUMMARY

OF TRANSIENTS i

A.

INCREASE IN HEAT REMOVAL BY SECONDARY SYSTEMS B.

DECREASE IN HEAT REMOVAL BY SECONDARY SYSTEMS C.

DECREASE IN REACTOR COOLANT FLOW RATE t

.i D.

REACTIVITY AND POWER DISTRIBUTION ANOMALIES E.

' INCREASE IN RCS INVENTORY F.

DECREASE IN RCS INVENTORY f

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SECTION 3 OVERVIEW NON-LOCA TRANSIENT ANALYSIS III.

- NON LOCA ' & COLSS/CPCS TRANSIENT ANALYSES OUTPUTS A.

COLSS CONSTANTS 1.-

REQUIRED OVER POWER MARGIN

~ 2.

UNDER FLOW FRACTION B.

CPCS CONSTANTS 1.

INPUT TO RELOAD DATA BLOCK CONSTANTS i

ANALYSIS l

2.

INPUT TO ADDRESSABLE CONSTANTS ANALYSIS

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SECTION 4 NON-LOCA TRANSIENT ANALYSIS RESULTS COMPARISONS ANALYSIS PARAMETER REF CE APS CEA EJECTION CALCULATED %' FUEL FAILURE

t PRE TRIP. STEAM LINE BREAK i

CALCULATED % FUEL' FAILURE t

1 POST TRIP STEAM LINE BREAK MINIMUM SUBCRITICALITY t

SEIZED ROTOR / SHEARED SHAFT CALCULATED % FUEL FAILURE r.,

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SECTION 3 OVERVIEW - COLSS/CPCS ANALYSIS I.

INPUTS TO COLSS/CPCS ANALYSIS

-A.

PHYSICS ANALYSIS INPUTS I.

NEUTRONICS MODEL 2.

POWER DISTRIBUTION - DATA B.

TRANSIENT ANALYSIS INPUTS C.

THERMAL HYDRAULIC ANALYSIS INPUTS f

i I.

THERMAL HYDRAULIC MODEL l

2.

DNBR LIMIT i

D.

FUEL PERFORMANCE ANALYSIS INPUTS 1.

LOCA KW/FT LIMIT

2..

CENTERLINE FUEL MELT LIMIT 1

E.

OTHER INPUTS

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's SECTION 3 1

OVERVIEW - COLSS/CPCS ANALYSIS II.

t COLSS/CPCS RELOAD ANALYSES A.

COLSS CONSTANTS ANALYSIS 1.

DATABASE CONSTANTS ANALYSIS 2.

OVERALL UNCERTAINTY ANALYSIS 3.

ADDRESSABLE CONSTANTS ANALYSIS 4

. B.

CPCS CONSTANTS ANALYSIS l.

RELOAD DATA BLOCK CONSTANTS ANALYSIS 2.

OVERALL UNCERTAINTY ANALYSIS 3.

ADDRESSABLE CONSTANTS ANALYSIS C.

COLSS AND CPCS RELATED TRANSIENT ANALYSIS i

1.-

COLSS ANALYSES f

2.

CPCS ANALYSES d

i

79

y.

.s SECTION 3 OVERVIEW - COLSS/CPCS ANALYSIS III.

' COLSS/CPCS RELOAD ANALYSES OUTPUTS A.

COLSS

'l.

DATABASE CONSTANTS 2.

ADDRESSABLE CONSTANTS B.

CPCS 1.

RELOAD DATA DLOCK CONSTANTS

- 2.

ADDRESSABLE CONSTANTS C.

OTHER COLSS/CPCS RELATED RELOAD PRODUCTS I-e 1

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SECTION 4 CPC/COLSS ANALYSIS RESULTS COMPARISON FOR CPCS AND COLSS DESCRIPTION OF ' DIFFERENCES IN DATA BASE CONSTANTS

=

DESCRIPTION OF DIFFERENCES IN ADDRESSABLE CONSTANTS

=

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SUMMARY

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APS~ is in the Final Phase of Obtaining CE

=

Reload technology Independent Analysis Unit 3

- Cycle 3 i

APS Currently Plans. on Submitting a Topical Report

. =

for NRC Review by 2nd Quarter of 1991 l

APS is Preparcd to Provide the NRC with all

=

Information Required to Demonstrate Licensibility

=

- NRC Approval of APS Reload Design Capability Requested

- by 1st Quarter of -1992 I

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NRCL FEEDBACK 4

APS PROPOSES AN ADDITIONAL WORKING LEVEL MEETING WITH NRC l

FIRST-QUARTER 1991 TO REVIEW TOPICAL REPORT FIRST DRAFT i

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