ML20214A215

Summary of 870508 Meeting W/Util & GE Re Licensee Plans for Fuel Reload.List of Attendees & Meeting Viewgraphs Encl

ML20214A215
Person / Time
Site:	River Bend
Issue date:	05/13/1987
From:	Stern S Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
Shared Package
ML20214A217	List: ML20214A215 ML20214F592
References
NUDOCS 8705190331
Download: ML20214A215 (47)

Text

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'o, UNITED STATES 8"

' 'o NUCLEAR REGULATORY COMMISSION

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WASHINGTON, D. C. 20666 e

May 13,1987

,g Docket No. 50-458 MEMORANDUM T0:

File FROM:

Stephen M. Stern, Project Manager ProjectDirectorate-IV Division of Reactor Proje~ cts - III, IV, V and Special Projects

SUBJECT:

MINUTES OF MEETING; NRC AND GULF STATES UTILITIES ON FUEL RELOAD OF MAY 8, 1987 The River Bend Station licensee, Gulf States Utilities, its contractor and the staff met to review licensee plans for fuel reload.

General Electric Corporation attended as the licensee's contractor.

The licensee is currently projecting the end of their first refueling outage for November 1987.

The licensee and contractor state that the application for fuel reload will be i

based on the General Electric standard fuel load application, GESTAR II.

All core reload materials will either be submitted on the River Bend docket or.by reference to GESTAR II which has earlier been docketed and approved by the staff.

The licensee and contractor stated that this application will be " conventional" with no special facets requiring additional staff resources for review.

The staff stated 'that the licensee must clearly specify and justify the applica-bility of all limiting events.

The staff advisea the licensee of resource limitations that might preclude timely review of a complex licensing application.

l Thelicensedwillfileforthisreloadapplicationbyendo July, early August l

1987.~

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trn,ProjectManager Pro t Directorate - IV Divl< sion of Reactor Pro;ects - III, l

IV, V and Special Pro;'ects gnost901 b

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2, ATTENDEES MAY 8, 1987 MEETING:

NRR & GS TOPIC:

PLAN FOR FUEL RELOAD Name Organization Title Phone Stephen M. Stern DRST/NRR Pro;ect Manager 492-7000 ~

Chuck Paone GE/FuelProj.

Pro ectManager(401)925-4696 Bill Zarbis GE/ Fuel Lic.

Sen;ior Engineer (408 925-5070 Eddie R. Grant GSU/ Licensing Director (504 381-4145 Wayne Hodges NRR/RSB Chief (301 492-7483 i

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O GENERAL ELECTRIC FUEL LICENSING GE FUEL LICENSING

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o PRE-APPROVED FUEL DESIGNS.

o PRE-APPROVED ANALYSIS METHODS I

o PRE-APPROVED GENERIC RESULTS

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PRE-APPROVED LICENSING o

GESTAR II o

LICENSING EVALUATION METHODS o

ELIMINATING OR BOUNDING ANALYSES-o FUEL DESIGN LICENSING o

OPERATING FLEXIBILITY OPTIONS t

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O W

O GESTAR II i

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GESTAR II GENERAL ELECTRIC STANDARD APPLICATION FOR REACTQR FUEL o-BASIS FOR ALL FUEL LICENSING o

CONTAINS RESULTS OF GENERIC NRC REVIEWS o

FULLY APPROVED BY NRC o

MAINTAINED UP TO DATE o

CONSISTS OF:

BASE DOCUMENT COUNTRY-SPECIFIC DOCUMENT STANDARDIZED PLANT-SPECIFIC SUBMITTAL

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GESTAR-il LICENSING Pre-Approved A

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.i Base Country Cycle Report Supplement Specific 1

• Fuel Design e Reactor Limits e Core Loading e Nuclear Models Methods Pattern e Thermal-

• Bounding

• Cycle-Specific Hydraulic Models Analyses Analyses Results e Country-Specific Requirements

• Plant-Unique i

information scoss.oo i

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GESTAR II CONTENTS VOLUME 1:

BASE DOCUMENT SECTION 1 INTRODUCTION SECTION 2 FUEL MECHANICAL DESIGN SECTION 3 NUCLEAR EVALUATION SECTION 4 STEADY-STATE HYDRAULIC ANALYSES APPENDIX A SAFETY ANALYSIS REPORT ROADMAP APPENDIX B SAFETY ANALYSIS REPORT VOLUME 2:

COUNTRY-SPECIFIC DOCUMENT SECTION 1 INTRODUCTION SECTION 2 REACTOR LIMITS DETERMINATION APPENDIX A SUPPLEMENTAL RELOAD LICENSING SUBMITTAL APPENDIX B RESPONSES TO NRC QUESTIONS APPENDIX C NRC SAFETY EVALUATION REPORTS i

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GESTAR II VOLUME 1 SECTION 1

INTRODUCTION o

REPORT PURPOSE AND SCOPE o

CORE LATTICE TYPE o

FUEL DESIGNATIONS AND LENGTHS I

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GESTAR II VOLUME 1 SECTION 2 - FUEL MECHANICAL DESIGN o

FUEL ASSEMBLY DESCRIPTION o

FUEL R0D LIMITS AND ANALYSES THERMAL MECHANICAL o

FUEL ASSEMBLY MECHANICAL EVALUATIONS HANDLING AND SHIPPING NORMAL AND TRANSIENT SEISMIC AND LOCA o

FUEL QUALITY ASSURANCE..

GESTAR II VOLUME 1 SECTION 3 - NUCLEAR EVALUATION METHODS o

LATTICE NUCLEAR CHARACTERISTICS REACTIVITY LOCAL PEAKING FACTORS D0PPLER COEFFICIENT VOID REACTIVITY o

REFERENCE LOADING PATTERN CALCULATIONS CORE REACTIVITY CONTROL SYSTEM WORTH SHUTDOWN MARGIN STANDBY LIQUID CONTROL SYSTEM TRANSIENT AND ACCIDENT INPUT PARAMETERS

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GESTAR II VOLUME-1 SECTION 4 - STEADY-STATE HYDRAULIC ANALYSES MODELS I

o PRESSURE DROP CALCULATIONS o

BYPASS FLOW DETERMINATION 4

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GESTAR II

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VOLUME 2 - COUNTRY-SPECIFIC DOCUMENT t

(UNITED STATES) t

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o PROVIDES SAFETY ANALYSIS METHODOLOGY UNIQUE TO EACH COUNTRY o

PROVIDES FORMAT FOR RELOAD CYCLE-SPECIFIC INFORMATION o

PROVIDES RESPONSES TO REGULATORY QUESTIONS AND REGULATORY-l SERs i

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GESTAR II VOLUME 2 - COUNTRY-SPECIFIC DOCUMENT (US).

SECTION.1 - INTRODUCTION

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o PROVIDES PLANT APPLICABILITY OF DOCUMENT i

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GESTAR II VOLUME 2 - COUNTRY-SPECIFIC DOCUMENT (US).

SECTION 2 - REACTOR LIMITS DETERMINATION i

o FUEL CLADDING INTEGRITY SAFETY LIMIT q

STATISTICAL MODEL i

B0UNDING ANALYSIS o

OPERATING MCPR LIMIT DETERMINATION MODEL DESCRIPTION - TRANSIENT ANALYSIS LIMITING EVENT DETERMINATION j

TURBINE TRIP WITHOUT BYPASS

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LOAD REJECTION WITHOUT BYPASS LOSS OF FEEDWATER HEATING-INADVERTENT HPCI STARTUP n

FEEDWATER CONTROLLER FAILURE PRESSURE REGULATOR DOWNSCALE FAILURE (BWR/6 ONLY)

CONTROL R0D WITHDRAWAL ERROR (BOUNDING ANALYSIS) i C0ASTDOWN OPERATION LOW FLOW EFFECTS PLANT-SPECIFIC INPUTS MCPR MARGIN IMPROVEMENT OPTIONS OPERATING MAP EXPANSION OPTIONS-1 o

OVERPRESSURE PROTECTION SAFETY VALVE SIZING BASIS I

~

MSIV CLOSURE - FLUX SCRAM o

B0UNDING STABILITY APPROACH 1 -

GESTAR II VOLUME 2 - COUNTRY-SPECIFIC DOCUMENT (US).

SECTION 2 - REACTOR LIMITS DETERMINATION o

ACCIDENT EVALUATIONS CONTROL R0D DROP ACCIDENT ANALYSIS METHODOLOGY B0UNDING ANALYSIS LOSS-0F-COOLANT ACCIDENT ANALYSIS MODEL DESCRIPTIONS MAIN STEAM LINE BREAK ANALYSIS B0UNDING ANALYSIS DOSE EVALUATION 1

FUEL HANDLING ACCIDENT BOUNDING ANALYSIS DOSE EVALUATION RECIRCULATION PUMP TRIP B0UNDING ANALYSIS DOSE EVALUATION LOADING ERROR ACCIDENT ROTATED EVALUATION METHOD MISLOCATED EVALUATION METHOD AND B0UNDING RESULTS -

i GESTAR II VOLUME 2 - COUNTRY-SPECIFIC DOCUMENT (US).

. APPENDICES o-APPENDIX A - SUPPLEMENTAL RELOAD LICENSING SUBMITTAL PLANT-SPECIFIC SUBMITTAL FORMAT REFERENCES ANALYSES DESCRIPTION o

APPENDIX B - RESPONSES TO NRC QUESTIONS APPROVED RESPONSES ONLY i

UPDATED CONSISTENT WITH NRC SER' o

APPENDIX C - NRC SAFETY EVALUATION REPORT l

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PLANT-SPECIFIC-SUBMITTAL (RELOADS);

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o RESULTS OF' RELOAD ANALYSES f

o RELOAD SPECIFIC INFORMATION o

ONLY TABLES AND FIGURES I

o NO TEXT l

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NEW INFORMATION HIGHLIGHTED - SEPARATE APPENDICES l

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SUPPLDENTAL RELOAD LICENSING SUBMITTAL FOR' O

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Verified:

Approved:

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1.

PLANT-UNIQUE ITEMS (1.0)*

List items different from or Appendix or Reference of Item not included in NEDE-24011:

h 2.

RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 and 4.0) t Fuel Type Qcle Loaded Number h

h Irradiated h

Q New Total 3.

REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle core average exposure at end of cycle:

h mwd /MT Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations:

Q mwd /MT Assumed reload cycle core average exposure at end of cycle:

Q mwd /MT J

Core loading pattern:

Figure 1 h

4.

CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO VOIDS, 20*C (3.3.2.1.1 and_3.3.2.1.2)

Beginning of Cycle, K,ff Uncontrolled 1.1 @

Fully Controlled 0.9 @

Strongest Control Rod out 0.9 @

R. Maximum Increase in Cold Core Reactivity with Exposure into Cycle, AK 0.0 @

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• ( ) Refers to area of discussion in " General Electric Standard Application for Reactor Fuel". NEDE-24011-P-A (latest approved revision); a letter "S" preceding the number refers to the appropriate country-specific l

supplement.

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5.

STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3) 1 Shutdown Margin (Ak)

(20 den. C. Xenon Free)

.m 0.0 @

6.

RELOAD-UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 and 5.2.2)

(COLD WATER INJECTION EVENTS ONLY)

EOC h EOC - h mwd /MT Void Fraction (%)

h h

Average Fuel h

h Temperature (,F)

Void Coefficient h/-h h /- Q N/A* (C/% Rg) j l

Doppler Coefficient

-0. Q /-0. @

-0. @ /-0. Q N/A* (C/*F)

Scram Worth N/A*(S) 7.

RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2) f Fuel Peaking Factors R-Bundle Bundle Flow Initia)

Design Local Radial Axial Factor Power (MWT)

(1000 lb/hr)-

MCPR i

Exposure: EOC hhhh h

h 1.@

Appro-i priate

]

Fuel Design (s)

Exposure: EOC h mwd /MT hhhh h

h 1.@

Appro-priate Fuel Design (s)

• N = Nuclear Input Data; A = Used in Transient Inalysis

• • Generic exposure-independent values are used as given in " General Electric Standard Application for Reactor Fuel," NEDE-24011 (latest revision).

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SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)

Transient Recategorization:

h Recirculation Pump Trip:

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Rod Withdrawal Limiter:

h

. Thermal Power Monitor:

h Improved Scram Time:

h Exposure Dependent Limits:

Q Exposure Points Analyzed:

9.

OPERATING FLEXIBILITY OPTIONS (S.2.2.3)

Single-Loop Operation:

h Load-Line Limit:

h Extended Load Line Limit:

h Increased Core Flow h

Flow Point Analyzed:

Feedwater Temperature Reduction: h ARTS Program; h

h Maximum Extended Operating Domain:

10.

CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1)

t. CPR

Flux, Q/A Appropriate Transient

(% NBR)

(% NBR)

Fuel Design (s)

Figure Exposure: BOC to EOC - h mwd /MT Limiting Pressure and

0. @

Power Increase Transient Limiting Coolant Temperature

0. @

Decrease Transient Feedwater Controller Failure 0.@

4 0

Flux, Q/A Transient

(% NBR)

(% NBR) Appropriate Fuel Design (s)

Figure Exposure: EOC - h mwd /MT to EOC

^

Limiting Pressure 0.@

and Power Increase Transient Limiting Coolant Q

0.@

Temperature Decrease Transient Feedvater Controller g

0. @

Failure 11.

LOCAL ROD WI'nIDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE)

TRANSIENT

SUMMARY

(S.2.2.1)

• Limiting Rod Pattern: Figureh**

Rod Position ***

8x8R/P8x8R/

8x8R/P8x8R/

Rod Block (Feet BP8x8R/CE8x8E/

BP8x8R/GE8x8E/

Reading Withdrawn)

GE8x8EB 8x8 8x8 GE8x8EB b

b l

b 4

Set Point Selected: @

• If plant has ARTS, this section is replaced with a reference to the ARTS document.

• • 1f the generic rod withdrawal error analysis is used, this figure is not reported.

• • • If the generic rod withdrawal error analysis is used, rod position and MLRCR are not reported since it will always be significantly below the plastic strain limit.

12.

CYCLE MCPR VALUES (S.2.2)

Non-Pressurization Events 4

Exposure Range: BOC h to EOC h Appropriate Fuel Design (s),

Limiting Coolant

1. @

Temperature Decrease Transient Fuel Loading Error 1.@

Rod Withdrawal Error 1.@

Pressurization Events Option A Option B Appropriate Fuel Design (s) Appropriate Fuel Design (s) i Exposure Kange:

BOC to EOC - h mwd /MT Limiting Pressure 1.@

1.@

and Power I

Increase Transient Feedwater Controller 1.@

1.@

Failure Exposure Range:

EOC - @ mwd /MT to EOC Limiting Pressure 1@

1.@

i and Power Increase Transient Feedwater Controller 1.@

1@

l Failure 13.

OVERPRESSURIZATION ANALYSIS

SUMMARY

(S.2.3)

P P,3 v

Transient (psig)

(psig)

Plant Response

'l MSIV Closure h

h Figure 6 (Flux Scram)

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14.

LOADING ERROR RESULTS (S.2.5.4)

Variable Water Cap Misoriented Bundle Analysis: h Event Initial CPR Resulting CPR j

15.

CONTROL ROD DROP ANALYSIS RESULTS (S.2.5.1) l Bounding Analysis Results:

Doppler Reactivity Coefficient:

Figureh Accident Reactivity Shape Functions:

Figureshandh Scram' Reactivity Functions:

Figureshandh t

Plant Specific Analysis Results:

Parameter (s) not Bounded, Cold:

Resultant Peak Enthalpy, Cold:

h Parameter (s) not Bounded, HSB:

l l

Resultant Peak Enthalpy, HSB:

Q 16.

STABILITY ANALYSIS RESULTS (S.2.4)*

Rod Line Analyzed: h Decay Ratio:

Figur'eh j

Reactor Core Stability Decay Ratio, x /x 0.@

y 0 Channel Hydrodynamic Performance Decay Ratio, x /*0 2

l Channel Type P8x8R 0@

8x8R 0.@

8x8

0. @

l

• This section is included only for BWR 4/5/6 plants which have not implemented i

NRC approved stability monitoring documentation. All other BWR 4/5/6 plants are exempt from reporting these values providing they utilise acceptable fuel j

types as outlined in Section S.2.4 of CESTAR and insert a note in this section stating: "SIL-380 recosamendations have been included in the (plant name) l operating procedures and, therefore, no stability analysis is required".

For BWR 2/3 plants, the following note should be used: "BWR 2/3 plants are exempt from performing cycle specific stability analyses".

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4 17.

LOSS-OF-COOLANT ACCIDENT RESULT (S.2.5.2)

(For Those Plants With An ECCS Report Which is Separate From the FSAR)-

IDCA Method Used:

SAFE /REFLOOD/ CHASTE: h SAFER /GESTR-LOCA:

h

Reference:

Individual Plant LOCA Document (as amended) o MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE (INCLUDE FOLI4 WING TABLE FOR i

EACH NEW BUNDLE DESIGN) h Fuel Type: h Plant:

i Average Planar Exposure MAPLHGR PCT

' Oxidation (mwd /t)

(kW/ft)

(*F)

Fraction Q

Q Q

j 200 1,000 4

i 5,000 j

i 10,000 15,000 20,000 25,000 30,000 35,000 40,000 Q

Q Q

j d

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• To account for the 2% uncertainty in bundle power required by Appendix K, i

j the SCAT calculation is performed with an MCPR of 1.18 (i.e., 1.2 divided by 1.02) for a bundle with an initial MCPR of 1.20.

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l RELOAD LICENSING VERSUS INITIAL CORE LICENSING o

INITIAL CORE LICENSING ALL EVENTS CONSIDERED APPROVED METHODS USED DETAILED. TEXT AND RESULTS PROVIDED I

REFLECTS INITIAL CORE CONDITIONS-o RELOAD-LICENSING ONLY LIMITING EVENTS CONSIDERED (JUSTIFICATION IN j

GESTAR) r APPROVED METHODS USED RESULTS PROVIDED ONLY (TEXT IN GESTAR)

REFLECTS RELOAD CYCLE CONDITIONS 4

6

GESTAR II o

PROVIDES~ SIMPLIFIED DOCUMENTATION ORGANIZED THE WAY FUEL AND CORE ANALYSES ARE PERFORMED APPLICABLE TO ALL 8X8 FUEL TYPES CONCISE PLANT-SPECIFIC SUBMITTAL o

ELIMINAT;ES REPETITIVE NRC REVIEW e

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LICENSING EVALUATION METHODS

LICENSING VEHICLES o-GESTAR'II o

LICENSING TOPICAL REPORTS 4

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EVALUATION MODEL LICENSING STATUS METHOD MQDEL APPLICATION GETAB APPROVED APPROVED ODYN APPROVED APPROVED GESTR-MECHANICAL APPROVED APPROVED NUCLEAR PHYSICS APPROVED APPROVED SAFE /REFLOOD APPROVED APPROVED i..

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O ELIMINATING OR B0UNDING ANALYSES..

r-ELIMINATING OR B0UNDING ANALYSES o.

EVENTS ANALYZED 7 TRANSIENTS 3 ACCIDENTS OVER-PRESSURIZATION o

ONLY LIMITING EVENTS REPORTED o

WHEN POSSIBLE, LIMITING EVENTS ELIMINATED B0UNDED

4 ELIMINATING OR BOUNDING ANALYSES o

ANALYSES ELIMINATED CONTROL R0D DROP ACCIDENT MISL0CATED BUNDLE LOADING ERROR j

STABILITY o

ANALYSES B0UNDED R0D WITHDRAWAL ERROR I -

e 6

e FUEL DESIGN LICENSING

EVOLUTION OF-GE FUEL' DESIGN 1

GE8X8NB

'1 GE8X8EB

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P8X8R i

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8X8R 8X8 7

7X7R ry 7X7 BARRIERC 4 -.... _. -

FUEL DESIGNS o

ALL 8X8 DESIGNS IN'GESTAR II o

REVIEW 0F NEW DESIGNS SIMPLIFIED REVIEWED ON GESTAR II. DOCKET PRE-APPROVED METHODS AND CRITERIA 0

v

t BARRIER FUEL DESIGN o

USED WITH P8X8R FUEL O

DESIGNED USING APPROVED--ANALYSES MODELS o

UNCONDITIONALLY APPROVED BY NRC, APRIL 1983 i

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e OPERATING FLEXIBILITY OPTIONS -

OPERATING FLEXIBILITY OPTIONS fD o

SINGLE LOOP OPERATION (SLO) o LOAD LINE LIMIT (EXTENDED LOAD LINE LIMIT IELLLA1) o INCREASED CORE FLOW (ICF) i FEEDWATER TEMPERATURE REDUCTION (FWTR) o E0C FEEDWATER HEATER OUT OF SERVICE o

MAXIMUM EXTENDED OPERATING DOMAIN (MEOD) d ELLLA ICF l

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OPERATING FLEXIBILITY OPTIONS PROGRAM WHAT IT IS ANALYSIS

- ELLLA OPERATION AB0VE 100% LOAD LINE TRANSIENTS ACCIDENTS SLO ONE RECIRC LOOP OUT OF SERVICE TRANSIENTS ACCIDENTS 4

+

ICF OPERATION AT HIGHER THAN RATED TRANSIENTS FLOW ACCIDENTS FWTR 1.

REDUCED FEEDWATER TEMPERATURE TRANSIENTS AT EOC ACCIDENTS 2.

OPERATION WITH FEEDWATER HEATER OUT OF SERVICE ME0D ELLLA AND ICF TRANSIENTS ACCIDENTS I

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STANDARD LICENSED OPERATING MAP I

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Power l

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i Flow l

i Operating Map Constrained i

by 100% Flow Control Line and 100% Pump Speed i

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STANDARD LICENSED FUEL CYCLE i-Power I

i Full Thermal Power Region with Coastdown Exposure i

i Coastdown to 40% Power Consistent with Fuel Mechanical Design Limits Approved in Gestar-Il sesse.oo

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l EXPANDED OPERATING MAP Power I

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Load Line Limit Analysis t

i Operating Map l

j Single increased l

Loop Core Operation Flow Flow f

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Operational Flexibility improvement to Expand Operating Map SC888.08 l

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l FUEL CYCLE EXTENSION i

Power

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'Feedwater i

Temperature i

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1 Load Line I

l Limit increased Full Thermal Power l

Core Flow l

i Region with Coastdown l

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Possible bFuel Mechanical

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Design Limited \\

g Exposure Improved Uranium Utilization t

i SC688.07 l

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