ML20247L401

Reload Safety Evaluation for Redesign of Sequoyah Unit 2 Cycle 4

ML20247L401
Person / Time
Site:	Sequoyah
Issue date:	03/31/1989
From:	Gergos B, Novendstern E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20247L052	List: ML20247L046 ML20247L401 ML20247L406
References
NUDOCS 8906020158
Download: ML20247L401 (12)

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RELOAD SAFETY EVALUATION g

FOR THE REDESIGN OF SEQUOYAH UNIT 2 CYCLE 4 .

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' March 1989

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,. Edited by: B. W. Gergos i

e Approved: ~'<N E. H. Novendstern, Manager T/H Design and Fuel Licensing Commercial Nuclear Fuel Divisions 8906020158 890523

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p , , TABLE OF CONTENTS l

,, Title Page 1.0 . INTRODUCTION AND

SUMMARY

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H s 1.1 Introduction 1

[ 1.2 General Description 1 2.0 - REDESIGN-2.1 '- Loading Pattern Alterations' 2 2.2 Safety Analysis Input Parameter Evaluation .2 2.3 Mechanical and Thermal and Hydraulic Design 3

3.0 J POWER CAPABILITY AND ACCIDENT EVALUATION

. (. 3.1 Power Capability 4 3.2 Accident Evaluation 4

4.0 REFERENCES

. 5 APPENDIX A RELOAD SAFETY EVALUATION OF SEQUOYAH UNIT 2, CYCLE 4

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( . LIST OF TABLES P Table ' Title Page-1' Fuel Assembly Design Parameters ' , '6-

'- '2 End-of-Cycle Shutdown Requirements and Margins - 7-LIST OF FIGURES Figure . ~ Title Page--

1 Core Loading Pattern c  : 8, s

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1.0 INTRODUCTION

AND

SUMMARY

1.1 Introduction q

The Sequoyah Unit 2 Cycle 4 nuclear design has been evaluated for the consequences.

of a loading pattern change. The loading pattern change was required when assembly '

P39 was discovered to have failed fuel rods.

This report presents the changes that have been made to.the' original design loading pattern and the subsequent evaluation of all input parameters to the safety analysis.

This evaluation demonstrates that the conclusions of the Power. Capability and Accident <

Evaluation of the original Reload Safety Evaluation'(RSE), Reference 1, is valid for the redesign. The original RSE is included as Appendix A to this report for completeness. It

.is concluded that the Cycle 4 redesign does not cause the acceptable safety limits for any incident to be exceeded.

1.2 ' General Description

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The Cycle 3 end of life burnup was 14483 MWD /MTU,.which was below the maximum allowable burnup of 14500 MWD /MTU. The redesign of Cycle 4 has a projected full power capability of approximately 15700 MWD /MTU. The safety analysis for the

._ redesign is valid up to a burnup of 16220 MWD /MTU, which includes a power coastdown. '

As a result of. damage to assembly P39, the assembly and its seven symmetric counterparts were exchanged with eight symmetric assemblies on the baffle. An assembly.from Region 4, originally scheduled for discharge, was designated as the replacement for P39. P39 was also from Region 4. The region burnups at the end of Cycle 4 are shown in Table 1. The redesign loading pattern is provided in Figure 1.

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2.0 REDESIGN 2.1 Loading Pattern Alterations i

i The. development of the new loading pattern considered the selection bf the best alternative for the damaged assembly and its' plucement in the core. No assembly with similar reactivity from among the Region 4 assemblies scheduled for discharge was available for direct substitution. The highest reactivity assembly from among the Region .

4 assemblies scheduled for discharge was selected to replace assembly P39. Although I the substitute assembly,.P54, was from the same region, it had an average burnup of approximately 10000 MWD /MTU greater. than P39. The reactivity difference was sufficient enough to require replacement of the P39 symmetric partners. In order to use '

the seven other assemblies, the eight Region 5A assemblies originally on the baffle were moved into the symmetric P39 locations. The corresponding Region 4 assemblies with the P39 substitute were assigned to the baffle locations. As the relative power in the  !

baffle position is low, the asymmetry caused by the substitution was negligible.

( The Region.5A assemblies are more reactive than the original Region 4 assemblies.

This caused local power distribution changes in the area of the substitution. The global  :

power. distribution changes were relatively small causing only slight increases in the peaking factors. As more reactive fuel was moved into the interior, the effect on cycle

- lifetime was mitigated.

l 2.2 Safety Analysis input Parameter Evaluation s

The redesign lo'ading pattern characteristics were very similar to the original loading pattern. The current nuclear design limits were evaluated and verified to be applicable for the redesign using the methods described in Reference 2. The one exception to this was the analysis to verify that the core remained subcritical on soluble boron following a j hypotheticallarge break LOCA. A separate analysis was performed for the redesign and the conclusions of reference 1 remain valid. ]

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p"4' ;The total controlirod worth increased for the redesign loading pattern.'.The shutdown margin summary presented inTable 3 of Reference'1 has been revised for the redesign

and is presented as Table 2 herein.

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?2.3 Mechanical and Thermal and Hydraulic Design: ,

The. redesign of.the loading pattern and subsequent safety evaluation did not impact the mechanical and thermal and hydraulic design for this cycle The' discussions in Sections .

' 2.1 and 2.3 of Reference 1 are applicable for the redesign.

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3.0 ACCIDENT EVALUA' TION.

3.1 Power Capability

.The plant power capability: for the Cycle 4 design was evaluated considering the consequences - of those~ incidents examined in the FSAR, Reference 3,' using the previously' accepted design basis. This evaluation is described 'in Reference 1 and

' remains valid for the redesign.

, 3.2 Accident Evaluation The accident evaluation discussion given in Reference 1 is applicable for the redesign except_as noted below. The conclusions presented in the FSAR remain valid for the redesign. .

' The Cycle 4 redesign resulted in changes with respect to the original design to the following Boron Dilution at Power parameters.

( 1) the critical boron concentration at the BOC, HFP, no xenon, rods to insertion

, limits condition 2)' the difference between the critical boron concentration of the above condition and the N-1 rods inserted, BOC, HZP, no xenon critical boron concentration Operator action time is directly proportional to a ratio of the boron concentrations used in the analysis. The redesign parameters are bounded by the assumptions of the licensing basis analysis, f

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4.0 REFERENCES

1. Gergos, B. W., (Ed.), "The Reload Safety Evaluation of Sequoyah Unit 2, Cycle 4,"

November,1988.

2. Davidson, S. L., (Ed.), " Westinghouse Reload Safety Evaluation Methodology,"

WCAP-9273, July,1985.

3. Sequoyah Nuclear Plant Final Safety Analysis Report, U.S. NRC Docket No. 50-327.

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SEQUOYAH UNIT 2 - CYCLE 4 REDESIGN

.t FUEL ASSEMBLY DESIGN PARAMETERS' Region 4 M 53 M Ef3

' Enrichment (w/o U 235)+ 3.503 3.802_ 3.604 3.405 13.602 Density (% Theoretical)+ 94.758 95.011. 94.828 95.212 95.351 Numberof Assemblies 45 40 28 44 36 Approximate Burnup at++ 24000 15800 16800 0 .0

' Beginning at Cycle 4 i

'(MWD /MTU) .

p Approximate Burnup at+++ 36100 32600 34400 '19100 15800 End of Cycle 4 (MWD /MTU) .

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+ As built data

++ Based on EOC3 = 14500 MWD /MTU  ;

+++ Based on EOC4 = 16220 MWD /MTU L

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.[.g. TABLE 2 '

SEQUOYAH UNIT 2 - CYCLE 4 END-OF-CYCLE SHUTDOWN REQUIREMENTS AND MARGINS:

Control Rod Worth (%op.1 Gvele3 ._ Cycle ~ 4 -

All Rods inserted Less Worst Stuck Rod. 5.86 6.25

- (1) Less 10% - Cycle 3 5.28 --

Less 7% - Cycle 4 --

5.81 GontLQ1 Rod Re_quirements Reactivity Defects (Doppler, T avg, . 3.27 3.13' Void, Redistribution)

Rod insertion Allowance 0.40 0.58 c{"

and Reposition Allowance- --

0.19 (2) Total Requirements 3.67 3.90 Shutdown MarginU1) - (2)]fA6p.) .1.61 1,91 Beguireft3Auldown Margin _( Aap) 1.60 1.60 l

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'( SEQUOYAH UNIT 2, CYCLE 4 REDESIGN CORE LOADING PATTERN 8 I

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( APPENDIX A i

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