ML20215E434

Rev 1 to Reload Safety Evaluation,Mcguire Nuclear Station, Unit 2 Cycle 4

ML20215E434
Person / Time
Site:	Mcguire
Issue date:	06/30/1987
From:	Dzenis E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20215E424	List: ML20215E419 ML20215E434
References
TAC-65072, TAC-65073, NUDOCS 8706190368
Download: ML20215E434 (20)

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i RELOAD SAFETY EVALUATION MCGUIRE NUCLEAR STATION q UNIT 2 CYCLE 4 REVISION 1

-June, 1987

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2 Edited by: C. R. Savage' l

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. l Approved: 4 51 Wed

. .: E.A.Drenis, Mand Core Operations e, Commercial Nuclear Fuel Division

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t-TABLE OF CONTENTS Title- .h 1$0 . INTRODUCTION-AND

SUMMARY

1 1.1 Introduction 1-

-1.2 General Description 2

.1.3 ' Conclusions -2 2.0; REACTOR DESIGN 4 l

2.1 yechanical Design 4 2.2 Nuclear Design 5 1

.E.3 .-Thermal a'nd Hydraulic Design 5 j

3.0 POWER CAPABILITY AND ACCIDENT EVALUATION 7

.. 1 3.1 Power Capability 7 3.2. Accident Evaluation 7 .-

3.2.1 Kinetic Parameters 8 3.2.2 Control Rod Worths '8 )

-3.2.3 Core Peaking factors 8 .]

.4,0 .. TECHNICAL SPECIFICATION CHANGES 9

5.0 REFERENCES

- 10 APPENDIX A - Technical Specification Page Changes l

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LIST OF TABLES l.-

Table' Title Pajgt

-1: Fuel Assembly Design' Parameters 11' 2 '- Kinetics Characteristics 12 3 End-of-Cycle Shutdown Requirements and Margins '13 4 Control Rod Ejection Accident' Parameters 14 LIST OF FIGURES ,

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11 Core'Loadin'g Pattern 15 t

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1.0 INTRODUCTION

AND

1.1 INTRODUCTION

This report presents an evaluation for McGuire Unit 2, Cycle 4, which j

. demonstrates that the core reload will not adversely affect the safety of the l plant. This evaluation was performed utilizing-the methodology described in I

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WCAP-9273-A, " Westinghouse Reload Safety Evaluation Methodology"(1) . This revised report replaces the original January 1987 Reload Safety Evaluation ,

'(RSE). The revision is necessary due to fuel assemblies P06 and P49 receiving ]

damage and not being usable for Cycle 4 operation. The core redesign does not l change the results of the safety evaluation or the conclusions of the January {

RSE. Changes from the original RSE are indicated with bars in the page margins. i m McGuire Unit 2 operated in Cycle 3 with Westinghouse 17x17 low parasitic (STD) I j and optimized fuel assemblies (OFA). For Cycle 4 it is plannad to refuel the N McGuire Unit 2 core with Westinghouse 17x17 optimized fuel assemblies. In the  ;

0FA transition licensing submittal (2) to the NRC, approval was requested for.

the transition from the STD fuel design to the OFA design and the associated proposed changes to the McGuire Units 1 and 2 Technical Specifications. The licensing submittal, which has received NRC approval, justifies the compatibility of the OFA design with the STD design in a transition core as well'as a full 0FA core. The OFA transition licensing submittal (2) contains mechanical, nuclear, thermal-hydraulic, and accident evaluations which are applicable to the Cycle 4 safety evaluation.

All of.the accidents comprising the licensing bases which could potentially be affected by the fuel reload have been reviewed for the Cycle 4 redesign i described herein. The results of new analyses and the justification for tne applicability of previous analyses are addressed in safety evaluations for the .

.RTD Bypass Elimination.(9) i n4a*e-s70eos }

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1 During the Cycle 3/4 refueling, fuel assembly / fuel rod damage was detected on

. two fuel assemblies. As a result, a core loading pattern was developed l l

excluding these damaged assemblies. The core loading pattern is provided in {

. Figure 1. l I

1.2 GENERAL DESCRIPTION The McGuire Unit 2, Cycle 4 reactor core will be comprised of 193 fuel assemblies arranged in the core loading pattern configuration shown in Figure 1. During the Cycle 3/4 refueling, 64 STD fuel assemblies will be replaced with 64 Region 6 optimized fuel assemblies. A summary of the Cycle 4 fuel inventory is given in Table 1 as shown on page 11, 1

Nominal core design parameters utilized for Cycle 4 are as follows:

Core Power (MWt) 3411

- System Pressure (psia) 2250 Core Inlet Temperature ('F) 558.5

- Thermal Design Flow (gpm) 382,000 Average Linear Power Density (kw/ft) 5.43 (based on 144" active fuel length)

1.3 CONCLUSION

S From the evaluation presented in this report, it is concluded that the Cycle 4 redesign does not cause any of the safety limits to be exceeded. This  ;

1 conclusion is based on the following: i

1. Cycle 3 burnup is limited to 10800 MWD /MTV. l l

2. Cycle 4 burnup is limited to 13500 MWD /MTU including a coastdown. {

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3. There is adherence to plant operating limitations in the Technical {

Specifications.

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.. 1 4.-:Th~e proposed Technical Specification change discussed in Section 4.0 of- {

this report'and provided in Appendix A is approved. . )

. . 5. The Large' Break LOCA. analysis for UHI Removal using the 1981 EM+ BASH is approved.

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l 2.0 REACTOR DESIGN I l

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2.1 MECHANICAL DESIGN The Region 6 fuel assemblies are Westinghouse 0FAs. The mechanical description and justification of their compatibility with the Westinghouse STD design in a transition core is presented in the OFA transition licensing 1

- submittal.(2)

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Region 6 has a smaller rod plenum spring than was used in previous fuel  !

regions. This new spring design satisfies a change in the non-operational 6g loading design criterion to'"4g axial and 6g lateral loading with dimensional stability." Notification of Westinghouse's plans to generically. incorporate this criterion change and the justification of no unreviewed safety questions j were previously transmitted to the NRC via Reference 13. The reduced spring l

. force reduces the potential for pellet chipping in the fuel rod.

• Table 1 presents a comparison of pertinent design parameters of the various fuel regions. The Region 6 fuel has been designed according to the fuel performance mode 1 I4) . The fuel is designed and operated so'that clad flattening will not occur, as predicted by the Westinghouse clad flattening model N) . For all fuel regions, the fuel rod internal pressure design  ;

basis, which is discussed and shown acceptable in Reference 6, is satisfied.

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Westinghouse has had considerable experience with Zircaloy clad fuel. This experience'is described in Reference 7. Operating experience for Zircaloy

- grids has also been obtained from six demonstration 17x17 0FAs(2) , four o

demonstration 14x14 0FAs(2) , three regions of 0FA fuel in the McGuire Unit 1 Cycle 2, 3, and 4 designs, and two regions of 0FA fuel in the McGuire Unit 2 Cycle 2 and 3 designs.

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2.2. NUCLEAR DESIGN The Cycle 4 core leading is designed to meet a F g (z) x P ECCS limit of

.. $' 2.32'x K(z). .

This' increased limit is justified by the analysis in Reference 12.

r Relaxed Axial Offset Control (RAOC) will be employed in Cycle'4 to enhance operational flexibility during non-steady state operation. 'The RAOC methodology.and application is fully described in Reference 8. The analysis for Cy'cle 4 includes a wider full power AI band than existed in Cycle 3.

.The RAOC Al band was widened by utilizing peaking factor margin which had become.'available.since the original McGuire RAOC analysis. No change to the safety parameters is required for Cycle 4 RAOC operation.

Table 2 provides a summary of Cycle 4 kinetics characteristics compared with c the current limits based.on previously submitted accident analyses.

.c Table'3 provides the control rod worths and requirements at the most limiting condition during the cycle (end of-life). The required shutdown margin is based on previously submitted accident analysis. The available shutdown margin exceeds the minimum required.

1 The loading pattern contains 320 wet annular burnable absorber (WABA)(14) l

-rods located in 52 fuel rod assemblies. Location of the WABA rods are shown l in Figure 1.

2.3 THERMAL AND HYDRAULIC DESIGN The thermal hydraulic methodology, DNBR correlation and core DNB limits used for Cycle 4, are consistent with the current licensing basis. No significant variations in thermal margins will result from the Cycle 4 redesign.

The thermal-hydraulic methods used to analyze axial power distributions

- generated by the RAOC methodology are similar to those used in the Constant 39487 6-670605 - 5 ,

Axial Offset Control (CAOC) methodology. Normal operation power distributions are evaluated relative to the assumed limiting normal operation power distribution used in the accident analysis. Limits on allowable operating-

' axial flux difference as a function of power level from these considerations -

l were found to be less restrictive than those resulting from LOCA Fg considerations.

' The Condition II analyses were evaluated relative to the axial power )

distribution assumptions used to generate DNB core limits and resultant-Overtemperature Delta-T setpoints (including the f(AI) function). No changes in the DNB core limits are required'for RAOC operation.

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. i 3.0 POWER CAPABILITY AND ACCIDENT EVALUATION.

3.1. POWER CAPABILITY The plant power capability has been evaluated considering the consequences of th'ose incidents examined in the FSAR(3) using the previously accepted design

- basis.- It is concluded that the redesign will~not adversely affect the ability to safely operate at the design power level (Section 1.0) during Cycle 4. For the overpower transient, the fuel centerline temperature limit

- of 4700'F can be accommodated with margin in the Cycle 4 core. The time dependent densification model(10) was used for fuel temperature evaluations. The LOCA limit at rated power can be met by maintaining Fg (z) at or below 2.32 x K(z).

3.2 ACCIDENT EVALUATION The effects of the reload redesign on the design basis and. postulated

- incidents analyzed-in the FSAR(3) were examined. In all cases, it was found that the effects were accommodated within the conservatism of the initial assumptions used in 1) the previous applicable safety analysis, or 2) the safety evaluation performed in support of the RTD Bypass Elimination licensing submittal (9) .

A core reload can typically affect accident analysis input parameters in the following areas: core kinetic characteristics, control rod worths, and core peaking factors. Cycle 4 parameters in each of these areas were examined as discussed in the following subsections to ascertain whether new accident analyses were required.

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3.2.1 KINETIC PARAMETERS

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Table 2 is a summary of the kinetic parameters current limits along with the 2 .

associated Cycle 4 calculated values. All of the kinetic values fall within the b'ounds of the current limits.

3.2.2' CONTROL ROD WORTHS

- Changes .in control rod worths may affect differential rod worths, shutdown margin, ejected rod worths, and trip reactivity. Table 2 shows that the-

- maximum differential rod worth'of two RCCA control banks moving together in their highest worth region for-Cycle 4 meets the current limit. Table 3 shows that;the Cycle.4 shutdown margin requirements have been satisfied. Table 4 is .

a summary of the current limit control rod ejection analysis parameters and the corresponding Cycle 4 values.

3.2.3 ' CORE PEAKING FACTORS Peaking factors for the dropped RCCA incidents were evaluated based on the NRC approved dropped rod methodology described in Reference 11. Results show that DNB design basis is met for all dropped rod events initiated from full power.

The peaking factors for steamline break and control rod ejection have been 4

evaluated and are within the bounds of the current limits. .l l

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'4.0 TECHNICAL SPECIFICATION CHANGES. 1 1

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.'To ensurt that plant. operation is cons istent w ti h the-design and safety- ,

. .L evaluation conclusion statements made in this report and to ensure that these -

conclusions remain valid, a Technical. Specifications change will be needed for l l

Cycle'4 to incorporate a wider full power RAOC AI band. .This change is presented in' Appendix A. . I 1-l ;

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5.0 REFERENCES

1. Davidson, S. L. (Ed), et. al., " Westinghouse Reload Safety Evaluation Methodology",-WCAP-9273-A, July 1985. .

2. Duke Power Company Transmittal _to NRC, " Safety. Evaluation for McGuire

. Units 1 and 2 Transition to Westinghouse 17x17 Optimized Fuel Assemblies",

December 1983.

3. "McGuire Final Safety Analysis Report."

4.-Miller,J.V.,(Ed.), " Improved Analytical Model used in Westinghouse Fuel Rod Design Computations", WCAP-8785, October 1976.

5. George, R.A., (et. al.), " Revised Clad Flattening Model", WCAP-8381, July 1974.

6. Risher, D. H., (et. al.), " Safety Analysis for the Revised Fuel Rod Internal Pressure Design Basis," WCAP-8964-A, August 1978.

7. Letter from E. P. Rahe Jr. (Westinghouse) to J. Lyons (NRC),

Subject:

transmittal of " Operational Experience with Westinghouse Cores" (through

• December 31,1985)(Non-Proprietary),NS-NRC-86-3184, November 26, 1986.

8. Miller, R. W., (et al.), " Relaxation of Constant Axial Offset Control-Fg Surveillance Technical Specification," WCAP-10217-A, June 1983

9. Westinghouse. Transmittal to Duke Power Company, "RTD Bypass Elimination Licensing Report: Final Version," September 1985.

10. Hellman, J.M. (Ed.),." Fuel Densification Experimental Results and Model i for, Reactor Operation", WCAP-8219-A, March 1975.

11. Morita, T., Osborne, M. P., et. al., " Dropped Rod Methodology for Negative Flux Rate Trip Plants," WCAP-10297-P-A (Proprietary) and WCAP-10298-A (Non .

Proprietary), June 1983.

12. Letter DAP-87-516, L. L. Williams (Westinghouse) to K. S. Canady (Duke Power), "McGuire BASH Analysis Submittal", March 1987.  ;

13. Letter from E. P. Rahe, Jr. (Westinghouse) to L. E. Phillips (NRC), April 12,1984,NS-EPR-2893,

Subject:

Fuel Handling Load Criteria (6g vs. 4g).

14. Letter from Thomas, C. 0., NRC, to Rahe, E. P., Westinghouse,

Subject:

Acceptance for Referencing of Licensing Topical Report WCAP-10021 (P),

Revision 1, and WCAP-10377 (NP), " Westinghouse Wet Annular Burnable Absorber Evaluation Report," August 9, 1983.

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4 TABLE'1-MCGUIRE UNIT 2 - CYCLE 4 FUEL ASSEMBLY DESIGN PARAMETERS Region 1 4* 5* 6A* 6B*

7 Enrichment (w/oU-235)+ 2.093 3.206 3.190- 3.20 3.40' Density (*/, Theoretical)+ 94.77 94.96 95.23 95.0 95.0:

I - Number of Assembiies 12 57 60 44 20 Approximate Burnup at++ 15294# ~18886 13360 0 0 Beginning-of Cycle.4 (MWD /MTU).-

--Approximate Burnup at++ 27045# 32544 24396 15374 15908 End of. Cycle 4' (MWD /MTU).

• Optimized Fuel - Zire grid

+'All fuel region values are as-built except Region 6 values which are nominal.

- ++ Based on EOC3 = 10800 MWD /MTU., EOC4 = 13500 MWD /MTU (coastdown included) h

1. The burnups noted are for the Region 1 fuel assemblies being used and are not

'+< an average for'the whole region.

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TABLE 2 i

C MCGUIRE UNIT 2 - CYCLE 4 KINETICS CHARACTERISTICS

• - Cycle 4 )

Current Limits Design Minimum Moderator- +7 <70% of RTP +7 <70% of RTP

' Temperature Coefficient +7 ramp to 0 from +7 ramp'to 0 from {

(pcm/*F)* 70% to 100% of RTP 70% to 100% of RTP {

Doppler Temperature -2.9;to -0.91 -2.9 to -0.91 ]

Coefficient (pcm/*F)* j i

Least Negative Doppler- -9.55 to -6.05 -9.55 to -6.05 i Only Power Coefficient, j 6

~ Zero to Full' Power, (pcm/% power)*-

F - Most Negative Doppler -19.4 to -12.6 -19.4 to -12.6 Only Power Coefficient,

" Zero to' Full Power (pcm/%

power)*

Minimum Delayed Neutron .44 >.44 Fraction seff' (%)

Minimum Delayed Neutron .50 >.50  ;

I Fraction 8 (%)

[EjectedR8$f,tBOL) a Maximum Differential Rod 100 ' <100 j Worth of Two Banks Moving i Together(pem/in)* j j

• pem = 10

-5 g

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TABLE-3~

1 ft- END-OF-CYCLE SHUTDOWN REQUIREMENTS AND MARGINS'

,MCGUIRE UNIT.2 - CYCLE 4- l Control Rod Worth (%Ap) Cycle 3 Cycle 4 All Rods Inserted 7.27. 8.03- j All Rods' Inserted Less Worst Stuck Rod 6.03 6.85 l

.(1)Less10%' 5.42 ~6.17 i Control Rod' Requirements (%Ap)

Reactivity Defects (Doppler, T avg, 3.06 3.87 Void,. Redistribution).

Rod Insertion Allowance- 0.50 0.50-(2) Total Requirements 3.56 4.37 Shutdown M 'argin ((1) - (2)) (%Ap) 1.86 1.80 Required' Shutdown Margin (%ap) 1.30 1.30 i

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TABLE-4

... . . MCGulRE UNIT 2 - CYCLE l

' CONTROL R0D EJECTION ACCIDENT PARAt;ETERS

+ HZP-BOC Current Limit * . Cycle 4-t ,

Maximum ejected rod . y 0.75 < 0.75-i:

worth, %Ap.

iMaximum Fg(ejected); 11 '. 0 - <11.0 HFP-BOC-Maximum ejected rod 0.23 <0.23 n worth,'%Ap' Maxi mum Fg(ejected)'- 4.5 <4.5 HZP-EOC L Maximum ejected rod' O.90 <0.90 worth, %Ap' Maximum Fg (ejected) 20.0 <20.0 HFP-EOC:

Maximumejected_ rod, 0.23 <0.23 worth, %Ap Maximum Fg (ejected) 5.9 <5.9

• Based on the safety evaluation performed in support of the RTD Bypass Elimination licensing'submitta1 I9) .

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MCGUIRE UNIT 2, CYCLE 4 l 15 l

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APPENDIX A TECHNICAL SPECIFICATION PAGE CHANGE Modification to Page:

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-50 -40 -80 -20 Axial Flux Difference (% Delta-Il FIGURE 3.2-1 AFD Limits as a Function of Rated Thermal Power McGuire Unit 2 Cycle 4 1

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