ML20204H655

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

ML20204H655
Person / Time
Site:	McGuire
Issue date:	07/31/1986
From:	Dzenis E DUKE POWER CO.
To:
Shared Package
ML20204H627	List: ML20204H624 ML20204H649 ML20204H655
References
TAC-61512, TAC-61513, NUDOCS 8608080172
Download: ML20204H655 (32)

Text

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I ATTACHpENT 2A RELOAD SAFETY EVALUATION MCGUIRE NUCLEAR STATION UNIT 1 CYCLE 4 REVISION 1 July, 1986 Edited by: 8. W. Gergos Contributors: L. G. Pilgrim D. S. Huegel P. J. Larouere .

J. R. Lesko P. Schueren C. M. Thompson Approved: /, y pfd E. A. Dzenis, Mandr Core Operations Nuclear Fuel Division

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! 8608090172 860804 PDR ADOCK 05000369 p PDR siw...- um

ATTACIDIENT 2A RELOAD SAFETY EVALUATION MCGUIRE NUCLEAR STATION UNIT 1 CYCLE 4 REVISION 1 July, 1986 Edited by: B. W. Gergos Contributors: L. G. Pilgrim D. S. Huegel P. J. Larouere .

J. R. Lesko P. Schueren C. M. Thompson Approved: g@ yard E. A. Dzenis, Mandr Core Operations Nuclear Fuel Division 8608090172 860804 PDR ADOCK 05000369 PDR P

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TABLE OF CONTENTS Title Pm 1

1.0 INTRODUCTION

AND

SUMMARY

1.1 Introduction 1 1.2 General Description 2 1.3 Conclusions 2 4

2.0 REACTOR DESIGN 2.1 Mechanical Design 4 2.2 Nuclear Design 4 2.3 Thermal and Hydraulic Design 6 7

3.0 POWER CAPABILITY AND ACCIDENT EVALUATION 3.1 Power Capability 7 7

3.2 Accident Evaluation 3.2.1 Kinetic Parameters 8 3.2.2 Control Rod Worths 8 3.2.3 Core Peaking Factors 8 9

4.0 TECHNICAL SPECIFICATION CHANGES l

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5.0 REFERENCES

APPENDIX A - Technical Specification Page Changes APPENDIX B - Safety Evaluation of Modified Fuel Assemblies siotts-soo730 j

LIST OF TABLES Table Title Page Fuel Assembly Design Parameters 11 1

Kinetic Characteristics 12 2

Shutdown Requirements and Margins 13 3

Control Rod Ejection Accident Parameters 14 4

LIST OF FIGURES Figure Title Page 1 Core Loading Pattern and Source and 15 Burnable Absorber Locations 1

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1.0 INTRODUCTION

AND

SUMMARY

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1.1 INTRODUCTION

This report presents an evaluation for McGuire Unit 1, Cycle 4, which demonstrates that the core reload will not adversely affect the safety of the plant. This evaluation was performed utilizing the methodology described in WCAP-9272, " Westinghouse Reload Safety Evaluation Methodology"(1) ,

McGuire Unit 1 is operating in Cycle 3 with Westinghouse 17x17 low parasitic (STD) and optimized fuel assemblies (OFA). For Cycle 4 and subsequent cycles, it is planned to refuel the McGuire Unit 1 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 0FA 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 (2'3) which could potentially be affected by the fuel reload have been reviewed for the Cycle 4 design described herein. The resuits of new analyses and the justification for the applicability of previous results for the remaining analyses are addressed in safety evaluations for a Positive Moderator Coefficient (10) and the UHI Elimination (14) licensing submittals.

' urSg u the cycle 2/3 refueling a problem was encountered in assembly ZV-1.

One removable rod was not reinserted because of mechanical interference. This asserably will remain in the core for Cycle 4. The safety impact for a rod removed with a water hole remaining is presented in Reference 4.

During the cycle 3/4 refueling, fuel assembly / fuel rod damage was detected on fuel assembly D03. As a result the core loading pattern was revised and fuel assembly 003 was not reinserted in the core for Cycle 4 operation. Also, two sien e-seem i

assemblies (D36 and ZV-1) were modified to increase their resistance to flow impingement through baffle joints and thereby reduce the potential for fuel damage. The safety impact for the modified assemblies is presented in Appendix 8. ,

1.2 GENERAL DESCRIPTION The McGuire Unit 1, 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 in Cycles 2 and 3, this cycle will contain one Region 4 demonstration assembly, designated in Figure 1 as 4A, of an intermediate flow mixer grid fuel assembly design. This assembly will be loaded into the core in a manner which satisfies the requirements given in Reference 13.

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

3411 Core Power (MWt)

System Pressure (psia) 2250 Core Inlet Temperature (*F) 558.9 Thermal Design Flow (gpm) 382,000 Average Linear Power Density (kw/ft) 5.44 l (based on 144" active fuel length)

1.3 CONCLUSION

S From the evaluation presented in this report, it is concluded that the Cycle 4 design does not cause the previously acceptable safety limits to be exceeded.

This conclusion is based on the following:

1. Cycle 3 burnup is between 11000 and 12127 MWD /MTU.

2. Cycle 4 burnup is limited to 13100 MWD /MTU including a coastdown.

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

4. The proposed Te:hnical Specification changes discussed in Section /.0 of this report and provided in Appendix A are approved.

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2.0 REACTOR DESIGN 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 submittal.(2)

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 model(5) . The fuel is designed and operated so that clad flattening will not occur, as predicted by the Westinghouse clad flattening model(6) . For all fuel regions, the fuel rod internal pressure design basis, which is discussed and shown acceptable in Reference 7, is satisfied.

Westinghouse has had considerable experience with Zircaloy clad fuel. This experience is described in WCAP-8183, " Operational Experience with Westinghouse Cores."(8) Operating. experience for Zircaloy grids has also been obtained from six demonstration 17x17 0FAs(2) , four demonstration 14x14 0FAsI ) and two regions of 0FA fuel in the McGuire Unit 1 Cycle 2 and 3 designs.

2.2 NUCLEAR DESIGN The Cycle 4 core loading is designed to meet a Fg (z) x P ECCS limit of In the event of UHI elimination, the F (z) x P ECCS limit 5 2.26 x K(z). n of 1 2.26* x K(z) will remain applicable to the Cycle 4 design.

Relaxed Axial Offset Control (RAOC) will be employed in Cycle 4 to enhance operational flexibility during non-steady state operation. RAOC makes use of i available margin by expanding the allowable AI band, particularly at reduced i

• Based on the LOCA analyses performed in support of the UHI elimination licensing submittal (14) ,

i 5101 L 6-860731 4

power. The RAOC methodology and application is fully described in Reference 9. The analysis for Cycle 4 indicates that no change to the safety parameters is required for RAOC operation. During operation at er near steady state equilibrium conditions, core peaking factors are significantly reduced due to the limited amount of xenon skewing possible under these operating conditions. The Cycle 4 Technical Specifications recognize this reduction in core peaking factors through the use of a Base Load Technical Specification.

Adherence to the gF limit is obtained by using the Fg Surveillance Technical Specification, also described in Reference 9. This provides a more convenient form of assuring plant operation below the Fg limit while retaining the intent of using a measured parameter to verify operation below Technical Specification limits. Fgsurveillance is only a surveillance requirement and as such has no impact on the results of the Cycle 4 analysis or safety parameters.

Table 2 provides a summp9 of Cycle 4 kinetics characteristics compared with the current limits based on previously submitted accident analyses.

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

L The loading pattern contains 304 burnable absorber (BA) rods located in 44 BA rod assemblies. Location of the BA rods are shown in Figure 1.

A more Positive Moderator Coefficient as compared to the current value will be utilized during Cycle 4. The safety evaluation is contained in Reference 10 and the associated Technical Specification changes are addressed in Section 4.0 of this report.

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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 (2) . The thermal hydraulic safety analyses used for Cycle 4 are based on a reduced design flow rate (15) in comparison to Reference 2. No significant variations in thermal margins will result from the Cycle 4 reload.

The thermal-hydraulic methods used to analyze axial power distributions generated by the RAOC methodology are similar to those used in the Constant 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 t were found to be less restrictive than those resulting from LOCA Fg considerations.

The Condition 11 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.

i secu e-eum 6

3.0 POWER CAPABILITY AND ACCIDENT EVALUATION ,

3.1 POWER CAPABILITY The plant power capability has been evaluated considering the consequences of those incidents examined in the FSAR(3) using the previously accepted design basis. It is concluded that the core reload 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(11) was used for fuel temperature evaluations. The LOCA limit at rated power can be met by maintaining FQ I*)

at or below 2.26 x K(z).

a 3.2 ACCIDENT EVALUATION The effects of the reload 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, 2) the safety evaluation I performed in support of the positive moderator coefficient (+7 pcm/*F) licensing submittal (10) or 3) the safety evaluation performed in support of the UHI Elimination licensing submittal (14) ,

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 three 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 Table 2 is a summary of the kinetic parameters current limits along with the associated Cycle 4 calculated values. All of the kinetic values fall within the bounds of the current limits except for the maximum moderator temperature coefficient. The safety evaluation for the Positive Moderator Coefficient is contained in Reference 10 and the associated Technical Specification changes are addressed in Section 4.0 of this report.

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 to'gether 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 12. 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 evaluated and are within the bounds of the current limits.

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4.0 TECHNICAL SPECIFICATION CHANGES To ensure that plant operation is consistent with the design and safety evaluation conclusion statements made in this report and to ensure that these conclusions remain valid, Technical Specifications changes will be needed for Cycle 4 to incorporate RAOC, the Positive Moderator Temperature Coefficient and the modified fuel assemblies. These changes are presented in Appendix A.

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5.0 REFERENCES

1. Davidson, S. L. (Ed), et. al., " Westinghouse Reload Safety Evaluation Methodology", WCAP-9272-P-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. " Reload Safety Evaluation McGuire - Unit 1 Cycle 3 - Revision 1," May 1985.

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

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

7. Risher, D. H., (et. al.), " Safety Analysis for the Revised Fuel Rod Internal Pressure Design Basis," WCAP-8964, June 1977.

8. Skaritka, J., lorii, J.A., " Operational Experience with Westinghouse Cores", WCAP-8183, Revision 14, July, 1985.

9. Miller, R. W., (et al.), " Relaxation of Constant Axial Offset Control-F0 Surveillance Technical Specification," WCAP-10217-A, June 1983

10. Westinghouse Transmittal to Duke Power Company, " Safety Evaluation for Operation of McGuire Units 1 and 2 with a Positive Moderator Coefficient",

January 1986.

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

12. 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.

13. Davidson, S. L., (Ed.), " Safety Evaluation for the Intermediate Flow Mixer Grid (IFM) Demonstration Fuel Assembly in McGuire Unit 1", February 1984.

14. Duke Power Company Transmittal to NRC, "McGuire Nuclear Station Safety Analyses for UHI Elimination", March 1986.

15. Duke Power Company Transmittal to NRC," McGuire 2 Cycle 2 0FA Reload",

November 1984.

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TABLE 1

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

2.108 3.205 3.204 3.20 3.40 Enrichment (w/o U-235)+

Density (% Theoretical)* 94.53 95.04 95.05 95.0 95.0 13 56 60 12 52 Number of Assemblies 16293# 20061 14604 0 0 Approximate Burnup at++

Beginning of Cycle 4 (MWD /MTU) 27891# 31899 27033 16767 14737 Approximate Burnup at++

End of Cycle 4 (MWD /MTU)

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• Optimized Fuel - Zirc grid

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

++ Based on EOC3 = 11560 MWD /MTU, E0C4 = 13100 MWD /MTV (coastdown included) l

1. The burnups noted are for the Region i fuel assemblies being used and are not an average for the whole region.

l 5101L 8-440731 g

TABLE 2 MCGUIRE UNIT 1 - CYCLE 4 KINETICS CHARACTERISTICS

- Cycle 4 Current Limits Design Maximum Moderator +5 < 70% of RTP +7 <70% of RTP Temperature Coefficient 0 > 70% of RTP +7 ramp to 0 from 70% to 100% of RTP (pcm/*F)*

Doppler Temperature -2.9 to -0.91 -2.9 to -0.91 Coefficiant(pcm/'F)*

Least Negative Doppler- -9.55 to -6.05 -9.55 to -6.05 Only Power Coefficient, Zero to Full Power, (pcm/% power)*

Most Negative Doppler -19.4 to -12.6 -19.4 to -12.6 Only Power Coefficient, Zero to Full Power (pem/%

power)*

Minimum Delayed Neutron .44 >.44 Fraction S,ff, (%)

Minimum Delayed Neutron .50 >.50 Fraction B (%)

(EjectedR88f,tBOL] a 100 <100 Maximum Differential Rod Worth of Two Banks Moving Together (pcm/in)*

-5

• pcm = 10 3, non e-sun' 12

TABLE 3 END-OF-CYCLE SHUTDOWN REQUIREMENTS AND MARGINS MCGUIRE UNIT 1 - CYCLE 4 Control Rod Worth (%Ap) Cycle 3 Cycle 4 All Rods Inserted 6.72 6.95 All Rods Inserted Less Worst Stuck Roli 5.90 5.95 (1) Less 10% 5.32 5.35 Control Rod Requirements (%A:)

Reactivity Defects (Doppler, T avg, 3.18 3.39 Void, Redistribution)

Rod Insertion Allowance 0.50 d.50 (2) Total Requirements 3.68 3.89 Shutdown Margin [(1) - (2)] (%ap) 1.64 1.46 Required Shutdown Margin (%Ap) 1.30 1.30 sioit e-..cm 13

TABLE 4 MCGUIRE UNIT 1 - CYCLE 4 CONTROL R0D EJECTION ACCIDENT PARAMETERS HZP-BOC ' Current Limit

• Cycle 4 Maximum ejected rod 0.75 <0.75 worth, %ap 11.0 <11.0 Maximum Fg (ejected)

HFP-BOC Maximum ejected rod 0.23 <0.23 worth, %Ap 4.5 <4.5 Maximum Fg (ejected)

HZP-E0C Maximum ejected rod 0.90 <0.90 worth, %Ap 20.0 <20.0 Maximum Fg (ejected)

HFP-EOC Maximum ejected rod 0.23 <0.23 worth, %Ap 5.9 <5.9 Maximum Fg (ejected)

• Based on the safety evaluation performed in support of the Positive Moderator Coefficient licensing submittal (10) ,

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180 R P N M L K J H G F E D C B A 5 6B 5 6B 5 6B 5 y 4 5 6B 4 6B 4 6B 4 6B 5 4 2 4 4 6B 4 6B 4 5 4 6B 4 6B 5 4 3 4A 5 8 SS 8 4

• # 4 6B 5 5 1 6A 5 6A 1 5 5 6B 5 5 ' 4 4 8 8 4 6B 4 5 4 6B 5 5 4 6B 4 4 6B 4 5 5 5 12 12 #

6B 4 5 4 5 4 6B 1 6B 4 6B 6 6B 4 6B 1 8 12 12 8 4 6A 4 5 1 6A 1 5 4 6A 4 6B 5 5 6B 7 8 4 8 4 4 ,

5 4 6A 1 6A 4 5 5 5 4 6B g 6B 4 5 5 90e 4 6A 4 5 1 6A 1 5 4 6A 4 6B 5 9 5 6B 8 4 8 4 4 ,

6B 4 5 4 5 4 6B 1 6B 4 6B 6B 4 6B 1 10 8 12 12 8 4 5 4 6B 5 5 4 6B 4 4 6B 4 5* 5 6B 11 12 12 6A 5 6A 1 5 5 6B 5 5 6B 5 5 1 12 4 8 8 4 4 6B 4 5 4 6B 4 6B 5 4 13 4 5 6B 4 8 SS 8 4 1

l 4 5 6B 4 6B 4 6B 4 6B 5 4 g 4 4 5 6B 5 6B 5 6B 5 15 1

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X Region Number

• Demonstration assembly with rod l

' #** V" Y BA's Assembly with stainless steel rods SS Secondary Source FIGURE 1 Core Loading Pattern McGuire Unit 1, Cycle 4 15

APPENDIX A TECHNICAL SPECIFICATION PAGE CHANGES Modifications to Pages:

3/4 1-Sa 3/4 2-4 5-6 l l

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REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT ,

LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be:

a. LesspositivethanthelimitsshowninFigure3.1-0,and

b. Less negative than -4.1 x 10

• delta k/k/ F for the all rods withdrawn, and of cycle life (EOL), RATED THERMAL POWER condition.

APPLICABILITY: Spacific tion: 3.1.1.3a. - MODES 1 and 2* onlyJ Specification 3.1.1.3b. - MODES 1, 2, and 3 only.f ACTION:

a. With the MTC inore positive than the limit of Specification 3.1.1.3a.

above, operation in MODES 1 and 2 may, proceed provided:

1. Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than the limits shown in. Figure 3.1-0 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT STAND 8Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6;

2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies ~ that the MTC has been restored to within its limit for the all rods withdrawn condition; and

\

3. A Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

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b. With the MTC more negative than the limit of Specification 3.1.1.3b.

above, be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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• With K,ff greater than or equal to 1.0.

1. See Special Test Exception 3.10.3.

I McGUIRE - UNITS 1 and 2 3/4 1-4 Amendment No.42 (Unit 1)

Amendment No.23 (Unit 2)

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REACTIVITY CONTROL SYSTEMS l

SURVEILLANCE REQUIREMENTS  !

4.1.1.3 T'he MTC shall be determined to be within its limits during each fuel

. cycle as follows:

a. The MTC shall be measured and compared to the BOL limit of

'Soec'ification 3.1.1.3a., above, prior to initial operation above 5% l of RATED THERMAL POWER, after each fuel loading; and

b. The MTC shall be measured at any THERMAL POWER and compared to I- -3.2 x 10 4 delta k/k/*F (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppe. In the event this comparison indicates the MTC is more negative than -3.2 x 10 4 delta k/k/'F, the MTC -

shall be remeasured, and compared to the EOL MTC limit of Specifica-tion 3.1.1.3b.. at least once per 14 EFPD during the remainder of the l fuel cycle.

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McGUIRE - UNITS 1 and 2 3/4 1-5 Amendment No.42 (Unit 1)

Amendment No.23 (Unit 2) i

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.'  % of Rated thermal Power FIGURE 3.1-0 MODERATOR TEMPERATURE COEFFICIENT V5 POWER LEVEL l

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-50 -40 -30 -20 Axial Flux Difference (% Delta-1)

FIGURE 3.2- 1 AFD Limits as a Function of Rated Thermal Power McGuire Unit 1 Cycle 4 l

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. DESIGH FEATURES -

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5.2.1.2 REACTOR BUILDING

a. Nominal annular space = 5 feet.

b. Annulus nominal volume = 427,000 cubic feet. I

c. Nomir,al cutsice heigtt (measurec f rom to; of f ouncation base tc the -

top of the dome) = 177 feet.

d. Nominal inside diameter = 125 feet. -

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e. Cylinder wall minimum thickness = 3 feet.

f. Dome minimum thickness =~2.25 feet. .

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g. Dome inside radius - 87 feet. t DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment is designed and shall be maintained for a maximum internal pressure of 15.0 psig and a temperature of 250*F.

5. 3 REACTOR CORE FUEL ASSEMBLIES -

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5.3.1 The core shall contain 93 fuel assemblies with each fuel assembly

' C Each fuel rod shall have a containing nominal active 264fuelfuel rod ofc 144 length M ,itt Zi :2.,and inches kontain a maxiaue total weight of

_ 1766 grams uranium. The initial cor~e loadipp shall haveta maximum 'enrichme'n t -

of 3.15 weight percent U-235. Re16ad fuel s iall be similar in physical design to the initial core loading and shall have maximum enrichment of 3.5 weight-percent U-235.

CONTROL ROD ASSEMBLIES ,  ;

i 5.3.2 The core shall c tain 53 full-length and no part-length cont'rol rod assemblies. The full- ngth control rod assemblies shall contain a nominal  :

142 inches of absorbe material. The nominal values of absorber material for Unit I control rods hall be 80% silver, 15% indium, and 5% cadmium. The nominal values of a orber material for Unit 2 control rods shall be 100%

boron carbide (B C for 102 inches and 80% silver,15% indium, and 5% cadmium for the 40-inch t' . All control rods shall be clad with stainless steel tubing.

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McGUIRE - UNITS 1 and 2 5-6

4 APPENDIX B MCGUIRE UNIT 1, CYCLE 4 SAFETY EVALUATION OF MODIFIED FUEL ASSEMBLIES

1. INTRODUCTION AND

SUMMARY

Inspection during the Cycle 3/4 refueling operations at McGuire Unit 1 disclosed fuel assembly 003 to be damaged. As a result 003 was not reinserted in the core and two fuel assemblies were modified. These modified fuel assemblies contain 255 and 256 fuel rods instead of 264 as i described in Section 5.3.1 of the Technical Specifications.

All of the accidents analyzed and presented in the FSAR which could potentially be affected by this change have been reviewed for the Cycle 4 design. As a result of these evaluations, it is concluded that the operation of the core containing the modified fuel assemblies does not cause the previously accepted safety limits for Cycle 4 to be exceeded and does not result in a significant hazards consideration. The sections which follow present a summary of the safety evaluations performed to determine the impact on the Cycle 4 core performance.

II. BACKGROUND Cycle 3/4 refueling operations at McGuire Unit 1, revealed fuel l assembly / fuel rod damage on fuel assembly 003. It was postulated that

! a contributing factor to this damage may have been flow through baffle joints impinging on rods resulting in induced vibration and subsequent damage to several fuel rods in fuel assembly 003. As a result, the core loading pattern for Cycle 4 was modified and fuel assembly 003 was not l

i reinserted in the core. In addition, two partially irradiated fuel assemblies were modified for insertion in the location previously l

occupied by fuel assembly 003, and as a precaution in another location

! where baffle gap measurements approximate those in the 003 location.

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III. MECHANICAL DESIGN The mechanical design of fuel assemblies ZV-1 and 036 are the same as the Region 4 assemblies with several exceptions. For these fuel assemblies eight fuel rods-(2 rows of four rods from corner nearest the baffle joints) were to be removed and replaced with stainless steel rods to eliminate their potential for failure. Replacement stainless steel rods have relatively the same outside dimensions as normal fuel rods.

In the case of fuel assembly ZV-1 eight fuel rods were removed and replaced with stainless steel rods as planned (See Figure B-1). As a result of difficulties encountered in loading replacement stainless steel rods in two end cells nearest the corner of fuel assembly D36, it was decided that these locations be left as open water channels (See Figure B-2). The modified fuel assemblies contain 255 and 256 fuel rods in fuel assemblies ZV-1 and 036 respectively.

The resulting geometrical irregularity of the fuel assembly modifications has been determined to not significantly alter the flow velocity profile and grid support conditions, and therefore not adversely affect the performance of the fuel. Stainless steel rods have similarly been used as replacements for fuel rods, and have successfully operated in a domestic plant for as long as 5 cycles. In conclusion, the open water channels and the stainless steel rods will not affect the mechanical integrity of fuel assemblies ZV-1 and 036. .

IV. NUCLEAR DESIGN The use of fuel assembly ZV-1 with one rod removed and eight stainless steel rods and fuel assembly 036 with two rods removed and six stainless steel rods has been evaluated. The Cycle 4 core loading is designed to meet a Fg (z) x P ECCS limit of <2.26 x K(z). These assemblies are located in the lower power region of the core. The limiting channel of 4

'Se core will not be located in either of these assemblies. This evaluation verifies that these assemblies will have no adverse affect on the parameters used in the accident analysis for Cycle 4.

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V. THERMAL-HYDRAULIC DESIGN The use of fuel assembly ZV-1 with cne rod removed and eight stainless steel rods and fuel assembly D36 with two rods removed and six stainless steel rods has been evaluated for thermal hydraulic effects. The effects of local peakir.g factors and hydraulic geometry on local flow and enthalpy distributions, and the effects of open water channels on core bypass flow are considered negligible. The effect of the reduced number of fuel rods on core average heat flux and kw/ft is also considered negligible. The prasant DNB core limits for Cycle 4 remain applicable.

VI. ACCIDENT EVALUATION The effects of the redesign on the design basis and postulated incidents analyzed in the FSAR have been examined, including the placement of eight stainless steel rods and the removal of one fuel rod from assembly ZV-1, and the placement of six stainless steel rods and the removal of two fuel rods from assembly D36. In all cases, it was found tha'. the effects can be accommodated within the conservatism of the assumptions used in the applicable safety analysis for Cycle 4.

VII. TECHNICAL SPECIFICATION CHANGES As a result of the modifications to fuel assemblies ZV-1 and D36 an amendment to Section 5.3.1 of the Technical Specifications will be needed. The proposed amendment is presented in Appendix A.

VIII. CONCLUSIONS Based on the evaluations presented above, it is concluded that operation during Cycle 4 with modified assemblies ZV-1 and D36, (1) does not involve a significant increase in the probability or consequences of an accident oreviously evaluated, does not create the possibility of an 5101L 6-660731

accident of a type different from any evaluated previously, does not involve a reduction in the margin of safety, and does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

5101L 6-860730

ss ss ss ss III I I ss ss ss ss l I I II I II I I (Xi I (X) I !X !l!

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l l l l l l ll I I I III I I I I WATER CHANNEL OFA FUEL RODS t 3.2 W/O )

GUIDE TUBES

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Q msmumimana nm STAINLESS STEEL ROD FIGURE B-1 IFM FUEL ASSEMBLY ZV-1 CONFIGURATION

WC WC SS SS X X X SS SS X. X SS SS X X X X X X X 0 X X X X X X X X X X X X l

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WATER CHANNEL 0FA FUEL RODS (3.20 W/0)

GUIDE TUBES INSTRUMENTATION TUBE SS STAINLESS STEEL R0D l

FIGURE B-2 FUEL ASSEMBLT D36' CONFIGURATION

ATTACHMENT 3 ANALYSIS OF SIGNIFICANT HAZARDS CONSIDERATION As required by 10CFR 50.91, this analysis is provided concerning whether the proposed amendments involve significant hazards considerations, as defined by 10CFR 50.92. Standards for determination that a proposed amendment involves no significant hazards considerations are if operation of the facility in accordance with the proposed amendment would not: 1) involve a significant increase in the probability or consequences of an accident previously evaluated; or 2) create the possibility of a new or different kind of accident from any accident previously evaluated; or 3) involve a significant reduction in a margin of safety.

Mr. H.B. Tucker's (DPC) May 15, 1986 letter (as supplemented via letters dated May 23, June 6. June 30, and July 10, 1986) presented an analysis of the significant hazards considerations for a proposed McGuire Unit 1/ Cycle 4 core reload and attendant RAOC and positive moderator temperature coefficient technical specification changes, concluding that the proposed amendments did not involve a significant hazards consideration. Subsequently, the NRC made a proposed determination that the amendment request involves no significant hazards consideration in a " Notice of Consideration of Issuance of Amendment to Facility Operating License and Proposed No Significant Hazards Consideration Determination and Opportunity for Hearing" published in the Federal Register on June 27, 1986.

However, as a result of damage to a fuel assembly discovered during Cycle 3/4 refueling operations it was decided to redesign the core without the damaged fuel assembly. The core loading pattern was altered and two partially irradiated fuel assemblies were modified to contain stainless steel rods and open water channels in place of certain fuel rods.

The McGuire Unit 1/ Cycle 4 reload safety evaluation (ref. Attachment 2A) was revised to reflect this core redesign, concluding that the previously acceptable safety limits remain unexceeded (although an additional technical specification change is required as a result of the two modified fuel assemblies). Consequently the previous no significant hazards determination related to the core reload's attendant RAOC and positive MTC technical specification changes remain valid (note that the proposed application of the positive MTC Tech. Spec. to McGuire Unit 2 is unaffected by this Unit 1/ Cycle 4 core redesign and consequently its no significant hazards determination remains valid). The safety impact for the modified assemblies is presented in Appendix B of the McGuire Unit 1/ Cycle 4 RSE which concludes that operation of the core containing the modified fuel assemblies does not result in a significant hazards consideration. The rewording of design features technical specification section 5.3.1 to generally allow certain modifications to fuel assemblies involving the fuel rod locations providing they are determined acceptable by cycle-specific reload analysis is an administrative type change enabling the specification to better describe the various potential fuel assemblies that could be shown acceptable for utilization in the reactor core. The specific fuel assembly modifications being utilized for McGuire 1/ Cycle 4 have been determined acceptable in the revised McGuire 1/ Cycle 4 RSE and shown not to involve a significant hazards consideration in Appendix B of that

cycle-specific RSE. The deletion of the sentence in the specification addressing the maximum enrichment of the initial core loading is an administrative change to remove an obsolete requirement. Both of these design features specification changes are also being proposed for McGuire Unit 2 and involve no significant hazards considerations on that unit for similar reasons.

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The commission has provided examples of amendments likely to involve no significant hazards considerations (48 FR 14870). One example of this type is

(vi), "A change which either may result in some increase to the probability or 2 consequences of a previously analyzed accident or may reduce in some way a safety

margin, but where results of the change are clearly within all acceptable criteria with respect to the system or component specified in the standard review plan

for example, a change resulting from the application of a small refinement of a previously used calculational model or design method". Because the evaluations previously discussed show that all of the accidents comprising the licensing bases which could poluulially be affected by the allered core loading pattern and modified fuel assemblies were reviewed for the Unit 1 Cycle 4 redesign and conclude that the reload design does not cause the previously acceptable safety i

limits to be exceeded, the above example can be applied to this situation.

Further, the NRC has made a proposed determination that a request for similar type changes for Beaver Valley Power Station Unit 1 involve no significant hazards consideration (ref. Federal Register dated July 25, 1986, pages 26783 and 26784). ,

l This example can also be applied to the rewording of the design features specification (applicable to Units 1 and 2) allowing certain modifications to fuel assemblies involving the fuel rod locations since the specification requires that

any results of a change be acceptable by cycle-specific reload analysis (i.e.

within acceptance criteria).

Another example of actions not likely to involve a significant hazards

{ consideration is (1), "A purely administrative change to technical specifications:

For example, a change to achieve consistently throughout the technical I

specifications, correction of an error, or a change in nomenclature". Accordingly the changes to the design features specification (applicable to Units 1 and 2) i deleting the sentence addressing the obsolete initial core loading maximum enrichment requirements involve no significant hazards considerations.

Based upon the preceding analyses, Duke Power Company concludes that the proposed amendments do not involve a significant hazards consideration.

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