ML20082D657

Safety Evaluation for McGuire Units 1 & 2 Transition to Westinghouse 17x17 Optimized Fuel Assemblies

ML20082D657
Person / Time
Site:	McGuire, Mcguire
Issue date:	11/14/1983
From:	WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20082D641	List: ML20082D639 ML20082D657
References
TAC-53319, TAC-53320, TAC-56336, TAC-56337, NUDOCS 8311220496
Download: ML20082D657 (26)

Text

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Safety Evaluation for McGuire Units 1 and 2 Transition to Westinghouse 17x17 Optimized Fuel Assemblies Duke Power Company McGuire Units 1 and 2 Docket Nos. 50-369 and 50-370 Westinghouse Electric Corporation i

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8311220496 831114 PDR ADOCK 05000369 P

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TABLE OF CONTENTS

'Section

1.0 INTRODUCTION

1 2.0

SUMMARY

AND CONCLUSIONS 3

.3.0 MECHANICAL EVALUATION 5

4.0 NUCLEAR EVALUATION 9

5.0 THERMAL AND HYDRAULIC EVALUATION 10 6.C ACCIDENT EVALUATION 12 7.0

SUMMARY

OF TECHNICAL SPECIFICATION CHANGES 19 8.0 -REFERENCES 20 LIST OF. TABLES Table No.

Title Page 1

.FSAR Chapter 15 Accident Analysis Sensitivity 22 to Proposed Changes 2

Summary of Affected Technical ' Specifications 23 L!ST OF FIGURES Figure No.

Title Page 1

17x17 0FA/STD Fuel Assembly Comparison 24 1

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1.0 INTRODUCTION

McGuire Units 1 and 2 have been operating with Westinghouse 17x17 low-parasitic (STD) fueled cores. It is planned to refuel both Units I and 2 with Westinghouse 17x17 Reconstitutable Optimized Fuel Assembly (OFA) regions. As a result, future core loadings would range from an approximately 1/3 0FA - 2/3 STD transition core to eventually an all 0FA fueled core. The OFA fuel has similar design features compared to the STD fuel which has had substantial operating experience in a number of nuclear plants. The major differences are the use of six intermediate (mixing vane) Zircaloy grids for the OFA fuel versus six intermediate (mixing vane) Inconel grids for STD fuel and a reduction in fuel rod diameter (see Table 1). Major advantages for utilizing the OFA are:

(1) increased efficiency of the core by reducing the amount of parasitic material and (2) reduced fuel cycle costs due to an optimization of the water to uranium ratio.

This report summarizes the evaluation performed on the region-by-region reload transition from the present McGuire Units 1 and 2 STD fueled cores to cores with all optimized fuel. This report examines the differences betweer the Westinghouse STD design and evaluates the effects of these 11fferences for the transition to an all 0FA core. The evaluation considers the standard reload design methods described in Reference 1,- and the transition effects described in Chapter 18 of Reference 2.

Reference 3 presents the operating experience through December 1981 of six 17x17 0FA demonstration assemblies which have the McGuire 1 and 2 design features. By early 1983, two 17x17 0FAs will have completed three cycles of irradiation to about 28,500 MWD /MTV, two have completed two cycles to about 19,400 MWD /MTU, and two have completed one cycle of irradiation with burnups in excess of 9,000 MWD /MTV. All demonstration 17x17 0FAs examined were in good condition. This provides evidence of favorable operation of Zircaloy grids and reduced fuel rod diameters which are the major new design features of the 17x17 0FA.

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e Sections 3.0 through 6.0 summarize the Mechanical, Nuclear, Thermal and Hydraulic, and Accident Evaluations, respectively.

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2.0

SUMMARY

AND CONCLUSIONS Consistent with the Westinghouse STD reload methodology for analyzing cycle specific reloads (Reference 1), parameters were chosen to maximize the applicability of the transition evaluations for each reload cycle and to facilitate subsequent determination of the applicability of 10 CFR 50.59. The objective of subsequent cycle specific reload safety evaluations will be to verify that applicable safety limits are satisfied based on the reference evaluation / analyses established in this report. The mechanical, thermal and hydraulic, nuclear, and accident evaluations considered the transition core effects described for mixed cores in Chapter 18 of Reference 2.

The summary of these evaluations

~ for the McGuire Units 1 and 2 transitions to an all 0FA core are given in tne following sections of this submittal.

The transition design and safetv. evaluations consider the following nominal operating conditions: 4411 MWt core power, 2250 psia system pressure, 559.2'F core inlet temperature (HFP) at 2250 psia, and 386,000 gpm RCS thermal design flow.

The results of evaluation / analysis and tests described herein lead to the following conclusions:

a.

The Westinghouse OFA reload fuel assemblies for McGuire 1 and 2 are mechanically compatible with the current STD design, control rods, and reactor internals interfaces. Both fuel assemblies satisfy the current design bases for the McGuire units.

b.

Changes in the nuclear characteristics due to the transition from STD to 0FA fuel will be within the range normally seen from cycle to cycle due to fuel management effects.

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c.

The reload 0FAs are hydraulically compatible with the current STD

design, d.

The accident analyses for the OFA transition core were shown to provide acceptable results by meeting the applicable criteria.such as, minimum DNBR, peak pressure, and peak clad temperature, as required. The previously reviewed and licensed safety limits are met. Analyses in support of this safety evaluation establish a

. reference design on which subsequent reload safety evaluations involving 0FA reloads can be based.

e.

Plant operating limitations given in the Technical Specifications will be satisfied with the proposed changes noted in Section'7.0 of this report.

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3.0 MECHANICAL EVALUATION The mechanical design requirements and criteria approved by the NRC for the 17x17 0FA design are described in Reference 2.

The OFA has been designed to be mechanically compatible with the STD design, reactor internal interfaces, fuel handling and refueling equipment, and spent fuel storage racks. Figure 1 presents a comparison of the OFA and STD designs. The grid ce,nterline elevations for both designs are identical, minimizing mechanical and hydraulic interaction.

The assembly envelopes and the top and bottom Inconel grids are the same

- for both designs. The design changes between the STD and 0FA design include a reduction in fuel rod, guide thimble, and instrumentation tube diameters, replacement of the six intermediate (mixing vene) Inconel grids with Zircaloy (mixing vane) grids, and the replacement of a standard bottom nozzle with a reconstitutable bottom nozzle. (The reconstitutable bottom nozzle is the current design used on all 17x17 fuel assembly designs, both 0FA and STD, with the exception of both McGuire 1 Cycle I and McGuire 2 Cycle 1, which utilized the 17x17 standard bottom nozzle.)

The OFA fuel rod will maintain the same overall length as the STD fuel rod. However, the fuel rod diameters are reduced to eptimize the water to uranium ratio.

The top and bottom Inconel (non-mixing vane) grids of ti.e OFA are nearly identical in design to the Inconel grids of the STD design. The only difference is that the spring and dimple heights have been modified to accommodate the reduced diameter fuel rod. The six intermediate (mixing vane) grids are made of Zircaloy material rather than of Inconel which is currently used in the STD design. The Zircaloy grids have thicker straps than the Inconel. Also, the Zircaloy grid height is 2.25 inches as compared to the Inconel grid which is 1.32 inches. These dimensional changes were made to compensate for differences 1.n material strength properties. The Zircaloy grid incorporates the same grid cell support k

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configuration as the Inconel grid (six support locations per cell, four dimples, and two springs).

With the exception of a reduction in diameters above the dashpot, the OFA guide thimbles are the identical to those in the STD design. An 8-m11 reduction to the guide thimble OD and ID is required due to the thicker Zircaloy grid straps, resulting in a reduced cell size.

The OFA guide thimble tube ID provides an adequate nominal diaretral clearance of 69 mils for the control rods. Even though there is a reduced clearance between the STD and 0FA designs, the rod drop time used for accident analysis, 3.3 seconds (to dashpot), is applicable for both the OFA and STD designs. The reduced 0FA thimble tube ID also provides sufficient diametral clearance for burnable absorber rods, source rods, and dually compatible thimble plugs; however, the STD thimble plugs used in McGuire Unit 1, Cycle 1 are not compatible with the OFA guide thimbles and must be replaced with dually compatible thimble plugs.

The OFA top nozzle design is identical to that of the STD top nozzle.

The 3-leaf holddown springs designed for the STD top nozzle. continues to meet all applicable design criteria for the OFA design, even though the OFA weighs less than the STD assembly. Since the nozzle design is identical, existing fuel handling tools will be compatible with the OFA design.

The OFA bottom nozzle design is identical to the STD design used on McGuire 1 and 2 Cycle 1 fuel asseblies with exception of a reconstitu-tion feature which involves a design modification to the thimble screw heads and mating surface with the underside of the bottom nozzle top plate. The instrumentation tube counterbore remains unchanged allowing a larger radial clearance with the instrumentation tube than in the STD bottom nozzle assembly. This additional clearance has been evaluated and shown to meet all applicable design criteria. The OFA bottom nozzle design has a reconst'itution feature which facilitates easy removal of the nozzle from the fuel assembly.

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The OFA design was extensively tested to determine the mechanical characteristics under lateral, axial, torsional, and dynamic loading conditions. The information obtained from these tests was used to verify fuel assembly design models and to provide fundamental mechanical constants for various accident analyses. Test summaries and results, used for the verification of th^e OFA design, are detailed in Reference 4.

The most probable causes of significant rod bow are attributed to grid spring and pellet-clad interaction forces. For the OFA design, reduced grid spring forces (due to Zircaloy grids), a larger fuel tube thickness-to-diameter ratio (t/d), and wider channel spacings between fuel rods as compared to the STD design, should tend to decrease rod bow. The current NRC approved methodology for comparing rod bow for two different fuel assembly designs is given in Reference 5.

Fuel rod wear is dependent on both the support conditions and the flow environment to which.the fuel rod is subjected. Due to the OFA and STD designs employing different grids, there is an unequal axial pressure distribution between the two assemblies. Crossflow resulting.from this unequal pressure distribution was evaluated to determine the induced rod vibration and subsequent wear. Hydraulic tests were performed to verify that the crossfl.ows were negligible and also to check hydraulic compati-bility of the OFA and the STD designs. Results from the tests showed that insignificant fuel rod wear occurred because of crossflow between the two assemblies.

The core components in McGuire I and 2 are designed to be compatible with both the OFA and STD designs. The reduced-diameter OFA thimble tube provides sufficient clearance for insertion of control rods, burnable poison rods, source rods, and dually compatible thimble plugs to assure the proper operation of these core components.

The thimble plugs utilized by the plugging devices, source assemblies, and burnable absorber assemblies for McGuire Unit 1, Cycle 1 are of a STD design which are not mechanically compatible with the OFA 300aF:6' 7

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design. A thimble plug has been designed to be compatible with both the STD.and the OFA designs from both a mechanical and thermal / hydraulic Eperspective and will be implemented for McGuire Unit 1, Cycle 2. All thimble plugs used at McGuire ' nit 2 are of the dually compatible J

design.

The fuel has been designed according to the fuel performance model in Reference 6.

The fuel is designed to operate so that clad flattening will not occur as predicted by the Westinghouse model, Reference 7.

The fuel rod internal pressure design basis, Reference 8, is satisfied for all' fuel regions.

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O

o 4.0 NUCLEAR EVALUATION The transition from the STD to the OFA design will not result in changes from the current ~ nuclear design bases given in the McGuire FSAR and applied to subsequent Unit 1 and Unit 2 Reload Safety Evalutions (RSE)s.

1 Although the physics characteristics are slightly different for the OFA fuel when compared to STD, evaluations show that differences are within the normal range of variations seen from cycle to cycle.

The methods'and core models used in the reload transition analysis are identical to those employed and described in References 1, 2, and 9.

These are the same methods.and models which have been used in other Westinghouse plants' reload cycle designs. No changes to the nuclear design philosophy, methods, or models are necessary due to the transition to 0FA fuel.

A number of changes to the McGuire 1 and 2 Technical Specifications will be proposed as part of the transition to 0FA fuel. Some of these changes, whether directly related to 0FA fuel or not, impact the core nuclear design. These changes include: (1) the positive moderator temperature coefficient (MTC) specification; (2) the 0.3 multiplier in the F limit function; (3) a reduction in the required shutdown AH margin (SDM) to 1.3% ap; and (4) F Surveillance.

g The key safety parameters evaluated for the conceptual designs show that the. expected ranges of variation for many of the parameters will lie within the normal cycle-to-cycle variations observed for past STD fuel reload designs. The parameters which fall outside of these ranges are those sensitive to fuel type, e.g.,

the MTC.

Power distributions and peaking factors are primarily loading pattern-dependent. The usual methods of enrichment variation, spent burnable poison (BP) usage, and fresh BP usage can be employed in the transition and full 0FA cores to ensure compliance with the peaking factor Technical Specifications.

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J 5.0 THERMAL AND HYDRAULIC EVALUATION

,The analysis of both STD and 0FA fuel includes the use of the WRB-1 DNB

correlation, Reference 10, and the Improved 1 Thermal Design Procedure, (ITDP), Reference 11. The WRB-1 DNB Correlation takes credit for the significant improvement in the accuracy of critical heat flux predic-tions over previous DNB correlations with a DNBR limit of 1.17 being applicable for this correlation.

The design method employed to meet the DNB design basis is the Improved Thermal Design Procedure. ' Uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95 percent probability that the minimum DNBR will be greater than or equal to 1.17 for the limiting power rod. Plant parameter uncertainties are used to determine the plant DNBR uncertainty. This DNBR uncertainty, combined with the DNBR limit, establishes a DNBR value which must be' met in plant safety analyses. Since the parameter uncertainties are considered in determining the design DNBR value, the plant safety analysas are performed using values of input parameters without uncertainties. For this application, the min:mim required design DNBR values are 1.32 for thimble coldwall cells (three fuel rods and a thimble tube) and 1.34 for typical cell (four fuel rods).

In addition to the above considerations, a specific plant DNBR margin has been considered in the analysis. In particular, the DNBR values of 1.47 and 1.49, for thimble and typical cells respectively, were employed in the safety analyses. The DNBR margin between the DNBRs used in the safety analyses and the design DNBR values (1.32 for thimble cells and 1.'4 for typical cells) will be used for the flexibility in the design, operation,' analyses, and transition core DNB effects of the fuel.

The STD and DFA designs have been shown to be hydraulically compatible in Reference 4.

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The phenomenon of fuel rod bowing, as described in Reference 5, must be accounted for in the DNBR safety analysis of Condition I and Condition II events for each plant application. Applicable generic credits for margin resulting from retained conservatism in the evaluation of DNBR and/or margin obtained from measured plant operating parameters (such as F r c re fl w) which are more restrictive H

than those required by the plant safety analysis, can be used to offset the effect of rod bow.

The safety analysis for McGuire 1 and 2 maintained sufficient margin between the safety analysis limit DNBRs (1.47 and 1.49 for thimble and typical cells, respectively) and the design limit DNBRs (1.32 and 1.34 for thimble and typical cells, respectively) to accommodate full-flow and-low-flow DNBR penalties identified in Reference 12, which are applicable t', 17x17 0FA analysis utilizing the WRB-1 DNB correlation.

The transition : ore DNB methodology given in Reference 2 has been approved by the USNRC. Using this methodology, transition cores will be analyzed as if they were full-core OFAs, applying the two percent DNBR transition core penalty. This penalty will be included in the safety analysis limit DNBR such that sufficient margin over the margin limit DNBR will exist to accommodate the transition core DNBR penalty and the appropriate rod bow DNBR penalty.

The fuel temperatures for use in safety analysis calculations for the OFA fuel are not significantly different from those for the STD fuel.

These small differences will not adversely affect the safety analysis calculations. Westinghouse uses the PAD fuel performance code, Reference 6, to perform both design and licensing calculations. When the code is used to calculate fuel temperatures to be used as initial conditions in safety analyses,'a conserva+'ve thermal safety model, Reference 13, is used.

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6.0 ACCIDENT EVALUATION This section addresses the impact on accident analyses of the following proposed changes for McGuire Units 1 and 2:

OFA Positive MTC F

Multiplier Change AH Reduction in Shutdown Margin 0.fA The principal mechanical design characteristic of the OFA design which could have an effect on accidents is the smaller fuel rod. This leads to a higher fuel rod temperature, surface heat flux, and a DNB penalty.

The larger hydraulic diameter and lower coolant flow velocity cause a reduction in heat transfer after DNB. The smaller fuel rod also leads to a faster heatup rate for severe reactivity transients such as Rod Cluster Control Assembly (RCCA) ejection.

As'a result of the smaller fuel rod, for the same power level, the OFA

- design will have a lower DNB ratio than the STD design.

The DNB penalty was offset for the 0FA core through the use of the WRB-1 DNB correlation, Reference 10,and the Improved Thermal Design Procedure, Reference 11. Those transients impacted by the OFA design are shown in Table 1.

A discussion of loss-of-coolant accidents (LOCA) is addressed later in this section.

Positive MTC The present McGuire Technical Specifications require the MTC to be zero or negative at all times while the reactor is critical. This requirement is overly restrictive, since a small positive coefficient at 3004F:6 12 L

reduced power levels could result in a significant increase in fuel cycle flexibility, but would have only a minor effect on the safety analysis of the accident events presented in the FSAR.

The proposed Technical Specification change, given in Section 7.0, allows a +5 pcm/*F MTC below 70 percent of rated power, changing to a 0 pcm/*F MTC at 70 percent power and above. A power-level dependent MTC was chosen to minimize the effect of the specification on postulated accidents at high power levels. Moreover, as the power level is raised, the average core water temperature becomes higher as allowed by the b

programmed average temperature for the plant, tending to make the moderator coefficient more negative. Also, the boron concentration can be reduced as xenon builds into the core. Thus, there is less need to allow a positive coefficient as full power is approached. As fuel burnup is achieved, boron is further reduced and the moderator temperature coefficient will become negative over the entire operating power range.

The impact of a positive MTC on the accident analyses presented in Chapter 15 of the McGuire FSAR, Reference 14, has been assessed. Those incidents which were found to be sensitive to minimum or near-zero moderator coefficients were reanalyzed. In general, these incidents are limited to transients which cause reactor coolant temperature to increase. With one exception, the analyses presented herein were based on a +5 pcm/ F MTC, which was assumed to remain constant for variations in temperature.

The control rod ejection analysis was based on a coefficient which was at least +E pcm/*F at zero power nominal average temperature, and which became less positive for higher temperatures. This was necessary since the TWINKLE computer code, on which the analysis is based, is a diffusion-theory code rather than a point-kinetics approximation and the moderator-temperature feedback cannot be artificially hold constant with 3004F:6 13

temperature. For all accidents which were reanalyzed, the assumption of a positive MTC existing at full power is conservative, since as noted in

.Section 7.0, the proposed Technical Specification requires that the coefficient be zero or negative at or above 70 percent power.

F Multiplier Change AH A proposed change from K=0.2 to K=0.3 in the following equation for the Nuclear Hot Channel Factor ((H) was evaluated with regard to

-its effect on accident analyses:

F

$ 1.49 [1.0 + K(1-P)] for ITDP transients H

$ 1.55 [1.0 + K(1-P)] for non-ITDP transients H

where P is the fraction of full power and the K multiplier is the power correction constant.

The effect on accident analyses is through the core safety limits at very high pressure and low power levels. Since the steam generator safety valves prevent the plant from reaching these limiting conditions, the protection setpoints are unaffected by this change. The change sometimes impacts the axial offset envelope such that the f(AI) changes. However, no credit for the f(AI)' protection is assumed in the accident analyses. Therefore, the safety analyses are not impacted by the proposed F multiplier char.ge.

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4 Reduction in Shutdown Margin The proposed Technical Specification change for Shutdown Margin from 1.6 percent Ap to 1.3 percent Ap has been determined to be acceptable for most Westinghouse 4-loop plants. Those accidents affected by this change are shown in Table 1.

The impact of the proposed changes as identified earlier in _this section has been assessed for the non-LOCA as provided in Chapter.15 of the McGuire FSAR. The following accidents have been reanalyzed:

Excessive Heat Removal Due to Feedwater System Malfunction Excessive Load Increase Main Steamline Depressurization Main Steamline Rupture Loss of Load / Turbine Trip Station Blackout Loss of Normal Feedwater Main Feedline Rupture Partial Loss of Forced Reactor Coolant Flow Complete Loss of Forced Reactor Coolant Flow Single Reactor Coolant Pump Locked Rotor Uncontrolled RCCA Bank Withdrawal from a Suberitical Condition Uncontrolled RCCA Bank Withdrawal at Power Startup of an Inactive Re~ actor Coolant Loop RCCA Ejection Inadvertent ECCS Operation at Power Reactor Coolant System Depressurization For each of the accidents analyzed, it was found that'the appropriate safety criteria are met.

Also, the RCCA misoperation events were not analyzed as part of the McGuire RTSR effort, but rather as part of a generic study, Reference 115, which is applicable to McGuire 1 and 2.

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4 The following two accidents have not been reanalyzed as part of the McGuire OFA transition:

Uncontrolled Boron Dilution Inadvertent Loading of a Fuel Assembly into an Improper Position The' Boron Dilution event contained in the McGuire FSAR was analyzed at Refueling, during Startup, and at Power. In each case the driving forces were the boron worth coefficient, the initial boron concentration, and the critical boron concentration. Each of these driving forces remain unchanged for the OFA design. Therefore, the conclusions reached in Section 15.2.4 of the McGuire FSAR for the Uncontrolled Boron Dilution event remain valid and no new analysis is required.

Assembly average power distributions with 0FA fuel are essentially the same as with STD fuel. Therefore, the power distribution anomalies seen Las the result of misloading a fuel assembly with 0FA fuel will be no different than those.seen~using STD fuel. Hence, the results and conclusions reached in Section 15.3.3 of the McGuire FSAR-for the Inadvertent Loading of a Fuel Assembly into an Improper Position remain-unchanged and no new analysis is requ. ired.

.The large break LOCA analysis applicable for transition and full 0FA core cycles of McGuire 1 and 2 was performed utilizing the OFA design.

This-is consistent with the methodology given in Reference 2 for the OFA transition. The currently approved UHI Large Break ECCS Evaluation Model was utilized for this analysis, and five cold leg breaks were reanalyzed.

-The LOCA analysis performed assumes a full core of 0FAs and conserva-tively applies to transition cores. The OFA design is very similar hydraulically to the STD design it replaces. Differences in total hydraulic resistance between the two designs is less than 1%.

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s Evaluation of hydraulic mismatches of less than 10% have shown an insignificant effect on blowdown cooling, such that the impact on reflood cooling alone needs to be considered.

Since the overall resistance of the two types of fuel is essentially identical, only the crossflows during core reflood due to the smaller rod size and different grid designs need be evaluated. The maximum flow reduction due to crossflow calculated tc occur in the OFA is ~2.9%.

Analyses have been performed which demonstrate that a 5% reduction leads to a maximum PCT increase of 19*F.

Therefore, the PCT increase due to crossflow between adjacent 0FA and STD assemblies would be approximately 11*F.

This effect can be offset in the McGuire 1 and 2 transition cores by considering the favorable UHI quench characteristics of the STD design. Quenching of fuel throughout the core during blowdown is calculated using UHIPOWERREGIONS LOCTA, with computed parameters then being input to UHIWREFLOOD. If the STD design is modeled the quench parameters significantly improve, leading to a faster reflooding of the core than is true for the OFA case. The magnitude of this benefit is several times the 11*F penalty identified for transition cyclesi because of this benefit no transition core penalty need be applied. Two further reasons why this method is indeed conservative for transition cores are:

1.

The increase in core flow area associated with 0FA due to the smaller rod diameter has an important impact on flooding rates during reflood. Full 0FA core representation decreases core flooding rates, which reduces heat transfer coefficients.

2.

The OFA design has a higher volumetric heat generation rate than STD design. The analysis assumes that the OFA has the hottest rod and maximum F which maximizes the calculated PCT.

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9

'For breaks up to and including the double ended severance of a reactor coolant pipe, the emergency core cooling system will meet the acceptance criteria as presented in 10 CFR 50.46. That is:

1.

The calculated peak fuel element clad temperature is below the requirement of 2200*F.

2.

The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1% of the total amount of Zircaloy in the reactor.

3.

The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. The localized clad-ding oxidation limit of 17% is not exceeded during or after quenching.

4.

The core remains amenable to cooling during and after the break.

5.

The core temperature is reduced and decay heat is removed for an extended period of time as requireu by the long-lived radio-activity remaining in the core.

(he Large Break 0FA LOCA analysis for McGuire 1 and 2 utilizing the currently approved UHI Evaluation Models resulted in a PCT of 2175'F at 2.15 F for the CD = 0.6 (Imperfect mixing) DECLG break. The small q

impact of crossflow for transition core cycles is offset by the presence of STD fuel in the core so that margin to 10 CFR 50.46 limits remains in transition cycles.

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7.0

SUMMARY

OF TECHNICAL SPECIFICATION CHANGES 1.

A list.of_the Technical Specifications affected by the use of the OFA

'esign and positive MTC is provided'in Table 2.

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8.0 REFERENCES

1. Bordelon, F. M., et al., " Westinghouse Reload Safety Evaluation Methodology," WCAP-9272 (Proprietary) and WCAP-9273 (Non-Proprietary), March 1978.

2. Davidson, S. L.; Iorii, J.

A., " Reference Core Report - 17x17 Optimized Fuel Assembly," WCAP-9500-A, May 1982.

3. Skaritka, J.; Iorti, J. A., " Operational Experience with Westinghouse Cores," WCAP-8183, Latest Revision.

4. Davidson, S. L.; Iorii, J. A., (Eds.), " Verification Testing and Analyses of the 17x17 Optimized Fuel Assembly," WCAP-9401-P-A and WCAP-9402, March 1979.

5. Skaritka, J., (Ed.), " Fuel Rod Bow Ev21uation," WCAP-8691, Revision 1, July 1979.

6. Miller, J. V., " Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations," WCAP-8720, October 1976.

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

8. Risher, D.

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

-9. Camden, T. M., et al., "PALADON-Westinghouse Nodal Computer Code,"

WCAP-9485A (Proprietary) and WCAP-9486A (Non-Proprietary), December 1979, and Supplement 1, September 1981.

10. Motley, F. E.; Hill, K. W., et al., "New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids," WCAP-8762, July 1976.

3004F:6' 20

11. Chelemer,- H. ; Boman, L. H.; Sharp, D. R., " Improved Thermal Design Procedure," WCAP-8567P, July 1975.

12. " Partial Response to Request Number-l'for Additional Information on WCAP-8691, Revision 1" Letter, E. P. Rahe, Jr. (Westinghouse) to J.

R. Miller (NRC), NS-EPR-2515, dated October 9, 1981; " Remaining Response to Request Number I for Additional Information on WCAP-8691, Revision 1," Letter, E. P. Rahe, Jr. (Westinghouse) to J. R. Miller (NRC), NS-EPR-2572, dated March 16, 1982.

13. Letter from E. P. Rahe (Westinghouse) to C. O. Thomas (NRC),

NS-EPR-2673, " Westinghouse Revised PAD Code Thermal Safety Model,"

WCAP-8720,. Addendum 2, October 27,1982,(Proprietary).

14. Duke Power Company, McGuire Nuclear Station Units 1 and 2, Final Safety Analysis Report, Docket Nos. 50-369/370, as amended.

15. Morita, T., et al., " Dropped Rod Methodology for Negative Flux Rate Trip Plants," WCAP-10297-P-A, June 1983.

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TABLE 1 FSAR CHAPTER 15 ACCIDENT ANALYSIS SENSITIVITY TO PROPOSED CHANGES Accident OFA Positive MTC Reduced SDM

. Feedwater Malfunction X

Excessive Load Increase -

X Steamline Depressurization X

X Steamline Rupture X

X Loss'of Load X

X Station Blackout X

Loss of Normal Feedwater X

- Feedline Rupture X

Partial Loss of Flow X

X Complete Loss of Flow X

X Locked Rotor X

RCCA Withdrawal from Subcritical X

X RCCA Withdrawal at Power X

Startup of an Inactive Loop X

RCCA Ejection X

X Inadvertent ECCS Operation X

RCS Depressurization X

X 3004F:6 22 L

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TABLE 2

SUMMARY

OF AFFECTED TECHNICAL ~ SPECIFICATIONS Section Description Reason for Change 2.1.1 Reactor Core Safety Limits New Core Limits for Opti-mized Fuel 2.2.1 Reactor Trip System Setpoints New Core Limits (OTAT and OPAT only) 3/4.1.1.1 Shutdown Margin

.T,yg _> 200*F Reduced Shutdown Margin 3/4.1.1.3 Moderator Temperature Positive Moderator Coefficient Temperature Coefficient 3/4.1.3.4 Rod Drop Time Longer rod drop time (bounding value) used in the analysis 3/4.2.2 Heat Flux Hot Channel Factor -

New LOCA Analysis for F (Z)

Optimized Fuel and g

Inclusion of Fg Surveillance 3/4.2.3 RCS Flow Rate Minimum Measured Flowrate Assumed with the Improved Thermal Design Procedure 3/4.5.1 Accumulators - Cold Leg Injection New LOCA analysis for 0FA 6.9.1.12 Radial Peaking Factor F Surveillance g

Limit Report 3004F:6 23

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17x17 RECONSTITUTABLE STD. FUEL ASSEMBLY I

FIGURE 1 - 17x17 0FA/STD FUEL ASSEMBLY COMPARISON 24

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