Text

Reload Safety Evaluation,Mcguire Nuclear Station Unit 1 Cycle 3
	ML20113H637
Person / Time
Site:	McGuire
Issue date:	12/31/1984
From:	Novendstern E, Schueren P; DUKE POWER CO.
To:	;
Shared Package
ML20113H604	List: ML20113H601; ML20113H622; ML20113H637;
References
	TAC-56746, TAC-56747, NUDOCS 8501250241
	Download: ML20113H637 (71)
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ATTACHMENT 2A P.ELOAD SAFETY EVALUATION MCGUIRE NUCLEAR STATION UNIT 1 CYCLE 3 December, 1984 Edited by:

Pc Schueren Approved:

M E.H.Novendster[ Manager Thermal Hydraulic Design Nuclear Fuel Division I

RBR2RE85*oIS8$h; P

l t

1831L:6/841218

._m TABLE OF CONTENTS Title Page

1.0 INTRODUCTION

AND

SUMMARY

1 1.1 Introduction 1

1.2 General Description 1

1.3 Conclusions 2

2.0 REACTOR DESIGN 3

2.1 Mechanical Design 3

2.2 Nuclear Design 3

2.3 Thermal and Hydraulic Design 4

3.0 POWER CAPABILITY AND ACCIDENT EVALUATION 6

3.1_ Power Capability 6

3.2 Accident Evaluation 6

3.2.1 Kinetic Parameters 7

3.2.2 Control Rod Worths 7

3.2.3 Core Peaking Factors 7

3.3 Reduced RCS Flow 7

3.4 LOCA Analysis 15 4.0 TECHNICAL SPECIFICATION CHANGES 18

5.0 REFERENCES

19 APPENDIX A - Technical Specification Page Changes I

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i LIST OF TABLES Table Titie Page 1

Fuel Assembly Design Parameters 20 2

Kinetic Characteristics 21 3

Shutdown Requirements and Margins 22 4

Control Rod Ejection Accident Parameters 23 LIST OF FIGURES I

Figure Title Page 1

Core Loading Pattern and Source and 24 Burnable Absorber Locations 11 1831L:6/841218 e

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1.0 INTRODUCTION

AND

SUMMARY

1,1, INTRODUCTION

-This report presents an evaluation for McGuire Unit 1, Cycle 3, 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-9273, " Westinghouse Reload Safety Evaluation Methodology"(I)

-McGuire Unit 1 is operating in Cycle 2 with Westinghouse 17x17 low parasitic (STG) and optimized fuel assemblies (OFA).

For Cycle 3 and subsequent cycles, it is planned to refuel the McGuire Unit 1 core with Westinghouse 17x17 optimized fuel assembly (OFA) regions.

In the OFA 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 s

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 3 safety evaluation.

All of the accidents comprising 'the licensing bases (2,3) which could potentially be cffected by the fuel reload have been reviewed for the Cycle 3 design described herein. The results of new analyses and the

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justification for the applicability of previous results for the S

remaining analyses is presented.

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.1.2 GENERAL DESCRIPTION I

' The McGuire Unit 1, Cycle 3 reactor core will be comprised of 193 fuel 4

-assemblies arranged in the core loading pattern configuration shown in Figure 1.

During the Cycle 2/3 refueling, 60 STD fuel assemblies will 1831L:6/841218 1

l be replaced with 60 Region 5 optimized fuel assemblies. A summary of the Cycle 3 fuel inventory is given in Table 1.

As in Cycle 2, this cycle will contain one Region 4 demonstration assembly, designa.ted 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 14.

Nominal core design parameters utilized for Cycle 3 are as follows:

Core Power (MWt) 3411 System Pressure (psia) 2250 Core Inlet Temperature (*F) 558.9 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 3 design does not cause the previously' acceptable safety limits to be. exceeded. This conclusion is based on the following:

1.. Cycle 2 burnup is between 10200 and 10577 MWD /MTU.

2.

Cycle 3 burnup is limited to 11700 MWD /MTU including a coastdown.

3.

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

4.

The proposed Technical Specification changes discussed in Section 4.0 of this report and provided in Appendix A are approved.

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2.0 REACTOR DESIGN 2.1 MECHANICAL DESIGN The Region 5 fuel assemblies are Westinghouse OFAs. The mechanical description and justification of their compatibility with the Westinghouse STD design ir. a transition core is presented in the OFA transition l'icensing submittal.(2)

Table 1 presents a comparison of pertinent design parameters of the various fuel regions. The Region 5 fuel has been designed according to the fuel performance model(4)

The fuel is designed and operated so that clad flattening will not occur, as predicted by the Westinghouse clad.flatteningmodel(5)

For all fuel regions, the fuel rod internal pressure design basis, which is discussed and shown acceptable in Reference 6, is satisfied.

Westinghouse has had considerable experience with Zircaloy clad fuel.

This experience is described in WCAP-8183, " Operational Experience with Westinghouse Cores."(7) Operating experience for Zircaloy grids has also been obtained from~six demonstration 17x17 0FAs(2), four

~ demonstration 14x14 0FAs(2) and a full region of 0FA fuel in the McGuire Unit 1 Cycle 2 design.

2.2 NUCLEAR DESIGN The. Cycle 3 core loading is designed to meet a F (z) x P ECCS limit of g

< 2.26 x K(z).

Relaxed Axial Offset Control (RAOC) will be employed in Cycle 3 to

enhance operational flexibility during non-steady state operation.

RAOC makes use of available margin by expanding the allowable AI band, particularly at reduced power. The RAOC methodology and application is fully described in Reference 8.

The analysis for Cycle 3 indicates that no_ change to the safety parameters is required for RAOC operation.

During operation at or near steady. state equilibrium conditions, core 1831L:6/841218 3

i peaking factors are significantly reduced due to the limited amount of xenon skewing possible under these operating conditions.

The Cycle 3 Technical Specifications recognize this reduction in core peaking factors through the use of a Base Load Technical Specification.

Adherence to the F limit is obtained by using the F Surveillance g

g Technical Specification, also described in Reference 8.

This provides a more convenient f;orm of assuring plant operation below the F limit g

while retaining the intent of using a measured parameter to verify operation below Technical Specification limits.

F surveillance is g

only a surveillance requirement and as such has no impact on the results of the Cycle 3 analysis or safety parameters.

Table 2 provides a summary of Cycle 3 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 352 burnable absorber (BA) rods located in 52 BA rod assemblies.

Location of the BA rods are shown in Fi~gure 1.

2.3 THERMAL AND HYDRAULIC DESIGN The thermal hydraulic methodology, DNBR correlation and core DNB limits used for ' Cycle 3, are consistent with the current licensing basis (2).

i The thermal hydraulic safety analyses used for Cycle 3 are based on a reduced design flow rate in comparison to Reference 2.

No significant l

l variations in thermal margins will result from the Cys le 3 re' load.

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The thermal-hydraulic methods used to analyze axial power distributions generated by the RXOC methodology are similar to thoset 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 imbalance as a function of power level from these considerations were found to be less restrictive than those resulting from LOCA F considerations.

g 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 con-sequences 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

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pwer level (Section 1.0) during Cycle 3.

For the overpower transient, the fuel; centerline temperature limit of 4700 F can be accommodated with margin in the Cycle 3 core. The time dependent densification model(9) was used for fuel temperature evaluations. The LOCA limit at rated _ power can be met by maintaining F (z) at or below 2.26 x K(z).

n 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 the previous applicable safety analysis, or reanalysis as described in Section 3.3.

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 3 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 KINETICS PARAMETERS Table 2 is a summary of the kinetics parameters current limits along with the associated Cycle 3 calculated values. All of the kinetics values fall within the bounds of the current limits.

3.2.2 CONTROL ROD WORTHS Changes in control rod worths may affect differential rod worths, shut-down margin, ejected rod worths, and trip reactivity. Table 2 shows that the maximum differential 1 rod worth of two RCCA control banks moving together in their highest worth region for Cycle 3 meets the current limit.

Table 3 shows that the Cycle 3 shutdown margin requirements have b'een satisfied. Table 4 is a summary of the current limit control rod

~ ejection analysis parameters and the corresponding Cycle 3 values.

3.2.3 CORE PEAKING FACTORS Peaking factors for.the dropped RCCA incidents were evaluated based on the NRC approved dropped red methodology described in Reference 10.

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.

3.3 REDUCED RCS FLOW

. The safety analyses performed in support of the OFA transition licensing

-submittal (2) assumed a Thermal Design Flow of 386,000 gpm.

For Cycle 3, the TDF will be 382,000 gpm. This represents an approximate 1 percent' reduction in the RCS flow used for the OFA transition licensing submittal (2),

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The following safety evaluation confirms the acceptability of operation at 100 percent of rated thermal power and 99 percent of the RCS flow assumed in the.0FA transition analyses. All of the affected FSAR Chapter 15 accidents and protection system setpoints have been reviewed to determine the impact of the proposed reduction in flow requirement.

In addition, Technical Specification changes required to support the reduced flow are included in Section 4.0.

3.3.1 DNB CONSIDERATIONS The core DNB limits have been verified to be unchanged from the current values, and the conclusion that the DNB basis is met for the following. transients remains valid:

Excessive Heat Removal Due to Feedwater System Malfunction Excessive Load Increase Main Steamline Depressurization Main Steamline Rupture Loss of Load / Turbine Trip Partial Loss of Forced Reactor Coolant Flow Complete Loss of Forced Reactor Coolant Flow Uncontrolled RCCA Bank Withdrawal from a Subcritical Condition Uncontrolled RCCA Bank Withdrawal at Power Startup of an Inactive Reactor Coolant Loop Inadvertent ECCS Operation at Power Reactor Coolant System Depressurization 3.3.2 NON-DNB CONSIDERATIONS In addit' ion to the DNB concern, the following evaluations are presented for those accidents which are not DNB related or for which DNBR is not the only safety criterion of interest.

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r Uncontrolled RCCA Bank Withdrawal from a Subcritical Condition A control rod assembly withdrawal incident when the reactor is subcritical results in an uncontrolled addition of reactivity leading to a power excursion (Section 15.2.1 of the FSAR). The nuclear power response is characterized by a very fast rise terminated by the reactivity feedback of the negative fuel temperature coefficient.

The power excursion caused a heatup of the moderator. However, since the power rise is rapid and is followed by an immediate reactor trip, the moderator temperature rise is small. Thus, nuclear power response is primarily a function of the Doppler temperature coefficient. An increase in temperature due to reduced RCS flow would result in more Doppler feedback, thus reducing the nuclear power excursion as calculated in the current analysis which partially compensates for the flow reduction.

The OFA transition analysis shows that for a reactivity insertion rate

-5 of 75 x 10 delta-K/sec, the peak hot spot heat flux achieved is 179.4 percent of nominal with a resultant peak feel average temperature of 2242 F, and a peak clad temperature of 726*F. A 1 percent reduction of reactor coolant flow'would degrade heat transfer from the fuel by a maximum of 1 percent. Thus, peak fuel and clad temperatures would also increase by a maximum 1 percent, yielding maximum fuel and clad temperatures which are still significantly below fuel melt (4800*F) and zirconium-H O reaction (1800 F) limits. Therefore, the conclusions 2

presented in the current licensing submittal (2) are still valid.

Boron ~ Dilution The results of the boron dilution transient will remain unchanged for all modes of operation due to a reduction in reactor coolant flow.

The maximum dilution flow rate, RCS active volumes, and RCS boron concentrations are not impacted by a reduction in flow.

Since these parameters determine the amount of time available to the operator to terminate the dilution event, the results presented in the FSAR remain unchanged.

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I Loss of Load The' loss of load accident is presented in Section 15.2.7 of the FSAR and can result from either loss of external electrical load or a turbine trip. The result of a loss of load is an increase in core power which exceeds'the secondary system power extraction, thus causing an increase in core water temperature. A reduction in RCS flow will result in a more rapid pressure rise than that calculated in the OFA transition analysis. The effect will be minor, however, since the reactor is tripped on high pressurizer pressure. Thus, the time to trip will be decreased, which will result in a lower total energy input to the coolant. The analysis shows a peak pressurizer pressure of 2567 psia.

-A 1 percent reduction in flow will lead to a conservative increase in system pressure to less than 2580 psia. The pressurizer will not fill, and the maximum pressures are within the design limits. Therefore, operation at reduced flow will not violate safety limits following a loss ~of load accident.

Loss of Normal Feedwater/ Station Blackout This transient is analyzed to demonstrate that the peak RCS pressure

.does not exceed allowable limits and that the core remains covered with water..These criteria are assured by applying the more stringent requirement that the pressurizer must not be filled with water. The effect of reducing initial core flow results in an initial more rapid heatup of the RCS..The resultant coolant density decrease increases the volume of: water in the pressurizer. These transients have been reanalyzed with the reduced flow assumption.

In addition, the-low-low steam generator level'setpoint will be' revised and a filter added to the channels 'to help prevent unnecessary reactor trips as a result.of load

. rejections. These changes'have been incorporated into the reanalysis, and appropriate Technical. Specification changes are identified in Section 4.0.

The results show considerable margin to filling the

. pressurizer.

Therefore, all safety criteria are met for the events.

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Steemline Break

-The steamline break transient is analyzed at hot zero power, end-of-life conditions for the following cases:

Inadvertent opening of a steam dump, safety, or relief valve (Section 15.2.13 of the FSAR)

Main steam pipe rupture with and without offsite power available (Section 15.4.2 of the FSAR)

A steamline break results in a rapid depressurization of the steam generators and primary side cooldown. This causes a large reactivity insertion due to the presence of a negative moderator temperature coefficient. A reduction in reactor coolant flow will result in a reduction in heat transfer from the fuel to the coolant. Therefore, the reactivity insertion and return to power in the double-ended rupture case for reduced flow conditions would be less limiting that the cases case, the time of

. presented in the FSAR.

For the double-ended ruptur:3 safety injection actuation is unaffected by reduced coolaat flow.

This, coupled with a slower return to power would result in a significant reduction in peak average power from the FSAR results.

The main steam depressurization case is bounded by the double-ended rupture.

Since the return to power is less severe and the DNB evaluations remain valid as

-previously stated, the conclusions presented in the current licensing submittal (2) are still valid for a 1 percent reduction in reactor coolant flow.

Rupture of a Main Feedwater Line This transient is analyzed to demonstrate that the peak RCS pressure does not exceed allowable limits and that the core remains covered with water. These criteria are assured by applying the more stringent requirement that bulk voiding does not occur at the outlet of the core.

The'effect of reducing initial core flow results in an initial more rapid heatup of the reactor coolant system-(RCS). This transient has 1831L:6/841218 11

been reanalyzed with the reduced flow assumption.

In addition, the low-low steam generator level setpoint will be revised and a filter added to the cha'nnels to help prevent unnecessary reactor trips as a result of load rejections. These changes have been incorporated into the reanalysis, and appropriate Technical Specification changes are identified in,Secticn 4.0.

The results show considerable margin to hot leg saturation. Therefore, all safety criteria are met for the event.

Locked Rotor Following a locked rotor, reactor coolant system temperature rises until shortly after reactor trip. A reduction in RCS flow will not affect the time to DNB since DNB is conservatively assumed to occur at the beginning of the transient. The flow reduction in the affected loop is so rapid that the time of reactor trip on low flow does not change due to the 1 percent reduction in reactor coolant flow. Therefore, the nuclear power and heat flux transients will not change from those presented in the FSAR. However, the reduction in flow will result in slightly higher system pressures and clad temperatures. The peak RCS pressure calculated in the 0FA transition analysis was 2593 psia. A 1 percent reduction in reactor coolant flow would cause a conservative increase in pressure to less than 2620 psia, which is still signiff-l cantly below the pressure at which vessel stress limits are exceeded.

The peak clad temperature calculated in the OFA transition analysis is i

1964 F, well below the limit of 2700 F, and shows that a slight increase in this parameter due to reduced RCS flow can be easily accommodated.

l Therefore, the conclusions presented in the current licensing submittal (2) are still valid.

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Control R'od Ejection l

The rod ejection transient is analyzed at full power and hot standby for l

both beginning and end-of-life conditions (Sections 15.4.6 of the f

FSAR). A reduction in core flow will result in a reduction in heat l

transfer to the coolant, which will increase peak clad and fuel temperatures and peak fuel stored energy. However, all cases have 1831L:6/841218

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margin to fuel failure limits. The effect of reducing reactor coolant flow is to increase the peak clad temperatures.

The analysis shows that, for the worst case, there is sufficient conservatism in the analysis assumptions and margin in the results such that the peak clad temperature limit (2700*F) is not violated with the reduced flow. This was verified by a reanalysis of the limiting end-of-life zero power case. The peak clad temperature calculated for this case in the OFA transition analysis was 2685'F. The reanalysis of this case assumed the reduced RCS flow, but used shorter time steps to remove some conservatism in the calculation of the nuclear power transient. The result was a peak clad temperature of 2683'F.

Thus, the limit is not violated. The fuel temperatures and peak fuel stored energy will also increase slightly due to the 1 percent decrease in reactor coolant flow. However, there is sufficient margin between the analysis results and the limits to accommodate the effects of the reduced flow.

Therefore, the conclusions presented in the current lic,ensing submittal (2) are still valid.

LOCA Analysis A LOCA analysis has been performed for McGuire Unit 1 that uses the reduced Thermal Design Flow.

Results of the analysis are given in Section 3.4.

Technical Specification Changes The necessary revisions to the Technical Specifications to support operation at the reduced flow are included in Section 4.0.

Each Technical Specification change is discussed below.

2.1 Safety Limits A new reactor core safety limits curve is provided. As discussed above, the DNB limits of the figure are unchanged. However, the Vessel Exit 1831L:6/841218 13

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Boiling limits become more restrictive since flow is reduced for a given power.

2.2 Limiting Safety System Settings The protection system setpoints have been reviewed for the reduced flow. The only setpoints which are impacted by the flow reduction are the Overtemperature Delta-T and Overpower Delta-T functions.

These setpoints are designed to protect the core by tripping the reactor before the core safety limits (Figure 2.1-1) are exceeded. The setpoint equations have been recalculated for the reduced flow and steam generator low-low level setpoint changes.

In addition, the time constants in the equations have been updated.

Specifically, the lag time constants in the delta-T and Tavg channels have been increased from 2 to 6 seconds, to accommodate operational considerations. The effect of this change has been evaluated by reanalyzing the limiting events that rely on Overtemperature Delta-T and Overpower Delta-T protection.

The limiting RCCA Withdrawal at Power cases from the OFA transition analyses have been reanalyzed with the increased time constants in the Overtemperature Delta-T setpoint equation. The results show that the DNB design basis is met.

The Overpower Delta-T trip is not relied upon for protection for any of the FSAR accident analyses.

However, a spectrum of steamline breaks were analyzed at various power levels in Reference 11 to determine the limiting cases that are presented in the FSAR.

Some of the small

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steamline breaks at power analyzed in this generic study rely on Overpower Delta-T for protection.

l A McGuire-specific analysis was performed that verifies that the DNB l

design basis is met for small breaks at full power with the increased time constants in the Overpower Delta-T setpoint equation.

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Also, the lead-lag compensation on Tavg is changed from 33/4 to 28/4.

The 28/4 compensation was used in the accident analyses and affords the plant more margin to an Overtempc9ture Delta-T trip on a load rejection.

3/4.2.3 RCS Flow Rate and F-delta-H, and Bases A new RCS flow vs. R figure is provided for Unit 1 to reflect the reduced flow.

3.4 LOCA Analysis 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 modified to incorporate BART(12) core reflood hea,t transfer models was utilized for the analysis.

BART(13) has been approved for use on non-UHI plants.

Four cold leg breaks were reanalyzed.

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 crcssflow calculated to 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 OFA and STD assemblies would be approximately I

11*F.

This effect can be offset in the McGuire 1 and 2 transition cores by considering the favorable UHI quench characteristics of the STO 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 1831L:6/841218 15

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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* penalty identified for transition cycles; 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.

AH 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*.

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.

j 3.-

The clad. temperature transient is terminated at a time when the

[

core geometry is still amenable to cooling.

The localized cladding 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 required by the long-lived radioactivity remaining in the core.

d 1831L:6/841218 16 t-

The,Large Break LOCA analysis for McGuire 1 and 2 utilizing the currently approved UHI Evaluation Models modified to incorporate BART technology resulted in a PCT of 2157*F at 2.26 F for the CD = 0.6 g

(perfect mixing) DECLG break.

The small impact 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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4.0 TECHNICAL SPECIFICATION CHANGES To ensure that pl' ant operation is consistent with the design and safety evaluation conclusion statements made in this report and to ensure that these conclusions remain valid, several Technical Specifications changes will be needed for Cycle 3.

These changes are given in Appendix A.

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5.0 REFERENCES

1. Bordelon, F.M., et. al., " Westinghouse Reload Safety Evaluation Methodology", WCAP-9273, March 1978.

2. Duke Power Company Transmittal to NRC, " Safety Evaluation for McGuire Units 1 and 2 Transition to Westinghouse 17x17 Optimized Fuel Assemblies."

3. "McGuire Final Safety Analysis Report."

4. Miller, J.V., (Ed.),

" Improved Analytical Model used in Westing-house 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, June 1977.

7. Skaritka, J., Iorti, J. A., " Operational Experience with Westinghouse Cores", WCAP-8183, Revision 13, September, 1984.

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

June 1983.

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

10. Letter from NRC, C. O. Thomas to E. P. Rahe, Jr., Westinghouse,

" Acceptance for Referencing of Licensing Topical Report WCAP-10297-(P), WCAP-10298 (NS-EPR-2545) Entitled Dropped Rod Methodology for Negative Flux Rate Trip Plants", March 31, 1983.

11. Hollingsworth, S. D. and Wood, D. C., " Reactor Core Response to Excessive Secondary Steam Releases," WCAP-9226, Revision 1, (Proprietary), January,1978, and WCAP-9227, Revision 1, (Non-Proprietary), January,1978.

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12. Schwartz, W. R., " Addendum to BART-A1: A Computer Code for the Best Estimate Ana?ysis of Reflood Transients," WCAP-9561, Addend;m 1, November 1984.

(Westinghouse Proprietary)

13. Young, M.-Y., et al., "BART-A1: A Computer Code for the Best Estimate Analysis of Reflood Transients," WCAP-9561-P-A, March 1984.

(Westinghouse Proprietary)

14. Davidson, S.

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

February 1984.

1831L:6/841218 19

2 TABLE 1 MCGUIRE UNIT 1 - CYCLE 3 FUEL ASSEMBLY DESIGN PARAMETERS t

Region-1 3

4*

5*

Enrichment (w/o U-235)+

2.108 3.106 3.205 3.20 Density (% Theoretical)+

94.53 94.84 95.04 95.0 Number of Ass'emblies 9

64 60 60

-Approximate Burnup at++

16401#

23620 11061 0

Beginning of. Cycle 3

_(MWD /MTU)

Approxim' ate Burnup at++

27097#

34925 20185 14856 End of Cycle 3 (MWD /MTU) i l

-Optimized Fuel - Zire grid

+ All fuel region. values are as-built except Region 5 values w'hich are nominal.

++ Based on E0C2 = 10577 MWD /MTU, EOC3 = 11700 MWD /MTU (coastdown included)

The burnups noted are for the' Region 1 fuel assemblies be'ing used and are not'an average for the whole region.

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TABLE 2 MCGUIRE UNIT 1 - CYCLE 3 KINETICS CHARACTERISTICS Cycle 3 Current Limits Design Minimum Moderator

+5 < 70% of RTP

+5 <70% of RTP Temperature Coefficient 0 3 70% of RTP 0 370% of RTP 0

(pcm/ F)*

Doppler Temperature

-2.9 to -0.91

- Coefficient (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 (pcm/%

-power)*

Minimum Delayed Neutron

.44

>.44 Fraction 6,ff, (%)

Minimum Delayed Neutron

.50

>.50 Fraction 6,ff, (%)

[ Ejected Rod at BOL]

Maximum Differential Rod 100

<100 Worth of Two Banks Moving Together (pcm/in)*

pcm-= 10-5 3, 18311:6/841218~

21 mmaa

1 TABLE 3

.END-OF-CYCLE SHUTDOWN REQUIREMENTS AND MARGINS MCGUIRE UNIT 1 - CYCLE 3 Control Rod Worth (%Ap)

Cycle 2 Cycle 3 All Rods Inserted 7.06 6.72 All Rods Inserted,Less Worst Stuck Rod 5.94 5.90 (1) Less 10%

5.34 5.32 Control Rod Requirements Reactivity Defects (Doppler, T,yg, 3.22 3.18 Void, Redistribution)

Rod Insertion. Allowance 0.50 0.50 (2) Total Requirements 3.72 3.68 Shutdown Margin [(1) - (2)] (%Ap) 1.62 1.64 Required Shutdown Margin (%ap) 1.30 1.30 i

1831L:6/841218 22

r TABLE 4 MCGUIRE UNIT 1 - CYCLE 3 CONTROL ROD EJECTION ACCIDENT PARAMETERS HZP-BOC Current Limit Cycle 3 Maximum ejected rod 0.75

<0.75 worth, %Ap Maximum F '(ejected) 11.5

<11.5 g

HFP-BOC Maximum ejected rod 0.23

<0.23 worth, %Ap Maximum Fg (ejected) 5.3

<5.3 HZP-EOC Maximum ejected rod 0.90

<0.90 worth, %Ap Maximum Fg(ejected) 20.0

<20.0 HFP-EOC Maximum ejected rod 0.23

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

<5.9 1831L:6/841218 23

~

180' R

P N

M L

K J

H G

F E

D C

8 A

4 4

4

_g 4

4 5

3 5

4 4

2 4

3

,5 3

5 3

4 3

5 3

4 3

4 8

SS 8

4 4

5 1

5 3

5 1

5 3

5 1

5 4

4 8

8 8

8 4

4 4

5 3

5 4

5 8

8 8

4 3

5 3

4 3

5 3

4 6

8 8

4 5

3 5

3 4

3 5

4, 7

4 8

8 4

90*

4A 3

4 1

5 3

4 1

4 3

5 1

4 3

4 8 ;'

R 9

4 5

3 5

3 4

3 5

4 g

4 8

8 4

4 3

5 3

4 3

5 3

4 i

8 8

10 4

5 3

5 4

g 8

8 8

(

4 5

1 5

3 5

1 5

3 5

1 5

4 12 4

8 8

4 4

3 5

3 4

3 5

3 4

4 8

SS 8

4 13 4

4 5

3 5

4 4

14 4

4 4

15 O'

X Region number Y

B A's SS Secondary Source FIGURE 1 CORE LOADING PATTERN MCGUIRE UNIT 1, CYCLE 3 24

c.

APPENDIX A TECHNICAL SPECIFICATION PAGE CHANGES Modificatians to Pages:

4 2-2 B 2-4

. 2-5 2-0 4

2-8 thru 2-11 3/4 2-1 B 3/4 1-2'

{

~

3/4 2-la B 3/4 2-1 thru 2-2a 3/4 2-6 thru 2-9b B 3/4 2-4 3/4 2-12 3/4 2-16 3/4 3-9 3/4 3-10 3/4 3-28 3/4 3-45 1831L:6/112884

o MODIFICATIONS TO 3/4.2.1 '

AXIAL FLUX DIFFERENCE LIMITS e

e 9

l 24.2 P0a'ER DIST :5UTICN LIMITS AFD 3/4.2.1 ' AXIAL FLUX DIFFERENCE 'U!?!T l' LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within:

a.

the allowed operational space defined by Figure 3.2-1 for RAOC operation, or 5

b.

within a 2 X percent target band about the target flux difference during base load operation.

APPLICABILITY: MODE I above 50% of RATED THERMAL POWER *.

ACTION:

~

a.

For RAOC operation with the indicated AFD outside of the Figure 3.2-1

limits, 1.

Either restore the fndicated AFD to within the Figure 3.2-1 limits within 15 minutes, or

^

i 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux -

i High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

For Base Load operation above APLND** with the indicated AXIAL FLUX DIFFERENCE outside of the applicable target band about the target flux difference:

1.

Either restore the indicated AFD to within the target band limits within 15 minutes, or 1

ND 2.

Reduce THERMAL POWER to less than APL of RATED T'HERMAL POWER and discontinue Base Load operation within 30 minutes.

c.

THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the Figure 3.2-1 limits.

~

See Special Test Exception 3.10.2.

- APLND'is the minimum allowable power level for base load operation and wi1T l.

be provided in the Peaking Factor Limit Report per Specification 6.9.1.9.

EU!"E - U!::T; 1 =d 2 3/4 2-1 l

S'.'WEILL ACE RE00!tEMEN15

~

4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 507. of RATED THERMAL POWER by:

a.

Monitoring the indicated AFD for each OPERABLE excore channel:

1.

At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and 2.

At least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitoring Alars to OPERABLE status.

b.

Monitoring and logging the indicated AFD for each OPERABLE excore channel at least'once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least cnce per 30 minutes thereafter, when the AFD Monitor Alare is inoperable. The logged values of the indicated AFD shall be assur.ed to exist during the interval preceding each logging.

4.2.1.2 _The indicated AFD shall be considered outside of its limits when at least two OPERABLE excore channels are indicating the AFD to be outside the limits.

4. 2.1. 3 When in Base Load operation, the target axial flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days. The provisions of Specification 4.0.4 are not applicable.

4.2.1.4 When in Base Load operation, the target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining the target flux difference,..c ;.;;.t te 0.2.1.0 above or by linear interpola-tion between the most recentlydmeasured value and 0 ;;;;n.t at the end of cycle life. The provisions of Specification 4.0.4 are[not applicable.

'the calca h+ed vake' l

A Q c,,,;aeh J, & ucube rey;<e,,,e.,ts af,spe@c,% 3/uS l

i

(

i l

tN w e

5a 8W

I E

.n-

(=,

\\

,4

,.h 5 -

+ 4

-Y

~

r r

e 9

. J-t e

?

MODIFICATIONS TO 3/4.2.2 HEAT FLUX HOT CH EfEL FACTOR. LIMITS O

h y

e T

O 9

4 e

n i

POWER DISTRIBUTION LIMITS HEAT. FLUX HOT CHANNEL FACTOR-F0(Z)_

LIMITING CONDITION FOR OPERATION F (z) shall be limited by the following relationships:

3.2.2 n

2.26 F (z) $ [ +.45.1 (K(Z)] for'P > 0.5 g

P 2.2S F (z) 1 [ M 3 [K(I)] for P 1.0.5 0

0.5 where P = THERMAL POWER ico THERMAL POWER and K(z) is 'the function obt'ained from Figure 3.2-2 for a 4

given core height location.

APPLICABILITY: MODE 1 ACTION:

With F (z) exceeding its limit:

g 1.

Reduce THERMAL POWER at least 1 percent for each 1 percent F (z), exceeds the limit within 15 minutes and similarly Q

reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided~ the Overpower AT Trip Setpoints (value of K ) have been reduced at least 1 percent (in AT 4

span) for each 1 percent Fg(z) exceeds the lim,it.

b.

Identify and correct the cause of the out of limit condi-tion' prior to increasing THERMAL POWER; THERMAL POWER may then be increased provided F (z) is demonstrated through Q

incore mapping to be within its limit.

4 e

O

~

i SL'WE1LLANCE REQUIREMENTS ' UNIT 1) 4.2.2.1 The provisions of Specification 4.0.4 are not applicable, 4.2.2.2 For RAOC operation, F (z) shall be evaluated to determine if F (z) q g

is within its limit by:

a.

Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

b.

Incr: t.:,ing the measured F (z) component of the power distribution q

map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.

%eify % re.3a;tements of Sfee:fic.nSir 3. 2.=2 are Safisfied.

c.

Satisfying the' following relationship:

M

U 2) for P > 0.5 Fq (z) $

gg)

M Fq (z) 5.

),0 for P 1 0.5 I

where F (z) is the measured F (z) increased by the allowances for g

manufacturing tolerances and measurement uncertainty,M is the Fg

_, limit, K(z) is given in Figure 3.2-2, P is the relative THERMAL POWER, and W(z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation. This function is given in the Peaking Factor Limit Report as per Specification 6.9.1.9.

M d.

Measuring Fq (z) according to the following schedule:

1.

Upon achieving equilibrium conditions after exceeding by 10% or more of RATED THERMAL POWER, the THERMAL POWER at which F (z) was last determined," or q

2.

At least once per 31 Effective Full Powi[ Days, whichever occurs first.~

"During power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved l

and a power distribution map obtained.

l F

WW4

..m.

r y.

SurvEILLA'CE REC..sEF.EC

M::' D (Continued) e.

With measurements indicating maximum F (z) over z (K(z)

M has increased since the previous determination of Fq (z) either of the following actions shall be taken:

M 1)

Fq (z) shall be increased by 2% over that specified in Specifi-

~ cation 4.2.2.2c. or M

2)

Fq (z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that maximum fF (z) is not increasing.

(K(z)/

over I f.

With the relationships specified in Specification 4.2.2.2c. above not being satisfied:

1)

Calculate the percent F (z) exceeds its limit by t'he following q

expression:

)b x 100 -

(maximum M

Fn (z) x W(z) j -l for P > 0.5

~

2.

xg(Z]

.J fmaximum M

1 F0 (Z)

WI*)

-1 x 100 for P < 0. 5

, over z A

2.4 M

L g x K(z)]j s

l 2)

One of the following actions shall be taken:

s-a)

Within 15 minutes, control the AFD'to within new AFD limits which are determined by reducing the AFD limits of 3.2-1 by 1% AFD for each percent F (z) exceeds its limits g

as determined in Specification 4.2.2.2f.1).

Within i

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , reset the AFD alarm'setpoints to these modified l-limits, or b)

Comply with the requirements of Specification 3.2.2 for F (z) exceeding its limit by the percent calculated q

above, or,

c)

Verify that the requirements of Specification 4.2.2.3 for Base Load operation are satisfied and enter Base Load operation.

"'CUIni-U""51 rd2 3/4 2-8 Y

i SUP'.EI'.L ANCE REC. :7.EME C # !' D (Continued)

g. ' The limits specified in Specifications 4.2.2.2c, 4.2.2.2e., and 4.2.2.2f. above are not applicable in the following core plane regions:

1.

Lower core region from 0 to 15%, inclusive.

2.

Upper core region from 85 to 100%, inclusive.

ND 4.2.2.3 Base Load operation is permitted at powers above APL if the following conditions are satisfied:

a.

Prior to entering Base Load operation, maintain THE ove ND APL and less than or equal to that allowed by Specification 4

.2 for at least the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Maintain Base Load operation s'urveillance (AFD within 25$ of target flux difference) during this time period. Base Load operation is then permitted providing THERMAL ND 0

and APL ' er between APLND POWER is maintained between APL and 100% (whichever is most limiting) and F surveillance is maintained g

BL pursuant to Speciff a ion 4.2.2.4.

APL is defined as:

APLOL = " "I""" [

~

3 x 100%

over Z F (Z) x W(Z)8L "where: F (z) is -the measured F (z) increased by the allowances for q

manufacturing tolerances and measurement uncertainty. The F limit q

is N K(z) is given in Figure 3.2-2.

W(z)BL is the cycle dependent function that accounts for limited power distribution transients encountered during base load operation. The function is given in the Peaking Factor Limit Report as per Specification 6.9.1.9.

t b.

During Base Load operation, if the THERMAL POWER is decreased below ND APL then the conditions of 4.2.2.3.a shall b2 satisfied before re-entering Base Load operation.

4.2.2.4 During Base Load Operation F (Z) shall be evaluated to determine if 9

F (Z) is within its limit by:

9 Using the movable incore detectors to obtitin a power distribution a.

map at any THERMAL POWER above APL"U l

. b.

Increasing the measured F (Z) component of the power distribution q

map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.

+ke re9 0. cats of.Sju.'hea+;s-3 2.1 are Nisfiel Veri 4

i CIZ ONITS 1.nw 2 3/42-9 l

S'X.'ElttA' CE REC '?.EPE';'s PJ': :) (Centinued) c.

Satisfy efollowingrelationbhip:

F (Z) g

) for P > APLND p,

7 2.26 where: F (Z) is the measured F (Z). The F limit is-Gr1*r.

q q

K(2) is given in Figure 3.2-2.

P is the relative THERMAL POWER.

(W(Z)8L is the cycle dependent function that accounts for limited power distribution transients encountered during normal operation.

This function is given in the Peaking Factor Limit Report as per Specification 6.9.1.9.

d.

Measuring F (1) in conjunction with target flux difference deter-mination according to the following schedule:

1.

Prior to entering BASE LOAD operation after satisfying Section 4.2.2.3 unless a full core flux map has been taken in the previous 31 EFPD with the relative thermal power having been ND maintained above APL for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to mapping, and 2.

At least once per 31 effective full power days.

i With measurements indicating e.

F (2) maximum [

]

over Z has increased since the previous determination F (Z) either of the l

following actions shall be taken:

1.

F (Z) shall be increased by 2 percent over that specified in 4.2.2.4.c, or

/

2.

F (Z) shall be measured at least once per 7 EFPD until 2 successive maps indicate that Fh(Z) maximum [ g ) is not increasing.

over Z

f.

With the relationship specified in 4.2.2.4.c above not being satisfied, either of.the following actions shall be taken:

1.

Place the core in an equilbrium condition where the limit in 4.2.2.2.c is satisfied, and remeasure F (2), or l

3/4 2-9a l-m

p(+f'

.i M~;il f!N-w

~.

y;;

t 5'J:xE:LJ'::E F.EC'.1REMEr. 5 ('J":' M (Continued) 2.

Comply with the requirements of Specification 3.2.2 for F (Z) n exceeding its limit by the percent calculated with one of the following expressions:

Ff(Z)xW(Z)gt ND

[(max. over z of [

] ) -1 ] x 100 for P > APL a.26 x K(Z)

}

[(max. over z of [

]

x 100 for 0.5 < P < APL

)

g.

T,he limits specified in 4.2.2.'4.c, 4.2.2.4.e, and 4.2.2.4.f above are not applicable in the following core plan regions:

1.

Lower core region 0 to 15 percent, incl.usive.

2.

Upper core region 85 to 100 percent, inclusive.

4.2.2.5 When F (Z) is measured for reasons other than meeting the requirements q

of specification 4.2.2.2 an overall measured F (z) shall be obtained from a power q

distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

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BO OM CORE HEIGHT ( FEET )

' TOP K(Z)- NORMALIZED F (Z) AS T ON CORE HEIGHT (UNIT 1) 9 NCUIE - UNITO 1 ed 2--

3/4 2-12

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34.2 g:ER 0:ST: *KT!?'. L!c1TS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (1) maintaining the calculated DNBR in the core at or above the design limit during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical prop-

'erties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.

The definitions of certain het channel and peaking factors 'as used in these specifications are as follows:

F (Z)

Heat Flux Hot Channel factor, is defined as the maximum local 0

heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing toler-ances on fuel pellets and rods; N

F Nuclear Enthalpy Rise Hot Channel Factor, is defined as the rat.io of M

the integral _of linear power along the red with the highest integrated power to the average rod power; and a

F,y(Z) a ia defined a 3

to avera e

re elevation Z.

l:

3/4.2.1 AXIAL FLUX DIFFERENCE ThelimitsonAXIAi. FLUX FER NCE (AFD) assure that the FQ(Z) upper bound envelope of 2.32 ("nh 2,, 9-15 (U..M 1) times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes.

Target flux difference is determined at equilibrium *' xenon conditions.

l The full-length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels.

The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL P0a'ER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

i 4

e f

er 4

8

--ew

J FO.:ER DISTRIBUT;0'. LIMITS l

BASES AXIAL FLUX DIFFERENCE (Continued)

Although it is intended that the plant will be operated with the AFD wit n the target band required by Specification 3.2.1 about the target f1 differ ce, during rapid plant THERMAL POWER reductions, control rod eo on will caus the AFD to deviate outside of the target band at reduced RMAL PDWER levels. This deviation will not affect the xenon redistrib ton suffi-ciently to chan the envelope of peaking factors which may be eached on a subsequent return RATED THERMAL POWER (with the AFD with the target band) provided the time du ion of the deviation is limited.

ccordingly, a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation limit mulative during the previou 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is previded for operaticn outside of the ta et band but within th imits of Figure 3.2-1 d-while at THERMAL P0n'ER levels tween 50% and 9 of_ RATED THERMAL PC.;ER.

For THERMAL POWER levels between and 50%

RATED THERMAL POWER, devia-m 1

tions of the AFD outside of the tar ba are less s,ignificant.

The penalty of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months actual time reflects this r ed significance.

i '

Provisions for monitoring AFD on an omatic basis are derived from the plant process computer t ugh the AFD Monit Alarm. The computer deter-1 mines the 1 minute avera f each of the OPERABLE ore detector outputs and i-provides an alarm mess immediately if the AFD for t or more OPERABLE excore channels are tside the target band and the THE OWER is greater than 90%-of RAT HERMAL P0WER. During operation at THERMA ER levels between 50% i 90% and between 15% and 50% RATED THERMAL POWER, e computer outputs a arm message when the penalty deviation accumulates bey the limits I hour and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , respectively.

Figure B 3/4 2-1 shows a typical monthly target band.

N E t 1, A;t. power levels below APL, the limits on AFD are defihed by -

ND Figures 3.2-1, i.e. that defined by the RAOC operating procedure and limits.

l These limits were calculated in a manner such that expected operational l

transients, e.g. Ioad follow operations, would not result in the AFD deviating I

outside of those limits. However, in the event such a deviation occurs, the short period of titie allowed outside of the limits at reduced power levels will not result it. significant xenon redistribution such that the envelope of peaking factors would change sufficiently to prevent operation in the vicinity of the APL"O power level.

ng At power levels greater than APL, two modes of operation are permissible; s

1) RAOC, the AFD limit of which are defined by Figure 3.2-1, and 2) ase Load operation, which is defined as the maintenance of the AFD within a band o

ND I

3 about a target value. The RAOC operating procedure above APL is t e same as NU that defined for operation below APL However, it is possible wtn, l-following extended load following maneuvers that the AFD limits may result in

+

restrictions 'in the maximum allowed power or AFD in order to guarantee operation with F (z) less than its limiting value. To allow operation at the g

maximur. permissible value, the Base Load operating procedure restricts the

"- C : r.-

..t ! ", 1.

2 2 3/4 2-2

r: E e c I S T E'.'T I**. L It'I' S EASES AYIAL FLUX DIFFERENCE (Continued) indicated AFD to relatively small target band and power swings (AFD target bandof2k,APLND < power < APLBL or 100% Rated Thermal Power, whichever is lower). For Base Load operation, it is expected that the plant will operate within the target band. Operation outside of the target band for the short time period alloved will not result in significant xenon redistribution such that the envelope of peaking factors would change sufficiently to prohibit continued operation in the power region defined above.

To assure there is no residual menon redistribution impact from past operation on the Base Load 0

operation, a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months waiting period at a power level above APL and allowed cy RAOC is necessary. During this time peried load changes and red rrotion are restricted to that allowed by the Base Loac procedure.

After the waiting j

period extended Base Load operation is permissible.

T

,a the cor:puter determines the one minute average of each of the

5 OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excgre channels are: 1) outside the allowed AI power operating space (for RAOC operation), or 2) outside the allowed AI target band (for Base Load operation). These alarms are active when power is greater than: 1) 50% of RATED THERMAL POWER (for RAOC operation),

or 2) APLND (for Base Load operation). Penalty deviation minutes for Base Lead operation are not accumulated based on the short period of time during which operation outsica of the target band is allowed.

l 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor, RCS flow rate, and nuclear l

enthalpy rise hot channel factor ensure that: (1) the design limits on peak l

local power density and minimum DNBR are not exceeded, and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS accep-tance criteria limit.

Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3.

This periodic surveillance is

'suf ficient to insure that the limits are maintained provided:

(

l l

r l

l l

" C. : T. *

..J U 1.% 2 33/t. 2-2a

~

__..~_m,._,

,,,,___-e,

_ _. =, _,,,,

,n m.,

,m,

,ww--.-

i a

t THIS FIGURE DELETED -

4 gp, g 4 21 TYPICAL INDICATED AXtAL FLUX CIFFE.9ENCE VERSUI 3

T)iERMAL power 4

c:

FURTHER MODIFICATIONS l

DUE TO :

Reduced RCS Flow Increased t in OTAT and OPAT Equations Revised SG Low-Low Level Setpoint L

i i

i t

e I

r 9

-,r-~

..-..--w

e t'

i O. C C

W

r Qiere

.ro llo "I"j Jo 668 w Per 1.000

98.400 g m

. 6:

653 645 Unace table 00er s1on 640 4

635 633 4-si, 625- -

g 629 2

,,, 615

bat, /,

6te 625- -

tctet le 6C0 One tan

\\

595 -

\\

580 -

g

\\

Ses>

.I J

.5 4

.5

.6 7

.8 1.

I.t t.2' PC'.'CR Ifesction or nominall

. FIGURE 2.1-la UNIT 1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION

--fo GMRC 0"ITS'i end 2 -'

2-2

%vndmenLMoj.2_41n4647~~~

c Amendmemt'No. D'(UnklL

660 Flow Per Loor

97220 3p-650 T' o o #*I

"* C'* NAN

645 n

640 22Q 655 650 k625 2%

a Pa

$ 629 9

N3y s.

615

/,

6IO 605 Accepblote.

- OperaMoa 595 590 585 9.

I

.2

.5 4

.5

.6

.7

.8

.9 1.

1.1 1.3 POWER (fraction of nominell UNIT 1 REACTost. CORE SAFETY L lM tTS Fou R.

LocPs N OPEltATioM

-_..___-.-.,.,.--____,__m..,.

--.-w-

,. =-

v-

TABLE 2.2-1 if 8

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS u

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

1. Manual Reactor Trip N.A.

N.A.

2. Power Range, Neutron Flux Low Setpoint - < 25% of RATED Low Setpoint - < 26% of RATED

[

THERMAL POWER -

E High Setpoint - 1 109% of RATED High Setpoint 5 110% of RATED THERMAL POWER THERMAL POWER

3. Power Range, Neutron Flux,

< 5% of RATED THERMAL POWER with 5 5.5% of RATED THERMAL POWER High Positive Rate a time constant 1 2 seconds with a time constant 2 2 seconds

4. Power Range, Neutron Flux, 5 5% of RATED THERMAL POWER with

< 5.5% of RATED THERMAL POWER High Negative Rate a time constant > 2 seconds 7

with a time constant 1 2 seconds u

5. Intermediate Range, Neutron 1 25% of RATED THERMAL POWER

$ 30% of RATED THERMAL POWER Flux 5

5

6. Source Range, Neutron Flux i 10 counts per second 5 1.3 x 10 counts per second

7. Overtemperature AT See Note 1 See Note 3

8. Overpower AT See Note 2 See Note 3

9. Pressurizer Pressure--Low

> 1945 psig 1 1935 psig h

l

10. Pressurizer Pressure--High 1 2385 psig i 2395 psig

11. Pressurizer Water Level--High 1 92% of instrument span 5 93% of instrument span z

12. Low Reactor Coolant Flow 1 90% of design flow per loop

1 89% of design flow per loop

f

Design flow is 03,'00 gpa per loop f;r Ur.it 1 cr.d 05,500 ;;p pr 10 p for Ur.it 2.

2 97,200 l

\\

TABLE 2.2-1 (Continued) j REACTOR TRIP SYSTEM INSTRUMENTATION IRIP SETPOINTS N

[

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

13. Steam Generator Water

> 12% of span from 0 to 30% of

> 11% of span from 0 to 30% o'f

[

Level--Low-Low EATED THERMAL POWER, increasing NATED THERMAL POWER, increasing linearly to > 51."%

f span at to -53.^% of span at 100% of RAIED U

100% of RATED THERMAL POWER.

THERMAL} POWER.

40. O */,

39.O '/,

14. Undervoltage-Reactor

> 5082 volts-each bus

> 5016 volts-each bus Coolant Pumps

15. Underfrequency-Reactor

> 56.4 Hz - each bus

> 55.9 Hz - each bus Coolant Pumps

16. Turbine Trip

){

a.

Low Trip System Pressure

> 45 psig

> 42 psig b.

Turbine Stop Valve Closure

> 1% open

> 1% open i

17. Safety injection Input N.A.

N.A.

i from ESF 18.

Reactor Trip System Interlocks

-11 a.

Intermediate Range Neutron Flux, P-6,

> 1 x 10'I amps

> 6 x 10 amps Enable Block Source Range Reactor Trip b.

' Low Power Reactor Trips Block, P-7 1)

P-10 Input 10% of RATED

> 9%, i 11% of RATED THERMAL POWER THERMAL POWER 2)

P-13 Input 5 10% RTP Turbine 5 11% RTP Turbine impulse Pressure Impulse Pressure Equivalent Equivalent

TABLE 2.2-1 (Continued) l

{ {

~

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS i

NOTATION C

i NOTE 1: OVERTEMPERATURE AT

~

o 1

1*I 1

AT (1 + t S) (1 + 1 5) < ATo (K1 4

2 (1 + tsS)[T(1 + TsS)-T'] + K (P-P') - f (AI))

-K 1+T 5

.2 3

1 2

3 Where:

AT Measured AT by RTD Manifold Instrumentation,

=

i 1+t 5 lead-lag compensator on measured AT.

=

y,T 3

= Time constants utilized in the lead-lag controller for 13 T2 m.

AT, T 3? 8 sec., 1 di 3 sec.,

2 om 3

y, Lag compensator on measured AT',

=

j r3 Time constants utilized in the lag compensat.or for AT. Is S I 5eC.

=

i s

j AT, Indicated AT at RATED THERMAL POWER,

=

1 g

I I.SS 2 (UGU1 2), 1.$$$2 (2231 1),

l, 2,QQ.

i r 7

K2 C'2 }3 I *it 2I' O.0222 (Un't 1),

v' I,\\ l 1

i t' t 1*tS l

$ [ h 1 + tsS The function generated by the lead-lag controller for T,,g dynam,1c compensation,

=

(

t: e j

- lh l

'U Time constants utilized in the lead-lag controller for T*'8, T4

=

-;. ~ 2^.

CJ..;; 1), t h.33'sec.-(Un't 2), is 4 4 sec.

25 a.,

i

'T

=

Average temperature, 'F, l

3

]

Lag compensator on measured T,,g,

=

y,g3 g s

i i

7 TABLE 2.2-1 (Continued) i REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS i

r -

NOTATION (Continued) i I

[ [

NOTE 1: (Continued)

~

i l 6

Time constant utilized in the measured T ts

=

E_ lag compensator, r 4/sec s

1 i

--(5! ts 1 3. 2),

1 2

T'

-<588./*FReferenceT at RATED THERMAL POWER, i.,

j avg K

=

3

0. 50 G..R O, 0. MMS W..f. O,- /

i i

P

=

Pressurizer pressure, psig,

.P'

=

2235 psig (Nominal RCS operating pressure),

i m

4 S

=

-1 Laplace transform operator, sec i.

j and f (AI) is a function of the indicated difference between top and bottom detectors 3

I of the power-range nuclear fon chambers; with gains to be selected based on measured i

instrument response during plant startup tests such that:

1

-29%

+9.0%

(1) for q q,, between -36% and 4. 0

'"i.'t 2), - iL*' : e -i.a*' (5f t 1); f (al) = 0, g

j where q alid q are Percent RATED THERMAL POWER in the top and bottom 3

g b

gj,,3 og 7' 7 the core respectively, and q

+q is total THERMAL POWER in percent of RATED g

b

{

THERMAL POWER;

, c, m

-29%

l{

(ii) for each percent that the magnitude of q q exceeds -365 (" 'i 2), -11*' ("a 8' 1),

g b

the AT Trip Setpoint shall be automatically reduced by 1.1"" (" 't 2),'3.151% ($5i 1) i is_

of its value at RATED THERMAL POWER; and 4

i

+9.0%

4 (111) for each percent that the magnitude of q i

exceeds +0dlE (5't 2), - i.'G%

l 2i E (5't1),theATTripSetpointsha11bekut tica11y reduced by t:90tN

'f h

(" 2),1.00"' (i't 1) of its value at RATED THERMAL POWER.

1.50 %

m c,

t

Y TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS a

3' 5

NOTATION (Continued)

NOTE 2:

OVERPOWER AT i

A T (I *

SI * *' ) (

) < AT {K

-K

(

)(

--) T -K [T(

)- T"] - f (AI))

I 1+T5 o

4 5 1 + tyS 1 + TsS 6

1+t S 3

D 2

g Where:

AT

=

As defined in Note 1, 1 +t 5

= As defined in Note 1 2

ti,12

= As defined in Note 1 3,,3 As defined in Note 1

=

T w

AT

=

As defined in Note 1, o

o K

$ 1.090gP.;;.i},i.G.~s..... 1}g !

4 K

= 0.02/*F for increasing average temperature and 0 for decreasing average S

temperature.

1 5 7

The function generated by the rate-lag controller for T

=

l 3,

3

}.j compensation.

dynamic s

t,

=

Time constant utilized in the rate-lag controller for T,,g, 1 4

5 sec (Unus -l-8 ?),

7

!!! L y

.iFi F 3,,3 As defined in Note 1,

=

e is As defined in Note 1

=

..r >

'd 3 K

=

y 6

^ "^120/

'U;!t 2), 0.00169/*F (U;'i 1)- for T > T" and

/

K 0 for T 1 T",

f 6

T TABLE 2.2-1 (Continued) c Ih REACIOR TRIP SYSTEM INSTRUMENTATION TRIP SETP0INTS

~

NOTATION (Continued)

I 5

T

=

As defined in Note 1 A

2 a

-<588./*FReferenceT at RATED THERMAL POWER, 1"

=

avg E

S

=

As def t ed in Note 1, and

,2 f (aI) 0 for aH AI.

=

2 Note 3:

The channel's maximum Trip Setpoint sha11 not exceed its computed Trip Setpoint by more than 2%.

7 M

G l

C.I L

i q,y

. so L DP C

?

m i

LIMITING SAFETY SYSTEM SETTINGS BASES i

Power Rance. Neutron Flux (Continued)

The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint.

Power Range Neutron Flux High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power level.

Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power.

The Power Range Negative Rate, trip provides protection for control rod drop accidents. At high power, a rod drop accident of a single or multiple rods could cause local flux peaking which could cause an unconservative local DN8R to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor.

No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DN8R's will be greater than D68: & Jasip limif DN82 valog.

Intermediate and Source Rance. Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a subcritical condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux chgnnels. The Source Range channels will initiate a Reactor trip at about 10 s counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 2SE of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

c l

l i

f 4

+

}

l l

i l

MT:

LC: 1F 8 2-4

1 50 PENALTIES OF 0.1% FOR UNDETECTED FEED-OPE AT O

[

TER VENTURI FOULING AND MEASUREMENT REGION

~

4g UN TAINTIES OF 1.7% FOR FLOW AND 4%

o r[

j.1.M. 47.6 [

IN E MEASUREMENT OF F ARE AH I

IN DED HIS FIGURE.

j l

it g'

J N

/

b /f/

i, N

T x>

~

ACCEPhtE

[

b h

o OPERATIO O D

a REGION E

Y UNACCEPTABLE 3

O g

42 w

7 OPERATION D

}

p (b

REGION 5

- /

8 5.

y x

/

N N

% 2:

ACCEPTAlstE OPERA 1 ' ION REGION FOR <98E nur (1.0, 39.

p

/

694% RTP n

38

~

$93 RW (1.0, 37.79)

N 6

/

se n1P (1.0, 37.39) y,

/

,o,o 0.09 0.92 0.M 0.M 0.98 1.00 1.02 1.04 1.06 08 s

' 3 R. = F 11.49(1.0 + 0.3 (1.0-P)]

AH f

FIGURE 3.2-3a RCS TOTAL FLOW RATE VERSUS R (UNIT 1) l

p g'..I E

)

.s l

g PENALTIES OF 0.1% FOR UNDETECTED FEED-l C

46 WATER VENTURI FOULING AND MEASUREMENT s

l

l UNCERTAINTIES OF,1.7% FOR FLOW AND 4%

FOR INCORE MEASUREMENT OF F ARE i

g INCLUDED IN THIS FIGURE.

MCCEPTABL[

k'<*

g y

-OPERATION-

)

J' n.

-REGION-FOR-O

/-BrONL%

i C

f.

l naa

\\

y ACCEPTABLE

)

a i

et OPERATION 3:

42 REGIONHHl, I 3123h.co{

9

-av + Rr

/

l A

'J g

UNACCEPTABLE gs s

OPERATION us

~

REGION l

f E

40 1

l ACCEPTABLE OPERATION HliGION FOR 598% RTP (1.0, 38.888) i j

$96% RTP (1.0, 38.4991 38 594% RTP (1.0, 38.11h

$92% RTP

'(1.0, 37.72,1)

I i

590% RTP (1.0, 37.33p g

R-(1.0,37A44134.944 a.2 36 z

PP 0.09 0.92 0.94 0.96 0.98 1.00 1.02 1.04 1.06 o.3

,% = F 11.49[1.6 +-e&(1.0 P)]

Figure 3.2-3)( RCS FLOW RATE VERSUS Rx and4g.- FOUR LOOPS IN OPERATION (Unit. 2)-

F

N.

A~,

es, a

p ih TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUNENTATION RESPONSE TIMES EE RESPONSE TIME

}

FUNCTIONAL UNIT 1.

Manual Reactor Trip N.A.

I 2.

Power Range, Neutron Flux 5 0.5 second*

'[

3.

Power Range, Neutron Flux, N.A.

High Positive Rate 4.

Power Range, Meutron Flux, 1 0.5 second*

High Negative Rate N.A.

5.

Intermediate Range, Neutron Flux N.A.

Y 6.

Source Range, Neutron Flux 8.0 7.

Overtemperature AT 5-6:4 seconds

8.0 14h4 seconds

8.

Overpower AT

< 2,0 seconds 9.

Pressurizer Pressure--Low 10.

Pressurizer Pressure--High 1 2.0 seconds N.A.

11.

Pressurizer Water Level--High l

l Neutron detectors are exempt from response time testing. ' Response time of the neutron flux signal portion a

of the channel shall be measured from detector output or input of first electronic component in channel.

i I

9 TABLE 3.3-2 (Continued)

C L

REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES i$

FUNCTIONAL UNIT RESPONSE TIME

!i

q 12.

Low Reactor Coolant Flow 3 is a.

Single Loop (Above P-8)

$ 1.0 second t **

s; b.

Two Loops (Above P-7 and below P-8)

$ 1.0 second 8'

3.5 sa-13.

Steam Generator Water Level--Low-Low 1-Br& seconds 4

i 14.

Undervcitage-Reactor Coolant Pumps

< l.5 seconds 15.

Underfrequency-Reactor Coolant Pumps

< 0.6 second I

16.

Turbin'e Trip a.

Low Fluid Oil Pressure N.A.

. g>

b.

Turbine Stop Valve. Closure l

N.A.

i 0 17.

Safety Injection Input from ESF N.' A.

18.

Reactor Trip System Interlocks N.A.

l 19.

Reactor Trip Breakers N.A.

I 20.

Automatic Trip and Interlock Logic N.A.

1

7

^

~

!g TABLE 3.3-4 (Continued)

n li ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS Ii y

-FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

.55

!;i 7.

Auxiliary Feedwater

[

a.

Manual fnitiation N.A.

N.A.

b.

Automatic Actuation Logic N.A.

N.A.

and Act.uation Relays 2

i c.

Steam Generator 59.07.

Water Level--Low-Low 1)

Start Motor-Driven Pumps

> 12% of span from 0 to 13 of span from 0 to 40.QQ>30% of RATED THERMAL POWER

>0% of RATED THERMAL POWER, i

3

!)w ncreasing linearly to creasing linearly to 4W9% of span at 100% of

_ GiMJN of span at 100% of lwA RATED THERMAL POWER.

RATED THERMAL POWER.

I" 2)

Start Turbine-Driven Pumps-

> 12% of span from 0 to

> 11% of span from 0 to l

30% of RATED THERMAL POWER, 30% of RATED THERHAL POWER, 40.0!g> increasing linearly to increasing linearly to i

% 9% of span at 100% of

> SiMJ% of span at 100% of TIATED THERMAL POWER.

TtATED}THERMALPOWER.

31.0 7, d.

Auxiliary Feedwater

> 2 psig

> 1 psig Suction Pressure - Low (Suction Supply Automatic Realignment) e.

Safety Injection -

See Ites 1. above for all Safety Injection Trip Setpoints Start Motor-Driven Pumps

,and Allowable Values i

f.

Station Blackout - Start 3464 i 173 volts with a

> 3200 volts Motor-Driven Pumps and 8.5

0.5 second time Turbine-Driven Pump delay g.

Trip of Main Feedwater Pumps -

N.A.

Start Motor-Driven Pumps i

O O'

J IN,STRUMENTATION MOVABLE INCORE DETECTORS LIMITING CONDITION FOR OPERATION 3.3.3.2 The Movable Incore Detection System shall be OPERABLE with:

a.

At least 75% of the detector thinibles, b.

A minimum of two detector thimbles per core quadrant, and c.

Sufficient movable detectors, drive, and readout equipment to map these thimbles.

APPLICABILITY: When the Movable Incore Detection Syster is used for:

a.

Recalibration of the Excore Neutron Flux Detection Systen, b.

Monitoring the QUADRANT POWER TILT RATIO, or MeasurementofFh,F(Z)r.dy c.

q ACTION:

With the Movable Incore Detection System in.operat'ie, do not use the system for the above applicable monitoring or calibration functions.

The provisions of Specifications 3.0.3 and 3.0.4 arg not applicable.

~

SURVEILLANCE REQUIREMENTS-4.3.3.2 The Movable Incore Detection System shall be demonstrated OPERABLE at

_least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by normalizing each detector output when required for:

~

a. * 'Recalibration of the Excore Neutron Flux Detection System, or b.

Monitoring the QUADRANT. POWER TILT RATIO, or MeasurementofFh,F(Z),rdF _

~

c.

q xy j-

- " CUI."C

.U"ITS 1 end 2 3/4 3-45

s REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued)

The Surveillance Requirements for measurement of the MTC at the beginning and near the and of the fuel cycle are adequate to confirin that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be inade critical with the Reactor Coolant System average temperature less than 551*F.

This limitation is required to ensure:

(1) the moderator temperature coefficient is within it analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in -

an OPERA 8LE status with a steam bubble, and (4) the reactor vessel is above its minimum RT temperature.

NOT 3/4.1.2 80 RATION SYSTEMS The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include: (1) barated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, (5) associated Heat Tracing System's, and (6) an emergency power supply from OPERA 8LE diesel generators.-

g3 4 With the RCS average temperature above 200*F, minimum of two boron injection flow paths are required to ensure single unctional capability in the event an assumed failure renders one of the f ow paths inoperable. The boration capability of either flow path is suff tent to provide a SHUTD0kN

-MARGIN from expected operating conditions of delta k/k after xenon decay i

and cooldown to 200*F.

The.raximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 16,321 gallons of 7000 ppe borated water from the boric acid storage tanks or l

75,000 gallons of 2000 ppe borated water from the refueling water storage tank L

(RWST).

With the RCS' temperature below 200*F, one Boron Injection System is acceptable without single failure consideration on the basis of the. stable reactivity condition of the reactor and the additional restrictions prohibiting l

CORE ALTERATIONS and positive reactivity changes in the event the single Baron Injection System becomes inoperable.

The limitation for a maximum of one centrifugal charging pump to be l

OPERA 8LE and the Surveillance Requirement to verify all charging pumps except i:

the required OPERA 8LE pump to be inoperable below 300*F provides assurance j.'

that a mass addition pressure transient can be relieved by the operation of a single PORV.

I

'C u 4 anu 4 8 3/4 1-2 l

-w m

'~w-

*""**"'"' '"= ~ * ' ' " ' ~ ~ " * " " ' " ' ' ' ' ' ' * " ' ' ' ' ' ^

'~

J POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR. and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) a.

Control rods in a single group move together with no individual rod insertion differing by more than + 13 steps from the group demand position; b.

Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6; c.

The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and d.

The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

N F

will be maintained within its limits provided Conditions a. through i

d. above are maintained. As noted on FigureX 3.2-3 6. RCS flow rate 0ower and ay be " traded off" against one~another rate is acceptable if the _"___..d..N is dun;easeJi.e., a low measure P* te lev *l h

. :: icu) to ensure that the calcu-g lated DNBR will not be below the design DNBR value.

The relaxation of F as a function of THERMAL PDWER allows changes in the radial power shape for all permissible rod insertion limits.

Rg as calculated in Specification 3.2.3 and used in Figure 3.2-3, accounts N

for F less than or equal to 1.49.

This value is used in the various accident analyses where F influences parameters other than DNBR, e.g., peak clad tem-perature, and thus is the maximum "as measured" value allowed.

S; -- 2d'.4 a

s for the inclusion of a penalty for Rod. Bow on DNBR only. Thus, know the "as ured" values of F" and RCS flow allows for " tradeoffs" xcess of R equal to 1.

r the purpose of offsetting the Rod Bow D penalty.

Fuel rod bowing reduc he value of DNB ratio redit is available tc partially offset this reduction.

his credit s from a gaeric or plant-specific design margin.

For McGuire

, the margin used to partially offset rod bow penalties is 5.1%.

T m

n breaks down as follows:

1)

Design limit DNBR l'.6%

2)

Grid spacing 2.9%

3)

-Therma ffusion Coefficient 1.2%

4)

D ultiplier 1.7%

4 5

itch Reduction 1.7%

-I

=:-- _

-na /_

B 3/4 2-4 No. M (M

_ " -... m n 1.

o.

~

.~

~..

ATTACHMENT 2B PEAKING FACTOR LIMIT REPORT FOR MCGUIRE UNIT 1 CYCLE 3 RAOC AND BASE LOAD OPERATION This Peaking Factor Limit Report is provided in accordance with Paragraph 6.9.1.9 of the McGuire Unit 1 Technical Specifications.

The McGuire Unit 1, Cycle 3 elevation dependent W(z) values for RAOC operation at beginning, middle, and near end-of-life are shown in Figures 1 throu This information is sufficient to determine W(z) gh 3 respectively.

versus core height for Cycle 3 burnups in the range of 0

. MWD /MTU to 11700 MWD /MTU through the use of three point interpolation.

The McGuire Unit 1, Cycle 3 elevation dependent W(z) values for base load operation between 80% and 100% of rated thermal power with a 5 percent AFD about a measured target value at 150, 6000, and 10000 MWD /H7U Cycle 3 burnups are shown in Figures 4 through 6 respectively. This information is sufficient to determine W(z) versus core height for Cycle 3 burnups in the range of 0 MWD /MTU to 11700 MWD /MTU through the use of three point interpolation.

W(z) values for RAOC and base load operation were calculated using the method described in Part B of Reference 1.

The minimum allowable power level for base load operation, APLE, for McGuire 1 Cycle 3 is 80 percent of rated thermal power.

The appropriate W(z) function is used to confirm that the heat flux hot channel factor, Fq(z), will be limited to the Technical Specificktion values of:

Fq(z) 1

[K(z)] for P > 0.50 and Fq(z) 1 52 [K(z)] for P 1 0.50 4

l l

The appropriate elevation dependent W(z) values, when applied to a power distribution measured under eouilibrium conditions, demonstrates that the initial conditions assumed in the LOCA are met, along with the ECCS acceptance criteria of 10CFR50.46.

l (1) WCAP-10216-P-A, Relaxation of Constant Axial Centrol - Fq

~

Surveillance Technical Specification t

L i

HEIGHT iaAX (FEET)

W12)

BOTTOII--> 0.00 1.000 e

.20 1.000 m 40 1.000 e

.00 1.000 *

.80 1.000 s 1.00 1.000 +

1.20 _

1.000 a s.se 1.40 1.000 +

1.00 1.000 +

8.#

1.80 1.339 8*

2.00 1.313

)

2.20 1.207 1

g,,

2.40 1.281 2.00 1.236 s.,

2.80 1.200 3.00 1.105 s.es 3.30 1.190 3.40 1.195 8.38 3.90 1.187

'*8 3.80 1.188 4.00 1.197 s.3.

4.20 1.185 4.40 1.101 a.m 4.00 1.158 4.80 1.149

@ lM 5.00 1.142 z

5.20 1.140 h a*8 8

S.40 1.139 5.00 1.138 b

5.80 1.137 3,,

.00 1.14 E

s S.20 1.100 8

1.2 g

5.40 1.195 8* 8 S.90

' 1.188 S.00 1.175 8.to I

..888=

7.00 1.180 s

8 x'

7.20 1.184 EsEE ls.

s 3.go

's 7.40 1.188

=

7.00 1.105 8.14 7.00 1.103 8.12 8.00 1.180 8.20 1.178 g,,

8.40 1.170 8.00 1.192 3.se 8.00 1.151 0.00 1.141 8.es 3.20 1.138 9.40 1.190 3*#

9.00 1.183 g,

9.80 1.190 10.00 1.218 3.es:--- :---.

10.20 1.232 e.se s.m a.es Les 4.e 5,es se i.es e,y g.se te.se st.es I'ea.co

.00 s

10.80 1.000

EINE DEI 13ff FEET) 11.00 1.000

11.20 1.000

11.40 1.000

11.80 1.000

11.80 1.000

TOP.-> 12.00 1.00C

l e TOP Ape BOTTool 15% EXCLUDE 0 AS PER TEC39 DICAL SPECIFICATI0'14.2.2.2.8 Figure 1 McGuire Unit 1 Cycle 3 RAOC W(z) at 150 MWD /MTU l

F HEIGHT ttAX I

4 FEET)

WIZ)

B07 TOM--> 0.00 1.000 *

.20 1.000 *

.40 1.000 *

.50 1.000 *

.50 1.000

1.00 1.000

1.20 1.000

I.m 1.40 1.000

1.80 1.000

s.e 1.80 1.102 2.00 1.178 2.20 1.187 2.40 1.155 a.

2.00 1.145 her 2.80 1.128 3.00 1.135 s.es 3.20 1.135 3.40 1.140 3.00 1.148 3.80 1.157 i,3 4.00 1.167

s. >

4.20 1.174 4.40 1.179 Lsr 4.90 1.184 4.80 1.194 Sf 5.00 1.208 8','

5.30 1.218 l

2 8

' ~

3,3 5.40 1.228 E

s E

5.00 1.222 La 5.80 1.244 g

S.C0 1.259 8*M

,o a

I S.30 1.274 d '*"

5.40 1.281

=

I 8

S.00 1.285 a.m 3

5 5

E S.40 1.280 7.00 1.289

,8.'

5.8e 7.20 1.208 8*88

'7.40 1.200

_I 7.30 1.271

=l s-7.80 1.258 8.00 1.241 s.u 8.20 1.220 3.3e S.40 1.194 S.90 1.171 t

8*8 8.80 1.170 0.00 1.170 9.30 1.189 9.40 1.196 s.m 9.90 1.309 l

Le 0.80 1.229 10.00

'1.249 l

sm 5 5 5 t' 5: -

.:0!!!

10.20 1.272 a.m s.m 3.m ses se La e.m 1.as e.es s.es es.m sn.m u.as 10.40 1.000

10.90 1.000

M EIN NI 000

11.20 1.000

l

t.40 1.000

11.90 1.00C

11.80 1.000

TOP--> 12.00 1.000

e TOP Afe 90TT00615% EXCt.U0E0 AS PER TEO9 DICAL EPEf!FICATION 4.2.2.2.8 Figure 2 McGuire Unit 1 Cycle 3 RA0C W(7} tit 6000 MWD /MTU e

HEIGHT 14AX (FEET) t:(Z )

SOTTON--> 0.00 1.000 e

.20 1.000 e

.40 1.009 e

.60 1.000 e

.80 1.000 e 1.00 1.000 e 1.20 1.000 e 1.40 1.000 e t.te 1.50 1.000 e s.e 1.80 1.188 2.00 1.100 14 2.20 1.172 2.40 1.184 hw 2.90 1.157 2.80 1.148 h4 2.00 1.142 8.ee 2.20 1.138 2.40 1.148 km 2.90 1.193 2.80 1.178 km 4.00 1.181 4.20 1.202

8. 2 4.40 1.211 8.m 4.80 1.217 4.80 1.224 g,

5.00 1.222 5.20 1.237 sm 5.40 1.243 g km 5.00 1.248 5.80 1.257 b

,8 8.00 1.288 3*"

=

1 S.20 1.278 o

l m

d 3,3 S.40 1.381 5.00

1.282 g

s.m 8.80 1.283 s

8 7.00 1.281 kne

=I

~

7.20 1.270 m

n

'ar._

7.40 1.298 3'"

=

7.90 1.258 s.it 7.80 1.244 8.00 1.228 1.12 8.20 1.207 8.40 1.180 l.H 8.90 1.178 4

8.80 8.80 1.173 0.00 p 1.198 1.5 8.20 1.198 8.40 1.184 5.m 8.90 1.191 0.80 1.158 l.82 10.00 1.195 10.20 1.185 a.se er-t r!rr t r '

10.40 1.000 e e.ee s.m 3.es Em he Em s.m 1.m e.m t.m le.m II.m t&B 10.00 1.000

10.80 1.000

j EISE IEll8f1 FEET) 11.00 1.000 e 11.20 1.000 e 11.40 1.000 e 11.90 1.000 e 11.80 1.000 e 70P--> 12.00 1.000 e

[

e TOP ApS BOTTON 15% EXCLUDE 0 AS PER TEC3ed! CAL SPECIFTCATION 4.2.2.2.8 Figure 3 i

McGuire Unit 1 Cycle 3 RA0C W':) at 10000 MWD /MTU-

HEIGHT ISAX IFEETI W12)

SOTTOII--> 0.00 1.000 e

.30 1.000 e

.40 1.000

.50 1.000 e

.80 1.000 e 1.00 1.000 a 1.20 1.000 e e.m 1.40 1.0tk =

1.80 1.000 e 6.8 1.80 1.007 2.00 1.005 8'"

2.30 1.003 2A0 1.001 g

2.90 1.000 2.80 1.007 s,e 3.00 1.084 s.e 3.30 1.000 3.40 1.070 som 3.30 1.074 3.80 1.073 8*8 4.00 1.072 4.20 1.071

,,3 4.40 1.070 hs 4.00 1.089 4.00 1.088 C L2 S.00 1.005 i

5.20 1.083 l'8 S.40 1.000 m

S.00 1.087 R

g,g S.80 1.005 b

S.00 1.055 s.>

5 0.20 1.000 8

h3 S.40 1.083 3

0.00

1.005 i

I

,*8 9.00 1.081 7.00 1.000 gg 7.20 1.070 s.es 7.40 1.075 a

i 7.00 1.078 I

hh 7.00 1.002 8.00 1.005 3 12 8.20 1.003 8.40 1.000 g,,

8.90 1.001 I

3g35 l3 g g3,l3 E B s 'J-

=1 8.40 1.002 l

3.m 8"

'88 lEss,

,3 9.00 1.003

' 8.8

'".33,,,"

O.20 1.W2 9.40 1.000 i

9.00 1.000 0.80 1.001 10.00 1.001 som

^ ^

^

^^

^

rre-10.20 1.091 10.40 1.000 e e es 8.5 a.e km 4.5 km E.e 1.as ene 0.5 48 5 la.m sa.m 10.00 1.000 e 10.80 1.000 e 83BE 8ElGMT Reil 11.00 1.000 e 11.30 1.000 m 11.40 1.000 e 11.00 1.000 e 11.80 1.000 e TOP--> 12.00 1.000 e i

e TOP Afe SOTTON 10% EXCLUDE 0 AS PER TE00tICAL SPECIFICATION 4.2.2.2.8 Figure 4 McGuire Unit 1 Cycle 3 Basela0d W(z) for P0wers Between 80T and 1007 of Rated Themal P0wer Within +55 AFD Of Measured Target 150 MWD /MTU e

,,-n

--.er

+,

,w

-,g-

,.,n-

l HEIGHT CAX (FEETl Wl2 )

SOTTON--> 0.00 1.000 e

.20 1.000 e i

40 1.000 e l

.60 1.000 m

.80 1.000 e

{

1.00 1.000 =

i 1.20 1.000 v i

4.M 1.40 1.000 m 1.90 1.000 e s.e 1.00 1.122 i.e 2.00 1.117 2.20 1.112 km 2.40 1.105 2.00 1.100 8.4 2.90 11093 2.00 11006 l.as 2.20 1.000 8.5 3.40 1.077 3.30 1.078 km 3.80 1.078 4.00 1.075 s se 4.20 1.075 4.40 1.074 8.2 4.00 1.073 g

g,3 4.80 1.071 5.C0 1.000 som 5.20 1.087 g kB 5.90 1.002 5.40 1.085 b ***

5.80 1.000 S.00 1.000 od g,g 0.20 1.002 S.40 1.002 1 a.m 8.00 1.054 S.00 1.007 ble 7.00 1.071 7.30 1.078 7.40 1.001 8.t*

7.00 1.005 7.00 1.008 s,33 8.00 1.000

,3 8.20 1.003 s.es 3-8.40 1.004 seau:::g, 8.90 1.005 1

8, lE l*#

. 3 8.80 1.000

==u:

=

I 8:e;..m8:

0.00 1.095 9.30 1.004 s.se S.40 1.002, 9.50 1.001 s.m 9.80 1.090 10.00 1.000 g,g,.--

y see m :

10.20 1.007 6.M l.5 3.W 3.5 45 hm 4.5 1.m e.m s,a

34. s pa.m 83.g 10.40*.

1.000 e 10.00 1.000 a 10.80 1.000 s (INE ElliNT FEET) 11.00 1.000 11.20 1.000 m

11.40 1.000 a 11.90 1.000 e 11.80 1.000 e

70P--> 12.00 1.000 e e TOP ' Age 0077018 ISE EXCLUDE 0 As PER TEOSdICAL SPECIFICATION 4.2.2.2.8 Figure 5-McGuire Unit 1 Cycle 3 Baseload W(z) for Powers Between 80% and 100% of Reted Thermal Power Within +5% AFD of Measured Target 6000 MWD /MTU r--

vy e-p,

,m r

......m-

.-,-.-r

,---w-,

MEIGHT r.Ax IFEET)

Ut21 80770m--> 0.00 1.000

.20 1.000 e

.40 1.000 n

.50 1.000 s

.50 1.000 e

{

1.00 1.000 m i

1.20 1.000 e 1.40 1.000 =

s..

1.50 1.000 m 1.80 1.109 s.e 2.00 1.105 2.20 1.101 2.40 1.007 2.00 1.002 s.

2.80 1.008 s.es 3.00 1.023 3.30 1.078 s.es 3.40 1.073 3.00 1.000 g,

3.00 1.003 4.00 1.057 g.3 4.20 1.082 a.se 4.40 1.040 4.90 1.048 3..

4.80 1.043 5.00

't.041 g,

5.30 1.037 2

5.40 1.034 g.gs x

5.00 1.035 E

5.80 1.041 4

S.00 1.051 gW S.20 1.081 S.40 1.005 g

g,,

0.00

1.070 8.00 1.074 3.3 T.00 1.078 8.18 7.30 1.002 7.40 1.088 8.as 7.90 1.008 7.80 1.001 8.00 1.084 8.30 1.008 s.12 8.40 1.008 s

8.90

-1.008 s.se

= 's,

==

8

.u E rl lr,P 8.00 1.007 n

S.00 1.085

=8 g

S.20 1.003 a

m I

l 8'"

9.40 1.001

's,' * :

8.00 1.000 g,,

9.80 1.088 10.00 1.000 s.e 10.20 1.003 s

r--

10.40 1.000 *

,,,s

'O. =

10.80 1.000

11.00 1.000

t R IEIIBfr FEET) 11.20 1.000

11.40 1.000

11.90 1.000

11.80 1.000

TOP--> 12.00 1.000

e TOP Afsp 3077058195 'iern unen Att Prft TEOctICAL SPECIFICATION 4.2.2.2.0 l

Figure 6 l

l McGuire Unit 1 Cycle 3 Base 10ad W(z) for P0wers Between 80% and 100% of Rated' Thermal P0wer l-Within155 AFD of Measured Target 10000 MWD /MTU l

l l

I l

Attachment 3 Analysis.of Significant Hazards Consideration 1

JAsfrequired by 10L CFR 50.91, this analysis is provided concerning whether the: proposed amendments involve significant hazards considerations, as defined by 10 CFR 50.92.

Standards for determination that a proposed.

-amendment involves no significant hazards considerations are if operation fof.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..

.The' proposed amendments ensure that plant operation is consistent with the.

' design and safety evaluation conclusion statements made in the McGuire Unit 1 (Cycle 3 reload safety evaluation and ensure that those conclusions remain valid. The reference safety cvaluation report submitted by Mr. H. B. Tucker's

, November 14, 1983 letter to Mr.' H. R. Denton summarized the evaluation,

. performed-on-the region-by-region reload transition from the McGuire Units 1

~

and 2 standard (STD) fueled cores to_ cores with all optimized fuel (OFA).

.The report examined the differences between the Westinghouse STD. design and

$3 e OFA' design and evaluated the effects of these differences for the transition

g;

.to an.all'0FA core. The report justifies the compatibility of the OFA design with the: STD design in a transition core as well as a full OFA-core. The

' report-also contained-summaries of the mechanical,-nuclear, thermal-hydraulic,

'and accident evaluations.

.The McGuire Unit 1/ Cycle 3 reload safety evaluation (Attachment 2A) presents an evaluation which demonstrates that the core reload will not ' adversely affect -

(the safety of the. plant. All of the accidents comprising the licensing bases i

~ which could potentially be affected by the fuel reload were reviewed for.the

' Unit 1, Cycle 3 design. : The results of new analyses and the justification for the applicability of cprevicus results for the remaining analyses is presented

.in the cycle specific reload safety. evaluation. The results of evaluation /'

~

analysis and tests lead to~the following' conclusions:

a.

The Westinghou'se OFA. reload fuel assemblies for McGuire 1 and

~

-2 are mechanically compatible with the STD design, control rods, and: reactor internals interfaces. Both fuel. assemblies-satisfy the design bases-for the McGuire units.

Lb. ' Changes in the nuclear characteristics'due to the transition-

- from STD to 0FA fuel will be within-the range normally seen-

. a frem cycle to cycle due to fuelimanagement effects.

o c.. The reload 0FAs are hydraulically compatible with the STD 1 design.-

o

, ~

d._"The accident analyses for: the.0FA. transition core'were shown

to provide acceptable results by meeting the applicable criteria, such as,' minimum DNBR, peak pressure, and peak clad temperature,

~ he;previously reviewed and licensed safety limits-

=asLrequired.

T lare met..

+

N A

f

.wmer..om v

.,-.--.w--.~

m -

u.am..

...~..+,4.,

m Page 2 Plant operating limitations given in the Technical Specifications e.

affected by the reload will be satisfied with the proposed changes.

- From these evaluations, it is concluded that the Unit 1 Cycle 3 design does not cause the prevfously acceptable safety limits to be exceeded.

The commission has provided examples of amendments likely to involve no signi-ficant 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 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 potentially be affected by the fuel reload were reviewed for the Unit 1 Cycle 3 design and conclude that the reload design does not cause the previously acceptable safety limits to be exceeded, the above example can be applied to this situation.

In addition, the NRC has proposed a no significant hazards. consideration determination for the McGuire Unit 2 Cycle 2 reload amendments (currently under review) which include among them changes similar to these (Ref. 49 FR 50802).

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

v t e Letter Sequence Other
TAC:56746, (Open) TAC:56747, (Open)
Initiation Request, Request Acceptance... Administration Questions Answers Results Approval Other: ML20113H622, ML20113H637, ML20128M056
MONTHYEAR1985-01-11 [Table View]

ML20113H637

Text

1.0 INTRODUCTION

SUMMARY

5.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

1.3 CONCLUSION

5.0 REFERENCES

=.=v

==

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