Text

Reload Safety Evaluation,Salem Nuclear Plant Unit 2, Cycle 3
	ML18092A508
Person / Time
Site:	Salem
Issue date:	11/27/1984
From:	Kapil S, Piplica A, Weber M; Public Service Enterprise Group
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ML18092A506	List: ML18092A505; ML18092A508;
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1811L: 6/841127 RELOAD SAFETY EVALUATION SALEM NUCLEAR PLANT UNIT 2 CYCLE 3 Edited by:

M. M. Weber A. N. Piplica E. F. Pulver M. E. Pohlman

. Approved: _l~<rWJ..;._.::;_.

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S. K. Kapil, Ma;°~

Core Engineering Nuclear Fuel Division

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TABLE OF CONTENTS Title

1.0 INTRODUCTION

AND

SUMMARY

1.1 Introduction 1.2 Generic Description 1.3 Conclusions 2.0 REACTOR DESIGN 2.1 Mechanical Design 2.2 Nuclear Design 2.3 Thermal and Hydraulic Design 3.0 POWER CAPABILITY AND ACCIDENT EVALUATION 3.1 Power Capability 3.2 Accident Evaluation 3.2.1 Kinetic Parameters 3.2.2 Control Rod Worths 3.3.2 Core Peaking Factors 3.3 Incidents Reanalyzed 4.0 TECHNICAL SPECIFICATION CHANGES 4.1 Specification 3/4.1 Reactivity 4.2 Specification 3/4.2 Power Distribution Limits

5. O REFERENCES 1811L:6/841128 1

1 1

2 3

3 3

4 5

5 5

5 6

6 7

8 8

8 9

Table 1

2 3

Figure 1

2 3

4 LIST OF TABLES Title Fuel Assembly Design Parameters Kinetic Characteristics End-of-Cycle Shutdown Requirements and Margins LIST OF FIGURES Title Core Loading Pattern and Source and Burnable Poison Locations Revised Rod Bank Insertion Limits vs. Thermal Power Acceptable with 0.3 FAH Multiplier Revised Rod Cluster Control Assembly Pattern Revised K(Z) Normalized Fq(z) as a Function of Core Height 1811L: 6/841130 10 11 12 13 14 15 16

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1.0 INTRODUCTION

ANO

SUMMARY

1.1 INTRODUCTION

This report presents an evaluation for Salem Unit 2, Cycle 3, which demonstrates that the core reload will not adversely affect the safety of the plant.

This evaluation was accomplished utilizing the method-ology described in WCAP-9273, "Westinghouse Reload Safety Evaluation Methodology" (1).

Based upon the above referenced methodology, only those incidents analyzed and reported in the FSAR( 2) which could potenti~lly be affected by this fuel reload have been reviewed for the Cycle 3 design described herein.

The justification for the applicability of previous results. is provided.

1.2 GENERAL DESCRIPTION The Salem Unit 2 reactor core is comprised of 193 fuel assemblies arranged in the core loading pattern configuratiori shown in Figure 1.

During the Cycle 2/3 refueling, 68 fuel assemblies were replaced with Region 5 fuel.

A summary of the Cycle 3 fuel inventory is given in Table 1.

Nominal core design parameters util.ized for Cycle 3 are as follows:

Core Power (MWt)

System Pressure (psia)

Core Inlet Temperature (°F)

Thermal Design Flow (gpm)

Average Linear Power Density (kw/ft) 1811L: 6/841128 1

3411 2250 545.0 349;200 5.43

1.3 CONCLUS~ONS From the evaluation presented in this report, it is concluded that the Cycle 3 design does not cause the previously acceptable safety limits for any incident to be exceeded.

This conclusion is based on the following:

1.

Cycle_2 burnup of 5,660 MWD/MTU.

2.

Cycle 3 burnup is limited to 16,700 MWO/MTU.

3.

There is adherence to plant operating limitations given in the Technical Specifi cations.

_?

1811L: 6/841128 2

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I 2.0 REACTOR DESIGN 2.1 MECHANICAL DESIGN The mechanical design of the Region 5 fuel assemblies is the same as the Region 4 assemblies with exception of an end plug design change which was implemented to facilitate modified fuel rod loading_ techniques.

Table 1 compares pertinent design parameters of the various fuel regions.

The Region 5 fuel has been designed according to the fuel performance model in Reference 3.. The fuel is designed to operate so that.clad flattening will not occur as predicted by the Westinghouse mod~l~ Reference 4~ The fuel rod internal pressure d~sign basis, Reference 5, is satisfied for all fuel regions.

Westinghouse 1 s experiente with Zircaloy clad fuel* is described in WCAP-8183, 110perational Experience with Westinghouse Cores, 11 Reference* 6, which is updated annually.

2. 2 NUCLEAR DESIGN The Cycle 3 core loading is designed to meet a FQ x P ECCS limit of

~ 2.32 x K(Z)* for a flux difference (~I) bandwidth during normal operation conditions of +6, -9 percent ~I.

Table 2 provides a summary of the Cycle 3 kinetics characteristics compared with the current limit based on previously submitted accident -

analyses, Reference 2.

Table 3 provides the control 'rod worths and requirements at the most limiting condition during the cycle.

The required shutdown margin is based on*previously submitted accident analyses, Reference 2.

The available shutdown margin exceeds the minimum required..

K(Z) ~ Figure 3.2-2 in Technical Specifications.

1811L: 6/841130 3

.e The loading c;pntains a total of 1424 fresh burnable absorber rods*

located jn Region 5 fuel assemblies and 240 fresh burnable absorber rods located in Region 4 fuel assemblies.

The locations of the burnable absorber and source rods are shown in Figure 1.

2.1 THERMAL AND HYDRAULIC DESIGN No significant variations in thermal margins will result from the Cycle 3 reload.

The present DNB core limits have been found to be applicable for Cycle 3.

1811L:6/841128 4

3.0 POWER CAPABILITY AND ACCIDENT EVALUATION 3.1 POWER CAPABILITY The plant power capability has been evaluated considering the consequences of those incidents examined in the FSAR,( 2) using the previously accepted design basis. It is concluded that the core reload will not adversely affect the ability to safely operate at 100 percent of rated power during Cycle 3.

For the evaluation performed to address overpower concerns, the fuel centerline temperature limit of 4700°F can be accommodated with margin in the Cycle 3. core using the methodology described in Referente 1. The time dependent densification model(l) was used ~or these fuel temperature evaluations.

The LOCA limit at rated power can be met by maintaining FQ at or below 2.32.

3.2 ACCIDENT EVALUATION The effects of the reload*on the design basis and postulated incidents analyzed in the FSAR( 2) were examined.

With the e~ception of the scram curve, it w~~ found that the effects were accommodated wi~hin the conservatism of the assumptions used in the previous applicable safety analyses. *Accidents affect~d by the change in the scram curve are discussed in Section 3.3.'

A core reload can typically affect accident analysis input parameters in the following areas:

core kinetic characteristics, control tad worths, and core peaking factors.

Cycle 3 parameters in each of these three areas were examined as discussed below to ascertain *whether new accident analyses were required.

3. 2.1 KINETICS PARAMETERS A comparison of the range of values encompassing.the Cycle 3 kinetics parameters with the current limits is given in Table 2.

All kinetic values fall within the bounds of the current limits.

1811L: 6/841128 5

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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 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 are satisfied.

Cycle 3 has a normalized trip reactivity insertion rate which 1s slightly different from the current limit,. Reference 11.

The effects of this reduced normalized trip reactivity rate have been evaluated for those accidents affected and compared to previous analyses.

The only significant non-conservative deviations, with re~pect to the current limit (Reference 11) between the two trip insertfon curves, occur for the first 10 percent of rod insertion. The remaining portion of the trip insertion curve is conservative with respect to the current limit, Ref ere nee 11.

Slow transients are rel.3.tively insensitive to the trip react1vity i~sertion rate.

Fast tran~ients are evaluated to confi~m that the*

limiting transient.conditions are unchanged.

The only accident impacted

. by this change to the scram curve is loss of flow.

3.2.3 CORE PEAKING FACTORS Peaking factors for the dropped RCCA incidents were evaluated based on the approved dropped rod'methodology described in Reference 8.. Peaking factbrs following control rod ejection are within the bounds of the currents limits. The peaking factors for steamline break have been evaluated and are within the bounds of the previous safety analysis limits.

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.e 3.3 INCIDENTS REANALYZED An investigation of the transients affected by the change in the Cycle 3 reactivity insertion curve has shown that the conclusions of the FSAR will not change, therefore.no reanalysis was performed.

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.e 4.0 TECHNICAL SPECIFICATION CHANGES This section contains the technical content of the proposed changes t6 the Salem Unit 2 Technical Specifications.

These changes are consistent with the plant operation necessary for the des*ign and safety evaluation conclusions stated previously to remain valid.

4.1 SPECIFICATION 3/4.1 REACTIVITY 4.1.1 Control Rod Insertion Limits Figure 3.1-1 A proposal has been made to modify the power dependent rod bank insertion limits to be identical to the Salem Unit 1 rod insertion limit (figure 2).

This change is consistent with the proposed reidentification of the control banks A, 8, C and D tti closely resemble the Unit 1 control rod pattern (figure 3).

4.2 SPECIFICATION 3/4.2 POWER DISTRIBUTION LIMITS 4.2.1 Section 3/4.2.1 It has been proposed that the K(Z) curve for Salem Unit 2 be modified so that it is identical to Salem Unit 1 K(Z) curve (figure 4), based on the small break LOCA analysis performed for Salem Unit 1 (referente 10).

1811L:6/841128 8

5.0 REFERENCES

1. Bordelon, F. M., et al.,

11Westinghouse Reload Safety Evaluation Methodology 11, WCAP-9273, March 1978.

2.

11Salem Unit No. 2 Updated Final Safety Analysis Report, 11 USNRC Docket Numbers 50-272 and 311, July 22, _1983..

3. Miller, J. V. (ed.),

11 Improved Analytical Model used in Westinghouse Fuel Rod Design Computations 11, WCAP-8785, October 1976.

4. George, R. A., et al.,

11R"evised Clad Flattening Model 11, WCAP-8381, July 1974.

5. Risher, D. H., et al.,

11Safety Analysis for the Revised Fuel Rod Internal *Pressure Design Basis 11, WCAP-8964, June 1977.

6. Iorii, J. A., Skaritka, J.,

110perational Experience with Westinghouse Cores 11, WCAP-8183, Revision 13, September 1984.

7. Hellman, J.M., (ed.),

11 Fuel Densification Experimental Results and Model for Reactor Operation 11, WCAP-8219-A, March 1975.

8. Letter from Cecil D. Thomas (NRC) to E.P. Rahe, Jr.,~

Subject:

Acceptance for Referencing of Licensing Topical Report WCAP-10297-A(p), WCAP-10248-A (NS-EPR-2545) Entitled 11Dropped Rod Methodology for* Negative Flux Rate Trip Plants, 11 March 31, 1983.

9. Pertraca, D.J.,

11Reload Safety Evaluation Salem Nuclear Plant Unit 2 Cycle 2, Revision 111, April 1983.

10. Letter from D.W. Williams (Westinghouse) to R.A. Uderitz (PSE&G)

FP-PS-229, April 24, 1979.

11. Pohlman, M. E., Piplica, A. N.,

11Reload Safety Evaluation Salem Unit 1 Cycle 6, Revision 111, July 1984.

12. Letter from T. R. Croasdaile (Westinghouse) to K. Brenner and J. T. Boettger (PSE&G) 84PS*-G-058, June 28, 1984.

. 1811L: 6/841130 9

_Region TABLE 1 FUEL ASSEMBLY DESIGN PARAMETERS SALEM UNIT 2 - CYCLE 3 3

4 Enrichment (w/o U-235) 3.12*

3,42*

Geometric Density 94.55*

94.54*

% theoretical)

. N'umber of 53 72 Assemblies Approximate Burn up 19,350 5,500 at Beginning of Cycle 3 (MWD/MTU)+

v~lues are as-built value~.

+ Based on-EOC2 *of 5,660 MWD/MTU.

1811L: 6/841128 10 SA SB 3.80 3:40 95.00 95.00 64 4

0 0

TABLE 2 KINETICS CHARACTERISTICS SALEM UNIT 2 - CYCLE 3 Moderator Density Coefficient (Ap/gm/cc)

Doppler Temperature Coefficient (pcm/°F)

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

Most Negative -Doppler - Only Power Coeff1cient Zero to Full Power (pcm/%power)

Minimum Delayed-Neutron Fraction aeff' (percent) aeff' (percent) minimum (Rod ejection only)

Maximum Prompt Neutron Lifetime (µ sec)

Maximum Differential Rod Worth of Two Banks Moving Together (pcm/in.)*

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pcm = 10-5 Ap

'1811L: 6/841130 Current Limit (2) (9) (12)

.43 to 0

-2. 9 to -1. 0

-10.2 to -6.7

-43.5 to -14.7 0.44 to 0.75 0.55 26 100 11 -

Cycle 3

.43 to 0

-2. 9 to -1. 0

-10.2 to -6.7

-43.5 to -14.7 0.44 to 0.75

-0.55 26 100

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TABLE 3 -

END-OF-CYCLE SHUTDOWN REQUIREMENTS AND MARGINS SALEM UNIT 2 - CYCLE 2 AND 3 Cycle 2 Control Rod Worth (%'1e)

All Rods Inserted 6.55 A 11 Rods Inserted Less Worst Stuck Rod 5.74 (1) Less 10%

5.17 Control Rod Requirements (%'1p)

Reactivity Defects (Doppler, T 2.76 avg

. Void, Redistribution).

Rod Insertion Allowance 0.50 (2) Total Requirements 3.26 Shutdown Margin [(1)-(2)] (%'1p)

1. 91 Required Shutdown Margin (%'1p)

1. 60 1811L: 6/841128 12 Cycle 3 6.95 5.97 5.37 2.90 0.60 3.50 1.87 1.60

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FIGURE 1 CORE LOADING PATTERN SALEM UNIT 2 CYCLE 3 1r 1so*

R p

N M

L K

J H

G F

E D

c B

A 1

4 4

4 SA 4

4 4

. 12 2

3 4

ML18092A508

Text

1.0 INTRODUCTION

SUMMARY

1.0 INTRODUCTION

SUMMARY

1.1 INTRODUCTION

5.0 REFERENCES

Subject:

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