ML20085L003

Rev 1 to Reload Safety Evaluation,Beaver Valley Nuclear Plant Unit 1,Cycle 4

ML20085L003
Person / Time
Site:	Beaver Valley
Issue date:	09/30/1983
From:	Petrarca D DUQUESNE LIGHT CO.
To:
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ML20085K995	List: ML20085K992 ML20085L003
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0723L:6, 723L:6, NUDOCS 8310210240
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RELOAD SAFETY EVALUATION BEAVER VALLEY NUCLEAR PLANT UNIT 1 CYCLE 4 Revision 1 September, 1983 Edited by:

D. J. Petrarca d-Approved:

s M. G. Arlotti, Manager Fuel Licensing and Program Support Nuclear Fuel Division f

i 8310210240 831011 PDR ADOCK 05000334 i

P PDR 0723L:6

TABLE OF CONTENTS Title Page

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1.0 INTRODUCTION

AND

SUMMARY

1 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 5

3.1 Power Capability 5

f" 3.2 Accident Evaluation 5

3.2.1 Kinetic Parameters 5

3.2.2 Control Rod Worths 6

3.2.3 Core Peaking Factors 6

4.0 TECHNICAL SPECIFICATION CHANGES 7

5.0 REFERENCES

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9 LIST OF TABLES Table Title Page t

1 Fuel Assembly Design Parameters 9

2 Kinetic Characteristics 10 3

' Shutdown Requirements and Margins 11 LIST OF FIGURES Figure Title Page 1

Core Loading Pattern and Source and 12 Burnable A sorber Locations 2

Heat Flux Hot Channel Factor -

13 hormalized Operation Envelope (N-loop) ii 0723L:6

1.0 INTRODUCTION

AND

1.1 INTRODUCTION

This report presents an evaluation for Beaver Valley Unit 1, Cycle 4, which demonstrates that the core reload will not adversely affect the safety of the plant.

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

Based upon the above referenced methodology, only those incidents ana-lyzed and reported in the FSAR(2) which could potentially be affected by this fuel reload have been reviewed for the Cycle 4 design described herein. The justification for the applicability of previous results is provided.

1.2 GENERAL DESCRIPTION The Beaver Valley Unit I reactor core is comprised of 157 fuel assemblies arranged in the core loading pattern configuration shown in Figure 1.

During the cycle 3/4 refueling, 52 fuel assemblies will be r?placed with 51 Region 6 and one Region 6A fuel assemblies, and one l

Region 1 assembly will be replaced with another Region 1 assembly. A summary of the Cycle 4 fuel inventory is given in Table 1.

As in Cycle 3, this cycle will contain two Region 4 demonstration assem-blies, designated in Figure 1 as 4A, of an optimized fuel assembly design. These assemblies will be loaded into the core in a manner that prevents them from becoming lead assemblies during normal operation or leading to more limiting conditions during transient conditions than analyzed for the standard fuel assemblies.

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

^

Core Power (MWt) 2652 System Pressure (psia) 2250 Core Inlet Temperature (*F).

542.5 Core Average Temperature (*F) 579.3 Thermal Design Flow (gpm) 265,500 Average Linear Power Density (kw/ft) 5.19

1.3 CONCLUSION

S From the evaluation presented in this report, it is concluded that the Cycle 4 design does not cause the previcusly acceptable safety limits for any incident to be exceeded. This conclusion is based on the following:

C 1.

Cycle 3 burnup is between 9300 and 11300 MWD /MTU.

2.

Cycle 4 burnup is limited to the end-of-life full power capability

• plus a 1000 MWD /MTU power coastdown.

3.

There is adherence to plant operating limitations given in the Tech-nical Specifications.

Definition: With control rods fully withdrawn and zero ppm residual boron.

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2.0 REACTOR DESIGN 2.1 MECHANICAL DESIGN The mechanical design of the 51 Region 6 fuel assemblies is the same as l

the Region 5 assemblies except that Region 6 assemblies incorporate: (1) the reconstitutable bottom nozzle design; (2) grid modifications consisting of additional guide vanes and tabs along with the corner surface condition already incorporated into Region 5 fuel assemblies; and (3) fuel rod backfill pressure of 350 psig. The mechanical design of the one Region 6A feel assembly is the same as the Region 5 assemblies, including fuel rod backfill pressure of 450 psig. Table 1 compares pertinent design parameters of the various fuel regions.

The Region 6 and Region 6A 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 model, Reference 4.

The fuel rod internal pressure design basis,

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Reference 5, is satisfied for all fuel regions.

Westinghouse's experience with Zircaloy clad fuel, is described in WCAP-8183, " Operational Experience with Westinghouse Cores," Refer-I ence 6, which is updated annually.

l 2.2 NUCLEAR DESIGN The Cycle 4 core loading is designed to meet a F x P ECCS limit of g

5 2.32 x K(z)* for a flux difference ( AI) band width during normal operation conditions of +7%AI.

Table 2 provides a summary of changes in Cycle 4 kinetic characteristics compared with the current limit based on previously submitted accident analyses.

K(z) - Figure 2 a

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Table 3 provides the control rod worths and requirements at the most l

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.

The loading pattern contains 560 burnable absorber (BA) rods located in 44 BA rod assemblies.

Th( secondary sources (with associated absorber rods) located in positions H3 and H13 in Cycle 3 will be removed.

Secondary sources (without bas) in positions J6 and G10 in Cycle 3 will be placed in H3 and H13 in Cycle 4.

Location of the burnable absorber rods and secondary sources are shown in Figure 1.

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.2.3 THdRMAL AND HYDRAULIC DESIGN No significant variations in thermal margins will result from the Cycle 4 reloac. The DNB core limits and safety analyses used for Cycle 4 are based on the conditions given in Section 1.0.

The thermal l

hydraulic design consideratior.s of the test assemblies are described in Reference 7.

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l 3.0 p0WER CAPABILITY AND ACCIDENT EVALUATION

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3.1 POWER CAPABILITY The plant power capability has been evaluated considering the con-sequences 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 the design power level (Section 1) during Cycle 4.

For the overpower transient, the fuel centerline temperature limit of 4700 F can be accommodated with margin in the Cycle 4 core. The time dependent densification model(8) was used for fuel temperature evaluations.

The LOCA limit at l

rated power can be met by maintaining F at or below 2.32 according to g

the normalized F envelope (Figure 2).

g 1

4 j

3.2 ACCIDENT EVALUATION i

The effects of the reload on the design basis and postulated incidents l

analyzed in the FSAR were examined.

In all cases, it was found that the effects were accommodated within the conservatism of the initial assump-

{

tions used in the previous applicable safety analysis.

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

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

Cycle 4 parameters in each of thase three areas were examined as discussed in the following subsections to ascertain whether new accident analyses were required.

3.2.1 KINETICS PARAMETERS Table 2 is a summary of the Cycle 4 kinetics paramete,rs current limits.

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

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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 4 meets the current limit. Table 3 shows that the Cycle 4 shutdown margin requirements are j

satisfied.

Ejected rod worths for the Cycle 4 design are also within the bounds of the current limits.

3.2.3 CORE PEAKING FACTORS Peaking factors for the dropped RCCA incidents were evaluated based on the NRC approved dropped rod methodology described in Reference 9.

Results show that DNB design basis is met for-all dropped rod events initiated from full power and therefore the interim operating restrictions may be removed.(9) Peaking factors following control rod ejection are within the bounds of the current limits.

The peaking factors for steamline break ave been evaluated and are within the bounds of the previous safety analysis limits.

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4.0 TECHNICAL SPECIFICATION CHANGES No changes to the Beaver Valley Unit 1 Technical Specifictions are required for Cycle 4.

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5.0 REFERENCES

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

2. " Beaver Valley Unit No. 1 Final Safety Analysis Report", Docket Number 50-334.

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

" Improved Analytical Model used in Westing-house Fuel Rod Design Computations", WCAP-8785, October 1976.

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

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

6. Jones, R. G., Iorii, J.A., " Operational Experience with Westinghouse Cores", WCAP-8183, Revision 11, May 1982.

7. O'Hara, T.L. (Ed.), " Optimized Fuel Assembly Demonstration Program",

WCAP-9286, July 1978.

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

9. 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.

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TABLE 1 BEAVER VALLEY UNIT 1 CYCLE 4 FUEL ASSEMBLY DESIGN PARAMETERS REGION I

4 4A*

5 6

6A 1

f 4

Enrichment (w/o U-235)+

'2.106 3.199 3.203 2.999 3.25 3.10 Density (% Theoretical)+

94.80 94.38 94.38 94.34 94.5 94.5 Number of AssemL11es 1

50 2

52 51 1

Approximate Burnup at++

13,800 21,100 22,100 9,400 0

0 l

4 Be. ginning of Cycle 4

)

(MWD /MTU) i Approxinate Burnup 25,200 30,000 35,000 23,200 13,000 7,730 Predicted for EOC 4 Region Average 0

Optimized Fuel Zirc grid All fuel regions are as-built values except Region 6 and 6A which are nominal values.

+

++

Based on E0C3 = 10637 MWD /MTU l

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TABLE 2 KINETICS CHARACTERISTICS BEAVER VALLEY UNIT 1 - CYCLE 4 Cycle 4 Changes Current Limit (2) to Current Limits Moderator Density **

Coefficient (Ap/gm/cc)

O to 0.43 Doppler Temperature Coefficient (pcm/ F)*

-2.9 to -1.4 0

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

-6.68 Most Negative Doppler - Only Power Coefficient Zero to Full Power (pcm/% power)*

-19.4 Delayed Neutron Fraction B,ff,(%)

0.44 to 0.75 Minimum Delayed Neutron Fraction Rod Ejection BOC 6,ff,(%)

0.52 Rod Ejection EOC 6,ff,(%)

0.47 Maximum Prompt Neutron Lifetime (y sec) 26 Maximum Differential Rod Worth of Two Banks Moving Together (pcm/in.)*

100

• pcm = 10-5,

3

• • The moderator density coefficient for the hot zero power, all rods out physics test condition may be negative at the BOC 4.

The coefficient will be kept positive at that zero power by administrative controls (with appropriate D bank position and boron concentration).

--Indicates no change.

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TABLE 3 l

END-OF-CYCLE SHUTDOWN REQUIREMENTS AND MARGINS BEAVER VALLEY UNIT 1 - CYCLE 4 Control Rod Worth (%Ap)

Cycle 3 Cycle 4 All Rods Inserted 8.14 8.59 All Rods Inserted Less Worst Stuck Rod 6.47 7.65

.(1) Less 10%

5.82 6.89 Control Rod Reouirements Reactivity Defects (Combined Doppler, T,yg, Void and Redistribution Effects) 2.86 3.06 Rod Insertio.n Allowance 0.50 0.50 (2) Total Requirements 3.36 3.56 Shutdown Margin [(1) - (2)] (%Ap) 2.46 3.33 Required Shutdown Margin (%Ap) 1.77 1.77 i

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Figure 1 REVZSED CORE LOADING PATTERN BEAVER VALLEY UNIT 1 CYCLE 4 R

P N

M L

K J

H G

F E

D C'

B A

4 6

4 1

4 6

6 4

6 6

4 (4)

(4) 2 4

6 5

5 4

5 5

6 4

3 (12)

SS (12) 4 5

6 5

6 5

6 5

6 5

4 4

(12)

(20)

(20)

(12) 4 6

6 4

4 5

5 5

4 4

6 6

4 5

(12)

(4)

(12) 6 5

5 4

5 6

4 6

5 4

5 5

6 6

(12)

(20)

(20)

(12) 4 6

5 6

5 6

4 5

4 6

5 6

5 6

4 7

(4)

(20)

(20)

20)

(20)

(4) 6A 4

4A 5

5 4

5 1

5 4

5 5

4A 4

6 8

(4)

(4) 4 6

5 6

5 6

4 5

4 6

5 6

5 6

4 9

(4)

(20)

(20)

(20)

(20)

(4) 6 5

5 4

5 6

4 6

5 4

5 5

6 10 (12) 20)

(20)

(12) 4 6

6 4

4 5

5 5-4 4

6 6

4 11 (12)

(4)

[12) 4 5

6 5'

6 5

6 5

6 5

4 12 (12)

(20)

(20)

(12) 4 6

5 5

4 5

5 6

4

'3 (12)

SS (12) 4 6

6 4

6 6

4 14 (4)

(4) 4 6

4

-15 X

- Region Number (Y)

- Number of Absorber Rods

?S

- Secondary Source Rods 12

FIGURE 2 K(Z) - NOPJ4ALIZED FQ(Z)

AS A FUNCTION OF CORE HEIGHT N-LOOP BEAVER VALLEY - UNIT 1 1.0

_ _ _ - ~ ~ - -

~~

' (lDiS r-07.94)

~~^

=_

0.8 Gg A

S 30.6 s

$ 0.4

~

.._.c =,r f; % 02 15 I

0.2

~_-

._____.-.r__-..___.__.

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l l

0' 2.

4 46 8:

10 12 14 CORE HEIGHT (FT)

+

13