ML20098E922

Reload Safety Evaluation,Beaver Valley Nuclear Plant Unit 1 Cycle 5

ML20098E922
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Site:	Beaver Valley
Issue date:	07/31/1984
From:	Mary Woods WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
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ML20098E920	List: ML20098E919 ML20098E922
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NUDOCS 8410020121
Download: ML20098E922 (18)

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RELOAD SAFETY EVALUATION BEAVER VALLEY NUCLEAR PLANT UNIT 1 CYCLE 5 July, 1984 Edited by: M. D. Woods Approved: N 1M E. A. Dzenis, dnager Thermal Hydraulic Design Nuclear Fuel Division fh P

DO O j j

Ir TABLE OF CONTENTS Title Page

1.0 INTRODUCTION

'AND

SUMMARY

I

-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 4-2.3 Thermal and Hydraulic Design 4 3.0' POWER CAPABILITY AND ACCIDENT EVALUATION 5 3.1 . Power Capability -5 3.2 Accident Evaluation 5 3.2.1 Kinetic Parameters 5 3.2.2 Control Rod Worths 5 3.2.3 Core Peaking Factors 6 4.0' ' TECHNICAL SPECIFICATION CHANGES 7

5.0 REFERENCES

8 i

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LIST OF TABLES Table Title Page 1 Fuel Assembly Design. Parameters 9 2 Kinetic Characteristics '0 1

3 Shutdown Requirements and Margins 11 LIST OF FIGURES Figure Title Page 1- Core Loading Pattern and Source and 12 Burnable Absorber. Locations 2 Heat Flux Hot Channel Factor - 13 Normalized Operation' Envelope

.(N-loop) 3 Heat Flux Hot Channel Factor - 14 Normalized Operation Envelope (N-1 loop) 11

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1.0 INTRODUCTION

AND

1.1 INTRODUCTION

This report presents an evaluation for Beaver Valley Unit 1, Cycle 5, which demonstrates that the core reload will not adversely affect the safety of the plant. Both three loop and twc loop operation were evaluated. This evaluation was accomplished utilizing the methodology described in WCAP-9272, " Westinghouse Reload Safety Evaluation Methodology" (Reference 1).

Based upon the above referenced methodology, only those incidents ana-lyzed and reported in the FSAR (Reference 2) and N-1 loop safety analyses (References 3-5), which could potentially be affected by this fuel reload, have been reviewed for the Cycle 5 design described herein. The justification for the applicability of previous results is provided.

1.2 GENERAL DESCRIPTION The Beaver Valley Unit 1 reactor core is comprised of 157 fuel assemblies arranged in the core loading pattern configuration shown in Figure 1. The Cycle 5 core configuration features a low leakage pattern. During the Cycle 4/5 refueling, all fifty (50) of the Region 4 assemblies, two (2) of the region 4a assemblies, and twenty-four (24) of the Region 5 assemblies will be replaced with seventy-six (76) Region 7 assemblies. One Region 1 assembly will be replaced with another Region 1 assembly. A summary of the Cycle 5 fuel inventory is given in Table 1.

A new Wet Annular Burnable Absorber (WABA) rod design will be utilized for Cycle 5. The WABA design provides significantly enhanced nuclear characteristics, when compared with the borosilicate absorber rod design. Use of the WABA rods has been approved by an NRC SER which is incorporated into the approved version of the Westinghouse WABA evaluation topical report (Reference 6).

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Nominal core design parameters utilized for Cycle 5 are as follows:

Three Loop (N) Two Loop (N-1)

~ Core Power (MWt) 2652 1724 System Pre',sure (psta) 2250 2250 Core-Inlet Temperature (*F) 542.5 534.4 Core Average Temperature (*F) 579.3 568.5

-Thermal Design' Flow (gpm) 265,500 187,800 Average Linear Power Density (kw/ft) 5.19 3.38 1.3 C,0NCLUSIONS From the evaluation presented in this report, it is concluded that the Cycle'5 design does not cause the previously acceptable safety limits

. for:any incident to be exceeded for three loop or two loop operation.

These conclusions are based on the following assumptions: .

.1. -: Cycle 4 burnup is between 11100 and 13100 MWD /MTV.

2. Cycle 5 burnup is limited to the end-of-life ful1 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 approximately 0-10 ppm residual boron.

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2.0 REACTOR DESIGN 2.1 MECHANICAL DESIGN The mechanical design and fuel rod backfill pressure of the 76 Region 7 fuel assemblies is the same as the Region 6 assemblies, except for the implementation of the IGF/ RECON fuel roa end plugs.* Table 1 compares pertinent design parameters of the various fuel regions. The Region 7 fuel i.as been designed according to the fuel performance model in .

Reference 7. The fuel is designed to operate so that clad flattening will not occur, as orecicted by the Westinghouse model (Reference 8).

The fuel rod internal pressure design basis, Reference 9, is satisfied for all fuel regions.

Westinghouse's experience with Zircaloy clad fuel is described in '

WCAP-8183, " Operational Experience with Westinghouse Cores," Refer-ence 10, which is updated annually.

Wet Annular Burnable Absorber (WABA) rods will be used instead of the .-

standard borosilicate glass absorber rods. The WABA rod design consists of annular pellets of aluminum oxide-boron carbide (A1 023-8 C) 4 <

4 burnable absorber material encapsulated within two concentric Zircaloy tubings. The reactor coolant flows inside the inner tubing and outside the outer tubing of the annular rod. Details of the WABA design are described in Reference 6.

• IGF/ RECON-Internal grip feature pull load (bottom)/reconstitutable (top) end plugs.

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2.2 NUCLEAR DESIGN The Cycle 5 core loading is designed to meet a Fg x P ECCS limit of 5 2.32 x K(z)* for three loop operation and 5 3.03 x K(z) for two loop operation. The two loop (N-1) Fg is an increase from 5 2.77 x K(z) specified for Cycle 4 based on an updated LOCA analysis for N-1 loop operation. The flux difference ( AI) band width during normal operating conditions is + 7% for both two and three loop operation.

Table 2 summarizes the current limits for kinetics characteristics which are based on previously submitted accident analyses. Ncne of these limits is exceeded in Cycle 5.

Cycle 5 control rod worths and requirements are compared in Table 3 with those for Cycle 4 at the most limiting condition (end-of-life). The available shutdown margin exceeds the minimum required margin.

The loading pattern for Cycle 5 is shown in Figure 1. It contains 880 WABA rods located in 72 WABA rod assemblies. Two secondary sources, retained from the Cycle 4 core, are located in positions 43 and H13.

2.3 THERMAL AND HYDRAULIC DESIGN No significant variations in thermal margins will result from the Cycle 5 reload. The DNB core limits and safety analyses used for Cycle 5 are based on the conditions given in Section 1.0.

1

• K(z) - See Figures 2 and 3 1427L:6/840626 4 m ,.

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~

==

3.0 POWER CAPABILITY AND ACCIDENT EVALUATION q 3.1 POWER CAPABILITY g The plant power capability for two and three loop operation is evaluated considering the consequences of those incidents examined in the FSAR and References 3 - 5 using the previously accepted design basis. It is 2-concluded that the core reload will not adversely affect the ability to safely operate at the two and three loop design power levels (Section 1) ,

during Cycle 5. For the overpower transient, the fuel centerline E temperature limit of 4700 F can be accommodated with margin in the Cycle imm 5 core. The time dependent densification model (Reference 11) was used $

for fuel temperature evaluations. The LOCA limit at rated power for l three and two loop can be met by maintaining Fq at or below 2.32 and 3.03, respectively, according to their normalized F genvelope (Figures j 2 & 3). [

=

3.2 ACCIDENT EVALUATION d

The effects of the reload on the design basis and postulated incidents _-

analyzed for 3 loop operation in the FSAR (Reference 2), and for two _

loop operation (References 3 - 5), were examined. In all cases, it was e found that the effects were accommodated within the conservatism of the 7 initial assumptions used in the previous applicable safety analysis. 3 ia 3.2.1 KINETICS PARAMETERS d 2

Table 2 is a summary of the current limits for kinetics parameters. All the Cycle 5 kinetic values fall within the bounds of the current limits.  ;

.2 3.2.2 CONTROL ROD WORTHS Changes in control rod worths may affect differential rod worths, shut- 3 down margin, ejected rod worths, and trip reactivity. Table 2 shows 3 that the maximum differential rod worth of two RCCA control banks moving h together in their highest worth region for Cycle 5 meets the current $

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t I I

limit. Table 3 shows that the Cycle 5 shutdown margin requirements are satisfied. Ejected rod worths for the Cycle 5 design are also within the bounds of the current limits.

t 3.2.3 CORE PEAKING FACTORS Peaking factors for the dropped RCCA incidents were evaluated based on the NRC approved dropped rod methodology described in Reference 12.

Results show that DNB design basis is met for all dropped rod events initiated from full power. Peaking factors following control rod ejection are within the bounds of the current limits. The peaking factors for steamline break have been evaluated and are within the bounds of the previous safety analysis limits.

The Fg of 3.03 for twc ' cop operation (Reference 5) is an increase from 2.77 in Cycle 4 and results in increased initial fuel temperatures for use in accident analyses. This condition was evaluated and the increased temperatures were confirmed to be acceptable.

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I l .:

4.0 TECHNICAL SPECIFICATION CHANGES No changes to the Beaver Valley Unit 1 Technical Specifications are required for Cycle 5.

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5.0 REFERENCES

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

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

3. Letter from J. D. Woodward (Westinghouse) to J. A. Werling (D. L.

Co.), " WRAP N-1 Loop Operation, Non-LOCA Analyses," RPFP-78-34, October 5, 1978.

4. Letter from J. D. Woodward (Westinghouse) to J. A. Werling (D. L.

Co.) " WRAP-N-1 Loop Operation [ECCS] Analyses," RPFP-78-16.

5. Letter from J. A. Triggiani (Westinghouse) to J. D. Sieber (D. L.

Co.), "N-1 Loop Operation," DLW-83-654, June 30, 1983.

6. Skaritka, J., (et. al.), " Westinghouse Wet Annular Burnable Absorber Evaluation Report," WCAP-10021-P-A, Revision 1, (Proprietary),

October, 1983.

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

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

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

10. Skaritka, J. , Iorti, J. A. , Operational Experience with Westinghouse Cores," WCAP-8183, Revision 12, August, 1983.

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

12. 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 5 FUEL ASSEMBLY DESIGN PARAMETERS Region 1 5 6 6A* 7 Enrichment (w/o U-235)* 2.107 2.999 3.248 3.114 3.250 Density (% Theoretical)* 94.80 94.34 94.73 94.38 95.G Number of Assemblies 1 28 51 1 76 Approximate Burnup at 13800 21600 13400 7700 0 BOC 5 (MWD /MTU)

+ All fuel regions are as-built values except Region 7 which is nominal value.

++ Based on EOC4 = 12100 MWD /MTU

• Cycle 4 redesign replacement fuel assembly from Texas Utilities, Comanche Peak Plant.

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, t TABLE 2 i

, i KINETICS CHARACTE915 TICS BEAVER VA,LLEY UNIT 1 - CYCLE 5 NN

,N and N-1 Loop Operation Cycle 5 Changes Current Limits to Current Limits -

Moderator Density **

Coefficient (ap/gm/cc) 0 to 0.43 --

Doppler Temperature Cooefficient (pcm/ F)* -2.9 to -1.4 --

m Least Negative Doppler - Only .

Power Coefficient, Zero to Full Power (pcm/% power)* -6.68 --

Most Negative Doppler - Only '

Power Coefficient Zero to s Full Power (pcm/% power)* -19.4 '

• s i

Delayed Neutron Fraction ,

S,ff,(%) -

0.44 to 0.75 --

,, e Minimum Delayed Neutron Fraction s

) i0.52 (3 loop) --

Rod Ejection BOC 8,y'.f,(%) s l '

, i 0.537 (2 loop)' --

NJs '

! l --

Rod Ejection E0C S'eff,(%) '

0.47,(3 loop)

<- \ O . ! 4, %,' 2 t i oo p ) -- '-

N *

/) / t .

Maximum i'rpmp,t Neutron 'Liff time \ =

(p sec),('- s i )

,' 2f, s .

Maximum Differintial Rod-Worth -

of Two Banks '.'k,ving Together'\ '

(pcm/in.)' i t

, 100 --

f

's ,

\' * '

o'r

, , i

• pcm = 10' ap \ f g

• • The mode ator dens t/ coef/ t :ted\l for the hot zero power, all rods out physics tist condf hion may,pe nehtive tx ty EX 5. The coefficient will be kept positive at that,.nro power by adr&ibtrativ; controls (with r.ppropriate D bank positic,t and borcn concaritratior.). ,

--Ind' cates no change. _

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TABLE 3 END-OF-CYCLE SHUTDOWN REQUIREMENTS AND MARGINS BEAVER VALLEY UNIT 1 - CYCLE 5 3 Loop (N) 2 Loop ( A-1)

Control Rod Worth (%Ap) Cycle 4 Cycle 5 Cycle 5 All Rods Inserted 8.48 8.19 8.19 All Rods Inserted Less Worst Stuck Rod 7.54 7.32 7.32 (1) Less 10*4 6.79 6.59 6.59 i

Control Rod Requirements l

l Reactivity Defects (Combined Doppler, T, g, Void and Redistribution Effects) 2.94 2.91 2.26 l

! Rod Insertion Allowance 0.50 0.50 0.50 l (2) Total Requirements 3.44 3.41 2.76 Shutdown Margin [(1) - (2)] (*4ap) 3.35 3.18 3.83 Required Shutdown Margin (!;ap) 1.77 1.77 2.40 Note: Cycle 4 has standard bps Cycle 5 has Wet Annular Burnable Absorbers (WABAs) 1427L:6/840802 11 i ii i

I FIGURE 1 CORE LOADING PATTERN BEAVER VALLEY UNIT 1 CYCLE 5 73 K J 'r G r N iV _ E J C 3 A 6 7 6 5 7 7 6 7 7 5 9 (8) (8) C 5 7 7 6 6 6 7 7 5 -;

(4) (12) SS (12) (4) Q 5 7 7 6 7 5 7 6 7 7 5 f (16) (?.0) (20) (16) .C., -

5 7 7 5 7 6 7 6 7 5 7 7 5 (4) (16) (16: (4) (20) (4) (16) (16) (4) 7 6 7 5 7 6 7 5 7 6 7 7 h

7 (12) (16) (16) (16) (16) (12) 6 7 6 7 6 7 6 6 6 7 6 7 6 7 6 7 (8) (20) (4) (16: (8) (16) (4) (20) (8) I 7 6 6 5 7 6 6 1 6A 6 7 5 6 6 7 O (20) (8) (8) (20) O 6 7 6 7 6 7 6 6 6 7 6 7 6 7 6 (8) (20) (4) (16' (8) (16) (4) (20) (8) 7 7 6 7 5 7 6 7 5 7 6 7 7 (12) (16) (16) (16: (16) (12) 5 7 7 5 7 6 7 6 7 5 7 7 5 (4) (16) (16' (4) (20) (4) (16) (16: (4) 5 7 7 6 7 5 7 6 7 7 5 (16) (20) (20) (16:

5 7 7 6 6 6 7 7 5 (4) (12) ,SS (12) (4) l 5 ,

(8) 7 6 (8) 7 7 5 gt 6 7 6

}

X REGION NUMBER (Y) NUMBER OF ABSORBERS SS SECONDARY SOURCE RODS 12

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FIGURE 2 -.

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