ML20071J949

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

ML20071J949
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Site:	Beaver Valley
Issue date:	02/28/1983
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RELOAD SAFETY EVALUATION BEAVER VALLEY NUCLEAR PLANT UNIT 1 CYCLE 4 February, 1983 l

8305270190 830523 PDR ADOCK 05000

l RELOAD SAFETY EVALUATION BEAVER VALLEY NUCLEAR PLANT UNIT 1 CYCLE 4 February, 1983 Edited by:

D. J. Petrarca Approved:

/

t

[.G.Arlotti, Manager 4

Fuel Licensing and Program Support Nuclear Fuel Division 0231L:6

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TABLE OF CONTENTS Title Page

/

1.0 INTRODUCTION

AND

SUMMARY

~1 2.0 REACTOR DESIGN 3

2.1 Mechanical Design 4

2.2 Nuclear Design 4

2.3 Thermal and Hydraulic Design 5

3.0 POWER CAPABILITY AND ACCIDENT EVALUATION 6

6 3.1 Power Capability ~

3.2 Accident Evaluation

~

6 3.2.1 Kinetic Parameters 7

4 3.2.2 Control Rod Worths 7

3.2.3 Core Peaking Factors 7

4.0 TECHNICAL SPECIFICATION CHANGES 8

5.0 REFERENCES

9 h

l.

i 0231L:6

l LIST OF TABLES Table Title Page

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1 Fuel Assembly Design Parimet.crs 10 2

Kinetic Characteristics 11 3A Shutdown Requiremeritd and Margins 12 W/.BA Design 3B Shutdown Requ2rements and Margins 13 Standard Burnable Abrorber Design LIST OF FIGURES Figure Title Page 1

Core Loading Pattern and Source and 14 Burnable Absorber Locations 2

Heat Flux Hot Channel Factor -

15 Normalized Operation Envelope (N-loop) h 1

ii 0231L:6

A j

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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 i

safety of the plant. This evaluation was accomplished utilizing the methodology described in WCAP-9273, " Westinghouse Reload Safety EvaluationMethodology"(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 assem-blies arranged in the core loading pattern configuration shown in Figure 1.

During the Cycle 3/4 refueling, 52 fuel assemblies will be replaced with Region 6 fuel and one Region 1 assembly will be replaced with another Region 1 assembly. A summary of the Cycle 4 fuel inventory is given in Table 1.

A new Wet Annular. Burnable Absorber (WABA) rod design may be utilized for Cycle 4.

The WABA design provides significantly enhanced nuclear characteristics, when compared with the borosilicate absorber rod design. Fuel cycle benefits result from the reduced parasitic neutron absorption of Zircaloy compared to stainless steel tubes, increased water fraction in the burnable absorber cell, and a reduced boron penalty at the end of each cycle.

0231L:6 1

The materials, mechanical, thermal hydraulic, and nuclear design evaluations of the WABA rods are presented in WCAP-10021, Revision 1,

" Westinghouse Wet Annular Burnable Absorber Evaluation Report,"

Reference 3, which has been submitted to NRC for generic review and approval.

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.

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 previously acceptable safety limits for any incident to be exceeded. This conclusion is based on the fellowing:

1.

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

K 0231L:6 2

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.

'f

/

Definition: With control rods fully withdrawn and approximately 0 to 10 ppm residual boron.

0231L:6 3

2.0 REACTOR DESIGN 2.1 MECHANICAL DESIGN The mechanical design of the Region 6 fuel assemblies is the same as 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 inte Region 5 fuel assemblies; and (3) fuel rod backfill pressure of 350 psig. Table 1 compares pertinent design parameters of the various fuel regions. The Region 6 fuel has been designed according to the fuel performance model in Reference 4.

The fuel is designed to operate so that clad flattening will not occur as predicted by the Westinghouse model, Reference 5.

The fue.1 rod internal pressure design basis, Reference 6, is satisfied for all fuel regions.

WABA rods may be used instead of the standard borosilicate glass absorber rods (see Figure 1). The WABA rod design consists of annular pellets of aluminum oxide-boron carbide (Al 0 -B C) burnable 23 4 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 WCAP-10021, Revision 1, " Westinghouse Wet Annular Burnable Absorber Evaluation Report," Reference 3.

Westinghouse's experience with Zircaloy clad fuel, is described in WCAP-8183, " Operational Experience with Westinghouse Cores," Refe-rence 7, which is updated annually.

d 2.2 NUCLEAR DESIGN Two core designs are addressed in this report. One design utilizes all standard (borosilicate glass) burnable absorber rods and the other features Wet Annular Burnable Absorber (WABA) rods. The loading 0231L:6 4

pattern, identical for both designs as shown in Figure 1, contains 560 burnable absorber (BA) rods located in 44 BA rod assemblies.

Fuel assemblies and BA assemblies are placed in identical locations for both designs differing only in the type of burnable absorber used.

The C:.le 4 core loading is designed to meet a F x P ECCS limit of g

< 2.32 x K(z)* for a flux difference (AI) band width during norma'l 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. Table 2 is applicable for the Cycle 4 core using either the WABA design or the standard burnable absorber design.

Table 3A provides the control rod worths and requirements at the most limiting condition during the cycle (end-of-life) for the WABA design.

Table 3B provides the same information for the standard burnable absorber design. The required shutdown margin is based on previously submitted accident analysis. The available shutdown margin for both burnable absorber designs exceeds the minimum required.

The 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.

2.3 THERMAL AND HYDRAULIC DESIGN t,

No significant variations in thermal margins will result from the Cycle 4 reload. The DNB core limits and safety analyses used for Cycle l

4 are based on the conditions given in Section 1.0.

The thermal hydraulic design considerations of the test assemblies are described in Reference 8.

K(z) - Figure 2 0231L:6 5

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(2) using the previously accepted design basis.

It is concluded that the core reload will not adversely affect the abi,lity 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(9) was used for fuel temperature evaluations. The LOCA limit at rated power can be met by maintaining F at or below 2.32 according to g

the normalized F envelope (Figure 2).

g 3.2 ACCIDENT EVALUATION The effects of the reload on the design basis and postulated incidents 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.

The new dropped rod methodology was instituted for this cycle as described in Reference 10.

Formal notice to the NRC will be made relating:

(1) the new dropped rod evaluation has been successfully applied to Cycle 4, (2) the interim restriction on rod control and insertion are no longer necessary, and (3) requesting that the NRC approve the removal of the interim operational restrictions. Plant operation should continue under the interim operating restrictions until explicit NRC approval is received.

0231L:6 6

i

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 these 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 parameters current limits.

All the kinetic values fall within the bounds of the current limits; Table 2 is applicable to both Cycle 4 designs.

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. Tables 3A and 3B show that the Cycle 4 shutdown margin require-ments are satisfied for both designs.

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

3.2.3 CORE PEAKING FACTORS Evaluation of peaking factors for the rod out of position and dropped RCCA incidents show that DNBR is maintained above 1.3.

A peaking factor evaluation for the hypothetical steamline break transient showed that the DNBR is maintained above 1.3.

The peaking factors following control rod ejection are within the limits of previous analysis.

0231L:6 7

4.0 TECHNICAL SPECIFICATION CHANGES l

l No changes to the Beaver Valley Unit 1 Technical Specifictions are required for Cycle 4.

0231L:6 8

5.0 REFERENCES

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. Rahe, E.

P.', Westinghouse letter to C. Thomas, NRC, Number NS-EPR-2670, October 18, 1982,

Subject:

Westinghouse Wet Annular Burnable Absorber Evaluation Report," WCAP-10021, Revision 1, (Proprietary).

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

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

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

WCAP-9286, July 1978.

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

10. Letter No. NS-EPR-2545, E. P. Rahe (Westinghouse) to C. H. Berlinger (NRC), January 20, 1982.

l i

0231L:6 9

l TABLE 1 BEAVER VALLEY UNIT 1 - CYCLE 4 FUEL ASSEMBLY DESIGN PARAMETERS **

Region 1

4 4A*

5 6

Enrichment (w/o U-235)*

2.106 3.199 3.203 2.999 3.25 Density (% Theoretical)*

94.80 94.38 94.38 94.34 94.5 Number of Assemblies 1

50 2

52 52 Approximate Burnup at 13800 20700 21700 9200 0

Beginning of Cycle 4 (MWD /MTV)

• Optimized Fuel - Zirc grid

+ All fuel regions are as-built values except Region 6 which are nominal values.

• • Applicable to either the WABA design or the standard burnable absorber design.

0231L:6 10

TABLE 2 KINETICS CHARACTERISTICS +

BEAVER VALLEY UNIT 1 - CYCLE 4 Cycle 4 Changes Current Limit to Current Limits Moderator Density **

Coefficient (Ap/gm/cc) 0 to 0.43 Doppler Temperature Cooefficient (pcm/ F)*

-2.9 to -1.4 Least Negative. Doppler - Only Power Coefficient, Zero to Full Power (pcm/% power)*

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

-19.4 Delayed Neutron Fraction Oeff,(%)

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

0.52 Rod Ejection EOC S,ff,(%)

0.47 Maximum Prompt Neutron Lifetime (p 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 0 bank position and boron concentration).

+ Applicable to either the WABA design or the standard burnable absorber design.

--Indicates no change.

0231L:6 11

TABLE 3A END-OF-CYCLE SHUTDOWN REQUIREMENTS AND MARGINS BEAVER VALLEY UNIT 1 - CYCLE 4 WABA DESIGN Control Rod Worth (%Ap)

Cycle 3 Cycle 4 All Rods Inserted 8.14 8.49 All Rods Inserted Less Worst Stuck Rod 6.47 7.56 (1) Less 10%

5.82 6.80 Control Rod Requirements Reactivity Defects (Combined Doppler, T,yg, Void and Redistribution Effects) 2.86 2.98 Rod Insertion Allowance 0.50 0.50 (2) Total Requirements 3.36 3.48 Shutdown Margin [(1) - (2)] (%Ap) 2.46 3.32 Required Shutdown Margin (%ap) 1.77 1.77 0231L:6 12

TABLE 3B END-OF-CYCLE SHUTDOWN REQUIREMENTS AND MARGINS BEAVER VALLEY UNIT 1 - CYCLE 4 STANDARD BURNABLE ABSORBER DESIGN 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%

5.62 6.89 Control Rod Requirements Reactivity Defects (Combined Doppler, T,yg, Void and Redistribution Effects) 2.86 3.06 Rod Insertion 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 0231L:6 13

Figure 1 CORE LOADING PATTERN

• BEAVER VALLEY UNIT 1 CYCLE 4 R

P N

M L

K 4

H G

F E

D C

B A

4 6

4 1

4 6

6 4

6 6

4 2

(4)

(4) 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) 6 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)

~

l12) 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 13 (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 *

-.SS

- Secondary Source Rods

• Applicable for both the WABA rod and the standard burnable absorber rod usage 14

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