ML20207N821

Safety Evaluation Supporting Amend 123 to License NPF-3

ML20207N821
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Site:	Davis Besse
Issue date:	10/03/1988
From:	Office of Nuclear Reactor Regulation
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NUDOCS 8810190380
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.

123 TO FACILITY OPERATING LICENSE NO. NPF-3, TOLED0 EDISON COMPANY AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 DOCKET NO. 50-346

1.0 INTRODUCTION

In a letter dated May 18, 1988 Toledo Edison Company submitted an application (Ref. 1) to amend the Appendix A Technical Specifications (TS's) for the Davis-Besse Nuclear Power Station, Unit No.1, to pennit operation for Cycle 6.

The safety analyses performed and the resulting modifications to the TS's are described in the Cycle 6 reload report (Ref. 2).

The reference cycle for this reload is Cycle 5.

All accidents analyzed in Chapter 15 of the Updated Safety Analysis Report have been reexamined with respect to Cycle 6 parameters.

1.1 Description of the Cycle 6 Core The Cycle 6 core consists of 177 fuel assembles (FA's), each of which is a 15x15 array containing 208 fuel rods,16 control rod nide tubes and one incore instrument guide tube. Cycle 6 will operate 1' a feed-and-bleed mode.

The core reactivity will be controlled by 53 full length Ag-In-Cd control rod assemblies (CRA's), 64 burnable poison rod essemblies (BPRA's) and soluble boron.

Eight Inconel-600 axial power shaping rods (gray APSR's) are provided for additional control of the axial power distribution. The licensed core full power level is 2772 L't.

1.2 Significant Areas of Review for this Reload For the most part, Cycle 6 will be identical to Cycle 5 and many of the TS changes are the result of the changes associated with the insertion of new fuel, cycle lifetime, and the time of withdrawal of APSR(s) which are of ten made in Babcock & Wilcox (B&W) reactors. However, there are several significant changes associated with this reload:

(1) the APSR's will be changed from the "black" (Ag-In-Cd) APSR's used in previous cycles to "gray" (Inconel-600) APSR's; (2) the NOODLE code was used to generate the core physics parameters while previous cycles used the PDQ07 code; (3) the LYNXT crossflow codes were used in the thermal-hydraulic design evaluation; (4) the revised power imbalance detector correlation (PIOC) was used; (5) a reduced physics testing program will be implemented; and (6) the two regenerative neutron sources (RNS) will be removed from the core. This last change was addressed in a separate 10 CFR 50.59 review and was not reviewed or evaluated by the NRC.

All of these changes have been previously made end approved on other B&W

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2.0 EVALUATION OF THE FUEL SYSTEM DESIGN 2.1 Fuel Assembly Mechanical Design, The feed batch Batch 8, consists of 64 assemblies with uranium enrichment of 3.13 wt %.

Cycle 6 will consist of 64 Batch 8, 64 Batch 7, 48 Batch 6 and 1 Batch 1A assemblies. The Batch 7 and 8 FA's are the Mk-B5 design; while the other batches are the Mk-B4 design. The Mk-B5 FA's are identical to the Mk-84 with only a change to the upper end fitting design which eliminates retainers FA's (Table 4-1 Ref. 2)parison of the fuel design parameters for the various for BPRA holddown. Com revealed no mechanical differencu between the feed I

batch and previous batches.

l 2.2 Fuel Rod Design The cladding stress, strain, and collapse analysis methods used for the Cycle 6 fuel rod design are the same ones used and found acceptable for previous cycles.

The staff finds that no further review in these areas is necessary.

2.3 Fuel Thennal Design All fuel in the Cycle 6 core is thennally similar.

The fresh Batch 8 fuel inserted for Cycle 6 operation introduces no significant differences in fuel l

thennal performance relative to the fuel remaining in the core. The analyses i

perfonned for Cycle 6 demonstrate that 20.5 kw/ft is a conservative limit to l

preclude centerline fuel melt (CFM) for all fuel batches. The maximum fuel i

rod burnup at End of Cycle (EOC) 6 is predicted to be less than 38,300 mwd /mtU.

l Fuel rod internal pressure for the highest burnup fuel rod is predicted to be i

less than the reactor coolant pressure of 2200 psis at core outlet. All fuel thennal design analyses have been performed using TACO 2 which has been reviewed and approved previously by the staff (Ref.3) and is, therefore, acceptable.

2.4 Gray APSR Design The gray APSR design was analyzed for cladding stress due to pressure, temperature, and ovality.

It was found that the gray.APSR had sufficient claddira and weld stress margins. The gray APSR was also analyzed for cladding strain due to thenral and irradiation swelling. The results of B&W analysis showed that no cladding strain is induced due to thenral expansion or irradiation swelling of the Inconel absorber.

The staff has reviewed and accepted previously (Ref. 4) the mechanical design of the gray APSR's. This design has been used in several B&W reactors.

3.0 EVALUATION OF THE NUCLEAR DESIGN Toledo Edison Company has provided a comparison of the core physics parameters for Cycles 5 end 6.

The core design changes listed below explain the differences in the physics parameters.

4 3-1) the increase in cycle lifetime to 405 effective full power days (EFPD) 2 an increase in the BPRA concentration 3

the variation in the loading pattern between Cycles 5 and 6 4

the second transition cycle to the lumped burnable poison, low leaking fuel cycle design 5

the revised control rod groupings 6

the removal of the regenerative neutron powers 7

the change from black to gray APSRs.

The parameters for Cycle 5 were generated using POW 7, while the parameters for Cycle 6 were generated using the NOODLE code (Ref 5). The two codes give comparable results when compared to measured data. The NOODLE code has been reviewed and approved by the NRC staff. Several other B&W reactors have used the N000LE code.

The staff concludes Toledo Edison Compcny's predicted neutronic parameters are acceptable because they were obtained using approved trethods, the validity of which has been demonstrated through many cycles of predictions, including startup tests, for this reactor and others. As a result of this review of the neutronic parameters compared to previous cycles, the staff concurs with Toledo Edison Company's conclusions regarding th( Cycle 6 transient and accident analyses.

4.0 EVALUATION OF THE THERMAL-HYDRAULIC DESIGN

'The Cycle 6 design analysis is the first application of crossflow methodology i

for the Davis-Besse Station.

The use of crossflow codes, which can predict the flow redistribution effects in an open lattice reactor core, provides significant departure from nucleate boiling ratios (DNBR) margin improvements relative to the traditional closed-channel codes. The reactor coolant flow, bypass flow, and design axial flux shape were revised for the Cycle 6 analysis.

The bypass flow for Cycle 6 decreased to 8.1% from 10.17. for Cycle 5. The themal-hydraulic analyses used a conservative value of 8.6% for bypass flow.

The LYNX series of codes have been reviewed and approved by the NRC s* :ff, and severs) other B&W reactors have been using this methodology. Based on the similarity with Cycle 5 and the use of approved models and methods, the staff concludes that the thermal-hydraulic design of Cycle 6 is acceptable.

5.0 EVALUATION OF TRANSIENT ANC ACCIDENT ANALYSES Toledo Edison Company has examined each USAR accident analysis with respect to changes in the Cycle 6 parameters to determine the effects of the Cycle 6 reload and to ensure that themal perfomance during hypothetical transients is not degraded. The effects of fuel densification on USAR accident results have been evaluated and reported in Reference 6.

. The radiological dose consequences of the USAR Chapter 15 accidents have been evaluated using conservative radionuclide source tems that bound the cycle-specific source tem for Cycle 6.

The dose calculations were performed consistent with the assumptions described in the USAR but used the more conservative source tems (which bound future reload cycles). The results of the dose evaluations showed that offsite radiological doses for each accident were below the respective acceptance criteria values in the current NRC Standard Review Plan (NUREG-0800).

The key paaaneters that have the greatest effect on detemining the outcome of a transient tjpically can be classified in three major areas:

(1) core thermal, (2) therml-hydraulic, and (3) kinetics paraceters including the reactivity feedback coefficients and control rod vorths. A generic loss-of-coolant accident (LOCA) analysis for B&W 177-FA raised loop nuclear steam systems (NSS's) has been perfomed using the Final Acceptance Criteria ECCS Evaluation Model. The bounding values for the allowable LOCA maximum linear heat rates (LHR's) were reviewed for Cycle 6.

The values of the maximum LHRs at the 4, 6 and 10 foot elevations for Cycle 6 were calculated using methods acceptable on an interim basis. Only a 6 foot elevation calculation was performed, specifically, for Cycle 6 using the approved (Ref. 7) evaluation model. The staff believes that Toledo Edison Company's submittal provides reasonable assurance that the perfomance criteria of 10 CFR Part 50.46 are satisfied. However, since the new evaluation model is now fully approved, it is the staff position that the Toledo Edison Company should provide Cycle 6 confirmatory calculations at the 4, 8 and 10 foot elevations using the approved mde'..

Toledo Edison Company has agreed to provide the results of these calculations within six mov 's af ter the start of Cycle 6.

6.0 RELOAD PHYSICS TESTS Toledo Edison Company has proposed to reduce the physics startup test program by eliminating the zero power ejected rod worth test, the zero power rods-in temperature coefficient measurerent and one of the intercediate core power distribution test plateaus. The NRC staff previously has reviewed and approved similar changes and finds this change acceptable.

7.0 TECHNICAL SPECIFICATIONS Many of the proposed TS changes are a result of change in power peaking and control rod worths for Cycle 6.

The TS's also have been revised to incorporate the changes due to the crossflow rethodology. The stcff finds acceptable these changes as identified below.

1.

Section 3.1.3.6, Figures 3.1-2a, 3.1-2b, 3.1-3a and 3.1-3b end deletion of 3.1-2c, 3.1-2d, 3.1-3c and 3.1-3d:

These changes and deletions are due to rod group insertion limits as a result of Cycle 6 analysis.

2.

Section 3.1.3.7 and Figure 3.1-4:

These changes are a result of changing rod groupings as part of the Cycle 6 analysis.

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Section 3.1.3.9, Figure 3.1-Sa, 3.1-5b, and 3.1-Sc, and deletion of 3.1-5d, 3.1-Se, 3.1-5f, and 3.1-5g:

These changes are a result of APSR grouping and insertion limits due to the Cycle 6 analysis.

4.

Section 3.2.1, Figures 3.2-1 and 3.2-2, and deletion of Figures 3.2-la, 3.2-Ib, 3.2-Ic, 3.2-1d, 3.2-2a, 3.2-2b, 3.2-2c, and 3.2-2d:

These changes in the axial power imbalance limits are a result of the Cycle 6 analysis.

5.

Sections 3.1.2.8, 4.1.2.8, 3.1.2.9, 4.1.2.9, 3.5.4. 4.5.4, Figure 3.1-1, Bases 3/4.1.2 and 3/4.5.4:

These changes are all related to a review of the boron capability requirement to provide a 1% shutdown margin when cooling from 200*F.

Also, to clarify, the wording is being changed from "contained" volume to "available" volume.

6.

Sections 3.2.4 and 3.2.5, Tables 3.2-1, 3.2-2, and 4.3-1:

These changes are a result of revising the power in. balance detector correlation (PIDC) for Cycle 6 and introducing the power dependent quadrant power tilt (QPT). The PIDC test calibrates the excore detectors to the incore detector measurements of core offset. With the revised method, the excore/incore calibration is tightened and there is less conservatism in the correlation. This change has been approved and used on other B&W reactors.

7.

Section 3.4.1.1, 4.4.1.1, Figures 2.1-1, 2.1-2, 2.2-1, Table 2.2-1, Bases 2.1.1, 2.1.2 and 2.2.1, and Bases Figure 2.1:

These changes pertain to operation with three reactor coolant numps (RCP's). As a result of using the crossflow methodology, the thermal power restriction for three pump operation is increased from 79.7% to 80.6%.

The wording changes clarify this section of the TS's.

3.

Section 5.3.2: These changes are related to the changed APSR design.

8.0

SUMMARY

The staff has reviewed the fuel system design, nuclear design, thennal-hydraulic design and the transient and accident analysis information presented in the Davis-Besse Unit 1 Cycle 6 reload submittal. The staff has concluded that the proposed reload and associatea modified TS's are acceptable. Toledo Edison Company has agreed to submit confirmatory calculations of the LOCA maximum LHR's for the 4, 8 and 10 foot core locations within 6 months.

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9.0 ENVIRONMENTAL CONSIDERATION

An Environmental Assessment and Finding of No Significant Impact has been issued for this amendment (53 FR 37661. September 27,1988).

10,0 CONCLUSION The staff has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the hecith and safety of the will not be endangered by operation in the proposed manner, and (2) public such activities will be conducted in compliance with the Commission's regulations, and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

Prinicipal Contributor:

M. Chatterton Dated:

October 3, 1988 I

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, REFERENCES 1.

D. C. Shelton (Toledo Edison) letter to NRC,

Subject:

Cycle 6 Reload ReportLicenseAmendmentRequest(TACNo.66730),datedMey 18, 1988.

2.

Davis-Besse Nuclear Power Station Unit 1, Cycle 6 - Reload Report, BAW-2038, dated April 1988, transmitted with Reference 1 above.

3.

Y. Hsii, et al., "TAC 02 - Fuel Pin Performance Analysis," Babcock &

Wilcox Company Report BAW-10141P-A, Rev. 1, dated June 1983.

4.

H. Silver (NRC) to W. W11gus, Florida Power Corporation, Amendment No. 77 and Supporting Safety Evaluation of the Crystal River Unit 3 Cycle 6 Reload, dated July 16, 1985.

5.

"N0ODLE - A Multi-Dimensional Tro-Group Reactor Simulator," Hays, C. W.,

et al., BAW-10152A, Babcock & Wilcox, dated June 1985.

6., Davis-Besse Unit 1 Fuel Densification Report, BAW-1401, dated April 1975.

7.

A. C. Thadani (NRC) to C. H. Turk (B&W Owners Group), Safety Evaluation Report on B&W-10104, Rev. 5, dated December 27, 1987.

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