ML20058E040

Safety Evaluation Supporting Amend 137 to License DPR-51

ML20058E040
Person / Time
Site:	Arkansas Nuclear
Issue date:	10/29/1990
From:	Office of Nuclear Reactor Regulation
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ML20058E037	List: ML20058E034 ML20058E040
References
NUDOCS 9011060411
Download: ML20058E040 (6)

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\\.....,h SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.137 i

TO FACILITY OPERATING LICENSE NO. DPR-51 ENTEPGY OPERATIONS. INC.

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ARKAllSAS NUCLEAR ONE. UNIT NO.1 t

DOCKET NO. 50-313

1.0 INTRODUCTION

By letter dated August 8,1990, Entergy Operations, Inc., the licensee, sub-mitted proposed Technical Specification (TS) changes for Cycle 10 operation of Arkansas Nuclear One, Unit 1 (ANO-1). A Cycle 10 Reload Report, BAW-2114, I

dated June 1990, was also submitted to support Cycle 10 operation and TS l

changes, j

L 2.0 EVALUATION 4

l 2.1 Fuel Design Cycle 10 will contain 1 Batch 6E (Mark-B4), 60 Batch 10B (Mark-84)l assemblies 60 Batch 11 (Mark-B6),and56 Batch 12(Mark-B8) assemblies. All of these fue are mechanically interchangeable. The Mark-B8 fuel incorporates slightly longer fuel rods and a shorter lower end fitting. The longer fuel rods have increased plenum volume allowing for' higher fuel burnup. The lower end plug solid portion was also extended in length and the lower spacer grid location i

was dropped causing the solid end plug to extend through~ the lower spacer grid.

The intention of this change is to trap any debris capable of fuel rod fretting below the bottom spacer grid where the solid lower end plug would prevent failure.

Because of the previous incore exposure of Batch 10B fuel, it is the most-I limiting in terms of cladding creep collapse. The licensee has stated that the I

cladding collapse time for the nest limiting Cycle 10 assembly was conserva-l.

tively determined to be greater than the reximum projected residence time for any Cycle 10 assembly. The methods and procedures used for the analyses have been previously reviewed and approved by the staff. The staff concludes that cladding collapse has been appropriate 1y' considered and will not occur for Cycle 10 operation.

l The cladding stress and strain analyses for the Cycle 10 fuel designs were calculated using methods and limits previously reviewed and approved by the NRC. The staff concludes that the analysis of cladding stress and strain has been appropriately considered for Cycle 10 operation and is acceptable.

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4 The thermal behavior of all fuel in the Cycle 10 core is virtually identical.

The thermal analysis was performed with the approved TACO 2 code and the Cycle 10 core protection limits were based on the calculated linear heat rate (LHR) to centerline fuel melt limit of 20.5 kw/ft. This limiting value is satis-factor 11y incorporated into the TS for Cycle 10 through the operating limits on rod index and axial power imbalance.

StandardReviewplan(SRp)4.2,Section11.A.1(f),containstherequirement that the fuel rod internal gas pressure should remain below nors.a1 system pressure during normal operation unless otherwise justified. Based on TACO 2 analyses, the licensee has stated that the internal pressure in the highest burnup rod of each fuel type will not reach the nominal reactor coolant system (RCS)pressureof2200 psia. The staff finds _this acceptable and concludes that the fuel. rod. Internal pressure limits have been adequately considered for Cycle 10 operation.

Based on its review, the staff concludes that approved methods have been used, that the fuel design parameters meet applicable criteria and that the fuel design for ANO-1 Cycle 10 is acceptable.

2.2 Nuclear Design The nuclear design parameters characterizing the ANO-1 Cycle 10 core have been computed by methods previously used and approved for Babcock and Wilcox (B&W) i reactors.

Comparisons between the parameters for Cycle 9 and Cycle 10 have shown little change.

Shutdown margin calculations = for Cycle 10 include the effects of poison material depletion,.a 10% calculational uncertainty, allow-ance for rod bite, and neutron flux redistribution as well as a maximum worth stuck rod.- Beginning-of-cycle (BOC) and end-of-cycle (EOC) shutdown margins show that adequate.. reactivity worth exists above the total recuired worth L

during the cycle. Shutdown margins at BOC and EOC are 4.02% celta k/k and 1

3.34f delta k/k, respectively, compared to-the minimum required value of 1.0f delta k/k.

Based-on its review, the staff concludes that approved methods have been used.

- that the nuclear design parameters meet applicable criteria, and that the nucleer. design of ANO-1 Cycle 10 is acceptable, j

2.3 Thermal-Hydraulic Design Cycle lo is' the second cycle in which a mixture of Mark B and Mark BZ fuel assen611es coexist in the core. Although a full Mark BZ core and a full Mark B H

core provide practically the same departure from nucleate boiling (DNB) margin for both' steady-state and transient conditions, incompatibility in the hydraulic characteristics has an effect on thermal' margin during transitional mixed core cycles when both Mark BZ and Mark B fuel assemblies coexist in the core.: Since the Mark BZ assemblies have a higher hydraulic resistance due to the burnable poison rod assembly (BPRA) retainers and the Zircaloy intermediate spacer grids, some of the coolant flow is diverted from the Mark BZ fuel to the lower power Mark B fuel.

The fact that the Mark BZ assemblies have less flow

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3 in a mixed core results in lower maximum allowable power peaking and a lower enthalpy rise factor required in order to maintain the same DNBR limit compared to a whole core of Mark BZ fuel. A cycle-specific analysis which modeled the

. actual Cycle 10 core configuration and bypass flow value has shown that the Cycle 9 thermal-hydraulic reference analysis remains applicable for Cycle 10 i

operation.

2.4 Accident and Transient Analysis The important-physics, thermal-hydraulic, and Linetics parameters for Cycle 10 have been compared to the values used in the srevious cycle, the FSAR, and/or

.the fuel densification report. The licensee las shown that the Cycle 10 values are bounded by those previously used and, therefore, the transient and accident l

evaluation of Cycle 10 are considered to be bounded by previously accepted l

analyses.

Babcock and Wilcox (B&W) has performed a generic loss-of-coolant accident (LOCA) analysis for the B&W 177FA lowered-loop nuclear steam supply system (NSSS) using the final acceptance criteria emergency core cooling system (ECCS) evaluation model, updated with an upgraded fuel performance model and the B&W modified version of FLECSET. The combination of average fuel temperature as a function of LHR and the lifetime pin pressure data used is conservative relative to those calculated for this cycle. Two sets of bounding values for allowable LOCA peak LHRs for Cycle 10 are given as a function of core height.

These limits apply during the periods from 0 to 1000 MWD /MTU and for the balance of the cycle. The B&W analysis has been approved by the NRC and the LHR limits are satisfactorily incorpcrated into the Technical Specifications for Cycle.10.

Themostnegativemoderatortemperaturecoefficient(MTC)occursatEOCand.is based on the steam line break accident.

Because'of longer fuel cycles, some B&W plants are approaching their predicted EOC limit on MTC.

Evaluations made by B&W have revealed a small nonconservative (positive) bias in the MTC

-calculational methodology which has not been incorporated into the Cycle 10 reload analyses. To account for this in Cycle 9. Entergy Operations has administrative 1y limited operations to a minimum soluble boron concentration sufficient to conservatively ensure that MTC will not exceed the limiting value used in the safety analyses. However, the steam line break event is being reanalyzed for Cycle 10 using the latest approved methodology. The results of similar analyses performed for other plants using the newer methodology have shown a significant relaxation in the EOC MTC limit, compared to the FSAR i

results, which more than compensated for the MTC calculational bias mentioned above.- The results of this reanalysis and its effect on the EOC MTC limits will be submitted to the NRC well before the Cycle 10 EOC MTC limit is

-approached.

Based on this, initial Cycle 10 operation with the current negative MTC limit is acceptable.

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4 3.0 TECHNICAL SPECIFICATION CHANGES ANO-1 Technical Specifications have been modified for Cycle 10 operation to reflect changes in core reactivity, power peaking, and control rod worths. The specific changes are evaluated below.

(1). The Core Protection Safety Limits specified in Figure 2.1-2 are modified.

The ch6nges reflect the Cycle 10 core design which has been reviewed and approved by the staff in this safety evaluation report.

(2) The Protective System Maximum Allowable Setpoints specified in Figure 2.3-2 are modified.

The changes reflect the Cycle 10 core design which hr.s been reviewed and approved by the staff in this safety evaluation report.

(3) TheModeratorTemperatureCogfficient(HTC)specifiedinTS 3.1.7 would be increased from +0.5 X 10' delta k/k/*F to *0.9 X 10'4 delta k/k/*F at p(ower levels less than or equal to 95 percent of rated thermal power RTP). The MTC would still be required to be nonpositive at power levels above 95 percent of RTP.

The most limiting accident adversely affected by the positive MTC is the startup accident in which an uncontrolled addition of reactivity is caused by a rapid withdrawal of control rods from a subcritical or low power ccndition. A positive MTC would yield the maximum peak heag flux. The accident analyses presented in the FSAR included 40.9 X 10' delta k/k/*F in a sensitivity study which examined the effects of variations in the MTC -

for all pertinent accidents including the startup accident. The resulting peak thermal power and reactor coolant system peak pressure remained well below the ac:eptance criteria.

In addition, no adverse im)act is expected on the hydraulic and neutronic stability of ANO-1 due to tie larger positive MTC.

The hydraulic and neutronic stability of ANO-1 is not expected to be different from the other B&W reactors such as Crystal River

3. Oconee 1, 2, and 3. Rancho Seco, Davis-Besse, and Thre whicharealreadylicensedforanMTClimitof40.9X10'gMileIsland1, delta k/k/*F from 0 to 95 perce;.s of-RTP. The staff concludes that the proposed increase-in the MTC limit for power levels'less than 95 )ercent of rated power has been adequately considered and is bounded by tie existing FSAR safety analyses and is,-therefore, acceptable.

-(4) TS Section 3.5.2.4 would be modified to reflect a quadrant power tilt limit of 4.241.

The quadrant power tilt revision is based on incore detector sensitivity depletion.

The tilt specified is the bounding value, derived from the detector sensitivity depletion at EOC 10. of the full incore quadrant tilt setpoint values determined for Cycle 10. The change is, therefore, acceptable.

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a The staff notes, however, that there are no required completion times specified for the power reduction if the quadrant tilt exceeds 4.24% (T5 3.5.2.4.1) or for the shutdown requirament if the quadrant tilt exceeds 25%(TS3.5.2.4.3).

We recomend that 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, respectively, be inserted to conform to present accepted practice.

(5) TSSection3.5.2.5.4andassociatedFigures3.5.2-1(A-C),3.5.2-2(A-C),

and 3.5.2-3 (A-C) would be modified to accomodate Cycle 10 changes to the control rod insertion limits and operational limits on the gray axial powershapingrods(APSRs).

The revised rod insertion limits provide assurance of achieving hot shutJown by reactor trip at any time during Cycle 10 and ensure that power praking criteria are not exceeded.

These limits also preclude insertion of rod groups which could result in any single rod worth greater than the 7,afety analysis assumption for the rod ejection accident. The gray APSR position limits have been derived using worst case conditions and time of core life such that core power avakina limits are not violated. These revised limits were derived witi E approved methods and computer codes and are, therefore, acceptaMar.

(6) TSFigures3.5.2-4(A-C)powerimbalancelimitswouldbemodified.

The power imbalance envelope defined in the revised Figures is based on LOCAanalyseswhichhavedefinedthemaximumlinearheatrate(Figure 3.5.2-5) to assure that the maximum cladding temperature will not exceed the final acceptance criteria in 10 CFR 50, Appendix K, assuming worst case power distributions. These revised setpoints have been derived using~

NRC approved methodology and computer codes, taking into account all perceived uncertainties, worst case conditions and core burnup. The proposed changas are, therefore, acceptable.

(7) TS Figure 3.5.2-5'would be modified to provide provide new loss-of-coolant accident (LOCA) linear heat rate (LHR) limits for Cycle 10.

The revised LHRs are based on a worst case range of power distributions, power shapes and ?ower peaking factors within the core during Cycle 10 which result in tie most severe calculated consequences for the spectrum of postulated accidents including LOCA. These revised LHRs continue to provide assurance th6t the fuel rod cladding temperature remains below the final-acceptance criteria of 10.CFR 50, Appendix K.

The revised values have been derived using NRC approved ECCS evaluation codes and methodology and are, therefore. acceptable.

4.0

SUMMARY

The NRC sr.aff has reviewed the proposed ANO-1 TS changes submitted by letter dated August 8,1990, from Entergy Operations, Inc., to support operation for a 10th cycle and finds them acceptable.

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The results of a reanalysis of the steam line break accident will be submitted i

to the NRC early in Cycle 10 to assure that the EOC MTC limit is adequate and will not be violated during Cycle 10 operation.

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5.0 ENVIRONMENTAL CONSIDERATION

The amendment involves a change in a requirement with respect to the installa-tion er use of a facility component 1ccated within the restricted area as defined in 10 CFR Part 20 and changes in surveillance requirements.

The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposures. The Commission has previously issued a proposed finding that the amendment involves no significant hazards I

consideration and there has been no public comment on such finding. Accordingly.

l the amendment meets the eligibility criteria for categorical exclusion set forthin10CPRSection51.22(c)(9).

Pursuant to 10 CTR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

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6.0 CONCLUSION

The staff has concluded, based on the considerations discussed above, that.

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

security or to-the health and safety of the public.

l Dattd: October 29, 1990 Principal Contributor:

L. Kopp, SRXB i

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