ML20217E320
| ML20217E320 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 09/22/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20217E317 | List: |
| References | |
| NUDOCS 9710060453 | |
| Download: ML20217E320 (5) | |
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 3088H001
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MFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION I
RELATED TO AMENDMENT NO.110 TO FACILITY OPERATING LICENSE NO. NPF-42 WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482 1
1.0 INTRODUCTION
By letter of July 3,1997. Wolf Creek Nuclear Operating Corporation (WCNOC),
the licensee, submitted a request for changes to the Wolf Creek Generating Station (WCGS) Technical Specifications (TSs) 5.3.1 and 6.1.9.6.
The requested changes would modify WCGS TSs 5.3.1, " Fuel Assemblies". and 6,1.9.6.
' Core Operating Limits Report (COLR)" to add ZIRLO as fuel cladding material, the use of limited zirconium alloy filler rods in place of fuel rods, and 1
allow the limited use of lead test assemblies in non limiting core positions.
The Westinghouse ZIRLO fuel was described in Topical Report WCAP-12610
" VANTAGE + Fuel Assembly Reference Core Report." and was approved by the NRC staff for irradiation up to 60,000 MWD /MTJ rod average burnup (Wolf Creek's current burnup limit),
Extensive testing has been conducted by Westinghouse through lead test assembly (LTA) programs and was selected as reload fuel by other utilities.
2.0 EVALUATION TS 5.3.1 requires fuel rods to be constructed with Zircaloy.
Zircaloy or stainless steel filler rods may be substituted in place of fuel rods and justified by cycle-specific reload analysis. The proposed amendment would modify TS 5.3.1 to allow fuel rods to be constructed with ZIRL0 and to allow fuel assembly reconstitution with Zirconium alloy, i.e.. Zircaloy or ZIRLO, filler rods and allow a limited number of LTAs in non limiting core positions.
The use of other zirconium alloys would require an exemption from 10 CFR 50.46 in that only Zircaloy and ZIRLO are identified in that regulation.
TS 6.1.9.6 lists the analytical methods, which have been reviewed and approved by the NRC, that are used to determine the core operating limits. The licensee 3roposes to add a reference to the NRC issued Safety Evaluation Reports tlat approved the use of Westinghouse report WCAP-12610-P-A, " VANTAGE +
Fuel Assembly Reference Core Report," to TS 6.1.9.6 to reflect the methodology used for the rod heatup calculation in the LOCA evaluation models with ZIRLO clad fuel, The staff approved the ZIRLO fuel design in a safety evaluation dated July 1, 1991, of Westinghouse Topical Report WCAP-12610. The NRC staff also approved loss of coolant accident (LOCA) methodologies in another safety evaluation. dated October 9,1991, of Westinghouse Topical Reports WCAP-12610, FM 188 u M88He2 P
2 Appendix F. "LOCA NOTRUMP Evaluation Model:
ZlRLO Modifications." and Appendix G.' "LOCA Plant-Specific Accident Evaluation." The July 1. 1991, safety evaluation concluded that:
a.
The mechanical design bases and limits for ZlRLO clad fuel assembly design are the same as those for the previously licensed Zircaloy-4 clad fuel assembly design, except those specified for clad corrosion.
b.
The neutronic evaluations have shown that ZlRLO clad fuel nuclear design bases are satisfied and that key safety parameter limits are applicable. The nuclear design models and methods accurately describe the behavior of ZlRLO clad fuel.
c.
The thermal and hydraulic design basis for ZlRLO clad fuel is unchanged.
d.
The methods and computer codes used in the analysis of the non LOCA licensing-basis events are valid for ZIRLO clad fuel, and all licensing-basis criteria will be met.
e.
The large-break LOCA evaluation model was modified to reflect the behavior of the ZIRLO clad material during a LOCA. Consequently, the revised evaluation model satisfies 10 CFR 50,46 and Appendix K of 10 CFR Part 50.
In the October 9, 1991, safety evaluation for WCAP-12610. Appendices F and G, the NRC concluded that the LOCA analyses and methods used demonstrated conformance with the criteria given in 10 CFR 50.46 and 10 CFR Part 50.
A]pendix K.
The safety evaluation stated that its conclusions were based upon t1e close similarity between the material properties of the ZIRLO alloy of zirconium to those of other zirconium materials that have been previously licensed for use as cladding material.
Based on this similarity, the NRC staff found that it is a)propriately conservative to apply the criteria of
-10 CFR 50.46 and 10 CFR 3 art 50. Appendix K. when reviewing VANTAGE + (ZlRLO) fuel a)plications, including WCAP-12610. Appendices F and G. The staff finds that tie cited findings from the July 1 and October 9, 1991, safety evaluations apply to the use of ZIRLO at WCGS.
The proposed change to allow the use of ZlRLO is intended to remedy the phenomenon of incomplete rod insertion, which has been experienced at the Wolf Creek Generating Station (WCGS).
In-vessel compressive loading and irradiation growth of the fuel assembly guide tubes have been determined to be the cause of incomplete insertion. The material of the guide tubes is being changed to ZlRLO for better dimensional stability and corrosion resistance, as 1
well as compatibility with the fuel assembly skeleton.
Changing to ZlRLO cladding will alst inhibit in-cme fuel rod corrosion, which studies have shown to be of concern relative to h19h burnup fuel and longer
-cycles.
l a t
The bounding analysis for large and small break LOCA rod heatup cases were i-evaluated by the licensee for the WCGS.
In all cases, the acceptance criteria were met, includin those in 10 CFR 50.46. and an adequate margin to the peak clad temperature 1 mit of 2200 'F is maintained The effect of ZIRL0 on non LOCA analyses were also evaluated by the licensee.
Two events, rod ejection and locked rotor, were determined to be potentially affected.
The results of the evaluations demonstrated that all acceptance criteria continue to be met.
The change from Zircaloy-4 to ZIRLO is consistent with 10 CFR 50.44 and 10 CFR 50.46 which contains standards and criteria for fuel clad with Zircaloy or ZIRLO.
The change is also consistent with NRC-approved topical report WCAP-13060 " Westinghouse Fuel Assembly Reconstitution Evaluation Methodology." which sets forth the methodology used to evaluate the applicable design criteria associated with filler rods and, meets the intent of Supplement 1 of Generic Letter 90-02. " Alternative Requirements for Assemblies in the Design Sections of Technical Specifications." NUREG-1431. " Standard Technical Specifications for Westinghouse Plants." specifically includes ZIRL0 as an acce3 table cladding material.
Thus, the NRC staff concludes that the a of ZIR.0 clad fuel at WCGS is acceptable.
Technical SDecification Chanaes Section 5.3 Reactor Core. Fuel _ Assemblies 5.3.1 Technical Specification 5.3.1 has been rewritten to agree with Section 4.2.1 of the Westinghouse Standard Technical Specifications (NUREG-1431. Rev.1) as follows:
"The reactor shall contain 193 fuel assemblies.
Each assembly shall consist of a matrix of Zircaloy or ZlRLO clad fuel rods with an initial composition of natural or slightly enriched urarium dioxide (U0 ) as fuel 2
material.
Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods in accordance with approved a)plications of fuel rod configurations. may be used.
Fuel assemblies slall be limited to those fuel designs that have been analyzed with applicable NRC approved codes and methods and shown by test or analyses to comply with all fuel safety-design bases.
A limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions."
The staff interprets the term " zirconium alloy" in the discussion of-the substitution of filler rods for fuel rods to include only ZIRLO and Zircaloy.
- As required by the TS. lead test assemblies that are inserted in the reactor core must have been analyzed using NRC approved codes and methods to show that all fuel safety design bases will be met. This 1s in agreement with the staff's guidance in Generic Letter 90-02. Supplement 1. " Alternate Requirements for Fuel Assemblies in the Design Features Section of Technical Specifications. " The staff finds this change acceptable.
4 On the basis of its evaluation of the acceptability of ZlRLO for WCGS, the NRC staff concludes that the proposed changes to WCGS TS 5.3.1 are acceptable.
Core Ooeratina limits Reoort. Section 6.9.1.6b The licensee also proposed to add references to the NRC issued Safety Evaluation Reports, dated July 1. 1991 and September 15, 1994, that approved the use of Westinghouse Topical Report WCAP-12610 for a LOCA evaluation model with ZIRLO clad fuel for rod heatup calculation (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor).
This proposed reference will provide that the analytical methods used to determine the core o)erating limits shall be reviewed and a) proved by the NRC staff.
The use of NRC-approved methodologies will ensure tlat values for-cycle specific parameters are determined such that applicable limits (e.g.,
fuel thermal-hydraulic limits, core thermal-hydraulic limits. ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
The added topical report appropriately models the core performance at WCGS.
With this addition, the analytical methodologies referenced in Section 6.1.9 of the Wolf Creek TS are appropriate for calculating the limits associated with ZlRLO clad fuel and reconstituted fuel using ZIRLO or stainless steel filler rods.
Therefore, the staff finds the modification to Section 6.1.9.6b acceptable for WCGS.
The NRC staff has reviewed the licensee's submittal regarding the use of ZIRLO clad fuel and the associated TS changes and, on the basis of its evaluation, the staff concludes that the aroposed changes to the TS and the COLR meet the related requirements of 10 CF1 50.46: 10 CFR 50. Appendix K: and 10 CFR 50.44, and are therefore acceptable.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations. the Kansas State Official was notified of the proposed issuance of the amendment.
The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC 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 exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (62 FR 40860).
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). The amendment also involves changes in recordkeeping. reporting or administrative procedures-or requirements.
Accordingly, with respect to these items. the amendment meets the eligibility criteria for categorical
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f' 5-exclusion-set forth in 10 CFR 51.22(c)(10).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or. environmental assessment need be prepared in connection with the. issuance of the amendment.
5,0 CONCLUSION The Commission has concluaed, based on tne considerations discussed above.
that: -(1) there is reasonable assurance that the. health and safet of the public will not be endangered by operation in the proposed manner,y(2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendnent will not be inimical to the common
. defense and security or to the health and safety of the public, Principal Contributor:
A, Attard Date:
September 22, 1997 1
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