ML20195H950
| ML20195H950 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 11/22/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20195H948 | List: |
| References | |
| NUDOCS 8812010127 | |
| Download: ML20195H950 (5) | |
Text
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UNITED 8TATES 8
NUCLEAR REGULATORY COMMISSION o
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- I wAsmoToN. D. C. 20066
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'i SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 3
RELATED TO AMENDMENT NO.154 TO FACILITY OPERATING LICENSE h0. OPR-49 IOWA ELECTRIC LIGHT AND POWER COMPANY 1
CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DUANE ARNOLD ENERGY CENTER 1
00CKET NO. 50-331 b*
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1.0 INTRODUCTION
By letter dated August 19,1988(Reference 1),IowaElectricLightandPower Company (IELP)(etal.,submittedanapplicationtoamendtheDuaneArnold Energy Center DAEC)TechnicalSpecifications(TS's). The changes were proposed to support the DAEC fuel reload and operation for Cycle 10, and to incorporate administrative changes reflecting revisions to figure numbers and references.
In support of these changes, IELP also submitted two General Electric Company reports; "Supplementary Reload Licensing Submittal For Duane Arnold Atomic Energy Center Unit 1 Reload 9. Cycle 10," (Reference 2) and "Duane Arnold Energy Center SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis.
E & A No. 1 " (Reference 3).
t The licensee proposes to change the TS's by updating the fuel thermal limits of Section 3.12 and revising the Minimum Critical Power Ratio (MCPR) Safety Limit in Section 1.1.A.
The analytical methods used in References 2 and 3 to l
support the requested changes have been generically approved by the NRC staff I
l and were used in the analysis of the previous reload for the DAEC.
I 2.0 EVALUATION i
2.1 Reload Description l
t l
l For Cycle 10,120 irradiated fuel assemblies will be removed from the reactor l
coro and replaced by 64 BD324B and 56 B0303A new GE8x8EB fuel assemblies. The i
248 fuel assemblies retained for Cycle 10 are identified in Reference 2.
The l
reload is based on a previous end-of-cycle core nominal average exposure of
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21,361 mwd /MT and Cycle 10 end-of-cycle assumed core average exposure of 20.921 l
1 mwd /MT. The core loading pattern will be a conventional scatter pattern with lower enrichment fuel on the periphery, as inoicated in Reference 2.
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2.2 Fwl Mechanical Design The new fuel to be used for Cycle 10 (GE8x8EB) is described in References 2 and 3, and is the same fuel type previously added for use in the DAEC Cycle 9 reload.
This fuel type is customarily used in BWR reloads and has been approved in the NRC Staff Safety Evaluation Report for Amendment 10 to GESTAR-II (References 4 and 5).
The specific fuel description for Cycle 10 is provided in Reference 3 and is acceptable.
The uechanical design methodology is described in Reference 6 and was used in this design for the GE8x8EB fuel.
Reference 6 has been approved by the staff in Reference 7 and its supplements.
The staff therefore concludes that the fuel mechanical design for the DAEC Cycle 10 fuel is acceptable.
- 2. 3 Nuclear Design The nuclear design and analysis of the Cycle 10 reload have been performed using methods and techniques described in Reference 6, which are used in all reload analyses performed by General Electric.
The results of the analyses for DAEC Cycle 10, presented in References 2 and 3, are within the range of those reload cores previously reviewed by the NRC staff and found to be acceptable.
The staff therefore concludes that the nuclear design and analysis of the DAEC Cycle 10 reload is acceptable.
2.4 Thermal-Hydraulic Design the methods and procedures employed in the thermal-hydraulic design and analysis of the Cycle 10 core are described h Reference 6.
The value of 1.04 for the MCPR Safety Limit, previously reviewed and approved on a generic basis by the NRC staff in Reference 8, is used for Cycle 10.
Because the Cycle 10 reload for the DAEC meets the necessary criteria (D-lattice plant, second successive reload core of GE8x8E8 fuel type with bundle R-factors all 2 1.04),
the staff's generic approval of the revised MCPR limit is applicable.
In addition, the methods and procedures used to obtain the operating limit MCPR are those described in Refersnce 6 and approved in Reference 7, and are acceptable.
- 2. 5 Loss-of-Coolant Accident Ana_ lyses The LOCA analyses were perfctmed using the SAFER /GESTR code and the application methodology described in Reference 9.
Reference 10 approved that methodology and specified the conditions necessary for demonstrating applicability of the methodology to plant-specific analyses.
References 3 and 11 adequately demonstrate that the necessary conditions are met, thur the accident analyses have been purformed using approved methods.
The results of the LOCA analyses meet the acceptance criteria of 10 CFR 50.46 and 6re therefere acceptable.
2.6 MCPR and MAPLHGR Limits A safety limit MCPR has been imposed to assure that 99.9 percent of the fuel rods in the core will not experience boiling transition during normal operation and anticipated operational transients.
The proposed TS change for Cycle 10 operation revises the MCPR safety limit to 1.04 (1.07 for single loop' operation).
To assure that the fuel cladding integrity safety limit MCPR will not be violated during any anticipated transient, the most limiting events were iuanalyzed for this reload (keference 2) to determine which events result in the largest reduction in critical power ratio (CPR).
The operating limit MCPR was then established by adding the largest reduction factor in the CPR to the safety limit MCPR.
Since acceptable methods described in Reference 6 have been used, the staff finds the MCPR TS changes to be acceptable.
As for the previous DAEC reload, the GE8x8EB fuel will be assigned a number of axial lattice regions and appropriate maximum average planar linear heat generation rate (MAPLHGR) limits.
The MAPLHGR limits have been determined by approved thermal-mechanical and loss-of-coolant accident (:.0CA) analyses calculations and will be applied to each of these regions.
The plant process computer contains, and acts on, full details of the MAPLHGR information.
The proposed TS's for the BD 324B fuel assemblies present the lowest (and most limiting) lattice MAPLHGR as a function of burnup. When hand calculations of MAPLHGR are required (process computer inactive), the most limiting values are used for 611 limits.
This method was approved by the staff for the OAEC Cple 9 reload (Reference 12), and is also acceptable for Cycle 10.
The MAPLHGR limits specified in the proposed TS changes are less than or equal-to the bounding MAPLHGR limits used in the SAFER /GESTR-LOCA analyses of References 3 and 11 and are, thm fore, acceptable.
2.7 Technical Specification Changes The TS changes proposed by the licensee reflect the new fuel for Cycle 10.
These changes include:
revising the MAPLHGR curve of Figure 3.12-9 to reflect the operating Ifmits for the new bundle type (BD 3248) being added to the core; revising the MCPR safety limit of Section 1.1.A to 1.04 for two recirculation loop operation (and to 1.07 for single loop operation); and revising the operating limit MCPR curve of Figure 3.12-3.
These proposed changes are acceptable, since they are based on approved analytical methods as discussed above.
The staff has also reviewed the administrative changes supportitig the Cycle 10 reload and finds them acceptable.
3.0 ENVIRONMENTAL CONSIDERATION
S This 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 or changes a surveillance requirement.
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 exposure.
The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public con.?ent on such finding.
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
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.
4.0 CONCLUSION
Based on the review discussed above, the staff concludes that the Ouane Arnold Energy Center may be loaded and operated for Cycle 10.
This conclusion is based on the following:
1.
The safety analyses have been performed by previously approved methods and procedures, 2.
The Cycle 10 core meets all of the staff's t.cceptance criteria.
The staff also concle:Jes that the associated charges to the TS's for Cycle 10 operation, includird revision of the MCPR safety limit, are acceptable.
The staff has a'so concluded, based on the considerations discussed above, that:
(1) ths.re is reasonable assurance that the health a'id safety of the public will.iot 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 this amendment will not be inimical to the comon defense and security or to the health and safety of the public.
Principal Contribut, ~
J. R. Hall Dited:
November 22, 1988
5.0 REFERENCES
1.
Letter, William C. Rothert (Iowa Electric Light and Power Company) to Thomas E. Murley (NRC), August 19, 1988 (NG-88-2298).
2.
Supplemental Reload Licensing Submittal for Duane Arnold Atomi'c Energy Center, Unit 1 Reload 9, Cycle 10, General Electric Company, 23A5906, June 1988.
I 3.
Duane Arn,1d Energy Center SAFER /GESTR-LOCA Loss-af-Coolant Accident Analysis, NEOC-31310P, E & A No. 1, General Electric Company, June 1988.
(Proprietary) 4.
Letter, Cecil 0. Thomas (NRC) to J.S. Charnley (GE), May 28, 1985, Amendment 10 to General Electric Standard Application for Reactor Fuel.
5.
Letter, H.N. Berkow (NRC) to J.S. Charnley (GE). December 3, 1985, Amendment 10 to General Electric Standard Application for Reactor Fuel (Extended Burnup Operation).
6.
Licensing Topical Report NEDE-24011-P-A-8, General Electric Standard Application for Reactor Fuel, (GESTAR-II) Revision 8 July 1986.
7.
Letter, D.G. Eisenhut (NRC) to R. Gridley (GE) dated May 12, 1978, and supplements thereto, Forming Appendix C to Reference 6.
8.
Letter, A. Thandani (NRC) to J.S. Charnley (GE) December 27, 1987, Amendment 14 to General Electric Standard Application for Reactor Fuel.
9.
NEDE-23785-1-PA, The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Vols I, II and III, General Electric Company, June 1984.
10.
Letter, Cecil 0. Thomas (NRC) to J.F. Quirk (GE), Acceptance for Referencing of Licensing Topical Report NEDE-23785, Revision 1, The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident.
November 2, 1983, 11.
Duane Arnold Energy Center SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis. NEOC-31310P, August 1986, General Electric Company (Proprietary).
12.
Letter, A.J. Cappucci (NRC) to L. Liu (IELP) License Amendment No. 142 Cycle 9 Reload, May 7, 1987.
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