ML20213A718
| ML20213A718 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 04/10/1987 |
| From: | Diianni D Office of Nuclear Reactor Regulation |
| To: | Musolf D NORTHERN STATES POWER CO. |
| References | |
| TAC-64933, TAC-64934, NUDOCS 8704280224 | |
| Download: ML20213A718 (6) | |
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APR 101987 -
50-282' Docket Nos.:
and 50-306 Mr. D. M. Musolf, Manager.
Nuclear Support Services Department Northern States Power Company 1
414 Nicollet Mall - 8th Floor Minneapolis, Minnesota 55401
Dear Mr. Musolf:
J
SUBJECT:
CHANGES TO THE 1981 LOCA MODEL FOR OPERATION OF THE PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NO. 1 CYCLE 12 The Office of Nuclear Reactor Regulations (NRR) has completed the review of i
the changes to the 1981 model of the Loss-of-Coolant Accident (LOCA) analysis, described in your letters dated March 3, and April 8,1987. This analysis was-performed in support of the Prairie Island Unit No.1 Cycle 12 fuel reload.
The staff has reviewed only the modifications to the model since the 1981 evaluation model has been. approved by the staff for plants with upper plenum s
injection.
l l
The staff concludes that the changes to the 1981 model of LC0A analysis are acceptable for the Prairie Island Unit No. 1 Cycle 12 and the core will be adequately protected in the event of a LOCA when the reactor is operated within the maximum peaking factor of 2.3 allowed by the technical i
specifications. This action completes our efforts under TAC Nos. 64933 and l
64934.
i A copy of our safety evaluation addressing this matter is enclosed.
/J/
3 Dominic Dilanni, Project Manager i
PWR Project Directorate #1 Division of PWR Licensing-1 1
Enclosure:
i As stated cc: w/ enclosures i
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o Mr. D. M. Musolf Prairie Island Nuclear Generating Northern States Power Company Plant cc:
Gerald Charnoff, Esq.
Shaw, Pittman, Potts and Trowbridge 2300 N. Street, N.W.
Washington, DC 20037 Dr. J. W. Ferman Minnesota Pollution Control Agency 520 Lafayette Road St. Paul, MN 55155 Mr. E. L. Watzl, Plant Manager Prairie Island Nuclear Generating Plant Northern States Power Company Route 2 Welch, Minnesota 55089 Jocelyn F. Olson, Esq.
.Special Assistant Attorney General g.,,
Minnesota Pollution Control Agency 1935 W. County Road, B2 Roseville, Minnesota 55113 U.S. Nuclear Regulatory Commission Resident Inspector's Office 1719 Wakonade Drive East Welch, Minnesota 55089 Regional Administrator, Region III U.S. Nuclear Regulatory Commission Office of Executive Director for Operations 799 Roosevelt Road Glen Ellyn, Illinois 60137 Mr. William Miller, Auditor Goodhue County Courthouse Red Wing, Minnesota 55066 i
e Enclosure i
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO CHANGE TO 1981 LOCA MODEL TO FACILITY OPERATING LICENSE NOS. DPR-42 and DPR-60 NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-282 AND 50-306 Introduction Xs described in letters dated March 3,1987 and April 8,1987 the licensee revised the large break LOCA analysis for Prairie Island Unit I and in support of Cycle 12 which is scheduled to begin May 12, 1987. The calculations were done using a modified version of the Westinghouse 1981 ECCS Evaluation Model for plants with Upper Plenum Safety Injection.
Since the 1981 Evaluation Model has already been reviewed and approved by the staff for plants with upper plenum injection, this report only evaluates the modifications.
Discussion and Evaluation The modifications are in the input to the WREFLOOD code which calculates the inlet core flows and enthalpies during the period after a LOCA when emergency coolant is reentering the core. The core inlet conditions calculated by WREFLOOD are then input into the LOCTA-IV code which calculates the local heatup of the fuel and the cladding.
The first modification was required because the water volume which flows into the control rod gpide thimbles during core reflooding was previously neglected. Allowing the guide tubes to fill produces a small increase in peak cladding temperature (6-12*F) for plants analyzed with the 1981 version of the Westinghouse ECCS evaluation model.
. The second modification involved a more realistic calculation of the heat flow to the coolant in the downcomer and lower plenum of the WREFLOOD code and compensates for the increase in peak cladding temperature produced by consideration of control rod guide thimble filling. The original version of WREFLOOD contained a simplified heat transfer model which resulted in overly conservative heating of the reflooding water before it entered the core. This version is utilized in the 1981 Evaluation Model. A more recent version of the WREFLOOD code which has also been approved by the staff utilizes more detailed and realistic modeling of downcomer and lower plenum heat transfer. The newer WREFLOOD calculates less heating of the reflooding water entering the core.
Rather than replace the original version of WREFLOOD in the 1981 Evaluation Model with the newer WREFLOOD, Westinghouse chose to modify the input to the original version to produce approximately the same heat transfer to the water entering the core as the newer version.
This was done using sensitivity studies so that for Prairie Island the original WREFLOOD with the modified input calculates slightly more heat transfer to the core inlet water than the newer WREFLOOD, which is conservative. The staff concludes that the modifications to the Westinghouse 1981 Evaluation Model are acceptable for the ECCS analysis for Prairie Island Unit 1.
Cycle 12 contains mixed fuel of which 2/3 is Westinghouse Optimized and 1/3 is Exxon TOPR00. The staff questioned the assumption that the hot channel would be in the Westinghouse fuel. The licensee provided a core loading diagramandpeakingfactoranalysiswhichshowedthat,ttypeakpowerofthe Westinghouse fuel will be at least 25% higher than thatifor the Exxon fuel.
The Exxon fuel has been through two core cycles and has been depleted of fissionable material and contains more neutron absorbing fission products relative to the Westinghouse fuel.
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1 3-l Conclusion The peak cladding temperature was calculated to be 2197*F which is acceptable.
since it falls below the limit of 2200*F specified in 10 CFR 50.46. The calculated peak cladding temperature includes adjustments to apply additional conservatism to account for upper plenum injection (54*F) and to account for a
.i core with mixed Westinghouse and Exxon fuel (10*F).
l The staff concludes that the Prairie Island Cycle 12 core will be adequately protected in the event of a LOCA when operated with the maximum peaking factor of 2.3 allowed by the Technical Specifications.
e-Principal Contributor:
W. L. Jenson i
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