ML20238D766
| ML20238D766 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 12/22/1987 |
| From: | Fay C WISCONSIN ELECTRIC POWER CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| CON-NRC-87-131 VPNPD-87-542, NUDOCS 8801040426 | |
| Download: ML20238D766 (6) | |
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WISCONSIN Electnc eaara coursur 231 W. MICHIGAN,P.O BOX 2046. MILWAUKEE.Wl 53201 (414)277-2345 VPNPD-87-542 NRC-87-131 December 22, 1987 U. S. NUCLEAR REGULATORY COMMISSION Document Control Desk Washington, D.C.
20S55 Gentlemen:
DOCKET NUMBERS 50-266 and 50-301 INCREASED ALLOWABLE CORE POWER PEAKING FACTORS POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 Wisconsin Electric has begun a program of accident reanalysis with Westinghouse Electric Corporation to support a license amendment which will increase the Point Beach Nuclear Plant (PBNP) Technical Specifications allowable core power peaking factors (F and FoH).
Higher peaking factors will allow the use g
of low-low leakage loading pattern (L4P) fuel management, which will result in decreased neutron fluence to the reactor vessel.
The higher peaking factors will also allow measures to be pursued which will result in additional reduction of fluence to the reactor vessel walls.
This fluence reduction will address the current reactor vessel issues dealing with pressurized thermal shock, low upper shelf material toughness, and heatup and cooldown pressure-temperature restrictions, and will enhance our ability to extend the useful life of the reactor vessels at the Point Beach Nuclear Plant.
A license amendment to permit the adoption of higher allowable peaking factors requires the reanalysis of the design-basis accidents presented in Chapter 14 of the PBNP Final Safety Analysis Report (FSAR).
This accident reanalysis falls into three categories: large-break Loss of Coolant Accident (LOCA),
small-break LOCA, and non-LOCA transients.
The large-break LOCA analysis is being conducted jointly with Northern States Power Company as part of resolution of the Upper Plenum Injection (UPI) issue using the Westinghouse best-estimate W COBRA / TRAC model.
The anticipated schedule for this work is presented in the attached figure.
Volume 2 of WCAP-10924,
" Application to Two-Loop PWRs Equipped with UPI" with the Prairie 8801040426 AD BH222 PDR pygCK 05000266 I
P Ag C ttMX 70 M
A NRC Document Control Desk December 22, 1987 Page 2 Island " Lead-Plant" analysis, was submitted to the Nuclear Regulatory Commission (NRC) in March 1987.
Responses are now being prepared to NRC questions on Volume 2.
Upon receipt of the NRC Safety Evaluation Report (SER) for the lead plant analysis which is expected in July 1988, the PBNP-specific analysis will be submitted.
The NRC SER for the PBNP-specific analysis is i
expected in October 1988.
The small-break LOCA' analysis of Point Beach will use the NOTRUMP code described in WCAP-10054-P-A, " Westinghouse Small Break LOCA ECCS Evaluation Model Using the NOTRUMP Code," submitted to the NRC on December 21, 1982.
The use of NOTRUMP for small-break LOCA analyses of Westinghouse reactors was accepted by the NRC in an SER dated May 21, 1985.
We believe we can justify the analysis of less than a three-break spectrum based on previous analysis done for PBNP and NOTRUMP generic studies described in WCAP-11145, " Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study with the NOTRUMP Code," submitted to the NRC on June 11, 1986.
We would like to discuss this matter with the NRC staff before February 1988 when a decision must be made regard-ing the break or break spectrum to be analyzed.
[
The final analysis category comprises the reanalysis /re-evalua-tion of all the non-LOCA transients presented in Chapter 14 of the PBNP FSAR.
The most recent' transient reanalysis for Point Beach was performed for the licensing of Westinghouse Optimized Fuel Assemblies (OFA), which was approved by the NRC on October 5, 1984 (Unit 2 License Amendment No. 90) and May 22, 1985 (Unit 1 License Amendment No. 86).
The only new methodologies to be used in the present reanalysis are the Westinghouse Revised Thermal Design Procedure (RTDP), WCAP-ll397, submitted by Westinghouse to the NRC on March 16, 1987, and the Westinghouse Owner's Group (WOG) Dropped Rod Generic Methodology, WCAP-ll394, submitted by the WOG on May 22, 1987.
It would be beneficial to both the NRC and WE if these generic methodologies could be reviewed and SERs issued by the NRC by February 1988, if possible, to support our reanalysis effort.
The non-LOCA and small-break LOCA analyses should be completed by June 30, 1988.
Following Wisconsin Electric's internal review, the license amendment request for higher allowable peaking i
factors is expected to be submitted to the NRC by August 1988.
NRC approval of this submittal is required by February 1989 to implement an L4P core design in spring 1989.
Uranium enrichment for the feed region for the spring 1989 L4P core is scheduled to be ordered in February 1988.
j t
NRC Document Control Desk
~
December 22, 1987 Page 3 Our proposed schedule for increased peaking factor reanalysis and approval is attached.
Please contact us should you have any questions or comments regarding this schedule.
Very truly yours,
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C.
W. Fay Vice President Nuclear Power Attachment Copy to NRC Resident Inspector
'NRC Regional Administrator, Region III j
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GANTT CHART INCREASED PEAKING FACTOR SCHEDUL.
REB '8 MAR '87 APR '87 MAY '87 JUN '87 JUL '87 AUG '87 SEP '87OCT '87 NOV '87 EC '87 JAN '88 I. LARGE BREAK LOCA ANALYSIS SUBMIT UPI WCAP VOL. 2 _
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SUBMIT UPI VOL. 2. CH. 7 RESUBMIT UPI VOL. 2. CH. 3.
NRC REVIEW AND GENERIC SER.
E IE PBNP - SPECIFIC ANALYSIS NRC REVIEW AND P8HP SER _
II.0THER WESTINGHOUSE ANALYSIS I
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R SUBMIT OROPPED ROD WCAP DROPPED ROD. ROD EJECTION SMALL BREAK LOCA ANALYSIS
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_l OT}ER TRANSIENTS WE INTERNAL REVIEW NRC REVIEW AND SER III. L4P IMPLEMENTATION ORDER U1C17 ENRICHHElT __
LICENSE AMEN 0 MENT REQ' JEST ____
FINAL UIC17 CORE DESIGft __
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