ML20138B805
| ML20138B805 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 03/20/1986 |
| From: | Fay C WISCONSIN ELECTRIC POWER CO. |
| To: | Harold Denton, Lear G Office of Nuclear Reactor Regulation |
| References | |
| CON-NRC-86-025, CON-NRC-86-25 VPNPD-86-133, NUDOCS 8603250170 | |
| Download: ML20138B805 (3) | |
Text
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4 Wisconsin Electnc eaara coursur 231 W. MICHIGAN, P.O. BOX 2046. MILWAUKEE. WI 53201 VPNPD-86-133 NRC-86-025 March 20, 1986 Mr.
H.
R. Denton, Director Office of Nuclear Reactor Regulation U.
S. NUCLEAR REGULATORY COMMISSION Washington, D.
C.
20555 Attention:
Mr.
G.
Lear, Project Director PWR Project Directorate No. 1 Gentlemen:
DOCKET NO. 50-266 CYCLE 14 RELOAD POINT BEACH NUCLEAR PLANT, UNIT 1 A refueling shutdown for the Point Beach Nuclear Plant Unit 1 is presently planned to begin in early April 1986.
This shutdown will be at the end of Unit 1 Cycle 13 operation.
Cycle 13 burnup will be approximately 8,850 mwd /T.
Point Beach Unit 1 startup for Cycle 14 is expected to occur in May 1986 following a nominal seven-week refueling and maintenance outage.
The reload region for Unit 1 Cycle 14 operation will contain 28 Westinghouse 14x14 Optimized Fuel Assemblies (OFA).
This will be the second reload region of OFA fuel inserted into a Unit 1 core.
The use of OFA fuel in both Point Beach Nuclear Plant units was revieweQ and approved as reported in the NRC Safety Evaluation Report issued on October 5, 1984 in support of License Amendments 86 and 90 for Units 1 and 2, respectively.
8 Fuel Region IC, the Point Beach Unit 1 Cycle 14 reload, contains f
a Wesitinghousi design improvement that replaces the standard fuel
^
pellet design'with an improved chamfered pellet design.
This design improvement consists of a small edge chamfer and a reduced dish diameter depth.
These changes conserve the pellet land area with a slight increase in pellet weight.
The benefits of the chamfered pellet include improved pellet chip resistance uring N6 8603250170 860320 6
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W Mr. H. R. Denton, Director j
March 20, 1986 Page 2 manufacturing and handling which results in a greatly reduced potential for pellet chipping and stress-induced type failures during i
reactor operation.
This design change meets all safety and design limits, results in no Technical Specification changes, and no unreviewad safety questions as defined by 10 CFR 50.59.
4 The Cycle 14 core will also contain four 12-rodlet " water displacer" assemblies.
The " water displacer" rod assemblies are essentially the same as the burnable absorber tubes (BPRA's),
which contain borosilicote glass as a burnable absorber, except that the " water displacer" tubes contain no absorber and are pressurized with helium.
Engineering review of the " water displacer" design has verified that even if leakage occurs no design or safety concerns are posed.
The mechanical and thermal-hydraulic design for the Unit 1 Cycle 14 reload core is similar to that of previously reviewed i
and accepted reload designs containing OFA fuel.
This core is j
designed to operate under nominal design parameters and the approved Technical Specifications, including those provided with License Amendment 86 for Unit 1, so that the core characteristics will be less limiting than previously reviewed and accepted.
For those postulated accidents presented in the FSAR which could be affected by the reload core, reevaluation has demonstrated that the results of the postulated events are within allowable limits.
ThereloadcoremeetsF]xPlimitoflessthan2.21andthecurrent FAQ limits of less than 1.58.
j In accordance with past practice, the Westinghouse reload safety evaluation report relies on previously reviewed and accepted analyses as reported in the FSAR, the OFA transition safety reports, and in earlier reload cycle safety evaluation reports.
The reload safety evaluation report demonstrates that no unreviewed 3
safety questions, as defined in 10 CFR 50.59, are involved in operation of the reactor during Cycle 14; therefore, no application for license amendments, beyond those already approved by the NRC in License Anendment No. 90, are required for Cycle 14 operation.
This 10 CFR 50.59 evaluation will be confirmed by the Manager's Supervisory Staff prior to startup of Cycle 14.
Verification of the core design will be performed by means of the standard startup physics tests normally performed at the beginning of each cycle.
i l
Mr.
H.
R. Denton, Director March 20, 1986 Page 3 Please contact us if you have any question regarding the Cycle 14 reload design or operation.
Very truly yours, i
{' u C. W. Fay' Vice President Nuclear Power Copy to NRC Resident Inspector