ML20246N038

Forwards Mod to 890808 Application for Amends to Licenses DPR-24 & DPR-27,consisting of Tech Spec Change Request 134, Removing Unit 2 Capsule in Fall 1990 Refueling Outage & Capsule P in Fall 1996 & Modifying Page 15.3.1-8a

ML20246N038
Person / Time
Site:	Point Beach
Issue date:	08/31/1989
From:	Britt R WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20246N044	List: ML20246N038 ML20246N294
References
CON-NRC-89-104 NUDOCS 8909080023
Download: ML20246N038 (3)

Text

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Wisconsin

' Electnc POWER COMPANY 231 W Michigan. Po Box 2046. Milwaukee, WI 53201 (414)291-2345

' NRC 104 August- 31, 1989 10 CFR 50.59 10 CFR 50.90 U. S. NUCLEAR REGULATORY COMMISSION Document Control Desk Mail' Station ~Pl-137 Washington,-D.C. 20555 Gentlemen:

DOCKETS 50-266 AND 50-301 MODIFICATION TO TECHNICAL SPECIFICATION CHANGE REQUEST 134 REACTOR VESSEL SURVEILLANCE CAPSULE REMOVAL SCHEDULE POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 By letter dated August 8, 1989, Wisconsin Electric Power Company (Licensee) submitted Technical Specification Change Request 134.

That request proposed that Unit.2 reactor vessel surveillance Capsule S be withdrawn in Fall 1995 and Capsule P be withdrawn in Fall 1998. It also proposed a change to Technical Specification page 15.3.1-8a to acknowledge our participation in the Babcock and <

Wilcox Master-Integrated Reactor Vessel Surveillance Program and briefly describe our current capsule withdrawal schedule.

That schedule was based on the requirements of 10 CFR 50, Appendix H, and ASTM E185-82. ASTM E185-82 states "...the withdrawal schedule of the final two capsules is adjusted by the lead factor so the exposure of tne second to last capsule does not exceed the peak end-of-life (EOL) fluence on the inside surface of the vessel, and so~the exposure of the final capsule does not exceed twice the.EOL vessel inside surface peak fluence." At the time of our August 8 submittal we believed the Unit 2 estimated maximum EOL fluence at the insidggsurfage of the reactor vessel to be approximately 3.5x10 n/cm . Capsule S would see that fluence about Fall 1995 and Capsule P would see 150% of that fluence in 1998.

We subsequently have learned of calculations done by Westinghouse,

.which'show that implementation of super low leakage cores (L4P) with hafnium absorbers beginning with the Unit 2 Fall 1989 Outage, will result in a significant reduction in EOL fluence not only in the vessel weld region but also in the remainder of the beltline vessel bh .6 [ 00I i i  !

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U.S. NRC Document Control Desk August 31, 1989 Page 2 plate / forging regions. Accordingly, we now estimate that maximumUnit2EOLfastneutronexposurewillbe2.92x10}gen/cm2 .

Thus, in order to satisfy the requirements of 10 CFR 50, Appendix H and ASTM E185-82, a modification to the proposed withdrawal schedule is necessary.

1 Accordingly, Wisconsin Electric hereby submits a modification to Technical Specification Change Request 134, substantially revising aur original request. We propose to remove the Unit 2 Capsule S in Fall 1990 refueling outage and Capsule P in Fall 1996 refueling outage. These withdrawal dates closely correspond to the estimated recc+0r vessel beltline forging fast neutron exposure levels at EOL and 150% EOL (See Table 1 for fluence comparisons). You may note that the target fluence for Capsule S slightly exceeds the projected EOL fluence for the vessel. This will allow some margin in the l event that the Capsule S exposure is overestimated as has occurred with some other capsules. With this margin there should be little i risk that the capsule data would represent a fluence level significantly less than the expected vessel EOL maximum. This is consistent with a recommendation made by Westinghouse.

Additionally, we propose to modify page 15.3.1-8a to acknowledge the fact that we are participants of the Babcock and Wilcox Master Integrated Reactor Vessel Surveillance Program and to briefly i describe the data points to which the new schedule corresponds.

Please note that this page is also affected by Technical Specification Change Request 126 submitted August 3, 1969; however,

. that change is not reflected here. The proposed Technical l

Specification pages are attached with changes indicated by margin bars.

As required by 10 CFR-50.91(a), we have evaluated these changes in accordance with the standards specified in 10 CFR 50.92(c) to determine if the change constitutes a significant hazards consideration. The revised surveillance capsule withdrawal schedule

.will provide reactor vessel materials data more representative of that predicted at EOL and 150% EOL. These changes also satisfy the requirements of 10 C?R 50, Appendix H, and ASTM E185-82. Therefore, these changes will not increase the probability or consequences of an accident previously analyzed nor will they involve a reduction in a margin of safety. These changes cannot create a new or different kind of accident since they involve only a schedule change which brings Point Beach into compliance with the regulations. Therefore, these changes do not involve a significant hazards consideration.

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Very.truly yours, u

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1 .R..W. Britt Chairman of.the Board;&

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Chief Executive. Officer Enclosures gopiesito NRC;' Regional Administrator,' Region III

.g NRC Resident Inspector L:  ? Subscribed and.swo n'to!before me b: ,

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