ML20245G098

Application for Amend to License DPR-27,consisting of Tech Spec Change Request 134,revising Tech Spec Table 15.3.1-2, Point Beach Nuclear Plant,Unit 2,Reactor Vessel Surveillance Capsule Removal Schedule

ML20245G098
Person / Time
Site:	Point Beach
Issue date:	08/08/1989
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20245G102	List: ML20245G098 ML20245K221
References
CON-NRC-89-097, CON-NRC-89-97 VPNPD-89-431, NUDOCS 8908150205
Download: ML20245G098 (4)

Text

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v Wisconsin Electnc eom couem 2'll W. MICHIGAN, P.O. BOX 2046, MILWAUKEE, Wl 53201 (414)221-2345 1

VPNPD-89-431 10 CFR 50.59 NRC-89-097 10 CFR 50.90-August 8, 1989.

U. S. NUCLEAR REGULATORY COMMISSION Document Control Desk Mail Station F1-137 Washington, D. C. 20555 Gentlemen:

DOCKET 50-301 TECHNICAL SPECIFICATION CHANGE REQUEST 134 REACTOR VESSEL SURVEILLANCE CAPSULE REMOVAL SCHEDULE POINT BEACH NUCLEAR PLANT, UNIT 2 In accordance with the requirements of 10 CFR 50.59 and 50.90, Wisconsin Electric Power Company (Licensee) hereby submits an application for amendment to Facility Operating License DPR-27 l

for Point Beach Nuclear Plant, Unit 2. The proposed amendment will revise Technical Specification Table 15.3.1-2, " Point Beach Nuc. lear Plant, Unit No. 2 Reactor Vessel Surveillance ,

Capsule Removal Schedule", to correspond to our plans regarding future surveillance capsule removal and testing, as discussed in our letter dated May 24, 1989. Specifically, it is our plan i to delay the removal and testing of-Unit 2, Capsule P from the l fall 1989 outage until the fall 1998 outage. This change is l identified by a margin'bar on the attached Technical Specification Table 15.3.1-2.

In August 1988 we became participants in the Babcock and Wilcox L Owners Group (BWOG) Reactor Vessel Integrity Program. We i joined the BWOG to obtain the fracture toughness data necessary to address 10 CFR 50, Appendix G requirements. Toward this i

end, all Westinghouse plant owners with reactor vessels manufactured by Babcock and Wilcox (B&W) and all of the B&W plant owners have been developing a master integrated reactor 9908250205 DR 890808 > i p ADOCK 0500 2 ggy  !

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a NRC bocument Control Desk

' August 8, 1989 Page 2 vessel surveillance program. As a result of this integration process, the plant-specific surveillance capsule test schedule for Point Beach is expected to be as shown on the revised Table 15.3.1-2. The revised scheduled removal dates accommodate the weld data needs of all the participants in the integrated program, as well as provide plate / forging material data and fluence data directly applicable to Point Beach Nuclear Plant, Unit 2.

The revised removal and testing schedule satisfies the governing requirements of 10 CFR 50, Appendix H and the incorporated requirements of 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 the 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." The revised schedule takes into account the vessel flux reductions we are now implementing, utilizing super low-leakage cores with hafnium absorbers in the guide ,

tubes of peripheral fuel assemblies. This will allow the remaining capsules to be pulled near peak vessel fluences corresponding to the expiration of our current license and the

.end of a twenty-year license extension.

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

The revised schedule is the result of participating in the ,

BWOG's Master Integrated Reactor Vessel Surveillance Program.

I Additionally, due to flux reductions which we are now

! implementing, the revised schedule will satisfy the  !

requirements of 10 CFR 50, Appendix H and ASTM E185-82. The delayed evaluation of vessel fluence (1) has no bearing on the j probability or consequences of an accident previously analyzed; (2) cannot create the possibility of a new or different kind of accident than any accident previously analyzed; or (3) does not l l

involve a reduction in a margin of safety. The proposed

} amendment, therefore, involves no significant hazards l consideration.

The proposed changes to the Technical Specifications are  !

attached.

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~ NRC Document Control Desk

. August 8,.1989 Page 3 L . . .

Please.. contact us if you have any. questions regarding this submittal.

-Very truly.yours,

.g hs y C. W,. Fay Vice' President-

. Nuclear Power Attachments-Copies to NRC Regional' Administrator Region III NRC-Resident Inspector-

' Subscribed and' sworn to before me

'this'4 D day of Avaa T

, 1989.

.3 JJ MM Notary Public, Stats of Wiscons2".n My ' Commission expiresj- E 7- 7 d =.

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dates accomodate the weld' data needs of all the participants in the Babcock.and 1Wilcox Master Integrated Reactor Vessel Surveillance Program. Additionally, the schedule will provide plate / forging material data as well as-fluence' data corresponding to the expiration of the current Unit 2 license and the end of a twenty year license extension.

References 4 (1) FSAR, Section 4.1.5 (2) Westinghouse Electric Corporation, WCAP-10638 (3) Westinghouse Electric Corporation, WCAP-8743 (4) Westinghouse Electric Corporation, WCAP-8738 Unit 1 Amendment Unit 2 Amendment 15.3.1-8a j i

t