ML20246N038
| 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: |
| References | |
| CON-NRC-89-104 NUDOCS 8909080023 | |
| Download: ML20246N038 (3) | |
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' 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That request proposed that Unit.2 reactor vessel surveillance Capsule S be withdrawn in Fall 1995 and Capsule P be withdrawn in Fall 1998.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
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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}ge 2
n/cm.
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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U.S.-NRC Document Control: Desk-L7 August 31, 19.89 Fage:3E F
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.Please contact us.if you have'any questions concerning this request; l i, +, '
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Very.truly yours, u
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.R..W.
Britt 1
Chairman of.the Board;&
Chief Executive. Officer i
1 Enclosures gopiesito NRC;' Regional Administrator,' Region III
.g NRC Resident Inspector L:
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Subscribed and.swo n'to!before me b:
this y day'of'./
mM: 1989.
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