ML20195C293
| ML20195C293 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 10/28/1988 |
| From: | Fay C WISCONSIN ELECTRIC POWER CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML20195C298 | List: |
| References | |
| CON-NRC-88-103 VPNPD-88-526, NUDOCS 8811020434 | |
| Download: ML20195C293 (2) | |
Text
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i Wisconsin Electnc aca cwn 231 W MICHIGAN. P.O. BOX 2046. MILWAUKEE.W153201 (414)221 2345 VPNPD-88-526 NRC-88-103 October 28, 1988 U. S. NUCLEAR REGULATORY COMMISSION Document Control Desk Mail Station Pl-137 Washington, D.
C.
20555 Gentlemen:
DOCKETS 50-266 AND 50-301 ADDITIONAL INFORMATION FOR TECHNICAL SPECIFICATION CHANGE REQUEST 127 INCREASE ALLOWABLE CORE POWER PEAKING FACTORS POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 The enclosed Leport provides additional information requested by your staff during our October 6, 1988 meeting regarding Technical Specification Change Request 127.
The proposed changes provide for the design and operation of the Point Beach Nuclear Plant cores with enhanced Optimized Fucl Assembly Fuel and at higher core peaking factors than are allowed by current plant Technical Specifications.
This report provides more detail on the reference core design used in the analy'acs, as well as on the results of the safety analyses.
The remainder of this letter briefly addresses questions raised at our October 6 meeting regarding the methodologies used in the analyses:
1.
The steam generator tubo rupturo event reanalysis uses the same methodology presently used in Chapter 14 of the Point Beach Final Safoty Analysis Report (FSAR) with somo revised input assumptions, as outlined in the enclosed report.
In addition, the tubo uncovery issue for this event is being addressed generically through the ao Westinghouse Owners' Group (WOG) program presented to the
@Ng NRC on July 27, 1988.
Preliminary results of the WOG obs program are expected by January 1989.
mb QO 2.
Use of periphoral power suppression assemblics (PPSAs) as g$
they affect the accident analyses has been bounded in the wg reanalyses, as described in the enclosed report.
Other offects regarding the PPSAs and the specifics of their use o
will be discussed in a supplemental submittal to be mado h[
-m when further analysis is completed.
Wo expect this C
@@a supplement to be submitted in January 1989.
i,
NRC Document Control Desk October 28, 1988 Page 2 3.
New or revised methodologies employed in these reanalyses include the revised thermal design procedure (RTDP) and 4
the WOG dropped rod methodology.
The event analyses affected by the RTDP methodology include: Uncontrolled RCCA Withdrawal at Power, RCCA Drop, Excessive Load Increase Incident, Loss of External Electrical Load, and Loss of Reactor Coolant Flow.
The WOG dropped rod methodology affects only the RCCA drop analysis.
The WCAPs for these methodologies have been submitted for NRC review.
A Safety Evaluation Report (SER) for the RTDP methodology is expected shortly.
However, review of the l
' OG dropped rod methodology has apparently been delayed.
I Since it is used in our justification for this Technical Specification change, we again emphasize the importance of NRC approval of the WOG dropped rod methodology.
i I
4.
The small-break LOCA analysis used the NOTRUMP code, which has been generically approved by the NRC for Westinghouse plants.
i 5.
All the other analysis methodologies used, with the exception of the SGTR input changes already mentioned, are the same as those currently used in the FSAR analyses.
It i
should be noted that the locked rotor reanalysis employed i
the same methodology and mot-the same acceptance criteria i
as our current FSAR analysis, which was accepted as a part i
of our OFA fuel submittal in 1984.
t i
Proposed changes to the safety analysos, as described in Chaptor 14 of the FSAR, and the results of the largo-break LOCA i
j best-estimato analysis will be submitted separately.
These submittals will provide additional information supporting the 1
i request and may help you in your ovaluation.
Please contact us should you have any questions regarding the l
i information provided.
Very truly yours, j
(2d/; D W. Fay 2
C.
Vice President I
Nuclear Power Enclosure i
Copics to NRC Regional Administrator, Region III j
NRC Resident Inspector l
R.
S. Cullen, PSCW j
3 t
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