ML20207R870
| ML20207R870 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 03/12/1987 |
| From: | Fay C WISCONSIN ELECTRIC POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20207R871 | List: |
| References | |
| CON-NRC-87-26 VPNPD-87-102, NUDOCS 8703180233 | |
| Download: ML20207R870 (4) | |
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Wisconsin Electnc powea couesur 231 W. MICHIGAN, P.O. BOX 2046, MILWAUKEE, WI 53201 (414)277-2345 VPNPD-87-102 NRC-87-26 March 12, 1987 CERTIFIED MAIL U.S.
NUCLEAR REGULATORY COMMISSION Document Control Desk Washington, D.C.
20555 Gentlemen:
DOCKETS 50-266 AND 50-301 TECHNICAL SPECIFICATION CHANGE REQUEST 117 FUEL ROD SUBSTITUTION IN FUEL ASSEMBLIES POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 In accordance with the requirements of 10 CFR 50.59 and 50.90, Wisconsin Electric Power Company (Licensee) submits an appli-cation for amendments to Facility Operating Licenses DPR-24 and DPR-27 for the Point Beach Nuclear Plant, Units 1 and 2.
These amendments would incorporate a change to Technical Specification 15.5.3.A.1,
" Reactor Core-General."
The proposed change modi-fies the specification addressing the substitution of fuel rods suspected of leaking.
We have attached the Technical Specifi-cation page containing the proposed revision and identified the change with a margin bar.
Technical Specification 15.5.3.A.1 provides a general descrip-tion of the reactor core and the fuel assemblies.
It also permits the repair of a fuel assembly wnich is suspected of leaking.
As presently worded, repair can be accomplished by substitution of an inert rod for a leaking rod, or removal of a leaking rod and leaving the assembly with an extra vacancy or water hole.
The specification presently limits substitution to no more than one fuel rod in any single assembly.
No more than six (6) such modified assemblies may be in the core at any time, g lS6DO Sed) Check W#
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i NRC Document Control Desk
. March 12, 1987-Page 2 Our proposed change to this specification will provide for more flexibility in.the repair of fuel assemblies with damaged or.
leaking fuel rods or other core design considerations.
The proposed wording allows for the substitution of multiple fuel rods in a fuel assembly by either substituting filler rods or leaving vacancies, provided the substitution is justified by a cycle specific reload analysis.
For'each fuel cycle an analysis is conducted to ensure that, with each reload of fuel, all core design safety criteria are met.
Westinghouse, our current nuclear core designer, conducts _
these cycle specific core design and safety analyses in accor-dance with their safety evaluation methodology (WCAP 9272-P-A) to verify that previous safety analyses, as presented in the Point Beach Final Safety Analysis Report (FSAR), are still applicable.
Under the proposed rewording of the specification, the core designer would employ a methodology which will specifically account for both the locations and the type of fuel rod sub-stitution.
Appropriate safety analyses will be conducted in conjunction with the normal reload analysis, verifying that all applicable core safety limits for fuel rods in the vicinity of the missing or substituted rods are still met.
By modeling based on the exact substitution, an accurate and complete safety analysis can be achieved, and conformance'with established safety margins will be ensured.
These analyses and the core reload changes are also reviewed as required by 10 CFR 50.59 by the licensee's staff.
The Commission has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing certain examples in 51 FR 7751.
The examples of actions which involve no significant hazards consideration include Item lii which states:
"For a nuclear power reactor, a change resulting from a nuclear reactor core reloading, if no fuel assemblies are significantly differ-ent from those found previously acceptable to the NRC for a previous core at the facility in question are involved.
This assumes that no significant changes are made to the acceptance criteria for the technical specifications, that the analytical methods used to demonstrate conformance with the technical specifications and regulations are not significantly changed, and that the NRC has previously found such methods acceptable."
10
t flRC Documentation Control Desk March 12, 1987 Page 3 We believe that the proposed change to the Technical Specifica-tions falls within the bounds of this example.
We emphasize that there are no changes to the acceptance criteria for the Technical Specifications.
All safety-related criteria presently existing and found acceptable by the NRC will remain in effect.
As further required by 10 CFR 50.91(A), we have evaluated this change request in accordance with the standards specified in 10 CFR 50.92 to determine if the proposed change constitutes a significant hazards consideration.
A proposed amendment involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or conse-quences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety.
The first criterion is not violated.
The present Technical Specification allows for fuel rod substitution or vacancies.
While the proposed change alleviates the discrete limits specified for such cases, the more important requirement of satisfying a core specific reload analysis remains in effect.
By taking into account any fuel rod substitutions or vacancies, that analysis will verify that all applicable safety margins as defined in the licensing documents are not reduced.
Therefore, there can be no increase in the probability or consequences of an analyzed accident.
Similarly, the second criterion is not violated.
While fuel assemblies containing the rod substitution or vacancies repre-sent a change in the physical core configuration, it is a physical change which is no more significant than, for example, using fuel of a different enrichment from a previous cycle.
Any such changes will be accounted for in the reload analysis.
Our proposed change merely states that rod substitution or vacancies must be justified by reload analysis.
Given successful comple-tion of such an analysis, it is not possible to create a new or different kind of accident.
The third criterion is also not violated for the same reasons described above.
If the physical parameters of the reload core are evaluated as being within previously defined acceptance criteria, then a reduction in the margin of safety is precluded.
4-e
. Document Control Desk
. March 12, 1987 Page 4 We request that the NRC review and approve this change ~ request by May 25, 1987, to. coincide with the proposed completion of the scheduled Unit 1 outage.
We have enclosed a check in the amount of $150 for the applica-tion fee. prescribed in 10 CFR 170.
Please contact us at once if you have any questions concerning this request.
Very truly yours, f
hf/
C. W.
Fa Vice President-Nuclear Power Enclosures (Check 946912 )
Copies to NRC Regional Administrator, Office of Inspection and Enforcement, Region III; NRC Resident Inspector; R. S. Cullen, PSCW i
t Subscribed and sworn to before me this l,2 % day of March 1987.
4 Notary Public, State of Wisconsin J
l My Commission expires I'27*30 i
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