ML20203M883

Application for Amends to Licenses DPR-24 & DPR-27, Incorporating Tech Spec Change Request 111 to Revise Surveillance Requirements for Main Steam Stop Valves,Safety Valves & Pressurizer Safety Valves.Fee Paid

ML20203M883
Person / Time
Site:	Point Beach
Issue date:	08/26/1986
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	Harold Denton, Lear G Office of Nuclear Reactor Regulation
Shared Package
ML20203M886	List: ML20203M883 ML20203M888
References
CON-NRC-86-76 VPNPD-86-334, NUDOCS 8609050129
Download: ML20203M883 (4)

Text

,s Wisconsin Electnc eowca couraur 231 W. MICHIGAN. P.O. BOX 2046. MILWAUKEE, WI 53201 (414)277-2345 VPNPD-86-334 NRC-86-76 August 26, 1986 CERTIFIED MAIL Mr. H. R. Denton, Director Office of Nuclear Reactor Regulation U. S. NUCLEAR REGULATORY COMMISSION Washington, D. C. 20555 Attention: Mr. George Lear, Project Director PWR Project Directorate 1 Gentlemen:

DOCKETS 50-265 AND 50-301 TECHNICAL SPECIFICATION CHANGE REQUEST 111 SURVEILLANCE REQUIREMENTS - MAIN STEAM STOP VALVES AND SAFETY VALVES, PRESSURIZER SAFETY VALVES 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) hereby submits an application for amendments to Facility Operating Licenses DPR-24 and DPR-27 for the Point Beach Nuclear Plants, Units 1 and 2. The purpose of these amendments is to incorporate changes to Technical Specification 15.4.1,

" Operational Safety Reviews," and Technical Specification 15.4.7, " Main Steam Stop Valves." The proposed changes revise the surveillance requirements for main steam stop valves, main steam safety valves, and pressurizer safety valves. Technical Specification pages containing these proposed changes which are identified by margin bars are attached.

The proposed change to Technical Specification 15.4.1,

" Operational Safety Review," involves Items 11 and 12 of Table 15.4.1-2 and revises the set point testing periodicity of the main steam safety valves and the pressurizer safety 3 valves from each refueling to once every five years for 3 i cach valve. A note has been added which would require testing of additional valves if valves fail the scheduled test. The specification would state that an approximately 60 G609050129 860826 h PDR ADOCK 05000266 knQ\

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Mr. H. R. Denton '

August 26, 1986 Page 2 equal number of' valves will be tested each refueling outage so that all valves will be tested within a five-year period.

This periodicity, and provision for additional testing in the case of valve failure, is consistent with the ASME -

• Section XI code for inservice testing of Category C safety and relief valves and is consistent with the Point Beach Nuclear Plant Inservice-Testing Program presently under review by the NRC. We have received written relief from the Wisconsin Department of Industry, Labor, and Human Relations to perform the setpoint testing in accordance with the-applicable ASME Section XI code requirements in lieu of performing the state-required annual lift test.

The proposed change to Technical Specification 15.4.7, " Main Steam Stop Valves," wouldfrevise the test conditions for the main steam stop valves. Presently, valve closure times are verified under no-flow conditions during reactor shutdowns for major refueling. Our proposed specification would

- change the test. conditions to " low-flow conditions." The i basis for the specification would remain the same.

This surveillance serves to test a component which.is addressed in the safety analysis of the steam line rupture accident. This accident analysis is described in Section l

14.2.5 of the Point Beach Nuclear Plant Final Safety -

Analysis Report (FSAR). Successful shutting of a main steam l

stop valve or the shutting of a downstream non-return check valve is necessary to prevent the blowdown of both steam generators in the event of a steam line break. The analysis i

assumes the closure of a main steam stop valve at the five-second point in the transient. The analysis also states that while the valves are designed to be fully closed in five seconds with no flow through them, the high flow existing during a' steam line rupture will cause the valves to close considerably faster. This is attributed to the design of the valves. The valves are classified as self-aligning, swinging disc, inclined seat-type " check" r valves. The disc operating mechanism consists of an air cylinder installed on the outside of the valve body and connected to the valve shaft by means of a piston rod and linkage. The orientation of the valve is such that steam flow will act to aid in seating the valve disc, thereby decreasing closure time.

We desire to test the valves in the condition in which they will have to perform their safety function, that is, hot and in a steam environment. This can be most readily accomplished following a refueling shutdown during main steam system warmup. At this time the main steam stop valves are bypassed to warm up the steam system in the

Mr. H. R. Denton August 26, 1986 Page 3 turbine hall. Upon equalization of pressure across the main steam stop valves, the valves can be exercised and closure times verified toInbethis within the limits of the Technical condition, however, absolutely zero Specification.

steam flow cannot be maintained, although existing flow would be minimal.

Testing the valves at zero flow is conservative in view of the argument that if the valves close in less than five seconds with no flow, they will close even more rapidly in the expected high flow conditions of a main steam line rupture. We maintain that the same argument is true for testing the valves in the low flow condition described above. The existing FSAR analysis would still be valid.

Some minor changes to the descriptive wording in FSAR Section 14.2.5 would, however, be necessary to reflect low flow vice no-flow testing. These changes would be included in the 1987 update of the FSAR.

As required by 10 CFR 50.91(a), we have evaluated these changes in accordance with the standards specified in 10 CFR 50.92 to determine if those proposed changes constitute a significant hazards consideration. 10 CFR 50.92 states that 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 consequences 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.

Changing the main steam and pressurizer safety valve testing periodicity does not significantly increase the probability or consequences of an accident previously evaluated in that we are requesting a periodicity which is in compliance with j

the guidelines of a nationally accepted standard. l similarly, the changing of the test conditions for main steam stop valve surveillance does not alter the initial conditions or consequences of the analyzed main steam line rupture for the reasons previously described.

Regarding the second criterion, these changes are revisions to surveillance requirements and conditions. Thus, no new or different accident can be created as no changes or modification to the physical plant have occurred.

Mr. H. R. Denton August 26, 1986 Page 4 Lastly, a significant reduction in a margin of safety is not applicable to these changes. Again, the changes relative to main steam and pressurizer safety valve testing are a request for adherence to the guidelines of the ASME Section XI code for inservice testing of safety valves. The purpose of this section of the code is to ensure a sufficient margin of safety exists relative to the proper functioning of these components, verifiable through a specified testing periodicity. Also, no reduction in the margin of safety will occur with the new test conditions for main steam step valve surveillance. Since the applicable accident analysis remains unchanged, the margin of safety remains unaffected.

We, therefore, conclude that the proposed changes do not involve a significant hazards consideration.

In accordance with the Commission's regulations, we are enclosing three signed originals and, under separate cover, forty copies of this amendments application. We have also enclosed a check in the amount of $150 for the application fee prescribed in 10 CFR 170. Please contact us at once if you have any questions concerning this request.

Very truly yours,

/.

N C. W.iF ay Q'

Vice President Nuclear Power Enclosures (Check 920876 )

Copies to NRC Resident Inspector R. S. Cullen, PSCW Subscribed and sworn to before me this R6 ih day of d o n..#, 1986.

6 s mo WA s vo ed Notary Public, State of Wisconsin l

My commission expires May 27, 1990.

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