ML20207N513

Application for Amends to Licenses DPR-24 & DPR-27,changing Tech Spec 15.4.1-1 to Clarify Requirements for Reactor Coolant Flow Logic Testing & Specifying Conservative Conditions to Perform Surveillance.Fee Paid

ML20207N513
Person / Time
Site:	Point Beach
Issue date:	01/06/1987
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF ADMINISTRATION (ADM)
Shared Package
ML20207N515	List: ML20207N513 ML20207N557
References
CON-NRC-87-3 VPNPD-87-3, NUDOCS 8701140263
Download: ML20207N513 (3)

Text

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Wisconsin Electnc eom couem 231 W. MICHIG AN. P.O. BOX 2046, MILWAUKEE, WI 53201 (414)277-2345 VPNPD-87-7 NRC-87-3 January 6, 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 115 SURVEILLANCE REQUIREMENTS - REACTOR TRIP LOGIC TESTING ON LOSS OF FLOW IN BOTH LOOPS POINT DEACH NUCLEAR PLANT, UNITS 1 AND 2 In accordance with the requirements of 10 CFR 50.59 and 59.90, 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 Plant, Units 1 and 2.

These amendments would incorporate a change to Technical Speci-fication Table 15.4.1-1, " Minimum Frequencies for Checks, Cali-brations and Tests of Instrument Channels." The proposed change clarifies the requirements for Reactor Coolant Flow Logic test-ing. We have attached the Technical Specification page contain-ing the proposed change which is identified by a margin bar.

The Point Beach Nuclear Plant Reactor Protection System (RPS) incorporates provisions for a reactor protective action in the event of a partial or full loss of flow under various power conditions. The protective action - reactor trip - occurs when instrumentation senses a loss of flow in both loops "A" and "B" while reactor power is between 9.5% and 49% of rated thermal power and when a loss of flow occurs in either loop "A" or "B" and reactor power is greater than 49%.

f O 66 PDR QU

l NRC Document Control Desk January 6, 1987 Page 2 Reactor Coolant Flow Instrumentation provides redundant indica-tion as well as redundant input to the RPS. Flow is measured by sensing the differential pressure across a primary piping cold leg elbow. Three differential pressure cells connect to the elbow tap (one in loop "A", one in loop "B") and are part of three independent flow channels. Each channel provides indica-tion (visible in the control room) and inputs to RPS when a bistable trips at the setpoint of 93% rated flow (decreasing).

RPS logic circuitry will interpret that a low flow condition exists in a loop when a 2/3 channel coincidence for that loop is satisfied.

Item 5 of Table 15.4.1-1, " Reactor Coolant Flow", delineates check, calibration, and test periodicities for the reactor cool-ant flow channels. The monthly test periodicity required for flow instrumentation is met through analog channel and logic channel testing. Logic channel testing involves checking the ability of the RPS to cause a reactor trip with all possible combinations of low flow on 2 of 3 flow channels in either loop "A" or "B". This testing is performed with the reactor critical by testing only one train of instrumentation at a time. The reactor trip breaker associated with the train being tested is bypassed by racking in and shutting its associated bypass breaker. This bypass breaker would be tripped automatically during this time should an actual condition occur requiring a trip as sensed by the other instrumentation train.

The circuitry used for logic testing was not designed to allow testing at power of the contacts which initiate the double loss of flow trip. Although the same bistables and relays are used to actuate the single and double loss of flow trips, the relay contacts used are different. Consequently, these contacts can-not be tested without actually generating a low RCS flow signal in two of three flow channels associated with one loop. This would result in tripping bistables which actuate relays in both trains of reactor protection. If this testing is performed above 49% power, a reactor trip will occur because the single loss of flow trip in the on-line train is unblocked above 49%

power.

Our change to the Technical Specifications adds a statement to the " Remarks" column for Item 5 of Table 15.4.1-1. The state-ment directs logic channel testing on loss of flow in both loops to be tested each refueling interval. This will ensure complete testing of the circuitry for a double loss of flow reactor trip.

The bistables and relays will continue to be tested monthly as part of the single loss of flow trip test.

As required by 10 CFR 50.91(a), we have evaluated this change in accordance with the standards specified in 10 CFR 50.92 to determine if the proposed change constitutes a significant

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NRC Documentation Control Desk January 6, 1987 Page 3 hazards consideration. 10 CFR 50.92 states that a proposed I amendment involves no significant hazards consideration if oper-ation 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.

The proposed change is purely administrative in nature, involv-ing no physical change to the plant or procedural change which will affect operation of the plant. The change merely provides clarifying guidance relative to the conditions necessary for performing a surveillance. Therefore, the amendment will not cause an increase in the probability or consequences of an accident previously evaluated.

The same argument can be applied to the second criterion of 10 CFR 50.92. The change adds a surveillance requirement and conservative conditions in which to perform the surveillance.

Thus, no new or different accident from any previously evaluated accident can be created.

Lastly, a significant reduction in a margin of safety is not applicable to this change. The change both clarifies surveil-lance requirements and specifies conservative conditions in which to perform the surveillance. In this light, margin of safety is actually increased.

We have enclosed a check in the amount of $150 for the appli-cation fee prescribed in 10 CFR 170. Please contact us at once in you have any questions concerning this request.

Very truly yours, i%:[-

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C. W.'Fa Vice President-Nuclear Power Enclosures (Check 936744)

Copies to NRC Regional Administrator, Office of Inspection and Enforcement, Region III; NRC Resident Inspector; R. S. Cullen, PSCW Subscribed and sworn to before me this 9 Mi day of January 1987.

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L ccDJ (A M f Notary Public, StateDpf Wisconsin My Commission expires JI- 3 7 - 9 0

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