ML20138N195

Application for Amend to License NPF-39,changing Tech Specs to Extend,By 14 Wks,Surveillance Testing Interval for Reactor Instrumentation Line Excess Flow Check Valves

ML20138N195
Person / Time
Site:	Limerick
Issue date:	12/18/1985
From:	PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20138N178	List: ML20138N176 ML20138N195 ML20138N213
References
NUDOCS 8512230403
Download: ML20138N195 (8)

Text

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BEFORI: THE UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of  :

PHILADELPl!IA ELECTitIC COMPANY APPLICATION FOR AMENDMENT OF FACILITY OPERATING LICENSE NPF-39 Edward G. Dauer, Jr.

' Eugene J. Bradley 2301 Market Street Philadelphia, Pennsylvania 19101 At to rneys fo r Philadelphia Electric Company 2

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M BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of  :

Docket No. 50-352 PHILADELPHIA ELECTRIC COMPANY APPLICATION MR AMENDMENT OF FACILITY OPERATING LICENSE NPF-39 Philadelphia Electric Company, Licensee under Facility Operating License NPF-39 for Limerick Generating Station Unit 1, hereby requests that the Technical Specifications contained in Appendix A to the Operating License (NUREG-1149) be temporarily amended to provide an extension of fourteen weeks to the surveillance testing interval for the reactor instrumentation line excess flow check valves contained in Technical Speci fication 4.6.3.4 (page 3/4 6-10) .

In order to meet the requirements of the Technical Speci fications , it will be necessary to shutdown the plant prior to February 19, 1986 to perform the necessary testing. A

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shutdown is necessitated because the valves in question, which are functionally tested by opening a line downstream of the valve with the reactor pressurized, serve one or more components which I-l must be removed from service during testing. This action could result in Emergency Core Cooling System, Reactor Protection system or Nuclear Steam Supply Shutof f System actuations, or in a condition prohibited by Technical Specifications. To do this testing at power also poses a risk of personnel injury, in the unlikely event that one of the valves fails to check, due to high temperature water or radiation hazard. The estimated duration of this testing would be approximately fourteen days as necessitated by the operational requirement to cool the reactor to a decay heat level consistent with the heat removal capabilities of the Reactor Water Cleanup (RWCU) system.

The long time associated with obtaining the full power license led to the need for this extension. A normal schedule for low power testing, startup testing and 100-hour full power warranty run would not have resulted in a requirement to extend the testing interval. All low power (less than 5% thermal power) testing was completed prior to late April,1985. Circumstances beyond licensee's control delayed the issuance of the full power license until Augus t, 1985. During this period of time, the unit was maintained in a 48-hour standby condition to demonstrate its availability for operation. This action precluded testing the excess flow check valves.

The current schedule is for a maintenance and surveillance testing outage to begin on or before May 26, 1986.

During this outage, maintenance activities, surveillance testing, w - __-_ -__ - _-________-_- _

and minor plant modifications will be performed Which will allow l ,

the plant to operate through the first refueling outage. The fourteen-day outage required to perform the testing of the excess flow check valves would result in a net increase in overall outage time if an extension was not permitted. This additional outage would impose an economic penalty of greater than 6 million dollars to area customers as a result of the cost of replacement generation and would also subject plant equipment and systems to the detrimental ef fects inherent in an additional shutdown and startup operation.

The refo re , lice nsee requests an extension of fourteen weeks to the surveillance testing interval for reactor instrumentation line excess flow check valves for the first cycle so that this testing may be performed concurrent with a maintenance outage currently scheduled for late May,1986.

Signi ficant Hazards Determination The Commission has provided guidance concerning the application of standards in 10 CFR 50.92 for determining whether license amendments involve a significant hazards consideration by providing certain examples which were published in Federal

• Register on April 6,1983 (48 FR 14870) . One of the examples (vi) of an action involving no signi ficant hazards consideration is a change which may in some way reduce a safety margin, but Where the results of the change are clearly within all acceptable criteria. The requested change fits this example. Postponing 3

l the aforementioned surveillance testing until an outage commencing in late May,1986 would allow for continued operation of the plant and would have little or no ef fect on containment integrity for the following reasons:

1. The following design features would limit inventory loss in the event of a reactor instrument line rupture coincident with the f ailure of the excess flow check valve to close:

a) The lines in question are one-inch in diameter or less.

b) These lines are equipped with one-quarter inch restricting orifices, inside containment, which serve to limit flow.

c) The line rupture, in order to pose a hazard, would have to occur outside of primary containment, where the majority of the line is only 3/8" diameter.

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d) The excess flow check valves are designed so that should they f ail to close the main flow path through the valve has a flow resistance equivalent to a sharp edged orifice of 0.375 inch diameter.

2. Manual valves are available to shut of f the protected line, outside of primary containment, should any indication be present concerning excess flow check valve in ope rabili ty.

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3. The excess flow check valves are located outside of primary containment therefo re, they are available for t

periodic visual inspection, i f neces sa ry.

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4. The lines which are protected by the excess flow check valves are located within Ehe reactor enclosure Which is served by the standby ges treatment rystem which would filter and monitor any release.

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, 5. A rupture of a single instrument line, assuming the failure of the excess flow check valve t,o seat, will not result in a release of radioactivity in excess of 10 CFR Part 100 limits (PSAR Table 15.6-7) .

6. Excess flow check valves have exhibited a high degree of reliability in performing their " checking" function thus, the inspection interval which is designed to ,

provide a high probability of detection of a leaking valve is very conservative and the probability of detection will not be significantly reduced by the requested interval extension of less than 204.

A review of the Nuclear Plant Reliability Data System and a poll of several utilities having similar make and model valves revealed no instances of the valves failing to perform their safety-related functiph. During the first surveillance tests, all valves tested successfully. Philadelphia Electric's Peach Bottom Units 2 and 3 have valves which are similar in design, although by a dif ferent manufacturer, and have had a high degree of success with these valves checking properly.

! For these reasons, the proposed temporary amendment to the Limerick Operating License does not constitute a significant hasards consideration in that it would not:

1. Involve a signi ficant increase in the probability or consequences of an accident previously evaluated because the change extends the surveillance interval less than 20% beyond the current conservative surveillance requirements and has no ef fect on the assumptions of valve failure assumed in the present analysest or

2. Create the possibility of a new type of accident or a dif ferent kind of accident from any accident previously analyzed because current analyses assume valve failure concurrent with line rupture. No new accident scenarios are credible based upon scheduling of this testing aloner or

3. Involve a significant reduction in the marain of safety because the design addresses failures of the excess flow valves to function by the use of small lines, restricting orifices and valve body impediments to free flow.

The requested amendment will not result in a signi ficant change in the types or amounts of any ef fluents that may be released of f-site in that the chaivje is schedular in nature and af facte no systems concerning of fluents.

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There will be r.o significant increase in individual or cumulative occupational radiation'Npdrure as a result of the requested amendment which merely requests to delay testing which will be performed regardless of the outcose of the amendment request.

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The Plant operations Revkew Committee and the Nuclear Review Board have reviewed these proposed temporary changes to the Technical Speci fication~s and'have concluded that they do not involve an unreviewed safety question or a significant hazards consideration and will not 'andakter the public health and safety.

Hospectfully Submitted, Pl!ILAr:ELPHIA ELECTRIC COMPANY

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