ML20203N706

Safety Evaluation Supporting Amend 2 to License NPF-47

ML20203N706
Person / Time
Site:	River Bend
Issue date:	10/09/1986
From:	Office of Nuclear Reactor Regulation
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ML20203N703	List: ML20203N701 ML20203N705 ML20203N706
References
NUDOCS 8610170213
Download: ML20203N706 (6)

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$;f gC SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 2 TO FACILITY OPERATING LICENSE NO. NPF-47 GULF STATES UTILITIES COMPANY RIVER BEND STATION, UNIT 1 DOCKET NO. 50-458

1.0 INTRODUCTION

In River Bend Technical Specification 3.6.4, Table 3.6.4-1, Item a.1 identifies the reactor core isolation cooling (RCIC) steam supply valve 1E51*MOVF076 (M0VF076) as a primary containment automatic isolation valve.

By letters dated September 17, 1986 and September 19, 1986, the Gulf States Utilities Company (GSU) requested a one-time change to the Table by adding a footnote, applicable to isolation valve MOVF076, that would exempt the operability requirements for this valve from September 22, 1986 through October 4, 1986.

While the reactor was at 100% power on September 8, 1986 the M0VF076 valve was undergoing a functional test of the annunciator when the MOV76 breaker tripped due to a ground in the system. A subsequent attempt to stroke the valve indicated the valve to be fully open. As required by the River Bend Technical Specification 3.6.4 Action Statements, the outboard isolation valve 1E51*M0VF064 (M0VF064) was closed. Consequently, the RCIC system was rendered inoperable and by Specification 3.7.3, the plant was placed in a 14-day Action Statement that requires the plant to initiate a shutdown on September 22, 1986.

The GSU proposed technical specification revision will exempt valve MOVF076 from technical specification operability requirements, thereby allowing isolation valve F064 to be opened and returning the RCIC system to operable status. This reouest would be a one-time modification to the Technical Specification effective through October 4, 1986, at which time GSU has scheduled a shutdown and will repair the inoperable isolation valve (M0VF076).

2.0 EVALUATION The RCIC steam turbine supply line is an 8-inch line that branches off the main steamline "A". Inside the River Bend drywell the 8-inch line has an inboard isolation valve 1E51*M0VF063 (M0VF063) and circumventing this valve is a 3/4 inch bypass line consisting of a manual isolation valve and the motor-operated isolation valve M0VF076. The 3/4-inch bypass line is parallel to the 8-inch inboard isolation valve; its functions h id1No2 b e65'009 ' l PDR ADOCK 05000458

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w during start-up are to equalize the pressure around MOV63 and to warmup the RCIC steamline. The common out-board isolation valve (8-inch) is MOVF064. During power operation, the normal RCIC turbine steam supply line valve lineup is M0VF076 closed and inboard and outboard isolation valves MOVF063 and MOVF064 open.

The RCIC is a high pressure system primarily designed to maintain suffi-cient water in the reactor pressure vessel to cool the core and to maintain the reactor in the standby condition in the event the vessel becomes isolated and feedwater is not available. During a LOCA, RCIC (which augments other core injection systems) initiates on low vessel water level and delivers rated flow to the vessel through a connection in the vessel head.

1 Even though RCIC is not a part of the ECCS network, the RCIC system is designated as a safety-related system. During limiting conditions of operation (LCO) (i.e., when HPCS is inoperable), power operation is allowed to continue for a period of time provided RCIC is operable.

Valve MOVF076 is only used for RCIC turbine steam supply line warm-up before start-up of the system and the ability to close the valve is not required for the RCIC system to perform its intended design function.

Therefore, in the open position MOVF076 will not affect system operation.

The RCIC steam line isolation valves MOVF063 and M0VF064 are normally kept open and should be kept open in order to keep the steam inlet lines warmed-up and to minimize the potential for water hammer upon system startup. Accordingly, the RCIC steam supply line does not receive the standard containment isolation signals (e.g., high drywell pressure and low reactor water level). The isolation signals that serve the inboard and outboard isolation valve monitor conditions to reveal a break in the RCIC system.

The applicant in a letter dated September 19, 1986, also submitted a radiation dose assessment. In the event of a pipe break downstream of the outboard isolation valve (MOVF064), the leakage will be detected by the leakage detection systems. The leakage detection systems are the area i temperature monitoring system and the RCIC steam line flow monitoring system. Either leakage detection system will provide an isolation signal to MOVF063 and M0VF064. Once a signal is received, a total of a 20-second (a 10-second delay for diesel start plus a 10-second delay for valve closure) delay is assumed for closing of the valve. Full flow is assumed through the 8-inch line for the first 20 seconds. At the end of 20 seconds, the inboard-isolation valve F063 is assumed to close, but the outboard isolation valve fails to close (single failure). With the inoperable valve, F076, in the failed open position and the outboard F064 assumed to be failed open, there will be an unisolated leakage path (3/4-inch line) from the reactor to the auxiliary building. The maximum expected flow through the bypass .(F076) is 30 GPM. Feedwater flow could provide make-up for this fluid loss. If feedwater is not available, the

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l High Pressure Core Spray (HPCSI system could be used to provide cooling water to the vessel, thereby preventing uncovering of the reactor fuel.

Fuel failure would, therefore, be precluded. The only radioactivity available for release from the break is that which is present in the reactor coolant prior to the break. The bypass flow is assumed to be constant for the first 30 minutes of the transient and then is ramped down to zero in the next 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> as the reactor is brought down to a cold shutdown condition. The total mass released from the RCIC steam line break is estimated to be about 50,000 lbs compared with the-140,000 lbs estimated by the staff for Main Steam Line Break Accident Analysis. In section 15.6.4 of the River Bend SER, we concluded that.the doses for Main Steam Line Break are only a small fraction of 10 CFR Part 100 Guideline values. Since the RCIC piping break with unisolated bypass flow through the F076 valve is bounded by the Main Steam Line Break Analysis, the expected doses for the RCIC break also will be a small fraction of 10 CFR Part 100 guideline values, and hence within our acceptance criteria.

The staff concurs with GSU's assessment that there is no significant reduction .in the margin of safety relating to the temporary period (i.e.,

about 12 days) of having the M0VF076 valve in an inoperable condition.

The alternative approach would require reactor shutdown in order to enter the drywell and subject plant personnel to exposures from the harsh drywell environment. Therefore, such an action would induce unwarranted and inherent challenges to various plant systems and to the affected plant personnel.

3.0 BASIS FOR EMERGENCY CIRCUMSTANCES The Commission published guidance concerning emergency circumstances in the Federal Register on March 6, 1986 (51 FR 7762) which states, in part:

"The staff may receive an amendment request and find an emergency situation, where failure to act in a timely way would result in derating or shutdown of a nuclear power plant...(NRC) may proceed to issue the license amendment, if it determines, among other things, that no significant hazards considerations are involved." In addition, the Commission further states in 51 FR 7762, "...that it expects its licensees to apply for license amendments in a timely fashion..."

On September 8, 1986, while the licensee was performing a functional test on annunciator 1E51*M0VF076, the MOV76 breaker tripped with the valve open.

MOV76 is a 3/4" bypass line valve used for prewarming the RCIC steam line.

With this valve open RCIC was declared inoperable and the licensee entered into a 14-day LC0 within which time the licensee had to restore RCIC to operable status or shutdown.

The licensee determined that the breaker malfunction was due to an unintended electrical ground, apparently due to a wet valve motor. Since the valve is inside the containment, the licensee attempted to close the valve by remote means. By September 15, 1986, the licensee concluded that the valve could not be closed remotely, f

The licensee had three basic options:

(1) Derate to a very low power level, enter containment and open/ repair the valve (2) Shutdown (3) Request an amendment to their operating license to allow continued l operations. l l

The sta H finds that denial of the licensee's request for an amendment to their operating license would have placed the reactor in a shutdown or derated condition. Moreover, the licensee could not have foreseen this need for a license amendment in advance because it involved a component failure. Furthermore, even had the licensee requested an amendment of the license on the date the event occurred (September 8, 1986), the 14-day LC0 would have +/equired-NRC to issue the amendment on an emergency basis.

4.0 NO SIGNIFICANT HAZARDS CONSIDERATION The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amend-ment to an operating license for a facility involves no significant hazards considerations if operation of the facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (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 licensee has provided an analysis of significant hazards considerations in its September 17, 1986 request which was amplified by their letter of September 19, 1985 for a license amendment. The licensee has concluded, with appropriate bases, that the proposed amendment meets the three standards in 10 CFR 50.92 and, therefore, involves no~significant hazards

considerations. The Commission has also provided guidance concerning the i application of these standards by providing examples of- amendments i considered likely, and not likely, to involve a significant hazards consideration. These were published in the Federal Register on March 6,

'096 (5*. FR 7744). The NRC staff has made a preliminary review of the see's submittal. A discussion of these examples as they relate to aroposed amendment follows.

0: of the examples of actions involving no significant hazards consider-ation (vi) is a change which either may result in some increase to the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the standard review plan.

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The proposed amendment to the Technical Specifications may involve some increase, but in our judgment not a sionificant increase in the probability ,

or consequences of an accident previously evaluated because the potentially '

unisolated penetration is limited to a 3/a" line discharging to a closed system that returns to the containment or assuming a break in the RCIC steam line between valve M0v64 and MOV45,10CFp Part 100 limits will not be exceeded. '

This conclusion is based upon the fact that the fuel failure is not expected i because the make-up water from the fluid loss can be provided by the Feedwater i System or the High Pressure Core Spray system. The only activity available for release from the break is that which is present in the reactor coolant and steam lines prior to the break. Moreover, large break LOCA (Double-Ended Recirculation Line Break) with fuel failure assumes a closed loop in the RCIC system and filtration by the Standby Gas Treatment System for any valve packing leaks and meets the current accident analysis and 10CFR Part 100 limits.

The proposed amendment does not create.the possibility of a new or different kind of accident from any accident previously evaluated because line breaks such as may be postulated for the RCIC system have been considered in the analysis and design basis with many of the previous analyses bounding all the short-term consequences of the plant for this event. There may, however, be lonaer term environmental effects which would only impact the overall lifespan of the equipment and would be evaluated if such an event should occur. This change introduces no new mode of operation and no change in the plant design is being made.

The proposed amendment does not involve a significant reduction in the margin of safety because this event is bounded by a previous analysis and design basis. Therefore the margin of safety remains the same.

5.0 STATE CONSilLTATIONS In accordance with the Commission's regulations, consultation was held with officials of the State of Louisiana by telephone on September 19, 1986. The Manager, Nuclear Pro,iects of the Nuclear Energy Division, Office of Air Quality and Nuclear Energy, Department of Environmental Ouality of the State of Louisiana, had no objections to the proposed action.

6.0 PUBLIC COMMENTS None 0

7.0 ENVIRONMENTAL CONSIDERATION

This amendment involves a change to a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The staff has determined that the amendment involves no significant increase in the amour.ts, and no signifi-cant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has made a final no significant hazards consideration finding with respect to this amendment.

Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

8.0 CONCLUSION

The staff has concluded, based on the considerations discussed above that:

(1) the amendment does not (a) significantly increase tne probability or consecuences of an accident previously evaluated, (b) create the possibil-ity of a new or different kind of accident from any previously evaluated or (cl significantly reduce a safety margin and, therefore, the amendment does not involve a significant hazards consideration; (2) there is reason-able assurance that the health and safety of the public will not be endan-gered by operation in the proposed manner; and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of the amendment will not be inimical to the common defense and the security or to the health and safety of the public.

Principal Contributors: A. Notafrancesco C. Schulten S. Stern G. Thomas Dated: October 9,1986

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