ML20149E883

Safety Evaluation Accepting Interim Relief Request IRR-03 Re Drywell Isolation Check Valves in Equipment Drain Lines & Reactor Equipment Closed Cooling Water Sys

ML20149E883
Person / Time
Site:	River Bend
Issue date:	08/02/1994
From:	Office of Nuclear Reactor Regulation
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ML20149E881	List: ML20149E876 ML20149E883
References
NUDOCS 9408080142
Download: ML20149E883 (5)

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[ * (#Y UNITED STATES i[s j NUCLEAR REGULATORY COMMISSION g  %,p j WASHINGTON, D C. 20555 4 001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO INSERVICE TESTING PROGRAM INTERIM RELIEF RE0 VEST IRR-03 GULF STATES UTILITIES COMPANY RIVER BEND STATION. UNIT 1 DOCKET NO. 50-458

1.0 INTRODUCTION

The Code of Federal Regulations, 10 CFR 50.55a, requires that inservice testing (IST) of certain American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Class 1, 2, and 3 pumps and valves be performed in accordance with Section XI of the ASME Code and applicable addenda, except where relief has been requested and granted or proposed alternatives have been authorized by the Commission pursuant to 10 CFR 50.55a(f)(6)(i), (a)(3)(i), or (a)(3)(ii). In order to obtain authorization or relief, the licensee must demonstrate that: (1) conformance is impractical for its facility; (2) the proposed ilternative provides an acceptable level of quality and safety; or (3) compliance would result in a hardship or unusual difficulty without a compensatiug increase in the level of quality and safety. Section 50.55a(f)(4)(iv) provides that inservice tests of pumps and valves may meet the requirements set forth in subsequent editions and addenda that are incorporated by reference in 10 CFR 50.55a(b), subject to the limitations and modifications listed, and subject to Commission approval.

NRC guidance contained in Generic Letter (GL) 89-04, " Guidance on Developing Acceptable Inservice Testing Programs," provided alternatives to the Code requirements determined to be acceptable to the staff and authorized the use of the alternatives in Positions 1, 2, 6, 7, 9, and 10 provided the licensee follow the guidance delineated in the applicable position. When an alternative is proposed which is in accordance with GL 89-04 guidance and is documented in the inservice testing (IST) program, no further evaluation is required; however, implementation of the alternative is subject to NRC inspection.

Section 50.55a authorizes the Commission to grant relief from ASME Code requirements or to approve proposed alternatives upon making the necessary findings. The NRC staff's findings with respect to granting or not granting the relief requested or authorizing the proposed alternative as part of the licensee's IST program are contained in this safety evaluation (SE).

In rulemaking to 10 CFR 50.55a effective September 8, 1992, (see 57 federal Register 34666), the 1989 edition of ASME Section XI was incorporated in 10 CFR 50.55a(b). The 1989 edition provides that the rules for IST of pumps and valves shall meet the requirements set forth in ASME Operations and 9408000142 940002 PDR ADOCK 05000458 P PDR

Maintenance Standards Part 6 (OM-6), " Inservice Testing of Pumps in Light-Water Reactor Power Plants," and Part 10 (OM-10), " Inservice Testing of Valves in Light-Water Reactor Power Plants." Pursuant to (f)(4)(iv), portions of editions or addenda may be used provided that all related requirements of the respective editions or addenda are met, and subject to Commission approval.

Because the alternatives meet later editions of the Code, relief is not required for those inservice tests that are conducted in accordance with OM-6 and OM-10, or portions thereof, provided all related requirements are met.

Whether all related requirements are met is subject to NRC inspection.

The licensee submitted inservice testing Interim Relief Request IRR-03 in a letter dated August 2, 1994, to address testing which had not been conducted in accordance with their IST program during their last refueling outage. The missed surveillance was discovered during a systematic review of the River Bend IST program currently being conducted by the licensee. A safety evaluation of the interim relief request is included below.

The licensee's IST program covers the first ten-year interval from June 16, 1986, to June 16, 1996. The River Bend Station, Unit 1, IST program was developed in accordance with the 1980 Edition of ASME Section XI through the Winter 1981 Addenda. '

2.0 RELIEF RE0 VEST 2.1 Interin Relief Reauest Number IRR-03 The licenset has requested interim relief from the inservice test frequency and procedure requirements specified in ASME Section XI, Paragraphs IWV-3521 and IWV-3522, for the reactor plant closed cooling water supply to recirculation pump bearing cooler drywell isolation check valve ICCP*V119.

The interim relief has been requested because approved testing to verify closure of this check valve during refueling outages was determined by the Jicensee not to be adequate to verify closure. The licensee has proposed to defer testing of this valve until the next refueling outage which is currently scheduled for September of 1995 to perform either flow testing or valve disassembly.

2.1.1 Licensee's Basis for Reauestino Relief The licensee states:

Quarterly reverse flow testing of this check valve would isolate cooling water flow to the reactor coolant recirculation pump and motor bearings. Interruption of cooling water flow to the reactor coolant recirculation pump c'd motor bearings could result in -

extensive damage to the pump and/or mete- bearings. Additionally, the valve cann't be tested quarterly as the drywell is not accessible during power operation.

Rtlief Request VRR-2, submitted and approved with Revision 6 of the River Bend Station Pump and Valve Inservice Testing Program Plan, allowed the drywell bypass leakage test required by Technical

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Specification 4.6.2.2 to be used to verify the closure of the drywell isolation check valves. A recent review of testing activities for these valves has indicated that use of the drywell bypass leakage test to verify closure for the subject valve may not be adequate. Due to the fact that the valve is located in a closed loop cooling water system and is not open to the drywell atmosphere, the valve is not tested during the drywell bypass test. ,

An engineering evaluation has been performed which determined that in no credible combination of events would the non-tested ,

condition of check valve ICCP*V119 results in a degradation of drywell/ containment integrity.

The outboard motor-operated drywell isolation valve (ICCP*MOV142) t is an automatic drywell isolation valve. It has been stroke-time tested in the closed direction and confirmed operable. -

2.1.2 Alternate Testina The licensee does not propose any alternate testing. The valve will be tested at the next refueling outage which is scheduled to start in September of 1995.

2.1.3 Evaluation l The licensee is currently performing a systematic review of their inservice testing program. During this process, it was discovered on July 27, 1994, ,

that several check valves identified in Relief Request VRR-02 had not been leak-rate tested to verify closure in accordance with the requirements of the IST program during thc previous refueling outage which ended in late June of 1994. Tne eight reactor equipment drain drywell isolation valves were evaluated and reclassified in the licensee's IST program from A/C to C and Interim Relief Request IRR-02 was requested in a letter dated August 1, 1994, to request that testing of these eight valves be deferred until the next refueling outage. In addition, this prompted the licensee to conduct a closer evaluation of the remaining valves included in Relief Request VRR-02. The licensee then submitted Interim Relief Request IRR-03 in a letter dated August 2,1994, to request that the testing of drywell isolation check valve ICCP*V119 also be deferred until the next refueling outage which is currently .

scheduled for September of 1995.

Check valve ICCP*V119 is part of the component cooling water system and is physically located inside the drywell. It is a six-inch valve and has a safety function to close to isolate flow from the drywell to the containment.

Relief Request VRR-02 had stated that this valve was to be verified closed by performance of the drywell bypass leakage test. However, it is not possible for the drywell bypass leakage test to verify this valve closed.

~4-figure 9.2-2a of the River Bend Updated Safety Analysis Report shows that motor-operated valve (MOV) ICCP*MOV142 is directly upstream of check valve ICCP*V119 and located outside of the drywell. This MOV also performs a safety ,

function in the closed direction and, according to the licensee's submittal of '

August 2,1994, is automatically actuated on high drywell pressure or low reactor water level. The licensee's IST program states that this valve is a category A butterfly valve and is full-stroke exercised on a cold shutdown frequency (Note: the licensee changed this classification to category B during  ;

the evaluation of this relief request). This MOV is part of the licensee's GL  ;

89-10 program; however, testing in accordance with GL 89-10 has not been i performed to date. While the valves have not been positively verified to '

fully close, there have been no operational abnormalities which indicate that the valves would not close. However, for conservatism, the licensee has performed an engineering evaluation and determined that even if the M0V and the check valve both fail to close, and a line break results in a bypass of the drywell/ suppression pool, safety analysis limits are not exceeded. In the licensee's August 2, 1994, submittal, it is stated that " calculations show that in the unlikely event that this combination of events were to occur, the I overall contribution to post-accident drywell bypass leakage area would be approximately 0.164 square feet, well below the criterion stated in the RBS

[ River Bend Station) USAR [ Updated Safety Analysis Report] of 1.15 square feet for a small steam line break." The staff agrees with the licensee's analysis.

It would be a hardship or difficulty without a compensating increase in the level of quality and safety for the licensee to shut the plant down solely to perform closure testing of the reactor plant closed cooling water supply to recirculation pump bearing cooler drywell isolation check valve. In order for a bypass of the drywell (suppression pool) to occur, failure of the MOV and check valve, plus a pipe break upstream or down stream of the M0V must occur. l Assuming the valves cannot fully close, the probability of a concomitant pipe '

break is small. Therefore, for an interim period of time, the testing can be delayed until an outage of greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> occurs (the length of a cold l

shutdown for which OM-10 requires all cold shutdown testing to be performed) '

without impacting the safety analysis. Testing of the check valve or GL 89-10 testing of the MOV must be performed during the next cold shutdown of greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to provide assurance that at least one valve will fully close during the interim period until the refueling outage. If no shutdowns of  ;

greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> occur, at a minimum, both valves must be tested prior to ,

startup from the next refueling outage currently scheduled for September 1995.

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2.1.4 Conclusion The staff has determined that, based on the determination that compliance with the specified requirements results in a hardship or difficulty without a compensating increase in the level of quality and safety, the proposed alternative to the ASME Code requirements in Interim Relief Request IRR-03 is authorized pursuant to 10 CFR 50.55a(a)(3)(ii). The relief is applicable for an interim period from the date of this SE to the next refueling outage which is currently scheduled for September of 1995; however, if a plant shutdown of greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> occurs before the refueling outage, either the check ,

valve or the MOV must be adequately tested to demonstrate its capability to close.  ;

Principal Contributor: Joseph Colaccino l

Date: August 2, 1994 f

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