ML20056B441

Environ Assessment & Finding of No Significant Impact Re Exemption from 10CFR50,App J Requirement That Upon Two Consecutive Failures of App J Containment Type a Tests,Test Frequency Accelerated

ML20056B441
Person / Time
Site:	Sequoyah
Issue date:	08/21/1990
From:	Hebdon F Office of Nuclear Reactor Regulation
To:
Shared Package
ML20056B442	List: ML20056B441
References
NUDOCS 9008280292
Download: ML20056B441 (7)

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UNITED STATES HUCLEAR REGULA10RY COMMISSION j

TENNESSEE VALLEY AUTHORI_TY D_0,C,K,ET NO. 50-328 ENVIRONMENTAL ASSESSMENT AND FINDING OF NO SIGNIFICANT IMPACT The U.S. Nuclear Regulatory Commission (the Commission) is considering.

issuance of an exeniption from the requirements of Section III.A.6(b) of Appendix J to.10 CFR Part 50 to the Tennessee Valley Authority (the licensee) for the Sequoyah Nuclear P.lant, Unit 2.

The unit is located at the licensee's site in Familton County, Tennessee. The exemption was requested by the

-licensee in its. letter dated May 21, 1990.

ENVIRONitENTAL ASSESSMENT Iden,tj,fj,ca, tion, of Proposed Action:

The exemption would allow the-licensee relief from the provisions in:

Section III.A.6(b) of Appendix J with respect to the' requirement that upon two consecutive failures of Appendix J containment Type A tests there is an acceleration of the test frequency. If two consecutive Type A tests fail to 9

neet the acceptance criteria of 0.75La, a Type A ; test shall be performed at each refueling outage until two consecutive Type A tests meet the acceptance-l criteria. After this, the test frequency in Section III.D'of Appendix J, which is: performing three Type A tests at approximately equal intervals during e

each 10-year service-period, may resume. The relief would relax the t.cceleration of the Type A test frequency and the requirement to conduct a Type A test et Unit 2 in the Unit 2 Cycle 4 refueling outage scheduled for the

-fall-of 1990.

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At Unit 2, the licensee _ conducted Type A tests during the preoperational testing in 1981, the Unit 2 Cycle 2 refueling outage in November 1984,_ and the Unit 2 Cycle 3 refueling outage in March 1989.

Had the last two Type A tests-not been classified as failures, the next Type A test at Unit 2 would be conducted in the Unit 2 Cycle 5 refueling outage in 1992 to complete the three tests in a 10. year service period. With two consecutive failures, the licensee is required-to conduct a Type A test in each refueling outage until the unit passes two consecutive Type A tests. The first refueling outage that would be f

affected is the Unit 2 Cycle 4 refueling outage. The Unit 2 Cycle 5 refueling-outage is not affected by this relief because this outage is scheduled for the third Type A~ test of the 10-year service period-The history of the Type A tests conducted at Unit 2 is noted below:

Type A Tests As-found 0.75La 1.0La Status performed leak rate limit

. limit

(% per day)

(% per day)

(%perday) preoperational test (1981) 0.14 0.1875 0.25 pass Test 1 (1984) 0.22 0.1875 0.25-failure Test 2 (1989) 0.20 0.1875 0.251 failure The last two Type A-test results exceeded the acceptable-limit of 0.75La required by Appendix J but did not exceed the maximum allowable ratelof La.

I La is' the leakage rate assumed for the containment during a loss-of-coolant accident. The licensee stated that the root cause of the Cycle 2 Type A test failure was determined to be packing leakage from two outboard root valves on two containcent pressure sensing lines. The licensee performed maintenance on the pressure sensing lines during Cycle 2 refueling outage and repaired the root valves which resulted in an immediate reduction in the measured leak rate i

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e to'below the acceptance criteria. The licensee also implemented corrective actions to prevent the pressure sensing line leakage. These actions include:

(1). Programmatic review of the instrument maintenance and opration-activities to. identify potential impacts on containment integrity, and i

(2) Expansion of the local leak rate test (LLRT) program to require an LLRT following any maintenance performed on the pressure sensing lines. Post-maintenance leak rate _ testing is required and added to the Surveillance Instruction (SI)- 159.1, " Leak Rate Test on Containment Pressure Instruraentation. "

The primary cause of-the Cycle 3 Type A failure was due to excessive leakage through Penetration X-59.

The root cause was personnel error in connecting the hose from the test equipment to the test connection for the valves associated with Penetration X-59.

Another factor that contributed to the excessive leakage through Penetration X-59 involved a maintenance sequence

'thatoccurredwhentheoutboardcontainmentisolationvalve(FCV-67-88)was previously disessembled, cleaned, and reassembled during the outage. The t

licensee has impleraented-corrective actions for the-root causes of excessive leakage from Penetration X-59.

These actions include:

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(1) Revision of the LLRT program (SI-158.1) to include instructional steps t

that require the test hoses to be visually inspected to ensure that no restrictions or crimped conditions exist, and

,(2) Revision of the Maintenance Instructions (0-MI-HVV-000-008.0) to ensure

'T that when soft-seated butt; fly valves without internal disc stops are removed from the piping, the valve operator limits are set with the valve

' body attached to ensure that valve position is established prior to reinstallation.

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The staff has reviewed the licensee's submi:tal and agrees with the l

licensee that the root cause of each of the last wo Type A test failures was due to excessive leakage of a single component or pt etration in'the contain-ment boundary.

Even with the leakage, the Type A test results were found still within the maximum allowable leak' rate of 1.0La. The licensee has corrected and repaired the components that caused the Type A test failures and implemented 2

corrective actions to preve'nt future test failures. Additionally, the current Appendix J 1eak rate limit for Type A tests contain a 25% safety margin between the leak. rat: acceptance criteria and the leak rate assumed during the loss-

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of-coolant accident. A proposed revision to Appendix 0 currently under consideration would remove this margin. With the above corrective actions and the fact that the last two Type A test failures were below the maximum allowable leak rate of 1.0La,- the staff concludes that the requested exemption has no significant impact on containment integrity and no benefit wculd be gained by requiring the licensee to perform Type A tests on an accelereted test frequency.

' lh,e,7,c,ed,,fpor, _thp,,Prpppsed Action:

The proposed exemption is required to relieve the licensee from the requiteuent to conduct a Type A test of its Unit 2 containment in the Unit 2 Cycle 4 refueling outage scheduled for the fall of 1990.

Environtrental Impacts of the Prppposed Action:

With respect to the requested exemption, the relief from the above requirerrent would allow the licensee to avoid conducting an unnecessary Type A test at Unit 2 in the upcoming Unit 2 Cycle 4 refueling outage. The test is not needed to assure the integrity of the containment during an accident which is the purpose of the test. Consequently, neither the probability of accidents

x 5-nor.the radiological releases from accidents will be increased. _With regard to-other potential radiological environmental impacts, the proposed exemption does

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not, increase the radiological effluents from the facility and does not increase the occupational exposure at the facility. Therefore, the Commission concludes that there are no significant radiological impacts associated with the proposed exemption.

.With regard to potential nonradiological environmental impacts, the proposed exemption involves systems located within the restricted areas as defined in 10 CFR Part 20.

It does not affect nonradiological plant effluents and has no other environmental impact. Therefore, the Commission concludes that there are no significant nonradiological environmental impacts associated 1

with the proposed exemption.

Therefore, the proposed exemption does not significantly change the conclusions in-the licensee's " Final Environmental Statement Related to the Operation of Sequoyah fluclear Plant Units 1 and 2," (FES) dated February 21, 1974. The Commission concluded that operation of the Sequoyah units will not result in any environmental impacts other than those evaluated in the FES in its letter to the licensee ' dated September-15,1981 which granted the facility.

operating license DPR-79 for Unit 2.

Alternative to the Proposed Action:

Because the staff has concluded that there is no measurable environmental impact associated with the proposed exemption, any alternative to this exemption will have either no significantly different environmental' impact or greater environmental impact.

The principal alternative would be to deny the requested exemption.

'This would not reduce environmental impacts as a result of plant operations.

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41ternativeUseOfResources:

This actica does not involve the use of resources not previously considered in connection with the " Final Environmental Statement Related to the Operation of the Sequoyah Nuclear Plant, Units 1 and 2," dated February 21, 1974.

AleP.c,1,es, ag,,Pe,rsons Consulted:

The NRC staff has reviewed the licensee's request and the licensee's supplemental letters that support the proposed exempti.on. The NRC staff did not consult other agencies or persons.

.FINDIllC OF NO SIGNIFICANT IMPACT The Commission has determined not to prepare an environmental impact statu..ent for the proposed exemption, Pased upcn the foregoing environmental assessrnent, we conclude that the proposed action will not.have a significant effect on the quality of the human envircrs.ent.

q For details with respect to this action, see the licensee's request for en exer.ption dated May 21, 1990 which is available fcr public inspection at

- the Coraission's Public Document Room, Gelman Building, 2120 L Street, N.k'.,

Washington, D.C., and at the Chattanooga-Hanilton County Bicentennial Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Dated at Rockville, flaryland, this 21st day of August 1990.

FOR THE. NUCLEAR REGULATORY COMMISS10ll'

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- l Frederick J..HebCon, Director Project Directorate 11-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation L

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I-Mr. Oliver D. Kingsley, Jr.

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l Mr. Marvin Runyon, Chairman Mr. Joseph Bynum, Acting Cite Director Tennessee Valley Authority Sequoyah Nuclear Plant ET 12A 7A Tennessee Valley Authority 400 West Summit Hill Drive P. O. Box 2000 Knoxville, Tennessee 37902 Soddy Daisy, Tennessee 37379 Director Mr. Mark J. Burzynski Tennessee Valley Authority Site Licensing Manager

.ET 12A 11A Sequoyah Nuclear Plant 3

400 West Sunmit Hill Drive P. O. Box 2000 1

Knoxville, Tennessee 37902 Soddy Daisy, Tennessee 37379 Mr. John B. Waters, Director County Judge Tennessee Valley Authority Hamilton County Courthouse ET 12A 9A Chattanooga, Tennessee-37402 400 West Sumit Hill Drive i

Knoxville, Tennessee 37902 Regional' Administrator, Region II U.S. Nuclear Regulatory Commission l

Mr. W. F. Willis 101 Marietta Street, N.W.

Chief Operating Officer Atlanta, Georgia 30323 l

ET 12B 16B 400 West-Sumit Hill Drive Mr. Paul E. Harmon 1

Knoxville, Tennessee 37902 Senior Resident Inspector Sequoyah Nuclear Plant General Counsel U.S. Nuclear Regulatory Commission

' Tennessee Valley Authority 2600 Igou Ferry Road 400 West Summit Hill Drive Soddy Daisy, Tennessee 37379 ET llB 33H Knoxville, Tennessee 37902 Mr. Michael H. Mobley, Director Division of Radiological Health Mr. Dwight Nunn T.E.R.R.A. Building, 6th Floor-Vice President, Nuclear Engineering 150 9th Avenue North 1

- Tennessee. Valley Authority _

Hashville, Tennessee 37219-5404 6N 38A Lookout Place 1

1101 Market Street Tennessee Valley Authority Chattanooga, Tennessee 37402-2801 Rockville Office

.11921 Rockville Pike Dr. Mark 0. Medford Suite 402 Vice President and Nuclear Rockville, Maryland 20852 Technical Director Tennessee Valley Authority 6N 38A Lookout Place

-Chattanooga, Tennessee 37402-2801 Mr. Edward G. Wallace i

Manager, Nuclear Licensing and Regulatory Affairs Tennessee Valley Authority 1

.5N 157B Lookout Place l

Chattanooga, Tennessee - 37402-2801 i

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UNITED STATES -

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WASHINGTON, D. C. 20066

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August.8. 1990

- Docket t!o. 50-461 Mr. Frank A. Spangenberg Licensing and Safety Clinton Power Station P. O. Box 678 Mail Code V920.

e Clinton, Illinois 61727

SUBJECT:

. RESPONSE TO GENERIC LETTER 89-10. " SAFETY-RELATED MOTOR-OPERATEDVALVE(MOV)TESTINGANDSURVEILLANCE"(TACNO.75647)

On June 28, 1989, the_NRCissuedGenericLetter(GL)89-10r6questingthe establishment of a program to ensure the operability of all safety-related MOVs under design basis conditions. The program in GL 89-10 significantly expands the scope of the program outlined in NRC Bulletin 85-03 and its supplement.

o On December 29, 1989,.you submitted a response to GL 89-10, regarding the Clinton Power Station. The staff is Froviding the following comments on your i

submittal.

The GL' requested that licensees have available a description of the GL program for review by NRC personnel by June 28, 1990, or the first refueling outage

-after December-28, 1989, whichever was later.-

In your December 29 submittal, you state! that the design-basis reviews and required thrust determinations will be completed by January 1, 1991. Additionally you state that the scope and schedule for flow testing will-be available by February 28,.1991. The' staff considers the proposed completion dates for these activities to be acceptable, provided you

. prepare the-description of'the overall program by February 28, 1991.

At the L

public workshops held in September 1989 the staff indicated the information-that should be contained in the program description.- The_results of the workshops are provided in supplement 1-to the generic letter. The minutes a

are'of particular assistance in clarifying the scope of the generic letter L ~

and-the. staff's position with respect to the inadvertent operation of-motor-operatedvalves'(MOVs). You should compare'your program with the 4

staff positions on the GL in supplement 1 and should justify in your program description any differences with those positions.

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Additionally you state that torque switches on your rising stem MOVs have..seen tested to ensure that the delivered thrust meets or exceeds the valve manufac-turer's recommendations. As discussed at the workshops and demonstrated by plant experience, the recommendations of the valve manufacture have been shown in many instances to be inadequate to ensure the capability of the MOV under design-basis conditions. You should not accept the thrust calculations of the i

valve manufacturer without evaluating their accuracy.

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