ML20044E622

LER 93-006-00:on 930416,discovered That Outside CIV Not Leak Tested,Per TS 15.4.4.III.D Requirements Due to Review on 840202,recommending That Valve Be Tested During RHR Hydrostatic Testing.Subj Valve replaced.W/930517 Ltr

ML20044E622
Person / Time
Site:	Point Beach
Issue date:	05/17/1993
From:	Link B, Weaver D WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CON-NRC-93-064, CON-NRC-93-64 LER-93-006, LER-93-6, VPNPD-93-105, NUDOCS 9305250246
Download: ML20044E622 (7)

Text

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Wisconsin Electnc  ;

POWER COMPANY l 231 W fActugon. DO. Box 2046 thouLee Wi S3201 {4MJ 221-2345 VPNPD-93-105 10 CFR 50.73 NRC 064 May 17, 1993 Document Control Desk U.S. NUCLEAR REGULATORY COMMISSION Mail Station P1-137 Washington, DC 20555 Gentlemen:

DOCKETS 50-266 AND 50-301 LICENSEE EVENT REPORT 93-006-00 CONTAINMENT ISOLATION VALVE NOT LEAK TEST _Fa IN ACCORDANCE WITH TECHNICAL SPECIFICATION REQUIREMENTS POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 Enclosed is Licensee Event Report 93-006-00 for Point Beach Nuclear Plant, Units 1 and 2. This report is provided in accordance with 10 CFR 50.73 (a) (2) (i) (B) , "The licensee shall report...any operation or condition prohibited by the plant's Technical Specifications."

This report describes the discovery of a containment isolation valve which was not included in a leak testing program as required by Technical Specification 15.4.4.III, " Type C Tests."

Please contact us if any further information is required.

Sincerely, w m ,

Bo Link Vice President Nuclear Power DAW /jg Enclosure L'

cc: NRC Resident Inspector NRC Regional Administrator Rann39 9305250246 930537 PDR S ADOCK 05000266 k l pyy A.sm auhnwswDuurnoi;ueam

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On April 16, 1993, with Unit 1 in a refueling shutdown condition and l Unit 2 operating at 100% power, we discovered that outside Containment i Isolation Valve CV-00369A on Penetration P-10 for both units was not  !

included in our 10 CFR 50, Appendix J, " Primary Reactor Containment J

Leakage Testing for Water-Cooled Power Reactors," Type C leakage l 3

testing program as required by Technical Specification 15.4.4.III.D.  !

1 Redundant outside Containment Isolation Valve 1CV-00369B for PBNP i Unit 1 was also found with excessive leakage. Subsequent inspection I revealed that Check Valve 1CV-00369B had a 1/16 inch hole machined into i

, its disk (as specified by valve model design). The hole in the disk was '

determined to be the cause for the leakage previously identified. Further j investigation revealed that the wrong model valve was installed as part of a modification performed in 1972. The model number for the installed .

valve was virtually identical to the model number of the check valve ,

designed for the modification except for one character. Valve ICV-00369B  !

vas replaced and a test connection was added to allow leak testing of Valve ICV-00369A. Penetration 10 (encompassing Valves CV-00369A&B) also has inside containment automatic Isolation Valve CV-00371A which is  ;

Appendix J tested. A modification will be performed during the next ,

scheduled Unit 2 refueling outage to add a test connection to allow 1 Appendix J testing of Valve 2CV-00369A. Valve 2CV-00369B was inspected l

and it appears that the proper va'.ve was installed. However, this will ,

3' be verified during the next scheduled Unit 2 refueling outage. The NRC ]

Resident Inspector was notified of this event.  !

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EVENT DESCRIPTION (Please refer to the attached figure)

On April 6, 1993, Maintenance personnel initiated Maintenance Work Request (MWR) 931656 to investigate boric acid build-up on the socket weld between '

the pipe and the outboard side of Containment Isolation Valve 1CV-00369A.- 1 Upon inspection on April 15, 1993, workers found that a crack in the I socket weld was responsible for the boric acid build-up. Workers  ;

performing this MWR also recalled a previous MWR, initiated on April-16,  !

1992, which performed a special test to determine the leak rate through i Containment Isolation Valves ICV-00369A and 1CV-00369B. That test i revealed through-leakage of 10.5 gpm on Check Valve 1CV-00369B. At that? I time, Valve 1CV-00369B was not correctly identified as a redundant outside r containment isolation valve, and no further action was taken.

l As the workers inspected the cracked weld on Valve 1CV-00369A, they

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decided to also inspect Valve ICV-00369B (due to its close proximity to Valve 1CV-00369A) to determine the cause of the through-leakage detected t

.on April 16, 1992. The workers discovered that Check Valve 1CV-00369B'had a 1/16 inch hole machined into its disk (as specified by valve'model- i design). The hole in the disk was determined to be the cause for the  ;

leakage previously identified. Subsequent investigation revealed that the  :

wrong model valve was installed as part of a modification performed in- ,

1972. The model number for the installed valve was virtually identical to:  ;

the model number of the check valve designed for the modification except for a one character difference.

. Containment Isolation Valves 1CV-0.0369A and 1CV-00369B are isolation, {

valve's associated with Containment Penetration P-10 (Refer to FSAR ,, ,

Figure 5.2-10, attached). This penetration configuration is identical l for PBNP Units 1 and 2.

Penetration P-10 is classified as a Class 1 penetration as defined in.FSAR I Section 5.2, " Containment Isolation System." This class of penetration, normally operating outgoing lines connected to the reactor coolant system, l

is required to have at least one automatically operated trip valve and one-manual valve in series located outside containment. Normally closed 1 manually operated valves which are locked closed or under administrative  :

control during power operation qualify as automatic isolation valves. >

For Penetration P-10, Valve CV-00369A is normally locked closed which  !

qualifies it as an automatic trip valve in accordance with Note 3 in FSAR [

Section 5.2.2, " System Design." Therefore, Valve CV-00369A should be leak '

tested in accordance with Technical Specification Section 15.4.4.III,

" Type C Tests," Specification A.3.a, and 10 CFR 50, Appendix J, " Primary }

Reactor Containment Leakage Testing for Water-cooled Power Reactors," '

(hereinafter referred to as " Appendix J") requirements. Valve CV-00369B ,

is not required to be leak tested. '

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0 10 013 i rcxisso -. <. -*uc.wsw.amm The original configuration of the Penetration P-10 did not include automatic Containment Isolation Valve CV-00371A inside containment.

This valve was added as a post-TMI upgrade in response to NUREG-0737.

Plant records were reviewed to determine the leak rate history on Valve ICV-00369B. However, since this valve is not required to be Type C leak tested, leak rate history could not be determined.

We also discovered during this review that Containment Isolation Valve CV-00369A on both Units 1 and 2 was also not being Appendix J, Type C tested.

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The lack of proper Appendix J testing of Valve CV-00369A was reviewed for operability concerns. There is no operability concern for the following reasons:

1) This valve is locked closed during power operations (duplicating the effect of an automatic valve with a trip signal). Penetration P-10 also has an Appendix J tested automatic isolation valve, CV-00371A, inside containment.

2) While not individually Appendix J tested, Valve CV-00369A is closed and serves as part of the Appendix J test pressure boundary during the test of outside Containment Isolation Valve CV-00371. In addition, this branch of Penetration P-10 connects to the RHR system which is a closed system outside containment and serves as a containment boundary.

Our immediate corrective action was to replace Check Valve 1CV-00369B.

This~ modification also included installation of a test connection to allow Valve 1CV.-00369A to be Appendix J, Type C leak tested. The modification was performed and Valve 1CV-00369B was replaced with a new valve with a solid disk. A modification will be performed during U2R19 to add a test connection to allow Appendix J leak testing of the corresponding Unit 2 Valve 2CV-00369A.

Valve 1CV-00369A was subsequently leak tested with acceptable results.

Valve 2CV-00369B was inspected and it appears that the proper valve was installed. However, this will be verified during U2R19.

We evaluated the feasibility of classifying Check Valve CV-00369B as an automatic valve for this penetration. This would have allowed the opening of Valve CV-00369A when above 200*F. To allow the proposed re-classification, Valve ICV-00369B was required to pass a leak test in accordance with Appendix J. A leak test was performed. However, Valve 1CV-00369B leaked in excess of PBNP administrative limits (but less than Technical Specification limits). Since the Residual Heat Removal (RHR) system connection to the letdown line is not required to be open at

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Tcxv a , - . - e. -w ==:s.- an.e m m RCS temperatures greater than 200*F, we decided that the proposed change to re-classify Check Valve 1CV-00369B as an automatic valve for this penetration was not desirable. Therefore, locked-closed manual Valves 1&2CV-00369A will continue to be classified as the automatic valves outside containment for this branch of Penetration P-10.

Operating procedures were reviewed to determine if changes were needed to restrict the opening of Valve 1CV-00369A above 200*F RCS temperature. We determined that procedural guidance was adequate for cooldown operations '

but additional guidance was needed for heat-up operations. The necessary changes were incorporated prior to heat-up of Unit 1. Changes will also t be made to the Unit 2 procedures prior to the shutdown of Unit 2.

EOUIPMENT DESCRIPTION See FSAR Figure.5.2-10 (Attached).

i CAUSE f i

-On June 25, 1982, the NRC transmitted License Amendment Numbers 61 (Unit 1) and 66 (Unit 2) which provided primary containment integrated- i leak rate test requirements and schedules consistent in part to the '

requirements of Appendix J to 10 CFR Part 50. The accompanying Safety Evaluation Report (SER) states (in part):

" Periodic hydrostatic testing of the RHR system is an adequate substir.ute for the pneumatic (Type C) testing required by i Appendix J because the hydrostatic testing is utilized to ensure

.that the isolation valves are not relied upon to prevent the >

post-accident escape of containment air. Appendix J does not .

require further air (Type C) testing of these valves; therefore, an exemption from the requirements of Appendix J is acceptable."

On February 2, 1984, a review of all containment penetrations was conducted to determine which valves should be classified as Appendix J ,

containment isolation valves. The raview recommended that Valve CV-00369A '

be tested during RHR hydrostatic testing and not be Appendix J leak '

tested. This testing scheme was consistent with the above SER, and -

therefore deemed appropriate. However, a section of piping immediately I upstream of Valve CV-00369B is designated non-QA and non-seismic Class 1

• and does not qualify as a closed system. Therefore, Valve CV-00369A  !

should have been Appendix J leak tested in accordance with Technical Specification 15.4.4.III. l t

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0 l0 0l5 OF Dj 6 CORRECTIVE ACTIONS A. Immediate: ,

1. Plant records were reviewed to determine the history of leak rate tests performed on Valve 1CV-00369B

2. The lack of proper 10 CFR 50, Appendix J testing of Valve CV-00369A on both units was reviewed for operability concerns. No operability concerns were identified.

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B. Short Term:

1. Valve 1CV-00369B was replaced with a new valve with a solid disk and was tested. A test connection was also installed to allow Valve 1CV-00369A to be 10 CFR 50, Appendix J tested. .

2. A modification will be performed during U2R19 to add a test connection to allow 10 CFR 50, Appendix J testing I of Valve 2CV-00369A. l i

3. Valve 2CV-00369B was inspected and it appears that the proper I valve was installed. However, this will be verified during U2R19.

4. Operating procedures were reviewed to determine if 1

_ changes were needed to restrict the opening of i Valve CV-00369A above 200*F RCS temperature. It was .. !

determined that additional guidance was needed for l heat-up operations. The necessary changes were ,

incorporated prior to heat-up of Unit 1. Changes will  :

also be made to the Unit 2 procedures prior to the j shutdown of Unit 2. I j

C. Long Term: .

I

1. A review of containment penetrations and associated containment isolation valves is being conducted to determine appropriate testing requirements. This )

review will be completed by July 15, 1993. j I

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• e,-- % aer.a w crn REPORTABILITY This event is being reported under the requirements of 10 CFR 50.73(a)

(2) (i) (B) , "The licensee shall report...any operation or condition prohibited by the plant's Technical Specifications." The NRC Resident

. Inspector was also notified.

SAFETY ASSESSMENT The safety of the plant, and the health and safety of the public and plant employees, were not jeopardized by this event. Although not Appendix J leak tested, Valve CV-00369A was hydrostatically tested annually as part of Inservice Test Procedures IT-530 (Unit 1) and IT-535 (Unit 2), " Leakage Reduction and Preventive Maintenance Program Test of the Residual Heat Removal System." This testing method was accepted by the NRC as documented in their Safety Evaluation Report dated June 25, 1982.

Valve CV-00369A also serves as part of the test pressure boundary during the Appendix J test of outside Containment Iselation Valve CV-00371. This branch of Penetration P-10 connects to the RHR system which is a closed system outside containment and serves as a containment boundary. Any leakage past Valve CV-00369A would subsequently be contained in the RHR system.

GENERIC IMPLICATIONS No generic implications have been identified.

SIMILAR OCCURRENCES No similar. occurrences were identified.

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ISDLATIDN TEMP. i VALVES ,

LINE HDT)200 PENETRATION INSIDE CUTSIDE- PRANCH/ SYSTEM SIZE _ FLUID COL D<200 CLASS i'

10 CV-371A CV-37I LETDDVN LINE/RCS 2' V HDT I

. CV-204A .

. CV-371A CV-269A RHR PUMP DISCHARGE 2' V HDT I CV-3693 TD LETDDVN LINE/RHR FCR FURTHER INFORMATION REFER TD FSAR CHAPTER 9 L FIG. 92-L 92-2 . .

NDTE'

'1 D LETDDVN LINE ERANCH- THIS BRANCH MEETS CLASS I CONTAINMENT ISDLATIDN CRITERIA VITH AN AUTDMATIC TRIP VALVE (CV-37D AND A MANUAL VALVE (CV-204A> CDNNECTED IN SERIES DUTSIDE CDNTAINMENT TO A REMDTE OPERATED VALVE (CV-371A> INSIDE CDNTAINMENT. THIS ,

LINE IS CDNNECTED TO THE REACTOR CDDLANT SYSTEM. CVI-37IA VAS ADDED AS A TMI  !

COMMITMENT.

2) RHR PUMP DISCHARGE TD LETDDVN LINE- THIS BRANCH MEETS CLASS I CDNTAINMENT ISDLATION CRITERIA VITH A LDCKED CLDSED MANUAL VALVE (CV-369A) VHICH CUALIFIES AS AN AUTDMATIC TRIP VALVE PER NDTE 3 IN FSAR SECTIDN 5.22 AND CHECK VALVE CCV-3693) VHICH FULFILLS THE REDUNDANT ISDLATION CRITERIA DF CLASS 1. BDTH ARE LDCATED DUTSIDE CDNTAINMENT.

REMDTE DPERATED VALVE (CV-371A> IS LDCATED INSIDE CDNTAINMENT AND VAS ADDED AS A TMI +

CDMMIT MENT.

TIG. 5.2-10 ,

June 1992 i

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