ML18106A814

LER 98-010-00:on 980714,determined That Leakage from Boron Injection Tank Exceeded Max Allowable ECCS Leakage from Sources Outside Containment.Caused by Leaking 2SJ404 Manual Sample valve.2SJ404 Valve repaired.W/980813 Ltr

ML18106A814
Person / Time
Site:	Salem
Issue date:	08/13/1998
From:	Bakken A, Duca P Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-98-010-01, LER-98-10-1, LR-N980397, NUDOCS 9808210157
Download: ML18106A814 (5)

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• Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nucle'ar Business Unit AUG 13 1998 LR-N980397 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:

LER 311/98-010-00 SALEM GENERATING STATION - UNIT 2 FACILITY OPERATING LICENSE NO. DPR-75 DOCKET NO. 50-311 This Licensee Event Report entitled "ECCS Leakage Outside Design Basis Value" is being submitted pursuant to the requirements of the Code of Federal Regulations

• • • • 1 OCFR50.73 (a)(2)(ii)****.

Sincerely, A. C. Bakken Ill General Manager Salem Operations Attachment PJD/

c Distribution LER File 3.7 9808210157 980813 PDR ADOCK 05000311

• S PDR The power is in your hands.

95-2168 REV. 6/94

NRC FORM 366 U.S. NUCLEAR REGULATOR MISSION APPROVED BY 0. 3150-0104 EXPIRES 06/30/2001 (6-1998) Estimated burden p ponse to comply with this mandatory information collection request: 50 hrs. Reported lessons learned are incorporated into LICENSEE EVENT REPORT (LER) the licensing process and fed back to industry. Forward comments regarding burden estimate to the Records Management Branch (T-6 F33), U.S.

Nuclear Regulatory Commission, Washington, DC 20555-0001, and to the (See reverse for required number of Paperwork Reduction Project (3150-0104), Office of Management and digits/characters for each block) Budget, Washington, DC 20503. If an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

FACILITY NAME (1)

SALEM UNIT 2 DOCKET NUMBER (2) 05000311 !PAGET OF 4 TITLE(4)

ECCS LEAKAGE OUTSIDE DESIGN BASIS VALUE.

EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NUMBER 07 14 98 98 -010 00 08 13 98 FACILITY NAME DOCKET NUMBER Salem Unit 1 05000272 OPERATING 1 100% x NAME TELEPHONE NUMBER (Include Area Code)

Philip J. Duca Jr. , Salem Licensing Engineer (609) 339-2381 COMPLETE AILURE DESC BED IN THIS REPORT (13)

CAUSE SYSTEM COMPONE MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE NT TO EPIX TO EPIX EXPECTED MONTH DAY YEAR YES (If yes, complete EXPECTED SUBMISSION DATE).

ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)

On 7 /14/98 Operations personnel determined that leakage f.rom the Boron Injection Tank exceeded the maximum allowable ECCS leakage from sources outside Containment.

The leakage path was determined to be past the valve seat of a manual sample valve (2SJ404) from the Boron Injection Tank. The valve was subsequently isolated. With the valve isolated, ECCS leakage returned to within UFSAR limits.

The cause of the event was the leaking 2SJ404 manual sample valve.

During the period of leakage, there were no actual safety consequences to the health and safety of the public or to the plant staff. Estimated potential post accident offsite doses remain well within 10CFR100 limits. Estimated post accident onsite beta skin dose is within the General Design Criterion 19 (GDC 19) limit, while estimated whole body and thyroid doses are slightly above the GDC 19 limits based on conservative estimates. Estimated core damage frequency during the period of the leakage, at approximately 3.7xE-6, decreases potential significance.

This event is reportable pursuant to 10CFR50.73(a)2(ii) "any event or.

condi tion ............... that resulted in the nuclear power plant being:...... (B) in a condition that was outside the design basis of the plant".

. NRC FORM 366A U.S. NUCLEAR REG ATORY COMMISSION (6-1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET (2) LER NUMBER (6) PAGE (3)

NUMBER(2)

SALEM UNIT 2 05000311 YEAR I SEQUENTIAL NUMBER IREVISION NUMBER 2 0F 4 98 0 1 0 00 TEXT (If more space is required, use additional copies ofNRC Form 366A) (17)

PLANT AND SYSTEM* IDENTIFICATION Westinghouse - Pressurized Water Reactor High-Pressure Safety Injection System/Sample Valve {BQ/SMV}*

• Energy Industry ~dentification System {EIIS} codes and component function identifier codes appear as (SS/CCC)

CONDITIONS PRIOR TO OCCURRENCE At the time of identification, Salem Unit 2 was operating at 100% Power.

DESCRIPTION OF OCCURRENCE On 7/14/98 Operations personnel determined that leakage from the Boron Injection Tank exceeded the maximum allowable ECCS Leakage as defined by UFSAR Section 6.3.2.11. The leakage was determined to be approximately 0.25 gpm (60,000cc/hour) past the valve seat of a manual sample valve (2SJ404) from the Boron Injection Tank. The leakage was flowing into the RHR sump in the Auxiliary Building, outside of Containment. UFSAR Section 6.3.2.11 addresses leakage during the recirculation phase of an accident, and states that "The total leakage resulting from all sources is about 3800 cc/hour as described in UFSAR Section 15.4.1. Recirculation loop leakage sources are summarized in Table 6.3-12. Leakage is monitored by procedure to ensure this leak rate is not exceeded." Leakage in excess of 3800 cc/hour placed the plant in a condition outside that assumed for design basis accident conditions.

For about two weeks prior to the event, efforts were being focused on reducing an elevated reactor coolant system (RCS) unidentified leakage.

Initially this leakage was thought to be within containment. The continuing investigations included determining the source of increasing in-leakage to the RHR sump located in the Auxiliary Building.

These investigations used station procedures. One of these procedures provides a program for monitoring leakage from systems outside Containment that could contain highly radioactive fluids following an accident. The objective of the program is to detect and correct any degradation of the pressure boundaries for these systems and thereby reduce post-accident dose rates and airborne activity.in the Auxiliary Building in accordance with Technical Specification 6.8.4.a.

NRC FORM 366A (6-1998)

I~

• • ~========~*==========~*~====~

NRC FORM 366A (6-1998)

U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET (2) LER NUMBER (6) PAGE (3)

NUMBER(2)

SALEM UNIT 2 05000311 YEAR ISEQUENTIAL NUMBER IREVISION NUMBER 3 0F 4 98 0 1 0 00 TEXT (If more space is required, use additional copies ofNRCForm 366A) (17)

Following identification that 2SJ404 was the source of the leakage, 2SJ8 was closed, isolating 2SJ404. Valve 2SJ8 is the root valve in series with 2SJ404. Subsequent to closing 2SJ8, leakage into the RHR sump stopped. A follow-up RCS leak rate confirmed that the unidentified leakage was significantly reduced.

CAUSE OF OCCURRENCE The cause of the event was the leaking 2SJ404 manual sample valve.

PRIOR SIMILAR OCCURRENCES 1996, 1997 ano 1998 LERs were reviewed for similar occurrences. No similar events were identified.

SAFETY CONSEQUENCES AND IMPLICATIONS The activity level of the reactor coolant was slightly elevated due to a suspected single fuel pin leak. However, there was no appreciable change to radiological conditions in the area of the 2SJ404 due to the leakage through the valve. Routine effluent monitoring and dose assessment for the period showed the dose to be well within technical specification limits.

Therefore, during the period of leakage, there were no safety consequences to the health and safety of the public or to the plant staff.

In an effort to assess the potential safety consequences, dose estimates were conducted. The estimates were performed using the actual measured leakage through 2SJ404 (0.25 gpm). These estimates showed offsite whole body accident dose to be comparable with the plant design basis for both the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ) and well within 10CFR Part 100 limits. Offsite accident thyroid doses for the EAB and LPZ while greater than the design basis, were well within the Part 100 limits.

Potential onsite beta skin dose was estimated to remain within the General Design Criterion (GDC) 19 limit. However, whole body and thyroid doses, assuming a source term (the Salem design basis accident source) per the guidance of Regulatory Guide 1.4, are estimated to be slightly above the GDC 19 limits.

The potential consequences of the event depend on the core damage frequency, which for Salem is 4.45xE-5/year. While it is impossible to exactly

NRC FORM 366A (6-1998)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET (2) LER NUMBER (6) PAGE (3)

NUMBER(2)

SALEM UNIT 2 05000311 YEAR I SE~~ 1=~~ 4 OF 4 98 0 1 0 00 TEXT (If more space is required, use additional copies ofNRC Form 366A) (17) identify a time when the 2SJ404 leakage became a significant component of the RCS unidentified leakage, formal troubleshooting to identify the leakage began two weeks prior to identification of *the leak. If this period is increased to a month, the core damage frequency for the period is approximately 3.7xE-6.

CORRECTIVE ACTIONS

1. The leakage flow path was isolated restoring ECCS leakage from systems outside Containment to within UFSAR limits.

2. A work order has been initiated to inspect and repair the 2SJ404 valve.

3. Operator training will be provided on this event during an upcoming requalification cycle. This training will include discussion of the design basis leakage criteria for primary coolant sources outside containment. 1

4. The program to reduce leakage from primary coolant sources outside containment (required by Technical Specification 6.8.4.a) will be reviewed and appropriate revisions will be made~

NRC FORM 366A (6-1998)