ML18107A274

LER 99-002-00:on 990405,determined That Containment Isolation Valve Failed as Found Leakrate Test.Caused by Foreign Matl Blocking Valves from Closing.Check Valve Mechanically Agitated.With 990504 Ltr

ML18107A274
Person / Time
Site:	Salem
Issue date:	05/03/1999
From:	Garchow D, Nagle J Public Service Enterprise Group
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
LER-99-002-01, LER-99-2-1, LR-N990220, NUDOCS 9905110241
Download: ML18107A274 (4)

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IAY '1999 LR-N990220 Regional Administrator U.S. Nuclear Regulatory Commission Region 1 475 Allendale Road King of Prussia, PA 19406-1415 Gentlemen:

LICENSEE EVENT REPORT 311/99-002-00 SALEM GENERATING STATION-UNIT 2 FACILITY OPERATING LICENSE NO DPR 75 DOCKET NO. 50-311 This Licensee J:vent Report entitled CONTAINMENT ISOLATION VALVE FAILS LLRT is being submitted in accordance with the requirements of 10CFR 50.73(a)(2)(ii) which states that Licensees shall report: "Any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded;"

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Vlt>iv David F. Garct\ow General Manager-Salem Operations Attachment C U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

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NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 06/30/2001

~ (6-1998) Estimated burden per response to comply with this mandatory information collection request 50 hrs. Reported lessons learned are incorporated into the LICENSEE EVENT REPORT (LER) 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 Paperwork (See reverse for required number of Reduction Project (3150-0104), Office of Management and Budget, Washington, DC 20503. If an information collection does not display a airrentty digits/characters for each block) valid OMB control number, the NRC may not condud or sponsor, and a person is not required to respond to, the .information collection.

FACILITY NAME 111 DOCKET NUMBER 121 PAGE 131 SALEM GENERATING STATION UNIT 2 05000311 1 OF 3 TITLE (41 CONTAINMENT ISOLATION VALVE FAILS LLRT- DEGRADED CONTAINMENT INTEGRITY EVENT DATE 151 LEA NUMBER 161 REPORT DATE 171 OTHER FACILITIES INVOLVED 181 MONTH DAY YEAR YEAR I SEQUENTIAL NUMBER I REVISION NUMBER MONTH DAY YEAR FACILITY NAME DOCKET NUMBER 05000 FACILITY NAME DOCKET NUMBER 04 05 99 99 002 00 05 03 99 05000 I OPERATING MODE 191 I 5 I Tl-'*~ DCDnDT 110* .,,

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NAME TELEPHONE NUMBER (Include Area Codel John C. Naale Senior Licensina Enaineer 609-339-3171 P&H IO~P

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REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER TO EPIX CAUSE SYSTEM COMPONENT MANUFACTURER TO EPIX B LF ISV V085 N B LF ISV V085e N

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(If I NO EXPECTED SUBMISSION DATE 1151 06 30 99 ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (161 During performance of Type C local leak rate testing on April 5, 1999 it was determined that 2SA118 (containment service air outboard manua.l isolation valve) failed the as found leakrate test. Because this failure was the second valve for this penetration to fail (2SA-119 inboard check had also failed) containment integrity was considered to be degraded. At the time of discovery the Unit was shut down for refueling and containment integrity was not required. A 4-hour report was made to the NRC as required by the plant's Emergency Classification Guide and 10CFR50.72(b) (2) (i). These valves are on the service air supply to the containment and are normally not open except during periods when maintenance activities are being performed within containment. The cause of the leakage appears to be foreign material

.blocking the valves from closing. The manual valve was cycled several times and the leakage returned to measurable levels. The check valve was mechanically agitated and returned to the normal closed position. Both valves will be opened and inspected in order to determine failure cause prior to entry into mode 4 during restart from the current refueling outage.

Further corrective actions may be identified upon inspection. This report is being made pursuant to 10CFRS0.73 (a) (2) (ii) Licensees shall report: "Any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded; .. "

NRC FORM 36616-19981

NRC FORM 366A (6-1998)

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

TEXT CONTINUATION DOCKET 121 FACILITY NAME 111 NUMBER 121 LER NUMBER 161 PAGE 131 YEAR l SEQUENTIAL NUMBER l REVISION NUMBER Salem Generating Station Unit 2 05000311 99 0 02 00 2 OF 3 TEXT (If more space is required, use additional copies of NRC Form 366AJ 1171 PLANT AND SYSTEM IDENTIFICATION

?*.

Westinghouse - Pressurized Water Reactor Service air system/Isolation Valve {LF/ISV}*

• Energy Industry Identification System (EIIS) codes and component function identifier codes appear as {SS/CCC} in the text.

CONDITIONS PRIOR TO OCCURRENCE The unit was in cold shutdown in preparation for refueling prior to the event.

DESCRIPTION OF OCCURRENCE During performance of the local leak rate test, containment service air manual isolation valve 2SA118 failed. This valve is a three inch gate valve manufactured by Velan Corp. Personnel were unable to measure test pressure because of a leakrate which exceeded the capability of the available test equipment. Based upon the inability to establish test pressure the leakrate was estimated to be greater than 100,000 seem. This failure constituted the second gross leakrate valve failure on the station air penetration. The three inch Velan swing check on this penetration alsa failed the as found (Type C)test. Therefore, the as-found Type C leakage for the valves on this penetration caused the total Type B and C Technical Specification leakage limit to be exceeded. T.S. Section 6.8.4.f, requires Type B and C leakage to be less than or equal to 0.6 La.

• In addition, T.S Section 3.6.1.2 b. requires all Type B and C leakage rates to be in accordance with the containment leakage rate testing program in modes 1, 2, 3 and 4. This program requires that the leakage be below 0.6 La whenever containment integrity is required. At the time of discovery the ~

plant was in a mode where containment integrity was not required.

A 4-hour report was made to the NRC as required by the plant's Emergency Classification Guide and 10CFR50.72(b) (2) (ii), which requires reporting when one of the primary barriers is seriously degraded. The leakage was indeterminate because the rate exceeded that capabilities of the test equipment therefore it must be assumed that the leakage exceeded requirements. 1 NRC FORM 366A (6-19981

\~ NRC FORM 366A (6-19981 U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION DOCKET 121 FACILITY NAME 111 NUMBER 121 LER NUMBER 161 PAGE 131 YEAR 1 SEQUENTIAL NUMBER l REVISION NUMBER SALEM GENERATING STATION UNIT 2 05000311 99 0 02 00 3 OF 3 TEXT (If more space is required, use additional copies of NRC Form 366AJ 117)

CAUSE OF OCCURRENCE The event investigation can not be completed until the valve is opened and inspected. This work will be completed prior to entry into mode 4 upon return to service from the current refueling outage.

It is believed that foreign material, such as rust, may have been deposited in the seating area of the valve.

PRIOR SIMILAR OCCURRENCES A review of 1997 and 1998 Licensee Event Reports and Inspection Reports for Salem Units 1 and 2 has identified no similar incidences on barrier degradation. The 2SA118 valve has had several failures over the last nine years.

SAFETY CONSEQUENCES AND IMPLICATIONS Excess containment leakage results in an increase in the calculated post accident doses to the control room personnel and to the public. These dose calculations are performed assuming .a leakage rate of l.OLa, which is greater than the Tech Spec acceptance criteria of 0.6 La. However, the leakage for this penetration exceeded the capabilities of the test equipment therefore it is not possible to determine if the total leakage was less than La. The last record of successful testing for this penetration was August of 1998. This date represents the maximum period of time that this condition may have existed. These valves were opened for subsequent in-containment activities and it is assumed that the condition occurred at that time. A review of the tagging database reveals that the date that this valve was operated since the successful test was in December 1998.

CORRECTIVE ACTIONS The 2SA118 valve was cycled several times and the seat area was air blown in order to displace the foreign material. Re-testing determined that there was a measured leakage of approximately 6,500 seem which satisfies the Tech Spec Limit.

A review is being conduct~d to determine the feasibility of revising the testing program in order to require as left testing any time the SA118 valve is operated.

The 2SA118/119 valves will be disassembled, inspected, cleaned, repaired and re-tested prior to entry into mode 4 at which time further corrective actions may be identified.

NRC FORM 366A 16-19981