ML18106A853

LER 98-007-00:on 980730,reactor Coolant Instrument Line through-wall Leak Was Noted.Cause of Event Has Not Yet Been Determined.Assembled Root Cause Team & Replaced Affected tubing.W/980827 Ltr

ML18106A853
Person / Time
Site:	Salem
Issue date:	08/27/1998
From:	Bakken A, Manges C Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-98-007-01, LER-98-7-1, LR-N980413, NUDOCS 9809040073
Download: ML18106A853 (5)

Text

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OPS~G P~blic Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit

~UG 2 7 1998 LR-N980413 U.S. Nuclear Regulatory Commission

• Document Control Desk Washington, DC 20555

Dear Sir:

LICENSEE EVENT REPORT 98-007-00 SALEM GENERATING STATION - UNIT 2 FACILITY OPERATING LICENSE NO. DPR-75 DOCKET NO. 50-311 This Licensee Event Report entitled "Reactor Coolant Instrument Line Through-Wall Leak" is being submitted pursuant to the requirements of 10CFR50.73(a)(2)(ii).

Sincerely, A. C. Bakken Ill General Manager -

Salem Operations CEM C Distribution LER File 9809040073 980827 PDR ADOCK 05000311 S PDR The po\\*er is in your hands.

95-2168 REV. 6/94.

NRC FORM 366 U.S. NUCLEAR REGULA COMMISSION APPR BY OMB NO. 3150-0104 EXPIRES 06/30/2001 (6-1998) Estimated burden per response to comply with this mandatory information

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LICENSEE EVENT REPORT (LER) collection request: 50 hrs. Reported lessons learned are incorporated into 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 (See reverse for required number of to the Paperwork Reduction Project (3150-0104), Office of Management and Budget, Washington, DC 20503. If an information collection does not digits/characters for .each block) display a currently valid OMS control number, the NRC may not conduct or SIJ<?m;<:Jr, and a person is not required to respond to, the information FACILITY NAME (1 l DOCKET NUMBER (2) PAGE (3)

SALEM GENERATING STATION UNIT 2 05000311 1 OF 4 TITLE (4)

Reactor Coolant Instrument Line Through-Wall Leak

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

MONTH DAY YEAR YEAR SEQUENTIAL I REVISION MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NUMBER I 05000 07 30 98 98 -- 007 -- 00 08 27 98 FACILITY NAME DOCKET NUMBER 05000 OPERATING 5 II THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 1 o CFR ": (Check one or more) (111 MODE (9) 20.2201 (b) 20.2203(a)(2)(v) 50. 73(a)(2)(i) 50. 73(a)(2)(viii)

POWER LEVEL (10) 0 20.2203(a)( 1) 20.2203(a)(3)(i) x 50. 73(a)(2)(ii) 50. 73(a)(2)(x) 20.2203(a)(2)(i) 20.2203(a)(3)(ii) 50. 73(a)(2)(iii) 73.71 20.2203(a)(2)(iil 20.2203(a)(4) 50. 73(a)(2)(iv) OTHER 20.2203(a)(2)(iii) 50.36(c)(1) 50.73(a)(2)(v) Specify in Abstract below or in NRC Form 366A 20.2203(a)(2)(iv) 50.36(c)(2) 50. 73(a)(2)(vii)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER (Include Area Codel C. Manges - Senior Licensing Engineer (609) 339-3234 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO EPIX TO EPIX B AB TBG X999 N SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY Yt:AH SUBMISSION x IYES INO 10 15 98 (If yes, complete EXPECTED SUBMISSION DATE).

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

On July 29' 1998, indications of leakage through reactor coolant system (RCS) instrumentation tubing were discovered. The subject tubing is used for RCS flow indication and protection and is an ASME Code Class 2 component.

Following initial evaluation of the condition, on Ju.ly 30, 1998, the Technical Specification for ASME Code Class 2 leaks was entered, and the line was isolated. A four hour notification was made to the NRC in accordance with 10CFR50. 72 (b) (2) (i). Additional walk-downs resulted in the discovery of leakage indications on the tubing of five other RCS instrument lines and on tubing in the pressurizer liquid sample line delay coil. Small accumulations of dried boron on the outside of the tubing were the only indications of leakage. The affected tubing is Type 304 stainless steel tubing that contains reactor coolant. The affected lines are ASME Code Class 2 components and are designed to maintain the RCS pressure boundary. The failure mechanism and root cause have not yet been determined. Corrective actions include assembling a root cause team, replacing the affected tubing, sending samples of the affected tubing to Westinghouse for metallurgical analysis, and initiation of an action request to inspect Salem Unit 1 tubing during the next cold shutdown.

NRC FORM 366 (6-1998)

I; I

NRC FORM 366A (6-1998)

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

TEXT CONTINUATION FACILITY NAME 11 I DOCKET 121 LER NUMBER 161 PAGE 131 NUMBER 121 SALEM GENERATING STATION UNIT 2 05000311 YEAR I SEQUENTIAL NUMBER I REVISION NUMBER 2 OF 4 98 - 007 - 00 TEXT (If more space is required, use additional copies of NRG Form 366AJ 1171 PLANT AND SYSTEM IDENTIFICATION Westinghouse - Pressurized Water Reactor Reactor Coolant System (PWR) - EIIS Identifier {AB/TBG}*

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

IDENTIFICATION OF OCCURRENCE Event Date: July 30, 1998 CONDITIONS PRIOR TO OCCURRENCE At the time of discovery, the reactor was in Mode 5 to enable replacement of the pressurizer safety valves. Prior to the shutdown of the reactor, the plant had operated for several months following an extended outage that had lasted for more than two years. During the extended outage, the reactor coolant system (RCS) tubing and sample system (SS) tubing contained a solution of water and boric acid and dissolved oxygen. There were no structures, systems, or components that were inoperable at the beginning of the event that contributed to the event.

DESCRIPTION OF OCCURRENCE On July 29, 1998, during a walk-down of the Salem Unit 2 steam generators, indications of leakage through nearby reactor coolant system (RCS) instrument tubing were discovered. The subject tubing is used for RCS flow indication and protection and is an ASME Code Class 2 component. Following initial evaluation of the condition, on July 30, 1998, Technical Specification 3.4.11.1, Action b was entered for the ASME Code Class 2 leak, and the line was isolated. On July 30, 1998, at 1232, a four hour notification was made to the NRC in accordance with 10CFR50.72(b) (2) (i) as an event found while the reactor is shutdown that, had it been found while the reactor was in operation, would have resulted in the nuclear power plant, including its principal safety barriers, being seriously degraded.

NRC FORM 366A (6-1998)

I I

• NRC FORM 366A (6-1998)

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

TEXT CONTINUATION FACILITY NAME 111 DOCKET 121 LER NUMBER (6) PAGE (3)

NUMBER (2)

SALEM GENERATING STATION UNIT 2 05000311 YEAR I SEQUENTIAL NUMBER I REVISION NUMBER 3 OF 4 98 -- 007 - 00 TEXT (If more space is required, use additional copies of NRG Form 366AJ (17)

DESCRIPTION OF OCCURRENCE (CONTINUED)

In response to the identified condition, walk-downs were performed of selected portions of the Sample System tubing within the RCS pressure boundary and all accessible RCS tubing within the RCS pressure boundary. As a result, indications of leakage were discovered in five other RCS instrumentation lines. An indication of leakage was also observed on the pressurizer liquid sample line decay coil tubing. The affected lines were un-insulated. Walkdowns of selected insulated tubing were also performed with no indication of additional leakage.

Small accumulations of dried boron on the outside of the tubing were the only indications of leakage. The affected tubing is Type 304 stainless steel tubing that contains reactor coolant. The tubing outside diameters for the RCS instrument lines and sample system are 0.375 inches and 0.5 inches, respectively. The affected lines are ASME Code Class 2 components and are designed to maintain the RCS pressure boundary.

This event is being reported pursuant to 10CFR50. 73 (a) (2) (ii) as a condition that resulted in the nuclear power plant, including its principal safety barriers, being seriously degraded. Specifically, the through-wall leak is considered a degradation of the RCS pressure boundary.

CAUSE OF OCCURRENCE The failure mechanism of the tubing is not currently known. Samples of the tubing have been sent to Westinghouse for metallurgical analysis. A failure mechanism and cause will be determined after the results of the testing are obtained.

PREVIOUS OCCURRENCES A review of LERs issued in the last two years for Salem Generating Station did not identify any similar occurrences; however, an occurrence involving a through-wall leak on a core spray nozzle weld at Hope Creek Generating Station was reported in LER 354/97-023. The cause of the Hope Creek event was IGSCC in the Alloy-182 weld material and an error in evaluating indications from the previous ultrasonic test of the weld. The corrective actions from the Hope Creek event could not have been expected to prevent the occurrence of the Salem event due to the different materials, conditions, and circumstances.

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 GENERATING STATION UNIT 2 05000311' YEAR I SEQUENTIAL NUMBER I REVISION NUMBER 4 OF 4 98 -- 007 -- 00 TEXT (If more space is required, use additional copies of NRG Form 366AJ ( 17)

ASSESSMENT OF SAFETY CONSEQUENCES There were no actual consequences and no impact on public health and safety.

The potential safety consequences are considered minimal based on the following:

UFSAR Section 15.3.1.1 describes a rupture of small diameter piping. An RCS rupture approximately equal to a 0.375 inch diameter hole could be accommodated by a single centrifugal charging pump. With a rupture of this size, the operational level in the pressurizer would be maintained, permitting the operator to execute an orderly shutdown. The 0.375 inch instrument tubing has a wall thickness of 0.065 inches, for an internal diameter of 0.245 inches. The 0.5 inch sample tubing has a wall thickness of 0.065 inches, for an internal diameter of 0.370 inches. The area from a break in the 0.375 inch or 0.5 inch tubing is below the 0.375 inch diameter analyzed in Chapter 15 of the UFSAR and would therefore be well within the capability of a single centrifugal charging pump. The affected tubing lines are capable of being isolated from the RCS when the plant is shutdown.

The RCS flow instruments are designed with a common high pressure tap and individual low pressure taps. A failure of a high pressure sensing line would cause all three flow channels to fail low, conservatively actuating the reactor protection system. A failure of a low pressure tap would cause the affected channel to fail high, leaving two channels available for the reactor protection function. The sample line is designed for providing reactor coolant samples for analyzing pressurizer chemistry. There is no safety function associated with this line.

CORRECTIVE ACTIONS

1. A root cause team was assembled to evaluate the condition.

2. The tubing sections with identified leaks discovered during the extent of condition inspections of Unit 2 were replaced during the outage utilizing the original configurations and materials. These actions restored the structural integrity of the ASME Class 2 pressure boundary and allowed Technical Specification Action Statement 3.4.11.1 to be exited.

3. An action request was initiated to inspect selected Salem Unit 1 RCS and sample system tubing within the RCS pressure boundary for evidence of leaks during the next cold shutdown.

4. Samples of the removed tubing were sent to Westinghouse for metallurgical analysis. Further corrective action will be taken, as required, based on review of the metallurgical analysis.

NRC FORM 366A (6-1998)