ML18096A754

LER 92-008-00:on 920501,MSL Isolation Occurred on Low Temp Causing Sys to Trip.Caused by Design,Manufacturing, Construction & Installation inadequacy.On-going Assessment & Design Mod in progress.W/920528 Ltr

ML18096A754
Person / Time
Site:	Salem
Issue date:	05/28/1992
From:	Pollack M, Vondra C Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-92-008-01, LER-92-8-1, NUDOCS 9206120080
Download: ML18096A754 (5)

Text

• Public Service Electric and Gas Company P.O. Box 236 Hancocks Bri(:Jge, New Jersey 08038. Salem Generating Station u. s. Nuclear Regulatory Commission Document Control Desk Washington, DC. 20555

Dear -Sir:

SALEM GENERATING STATION LICENSE NO. DPR-75 DOCKET NO. 50-311 UNIT NO. 2 LICENSEE EVENT REPORT 92-008-00 May 28, 1992 This Licensee*Event Report is being submitted pursuant to the requirements of the Code of .Federal Reguiations lOCFR 50.73(a) (2)(iv). This report is required to be issued within thirty (30) days of event discovery.

MJP:pc .Distribution

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• ..,* '-' \._r I *b" U 9206120080 920528 PDR ADOCK 05000311 S PDR c. A.* Vondra General Manager Salem Operations 1/1)1 110M) 12-NRC FORM 366 16-89) U.S. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 3150-0104 LICENSEE EVENT REPORT (LER) EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (P-5301. U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON.

DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503. FACILITY NAME 111 l DOCKET NUMBER 121 I PAGE 131 Salem Generating Station -anit 2 o 15 I o 1 o I o I 31 11 l 1 loF a I 4 TITLE (41 ESF Main Steamline Isolation Signal -Channel Spike EVENT DATE (61 MONTH DAY a Is al1 OPERATING MODE (81 YEAR 9 2 LER NUMBER (61 REPORT DATE 171 OTHER FACILITIES INVOLVED (Bl YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACILITY NAMES DOCKET NUMBERISl NUMBER NUMBER Salem Unit 1 o 1s101010 1 21 71 2 912 -al a Is -al a Is 2js 91 2 a o1s1010101 I I THIS REPORT IS SUBMl.TTED PURSUANT TO THE OF 10 CFR §: (Chock on* or mor* of th* following)

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60.73(o)(2lliiil LICENSEE CONTACT FOR THIS LER (12.I NAME 60.73(o)(211vlll)(A) 60.73(olf211vliillBI

_60.73(olf2Hxl AREA CODE 366AI TELEPHONE NUMBER M. J. Pollack -LER Coordinator 6 1a I 9 3 13 I 91-I 2t a 12 I 2 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (131 CAUSE SYSTEM COMPONENT . I I I I I I I I MANUFAC* TUR ER I . I I I I I SUPPLEMENTAL REPORT EXPECTED (141 n YES (If yos, complete EXPECTED SUBMISSION DATE! lxl NO. (Limir ro 1400 spac6t, i.e., approximately fiftesn single-spactt rVpt1writt11n lines) 116) SYSTEM I I COMPONENT MANUFAC* TUR ER I I I 1 I I I I I I I I EXPECTED SUBMISSION DATE 1151 MONTH DAY YEAR I I I ori 5/1/92,.at 0410 hours0.00475 days <br />0.114 hours <br />6.779101e-4 weeks <br />1.56005e-4 months <br />,.a Main Steamline (MSL) isolation occurred on a low T v (< 543°F) coincident with high steamline flow signal following BF19 valve repairs (reference LER Jli/92-007-00).

The plant had entered Mode 4 at 0221 hours0.00256 days <br />0.0614 hours <br />3.654101e-4 weeks <br />8.40905e-5 months <br /> ori May 1, 1992. In Mode 4, Reactor Coolant. System Tavg ranges from 200°F to 350°F (actual temperature was approximately 250°F); therefore, the low T bistables are tripped providing half of the logic signal tor MSI .. The high stea.mline flow logic requires indication of high flow in 1 out of 2 channels per steam Generator (S/G) in 2 of the 4 S/Gs. The MSI occurred when the No. 22 Steam Generators (S/G) Steamline flow channel No. 1 and No. 24 S/G Steamline flow channel No. 1 bistables tripped. The root cause of this event is "Design, Manufacturing, Constructi,on/Installation inadequacy.

With the plant in Mode 4, condensation of steam occurs in the steamline flow reference legs resulting in channel spikes. This apparently occurred coincidently in the Nos. 22 and 24 S/G channels satisfying the logic for MSI. Assessment of this event, by Maintenance personnel, *was-that the false high steam flow signals were not caused by failed components.

The false signals cleared, on their own, after a .few hours. An in-depth study of main steamline flow instrumentation concerns was completed prior to this event. Engineering has initiated development of proposed design modifications to correct the main steamline flow sensing line concerns.

NRC Form 366 (6*891

• LICENSEE EVENT REPORT (LER) TEXT CONTINUATION

• Sal.em Generating Station Unit 2 DOCKET NUMBER 5000311 PLANT AND SYSTEM IDENTIFICATION:

Westinghouse

-* Pressurized Water Reactor LER NUMBER 92-008-00 PAGE 2 of 4 Energy Industry Identification System (EIIS) codes are identified in the text as {xx} IDENTIFICATION OF OCCURRENCE:

• Engineered Safety Feature signal actuatic:m; Main Steamline Isolation Event Date: 5/01/92 Report Date: 5/28/92 This report was initiated by Incident Report No.

CONDITIONS PRIOR TO OCCURRENCE:

Mode 4 (Hot Shutdown)

DESCRIPTION OF OCCURRENCE:

On May 1, 1992, at 0410 hours0.00475 days <br />0.114 hours <br />6.779101e-4 weeks <br />1.56005e-4 months <br />, a Main Steamline (MSL) isolation occurred on a low T (< 543°F) coincident with high steamline flow signal during following BF19 valve repairs (reference LER 311/92-007-00).

The plant had entered Mode 4 at 0221 hours0.00256 days <br />0.0614 hours <br />3.654101e-4 weeks <br />8.40905e-5 months <br /> on May 1, 1992. In Mode 4, Reactor Coolant System T ranges from 200°F to 350°F (actual temperature was approximateiy 9 250°F); therefore, the low T bistables are.tripped providing half of the logic signal required MSI. The high steamline flow logic requires indication of high flow in one (1) out of two (2) channels per Steam Generator (S/G) in two (2) of the four (4) S/Gs. The MSI occurred when the No. 22 steam Generators (S/G) steamline flow channel No. 1 and No. 24 S/G Steamline flow channel No. 1 bistables tripped. MSI is an Engineered Safety Feature (ESF). Therefore, on May 1, 1992, at 0536 hours0.0062 days <br />0.149 hours <br />8.862434e-4 weeks <br />2.03948e-4 months <br />, this event was reported to the Nuclear Regulatory Commission (NRC) in accordance with Code of Federal Regulations lOCFR 50.72(b) (2) (ii). APPARENT CAUSE OF OCCURRENCE:

The root cause of this event is "Design, Manufacturing, Construction/

Installation inadequacy.

With the plant in Mode 4, condensation of steam occurs in the steamline flow reference legs resulting in channel spikes. *This apparently occurr.ed coincidently in the Nos. 22 and 24 S/G channels satisfying th.e logic for MSI. Salem. Unit 1 has experienced similar MSI actuations (e.g., September 23, 1991; refeience LER . I

• LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Sal.em Generating Station Unit 2 DOCKET NUMBER 5000311 APPARENT.CAUSE OF OCCURRENCE:'

{cont'd) LER NUMBER 92-008-00 PAGE 3 of 4 Assessment of this event, by Maintenance personnel, was that the false high steam flow signals were not caused by failed components.

The. false signals cleared, on their own, after a few hours. The Salem design arrangement for main flow differential pressure measurement includes two (2) taps (to provide.redundancy) on the high and* low pressure side of the main steamline venturi.

• Attached.to the taps are 1 11 manual globe valves. Steam is through 1 11 pipe to.condensate pots *1ocated .near the high pressure tap. The condensate is then directed to a model 1153HD5 differential pressure transmitter via a 3/8" line.

• ANALYSIS OF OCCURRENCE:

MSI protection is applicable in Mode 1 (Power Operation), Mode 2 (Startup), and Mode 3 (Hot Standby).

It is provided to mitigate the consequences of various design base accidents including main steamline rupture. and steam generator primary to se*condary tube rupture. In Mode 4, the reactor is subcritical with T between 200°F and 350°F. Decay* heat is removed the Residual Heat Removal system or steaming from* the steam generators.

Makeup water to the S/Gs can be supplied by either a Condensate Pump or by an Auxiliary Feedwater Pump. In Mode 4, the Auxiliary Feedwater System {BA} is not required to be operable.

At the 'time of the actuation, decay heat removal was being accomplished using the Residual Heat Removal (RHR) System {BP}. All. valves which close on an MSI signal were already closed. Since the actuation was not the result of an actual plant need for Main Steam Isolation, this event did not affect the health or safety of the

• public.

since Main Steam Isolation is an ESF system, this event is reportable to the Nuclear Regulatory Commission in accordance

• with Code of Federal Regulations 10CFR50.73(a)

(2) (iv). After initiation of the MSI signal, Nos. 23 and 24 S/G main steamline isolation bezel and overhead'alarm indications were not received.

Investigation revealed one of the redundant position indication limit switches for the 23MS167 and 24MS167 valves required minor adjustment.

The 23MS167 and 24MS167 valves were already closed* at the time of the event. All four MS167 valves were successfully

• functionally tested (i.e., stroked) after this event. CORRECTIVE ACTION: As identified by Salem Unit 1 LER 272/91-031-00, assessment of the ma1n steamline flow instrumentation has been on-going.

An in-depth study was completed prior to this event. Engineering has initiated development of proposed design modifications to correct the main steamline flow sensing line concerns. ,

• *

• LICENSEE EVENT REPORT (LER) TEXT CONTINUATION°-

Salem Generating Station Unit 2 CORRECTIVE ACTION: (cont'd) DOCKET NUMBER 5000311 LER NUMBER 92-008-00 PAGE 4 of 4 The 23MS167 and 24MS167 valves limit switches were adjusted and the va.l ves were functionally tested (i.e. , stroked)

• MJP:pc SORC Mtg.92-063 General Manager -Salem Operations