ML18096A754

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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 PSEG icon.png
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)
v • d • e


Text

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PS~G Public Service Electric and Gas Company P.O. Box 236 Hancocks Bri(:Jge, New Jersey 08038.

Salem Generating Station May 28, 1992

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 This Licensee*Event Report is being submitted pursuant to the requirements of the Code of .Federal Reguiations 10CFR 50.73(a) (2)(iv). This report is required to be issued within thirty (30) days of event discovery.

c. A.* Vondra General Manager Salem Operations MJP:pc

.Distribution

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NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION 16-89) APPROVED OMB NO. 3150-0104 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER) 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.

l FACILITY NAME 111 DOCKET NUMBER 121 I PAGE 131 Salem Generating Station - anit 2 o 15 I o 1o I o I 31 11 l 1 loF a I 4 TITLE (41 ESF Signal*Actuation~ Main Steamline Isolation Signal - Channel Spike EVENT DATE (61 LER NUMBER (61 REPORT DATE 171 OTHER FACILITIES INVOLVED (Bl MONTH DAY YEAR YEAR  :[~ff SEQUENTIAL NUMBER ~tt~

REVISION NUMBER MONTH DAY YEAR FACILITY NAMES DOCKET NUMBERISl Salem Unit 1 o 1s101010 121 71 2 a Is al1 9 2 912

- al a Is - al a a Is 2js 91 2 o1s1010101 I I THIS REPORT IS SUBMl.TTED PURSUANT TO THE R~OUIREMENTS OF 10 CFR §: (Chock on* or mor* of th* following) (111 OPERATING MODE (81 x I al a a l 20.402(bl 20.406(cl 60.73foll2llivl 73.71(b)

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POWER LEVEL 1101 I 20.-(o)(1)(i) 20.4061*11111111

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'- b*low and in Taxt. NRC Form 366AI LICENSEE CONTACT FOR THIS LER (12.I NAME TELEPHONE NUMBER AREA CODE 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 MANUFAC* MANUFAC*

COMPONENT SYSTEM COMPONENT TUR ER TUR ER

.I I I I I . I I I I I I 1 I I I I I I I I I I I I I I I I SUPPLEMENTAL REPORT EXPECTED (141 MONTH DAY YEAR EXPECTED n YES (If yos, complete EXPECTED SUBMISSION DATE!

A~STRACT (Limir ro 1400 lxl NO.

spac6t, i.e., approximately fiftesn single-spactt rVpt1writt11n lines) 116)

SUBMISSION DATE 1151 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 during*~tartup 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 require~v 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 DOCKET NUMBER LER NUMBER PAGE Unit 2 5000311 92-008-00 2 of 4 PLANT AND SYSTEM IDENTIFICATION:

Westinghouse -* Pressurized Water Reactor 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. 92~288.

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 ~t~rtup 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 9

(actual temperature was approximateiy 250°F); therefore, the low T bistables are.tripped providing half of the logic signal required ibi:-

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 10CFR 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 272/91-031-00)~

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Sal.em Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 2 5000311 92-008-00 3 of 4 APPARENT.CAUSE OF OCCURRENCE:' {cont'd)

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 steamlin~ 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 di~ected through 1 11 pipe to.condensate pots *1ocated .near the high pressure tap. The condensate is then directed to a Ros~mount 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 9 between 200°F and 350°F. Decay* heat is removed eith~~ by 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. However~ since Main Steam Isolation is an ESF system, this event is reportable to the Nuclear Regulatory Commission in accordance

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 th~t 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 DOCKET NUMBER LER NUMBER PAGE Unit 2 5000311 92-008-00 4 of 4 CORRECTIVE ACTION: (cont'd)

The 23MS167 and 24MS167 valves limit switches were adjusted and the va.l ves were functionally tested (i.e. , stroked)

  • General Manager -

Salem Operations MJP:pc SORC Mtg.92-063

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