ML18096A933

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LER 92-017-00:on 920730,main Steam Line Isolation Occurred Concurrent W/High Steam Line Flow Signal During Heatup. Caused by Design,Mfg & Const Installation Deficiency.Design Mods to Correct Flow Sensing Line underway.W/920831 Ltr
ML18096A933
Person / Time
Site: Salem PSEG icon.png
Issue date: 08/31/1992
From: Pollack M, Vondra C
Public Service Enterprise Group
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-92-017, LER-92-17, NUDOCS 9209030188
Download: ML18096A933 (4)
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Text

CPS~~

  • Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station August 31, 1992
u. s. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Dear Sir:

SALEM GENERATING STATION LICENSE NO. DPR-70 DOCKET NO. 50-272 UNIT NO. 1 LICENSEE EVENT REPORT 92-017-00 This Licensee Event Report is being submitted pursuant to the requirements of the Code of Federal Regulations 10CFR 50.73(a) (2) (iv). This report is required to be issued within thirty (30) days of event discovery.

Sincerely yours, C. A Vondra General Manager -

Salem Operations MJP:pc Distribution 030020 The Energy People 9209030188 920831 95-2189 (10M) 12-89 PDR ADOCK 05000272 S PDR

NRC FORM 366 ! U.S. NUCLEAR REGULATORY COMMISSION (6-89) .. 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-530), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASH.INGTON, DC 20503.

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EVENT DATE 151 LER NUMBER 16) REPORT DATE (71 OTHER FACILITIES INVOLVED CBI MONTH DAY VEAR VEAR )} sez~~~~~AL }) ~~~~~~ MONTH DAV VEAR FACILITY NAMES DOCKET NUMBER(SI OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE Rl:QUIREMENTS OF 10 CFR §: (Check ono or more of the following} 1111

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LICENSEE CONTACT FOR THIS LER 112) ll0.73Coll2llvllil1Bl 60.731oll21Cxl NAME TELEPHONE NUMBER AREA CODE M .T 'Dl"ll 1 ,,,..,. - T.F.~ rnnrdinat:Or 610 I 9 3 13 I 91- 12 I 0 12 12 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 113)

MANUFAC* MANUFAC*

CAUSE SYSTEM COMPONENT CAUSE SYSTEM COMPONENT TUR ER TUR ER I I I I I I I I I I I I I I I I I I I I I I I I I I I I SUPPLEMENTAL REPORT EXPECTED 114) MONTH DAV VEAR EXPECTED n

SUBMISSION YES (If y11, compl*r. EXPECTED SUBMISSION DATE}

ABSTRACT (Limit to 1400 1poce1, i.*.* *pproxim*tely fifr.*n 1ingle-1poc* typewrimn lined 116) rx-i NO DATE 1151 I I I On July 30 1992, at 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br />, a main steamline isolation (MSI) occurred on a* low TIii v g (< 543°F) coincident with high steamline flow signal during hea~up. The plant had entered Mode 4 at 1146 hours0.0133 days <br />0.318 hours <br />0.00189 weeks <br />4.36053e-4 months <br /> on July 30, 1992. MSI is an Engineered Safety Feature (ESF). In Mode 4, Reactor Coolant System T ranges from 200°F to 350°F (actual temperature was approximatelyv~50°F); therefore, the low T 9 bistables are tripped providing half of the logic required f~~ MSI. The high steamline flow logic requires high flow indication in one (1) of two (2) channels per Steam Generator (S/G) in two (2) of the four (4) S/Gs. The MSI occurred when the steamline flow channel bistables tripped for several S/Gs. The root caus~ of this event is "Design, Manufacturing, Construc:t.ion/Installation" inadequacy. With the plant in Mode 4, condensation occurs in the steamline flow reference legs resulting in channel spikes. This occurred coincidently in several S/G channels satisfying the logic for MSI. Salem Units 1 and 2 have experienced similar MSI actuations (e.g., reference LERs 272/91-031-00 and 311/92-008-01). An in-depth study was completed. Engineering is developing design modifications to correct the main steamline flow sensing line concerns.

NRC Form 386 1~91

Salem Generating Station

  • DOCKET NUMBER LICENSEE EVENT REPORT (LER) TEXT CONTINUATION LER NUMBER PAGE Unit 1 5000272 92-017-00 2 of 3 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 actuation: main steamline isolation due to design Event Date: 7/30/92 Report Date: 8/31/92 This report was initiated by Incident Report No.92-475.

CONDITIONS PRIOR TO OCCURRENCE:

Mode 4 (Hot Shutdown)

DESCRIPTION OF OCCURRENCE:

On July 30 1992, at 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br />, a main steamline isolation (MSI) occurred on low T (< 543°F) coincident with high steamline flow signal during healtip. The plant had entered Mode 4 at 1146 hours0.0133 days <br />0.318 hours <br />0.00189 weeks <br />4.36053e-4 months <br /> on July 30, 1992.

In Mode 4, Reactor Coolant system T ranges from 200°F to 350°F 9

(actual temperature was approximately 250°F); therefore, the low T bistables are tripped providing half of the logic required for MSI:vg The high steamline flow logic requires high flow indication in one (1) of two (2) channels per Steam Generator (S/G) in two (2) of the four (4) S/Gs. The MSI occurred when the steamline flow channel bistables tripped for several S/Gs.

MSI is an Engineered Safety Feature (ESF). Therefore, on July 30, 1992, at 1750 hours0.0203 days <br />0.486 hours <br />0.00289 weeks <br />6.65875e-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 occurs in the steamline flow reference legs resulting in channel spikes. This occurred coincidently in several S/G channels satisfying the logic for MSI. Salem Units 1 and 2 have experienced similar MSI actuations (e.g., reference LERs 272/91-031-00 and 311/92-008-01).

Assessment of this event, by Maintenance personnel, concluded that the false high steam flow signals were not caused by failed components.

"" . . ~

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 1 5000272 92-017-00 3 of 3 APPARENT CAUSE OF OCCURRENCE: (cont'd)

The false signals cleared, on their own, after a few hours. Channel functional tests of the main steamline flow instrumentation were successfully completed.

The Salem design arrangement for main steamline 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 directed through 1 11 pipe to condensate pots located near the high pressure tap. Condensate is then directed to a Rosemount 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 secondary tube rupture.

In Mode o

4, the o reactor is subcritical with, Tavg between '

200 F and 350 F. Decay heat is removed either by the Residual Heat Removal (RHR) System {BP} 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.

At the time of the actuation, decay heat removal was being accomplished using the RHR System. 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 with Code of Federal Regulations 10CFR50. 73 (a) (2) (iv).

CORRECTIVE ACTION:

As identified in Salem Unit 1 LER 272/91-031-00 and Salem Unit 2 LER 311/92-008-01, main steamline flow instrumentation was being assessed. An in-depth study was completed. Engineering is developing design modifications to correct the main steamline flow sensing line concerns.

MJP:pc SORC Mtg.92-096

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