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#REDIRECT [[05000272/LER-1990-019-01, :on 900603,main Steam Line Isolation Actuation Occurred & Bistables for Low Steam Line Pressure Logic Tripped.Caused by Equipment/Design Concerns W/Sensing Lines. Design Mods Re Flow Instrumentation Underway]]
| number = ML18095A327
| issue date = 07/03/1990
| title = LER 90-019-00:on 900603,main Steam Line Isolation Actuation Occurred & Bistables for Low Steam Line Pressure Logic Tripped.Caused by Equipment/Design Concerns W/Sensing Lines. Design Mods Re Flow Instrumentation underway.W/900703 Ltr
| author name = Miller L, Pollack M
| author affiliation = PUBLIC SERVICE ELECTRIC & GAS CO. OF NEW JERSEY
| addressee name =
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000272
| license number =
| contact person =
| document report number = LER-90-019-01, LER-90-19-1, NUDOCS 9007060239
| document type = LICENSEE EVENT REPORT (SEE ALSO AO,RO), TEXT-SAFETY REPORT
| page count = 5
}}
 
=Text=
{{#Wiki_filter:1*
i e
()      PS~G Public Service Electri_c and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem G.enerating Station July 03, 1990 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 90-019-00 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 within thirty (30) days of discovery.
Sincerely yours, L. K. Miller General Manager -
Salem Operations MJP:pc Distribution 95*2189 (10M) 12-89
 
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I0.731111211111                          I0.7:1lall2llwllillll I0.731al!21UHI                        . I0.7:11al!211*1 LICENSEE CONTACT fOll THIS LEll (!II NAME                                                                                                                                                              TELEPHONE NUlo!llEll AREA CODE M. J. Pollack - LER Coordinator COMPLETE ONE LINE FOR EACH .COMPONENT FAILURE DEICllllED IN THll llll'ORT 11:11 CAUSE    SYSTEM      COMPONENT                MANUFAC.                                                                                          MANUFAC.
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 IUPPLEMENTAL REPORT EXPECTED IHI                                                                                    MONTH    .DAY    "fAll EX,ECTED SUtlM1$$10N f      YES (If yn. <0mplot1 EXPECTED SUtlMISSION DATE!
DATE 1151 .
I      I      I UITRACT (Lim/I ro    14()() ll>>C*. I.e.* epprox/mettly fir...n rl"fl**IPOCO rypewri,..n final 1111 On 6/3/90 at 0712 hours, a Main Steamline Isolation actuation occurred.
* At the time of the event, the Unit was in Mode 4 and heating up in preparation for startup. The Main Steamline Isolation signal was generated upon completion of the High Steamline Flow coincident with Low Steamline Pres~ure logic (by design). In Mode 4, the bistables for low steamline pressure are tripped due to plant conditions providing half of the logic signal required for Main Steamline Isolation. The high steamline ,flow logic was completed when the bistables for No. 11 and No. 12 Steam Generator (S/G) high steamline flow tripped. Main Steamline Isolation is an Engineered Safety Feature (ESF). The root cause of this event has been attributed to possible equipment/design concerns associated with the main steamline flow transmitter sensing lines. The initial investigation by Maintenance personnel, to determine why the Nos. 11 and 12 S/G high steamline flow bistables tripped, did not identify any failed components. A channel functional test, of the subject steam flow channels, was successfully completed .
  . Steamline flow measuring.design concerns have previously been identified via LER 272/88-017-01 (i.e., steamline flow measurement drift concern). Engineering believes that the drift concern and this recent event appear to have a related cause. Design modifications, to correct the main steamline flow instrumentation concerns, are under development.
NFIC Form JSe 19.aJ>
 
LICENSEE EVEN'!' REPORT (LER) TEXT CONTINUATION Salem Generating Station        DOCKET NUMBER    LER NUMBER    PAGE Unit 1                            5000272        90-019-00    2 of 4 PLANT AND SYSTEM IDENTIFICATION:
Westinghouse    - Pressurized Water Reactor Energy Industry Identification System (EIIS) codes are identified in the text as {xxl IDENTIFICATION OF OCCURRENCE:
Engineered Safety Feature Actuation; Main Steamline Isolation Due to Equipment/Design Concerns Event Date:    6/03/90 Report Date:    7/03/90 This report was initiated by Incident Report No. 90-369.
CONDITIONS PRIOR TO OCCURRENCE:
Mode 4 (Hot Shutdown); Reactor in preparation for startup Tevg at approximately 270&deg;F and increasing DESCRIPTION OF OCCURRENCE:
On June 3, 1990 at 0712 hours, a Main Steamline Isolation actuation
{JEI occurred. At the time of the event, the Unit was in Mode 4 (Hot Standby) and heating up in preparation for startup. The Main Steamline Isolation signal was generated upon completion of the High Steamline Flow coincident with Low Steamline Pressure logic (by design).
In Mode 4, the bistables for low steamline pressure are tripped due to plant conditions providing half of the logic signal required for Main Steamline Isolation. The high steamline flow logic was completed when the bistables for No. 11 and No. 12 Steam Generator (S/G) high steamline flow tripped ..
Main Steamline Isolation is an Engineered Safety Feature (ESF).
* Therefore, on June 03, 1990 at 0840 hours, the Main Steamline Isolation signal event was reported to the Nuclear Regulatory Commission in accordance with Code of Federal Regulations 10CFR 50.72(b) (2) (ii).
APPARENT CAUSE OF OCCURRENCE:
The root cause of this event has been attributed to possible equipment/design concerns associated with the main steamline flow transmitter sensing lines.
The initial investigation by Maintenance personnel, to determine why the Nos. 11 and 12 S/G high steamline flow bistables tripped, did not identify any failed ~omponents. A channel functional test, of the
 
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station      DOCKET NUMBER    LER NUMBER        PAGE Unit 1                          5000272        90-019-00        3 of 4 APPARENT CAUSE OF OCCURRENCE:  (cont'd) subject- steam flow channels, was successfully completed.
The Salem design arrangement of a main steamline flow differential pressure measurement system includes two (2) taps (to provide 100%
redundancy) on the high and low pressure side of the main steamline venturi. Attached to the taps are 1" manual globe valve~. Steam is directed through l" pipe to condensate pots located near the high pressure tap. The steam is then. directed to the Rosemount model 1153HD5 differential pressure transmitters via a 3/8" line.
Steamline flow measuring design concerns have previously been identified vi.a LER 272/88-017-01. That LER addresses steamline flow measurement drift when the Unit is at power. Engineering believes that the drift concern and this recent event appear to be related to design concerns associated with the transmitter sensing lines.
Possible causes of these events are discussed in a detailed study which had been completed in support of LER 272/88-017-01.
ANALYSIS OF OCCURRENCE:
Main Steamline Isolation protection is applicable only in Mode 1 Power Operation}, Mode 2 (Critical Operation}, and Mode 3 (Hot Standby}. It is designed to ~itigate the consequences of various design base accidents including main steamline rupture and steam generator primary to secondary tube rupture.
In Mode 4, the reactor is subcritical with Tavg greater than 200&deg;F but less than 350&deg;F. Decay heat is removed either by the Residual Heat Removal system or via 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 {BAJ is not required to be oper~ble.
During this event, decay heat removal was accomplished using the Residual Heat Removal (RHR) System {BPI via No. 11 RHR Pump. When*
the Main Steamline Isolation signal was actuated, the four MS7 Main Steamline Drain Valves, open in support of main steamline warmup *
(procedure IOP-2), went closed as ~esigned. The other valves which would close on a Main Steamline Isolation signal were already closed.
Since this event 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:
Design modifications to correct the main steamline flow instrumentation concerns are under development for implementation (as
 
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station      DOCKET NUMBER    LER NUMBER    PAGE Unit 1                          5000272        90-019-00    4 of 4 CORRECTIVE ACTION:  (cont'd) per the corrective actions to LER 272/88-017-01).
General Manager -
Salem Operations MJP:pc SORC Mtg. 90-079}}

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