ML18096B011

LER 92-014-00:on 920903,reactor Trip & Cooldown Occurred Upon Opening Reactor Trip Breaker 2A.Caused by Personnel Error.Individual Involved Disciplined & Administrative Procedure Re Operating Practices revised.W/920924 Ltr

ML18096B011
Person / Time
Site:	Salem
Issue date:	09/24/1992
From:	Pollack M, Vondra C Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-92-014-01, LER-92-14-1, NUDOCS 9210020224
Download: ML18096B011 (7)

Text

(_~,:~ PS~G

• Public Service Electric and _Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 Salem Generating Station September 24, 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 92-014-oo 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,

/1

(/tv11!U

//'

Z

c. A. Vondra

//_.

General Manager -

/

Salem Operations MJP:pc Distribution

(" r.ciu~;;

.:Jv

-. 0 9210020224 920924 PDR ADOCK 05000311 s PDR

NRC FORM366 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 (LERI 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, WASHINGTON, DC 20503.

FACILITY NAME (1) I DOCKET NUMBER (2) I PAGE 131 Salem Generati~g Statcbon - tJni t 2 o I 5 I o I o I o I 3 11 I 1 1 OF 016 TITLE (4)

Reactor trip from 100% power upon opening of the 2A Reactor Trip Breaker.

EVENT DATE (5) LER NUMBER (6) REPORT DATE 17) OTHER FACILITIES INVOLVED 18)

MONTH DAY YEAR YEAR HI SEQUENTIAL NUMBER H? AEV1SION NUMBER MONTH DAY YEAR FACILITY NAMES DOCKET NUMBERISI o1s1010101 I I q~ 013 9 2 91 2

- 011 14 0 10 019 214 91 2 0151010101 I I THIS REPORT IS SUBMITTED PURSUANT TO THE RtOUIREMENTS OF 10 CFR §:(Chock one or more of the following! 1111 OPERATING MODE 19) 1 20.402lb)

....__ 20.4051c)

,..x 50,73loll2)(iv) 73.71lb)

POWER LEVEL 11 0 0 - 20.4051ell1 )Ii) 20.4051o)(1llii) 50.38lc)l1) 50.38lc)(2)

,.____ 50.73lo)l2)(v) 50.731o)(2)(vii) 73.71 lc)

OTHER (Spadfy in Abstract 110)

• .:.:.- . .:-*  :.:.-.-: ,.**:-:-:-:-:::*:-:* .,....,. - ,.____ below and in T6J<t, NRC Form

\\ ::*:::;.

~ '.*'.*.*.:*:-*:-:.:.;.:.:*>>*:*; ""

20.40510)(1 )(iii) 20.40510)(1 )(iv)

- 50.73lo)(2)(i) 60.73(o)(2) (iii 50.731o)l2)(viii)IA) 50.731*) 121 lviiil IB)

.366A!

• -:-:*-:*:::::::::;:::::::::=*=*

-:*: *:*:-:*::::;:;:::;:;:;:::;::::*=*=

• '""" 20.40510)(1 )(v)

- 50.73l*ll2)(iii) 50.73lo)(2llx)

LICENSEE CONTACT FOR THIS LER 112)

NAME TELEPHONE NUMBER AREA CODE M.J. Pollack - LER Coordinator 610 19 3 13 ~ , - ,2 p ,2 ,2 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 1131 MANUFAC~ REPORTABLE ))'.;; ;.;..;.; *'.*'.

MANUFAC* REPORTABLE [=}?> :*:..:* ""

COMPONENT .;.;

TO NPRDS 1:::-:-:::*:-:::-:-:::-;.*'.*'.*'.* .:.-".*::-:

CAUSE SYSTEM COMPONENT CAUSE SYSTEM *:".

TUR ER TO NPRDS I*>'.'.*'.* :;:;:::::;:; :;-::*:* ;.;.:;:;;:;: 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 [/] *'.*'.""""" tt/....;.,..........

SUPPLEMENTAL REPORT EXPECTED 1141 MONTH DAY YEAR EXPECTED

~NO SUBMISSION I YES (If yes. comp/ere EXPECTED SUBMISSION DATE!

DATE 1151 I I I ABSTRACT (Limit ro 1400 spaces. i.e., approximately fifteen single-space typewritten _lines) (16)

On 9/3/92, at 0917-hours a reactor trip occurred due to the II AII Reactor Trip Breaker (RTB) opening. Following the reactor trip, a cool down occurred. Auxiliary Feedwater was reduced and the MSlO valves (atmospheric relief valves) were verified closed. However, the cool down continued and manual Main steamline Isolation (MSI) was initiated stopping the cooldown. After the MSI, Reactor Coolant System (RCS) pressure and temperature reached maximum levels of 2280 psig and 552° F.

The Pressurizer Master Controller was taken to manual. Pressurizer SP,ray was initiated and pressure reduced and stabilized at 2235 psig. At 0936 hours0.0108 days <br />0.26 hours <br />0.00155 weeks <br />3.56148e-4 months <br />, the 21MS15 and 22MS15 main steam.safety valves lifted. The 21MS10 and 22MS10 valves had not opened. At 1005 hours0.0116 days <br />0.279 hours <br />0.00166 weeks <br />3.824025e-4 months <br />, an "Unusual Event" was declared for, 11 SG Safety Failure to Reseat". The Unusual Event was terminated with the plant in Mode 3. Investigation of this event included testing of the "2A 11 .RTB which did not identify any RTB problems. The cause of the reactor trip is attributed to personnel error. An NEO exhibited poor judgement resulting in a sequence of events leading to the trip. Appropriate disciplinary action has been taken with the individual involved. The reactor trip and events following the trip will be reviewed by the Nuclear Training Center. Investigation of a Pressurizer pressure master controller concern is continuing. The MSlO control concerns have been investigated by engineering and design changes are.planned. The breaker in the II 2A II RTB cabinet is being sent to Westinghouse to verify that it functions per design.

NRC Form 366 (6-891

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION

. Salem Generating Station DOCKET NUMBER . LER NUMBER PAGE Unit 2 5000311 92-014-00 2 of 6 PLANT AND SYSTEM IDENTIFICATION:

Westinghouse - Pressurized Water Reactor Energy Industry Identification System (EIIS) codes are identified in the text as {xx}

IDENTIFICATION OF OCCURRENCE:

Reactor Trip from 100% power upon opening of the 2A Reactor Trip Breaker Event Date: 9/3/92 Report Date: 9/24/92 This report was initiated by Incident Report No.92-567.

CONDITIONS PRIOR TO OCCURRENCE:

Mode *1 Reactor Power 100% ~ Unit Load 1150 MWe DESCRIPTION OF OCCURRENCE:

On September 3, 1992, at .0917 hours0.0106 days <br />0.255 hours <br />0.00152 weeks <br />3.489185e-4 months <br />, during normal power.operation, a reactor trip occurred with a fiist out alarm of "power range neutron flux rate high". At the time of the event, no maintenance -was in

• progress which could be attributed to the trip.

Investigation revealed that the "A" Reactor Trip.Breaker (RTB) had opened resulting in the control rods dropping into the core causing the negative rate reactor trip signal. Per design, the "B" RTB opened approximately 22 cycles after event initiation.

• The turbine/

generator tripped following the reactor trlp.

Following the reactor trip, a cooldown occurred. In accordance with.

Emergency Operating Procedure EOP-TRIP.-2, Auxiliary Feedwater was reduced from 44E04 lbm/hr to 22E04 lb /hr and the MSlO valves *

(atmospheric relief valves).were verified closed. However, the cooldown continued and, at 0921 hours0.0107 days <br />0.256 hours <br />0.00152 weeks <br />3.504405e-4 months <br />, a_ manual Main Steamline Isolation (MSI) (an Engineered Safety Feature) was initiated stopping the cooldown, in accordance with the EOP procedure.

After.the MSI, Reactor Coolant System (RCS) {AB} pressure and temperature reached maximum levels of '22so psig and 552°F, respectively. Based upon an elevated Pressurizer Pressure Operated Relief Valve (PORV) tailpipe temperature of 191°F, and pressure spikes observed.in the Pressurizer Relief Tank (PRT), the Pressurizer Master Controller was taken to manual. Pressurizer Spray was initiated and pressure reduced and stabilized at 2235 psig. The Pressurizer Power Operated Relief Valves (PORVs) were not observed to hav*e lifted.

• LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 2 5000311 92-014-00 3 of 6 DESCRIPTION OF OCCURRENCE: (cont'd)

At 0936 hours0.0108 days <br />0.26 hours <br />0.00155 weeks <br />3.56148e-4 months <br />, the 21MS15 and 2~MS15 main steam safety valves .

lifted. Main steamline pressure was 1052 psig. This is above the 1000 psig setpoint fo_r the 21MS10 and 22MS10 valves; however, they .

did not open.

This allowed T.avg to rise subsequentty causing steam pressure to rise until the MS15 valve setpoints were reached.

• At 0*952 hours, the two (2) MS15 valves were still open with steamline pressure at 980 psig. consequently at 1005 hours0.0116 days <br />0.279 hours <br />0.00166 weeks <br />3.824025e-4 months <br />, an i*unusual Event" was declared in accordance with Section 2A of the Emergency Classification Guide, "SG Safety Failure to Reseat~.

At 1029 a,nd 1041 hours0.012 days <br />0.289 hours <br />0.00172 weeks <br />3.961005e-4 months <br />, respectively, -the 21 & *22MS15 valves were observed closed. Plant heatup was then initiated from an initial T~yg of 530°F to minimize risk of safety injection. The T v 9 se~point is~

• 541°F for*safety injection logic of low T

• a v.g a

coincident with high steam flow in two of four steamlines.

At 1235 hours0.0143 days <br />0.343 hours <br />0.00204 weeks <br />4.699175e-4 months <br />, the plant heatup was terminated due to observed steam release from the 21 and 22MS15 valves. Procedure IOP-6, "Hot Standby to Cold Shutdown" was initiated. _At 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br />, a plant cooldown commenced. Upon reaching a T vg of 530°F with the 21 and 22MS15 valves observed to:be fully c1osed, the Unusual Event was terminated. The plant remained in Mode 3 (Hot standby) pending results of investigations into the reactor trip and subsequent plant transient.

The Nuclear Regulatory Commission (NRC) was notified of the events as.sociated with the reactor trip, main steamline isolation and the declared Unusual Event in accordance with Code of Federal Regulations 10CFR 50.72.

APPARENT CAUSE OF OCCURRENCE:

Investigation of this event included testing of the "2A" RTB. This testing included: 1) physical manipulation of the breaker to cause it to trip; 2) performing portions_ of the procedure for "Reactor Trip Bypass AIR circuit Breaker Semi-Annual Inspection, Lubrication and Testing"; 3) performing the procedure for "Train-A Reactor Trip and Reactor Trip Bypass Breakers P-4 Permissive Test" several times; 4) removing the "2A" RTB from its cubicle and inspecting it relative to recent NPRDS data,; and 5) additional field testing which included monitoring during a breaker trip.*

Additional investigations included visual inspe~tion of the Solid State Protection System (SSPS) UV card and megger/resistance tests of the cable from the SSPS to the."2A" RTB cubicle.

The above testing and inspections did not identify any RTB problems which would have resu1ted in the breaker opening.

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating station DOCKET NUMBER LER NUMBER -PAGE Unit 2 5000311 92-01"4-00 4 of 6 APPARENT CAUSE OF OCCURRENCE: (cont'd)

Prior to the trip, a Nuclear Equipment Operator (NEO), was assigned to rack-in the Salem Unit 1 "lB" Reactor Trip Bypass Breaker (~TBB) in support of surveillance testing. After entering the Unit 1 switchgear Room, the NEO left the I&C technicians and went to the Unit 2 Switchgear Room to view the Unit 2 "2A 11 -RTBB. He did this to reassure himself of the appearance of a racked out breaker. Upon entering the Unit 2 Switchgear Room, he states that he opened the "2A" RTBB door and studied the position of the racked out breaker for several seconds._ While viewing the breaker, he heard the Unit 2 RTBs open.

Based on investigation, the cause of the reactor trip is attributed to personnel error. It was determined that the NEO exhibited poor judgement resulting in a sequence of events leading to the trip. The NEO did not inform the Control Room or seek supervisory guidance before going to the 11 2A" RTBB cabinet to open it. The cabinet is clearly marked as a trip hazard. Also, the NEO stated that he was not aware of procedure Sl.OP-SO.RCP-0002, "Reactor Trip or Reactor Trip Bypass Operations". It details the operation of the RTBs. - Due to the nature of this procedure, it is not required-to be at the job site in support of work since the procedure tasks involve routine equipment operation. - -

Further review of this event revealeq that the physical arrangement of the Salem Unit 1 and Salem Unit 2RTBB and RTB cabinets are reversed. This factor contributed to the determination that the NEO opened the n2A11 RTB _cabinet instead of the "2A RTBB capinet.

ANALYSIS OF OCCURRENCE:

There are two (2) reactor trip breakers ("A" and "B") in series, which connect the output of the rod drive motor generator sets to the rod control power cabinets. * -In the -event of a reactor trip signal, these breakers open, removing power from the control rod drive mechanisms allowing the control rods to drop into the reactor core.

The opening of either breaker will cause this to occur. Two (2)

RTBBs are provided to allow surveillance testing at power.

The sequence of events as recorded by the P-250 computer confirmed that the "2A" RTB opened first. The first out indication would be the "power range neutron flux rate high" and the 11 2B" RTB would open a few cycles later.

I The , reduction in- Tavg , re*quiring MSI,, has been experienced I

I I

during other reactor trips (e~g., Unit 2 LER 311/92-009-00).

Engineering has investigated Tavq reduction (during trips) and design modifications are being assessed. . .

A low pressurizer level signal occurred resulting in RCS letdown isolation. - This signal was due to the cooldown (prior to MSI) and

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE

Unit*2. 5000311 92-014-00 5' of 6 ANALYSIS OF OCCURRENCE: (cont'd) the low power history of the core. Charging. flow remained established resulting in an increase of pressurizer pressure to approximately 2280 psig. The pressurizer spray actuation setpoint is*

2260_psig .. As stated previously, the operators placed the master pressure controller in manual and initiated pressurizer spray.

Investigation of this concern included performance of applicable

• sections of procedure 2PD-2.l.082, "2PC-455K Pre~surizer Pressure Control". The controller was. calibrated and its operability was checked and found satisfactory. Investigation into the pressurizer master controller concern is continuing.

Following the main steamline isolation, the 21 and 22MS15 valves*

lifted with the RCS at a temperature of 552°F. This correlates to a steam generator header pressure of 1051 psig. Recorded steam pressures for the 21 and 22 steamlines were 1030 psig and 1036 psig, respectively. Each valve lifted at least twice prior to plant stabilization at 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br />. There are five (5) main steamline safety valves*per steamline with the MS15 valves having the lowest setpoint (1070 psig).

These valves operated for approximately one (1) hour during the .

event. The steam flow that passed through the_ valves during the time they were open heated the valve* body and springs. This results in

. lift set and reset.pressure reduction. Following the event, the valves* were lift set tested. The as .found data were 1047 psig for 21MS15 and 1017 psig for 22MS15. The 1017 *psig for the 22MS15 is questionable due to a failed air motor. Both valves were recalibrated and left within 1% of set pressure.

Further investigation revealed that the 23MS15 valve reached a

.pressure of 1045 psig; however, i t did not lift. It was lift set tested with as found readings of 1097 psig. It too was ~ecalibrated.

The valve manufacturer (Crosby) was contacted. No abnormalities or inconsistencies were noted with the MS15 valve operation or subsequent findings.

Main steamline safety valve lifting has occurred previously. On August 8, 1985 reference LER 311/85-005-00), a similar event occurred where the "2A" RTB opened* followed by the lifting of main steamline safety valves. Following the 1985 event, investigation was conducted to determine why the MSlO valves did not respond preventing the main steamline safety valve lift.' It was determined that the MSlO controllers (set at 1000 psig) experience saturation and a subsequent "reset windup" phenomenon. Design changes are planned to address this concern.

The reactor trip and subsequent opening of the MS15 valves did not affect the health and safety of the public. Operator actions were appropriate to mitigate the subsequent transient. The reactor trip

fl\

LICENSEE EVENT REPORT (LER) ~EXT CONTINUATION Salem Generating Station DOCKET NUMBER LER.NUMBER PAGE Unit 2* 5000311 92-014-00 6 of 6 ANALYSIS OF OCCURRENCE: *(cont'd) and manual main steamline isolation events are reportable to ~he NRC in accordance with Code of Federal Regulations 10CFR 50.73(a) (2) (iv) .

. CORRECTIVE ACTION:

Operations* management has reviewed the events associated with the reactor trip. Appropriate disciplinary action is being assessed regarding the individual involved in the event.

The reactor trip.and events following the trip will.be reviewed by the Nuclear Training Center. Licensed and non~licensed Operator training will be enhanced as appropriate.

Administrative procedure NC.NA-AP.ZZ-0005(Q), "Station Operating Practices", will be revised to require shift notification and -

approval for accessing equipment posted with "Trip Hazard"_signs.

As identified in the Description of Occurrence section~ _a Pressurizer pressure control concern occurred. Investigation included performance of applicable sections of procedure 2PD-2;1.os2, "2PC-455K Pressurizer Pressure Control". The controller was calibrated and its operability was checked and found satisfactory.

Investigation into the pressurizer master controller concern is continuing.

The MSlO control concerns .have been investigated by engineering.

Modifications are planned for MSlO controls. The effects of "reset wind-up" phenomenon will be accounted for so that the valves will operate at their proper setpoint.

'The breaker in the 11 2A" RTB cabinet, at the time of the trip, is being sent to Westinghouse to verify that it functions per design.

//d'./ I 6~/!{/l~

(

General Manager -

Salem Operations MJP:pc SORC Mtg.92-105