ML20029A911

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LER 91-001-00:on 910125,reactor Scram Occurred During Calibr of Feedwater Computer Point.Caused by Incorrect Procedure in Summary Sheet.Maint Procedures Reviewed & revised.W/910225 Ltr
ML20029A911
Person / Time
Site: Brunswick Duke Energy icon.png
Issue date: 02/25/1991
From: Harris T, Spencer J
CAROLINA POWER & LIGHT CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-91-001-03, LER-91-1-3, NUDOCS 9103050093
Download: ML20029A911 (19)
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m Carolina Powet & Light Company Brunswick Nuclear Project P. O. Box-10429 '

Southport, N.C, 28461 0429 February 25, 1991 FILE: B09 135100 10CFR50.73 SERIAL: BSEP/91 0080 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.- C. 20555 BRUNSWICK STEAM ELECTRIC PLANT UNIT 2 DOCKET NO. 50 324 LICENSE NO. DPR 62 LICENSEE EVENT REPORT 2-91-001 Centlemen:-

In-'accordance with Title 10 of the Code of Federal- Regulations, the enclosed

- Licensee Event Report is submitted. This report fulfills the requirement for a

. written report within thirty (30) days of a reportable - occurrence and is submitted in accordance with the format set forth in NUREG-1022, September 1983.

Very truly yours, u d dra1' pencfr Gen Manager-

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r iswick Nuclear Project TH/-

Enclosure:

cc: Mr. S. D. Ebneter Mr. N. B. Le f BSEP NRC Resident Office-9103050093 910225

_S PDR ADOCK 05000324 PDR Ih Q 1

q NRC eO*1V 366 U.5, NUCifAR REGULATORY CoMM:SuuN AFPROWD oMB NO. 31M0104

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F ACIUTY NAME (1) Brunswick Steam Electric Plant Unit 2 DOCMT NUMBER (?) PAGE (3) 05000324 01 OF 19 Tate <4) Unit 2 Turbine Trip / Reactor SCRAM Vhile Calibrating a Feedvater Computer Point _

EVENT DATE !S) LER NUMBER (d) REPORT DATE (h oTHE R F ACIJTIES iNWMNED (8)

MONTH DAY YEAR YEAR SEO.NO. FtEV. NO MONTH DAY iEAR F ACluTY NAME DOC *iT NUYBER 01 25 91 91 - 001 - 00 02 25 91 l THis RUORT iS SUBM!rlED PUR$UANT TO THE REGAREMENTS OF 10 CFR 1: (Check one or more of 19 fonowirg) (11)

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NAVE TONY HARRIS, REGULATORY COMPLIANCE SPECIALIST ttLEmoNE NUMBEw (919) 457-2038 COMPtfTE oNE UNE l'OR EACH COMEONENT F AILURE DtSCRIDED IN THl$ REPORT (13)

CAUSE SYSTEM COMFONENT MANUFACTURER REPORTABLE CAVSE SYSTEM COWONENT MANUFACTURER REPORTABLE TO NPROS #

TO NPROS SUPPLEMENTAL REPORT ExHCTEO (14) EXPEC'ED MONTH DAY YEAR I

SUBMsSS oN y, Yrs m yes, compwte EXPECTED SU8MiSSON DATO NO DATE (15) 04 01 91 AB5 TRACT Omit to 1400 spaces. Le. approntmatey fifteen smfe space typewntten imes) (16)

On January 25, 1991, Unit 2 reactor scramned from 100% power. The scram was due to a turbine trip on high reactor water level which resulted from the Feedwater Level Control System responding to a sensed loss of feed flow during performance of a Process Computer point calibration on the Feedwater flow logic system. Instrumentation and Control (I&C;

[ technicians performing the calibration failed to recognize a procedure prerequisite step which stated that the unit must be in cold shutdown or refuel to perform the procedure.

The event was due to the fallute of the work control process to prevent this activity from

, being performed, caused by inadequate reviews in the scheduling and implementation phases of

( the process and the failure of the technicians to ensure the prerequisites were met. In

! addition, the procedure Summary Sheet incorrectly stated that no special plant conditions were required for the performance of the procedure. Corrective actions included stoppep of the computer point calibrations, suspension of use of the procedure Summary Sheets, communication of the event with plant personnel, development of a Recovery Action Plan, and initiation of a working 1cvel Task Force to investigate this and similar recent personnci errors. Future corrective actions will be identified in a supplement to this LER pending l

completion of management review of the HPES and Task Force evaluations.

l The safety systems functioned as designed. Equipment response concerns were identified, but j nono posed a significant safety concern.

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WASHINGTON. OC 70555. AND TO THE PAPERWORK AEoVCTON PAQJECT 0150c104). oF FnCE or MANAGEMEvr AND BUDGET. WASHINGTON.DC 20503 FACluTY NAME (1) 00CKET Lf R NUMBER (6) PAGE (3)

NUMBER A Brunswick Stest Electric Plant Unit 2 050003.!4 YEAR NUMBEA NWBER 91 -

001 - 00 l TEF 0F MORE SPACE l$ REQUIRED. USE ADDITONAL NRC FORM 3E6A'S)(17)

EVENT Unit 2 Turbine trip / Reactor SCRAM during calibration of Feedwater Process Computer point B022.

INITIAL CONDITIONS Unit 2 was operating at 100% power. Reactor pressure was at 1010 psig, and reactor ver.sel level was at 186 inches. Reactor Feedwater Level Control was in automatic, readiness: three Highelement Pressure control. The following Coolant Inj ection(systems HPCI),were Reactor operable Core in standby Isolation Cooling (RCIC), Automatic Depressurization System (ADS), Reactor Protection System (RPS), Residual Heat Removal (RHR)/ Low Pressure Coolant Injection (LPCI),

Core Spray (CS), Standby Cas Treatment (SBGT), Standby Liquid Control (SLC), and Emergency Diesel Cenerator (EDC). A Process Computer point instrumentation calibration was in progress on computer point A1713 (B022), Feedwater Flow Loop A, in accordance with Process Instrument Calibration (PIC) OPIC CPU 001, Attachment 13 Data Sheet. Instrumentation and Control (160) technicians, in accordance with Step 7.1.1 of OPIC-CPU 001, were preparing to lift wire number C32 A 18 from terminal DD 84.

EVENT DESCRIPTI.QH On January 25,1991, at 0810, the technicians lif ted wire C32 A-18 from terminal DD 84 in the Feedwater logic loop. A detailed sequence of events for the resulting transient is provided in Attachment A. The lifting of wire C32 A-18 caused a loss of the "A" Feedwater flow signal into the Feedwater Level Control System (FWLCS). As a result, the WLCS increased the speed of both Reactor Feed Pumps (RFP). This caused an increased flow in the feedwater path, which increased Reactor Water Level to the High Level Turbine Trip setpoint. The Main Turbine and both turbine driven RFPs automatically tripped. As a result of the Turbine Trip, a Turbine Stop Valve closure occurred, initiating an automatic Reactor Scram.

Following the Reactor Scram, vessel pressure increased to a transient maximum of 1032 psig. Reactor Water level decrecsed to a translat minimum value of between 116.5" and 121", approachit.g the Low Level 2 (LL2) instrument trip setpoints.

As a result, both the HPCI and RCIC systems initiated, both Recirculation pumps tripped, and Croups 2, 6, and 8 isolation signals were generated. A partial Croup 3 isolation signal was also received. RCIC began injecting into the reactor vessel, and HPCI entered into a minimum flow pathway, not inj ecting into the reactor vessel. Level was being restored to above the LL2 setpount, and the Control Operator manually initiated HPCI to assist in restoring vessel to normal level. HPCI and RCIC were secured once vessel level was returned to normal range. The isolations were reset at 0822, and all rods were confirmed to be fully in. Normal recovery procedurer were then followed. The Site Incident Investigation Team (SIIT) was convened to begin investigation of the event.

EVENT INVESTIGATION The event investigation initially focused on determining the origin of the

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001 - 00 TEXT (tF MORE SF ACE l$ REQUIRED. USE ADO (TONAL NRC FORM X4A3) 07) feedwater flow mismatch. Review of ongoing work activities determined that the calibration being performed on the Process Computer Feedwater flow point was a possible origin of the Feedwater flow mismatch. Inte rviews with involved personnel and reviews of the controlling procedure determined that a Prerequisite step in the procedure (Step 3.2.1) requires that "the unit will have to be in cold shutdown or refuel condition to perform this test, due to the interlocks and controls of this loop (feedwater and recirculation)." the Precautions and Limitations section of the procedure identifies Additionally, the invo lved instrument as part of the single point failure analysis.

The I6C technicians performing the ptocedure lif ted a wire in the feedwater control system as specified by Step 7.1.1 of the procedure. The wire removal resulted ln the loss of part of the feedwater flow signal to the reactor level control system and created a falso steam flos/feedwater flow mismatch. The' level control system responded as designed by increasing feodwater flow to the reactor to compensate for the senred reduction in feedwater flow. This action increased the reactor feed pump speeds and actual feedwater flow. The increased flow increased reactor water level until the high reactor water level trip setpoint was exceeded. The reactor high water level tri driven reactor feedpumps and the main turbine. p initiated trips for the turbineThe main turbine trip initiated the reactor SCRAM.

The immediate investigation performed by the SIIT was initiated to determine the root cause of the SCRAM, and to reconcile potential problems associated with the SCRAM and recovery evolutions. Additional investigations, including a Pir.nt ,

Incident Report and Human Performance Enhancement System (HPES) evaluation, are being completed, The primary factors in the root cause analysis are presented below.

B_q0T CAUSE ANALYSIS The January 25, 1991 SCRAM was the result of the performance of a preventive maintenance calibration procedure (OPIC-CPU 001) during operation which had prerequisites r andating that the procedure be performed during cold shutdown or refuel conditions. The procedure involves the calibration of a signal to the process computer from the feedwater control system. Since performance of this procedure affects the operation of the feedwater loop flow input to the reactor level control system, the procedure should only be performed while the unit is shutdown or in the refuel condition.

The work control process for a preventive maintenance activity at the Brunswick plant involves three phases: procedure preparation and preventive maintenance (PM) route (a prearran ed sequence for performing the PM) development, work scheduling, and work i lementation. For this event, each phase of the work control process had primary barriers that failed.

The procedure prepared for this route was inaccurate in that the Attachment Summary Sheet incorrectly stated that the procedure could be performed under any plant condition. The reviews of this procedure revision did not detect the inconsistency between the Summary Sheet and the Prerequisites in the body of the procedure.

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001 - 00 TEXT (!F MORE LhACE t$ REoulRED USE ADOtTONAL Nsc FonM 36 ENS)(17)

The work scheduling phase did not keep or remove this item from the work schedule. This phase involves multiple reviews of potential work items to ensure that a particular item can be performed at the specified tirne given the anticipated plant conditions. The inadequacy of the reviews which support this phase led to the route being scheduled for work.

The work implementation phase of the work control process failed to prevent the calibration from being performed. The reviews associated with this portion of the process did not detect the prerequisite plant condition for performing this calibration. In addition, the involved technicians did not follow the calibration procedure when they failed to ensure the procedure prerequisites were satisfied prior to lifting the wire.

The focus of an operating unit work scheduling phase at the Brunswick plant is the Site Work Force Control Group (SWFCG) . Work iterns are scheduled through the SWFCG, which includes representatives from each site work organization. SWFCG, however, relies on a multitude of reviews to ensure that work presenteu to the group is acceptable for a given plant condition. For a typical maintenance work item, such as this process computer point calibration, the scheduling process is as follows:

1. The planning process for preventive maintenance activities begins with an automatic computer function which prompes a maintenance planner / analyst from en interval based computer display to generate a route sheet WR/J0. The route sheet lists affected components and procedure numbers necessary to complete the route. The planner does not review procedures or the effects that component manipulation will have on the plant. The designated route sheet is then given to the responsible maintenance foreaan for initial screening, procedure revia , and scheduling.
2. The responsible maintenance foreman performs the initial assessment of the item to determine plant conditions required for the item to be performed, and at what time these conditions will exist. The foreman reviews the work item, including the applicable system work schedule, prerequisites and precautions involved for performing the job, and plant conditions required for an item to be worked. Once tne foreman determines that an item can be worked with current plant conditions, he routes a package to the SWFCG describing the item, the system (s) affected, and the preferred work time if no specific system constraints exist.
3. The SWFCC scheduling coordinator develops a system-sorted list of potential work items to be reviewed by the SWFCG at a weekly input meeting. Each system sort is uniquely reviewed to determine the appropriate time for scheduling the involved work. Questions are raised by participating groups if the inp a sheets do not provide sufficient information to determine potennal plant affects from a proposed work item. Once an item is approved by the SWFCG for work, the item is placed on a work schedule for a given week, t

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4. Once scheduled, items on the weekly schedule are reviewed by the }

Control Room- Operations organization prior to the- scheduled' performance to ensure that the item remains acceptable for the current plant conditions. Once approved for work on a given day, the ~ work package is = given back to the responsible. foreman to implement. Implementation of a preventive maintenance activity beei ns with the responsible . maintenance foreman. The responsible foreman for the ob activity plans the work to be done on.a_given day. Part of the responsib1 foreman's implementation function is to ccuduct a pre job briefing with the technicians performing the dob. The discussions with the technicians include the precautions and-limitations associated with-a given task. The technicians are-then given the' i package to perform.  ; The work package given to the technician includes the k'R/JO developed for the -j ob .- The. k'R/JO -is taken to the Control Room, if necessary, to obtain Shift Foreman approval for beginning the job. Once the Shif t Foreman has approved the . job-start by reviewing-the ackage, the ob is begun by the technicians. The technicians performin5 the ob. complete to job in accordance with the process defined -by the k'R/J0. - l The work control' process at the Brunswick plant during. plant operation is a- . multi level', multi organization review , and implementation a process, = involving 1several individuals. -The process is dependent upon-~ the completion of these reviews and activities.in accordance with established standards and procedures. The standards Land - procedures for these reviews and activities, if - properly followed,:would have ensured that this procedure vould not have been performed. -However, the incorrect = reliance by some of the involved- individual reviewers allowed -a single, factor, the incorrect Summary Sheet, c to defeat tho' barriers established by the system. The January 25,:1991 SCRAM was, therefore the result of the~ failure of the reviews in this process to identify the required conditions - for performing this ob. A summary of the barriers in the work control process -and the failures wit in each barrier follows. - Onet of the " causal. factors contributing to-the inadequate reviews was . the . universal use of an incorrect procedure Summary Sheet. The body of the procedure = beingused:(OPICCPU001)-has,aPrerefuisite-Sectionwhichcorrectlyspecifies" that the procedure should on be per ormed while.the plant is in cold shutdown-or ? refuel condition.- Tho -- ummary - Sheet , . provided as an attachment to the procedure, summarizes the impact of the procedure and includes such items as a procedure description,- required plant conditions for : performance : of the procedure, alterations to plant. systems nas- a result of the procedure;: - annunciators : and affected ; indications , and ' possible Technical Specification-- Limited Conditions for Operation (LCOs) which may result from the performance of-othe procedure. The Summary Sheet is intended to provide - an' impact summary for tho' Shif t ' Foreman to use --in assessing shether the- proceduro can be performed under the conditions which . exist at- the time the procedure - is- planned to be - worked by the Maintenance organization. ' The Summary Sheet is not intended to be -substituted for the Prerequisites and Procautions stated in the procedure, and states so = in the heading- of- the page. The Summary Sheet for this procedure P . , - , , ,~ 4 ,> .----.4 v.7. c-,- . . , , - - - , , , . . . , , , ,, , -a N8C ^FOAM Mk U.S. NUCtEAR REOULAToHY COMMiS90N C# MOVED oVB NO. 3150 0104 ripi4[6: 4 DOS 2 , ESTIMAftD BURDEN (Y H FE99NSE 10 COMP YL WTTH THl$ INFORMAnON LICENSEE EVENT REPORT (LER) %E R ,LS'g g "M %*$',107ntntco,c,,yngem, TEXT CONTINUATION W sGIMM BRANCH PSMS. NMAR WAToHY CN$50N, WsHIN3 TON, DC 20555, AND 10 THE PAPERWORK HEDUCTON PFOJECT 13150 e1C4). of flCE Of MANAGEMENT AND BOO 3rf, W ASHINGTON.DC 2050) FAC1UTY NAME (1) DOCsIT LER NUMBER (6) FAGE (3) NUMBE R (2) ~ Brunswick Steam Electric Plant Unit 2 05000324 vtAR NvMoi a NuMor n 91 - 001 - 00 TEXT (tF MORE $ DACE ($ Ff00'REO, USE ADCNTONAL NRC FORM MAT 4 (17) incorrectly indicated there were no special plant conditions required for performance of the procedure, and this greatly influenced the mind set of the involved individuals both performing the procedure and reviewing the procedure for possible plant impact. The initial barrier in the scheduling phase that was defeated was the I6C foreman review of the work item. The I6C foreman initially submitted six similar calibration routes to the SWFCG for work. The calibration of computer point B022 was included in this package. The foreman had not appropriately reviewed the precautions and limitations for each package prior to submitting the SWFCG work request as required by Maintenance procedures. Once the calibration routes were input into the SWFCC system index, the items were reviewed at the weekly SWFCG " upcoming work" input meeting. Questions were raised by the cognizant Operations representative concerning these calibrations, and the Maintenance representative was requested to further investigate and provide input to the SWFCG on plant affects from these calibrations. The Maintenance representative reviewed the procedure Summary Sheets for the work and determined that two of the routes required the plant to be in either shutdown or refuel condition to perform. lie then returned the packages to the rasponsible I6C foreman to reevaluate the remaining packages for additional impact. The I6C foreman's second review of the work packages consisted of a review of the procedure Summary Sheets for the involved procedures. This was not an appropriate review of the involved work as dictated by Maintenance practices. Had the foreman reviewed the Prerequisites and Precautions section of the procedure during either his first or second review, as defined by existing Maintenance standards, he would have noted that the Prerequisites of the procedure for the calibration of computer point B022 required the plant to be in either a shutdown or refuel condition. The I6C foreman, following completion of his second review, returned the work of package the Summary to the Maintenance Sheet for theSWFCG representative. calibration of computer Thepoint packageB022. included SWFCCa copy!cw rev appropriately credited the work done by the I&C foreman, and thus relied on an inaccurate Summary Sheet in determining that the work was safe to be performed under the existing plant conditions. Once the calibration was placed on the SWFCG work schedule for the week of January 19 through January 25, 1991, the package was given to the Operations staff for a final review. The Senior Reacter Operator (SRO) reviewing the work for the upcoming day noted the required conditions as defined by the SWFCC package and the Procedure Summary Sheet. The SRO identified a potential concern 1 about the affects of the evolution on the Process Computer Periodic Core Performance log (P1), which monitors core performance parameters. A note was put on the package for the Technicians working the job to contact the dayshif t Shif t Foreman to identify the affects on P1 from this activity. The implementation phase of the work control process began on Friday morning, January 25, 1991. The I&C technicians received a pre-job briefing from the responsible I6C foreman for the calibration of the computer point. 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WASHINGTON. DC 20555. AND TO THE PAPERWORK REDUCT)oN PFOJECT - Q1$0-o104). of RCE OF MANAQEMENT AND BUDOCT. WASHINGTON.DC 70503 o FAC8tifY NAME (1) DOCKET LIR NUMBER {6) PAGE p) NuM6ER it) Brunswick Steam Electric-Plant Unit 2-05000324 YTAR NUMBER NUMBER 7 of 19 91 - 001 - 00 TTXT (IF MORG $ PACE IS REou' RED. U$[ ADDfTioME NRC FORM 3f4A'S)(17) technicians noted that the responsible 160 Foreman instructed them that thic was , a job with no plant impact. .The 160 foreman pre-job briefing did not include a -l review of the prerequisites and precautions section of the procedure, as required by Maintenance practice. The technicians proceeded to the Control Room to discuss the P1 note with the Shift Foreman. The Shif t Foreman had discussed the af fects of similar computer points with the Nuclear Engineering group earlier that week. The affects of these jobs were determined to be of no concern relative to the- P1 Process: Computer log. The j- Shift Foreman then reviewed the~ Summary Sheet, which stated that there were no plant required conditions,- and plant systems being altered were limited to making computer - points on the . Process Computer and Ernergency Response Facility Information System (ERFIS) inoperable. The Shif t Foreman signed the package and

  • instructed the individuals to obtain the concurrence of the Senior Shif t Forenan prior to starting work. The Senior Shift Foreman was also misled by the Summary

-Sheet.into believing that the job would not have any affect on the operating- , > unit, s The final barrier in the impicmentation phase was the I6C technicians performing the calibration. 1The two-I60 technicians assigned to perform the calibration focused on the procedure Summary Sheet for assessing the plant impact of-performing the procedure. =The mindsets--of-the technicians involved with the event were the result of the work on cornputer points earlier in the week, the . incorrect Summary Sheet statement, and the -pre job briefing held with the 16C = foreman just prior to starting- the job. As-a:renult, the technicians did not ensure the prerequisites of;the procedure were satisfied prior ~ to initiating the calibrationi ' The prerequisite step which e requires the plant to be in cold-shutdown or refuel condition prior to . performing the procedure immediately follows a. sign off for. shift foreman approval to begin the procedure. This step was overlooked by the -- technicians. As the final barrier, the failure of' the ' technicians to follow the procedure prerequisite step directly resulted in the: S CP.AM ,- the work control process of. the operating plant did not eliminate a In summary, work item t hat should have been performed only with the plant 'in- a shutdown or refuel-condition. The process-failure was a result of the inadequate reviews relied-upon in'the scheduling process.to ensure work-items.are performed only

during desired modes of-operation, 'and a failure of.the involved technicians and-foreman to ensure prerequisites of the procedure were satisfied. The inadequateL
roviews-were= due to a combination:of personnel failures to adhere to established s procedures and directives, and an incorrect procedure which led the reviewers to-believe-'that: this. particular calibration could be. performed durin6 unit-

. operation. . ABNORMAL TRANSIENT OCCURRENCES This section provides a summary of any system or component failures experienced, as well as. explanations of unusual occurrences that occurred during the plant-transient. The following items were reported in the initial red phone report as potential 'Y r w w .r - , , , . , ,. rpm- . , , - .,,y.#.., , m,~.4 , +, NFc 'FOAM M4 U S. NUCLt AR AEGOIATOAY COMM4SION AfvHOVED OMB 40,3150c104 E AfWS 4/30/92 , ESTIMATED BURDEN FT A Hf tron 5f TO CCAmLY W1TH Dc51NFOrtMAtiON """"""5'"*^'* LICENSEE EVENT REPORT (LER) "oMMENTS C FWGAFONQ BURD[N [$11MATI TO THE 40080 TEXT CONTINUATION M*== 58^~c".m u.s wu =wtAmnv umsm WA$HINGTON, DC 20555 AND TO THE PAFTRWOOA HfDUCnoN PRDECT (315DS104 of FtCE of MANAGEMENT AND BUDGTT. WALHINGTO%DC 20N." FACIUTY NAME 0) DOCKET LIR NUMBE A (6) PAGE [h -NUMM H (n Brunswick Steam Electric Plant Unit 2 trouErmAL REVWON 05000324 vrAR NUMNH NUMBER 8 of 19 91 - 001 - 00 TEXT OF MORE SPACE 15 40U1HCD. ULE ADCKnJNAL NHC FOAM %6A'S) (17) I l concerns that needed further investigation: l

1. Standby Cas Treatment (SBGT) System not starting and the Reactor Building ventilation dampers not closing.
2. Reactor Vater Cleanup (RCU) Group 3 isolations not occurring.
3. HPCI F006 valve not automatically opening and injecting.
4. Possible excessive closure time of the G16 F020 Dryvell Equipment Drain Outboard Isolation Valve.

Resolution of these items is discussed below. Additional concerns found during the event are also discussed. ShGl_ SYSTEM ACTUATION. CROUP 3 AND RB VENT DAMPfR ISOLAIIONS SBGT systern actuation, Reactor Building (RB) Ventilation damper isolation, and a Group 3 Primary Containment Isolation System (PCIS) valve isolation are initiated by the same instrumentation group for Reactor Vessel Vater level Low level. The involved instrumentation (four instruments: 2-B21 LT N024A-1, A 2, B-1, and B 2) provide trips at Reactor water level 118", decreasing. The logic is such that an "A" device must trip in conjunction with a "B" device to initiate the isolation or actuation. The as-left setpoints of the four trip instruments in this logic system range from 117. 6" to 117. 8" . A review of the data from the January 25, 1991 SCRAM for the wide range instrumentation determined that reactor water level dropped to be tween 116. 5" and 121" . P,ased on the fact that the level did not positively decrease to lens than the isolation / actuation setpoints during this event, and that simultaneous actuation of both the "A" and "B" channels would have to occur for the isolations/actuations to be completed, it is not considered unusual for the isolations and actuations to not take place as a result of the transient conditions resulting from this event. It is probable that more than one of the individual instruments did not concurrently trip to initiate the Group 3 and Reactor Building damper isolations, as well as SBCT system actuation; therefore, based on the predicted instrument responses and the lowest vessel level seen during this transient, these items are not considered unusual responses. HPCI/RCIC OPERATION A review of the Emergency Response Facility Information System (ERFIS) ~ data associated wich the operation of the HPCI and RCIC systems for this event has detemined that the systems performed as expected, with no concerns noted. The HPCI system automatically started, but did not inject into the vessel. This is an expected response due to the short duration (< $ second,) of the Low Level 2 (118" trip, 122" reset) initiation signal. HPCI cperated in the minimum flow mode until the F006 HPCI inj ec tion valve was manually opened for reactor level recovery, approximately one minute following the initiation signal. HPCI inj ec ted h4C FORM 36(d U.S NOCd AR RCOULAT04Y CoMMOON A14mJD CNB MO. 3150e104 , [ W 4tt:4/ES2 . ($TiMATfD BufCYN I4 R HEMONSE TO COMfiY Wf1H TH$ INFDRMA10N 5 LICENSEE EVENT REPORT (LER) M'yWlglv 'JRli"5JRrtly,10,y ,m.m ,1yo ,mn,3 TEXT CONTINUATION M^w^=' 68^* " m u 5 Nucu^a ntou'^' oat couve WAtH!N3 TON. DC 70555. AND TO THf PArtnwong fif DUCTON PFOXCT (3 %C104 orhCi of MANAGEMf NT ANO BUDGET. WASH 1NGTON DC FM) F AOUTV NAME (1) DTkri LI A NUM0ER tta i AGC 0) NUMitt A m Brunswick Steam Electric Plant Unit 2 ] , 05000324 YE^n N e rn Nmrn 91 - 001 - 00 ^' ext (IF MORE $ PACE 15 REQUI ACD. U$f ADO!fCNAL. N4C foHM Ef>A*$) (14 for approximately 2 minutes. Following level recovery, HPCI was transferred to the Pressure Control mode. HPCI was manually removed from service after a total operation time of 6.5 minutes. The RCIC system automatically started and injected into t.he vessel, as designed, in response to the nal. at 400 gpm until level wasLow Level 2 (approximately restored actuation sig/4.5 minutes) .RCIC injected RCIC flow was manually reduced over a period of four minutes. RCIC was manually secured and placed in standby readinces after a total operat.lon time of 8.5 minutes. 2-016-F020 IS01ATION TIME The Group 2 isolation signals received during this event initiate c?osure of the Dr valves (2ywell G16 Floor F003, and F004, Equipment F019, andDrainF020). Inboard Closureand Outboard of the F003, F004 luolation and F019 valves occurred within the expected closure times (approximate'.y 3.5 seconds). The F020 Equipment Drain Outboard Isolation valve did not indicate full closed until approximately 11 seconds af ter receiving the closure signal. The closure time was determined to be excessive compared to the other valves and exceeded the acceptance criteria found in Periodic Test (PT)-11.3, Drywell Drains System Valve Operability Test. The closure time was within the Technical Specification operability requirement of 20 seconds. An outstanding Gork Request / Job Order (WR/J0) exists for repidelag the limit switch lot the 2 G16-F020 valve. Troubleshooting nas determined that an intermittsnt problem with the switch causos a delay in the closed indication signal being received, due to a binding condition with the switch protective boot over the switch plunger. Periodic Test (PT)-ll.3 was performed to ensure the valve stroke tirae was within Technical Specification limiti and testing acceptance criteria. The testing determined that thu valve stroked in approximately 3.5 seconds, by indication. The switch worked nppropriately during the testing. The excess closure time of the F020 valve seen during the transient was determined to be the result of the intermittent problem with the valve limit switch. This is not believed to be an operability concern; however, parts are on order to facilitate the necessary repair. 2-B32 F031B VALYE Following the SCRAM, an attempt was made to restart the 2B Reactor Recirculation pump. The pump would not start as a result of Motor Generator Set 2B field breaker not opening. The 2 B32-F031B Reactor Recirculation pump discharge valve was being opened following the failed start attempt, when the valve bound up at approximately 70% open. The 2-B32 F0"lB valve is a normally open valve, but in closed in order to place a Recirculation pump in operation. The F031B valve is then opened after the pump is restarted. The s..ive failure occurred as the valve was l i N4C FORM 3Md U.S NUCJ AR Rt OVLATCHY COMMISSON APPfovtD oMB Mo 31n01D4 . CARRE$:CDOS2 . f 5TIMATED DURDIN Pt A AESIONSE TO CoWNY W11H TH'S INrOrwATON LICENSEE EVENT REPORT (LER) MC'$$&igfu""o',lf".$A7E Toinmcofes ANo emnts TEXT CONTINUATION MAN *utNT ea^Neesm u s NUcuan atovutonv eoMM4ssoN. WALH1NGTON. DC 20S55. AND TO THE PAIT RWoFW. REDUCTON P%ACT Q150C100 orhCE or MANAGEMENT AND BUDGET, WASHIN37+M 205A3 FADUTY NAME (1) DOCPIT u R NUMBER (6) FAGE Q) NUMbCR (2) Brunsvick Steam Electric Plant Unit 2 0$000324 vtan Nuvata Num n 91 - 001 - 00 TEXT (18 MDRE SPACE IS HEOth4ED. USE ADDrTcNAL NfC 7ofW X4A'$;(17) being opened following the failed 2B pump restart attempt. k'R/JO 91- ABRF1 was initiated to investigate this problem, An entry was made into the Unit 2 Drywell to inspect the F031B valvo. The valve showed signs of a packing Icak of approximately 120 140 drips per minute, and an inspection of the valve stem indicated galling near the packing gland, k'R/JO 91 ABTTl was initiated to further investir, ate and repair the valve. A valve manufacturer representative (Anchor Darling) assisted in the investigation and repair of the P032 valve. The stem of the F031B valve was found galled into the valve stuffing box area. The gall marks ran parallel with the stem travel and were located 180 degrees apart, oriented with the p pe run. The gall marks covered a distance on the stem of approximate y 20" on both the upstream and downstream sides of the stem. Based on available visual information and discussions with the valve Vendor, the BSEP Technical Support unit has determined the most probable root cause of the stem galling and subsequent valve failure to be the result of stem contact with the packing gland. The contact was most likely caused by gland misalignment and/or an introduction of foreign material into the gland / stem area. The initial galling occurred on one side of the stem, the downstream side, due to system flow and pressure. Stem galling on the upstream side of the stem was the result of a decreased radial clearance caused by the accumulation of metal shavings. Subsequent valve stroking would have increased the metal deposit in the failed area (s) of the stem to the point that the valve actuator would not deliver the necessary torque to open the valve. The valve travel thus stopped as a result of the provided over-torque protection. The F031B valve is closed each time the Recirculation pump trips to facilitate pump restart. The valve has been cycled approximately six times since the last refueling outage, The remaining three Recirculation pump inlet and discharge valves were inspected, with no signs of galling noted. No instances of stem galling associated with the now singic stage _ packing have been noted. The valves will be inspected during the upcoming Unit 2 outage to ensure that the root cause of the valve stem galling has been adequately assessed. RHR E11 F003A VALVE FAILURE DURING SHUTDOWN COOLING During shutdown cooling following the event, the RHR E11-F003 RHR llent  ; Exchanger Outlet Valve breaker tripped while adjusting the cooling flow during shutdown cooling. This is a motor operated gate valve used by Operations during shutdown cooling operations as a throttling valve to control the RHR flow to the vessel. Upon investigation, the valve motor was found shorted to ground, which resulted in the breaker trip. The l motor was subsequently replaced. i Use of the F003 motor operated gate valve as a throttling valve has been previously identified as a misapplication of use. Replacement of these _. . . ___ . _ . , . . ___ _ _ _ .__m_ - . . NT FORM 3f4k U A NUCLIAR REGULATORY COMMISSION ANDvfD OMB No. 31500106  ; DPIRtS: 4/20M2 - EST1 MATED BURDEN PE R RE5PONSE To COMPLY WtTH TWS INFoRMADON co g Coc LICENSEE EVENT REPORT.(LER) p,oygcgAg ro tnc ,, eon 9,,g9 nc , TEXT CONTINUATION MANAMNT BRANCH (% U.S. NUCLEAR RDWTORY CoMMISSeQN, WAthNGTON, DC 20$55. AND TO THE PAKR'ACRK REDUCTION PROJECT (315&e104). OFFICE of a4ANAGEMENT AND BUDGET, WASHINGTON DC 20503 ~ FACIUTY NAME (1) DOCKET LIR NUMBER (6) - PAGE (3) NUMBER (2) _ Brunswick Steam Electric Plant Unit 2 05000324 -YEAR NUMBER NUMBER 91 < 001 - 00  : TDT OF MORE SPACE IS REoVIRED, USE ADolTioNAL NT FORM 366A%) (17) , valves has been identified in proj ects G0010B -and G00100, Plant Modifications 90 035 and 90 034 for both Unit 1 and Unit 2. 2B MC SET ' n A field breaker in the 4160 volt breaker compartment of the 28 Motor Generator (MG) Set was found tripped following an attempted recovery restart during the - SCRAM recovery, and VR/JO 91 ABPW1 was initiated to investigate this_ concern, Troubleshooting was unable to determine the cause of the breaker failing to latch, Subse were successful.quent attempts to latch the breaker during troubleshooting 2A MG SFJ During the Unit 2 SCRAM recovery efforta. the 2A MG set motor breaker immediately - tripped open when the recircutation pump was - given a~ start - command. Troubleshooting identified a temperature switch -(mercury contact type)'which was closing each time the-switch contact saw vibration from the motor receiving "in rush" current, similar to that expected during a i start. -The switch was' repaired under WR/JO 91 ABPUl. SPURIOUS PROCESS COMPUTER OPEN INDICATION ON SRV B21-F013K During review of the Process Computer data for the 1/25/91. transient, the printout-showed a momentary-lifting of SRV B21 F013K;-however, no other indications were seen that- this : SRV lif ted during the transient. The maximum reactor pressure:seen for this event would not require a SRV to open. In addition,- the process computer had shown false SRV F013K opening indications-prior to-the SCRAM. The SRV lif ting indication was therefore . considered to.be a ' false indication generated .in the Process- Computer . 5 logic. The reason for_the false indication is being investigated. IMMEDIATE CORRECTIVE ACTIONS Immediate corrective actions taken as a result of this-event included:

1. Capturing-work in progress and organization of the SIIT to identify -

event anomalies and root cause,

2. Stoppage of Computer point calibrations. -i
3. Suspended' Shift Foreman reliance on the use of Summary' Sheets.
4. Communication of the event with plant personnel, including a

~ briefing on the event and the -personnel- errors involved with the event. To allow for Unit restart, a' Recovery Action Plan was developed by the SIIT, -including the following items: L NrC FORM Mh U fx NOCLE AR RCGJLA'ORY coMMS$ON APP 80VED oMB DD 31%e104 i KPtRf 6 4/30S7 . ESTIMATIO BUTOCN F-(R FE$80NSC TO CCMPL Y Wff H THit INFO 4MATON LICENSEE EVENT REPORT (LER) cE't$ '!$8$'$lu""IAs'$u'AS To THc ruonDs aD nn > TEXT CONTINUATION u^"^="1 *"c" e 5m u $ =^R acw'*'on' covano". WA!iH!NGTON, DC ?OM5, AND TO THE PAMHWOht Ftf.DJcTON PWUE C1 (315Ge104). oHICE OF MANAGF MENT AND buDGif, elASHtGTONIC 7:601 FAO UD MME (11 DCCKET LER NUMMR (6) FA3f(h NUMM R (7) Brunswick Steam Electric Plant Unit 2 05000324 YEAf4 NUMM R NUMM R 91 - 001 - 00 TEX'(1 MORE Sr' ACE lS I4EQUiRED, USE ADoITIONA, NlC 804M V4A'S)(17)

1. Troubleshoot and resolve probicas with the openinr, of the B32 F031B Recirculation pump discharge valve (WR/JO 91 ABTil 6 EER 91 0036).
2. Resolve problem with the 2A Recirculation pump motor breaker (VR/JO 91 ABPU1 6 TSM 91-074).
3. Resolve problem with the 2B Recirculation pump start sequence (WR/JO 91 ABPW1 6 TSH 91-074).
4. Restore the Unit 1 Startup Auxiliary Transformer (SAT) to service.
5. ' Repair the Suppression Pool Temperature Monitoring System (SPTMS).
6. Resolve the RHR F003A valve problem (VR/JO 91 ABSGl).
7. Once the above items have been satisfactorily performed, obtain duty plant manager approval for startup.

Those items were completed by 1/30/91, priot to restart. ADDITIONAL CORRECTIVE ACTIONS In addition to the corrective actions associated with the SIIT Startup Recovery Action Plan, additional long term corrective actiuns were considered by the SII'r as a result of this event.

1. Investigate the spurious open indication on the process computer for SRV B21- F013K.
2. Investigate the elimination of the 40% steam flow isolation, or other action to reduce the probability of spurious Group 1 isolations due to failure of the trip unit to reset. A half Group 1 isolation was received during the SCRAM recovery on this event due to the 40% steam flow isolation (a Unit 2 function only), and has previously been identified in LERs as the contributing factor to other spurious Group 1 isolations. The isolations are not considered to be a safety concern, but may create an annoyance during operator recovery efforts following a SCRAM.
3. Investigate whether the Recirculation pump discharge valves can be shut af ter a Recirculation pump trip to maintain Recirculation loop temperatures and facilitate a pump restart.

4 Review and revise maintenance procedures having summary sheets to ensure that the prerequisites and equipment actuations and limitations accurately reflect the procedure impact.

5. Maintenance will review and revise the periodic maintenance (PM) routes which presently have 18 month frequency requirements to

, " Refuel" ("R0") frequencies if determined that plant operating ! conditions other than "RUN" are required for the performance of the 1 NRC FORM 35 U $. NLCLEAR REQULATORY CoMMcSSION aPPROWO oMB No,3154e100 EXPIRES: Of5V32 ESTIMATED BURDEN FER RESPONSE TO COMPLY WTTH THIS INFoRMATION ^ LICENSEE EVENT REPORT (LER) gy(5*'f g g' g o" & ' g M ,i ,,oysencCono3 ,yo nc,cn,, TEXT CONTINUATION MANAGEM BRANCH F530). UhvCMAR RFoOMTORY CoMMISSON, WASHINGTON CC 20555, AND To THE PAFERWORK RE00CTioN PROACT n156e104). OFFICE OF MANAGEMENT AND Buo3CT, WASHINGTON.DC 20503 FAC1LITY NAME (1) DOCKET MR NUMBER (8) PACE (h NVMBER (2) Brunswick Steam Electric Plant Unit 2 BCOUENTIAL - REWJON 05000324 vtAR NUMBER NuMBcR 13'of 19 11 - 001 - 00 ITEXT (tF MOP 4 5 PACE IS REQUIRED, USE ADO (TONAL NRC FORM 366A'$) n1) procedure. The plant Technical Specifications equate "18 month" frequencies with " refuel" '("R0") frequencies in terms of required ' timetable for-the performance of surveillances/ testing. Some of the i Periodic Maintenance (PM) _ routes which are "18-inonth" frequencies would have no impact on the plant if performed during _ power operation. Otber testing / calibrations, such as the one in this-event, have -requirements that the unit be in an op"erating condition other than "RUN". Designating such-PM routes as P_O"-would add an additional barrier against the performance of maintenance activities -requiring that- the plant be in - a shutdown condition. during - operation. This designation is made by the Maintenance organization during development of a route code. 6.. Brunswick : Site - Procedure - (BSP)-43 is 'being issued to enhance existing SWFCG guidelines. The procedure has been strengthened to require more than one method of verifying plant impact of planned-work,_ such as review of drawinga, summary sheets, procedure prerequisites, and system descriptions.

7. A Human Performance 2nhancement System (HPES) evaluation'is being conducted. for the personnel errors involved with this event, to determine causal' factors associated with the personnel errors, as

-well as determining if additional corrective actions may be  ! necessary to prevent recurrence.

8. -The event and event causes. are to be communicated to appropriate  !

site personnel. 'rl 9 .- . -_ Investigate the feasibility of adjusting the setpoints of various actuations instruments to prevent the partial actuations that are - prevalent on a high power SCRAM. This is a. recurring problem during this type transient. -- -10. Plant Management established a Task Force compris d of working level-individuals c directly involved -in ' this and othar events involving-personnel errors, to investigate potential underlying causes-recent events ' involving personnel error. The Task Force composition is intended : to provide Plant Management -with a perspective of event -causal factors from persons actually-involved with an event, along with recommendations to help prevent recurrence of similar personnell errors at the Brunswick plant. . The Task Force was chartered to:  ; provide recommendations which would result in plant " personnel ' consistently performing scheduled and emergent = work _ activities -in. -compliance with accurate site-approved -work control. process requirements and expectations." A presentation of the Task Force-findings to Plant Management was made on: February 22, 1991. Upon completion of the HPES evaluation, _ Plant Management will review the recommendations of this effort and the Task Force to determine if additional corrective actions are necessary to help prevent recurrence of similar events. A supplement to this LlR will be submitted following Plant Management review and l NRC FORM 366k .. U.$. NUCLEAR REQULAToRY CoMMi$$CN AKPROVTD oMB NO. 3154e103 f; * - -E R RES:4 *t0/32 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THtA INFoRMATION LICENSEE EVENT REPORT (LER)' CWy c M jg'g j@ #$n $ 1oyuc, gono $wontmn3 TEXT CONTINUATION MANWMENT BRANCH PSR V.$. NWAR REWLAW MMMISSON. W%$HtNGTON. DC 20555, AND TO THE PAPERWORM Rt. DUCTION PRCUECT (3150 0104). OFFICE oF MANAaf MENT AND Buo0ET. WASHINGTON.DC 20503 - FACIUTY NAME (1) DOCKET LER NUMBER (6) PAGE (3) NUMBER (2) Brunswick Steam Electric Plant Unit 2 y . 05000324 YEAR NUMBER NUMBER 14 of 19 91 - 001 - 00 1 TEXT ($ MORE SPACE t$ REQu' RED. USE ADomONAL NRC FORM 366A'S)(17) assessment of the recommendations. The supplement will be. issued by April 1, 1991, and will also-provide an update and schedule for completion of outstanding corrective actions identified by the SIIT. EVENT ASSESSMENT 1 The Unit was operating at a maximum power level at the time of the event, The Unit response for this - event .w as within the bounding parameters of the corresponding FSAR Chapter 15 Feedwater Controller Failure event involving an increase in feedwater flow to the Reactor vessel. The equipment concerns identified during the event did not significantly hamper operator ability to achieve and maintain the reactor in a shutdown condition. This event would not-have been.more severe under any other credible and reasonable conditions. Other -SCRAMS in the past two years at Brunswick that have been: directly - contributed to personnel errors have been reported in LERs 2 89 09 and 2 90 09. This event is not similar to the other SCRAMS in that the work control process o that did not - stop the work from being performed - has significant existing a prograrmaatic barriers in place to prevent such incidents from occurring, and an excellent track record for preventing this type event. As identified in the Corrective Action section,. plant Management is disturbed with the number-of plant system challenges and transient occurrences that have occurred in the past year resulting, from personnel errors involving experienced and skilled personnel. The Task Jorce of event involved personnel is believed ~ to be an essential- step in determining causal factors involved in Brunswick i personnel-errors; as well au determining effective corrective actions which would reduce the frequency of occurrence of this type event. , - This event was initially reported to the NRC as a one hour red phone report under 10CFR50.72(b)(1)(ii), the -plant -being in a degraded; condition, and 10CFR50,72(b)(1)(iv), an event that results or should have resulted in Emergency Core Cooling System (ECCS) discharge into the reactor coolant system as a . result - of-. a ' valid signal, and - the four hour criteria, - 10CFR50;72(b)(2)(ii) . -an RPS actuation, ' including plant . SCRAM, . Additionally, immediate event review-determined that the event was also reportable under 10CFR50.72(b)(2)(ii), ESF actuations. These calls were made within=their required reporting times. Initial ~ event response conservatively reported the plant to be in a degraded condition relative to potential llPCI:and SBGT system failures, and that the event: 'should have resulted in ECCS injection into ttie vessel from a valid signal, ' As ~ described-in this LER, event review indicated that the=llPCI and SBCT systems-operated as designed during this event,- an expected result of partial instrument -logic actuation and instrument permissive time constraints. Therefore the plant = s was'not in a degraded condition at any time during this event, nor should llPCI have injected -into the vessel, as designed time permissives for injection were not satisfied (approximately 13 seconds) . This LER is not reporting this event relative to the plant being in a degraded condition or required ECCS actuation. The situation created by the partial actuations seen during this event and similar high power SCRAM transients that led to the initial conservative call of-degradation and ECCS injection is being addressed by Corrective Action item 8. .. .. . .-, ~ ~ . ~ . . . . . . . . . - . . . . . - . . . --- ~, - - . ~ - ._a. . . ~ . ~ . . . . . ~ . - NRC FORM 36(d . -_ , _U.s, NUCLEAR REQUL ATORY CoMMIS5 ton APPft0VEo CMB ko. 31104104 - N* EXFifES:4/J0.W . FSTIMATED BURDEN MR RESf0NSE To COMALY WITH 763 INFO 4MAllON LICENSEE EVENT REPORT (LER) co',yj"gnggt$iEVEN co t $$ To nie mCoRos ANo Rcroms ? TEXT CONTINUATlON MANAGEMENT BRANCH (P 530), U.S. NUCLEAR REGULATORY COMMIS5 ton. WASHINGTON, DC 205S5. AND TO T HE PAKRWORK REoVCTKW PFCK.CT (31540104). OFFICE oF MANAOCMENT AND BUDOCT, WASHINGTON.DC 20$03 FAC1UTY NAME (1) - o0CFIT LCR NUMDCR (6) PAGE (3) NUMBCR (2) Brunswick Steam Electric Plant Unit 2- g 05000324 vtAR NuMoto NuuerR . . 91 - 001 - 00 Text (ie uoRt spACc is ReoViRco. use AcomONAL NRC FORM 360A'S)(17) ' E11S CODES SYSTEM / COMPONENT fQE Automatic Depressurization System

Reactor Core Isolation Cooling System BN Residual-Heat Removal / Low Pressure Coolant Injection B0-Standby Gas Treatment System BH Emergency Diesel Generator EK Reactor Protection System- JE Reactor Water Cleanup System CE Standby Liquid Control BR Turbine Stop Valve TA/ISV Reactor Feed Pump SJ/P Startup Level Control Valve SD/LCV Process Computer 10/ CPU Reactor Recirculation Pump . . RR/P Emergency Response Facility Information System IQ Reactor Building Ventilation Isolation Dampers VA/ ISO -Drywell Equipment-Drain-Outboard Isolation Valve JH/ ISO-Reactor Recirculation Pump Discharge Valve RR/ ISO RHR 11 eat Exchanger Outlet Valve - B0/ ISO ' Reactor = Recirculation Motor Generator Sets RR/MG . Safety Relief Valve */RV

  • No EIIS-System. Identifier Found- l 1

I I l -1 l 1 I . . . ~ ~ . . . , _ - , - . _ - - . ~ . . , . , . . - - _ . - - , . , , ~ ~ ,.+. . . -NN FORM 3E$'A . , . U.S. NUCLEAR REGULATO64Y CcMMi$5CN ARMOVED OMS NO. 31540100 EXRRES: 4/IQ,92 - EST1 MATE 0 0UADEN PEA RESooNSE To COMPLY WITH T@$ 1NfoHMATON ~ -- LICENSEE EVENT REPORT (LER) C co 'yRM'g"gcl,,"nu'A TO THE AENNO REPORTS TEXT CONTINUATION MANWMENT B8ANCH (ASE U,$. NUCM AR HEQULATORY COMMSON, WASHINGTON, DC 20$$$, AND t'0 THE PAPERWOh HEDUCTON PRCUECT (31504104). OFFICE OF MANAGEMENT AND BUDGET. WA5%0 TON.DC 20503 FAOUTY NAME (1) 00Ckri MR NUMDEH (6) PAGE (3) NUMBER (2) Brunowick Steam Electric Plant-Unit 2 sEauE"At nEwsoN 05000324 vEAR NuMeEn uvyatn op 16 of 19 91 . 001 - TEXT (IF MORE 5 PACE 15 REQUtAED, U$E ADDITCNAL NRC FOAM 366A'S) (17) ATTACHMENT 1 SEQUENCE OF EVENTS--JANUARY- 25, 1991 SCRAM The following is a SEQUENCE OF EVENTS for the January 25,1991 SCRAM, The times shown below are referenced to the Process Computer clock. Event times taken from the Emergency Response Facility Information System (ERFIS) printout have been modified oy subtracting 34 seconds. 08:10:52. I&C technicians lift the wire on terminal board DD 84, which causes a loss of- the "A" Feedwater flow signal into the Feedwater Level Control System (WLCS) . WLCS increases the speed of both RFPs. 08:10154 Computer alarm received on ' Condensato Filter Demineralizer high differential pressure, No dueannunciatorto increased flow . in the was received, as-Condensate /Feodwater path, the card for that window was pulled, taking the annunciator out of service. 08:11:00- Alarm received on High Reactor Water 1.evel. 08:11:04 APRM upscale . alarm received. Flux increase is due to the feedwater flow increase and cooler moderator temperature. Peak APRM reading is 106%. Condensate Booster . Pump "C" .sutomatically started. - 08':11:09- Reactor Water Level reaches thE. high ? level Turbino Trip-setpoint. The- main Turbine and both "eedwater turbines are .- - tripped automatically. Peak Level reached during the event is 208". Turbine Stop Valve closure generates a Reactor SCRAM signal-and rod insertion begins. 08:11:10 - All .four Turbine Control Valve Fast Closure - (TCVCF) .subchannels' detect low pressure and also receive a load reject SCRAM signal (a Unit 2 function only), Reactor high pressure SCRAM signal on channel 'A2" received .5- seconds). Peak pressure- -momentarily observed is--_1032. psig,-(approximately Examination of calibration records reveals that reactor pressure trip unit--B21-N023C-1 for subchannel A2 is set slightly lower than the other three-subchannels, within the range of ' observed - pressure. This response is therefore acceptable and expected. Generator-Output Breakers open. . g w __e - '-u___ _ . .-a --a . :, , , - < - .w. , _ ., -,.,,_m-.,... ., +_.,,n ,..-,,,74,,,,y_.,7.,-,,a ,,, ,m,m,yn9 7 N40 FoliM 25(A . U.S. NUCLfAR hiOU.ATOHY COMM:S$ON AFP40Vf D OMO NO 31:6e104 LFL405 0/kVW (ETIMATED 3URDEN PCH HCtPONSE TO CoM5=LY wtTH THt$ INFORMAT)ON LICENSEE EVENT REPORT (LER) E'yQ,M5'gl"a?'t3%^a, c ,o ,st ntoonos ,No ni,yrrs

  • ^N^=w' ea^wc" esn u.s. NuCue nrouTony CouMeioN.

TEXT CONTINUATION WASHINGTON. DC 20555. AND TO THE F AFTRWof4K REDUCT;CN P5CJECT C3150e1D4). OFFirE Or MANAGEMENT AND BUDGET, WASH!NGTON,DC FW) F ADU1Y NAME (1) DOCAET LE A NUMBER ($) PA *.f, (3) N_ UMBER (?) ~ Brunswick Steam Electric Plant Unit 2 , 05000324 Vf^a N *"t " N=a 91 - 001 - 00 TEXT (if MDRE tPACE 15 REQVl4ED USE ADOmoNAL NRC FORM %f>A'S)(IT) ATTACHMENT 1 (CONT.) 08:11:12 Neutron Monitoring System SCRAM sinnals generated on channels A2 and Bl. This trip signal is the result of a combination I logic of APRM downscale and IRM " Upscale Inop" . Unit 2 has been maintaining IPJi F (A2) and IRM G B1) "Inop." due to J equiprcent problems; therefore, these neutron monitoring trips are anticipated responses when the associated APRM drops below the downscale alarm setpoint. Turbine bypass valves are fully open. 08:11:17 Loss of Feodwater turbines causes a decrease in Reactor water inventory. All four subchannels detect low reactor water level and generate scram signals, l.ow reactor water level generates the group 2, 6 and 8 isolation signals. The group 8 valves do not isolate, as they have been isolated prior to the SCRAM. 08:11:20 21 Valves 2 G16 F003, F004 and F019 close due to the isolation signal. 08:11:22 Following procedure, the Control Operator inserts a manual SCRAM in both channels and transfers the reactor mode switch from "RUN" to " SHUTDOWN". A half group 1 isolation signal is received in subchannel "A2". The signal is a 40% steam flow signal . enabled by removing the mode switch f rom "RUN" . Trip unit B21 N008C 2 is observed to be in the alarm state. 08:11:39-43 All four subchannels of the SCRAM Discharge Volume (SDV) level detect the HI HI levei, 08:11:40 Reactor water level reaches the Low Level 2 setpoint, and Alternate Rod Injection is initiated. 08:11:46 SDV Hi water level Rod Block is generated. At approximately this tine, HPCI and RCIC receive automatic initiation commands, and the Anticipated Transient Without SCRAM (ATWS) circuitry trips both Reactor Recirculation Pump Motor / Generator (MG) Set drive motors. 08:11:50 The RCIC injection valve begins to open. 08:11:51 Outboard Drywell Equipment Drain Isolation Valve 2 G16 F020 is closed. Closure time exceeds expected time (11 vs. 9 t,aconds) per PT-11.4 Acceptance Criteria. ~ Amuo oMa Na m cu,. u s. NoaEAn ncoutArony comeseN (mnts 4mr vc tonM 4 i EflMAitD BURDLN PEH HLSPONe,t 10 COMPLY W11H T>tt tNronyAioN c" #5 $""5* "*^"o "vc'NTs" cov m"cwoNo euma t5wAfr io THc ntconos ann am I LICENSEE EVENT REPORT (LER) mAmt enAs:" (nsx>L u 5 Nuett An ntauti30nv coMM 550N. l WASHfNGTON, DC 20%5, AND TO THE PAPLAWORK HC000 Tion F4 CEC 1 TEXT CONTINUATION (315N11D4L orFIC[ OF MANAGEM(NT AND BUDGri, W ASHINGTON.(C 2i603 FAnt (3) LE R NUMBf A (6) (XDD FA0ldTY NAME 0) NUMlO (?i SE QUEN fiat HCviscN Brunswick Steam Electric Plant Unit 2 05000324 W AR_ NUMBER NUMacR 18 of 19 i 001 - l 00 91 - TEXT (1F MD4L SPA 0015 nEQUIRED, USE A'O F-AAL MC FOAM 366A'S) (17) ATTACllMENT 1 (CONT.) The turbine begins to llPCI Turbine Stop Valve begins to open. injection valve does not 08:11:54 increase to rated speed. The llPCI open, as reactor level is above the initiation setpoint IIPCI, when per design, the " valve open" permissivF s are catisfied. 2 will not inject for approximately 13 seconds following initiation due to tt e permissives which must be satisfied to open the injectica valve. RCIC system reaches rated injection flow. 08:12:04 momentary opening of Safety ERTIS computer printout shows This is not an accurate 08:12:14 Relief Valve (SRV) B21 F013K. Valve setpoint was never accounting of the valve actuating. indication showed the valve opening. reached, and no other this indication prior to The computer had been printing out the event start. the llPCI injection valve to assist 08:13:05 Operator manually opens RCIC in restoring level. llPCI exceeds rated injection flow. 08:13:20 the low level 08:14:27-35 Reactor SCRAM setpoint. water icvei has been restored aboveThe four low le time. Operator observes that Reactor water level has been restored 08:15:08 above the low icvel SCPM setpoint and manually initiates closure of the HPCI injection valve. Reactor water level goes above the low level alarm actpoint. 08:15:33 08:16:06 Operator storage tank (CST). manually opens llPCI discharge path to the condensateT mode. Hi Reactor vessel level alarm setpoint is reached (192"). 08:16:51 08:17 Operator secures pressure control mode of ilPCI by closing the discharge to CST. 08:18 operator manually trips the llPCI turbine. Operator opens the RWCU reject valve to establish a method of level control. 08:20 Operator closes the RCIC injection valve. 08:22 Operator resets the group isolation signals. ~~ - - - - _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ 0 NEC* FonM $$f . U 5. NUCLEAR REGULAfDRY COMM$5 ton NTROWO OMB NO. 31h0104 LKIMIE $: 4!3081 y f 5TIMATLD BURDFN PCR RESIONLE TO COMPLY W1f H THit INFORMAliON LICENSEE EVENT REPORT (LER) gi$'*3 Ofj'gu""4',3 y!?t ,o mt nocOn33 Ayo nocy,, TEXT CONTINUATION *^N^"NT S"" m U.S NUCdAR NATORY CoM2$50N. WASHINGTON, DC 20%$, AND TO THE PAPL RWORK RCDUCTioN Pf0 JECT (31W0104). C# F'C0 0F MANAGCME NT AND BUDGET. WASH:NOTON.DC TJ$C) F ADUTY NAMF (1) CO3f7 L! R NUMBC11 (6) PAGE (3) NUMBE R (7) Brunswick Steam Electric Plant Unit 2 , , 05000324 YEAR NUMBER NUMfiCR 91 - 001 - 00 TEXT (1F MORE $ PACE l$ HLQUIREU, LISE ADCETONAL NRC FORM 3E4A*S)(17) ATTACHMENT 1 (CONT.) 08:24 Operator secures the RCIC turbine and the "A" and "C" heater drain pumps. 08:28 Hi Reactor level alarm resets. 08:29 Operator starts the 2A RPP and places the Startup Level Control Valve (SULCV) in service. 08:30 Rod position scan performed.; all ods confirmed to be fully inserted. 08:40 Operator attempts to start the 2A Reactor Recirculation pump. The M/G set drive motor breaker closes, then immediately re-opens. 08:45 Operator attempts to start the 2B Reactor Recirculation pump. The M/G set starts, but the field breaker does not close. The drive motor breaker is then re-opened. 08:47 Alternate Rod Injection is reset. Automatic function of SCRAM channels "A" and "B" are reset. Manual SCRAM Channels "A" and "E" are reset. 08:49 Operator momentarily places the modo . 4 witch in " REFUEL" . Operator realizes error, and immediately returns the switch to " SHUTDOWN". Manual SCRAM signal is received on both channels. 08:50 Operator resets the manual SCRAM signals. 08:51 Attempt made to open B32-F031B, the Reactor Recirculation pump discharge valvo. Valve stopped with intermediate position indication. Thermal trip reset, but valve would not continue open. Work Request / Job Order (WR/J0),91-ABRF1 initiated. 09:01 HPCI Auxiliary 011 pump is secured, and HPCI restoration to standby alignment is begun. 09:07 Drywell Floor and Equipment Drain Inboard and Outboard Isolation valves are opened, i 10:24 RFP 2A is secured in accordance with GP 05. Remain h of i shutdown proceeded in accordance with GP-05. I 1 1 I

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