ML17262A292

LER 90-017-00:on 901212,reactor Trip Relay de-energized & Reactor Tripped When Dc Switches in Distribution Panel Opened.Caused by Procedural Inadequacy.Procedure Change Process Being evaluated.W/910111 Ltr

ML17262A292
Person / Time
Site:	Ginna
Issue date:	01/11/1991
From:	Backus W, Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-90-017, LER-90-17, NUDOCS 9101160181
Download: ML17262A292 (20)

Text

ACCELERAII=D D!S~RIBU t ION DEMONSTP 4.TION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9101160181 DOC.DATE: 91/01/11 NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet 'Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION

'BACKUSjW.H. Rochester Gas & Electric Corp.

MECREDY,R.C. Rochester Gas & Electric Corp.

RECIP.NAME RECIPIENT AFFILIATION . R

SUBJECT:

LER'90-017-00:on 901212,reactor trip relay de-energized &

reactor tripped when dc switches in distribution panel D opened. Caused by procedural inadequacy. Procedure change process being evaluated,W/910111 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR I ENCL SIZE:

TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.

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NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244 A RECIPIENT COPIES. RECIPIENT COPIES ID CODE/NAME LTTR ENCL. ID CODE/NAME LTTR ENCL PD1-3 LA 1 .1 PD1-3 PD 1 1

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Cit, January 11, 1991 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Subject:

LER 90-017, Opening of DC Switches (Procedural Inadequacy) Disables Manual and Auto Actuation of Safeguards Sequence Initiation Causing a Condition Outside the Design Basis of the Plant R.E. Ginna Nuclear Power Plant Docket No. 50-244 In accordance with 10 CFR 50.73, Licensee Event Report System, item (a)(2)(ii)(B), which requires a report of, "any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded, or resulted in the nuclear power plant being in a condition that was outside the design basis of the plant", the attached Licensee Event Report LER 90-017 is hereby submitted.

event has in no way affected the public's health and I'his safety.

Very ix ugly youk. 8, Robert C. Me redy

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xc: U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna USNRC Senior Resident Inspector Pu g p(g.

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LIC INC C I CONTACT SOA THIS LCA (ill HAUC TCLCTHOHC HUMSIA Wesley H. Backus AAIACOOI Technical Assistant to the Operations COM5LCTC OHC LIHC SOA CACH Mana er COMM>>tNT SAILUAC OIICAIIIOIN THI~ ACMAT (ill 31 524- 446 CAU5C SYSTSU COMMHCHT UANUYAC C~"ILC "'">'.>~~:.-,1 Ull lrSICM COMMHCHT MANVSAC CMATAIL g g>~$~(5(ÃrP A~

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On December 12, 1990, at 2310 EST, with the reactor at approximately 3%

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full power, the Control Room Foreman opened two cd by a Mainte>>.rce procedure, cav~inL) disabling of manual (pushbutton) and automatic actuation of the safeguards sequence initiation.

The two DC switches were closed, as directed by the Maintenance procedure, approximately twenty (20) minutes later, restoring manual (pushbutton) and automatic actuation initiation.'he underlying cause of the event was procedure inadequacy due to insufficient attention to detail.

A Extensive corrective actions are being taken to prevent recurrence, including communication of management expectations, HPES evalua-tions, identifying procedural'nadequacies, and a comprehensive upgrade of the procedure change process.

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IIAC SIUttt SSSA V.S. IIVCLKAAASOVLATOAT COUUISSIOII 19451 LICENSEE EVENT REPORT ILER) TEXT CONTINUATION ASSAOV50 OU9 AO 515OMIOS 5IISIA55 '9/SI I95 SACILITY IIASIC ill OOCKST IIVUSSA ITI LTA AVSSICA ICI ~ AOS ISI SSOVSIITIAL IIS U IS IO It U 1 lI U R.E. Ginna Nuclear Power Plant TTXT III'IIOIS AUSS tI ISSIUSS. USs NASOSIUS ArAC IItttlt WS'll I ITI I Ol 24 49IQQ1 7 00 Qi5oF l i5 The Control Room operators immediately .performed the applicable actions of E-0 (Reactor Trip or Safety Injection) and ES-0.1 (Reactor Trip Response) and stabilized the plant in hot shutdown. I After completing the applicable steps of E-0 and ES-0.1, the Control Room operators completed their part of M-48.14, by closing the two DC switches that had been opened in step 5.5.1 of M-48.14. This was accomplished at approximately 2330 EST, December 12, 1990.

The oncoming SS, who had been in the Control Room during this event, resumed the evaluation of the consequences of alarm L-31 after'plant conditions had stabilized. (The cause of the alarm had already been determined. ) He performed another review of M-48. 14 and called other knowledgeable members of the plant staff at their homes (at approximately 0100 EST, December 13, 1990) to discuss his concerns about the effect of opening these two DC switches. After receiving confirmation that'his concerns were legiti-mate, he made the proper notifications to higher supervision and the Nuclear Regulatory Commission (NRC) .

3.NOPERABLL'TRUCTURAL&s COMPONEN'1'6 s OK SYSTEI'sh THA'J.

CONTRIBUTED TO THE EVENT:

None.

D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED-None..

E. METHOD OF DISCOVERY:

The event was made apparent during the oncoming SS review of the consequences of Control Board Alarm L-31 (Safeguards DC Failure) and subsequent discussions with knowledgeable plant staff.

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Factors that influenced operator actions, during the event were as follows: v The Control Room operators questioned step 5.5.1 in procedure M-48.14, but information in M-48. 14, the DC switch labels, and Alarm Response procedure AR-L-31 did not provide sufficient operational information to determine the consequences of opening these two switches.

o The Control Room operators had confidence in a Plant Operating Review Committee (PORC) approved procedure that had also been reviewed by the Electrical Planner.

As the event was over prior to discovery, no operator actions other than normal were performed.

G. SAFETY SYSTEM RESPONSES:

None.

XXI. CAUSE OF EVENT A. IMHEDlATE CAUSE:

A condition outside the design basis of the plant was caused by the disabling of manual (pushbutton) and automatic actuation of the safeguards sequence initiation (i.e. auto and manual SI).

B. INTERMEDXATE CAUSE:

The disabling of manual (pushbutton) and automatic actuation of the safeguards sequence initiation was-caused by switch gl2 in the 1A DC Distribution Panel and switch g9 in the 1B DC Distribution Panel being open at the same time. Both of these panels are on the back of the Main Control Board.

MAC >414 999A (9WI

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0 HRC >wiA SAEA V.t. IIVCLEAR REOULATORY COMMISSION IE4S I LICENSEE EVENT REPORT ILER) TEXT CONTINUATION OME HO i'ttROYEO SISO&I04 EIItIRES 'EPICS AACILITYIIAME (II COCKET iIUMEER IEI LER HVMEER Iti tAOE ISI SEQVERTiAL AtVitiOH M EA HQIJ EA R.E. Ginna Nuclear Power Plant 5Io(oIo gI4 OI1i7 0 90F15 TEXT IA'i>>itA>>i>> e ~. v>> AiAAOCrWNAC Aiiiii~'el I ITI o 4 9 0 0 The effect of the potential delay in actuating safeguards equipment upon those events analyzed in the UFSAR was evaluated. The accidents effected by this action are those accidents which result in depressurization of the primary system causing SI. These are primarily the following:

0 Feed Line Break (FLB) 0 Steam Generator Tube Rupture (SGTR) 0 Small Break Loss of Coolant Accident (SBLOCA) o Large Break Loss of Coolant Accident (LOCA) o Small Steam Line Break (Small SLB) o Large Steam Line Break (SLB)

An analysis of these accidents was performed to determine the effect of the disabling of manual (pushbutton) and automatic actuation of the safeguards sequence initiation with the following results:

Feed Line Break t

This accident was analyzed by the Ginna Updated Final Safety Analysis Report (UFSAR) as a heat up event with auxiliary feedwater available in ten (10) minutes. As a heatup event, RCS pressure never decreased below the SI setpoint, but rapidly increased above the SI pump shutoff head. Therefore, SI was not necessary and auxiliary feedwater, when available within ten (10) minutes, is sufficient to mitigate the event. Operator actions to start auxiliary feedwater within ten (10) minutes is consistent with the Ginna licensing basis. If the FLB was re-evaluated as a cooldown event from 34 power the results would be bounded by a SLB.

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9 ACILITY IIAM4 (Il OOCIIST IIVM44A (1I LSA MVM44II I ~ I ~ AQ4 ISI S ~ QUSHTIAL ASVIQl08 4UM A ~i Q 9A R.E. Ginna Nuclear Power Plant 90 017 00,10 oFl TEXT lll~ CWCe M ~, ~ AAAIMAMWIC AtW AM'llI Ill 0 5 0 0IO 2 4 4 5 Steam Generator Tube Ru ture SGTR is bounded by SBLOCA from the RCS depressurization standpoint. The leak rate from a SGTR is small compared to break flow for a SBLOCA. There is no significant effect due to lack of manual (pushbutton) or automatic SI since the main steps in the procedure deal with isolation of the ruptured SG, depressurization of the RCS, and termination of SI.

Small Break Loss of Coolant Accident When manual (pushbutton) and automatic SI was de-activated, the reactor was operating at 34 power. The reactor had been at 3% power for approximately ten (10) hours. Prior to that, the'eactor had been subcritical for twenty-two (22) hours following a trip.

Westinghouse Owner's Group letter WOG 90-113, dated July 2, 1990, "Shutdown LOCA Program Draf< Report", evaluated a mode 4 LOCA using a generic two (2) loop plant with a six (6) inch break assumed to occur two and a half (2.5) hours after shutdown. Acceptable results were obtained provided SI was started ten (10) minutes after the break.

Assumptions of the mode 4 LOCA analysis are compared with the Ginna Event conditions below:

WOG MODE 4 GINNA EVENT Decay Heat 1.34 Decay Heat 0.864 No accumulators available Accumulators available RCS pressure.1000 psig RCS pressure 2235 psig RCS temperature 425 F RCS temperature 547 F The availability of accumulators and the lower decay heat offset the higher RCS temperature and pressure. Sufficient time ~s available to manually start the.SI and RHR pumps and open appropriate valves from the Control Room, and to recover from the SBLOCA. In any case, SBLOCA is bounded by LOCA because less time is available for operator action during a Large Break LOCA.

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~. 0 5 I0 0 io 90 017 TSxr nr eeee Meee e we eeeMeew'AC ~ Xa4 Tv I ITI 2 4 4 0 IO 1 )1 os' I 5 Lar e Break Loss of Coolant Accident An assessment of disabling manual (pushbutton) and automatic SI at 3% power was performed by Westinghouse with respect to the LOCA analysis. The assessment assumed the RCS was at 547 F, 2235 psig. The fuel rods were assumed to be at 600 F which would be the approximate pellet and clad temperature at the end-of-blowdown phase. The vessel lower plenum and the lower portion of the core would be covered with accumulator water.

that SI must be initiated when the fuel rods are at 1800 F Further, it was assumed to turn around the cladding temperature before 2200 F. Decay heat is based on an approximation of power it reaches history prior to the event, using the 1971 ANS Model. An adiabatic heatup calculation was performed using properties for a 14 x 14 array Optimum Fuel Assembly (OFA). The calculation indicated SX was necessary in 5.5 to 6 minutes.

Simulations on the Ginna specific simulator indicate a 5 to 6 minute operator response during a LOCA is achievable.

Small Steam Line Break This accident is bounded by the Large SLB because longer times are available for operator response.

Lar e Steam Li:ne Break Westinghouse assessed the effect of no manual (pushbutton) or automatic SX on the Steam Line Break analysis. Based on their experiences with Steam Line Break analysis as well as a review of the available margin to the acceptance criteria, it was judged that analyzed at 34 power with no manual (pushbutton) or if the accident were re-automatic SI, acceptable results would be obtained.

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SACILITY IIAME III DOCKET IIU4HEA ITI LEII IILNNEII IEI PACE 1$ I SEOVS JJTJAL PEV>SIC U IvIJU 1 1 UVU SA R.E. Ginna Nuclear Power Plant TEXT JJJ eOrP JPPPP A newer. UPS PJ>>1>>MJ JTAC M Ja4'Il lltl 0 5 0 0 0 2 4 4 90 '017 00 12 OF Rochester Gas and Electric Corporation (RG&E) performed a computer analysis of the SLB using the Westinghouse LOFTRAN Code. A base case was compared to a case where SI was delayed for ten (10) minutes. The comparison indicated negligible change in, minimum DNBR. There was an insigni-ficant change in mass released to containment because mass release is dominated by initial steam generator level and auxiliary feedwater flow, neither of which are affected by delayed SI. Comparing energy out the break for both cases, showed negligible differences. Therefore, delaying SI has negligible effect on minimum DNBR and mass/energy out the break.

In conclusion, delay of manual (pushbutton) and automatic SI with the reactor at 34 power would not cause Non-LOCA events to'xceed the acceptance criteria. A delay of 5.5 to 6 minutes in the LOCA can be tolerated without unacceptable results. Based on operator training, this is sufficient time for operator response.

Based on the above, it can be concluded that the public's health and safety was assured at all times.

V. CORRECTIVE ACTION A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:

The affected system was restored to normal when the two (2) DC switches were closed twenty (20) minutes after they were opened.

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