ML17264A448

LER 96-003-00:on 960308,both Pressurizer Relief Valves Inoperable.Hpes Evaluation Is Being Conducted to Determined Cause of Event.C/As:Both PORVs restored.W/960408 Ltr

ML17264A448
Person / Time
Site:	Ginna
Issue date:	04/08/1996
From:	Mecredy R, St Martin J ROCHESTER GAS & ELECTRIC CORP.
To:	Andrea Johnson NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-96-003, LER-96-3, NUDOCS 9604150124
Download: ML17264A448 (10)

Text

~ CATEGORY 1 REGULATORY INFORMATION DISTRIBUTION SYSTEM

~ (RIDS)

ACCESSION NBR:9604150124 DOC.DATE: 96/04/08 NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION ST.MARTIN,J.T. Rochester Gas & Electric Corp.

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

RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A,R.

SUBJECT:

LER 96-003-00:on 960308,both pressurizer relief valves inoperable.HPES evaluation is being conducted to determined cause of event.C/As:both PORVs restored.W/960408 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ENCL SIZE:

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

NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72), 05000244 G 0

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-1 PD 1 1 JOHNSONgA 1 1 INTERNAL: QQp4&&~RAB 2 2 AEOD/SPD/RRAB 1

ILE CENTE 1 1 NRR/DE/ECGB 1 1 N E 1 1 NRR/DE/EMEB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRCH/HICB 1 1 NRR/DRCH/HOL'B 1 1 NRR/DRCH/HQMB 1 1 NRR/DRPM/PECB 1 1 NRR/DSSA/SPLB 1 1 NRR/DSSA/SRXB 1 1 RES/DS IR/EIB 1 1 D RGN1 FILE 01 1 1 0

EXTERNAL: L ST LOBBY WARD 1 1 LITCO BRYCE,J H 2 2 NOAC MURPHY,G'.A 1 1 NOAC POOREiW. 1 1 NRC PDR 1 1 NUDOCS FULL TXT 1 1 N

NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTETH CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN SD-S(EXT. 415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEEDl FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 25 ENCL 25

WHO ROCHESlfR GAS AVDEIfClRIC CORPORA70h/ ~ S9EASl AVENUE, ROCHESTfR, N. V ld64900CI ~ ARE'A CODE716 546.2700

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April 8, 1996 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Allen R. Johnson PWR Project Directorate I-l Washington, D.C. 20555

Subject:

LER 96-003, Both Pressurizer Relief Valves Inoperable, Results in Condition That Could Have Prevented Fulfillment of a Safety Function R.E. Ginna Nuclear Power Plant Docket No. 50-244 In accordance with 10 CFR 50.73, Licensee Event Report System, item (a) (2) (v), which requires a report of, "Any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to: ... (D) Mitigate the consequences of an accident", the attached Licensee Event Report LER 96-003 is hereby submitted.

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

Very ly yours, Robert C. Mecredy xc: U.S. Nuclear Regulatory Commission Mr. Allen R. Johnson (Mail Stop 14B2)

PWR Project Directorate I-1 Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 U.S. NRC Ginna Senior Resident Inspector t<P~94 9604ISOi24 960408 PDR ADOCK 05000244 8 PDR

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (4-95) EXPIRES 04/30/99 ESTIMATED BURDEN PER RESPONSE To COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HAS.

AEPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSEE EVENT REPORT (LER) UCENSING PROCESS AND FED BACK To INDUSTAY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE To THE (See reverse for required number of INFORMATION AND AECOADS MANAGEMENT BRANCH (T.6 F33),

digits/characters for each block) U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555 000), AND To THE PAPERWORK REDUCTION PROJECT FACILITY NAME (1) OOCKET NUMBER (3) pAoe (3)

R.E. Ginna Nuclear Power Plant 05000244 1 OF 7 ITLE (4)

, Both Pressurizer Relief Valves Inoperable, Results in Condition That Could Have Prevented Fulfillment of a Safety Function EVENT DATE IS) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (6)

FACILITYNAME OOCKET NUMSEA SEQUENTIAL REVISION MONTH DAY YEAR NUMBER MONTH OAY YEAR NUMBER FACILITY NAME OOCKET NUMBER 03 08 96 96 003 00 04 08 96 OPERATING THIS REPORT IS SUBMITTED PURSUANT To THE REQUIREMENTS OF 10 CFR E: (Chock one or more) (11)

MODE (9) 20.2201(b) 20.2203(e)(2) (v) 50.73(a)(2)(i) 50.73(a) (2) (viii)

20. 2203(e) (1) 20.2203(e)(3) (I) 50.73(a)(2)(ii) 50.73(a) (2) (x)

POWER LEVEL (10) 000 20.2203(e) (2)(i) 20.2203(e) (3)(ii) 50.73(e) (2) (iii) 73.71 20.2203(a)(2) (ii) 20.2203(e) (4) 50.73(e) (2)(iv) OTHER 20.2203(a)(2)(iii) 50.36(c)(1) X 50.73(e)(2)(v) Specify in Abstract be low or in NRC Form 366A 20.2203(a)(2) (iv) 50.36(c)(2) 50.73(e) (2) (vii)

LICENSEE CONTACT FOR THIS LER (12)

IIAME ELEPIIONE NUMBER (Include Area Code)

John T. St. Martin - Technical Assistant t7161 771-3641 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESC RIBED IN THIS REPORT (13)

REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER To NPRDS To NPAOS SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY YEAR YES SUBMISSION (If yos, complete EXPECTEp SUBMISSION DATE).

NO DATE (16) 05 17 96 ABSTRACT (Limit to 1400 spaces, i.o., epproximatoly 15 single.spaced typewrinen lines) (16)

On March 8, 1996, with the plant in Mode 3, it was identified that both pressurizer power-operated relief valves had been inoperable concurrently. This resulted in the plant being in a condition that could have prevented the fulfillment of a safety function, in the unlikely event that a steam generator tube rupture occurred coincident'ith a loss of off-site power event.

Immediate corrective action was to restore both PORVs to operable status.

A Human Performance Enhancement System (HPESI evaluation is being conducted to determine the underlying cause of this event. A supplement to this LER will be submitted, describing the root cause.

Corrective action to prevent recurrence is outlined in Section V.B. Additional corrective actions will identified after the root cause has been determined.

NRC FORM 366 (4-65)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUM()ER (6) PAGE (3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER 2 OF 7 R.E. Ginna Nuclear Power Plant 05000244 96 003 00 TEXT llfmore spaceis required, use additional copies of PlRC Form 366Al (17)

I. PRE-EVENT PLANT CONDITIONS:

I ~

On March 8, 1996, the plant was in Mode 3 (hot shutdown) as a result of a reactor trip that occurred on March 7, 1996 (reported in LER 96-002). (Refer to Ginna Docket No. 50-244, LER 96-002.) The reactor coolant system (RCS) was being maintained at a temperature of 547 degrees F and a pressure of 2235 PSIG in anticipation of plant startup. The opportunity to perform maintenance and testing on the two pressurizer (PRZR) power-operated relief valves (PORV) became available. This work needed to be performed to support entering the Low Temperature Overpressure Protection (LTOP) condition during the annual refueling outage, which was scheduled to begin April 1, 1996. Conditions were judged acceptable to perform these activities during this unplanned shutdown. These activities consist of adjusting and verifying the proper valve benchset for the two PORVs, and readjustment of valve limit switches. (The equipment identification numbers for the PORVs are PCV-430 and PCV-431C.)

The Instrument and Control (IRC) group was notified that this work would be performed with the RCS at normal operating pressure of 2235 PSIG (with a saturation temperature in the PRZR of approximately 650 degrees F). The IRC group expressed concerns regarding personnel safety for working in the ambient temperature conditions that existed in the PRZR cubicle in the Containment (CNMT) A meeting was held

~

to discuss ISC concerns and to review the general plan. Maintenance Procedure M-37.150 (Copes-Vulcan

/ Blaw-Knox Air Operated Control Valves Inspection and Refurbishment) was changed to allow work on both PORVs concurrently, in consideration of the heat stress in the PRZR cubicle in CNMT.

The Control Room operators acknowledged these changes to procedure M-37.150 as acceptable, since it was their understanding that even with the procedure change, the PORVs would remain operable per the requirements of Ginna Station Improved Technical Specifications (ITS) Limiting Condition for Operation (LCO) 3.4.11. This was based on their assumption that nitrogen would still be connected to the PORV actuators and available to open PORVs. Thus, the Shift Supervisor determined that LCO 3A.11 Condition A and LCO 3.4.11 Condition F did not apply, since both PORVs were to remain operable.

At approximately 1406 EST on March 8, 1996, the Control Room operators held the control switches for the PORV block valves (MOV-515 for PCV-431C, and MOV-516 for PCV-430) in the close position, with power available to the breaker for each MOV. Operators understood that in this configuration these holds could be removed, the block valves opened, and the PORVs could then be opened.

II. DESCRIPTION OF EVENT:

A. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES:

March 8, 1996, 1455 EST: Event date and time. Air and nitrogen is removed from both PORVs.

o March 8, 1996, 1632 EST: Air and nitrogen is restored to both PORVs.

NRC FORM 366A (4-95)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4.95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITYNAME (1) DOCKET LER NUMBER (6) PAGE (3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER 3 OF 7 R.E. Ginna Nuclear Power Plant 05000244 96 003 00 TEXT iifmore spaceis required, use additional copies of NRC Form 366A) (17)

March 8, 1996: 2018 EST: Air and nitrogen is removed from both PORVs.

March 8, 1996: 2140 EST: Air and nitrogen is restored to both PORVs.

March 8,1996: Discovery date.

March 28, 1996, 1408 EST: Having both PORVs inoperable under the conditions on March 8, 1996, is determined to be reportable to the NRC.

March 28, 1996, 1640 EST: The NRC Operations Center is notified of the events of March 8, as per 10 CFR 50.72 (b) (2) (iii) (D) ~

B. EVENT:

On March 8, 1996, the plant was in Mode 3. The benchsets and limit switches for both PORVs were being adjusted and verified per a temporary change to procedure M-37.150, and tested using Surveillance Test Procedure PT-2.6.5-SD (RCS Overpressure Protection System PORV Operability Verification), which verifies the timing of the PORV stroke for LTOP requirements. The ISC technicians entered the CNMT at approximately 1445 EST on March 8, to perform these activities.

By approximately 1455 EST, the flex hose that supplies both instrument air and nitrogen to the PORV was removed (per changed procedure M-37.150) for both PORVs. In hindsight, this action made both PORVs inoperable concurrently, and is the event date and event time.

At the completion of the adjustments to the benchset and limit switches for each PORV, the flex hose was reconnected. By approximately.1 632 EST this was completed (per changed procedure M-37.150) for both PORVs. Each PORV was stroked by the Control Room operators from the Main Control Board (MCB). The adjustments for both PORVs were verified as acceptable per procedure PT-2.6.5-SD (at approximately 1759 EST).

Operations supervision had previously determined that, since the benchset was changed to decrease the closing spiing tension of the PORVs, it would be prudent to perform a seat leakage check on the PORVs prior to the completion of the maintenance package. The holds on the block valve control switches were then cleared. Both PORVs were verified to be in the closed position and block valve MOV-516 (for PCV-430) was stroked open at approximately 1803 EST for the seat leakage test. As MOV-516 stroked open, a substantial flow noise was evident in the PRZR cubicle, and it was discovered that seat leakage through PCV-430 was in excess of the previously observed leakage of approximately 0.25 gallons per minute (GPM).

Since the benchset had been changed on both PORVs, it was decided not to perform a seat leakage test on PCV-431C by opening MOV-515, and to conservatively assume that the seat leakage through PCV-431C was comparable to the leakage through PCV-430. The Control Room operators closed MOV-516 at approximately 1804 EST, and initiated discussions concerning PORV operability with this seat leakage.

NAC FOAM 366A (4 95)

'J NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION I4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME I1) DOCKET LER NUMBER I6) PAGE I3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER 4 OF 7 R.E. Ginna Nuclear Power Plant 05000244 96 003 00 TEXT (lf more spaceis required, use addidonal copies of ifRC Form 366AJ <17)

Subsequent 'to the I&C technicians exiting the CNMT, details of the activities were discussed between the I&C technicians who had performed the activities inside the PRZR cubicle in CNMT and members of the plant staff. Later that evening, members of the plant staff then realized that, during the benchset activities, the nitrogen supply had been disconnected to both PORVs concurrently. Therefore, the discovery date is March 8, 1996, late in the evening.

With both PORVs inoperable, LCO 3.4.11 Condition F requires that immediate action be initiated to restore at least one PORV to operable status. However, with nitrogen being restored to the PORVs at approximately 1632 EST, the PORVs were operable in accordance with LCO 3.4.11 at the time of discovery. During the period of (undeclared) PORV inoperability, LCO 3 4.11 Required Action F.3 was not met in that power for the associated block valve was not removed.

Activities needed to reduce the PORV seat leakage were reviewed. The leakage indications were reevaluated by plant staff. Preliminary estimates indicated that leakage was greater than the RCS operational leakage of ten (10) GPM identified leakage, as specified by LCO 3.4.13. At approximately 1940 EST on March 8, both PORVs were declared inoperable, solely due to their inability to meet the RCS operational leakage limits of LCO 3.4.13. Actions were taken as specified by LCO 3.4.1'I Condition F. Since the block valves were already closed, the block valve control switches and breakers were held, and efforts to restore PORV seat leakage to less than 10 GPM began, in accordance with LCO 3.4.11 Required Action F.1.

The l&C technicians completed the readjustments to benchset and readjusted the limit switches for the PORVs as per procedure M-37.150. The holds were removed from the block valve control switches and breakers using the guidance of LCO 3.0.5. The block valves were stroked as needed, and the PORVs were tested using procedure PT-2.6.5-SD. Both PORVs demonstrated acceptable performance. A check of seat leakage was conducted, and leakage was evaluated to be less than 10 GPM. The PORVs were then declared operable at approximately 0123 EST on March 9, 1996, and returned to service.

Upon further evaluation later in the month, it was identified that the nitrogen supply to both PORVs had been concurrently disconnected on two separate occasions on March 8. This rendered the PORVs inoperable per LCO 3.4.11 each time, for a combined duration of approximately less than three (3) hours, without being declared inoperable.

C. INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:

None D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED:

None NIIC FOAM 366A I4-95)

NRC FORM 366A NUCLEAR REGULATORY COMMISSION (4.95 I LICENSEE EVENT REPORT (LER)

CONTINUATION'.S.

TEXT FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER 5 OF 7 R.E. Ginna Nuclear Power Plant 05000244 96 003 00 TEXT Iifmore spaceis required, use eddi rionel copies of iVRC Form 366AJ (17)

E. METHOD OF DISCOVERY:

The condition of the PORVs was investigated after a failed seat leakage check. During discussions between l&C and plant staff, members of the plant staff became aware of the full scope of the work that had been performed during the benchset, and that the PORVs had actually been inoperable during the benchset activities. This information was subsequently relayed to the Shift Supervisor during the evening of March 8, 1996.

OPERATOR ACTION:

The Control Room operators notified higher supervision. Nuclear Safety and Licensing (NS&L) was also consulted concerning the seat leakage through the PORVs. These discussions also included a reportability discussion, but focused on the reportability of PORV seat leakage. The Control Room operators declared both PORVs inoperable, solely due to their inability to meet the RCS operational leakage limits of LCO 3.4.13 while the block valve was open, and isolated the PORVs in accordance with LCO 3.4.11 Condition F.

Documentation was prepared to initiate a reportability evaluation. At approximately 1408 EST on March 28, 1996, plant staff determined that a non-emergency four hour notification, per 10 CFR 50.72 (b) (2) (iii) (D), should be made to the NRC Operations Center. This notification was made at approximately 1640 EST on March 28, 1996.

G. SAFETY SYSTEM RESPONSES:

None III. CAUSE OF EVENT:

A. IMMEDIATECAUSE:

The immediate cause of both PORVs being inoperable concurrently was due to the disconnection of the flex hose to both PORV actuators to install air-sets for the benchset and limit switch activities. Removal of this flex hose removed the supply of instrument air and nitrogen to both PORVs.

INTERMEDIATE CAUSE:

The intermediate cause of the removal of the nitrogen supply to both PORVs was the procedural direction that permitted the concurrent benchset of both PORVs, and an incorrect assumption by plant staff that in this configuration, nitrogen would still be available to open PORVs.

NRC FORM 366A (4-95)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4.95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER 6 OF 7 R.E. Ginna Nuclear Power Plant 05000244 96 003 00 TEXT (lf more spaceis required, use additional copies of NRC Form 386AJ (17)

C. ROOT CAUSE:

A Human Performance Enhancement System (HPES) evaluation is being conducted to determine the underlying cause for how and why this incorrect assumption occurred. The preliminary results of the HPES evaluation have been presented to plant management. When the HPES evaluation and Root Cause Analysis are completed, a supplement to this LER will be submitted, describing the root cause(s).

~

The NUREG-1022 Cause Code for this event will be provided in a supplement to this LER.

IV. ANALYSIS OF EVENT:

A reportability evaluation was performed by NS&L. NS&L identified that the concurrent inoperability of both PORVs could have rendered the plant unable to mitigate the consequences of a steam generator tube rupture (SGTR) in accordance with the Westinghouse analysis of record, if the SGTR had occurred coincident with a lossof off-site power event.

This event is reportable in accordance with 10 CFR 50.73, Licensee Event Report System, item (a) (2) (v),

which requires a report of, "Any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to: ... (D) Mitigate the consequences of an accident". Both PORVs inoperable under the conditions that existed on March 8, 1996, is an event that could have prevented RCS depressurization after a SGTR, coincident with a loss of off-site power event.

An assessment was performed considering both the safety consequences and implications of this event with the following results and conclusions:

There were no operational or safety consequences or implications attributed to both PORVs being inoperable because:

o The PORVs serve as the LTOP protection devices during conditions of low temperature in the RCS. The plant did not enter conditions where this protection would have been required.

o The PORVs provide a depressurization path for a SGTR when normal and auxiliary PRZR spray are not available (during a loss of off-site power event). The plant did not enter conditions where a loss of off-site power occurred. Thus, PRZR spray was available.

o In either of the above situations, the PORVs could have been restored to service to support these requirements.

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

NRC FORM 366A (4-95)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)

YEAR SEQUENTIAL RFVISION NUMBER NUMBER 7 OF 7 R.E. Ginna Nuclear Power Plant 05000244 96 003 00 TEXT /if moro speceis required, use edditionel copies of fVRC Form 366A/ (17)

V. CORRECTIVE ACTION:

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

The benchset for each PORV was readjusted to reduce seat leakage and limit RCS leakage.

The PORV limit switches were verified in the proper setup and configuration.

The PORVs were tested per procedure PT-2.6.5-SD.

The supply of instrument air and nitrogen was reconnected to the actuator for each PORV.

B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:

1. Lessons learned have been provided to Operators, PORC Independent Reviewers, and Maintenance Planners.

2. The HPES will be completed.

3. A supplement to this LER Yvill be submitted by May 17, 1996.

VI. ADDITIONALINFORMATION:

A. FAILED COMPONENTS:

None B. PREVIOUS LERs ON SIMILAR EVENTS:

A similar LER event historical search was conducted with the following results: Since the'r'oot cause has not been determined, it is not appropriate to conclude which similar LER events with the same root cause have occurred at Ginna Nuclear Power Plant.

C. SPECIAL COMMENTS:

None i

NRC FORM 366A (4.95)