ML18101B083

LER 95-010-01:on 950615,both Units RHR Pumps Inoperable for long-term Flow Requirements Due to RHR Flow Instrument Uncertainties.Further Evaluated New EOP Setpoint for RHR Pump operation.W/951031 Ltr

ML18101B083
Person / Time
Site:	Salem
Issue date:	10/31/1995
From:	Fregonese V, Warren C Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-95-010, LER-95-10, LR-N95193, NUDOCS 9511030194
Download: ML18101B083 (7)

Text

OPSEG *

• Putllic Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear. Business Unit OCT 311995 LR-N95193 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attn: Document Control Desk SALEM GENERATING STATION LICENSE NO. DPR-75 DOCKET NO. 50-272 UNIT NO. 1 LICENSEE EVENT REPORT NO. 95-010-01 This Supplemental Licensee Event Report entitled "Inoperability of Both Units' Residual Heat Removal (RHR) Pumps for Long-Term Flow Requirements Due to RHR Flow Instrument Uncertainties" is being submitted on a voluntary basis. Upon further review the.

occurrence has been downgraded to not reportable per 10CFR50.73.

However, the original LER promised a determination of root cause and identification of corrective actions in a supplemental report. Attachment A contains those commitments which are currently outstanding related to this issue.

Sincerely, LlfJij JiLLL-c1ay C. Warren General Manager -

Salem Operations Attachment A SORC Mtg.95-125 JMO C Distribution LER-File 9511030194 951031 PDR ADDCK 05000272 S PDR

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• Attachment A PSE&G Commitments for LER 272/95-010-01 The following item represents PSE&G commitments made to the Nuclear Regulatory Commission related tu LER 272/95-010-01. The commitments are as follows:

1. Adopt a new EOP setpoint for RHR pump operation based upon minimum flow requirements and instrument uncertainty prior to Salem Units restart.

2. Review other EOP setpoints for adequacy of instrumentation inaccuracies and update the instrument setpoint bases documents for affected EOP setpoints prior to restart of the Salem Units.

3. Revise affected Emergency Operating Procedures prior to restart of the Salem Units.

NRC FORM 366 14-96)

• U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

APPROVED BY OMB NO. 3150-0104 EXPIRES 04/30/98 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS. REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK TO INDUSTRY. FORWARD COMMENTS REGARDING BURDEN

- ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (T-6 F33J, U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC (See reverse for required number of 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC digits/characters for each block) 20503.

FACILITY NAME 111 DOCKET NUMBER 121 PAGE 131 SALEM GENERATING STATION UNIT 1 05000272 1 of 5 TITLE 141 INOPERABILITY OF BOTH UNITS' RESIDUAL HEAT REMOVAL IRHR) PUMPS FOR LONG-TERM FLOW REQUIREMENTS DUE TO RHR FLOW INSTRUMENT UNCERTAINTIES EVENT DATE 151 LER NUMBER 161 REPORT DATE 171 OTHER FACILITIES INVOLVED 181 MOl\ITH DAY YEAR YEAR I SEQUENTIAL NUMBER IREVISION NUMBER MONTH DAY YEAR FACIUTY NAME Salem Generating Station, Unit 2 DOCKET NUMBER 05000311 06 15 95 95 -- 010 -- 01 10 31 95 FACIUTY NAME DOCKET NUMBER OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or morel 1111 MODE 191 5 20.2201(b) 20.22031all2llvl 50. 731all211il 50. 73(a)(2)(viiil POWER 20.2203lall 1 I 20.2203(all311il 50. 731all211iil 50. 73la11211xl LEVEL 1101 000 20.2203lall2llil 20.2203(all311iil 50. 73lall211iiil 73.71 20.22031all211iil 20.2203(all41 50. 73(all211ivl x OTHER 20.2203lall211iiil 50.36lcll1 I 50. 73la112llvl Specify In Ab*tract below or In NRC Fonn 366A 20.2203lall211ivl 50.361cll21 50. 731all211viil LICENSEE CONTACT FOR THIS LER ;121 NAME TELEPHONE NUMBER (Include Area CO<MI Mr. V. Fregonese, Salem Instrumentation and Controls 609-339-1607 Engineering COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 1131 CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TO NPRDS :1!:::;1::::!:1::::::!:: TO NPRDS l1l1*111111111:1111111

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SUPPLEMENTAL REPORT EXPECTED (141 EXPECTED MONTH DAY YEAR SUBMISSION

'YES (If yee, complete EXPECTED SUBMISSION DATE). xlNO DATE (151 ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) 116)

On 6/15/95, the NRC was notified of a potential problem with the Residual Heat Removal (RHR) flow orifice installation and the potential for inaccurate operation of the RHR pump recirculation valves. A design basis concern was also identified that minimum continuous (long-term) flow requirements of the RHR pumps might not be assured due to unaccounted for instrument inaccuracies in an Emergency Operating Procedure (EOP) setpoint.

Subsequent testing satisfactorily demonstrated the operability and control function of the RHR flow orifices and the recirculation valves. A review of operator actions incorporated in the Salem EOPs has established that despite an incorrect EOP setpoint, the RHR pumps would not be operated with pump flows less than the long term minimum flow requirements.

This incident was originally classified as reportable per 10CFR50.73(a) ( 2)

{v) but based upon further review, this incident has been downgraded to not ' reportable per 10CFR50.73. However, the original LER promised a determination of root cause and identification of corrective actions in a supplement. Thus, this report is being filed to provide details to bring this issue to closure.

NRC FORM 366 14-96)

.. NRC FORM 366A (4-96)

• LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION U.S. NUCLEAR REGULATORY COMMISSION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE 131 YEAR I SEQUENTIAL NUMBER I ~N NUMBER SALEM GENERATING STATION UNIT 1 05000272 95 -- 010 -- 01 2 of 5 TEXT (If more apace i* required, use additional copies of NRC Form 366A) ( 1 7)

PLANT AND SYSTEM IDENTIFICATION Westinghouse - Pressurized Water Reactor Residual Heat Removal System {BP}*

• Energy Industry Identification System (EIIS) codes and component function identifier codes appear in the text as {SS/CCC}.

IDENTIFICATION OF OCCURRENCE Inoperability of both Units' Residual Heat Removal (RHR) pumps for lo~g term flow requirements due to RHR flow instrument uncertainties.

Event Date: June 15, 1995.

Discovery Date: June 15, 1995.

Report Date: July 14, 1995 for the original LER.

CONDITIONS PRIOR TO OCCURRENCE Both Units were in a self-imposed extended shutdown.

Mode 5 Reactor Power % Unit Load MWe DESCRIPTION OF OCCURRENCE At 0745 hours0.00862 days <br />0.207 hours <br />0.00123 weeks <br />2.834725e-4 months <br /> on June 15, 1995, the NRC was notified with a four hour report, in accordance with 10CFR50.72(b) (2) (i), of a potential problem with the RHR flow orifices 1(2)FE641A/B installation and potential for inaccurate operation of the RHR pump recirculation valves, which provide pump protection. Tes~ing on June 30, 1995 satisfactorily demonstrated the operability and control function of the RHR flow orifices and the RHR pump recirculation valves.

However, on July 10, 1995, a potential design basis concern was identified in that the minimum flow requirements of the RHR pumps could not be assured while implementing EOPs during small break Loss of Coolant Accidents (LOCAs). This determination followed recognition that instrument loop uncertainties were not factored into the RHR flow indication setpoints.

NRC FORM 366A (4-95)

NRC FORM 366A 14-95)

• LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION U.S. NUCLEAR REGULATORY COMMISSION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

YEAR I SEQUENTIAL NUMBER l PEW;ION NUMBER SALEM GENERATING STATION UNIT 1 05000272 95 -- 010 -- 01 3 of 5 TEXT (If more space ia required, use additional copies of NRC Form 366A) ( 17)

DESCRIPTION OF OCCURRENCE (cont'd)

Thus the scope of the investigation and reportability was redirected to address RHR operation during small break LOCAs. As described in PSE&G's responses to NRC Bulletin 88-04, a minimum flow of 300 gpm is required to provide for adequate pump performance for up to three hours of minimum acceptable flow operation. Subsequent re-examination of the vendor's pump curves has identified that 700 gpm is acceptable for long term (periods in excess of three hours) minimum flow.

Salem Station deviates from the Westinghouse Owners Group (WOG) Emergency Response Guidelines (ERGs) by using RHR flow and RCS pressure as the basis for continued RHR pump operation, while the WOG ERGs only use RCS pressure as a criteria for continued RHR pump operation. The basis for using flow indication as defined in the Salem EOP bases document is to provide a more positive indication that the RHR pumps are injecting into the RCS. Thus, the Salem EOPs direct the operator to start RHR pumps when RCS pressure drops below 340 psig and to leave the pumps running when the RCS cold leg injection flow is greater than 200 gpm (note: actual total pump flow is equal to recirculation flow of approximately 500 gpm plus the RCS cold leg injection flow). Upon examination, 200 gpm setpoint did not appear to reflect instrument uncertainties and thus was non-conservative with respect to assuring long term flow. Therefore, an operator observing greater than 200 gpm would make the decision to leave the RHR pumps running, when in fact the total flow through the line could be less than the minimum long term pump flow requirements defined as 700 gpm. Thus, the RHR pumps were postulated to experience damage during long term low flow situations which represents a event reportable per 10CFR50.73(a) (2) (v).

Subsequent to the initial LER submittal, and as a result of an apparent cause analysis and safety significance determination, it was determined that sufficient guidance exists within the EOPs, coupled with licensed operator knowledge and supplementary accident response procedures to cause periodic reevaluations of pump engineering parameters. These periodic reevaluations provide the needed check of pump performance curves and thus serve to alert operators to situations in which pump flow is too low for sustained operation. Therefore, this occurrence was downgraded to not reportable per 10CFR50.73.

APPARENT CAUSE OF OCCURRENCE The cause of this occurrence is attributed to personnel error, cause code A per Appendix B of NUREG 1022. The personnel responsible for deriving this setpoint failed to obtain adequate engineering oversight, thus issues such as instrument uncertainty and pump minimum flow requirements were not considered when this EOP setpoint was established.

NRC FORM 366A (4-95)

NRC FORM 366A 14-95)

• LICENSEE EVENT REPORT (LER)

~*

U.S. NUCLEAR REGULATORY COMMISSION TEXT CONTINUATION FACILITY NAME 111 DOCKET NUMBER 121 LER NUMBER 161 PAGE 1:1)

YEAR I SEQUENTIAL NUMBBI I ll'llEWllON NUMBBI SALEM GENERATING STATION UNIT 1 05000272 95 -- 010 -- 01 4 of 5 TEXT (If more *pace i* required, use additional copies of NRC Form 366AI ( 1 71 PRIOR SIMILAR OCCURRENCES There are no prior similar occurrences in which EOP setpoints have been improperly controlled. LER 311/95-015-00 reported a failure to maintain configuration control due to inadequate design review and inadequate administrative controls which is related to the deficiency cited.

SAFETY SIGNIFICANCE Operating the RHR pumps for sustained periods at minimum flow could cause the pumps to degrade. The current large break LOCA analysis assumes that the reactor rapidly depressurizes to approximately 30 psia in about 30 seconds. Thus, operation of the RHR pumps during large break LOCA scenarios does not pose a reduced flow pump concern. For the limiting small break LOCA Peak Centerline Temperature (PCT) analysis, near term operation of the RHR pumps is not credited since system pressure does not fall below the cut in RCS pressure during the transient. Thus the relevant concern for assessing the impact of the RHR EOP flow setpoint is for longer scenarios in which operator action is being taken per EOPs.

The RHR pumps start following a Safety Injection (SI) signal. A review of the Salem Station EOPS indicate that the RHR pumps would be stopped in less than 45 minutes from the initiation if RCS pressure is above the pump shutoff head pressure or if RCS cold leg injection flow is less than 200 gpm. If indicated RCS cold leg injection flow is greater than 200 gpm, then the RHR pumps are allowed to continue running. For scenarios in which RCS pressure is below the RHR initiation point, but close to the shutoff head of the pumps, there is a possibility that the RHR pumps will be running for an extended period with less than adequate flow.

However, the Salem EOPs make use of continuous action step coding associated with RHR pump operation to assure that the RHR pumps are checked periodically for both a sufficiently low RCS pressure and minimum flow.

Additionally, the Control Room operators relay pump operating data during accident events to the Technical Support Center (TSC) where pump performance is continuously evaluated to assure the availability of the pumps. Thus, for long term periods despite the inaccurate EOP setpoint, the pumps would be effectively monitored and damage precluded. Therefore, there is no safety significance to this occurrence.

NRC FORM 366A (4-95)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION 14-96)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME 111 DOCKETNU~~B~E~R~l=2~1!1-~~L=ER..:....:..:N~U~M~B~E~R~l~6~l~--l~~~P~A~G~E~l3~1'--~*

YEAR I SEQUENTIAL NUMBER I~

NUMBER SALEM GENERATING STATION UNIT L 05000272 95 -- 010 -- 01 5 of 5 TEXT (If more 11pace ie required, uae additional copies of NRC Form 366AI 1171 CORRECTIVE ACTIONS

1. Evaluate further, a new EOP setpoint for RHR pump operation based upon minimum flow requirements and instrument uncertainty prior to Salem Units restart.

2. Review other EOP setpoints for adequacy of instrumentation inaccuracies and update the instrument setpoint bases documents for affected EOP setpoints prior to restart of the Salem Units.

3. Revise affected Emergency Operating Procedures prior to restart of the Salem Units.

4. As some years had passed since the creation of the 200 gpm setpoint, there were no specific individuals identified to receive appropriate counseling regarding the personnel error for this event. Barriers that have since been established to preclude this type of failure in the future include a PSE&G Programmatic Standard, "Loop Uncertainty/Setpoint Management and Control Program for Salem Generating Station Units 1 and 2" which specifically identifies the Salem Instrumentation and Control Engineering Group as the owner of all EOP setpoints including bases and calculations, and Salem Station Operations "Emergency/Abnormal Operating Procedures Program" which directs EOP setpoints to be calculated by the PSE&G Engineering organization.

NRC FORM 366A (4-95)