IR 05000272/1989027
| ML18094B247 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 01/03/1990 |
| From: | Dudley N, Swetland P, James Trapp, Jimi Yerokun NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18094B244 | List: |
| References | |
| 50-272-89-27, NUDOCS 9001240309 | |
| Download: ML18094B247 (41) | |
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ENCLOSURE 4 Report N License N Licensee: Facility Name: Inspection At: U.S. NUCLEAR REGULATORY COMMISSION REGION I INSPECTION REPORT 50-272/89-27 DPR-70 Public Service Electric and Gas C P. 0. Box 236 Hancocks Bridge, N Salem Unit No: 1 Hancocks Bridge, N Inspection Conducted: Nove~~.; 29, l~ thru,?~~r 1, 1989 Inspectors: . ~PA1zd-~ Approved by: -' 3 N. F. Dudley, Project Engineer, DRP J. ~r~. Reactor Eng., DRS, EB Cf~ ~ J. Yero~ Eng., DRS, EB P. D. Swetland, Chief Reactor Projects Section 2A 1/3/70 Date .1 I /1 ~~iD
- Date Inspection Summary:
Reactive Unannounced Inspection on November 29, 1989 through December 1, 198 Areas Inspected: Review of events surrounding the November 9, 1989 entry into Technical Specification 3.0.3 during the performance of the Turbine Volumetric Flow Tes Results: The inspector found that some actions taken by the licensee during the Turbine Volumetric Flow Test did not conform with NRC Regulation This inspection resulted in two violation Three additional items were not resolved prior to the exit meeting held on December 1, 1989 and will be tracked as Unresolved Items. 9001240309 900108 PDR ADOCK 05000272 F'DC
- Details 1.0 Persons Contacted 1.1 PSE&G
- L. Curran, Operating Engineer
- D. Dodson, Principal Engineer, Licensing
- W. Grau, Licensing Engineer
- B. Gorman, Manager External Affairs M. Gwirtz, Senior Nuclear Shift Supervisor(SNSS)
- R. Heaton, System Engineer
- E. Krufka, Engineer
- S. LaBruna, Vice President Nuclear Operations D. Martrano, Engineer
- M. Metcalf, Project Manager
- L. Miller, General Manager Salem OperJtions H. Onorato, Licensing Engineer
- K. Pike, Reactor and Plant Perf. Engineer
- V. Polizzi, Operations Manager
- B. Preston, Manager, Licensing and Regulation
- R. Reichel, Engineer
' * W. Schulk, Manager Station QA J. Serwan, SNSS 1.2 U.S. NRC Personnel
- N. Dudley, Project Engineer
- K~ Gibson, Senior Resident Inspector
- S. Pindale, Resident Inspector
- P. Swetland, Chief, RPS-2A
- J. Trapp, Senior Reactor Engineer
- J. Yerokun, Reactor Engineer
- denotes present at exit meeting held on December 1, 1989 2.0 Introduction The purpose of. this inspection was to review the events surrounding the entrance into Technical Specification Limiting Condition for Operation (LCO) 3.0.3 during the conduct of the Turbine Volumetric Flow Tes Specifically the scope of the inspection included the following:
- 0 Establish a sequence of event Review the adequacy of the test procedure/conduc Review the root cause analysis for PlO jumper re-energizing trip ° Conduct interviews with key personne Review actions taken to shutdown plant following entry into Technical Specification LCO 3. Review corrective action Assess performance of plant staff associated with the tes The inspectors findings with regard to these issues are contained in this inspection repor.0 Summary of Events The sequence of events surrounding th£ performance of the Turbine Volumetric Flow Test, is provided in Attachment A, in chronological orde A summary of the events is provided belo On November 9, 1989 a Turbine Volumetric Flow Test, REM T-1, Rev. 0, was being conducted to collect data needed as part of the Rerating Feasibility Progra The objective of the test was to determine the minimum steam pressure and the corresponding average reactor coolant loop temperature (Tavg) for full power operation with the main turbine control valves in the full open positio The test procedure required Tavg to be reduced in 2°F increments by adding boron to the Reactor Coolant Syste At each 2°F step, a calormetric calculation was planned with the Power Range Nuclear Instrumentation (NI 1s) being adjusted as necessar Calibration of the NI 1 s was required because a decrease in reactor inlet temperature (Tcold),
corresponding to the decrease in Tavg, will cause the downcomer water to shield more neutrons; thus causing the NI indicated power to be less than actual reactor thermal powe At 569°F the calormetric calculation was performed, NI's were adjusted, and the second boration to decrease Tavg an additional 2°F was complete At 567°F a second calormetric calculation was performed and three NI channels N41-N43 were satisfactorily adjuste While attempting to adjust N44 the fine gain adjustment reached its lower limit stop before the indicated power could not be raised to equal the actual powe The difference between the indicated and a~tual power was.8% in a non conservative directio A difference of.8% satisfied the +/- 1% acceptance criteria provided in the calormetric calculation procedure, however a second criteria that the average of the four channels be greater than or
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equal to the.calormetric calculation could not be me Therefore the operators chose conservatively to decl~re N44 inoperable per Technical Specification. At this point the turbine governor valves were full open and data collection for the test was complet Nuclear Instrumentation Channel N44 was declared inoperable at 5:40 per Technical Specification 3.3.I, Reactor Trip System Instrumentatio The Limiting Condition for Operation (LCO) action statement for an in-operable NI requires the inoperable channel be placed in trip within one hou Shift operating personnel were confident that a course gain adjust-ment could be performed within one hour by I&C technicians supporting this test. If the course gain could be adjusted, calibration of N44 could be performed, and the LCO could then be exite Operations requested that I&C perform the course gain adjustment, but did not convey adequately the urgency required due to the one hour action statemen I&C encountered a number of delays in making the coarse gain adjustment, and the decision was made by the operators at 6:30 a.m. to expedite the initial step of the course gain adjustment procedure which would place the channel in a trip condition. One hour and one minute after declaring N44 inoperable the channel bistables were tripped*by I&C technicians. At this point a control room operator recognized that th~ I&C procedure did not include installation of a jumper to energize the PIO rela The in~tallation of the PIO jumper ~as included in the operations procedure for tripping an NI channe The operators made a decision to consider the NI not tripped until the PIO jumper was installed. Therefore the action statement for T~chnical Specification 3.3.I.I, which requires the inoperable channel to be tripped within one hour, was not met and Technical Specification 3.0.3 was entere Technical Specification 3.0.3 requires in part that within one hour action shall be initiated to place the unit in a MODE in which the specification does not apply, and be in hot standby within the next six hour The operators believed that the successful installation of the PIO jumper would place N44 in a tripped condition per Technical Specification 3.3.I.I, and Technical Specification 3.0.3 could then be exite At 7:36 a.m. the PIO jumper was installed by I& However, the installation of the PlO jumper, caused an unanticipated repowering of the previously deenergized bistables for the high flux rate trip, high flux trip (high setpoint), and the high flux trip (low setpoint).
This was not understood and placed the channel in a condition not allowed by Technical Specification At 7:40 a.m. the one hour requirement of Technical Specification 3.0.3, to initiate action to place the unit in a MODE in which Technical Specification 3. did not apply had been exceede At 7:50 a.m. the Senior Nuclear Shift Supervisor (SNSS) ordered I&C to remove the PIO jumpe ~t the same time the SNSS, with the concurrence of the Operating Engineer, initiated the steps of the Turbine Volumetric Flow Test which restored the plant to normal conditions. These actions included diluting the Reactor Coolant .,:.'., *,....
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System (RCS) to raise Tavg back to it 1s program valu By ra1s1ng Tavg, a calormetric calculation could be -performed, allowing recalibration of N4 Recalibration of N44 would allow exiting Technical Specification LCOs 3.0.3 and 3.3. The SNSS thought that using the approved turbine test procedure restoration steps to restore Tavg and recalibrate the Nis was prudent and would return the N44 channel to service sooner than a power reduction which would extend the calorimetric stabilization perio Tavg was raised to 569°F, and recalibration of the NI 1 s was completed at 9:40 Technical Specification 3.0.3 was exited three hours after entering the specification. _Technical Specification 3.3.1.1 was also exited and the plant was returned to it's normal 100% power conditio.0 Test Conduct The inspectors reviewed the test procedure, 11Turbine Volumetric Flow Test,
REM T-1, Rev. This procedure was approved by the Station's Operations Review Committee (SORC) and the General Manage The test objective was to determine th~ minimum steam pressure and the equivalent Tavg fo: full power operation with the turbine governor valves in the full open or near full open position. This information would be used as part of the Rerating Feasibility Program. The lOCFR 50.59 Safety Evaluation performed for this test was reviewed by the inspector The evaluation acknowledged that the test was not des-cribed in the FSA The evaluation also imposed certain limits on the test in order to remain within analyzed condition The test was to be terminated when either all the turbine control valves are full open or Tavg had been reduced by 14° Test duration was limited to 16 hours or les The impact of test performance on the active 16 reactor trips and five ESF actuation signals was analyzed in the Safety Evaluatio It was determined that the test would not reduce the margin of safety provided by these trip The inspectors concluded the Safety Evaluation was adequat The 11 Precautions and Limitations" section of the test contained necessary information for a safe test performance except it did not list explicitly all the limitations contained in the lOCFR 50.59 Safety Evaluatio For example, the imposed 14°F temperature reduction limit was not liste The inspectors acknowledged that this limit was mentioned elsewhere in the procedur The pre-test briefings did not include personnel from the I&C departmen While personnel in the I&C department were on site specifically for the test, they were not included in the briefing This was inadequate because the involvement of this department in the test was anticipated and mentioned in the procedure. Step 6.3 of the procedure indicated that briefings should be conducted with operators and test support personne. *,-: . .
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- The test procedure referenced I&C procedures IC-I4.4-02I,022,023 or 024 to be used for in~trument adjustment, if require These referenced procedures were not the most appropriate.procedures available to the I&C department for the applicable instrument adjustmen Because I&C personnel were not properly involved in pre-test activities, this oversight was not discovered or corrected prior to initiating the tes The resultant discussions to clarify the appropriate procedures delayed tripping the N44 channe Other sections of the procedure were found to be adequat The resto-ration section of the test procedure properly restored the plant to its pre-test conditio The Turbine Volumetric Flow Test and plant restoration were conducted in accordance with the test procedur.0 PIO Jumper Installation The PIO Permissive allows manual blocking of the Source Range Detector Voltage, Intermediate Range Detector High Flux Trip, and the Power Range High Flux Trip-Low Setpoint during power escalation when power is greater than IO%.. The logic for this permissive requires 2/4 Power Range Detectors to indic.ate a power level greater than IO% power to allow manual bypass of these trips by the operato During power decreases, the trips will auto-matically reinstate when 3/4 of the Power Range Channels indicate less than IO% powe TS 3.3 requires this reinstatement function to be operable in operating Mode The Power Range Detector drawer powers bistables which send signals to the Solid State Protection System (SSPS) indicating the status of the power range channe Bistables are provided for the reactor trips and for the permissives, one such bistable in the power range drawer is the PIO bistabl This bistable changes state at IO% power indicating to the SSPS when 2/4 channels are greater than IO% powe The reactor trips mentioned above may then be manually blocke The output of the PIO bistable is energized or closed when reactor power is below IO% and open or deenergized when power is above IO%.
When a power range channel is removed from service, Technical Specifications requires all the reactor trip bistables be placed in their deenergized (tripped) positio This is performed by removing the control power from the bistable This sends a trip signal to the SSPS and reduces the reactor trip logic from a 2/4 to a I/ Removal of the control power works satisfactorily for the reactor trip bistables, but would not be adequate for the PIO permissiv PIO bistables must energize below IO% power to reinstate the Intermediate Range, and Low Power Range Trip, and-Source Range Voltag When control power is removed, the PIO bistable*deenergizes and cannot re-energiz Therefore the logic for reinstating the trips goes from a 3/4 less than IO%, to a 3/3 less than 10%. If a single failure is assumed, it can be postulated that the low level trips bypassed by PIO, would not automatically reinstat Consequently, a jumpe~ must be installed to energize the PIO .. :.. :_.::*:
- signal from the tripped channel to the SSPS, thus changing the logic t6 reinstate the trips to a 2/ Detailed information regarding the PlO permissive when tripping a power range channel was provided in ~RC.
Information Notice 86-105, and Westinghouse Technical Bulletin 86-0 During the performance of the Turbine Volumetric Flow Test, the decision was made to remove Nuclear Instrumentation Channel 44 (N44) from service, due to the inability to calibrate this channe The I&C technician removed N44 from service using I&C procedure lIC-16.4.024, 11 Power Range Channel 1N44 Detector Current Adjustment.
This procedure was used in lieu of the normal operations procedure for removing NI's from service IV-10.3.1, 11 Removing, Returning to Service and Loss of Protective System Channel.
Upon completion of tripping the channel by I&C personnel, a control room operator recognized that the PlO jumper required by operations procedure IV-10.3.1, was not required or installed by the I&C technician Following identification by the operators that a PlO jumper was required, I&C was requested to acquire the required materials and install the PlO jumper in acccrdance with operations procedure IV-10. The PlO jumper cQnsisted of supplying external 115 VAC power to the output sidL of the PlO bistable and thus to the SSP When the PlO jumper was installed per procedure IV-10.3.1, the N44 trips which had previously been deenergized using the I&C procedure became re-energize N44 was not in a tripped condition per Technical Specification The intent of the PlO jumper was not to re-energize these trips and the reason the trips became re-energized was not understood by the I&C technician or the operators on shift. Twelve minutes following re-energizing the reactor trips, the SNSS ordered the jumper removed and the trips previously deenergized by removing control power, were once again deenergize After the event, the licensee's initial corrective action for the failure of the PlO jumper to function properly was to revise the procedure IV-10.3.1 to lift electrical leads between the PlO bistable and the SSPS, and install the jumper directly in the SSP This action eliminated the possibility of the jumper affecting the other N44 bistables, but did not determine whether the unexpected NI channel performance resulted from some unknown system defect which could affect system operabilit The licensee's subsequent root cause analysis for the failure of the PlO jumper to function properly determined the followin The top and bottom detector cables had been removed from the N44 instrumentation drawe-r as part of the I&C procedure for removing a channel from service (lIC-16.4.024).
This step was not included in the Operations procedure for removing a channel from service, which was ultimately used for installation of the PlO jumper because the PlO jumper was not referenced in the I&C procedur By removing the N44 detector cables the channel indicated power fell OFF-SCALE lo Indicated power less than 10% caused the PlO bistable
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to clos Closing of the PlO bistable allowed an inductive circuit to feed back power supplied by the PlO jumper, through the PlO bistable to re-energize control power to the entire NI channe Re-energizing the control power allowed the other reactor trip bistables associated with this channel to become re-energized even though the control power fuses were remove The PlO jumper was not installed in accordance with guidance provided in the Westinghouse bulletin. This issue remains unresolved pending further NRC review of the licensee program for review and implemen-tation of industry operational experience and vendor recommendations (UNR 272/89-27-01).
6.0 Assessment/Findings The licensee entered Technical Specification 3.0.3 at 6:40 a.m. on November 9, 198 After one hour, Technical Specification 3.0.3 required actions to be initiated to place the plant in a mode where Technical Specification 3.3.1.1 did not apply. 10CFR50.72 required that a notifi-cation be made to the NRC within one hour when initiation of any nuclear plant shutdown is required by the plant's Technical Specification This notification was not made by the licensee until after the NRC inspection team arrived onsite, when the licensee recognized a shut down should have been.initiate The licensee committed at the exit meeting to review the Emergency Classification Guidelines to assure that notifications will be made in the future when shutdowns are required by Technical Specifications. Technical Specification 3.0.3 was entered at 6:40 a.m. and exited three hours later at 9:40 Technical Specification 3.0.3 required within one hour actions be initiated to place the unit in a MOOE in which the specification does not appl These actions to be performed within on hour, as described in the bases for Standard Technical Specifications, include time for the operator to prepare for and coordinate the reduction in electrical generation to ensure the stability and availability of the electrical gri At no time during the three hour period were such pre-parations made to make a load reductio In a similar event on November 17, 1989, licensee management decided to enter Technical Specification 3.0.3 while processing an Emergency Operating Procedure revision to compensate for an identified Emergency Core Cooling System Design Deficienc The licensee decided not to initiate actions for a reactor shutdown during the hour and forty-five minute period that Technical Specification 3.0.3 would be in effec A report was made to NRC within one hour of the recognition of the design deficiency by the statio Details of the event are discussed in NRC Special Inspection Report No. 50-272/89-25; 50-311/89-23. These two events are considered an apparent violation of Technical Specification 3. (Violation 50-272/ 89-27-02).
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Station operators are provided guidelines in the form of Operations Directives (ODs) which document station management 1 s position and inter-pretation of selected Technical Specification OD-12, Revision 10 (2/10/86) provides such a position on Technical Specification 3.0.3, which states that 11to show intent of compliance with this requirement, load should be reduced immediately at a rate determined by the senior shift supervisor, however, if it is likely that compliance with the Action Statement can be achieved within one hour, load does not have to be reduce The licensee has revised OD-12 to require load to be reduced one hour after entering Technical Specification 3. The inspectors observed that the present Technical Specification Inter-pretations provided in OD-12 do not require SORC revie The Technical Specification Interpretations are presently approved by the Operations Manage The licensee committed in the NRC exit meeting to have all Technical Specification Interpretations SORC approved (Unresolved Item 272/89-27-03).
The licensed operators showed poor judgement in allowing the LCD action statement requirement of tripping the channel within one hour to be exceeded prior to taking actiono Licensed operators are trained and capable of removing inoperable channels from servic Communications on the urgency of removing the channel from service was not adequately stressed to the J&C personnel by the operator Using separate sections of I&C and Operations procedures for removing the NI channel from service was unacceptabl By removing the detector cables using the I&C procedure, which is not part of the operations procedure, the re-energization of the N44 trips was permitted to occu Following the root cause determination, it was further revealed that when reactor power was decreased below 10%, the re-enerization of the trip bistables would have occurred even if the detector cables were installe Therefore both I&C and Operations pro-cedures for placing an NI channel in trip per Technical Specifications were inadequat The root cause analysis for the reenergization of the trip bistables upon installation of the PIO jumper was not determined until after the NRC inspection_ team arrivedo This was twenty days after the even The lack of timely root cause analysis on the part of the licensee is a violation of NRC regulations (Violation 50-272/89-27-04).
Operator communications and actions taken during this event were reviewedo The normal on-shift operating crew had been supplemented by the unit Operating Engineer and two I&C personnel dedicated to support test performanceo
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Communication among all party's involved*in this test were viewed as being wea I&C technicians were not made cognizant of the one hour Technical Specification time limit until forty minutes into the LC I&C personnel were also not made aware of the required support requirements, by the Test Director, prior to starting the test. Communications by shift operators to plant management when Technical Specification 3.0.3 was entered was also wea The Operations Manager was not made cognizant of the details of entering Technical Specification 3.0.3 in a timely manne These weaknesses had been appropriately identified during licensee review of the occurrenc * The Calormetric Calculation Procedure, 11 Rx ENG. MAN. PART 2, 11 provides guiduance on when NI power must be adjusted following a calormetric calculatio This procedure requires NI adjustment if the thermal power is plus or minus 1% of the indicated NI powe A second requirement for adjustment is that the average of the four NI channels must be equal to or greater than the reactor thermal powe By not specifically requiring NI adjustment when an NI channel indicates greater than reactor thermal power, a maximum of 1% nonconservative difference may exist between actual and indicated powe Subsequent channel drift could exceed the 1% difference between th~ High Reactor Power Tri~ Setpoint and the allowable value required by the Technical Specification The significance of this event was minimal since no channel exceeded the 1% allowable value. During the inspection, NRC also questioned which channel must be considered inoperable when the criterion not met is the average of four channel The licensee stated at the exit meeting that an evaluation would be made as to the adequacy of the acceptance criteria and its implementation with regard to Technical Specification operability (Unresolved Item 272/89-27-05).
7.0 Exit Meeting A summary of the inspection findings was discussed with the licensee at the conclusion of the inspection on December 1, 198 Additional discussions with the licensee were held on December 11, 1989, and during a telephone discussion on January 8, 1990.
- Attachment A Sequence Of Events
_ Turbine Volumetric Flow Test November 9,1989 Time Event 0130 Operating shift verifies portions of the Turbine Volumetric Flow Test prerequisite First boration made to reduce Tavg 2°F from normal 100% Tavg of 571° Tavg @569°F first Calormetric Calculation being performe First adjustment to NI 1s mad Second boration made to reduce Tavg an additional 2° Turbine Governor Valves full ope Tavg 567°F, second Calormetric Calculation being performe Second adjustment of NI's being mad N41-N43 successfully adjusted N44 fine gain bottoms ou I&C 11paged 11 to perform course gain adjustment on N4 I&C did not hear page at this tim I&C contacted to adjust course gai SNSS declares N44 inoperable, starts one hour LCD clock to trip or restore N44, concurred by Test Director and NS I&C contacted to expedite N44 adjustmen Operations personnel make I&C supervisor aware of one hour Technical Specification requiremen I&C told by operations to place N44 in trip to comply with Technical Specification LCO Action Statement 3.3. One hour Technical Specification Action Statement 3.3.1.l exceeded Tech. Spec. LCO 3.0.3 entered.
0641 N44 bistables tripped in accordance with I&C procedur NSS notes that PlO jumper not installed per Operations procedure IV 10. I&C Technician leaves control room to acquire material required for PlO jumpe Channel still considered not tripped LCO 3.0.3 continue I&C Technician installs PlO jumper in accordance with Operations Procedure IV 10. Bi~tables for HIGH FLUX RATE TRIP, HIGH FLUX TRIP (high/low setpoint) reenergiz SNSS verifies proper placement of PIO jumpe SNSS orders removal of jumpe Dilution commenced to return Tavg to program. 0840 Tavg 569°F, *Calormetric Calculation being performe NI's adjusted N44 calibrated and returned to servic Technical Specification LCO 3.0.3 and 3.3.1.1 cleared.
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PS~G RESIDUAL HEAT REMOVAL COLD LEG INJECTION ISOLATION VALVES SJ49 DECEMBER 11, 1989
PUBLIC SERVICE El;.ECTRIC AND GAS NRC ENFORCEMENT.CONFERENCE SJ49 VALVES AGENDA INTRODUCTION PSE&G UNDERSTANDING OF NRC CONCERN SJ49 MODIFICATION HISTORY SYSTEM DESIGN LICENSING BASIS ORIGINAL CONTROL CIRCUIT DESIGN MODIFIED CONTROL CIRCUIT DESIGN USQ DISCOVERY SAFETY SIGNIFICANCE SHORT TERM ACTIONS IMMEDIATE COMPENSATORY ACTIONS
- soRC REVIEW AND EFFECTIVENESS LONGER TERM ACTIONS INVESTIGATION RESULTS CORRECTIVE ACTIONS INDEPENDENT REVIEW PSE&G ASSESSMENT OF POTENTIAL VIOLATION APPLICATION OF GENERAL ENFORCEMENT POLICY SUMMARY TECHNICAL SPECIFICATION 3.0.3 POLICY T. M. CRIMMINS J. BAILEY L. K. MILLER J. RONAFALVY B. A. PRESTON T. M. CRIMMINS S. LABRUNA
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NRC FINDINGS ~
APPARENT VIOLATIONS FAILURE TO IDENTIFY THAT OCR 1EC~2295 FOR UNIT 1 AND 2EC-2295 FOR UNIT
CONTAINED A USQ BY INTRODUCING A POTENTIAL SINGLE FAILURE WHICH COULD HAVE JEOPARDIZED THE ABILITY OF THE ECCS SYSTEMS TO PERFORM THEIR SAFETY FUNCTION DURING A LOCA FAILURE OF SORC TO IDENTIFY THAT PROPOSED CHANGES TO EOPs ON NOVEMBER 17 CONTAINED A USQ; SINGLE FAILURE VULNERABILITY NOT COMPLETELY RESOLVED.
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USO DISCOVERY SEPTEMBER 29, 1989 POTENTIAL SINGLE FAILURE CONCERN IDENTIFIED SENT TO PRA GROUP TO PRIORITIZE CONTINUED RESEARCH INTO LICENSING BASIS AND SINGLE FAILURE CRITERIA
NOVEMBER 13, 1989 DEF RETURNED TO EAG WITH NEAR TERM PRIORITY USING QUALITATIVE PRA REVIEW
NOVEMBER 14, 1989 EAG SUPERVISOR CONCURS WITH PRIORITY NOTIFICATION OF MANAGEMENT AND STATION AS PER PROCEDURE MANAGEMENT MEETING WITH STATION QUESTIONS OVER VALIDITY OF SINGLE FAILURE REQUEST FOR FURTHER RESEARCH WITH WESTINGHOUSE
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USO DISCOVERY (CONT.)
NOVEMBER 17, 1989 ALL RESEARCH COMPLETE NO CHANGE IN ORIGINAL CONCLUSIONS STATION NOTIFIED INCIDENT REPORT PREPARED TSAS ENTERED NOTIFICATION MADE TO NRC COMPENSATORY ACTION IMPLEMENTED ENGINEERING EVALUATION REQUESTED BY SORC INITIATED INCIDENT OSR INDEPENDENT INVESTIGATION
NOVEMBER 20, 1989 NRC QUESTIONS ON COMPENSATORY ACTIONS /* NOVEMBER 21, 1989 EOP REVISIONS REVIEWED AND APPROVED BY SORC
NOVEMBER 22, 1989 OF ENGINEERING EVALUATION PRESENTED TO SORC AND APPROVED
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LOCA ANALYSIS (EXISTING) LARGE LOCA LIMITING BREAK IS LOCA WITH CD = * LIMITING SINGLE FAILURE ONE RHR PUMP FAILURE OF SI PUMPS ON AFFECTED TRAIN ALSO ASSUMED
ANALYSIS ASSUMES FLOW TO ALL COLD LEGS (I.E. 3 INTACT LOOPS) CALCULATED FLOW FROM ONE SI TRAIN IS 3374 GPM AT 25 PSIA RCS PRESSURE
CALCULATED PCT IS 2091 F (BASH ANALYSIS)
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.... *,. LARGE LOCA SAFETY EVALUATION LIMITING BREAK IS LOCA WITH CD = LIMITING SINGLE FAILURE = CLOSURE OF ONE SJ49 ALL PUMPS INCLUDING BOTH RHR PUMPS ARE RUNNING EVALUATION ASSUMES RHR FLOW TO ONE COLD LEG CALCULATED TOTAL FLOW TO ONE LOOP 2864 GPM AT 25 PSIA RCS PRESSURE (BY WESTINGHOUSE) CALCULATED PCT PENALTY DUE TO SI DEGRADATION IS 29 (BY WESTINGHOUSE) NO ADDITIONAL PCT PENALTY FROM ASYMETRIC FLOW DELIVERY (BY WESTINGHOUSE) CALCULATED NEW PCT: 2091 + 29 = 2120 NEW PCT REMAINS LOWER THAN 2200 F MINIMAL SAFETY SIGNIFICANCE . ' ~*.
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PROBABILITY ANALYSIS
PROBABILITY OF OCCURRENCE COMBINED PROBABILITY OF LARGE LOCA AND RANDOM SINGLE FAILURE OF CONTROL CIRCUIT (2.5 x 10-ll)
DETECTABILITY OF CONTACT FAILURE INDEPENDENT VALVE POSITION INDICATIONS AND ALARMS OVERHEAD ALARM ON EITHER VALVE NOT FULLY OPEN
... IMMEDIATE ACTIONS
INCIDENT REPORT GENERATED
TECH SPEC ACTION STATEMENT 3.0.3 WAS ENTERED
NOTIFICATION TO NRC FOLLOWED PROMPTLY
CONVENED A SORC MEETING
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COMPENSATORY ACTIONS (TAKEN AT 11/17/89 SORC MEETING)
TAGGED SJ49 MOTOR BREAKERS IN OPEN POSITION TO ELIMINATE SINGLE FAILURE CONCER *
REVISED OPERATOR LOGS TO REQUIRE VERIFICATION OF SJ49 OPEN POSITION, EACH SHIFTo CONDUCTED BRIEFINGS WITH SHIFT PERSONNELo
REVISED EOP'S TO ENSUF,E SJ49 1 S ARE POWERED-UP FOR SWITCHOVER TO RECIRCULATIO..
T=O RX TRIP/SI DBA LOCA TIMELINE FOR SJ49 -ENERGIZATION 3 MIN 39 SEC EOP-TRIP-1 DISPATCH NEO'S TO CIRCUIT BREAKERS 5 MIN 9 SEC NEO REACHES 78 1 ELEV 8 MIN 9.SEC ELEC PEN CLOSES BREAKER NEO REACHES 84 1 ELEV AUX BLDG CLOSES BREAKER '*.*.... _ '"\\ *... .***:.... *...
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... 1 HR 4 MIN EOP-LOCA~J CLOSE SJ49 VALVE
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ROOT CAUSE ROOT CAUSE AND CORRECTIVE ACTIONS
FAILURE TO IMPLEMENT REQUIREMENTS OF AP-32 AND PERFORM 50.59 EVALUATIO CORRECTIVE ACTION
COMPLETE REVIEW OF INCIDENT.WITH ALL SORC MEMBERS
COMPLETE RE-EVALUATION OF SORC REVIEW PROCESS FOR POTENTIAL ENHANCEMENTS
INCIDENT WILL BE REVIEWED WITH ALL STATION PERSONNEL REQUIRED TO APPLY AP=3 * COMPLETE REVIEW AND REVISION OF NAP-32
TRAIN AND QUALIFY STATION QUALIFIED REVIEWERS TO NEW NAP~32 .. _.... *.
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I I SORC REVIEW NRC CONCERN
FAILURE OF SORC TO IDENTIFY USQ IN PROPOSED CHANGES TO EOP' SUMMARY OF ISSUE
SORC UNDERSTOOD SINGLE FAILURE POTENTIAL WAS ELIMINATED WITH SJ49 BREAKERS TAGGED OPE ADDITIONAL ENGINEERING REVIEW CONCLUDED THAT CIRCUIT COULD NOT BE PRE-CONDITIONED WITH A CREDIBLE SINGLE FAILUR * INITIAL EOP CHANGE - RESTORE BREAKER POWER ASA REVIEWED IN DEPTH BY SORC SORC RECOGNIZED MINIMAL SINGLE FAILURE POTENTIA QUALITATIVE PRA EVALUATION - MINIMAL RISK EARLY RESTORATION OF BREAKER POWER JUDGED TO BE PRUDENT FROM*A HUMAN FACTORS VIEW POIN o SWITCHOVER TO RECIRCULATION LESS COMPLICATE o ELIMINATES POTENTIAL FOR MISCOMMUNICATION TO NEO DURING CRITICAL EVOLUTION IN EOP 1 So o SIMILAR LOGIC APPLIED TO ACCOMPLISH OTHER EOP ACTION o UNNECESSARILY TIES UP NEO DURING EARLY PHASES OF ACCIDEN * SORC TOOK INCOMPLETE COMPENSATORY ACTIO * FINAL EOP CHANGE - STANDBY NEO TO CLOSE BREAKER FOR RECIRCULATION SWITCHOVE,.. *.. *....,-., -.* __ **: :*: '> **.,. ~-.,.-
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STATION OPERAT.IONS :REVIEW COMMITTEE fSORC) EFFECTIVENESS
SORC REVIEW ACTIVITY REVIEW OF LER' S, DCP 1 S, TECH SPEC CHANGES, PR~CEDURES, VIOLATION RESPONSESi ET > 100 OPEN ITEMS IDENTIFIED FOR RESOLUTION OVER LAST 4 YEAR OF 638 (8%) ITEMS REVIEWED IN 1989 TO DATE REJECTE * OSR PERFORMS INDEPENDENT REVIEW OF ALL 50. 59 SAFETY EVALUATION TO DATE, NO SIGNIFICANT SAFETY ISSUE IDENTIFIED AFTER SORC REVIE * QA AUDIT PERFORMED EVERY 2 YEARS WHICH COVERS ALL SORC ACTIVITIES.* LATEST AUDIT ( 8 / 8 8) IDENTIFIED NO NEGATIVE FINDINGS o CONCLUDED "SORC" FUNCTION IS EFFECTIVEo
THIRD PARTY REVIEW PERFORMED BY IMPELL (EARLY 1989) SORC REVIEW PROCESS FOUND TO BE ADEQUAT * NRC INSPECTION (3/89) CONCLUDED THAT SORC REVIEW OF DCP'S WAS ADEQUAT * AMERICAN NUCLEAR INSURERS (ANI) AUDIT (10/89) OF SORC ACTIVITIES - DETERMINED TO BE EFFECTIV * SRG MEMBERSHIP ON SORC PROVIDES FOR AN INDEPENDENT VIEWPOIN SORC REVIEW PROCESS IS EFFECTIVE * . *.:.. :,
INVESTIGATION OF EVENTS AND CAUSAL FACTORS THAT LED UP TO THE EVENT. REVIEWED:
APPLICABLE PROCEDURES USED IN 1987
PLANT DESIGN AND DESIGN CRITERIA
PEOPLE ISSUES: ENVIRONMENT TRAINING/QUALIFICATION
EXISTING PSE&G AND INDUSTRY EXPERIENCE ON SIMILAR ISSUE NO INFORMATION WAS-FOUND ON SIMILAR CIRCUIT QESIGN ISSUESo
OTHER CHANGES MADE ON POWER LOCKOUT CIRCUITS AND CHANGES MADE DURING THIS TIMEFRAM NO OTHER CONCERNS IDENTIFIEDo . '. ~.... >>:_ .. ~... -... " .: :..
. '....... CONCLUSIONS THE ERROR IN DESIGN WAS THE RESULT OF AN INADEQUATE REVIEW OF DESIGN BASE DOCUMENTATIO THE REVIEW FAILED TO IDENTIFY THE PECULIARITY OF THE CIRCUIT DESIGN REQUIREMENT FOR MITIGATING SINGLE FAILURE CRITERION IN THE INJECTION MODE OF THIS SYSTE WE BELIEVE THIS TO BE AN ISOLATED EVEN "*
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MAJOR CONTRIBUTING FACTORS ENVIRONMENT PLANT SHUTDOWN REORGANIZATION
THE UNIQUENESS OF THIS CIRCUIT 1 S CHARACTERISTIC FOR MITIGATING SINGLE FAILURE WAS NOT COMPLETELY UNDERSTOOD BY THE ENGINEER WHO WORKED ON THE DESIGN CHANGE
PROCEDURE IN 1987 DID NOT REQUIRE DOCUMENTATION OF THE DETAILS OF AN FSAR REVIEW SUBSEQUENT REVIEWERS DID NOT HAVE EXPLICIT KNOWLEDGE OF THE UNIQUENESS OF THE CIRCUIT
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CORRECTIVE*ACTIONS. THE FOLLOWING ACTIONS HAVE BEEN, ACCOMPLISHED TO PREVENT RECURRENCE OR WILL BE EXECUTE MODIFICATIONS TO SJ49 CIRCUITS TO REESTABLISH ORIGINAL DESIGN REQUIREMENT TO BE ACCOMPLISHED DURING THE NEXT REFUELING OUTAGE PROCEDURES/DOCUMENTS/PROCESSES: o 1987 PROCEDURE AND CURRENT DESIGN CHANGE PROCEDURE COMPLY WITH APPENDIX B REQUIREMENTS CURRENT PROCEDURE PROVIDES FOR BETTER ORGANIZATION AND INSTRUCTIONS FOR DEVELOPMENT OF THE DESIGN CHANGE AND FOR BETTER DOCUMENTATION OF THE CONCLUSIONS o PROCEDURE FOR DOING 50.59 EVALUATION WAS REVIEWE NSAC 125 AND OUR INDEPENDENT AUDIT ENHANCEMENTS HAD BEEN INCORPORATE DOCUMENTATION OF SECTIONS OF FSAR AND OTHER DOCUMENTS REVIEWED IS NOW A REQUIREMEN NO 'FURTHER ACTION REQUIRE o MODIFIED DEF RESOLUTION PROCESS TO PREVENT POTENTIAL TIME LAPSE o FSAR SECTIONS DEALING WITH THIS ISSUE WILL BE REVIEWED WITH REQUIREMENT THE INTENT TO CLARIFY
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CORRECTIVE ACTIONS (CONT.)
PERSONNEL o TRAINING ON 50.59 PROCEDURE AND PROCESS DCP PROCEDURE AND PROCESS ENGINEERING TRAINING PROGRAM CONTINUE OUR DEVELOPMENT OF ECCS, LOCA AND EOP ANALYSIS EXPERTISE IN ENGINEERING INITIATED AN INDEPENDENT INVESTIGATION BY OUR OFF-SITE SAFETY REVIEW GROUP ON THE DESIGN ERRO FURTHER* ACTIONS TO CORRECT OR STRENGTHEN OUR PROGRAM MAY RESULT NUCLEAR DEPARTMENT IMPROVEMENT PROGRAM INITIATED PARTICIPATING IN NUMARC SUBCOMMITTEE ON DISCREPANCY RESOLUTION PROCESS DISSEMINATE LESSONS LEARNED
- PSE&G ASSESSMENT OF *POTENTIAL VIOLATION APPLICATION OF GENERAL ENFORCEMENT POLICY (lOCFR PART 2. APPENDIX Cl
APPLICATION OF MITIGATING FACTORS~ IDENTIFICATION AND REPORTING o DCP DEFICIENCY WAS SELF IDENTIFIED BY PSE&G o PROMPTLY REPORTED THE VIOLATION TO NRC CORRECTIVE ACTION TO PREVENT RECURRENCE o IMMEDIATE COMPENSATORY ACTIONS TAKEN IN PLANT o IMMEDIATE INVESTIGATION UNDERTAKEN ON DCP PAST PERF9RMANCE o PERFORMANCE IN APPLICATION OF 50. 59 IN DCP PROCESS HAS BEEN VERY GOOD. ISOLATED CONCERN PRIOR NOTICE OF SIMILAR EVENTS
NO SPECIFIC INDICATION MULTIPLE OCCURRENCES NRC, o ISOLATED DEFICIENCY SAFETY SIGNIFICANCE INDUSTRY, OR
0 DETERMINISTIC EVALUATION ~ PCT < 2200 F PROBABILITY OF OCCURRENCE 2.5 x10=11 OTHER
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PSE&G ASSESSMENT OF POTENTIAL VIOLATION APPLICATION OF GENERAL ENFORCEMENT POLI ClOCFR PART 2. APPENDIX Cl
APPLICATION OF NRC DISCRETION PSE&G AGGRESSIVE IN IDENTIFYING, REPORTING, AND CORRECTING VIOLATIONS o NOT REASONABLY PREVENTABLE BASED ON PRIOR NRC, INDUSTRY, OR PSE&G EXPERIENCE OR NOTICE
0 NOT.WILLFUL DOES NOT REPRESENT A BREAKDOWN IN *MANAGEMENT CONTROLS BASED ON MITIGATING FACTORS AND APPLICATION OF NRC DISCRETION, PSE&G BELIEVES ESCALATED ENFORCEMENT SHOULD NOT BE APPLIED TO THE SJ49 ISSUEo
SoMMARY
A THOROUGH EVALUATION OF THE RHR COLD LEG INJECTION ISOLATION VALVE DEFICIENCY HAS BEEN PERFORMED
PRELIMINARY ROOT CAUSES HAVE BEEN DETERMINED
CORRECTIVE ACTIONS HAVE BEEN TAKEN AND ARE CONTINUING DEFICIENCY HAS MINIMAL SAFETY IMPACT
STUDY PERFORMED TO ENSURE THAT DESIGN CHANGES HAVE NOT VIOLATED SINGLE FAILURE CRITERIA FOR OTHER SIMILAR CONTROL POWER LOCK OUT CIRCUITS
COMPENSATORY ACTIONS HAVE BEEN TAKEN AND PLANT IS CURRENTLY IN COMPLIANCE
DCP DEFICIENCY WAS SELF-IDENTIFIED THROUGH A PROGRAMMATIC ASSESSMENT AND IS A POSITIVE INDICATION OF PSE&G's INTENTION TO IDENTIFY/CORRECT PROBLEMS
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<\\) ~ ... '. - .**..... ...... ; T/S 3.0.3 POLICY . CURRENT POLICY
UPON ENTRY INTO T/S 3.0.3 START PREPARING FOR SHUTDOWN
INITIATE POWER REDUCTION NO LATER THAN ONE HOUR AFTER ENTERING T/S 3. * MAKE 10 CFR 50.72 ONE-HOUR REPORT WITHIN 60 MINUTES OF INITIATING A POWER REDUCTIO PLANT MANAGEMENT HAS CLEARLY ARTICUIATED POLICY TO ALL LICENSED OPERATORS .. **.. -- ~...... :.'*;...,...:.. -:-* :**-.. *:. }}