IR 05000272/1986011
| ML18092B163 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 05/23/1986 |
| From: | Norrholm L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18092B162 | List: |
| References | |
| 50-272-86-11, 50-311-86-11, NUDOCS 8606090075 | |
| Download: ML18092B163 (41) | |
Text
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Report No Docket No License No Licensee:
U. S. NUCLEAR REGULATORY COMMISSION 50-272/86-11 50-311/86-11 50-272 50-311 DPR-70 DPR-75
REGION I
050311-860320 Public Service Electric and Gas Company 80 Par_k Plaza Newark, New Jersey 07101 Facility Name:
Salem Nuclear Generating Station - Units 1 and 2 Inspection At:
Hancocks Bridge, New Jersey Inspection Conducted:
April 1, 1986 - May 12, 198 Inspectors:
T~ J. Kenny, Senior Resident Inspector R. J. Summers, Project Engineer K. H. Gibson, Reactor Engineer-In Training Reviewed by:
R. J. Summers, Project Engineer Reactor Pro e ts ec ion No. 2B, DRP Approved by:
Inspection Summary:
C ief, Reactor Projects Projects Branch No. 2, DRP
~zi-J/n;;
date Inspections on April 1, 1986 - May 12, 1986 (Combined Report Numbers 50-272/86-11 and 50-311/86-11)
Areas Inspected:
Routine inspections of plant operations including:
operational safety verification, maintenance, surveillance, review of special reports, licensee event followup, interpretation of Technical Specifications for the new Westinghouse fuel assemblies, thinning of the thimble tubes used in the incore detector system, an employee concern and the Commissioner's tou The inspection involved 128 inspector hour ~
PDR ADOCK 05000272 G
- Results: During this inspection no violations were identified, however several items identified in paragraphs 2, 6 and 8 will require further review by the licensee and inspector.
- DETAILS Persons Contacted Within this report period, interviews and discussions were conducted with members of licensee management and staff as necessary to support inspection activit.
Operational Safety Verification Documents Reviewed Selected Operators 1 Logs Senior Shift Supervisor 1 s (SSS) Log Jumper Log Radioactive Waste Release Permits (liquid & gaseous)
Selected Radiation Exposure Permits (REP)
Selected Chemistry Logs Selected Tagouts Health Physics Watch Log 2.2 The inspector conducted routine entries into the protected areas of the plants, including the control rooms, Auxiliary Building, fuel buildings, and containments (when access is possible).
During the inspection activities, discussions were held with operators, technicians (HP & I&C), mechanics, supervisors, and plant managemen The purpose of the inspection was to affirm the licensee 1 s commitments and compliance with 10 CFR, Technical Specifications, and Administrative Procedure (1)
On a daily basis, particular attention was directed to the fo 11 owing a re as:
Instrumentation and recorder traces for abnormalities; Adherence to LC0 1 s directly observable from the control room; Proper control room shift manning and access control; Verification of the status of control room annunciators that are in alarm; Proper use of procedures; Review of logs to obtain plant conditions; and, Verification of surveillance testing for timely completion.
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(2)
On a weekly basis, the inspector confirmed the operability of selected ESF trains by:
Verifying that accessible valves in the flow path were in the correct positions; Verifying that power supplies and breakers were in the correct positions; Verifying that de-energized portions of these systems were de-energized as identified by Technical Specifications; Visually inspecting major components for leakage, lubrication, vibration, cooling water supply, and general operating conditions; and, Visually inspecting instrumentation, where possible, for proper operabilit (3)
On a biweekly basis, the inspector:
Verified the correct application of a tagout to a safety-related system; Observed a shift turnover; Reviewed the sampling program including the liquid and gaseous effluents; Verified that radiation protection and controls were properly established; Verified that the physical security plan was being implemented; Reviewed licensee-identified problem areas; and, Verified selected portions of containment isolation lineu.3 Inspector Comments/Findings/Observations The inspector selected phases of the units' operation to determine compliance with the NRC 1 s regulation The inspector determined that the areas inspected and the licensee 1s actions did not constitute a health and safety hazard to the public or plant personne The following are noteworthy areas the inspector researched in depth:
- 5 Unit 1 Unit 1 began this report period in Mode 6 with a refueling outage in progres On April 8, 1986, the licensee notified NRC that discrepancies had been identified while performing a QA audit on E.Q. equipment (details in section 6 of this report).
On May 6, 1986, the unit was paralleled to the grid at 6:00 The unit operated at 200 MWe for an eight hour temperature soak period to perform the main turbine overspeed tes The test was performed satisfactorily on May 7 at 2:56 A balance move was made on the main turbine and the unit was returned to service on May 6 at 8:39 On May 12, 1986 at 7:45 p.m., the unit tripped from 95% power due to #14 Steam Generator Feedwater flow and level lo The trip occurred when both feed pumps were lost due to a limit switch, on the common preoperational cleanup loop isolation valve (BF-65),
going off its normally closed positio The switch is part of a trip demand circuit designed to protect the 600 lb. piping downstream of the BF-65 valve by tripping the main feed pumps if the valve should be opene At the close of this report period the unit was in Mode 3 with licensee investigation in progres The resident inspector will follow the event and document it when the LER is reviewe During this outage period the resident inspector selected 4 Design Changes to observe selected portions of installation and testing to verify adherence to NRC regulations and licensee procedure The four design changes listed below were installed to address NRC identified concern Installation of separate low suction pressure trips on the three Auxiliary Feedwater Pumps (AFP).
This change removes the existing common trip and installs separate trips on each AFP (DCR-lEC-2073).
The NRC concern was the low suction pressure trip was a key operated switch which had to be manually manipulate The switch has been replaced by three independent automatic trip switches located in separate auxiliary feed pump panels.
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Installation of upgraded condensate pumps (3).
This change will increase the condensate pressure by approximately 100 psig. (DCR-1EC-1380A)
The NRC concern was the abnormally large number of reactor trips due to feedwater problem The installation of the pumps on Unit 1 completes the condensate upgrade for the Salem Statio Installation of i seismically qualified Gamma Metrics Neutron Flux Measurement System in order to provide indication of the entire Source, Intermediate and Power ranges to the control room and Hot Shutdown pane (DCR-1EC-1558A)
These monitors were installed to satisfy Appendix R (Fire Protection Program for Nuclear Facilities operating prior to January 1, 1979)
and Regulatory Guide 1.97 (Post Accident Monitoring)
requirement Installation of new connectors for the rod drive syste (DCR-lSC-1578)
The concern was that the connection made by the old design directly caused several rod drops due to improper contact within the connecto More details on the installation of the above listed design changes appear in Section 3 (Maintenance Observations) of this repor The inspector determined that the licensee installed and tested the design changes in accordance with approved procedures, Q.C. was utilized, and that the appropriate operating procedures and system operating prints were upgraded prior to the startup of the system In addition to the above, the inspector witnessed outage activities, procedures, and housekeeping in the following area Fueling activities (from the control room, fuel building and containment)
Tagging and valve lineups of isolated systems Scaffold erections and removal Welding and burning permits
H.P. activities including proper adherence to procedures with regard to; entry and egress for controlled areas, removal of anti contamination clothing, frisking for contamination, and use of RWP 1 No violations were identifie.
Unit 2 Unit 2 began this report period at 100% powe On April 16, 1986 at 4:30 p.m., the unit tripped from 45% power due to turbine trip/reactor trip which was caused by High High Steam Generator level in #23 Steam Generator (SIG).
At 4:27 p.m. #22 SIG feed pump tripped, apparently due to water in the control oil syste The operators reduced load and*
placed #23 and 24 SIG feed regulation valves in manual due to valve and level oscillation Steam dumps opened, causing a swell in #23 SIG level, activating the High High S/G level signal which tripped the turbin All support systems functioned normall On April 17 the unit returned to servic On May 1, 1986, the licensee completed a detailed inspection of Unit 1 motor operated valve operators to confirm environmental qualification features as a result of quality assurance inspector observation Of 560 items inspected, the following results were tabulated:
14 junction boxes not sealed, 11 Limitorque valves with no T-drain, 4 Limitorque operators with brakes still installed but discon-nected, 4 valves without approved Conax connector The unit was completing a refueling shutdown and all above conditions have been repaire (For additional details see Section 6 of this report)
Unit 2 was operating at full power and an inspection of acces-sible components was initiate Similar problems were identified and the licensee elected to shut down the reactor for a complete inspectio During the controlled shutdown, with reactor power at 5%, the generator breaker was opened at about 3:35 Immediately thereafter, a reactor trip and safety injection were automatically initiated due to high steam flow with low T-averag All systems functioned as designed and about 3000 to 3500 gallons of borated water were injected before terminatio An Unusual Event, declared at 3:35 a.m. was terminated at 4:00 On May 3, 1986, the unit was brought on the line at 11:33 p.m. following a shutdown to correct licensee identified Equipment Qualification (EQ)
deficiencie The licensee completed repairs in Limitorque motors, stuffing boxes, and Conax connectors. (See Section 6 for details) The April 16th trip was a direct result of water in the lube oil and control oil system of #22 SIG feedwater pum Investigation by the licensee identified an orifice in the oil system between the control oil and lube oil portions, that had become clogged by the water oil mixtur This decreased the pressure on the control oil side causing the pump to tri The licensee also identified that the water was entering the oil system through the gland sealing syste Newly installed condensate pumps with an increased output pressure (100 psi) supply the seal water to the gland Seals that have become worn have been attributed to the increased water in the oi Another identified problem was the incorrect installation of a six way valve that directed the oil to the installed water/oil separator, which is shared by the two main feed pump The valve was directing the oil from the top of the sumps to the separator, rather than from the bottom of the sump The licensee corrected the six way valve and has installed a new separator of a larger capacity to remove moisture from the oi The new separator is removing the water from the oil and recent sampling shows little water in the main feed pump oil reservoir The licensee has incorporated a continuous sampling program to monitor for water in the oi The safety injection plant trip of May 1, 1986, has been attributed to the following; Just prior to opening the generator output breakers the reactor operator drove the control rods 11 in 11 in order to reduce reactor coolant temperatur This action was done to bring the level down in one of the steam generators, which was near the high level trip settin The Tave of the reactor coolant system continued to decrease below the Tave setpoint for one of the signals to the safety injection signa When the main generator breakers were opened several steam flow high signals were experienced, which made up the rest of the signal which caused a safety injectio A review of logs and recorder traces support the above scenari The
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reason for the steam flow high signals has been attributed to flashing in the condensate pots which are used for the reference leg on the steam flow transmitter The licensee duplicated the high steam flow conditions on the Unit 1 startup when the generator was taken off the line to perform the overspeed tes The licensee plans to instrument the steam flow signals and monitor plant shutdowns in the futur The licensee is also looking into a better design for steam flow transmitter The results of the licensee's actions will be reviewed when the data is availabl This item will be an inspector follow item (50-272/311/86-11-01). Maintenance Observations The inspector reviewed the following design change activities to verify that they were installed in accordance with approved procedures and in compliance with NRC regulations and recognized codes and standard The inspector also verified that the replacement parts and Quality Control utilized on the repairs were in complianc~ with the licensee's QA progra Design Change Number lEC-1380 Condensate Pump Replacement Procedures Reviewed Performance test records from manufacturer Installation package Work Order Uncoupled motor run test criteria System Startup
- 11 Condensate pump could not achieve the desired bearing temperatures during the test run and the pump and motor were replaced with a replacement uni lEC-1558 Installation of Source Range Certification for detectors, cables and brackets Work Order Training records for welders, electricians, and other crafts that worked on the installation, including the foreman
- lEC-2073 AFW Low Suction Pressure Trip lSC-1578 Installation of New Rod Drive Connectors
Welding, burning and grinding permits Scaffolding permits Installation package Work Order Testing of the circuits Welding, burning and grinding permits Installation package Witnessed portions of rod testing Work Order The inspector verified that the above design changes were accomplished in accordance with the above mentioned procedures and that operating procedures and control room prints were updated prior to the systems startup.
No violations were identifie Surveillance Observations During this inspection period, the inspector reviewed in-progress surveillance testing as well as completed surveillance package The inspector verified that the surveillances were performed in accordance with licensee approved procedures and NRC regulation The inspector also verified that the instruments used were within calibration tolerances and that qualified technicians performed the surveillance The following surveillances were reviewed:
Unit 1 SP(0)4.5.2H PD2.4.065 Throttling valve flow balance verification which verifies proper flow balance and flow rates to the ECCS subsystems; Low head and safety injection and high head safety injectio Channel functional test of 1R41B plant vent iodine monitor O.I. III-1.3.2 Turbine overspeed test No violations were identifie.
Review of Periodic and Special Reports Upon receipt, the inspector reviewed periodic and special report The review included the following:
inclusion of information required by the NRC; test results and/or supporting information consistent with design predictions and performance specifications; planned corrective action for resolution of problems, and reportability and validity of report informatio The following periodic reports were reviewed:
Unit 1 Monthly Operating Report - March 1986 Unit 2 Monthly Operating Report - March 1986 In addition, the inspector reviewed; Special Report 86-2 This report deals with the degradation of fire doors and the planned impairment of fire doors during the outage to prevent damage to doors from frequent use and the movement of equipment through the doors during the outag During the outage the fire doors that were impaired were periodically verified by the inspector to insure that the licensee was taking the proper compensatory measure In the report the licensee makes the following statement:
11 In the
. future, reports will not be submitted for 11 planned evolutions" involving the inoperability of fire detection, suppression or protection equipment for Unit 1 or Unit 2, as long as the action requirements are me These occurrences will, however, be documented and controlled in accordance with existing Administrative Procedures.
The resident inspector had discussions with the licensee to clarify the statement which appeared to be in conflict with Technical Specification As a result of the discussions the licensee will continue to submit reports in accordance with Technical Specifications even though they are 11 planned evolutions
- The inspector considers this item close Special Report 86-4 This report was submitted to document a diesel failure due to a jacket cooling water hose ruptur The inspector has verified that the licensee has taken the appropriate action and considers this item close No violations were identifie.
Licensee Event Report Followup The inspector reviewed the following LERs to determine that reportability requirements were fulfilled, immediate corrective action was taken, and corrective action to prevent recurrence had been accomplished in accordance with Technical Specification Unit 1 86-007 Environmental Qualification Discrepancies During the outage on Unit 1 the licensee identified the following discrepancies as a result of a Q.A. walkdown to verify the equipment previously installed; The conduit connectors on seven (7) junction boxes were not properly seale Four (4) solenoid operated reactor head vent valves did not contain the required Conax connector Although the brakes on four motor-operated valves (located outside containment) were electrically disconnected, the brake assemblies had not been physically removed fro~ the valve operator Eleven (11) motor-operated valves (located inside containment)
di~ not have the required T-drains installed in the operator housing The licensee performed a similar check in Unit 2 and began to identify similar discrepancies and elected to shut down the unit to perform a complete evaluation with the following results; The conduit connectors on nine (9) junction boxes were not properly seale The four (4) solenoid operated reactor head vent valves did not contain the required Conax connector Eight (8) motor-operated valves located inside containment did not have the required T-drains installed in the operator housing All of the above discrepancies have been corrected and the licensee is performing an evaluation to determine if similar conditions exist in other areas such as Appendix R modification The inspector does not have any more questions at this time; however, leaves an open item to followup on the E.Q. program during a later inspection (50-272/311/86-11-02).
Unit 2 86-001 Waste Gas Holdup System Not Continuously Sampled for Oxygen The licensee identified that Technical Specification 3.11.2.5 was violated when the waste gas holdup tank was not continuously sampled to verify oxygen content less than 2% by volum This condition existed for two days prior to discovery by the license The samples prior to the inoperability of the sampling system were less than 2% and the samples after the system was returned to normal were less than 2%.
The reason for the inoperability has been attributed to operating procedure OP II-12.3.3 which had not been updated when the method for sampling was changed to allow direct sampling of the waste holdup tanks without changing the position of the automatic sampling switc The procedure has been changed to preclude the mispositioning of the automatic sampling switc The inspector considers this item close No violations were identified by the resident inspecto.
Interpretation of Technical Specifications for the New Westinghouse Fuel Assemblies WEstinghouse informed the licensee that portions of the new fuel provided may contain greater than 1766 grams of uranium per fuel pi The 1766 grams per fuel pin is referenced in Section 5 of Technical Specification As a result the licensee will submit a new interpretation to Section 5 of Technical Specifications. The new interpretation is based on the following facts:
The local distribution of weight among the fuel pins within an assembly, and among assemblies, has a negligible effect on the core physics parameters and is not a significant factor in any safety analyse The average pin weight in a reload batch has a very small effect on the core physics, and is considered in the design of each fuel cycl Westinghouse has provided, to the licensee, a safety evaluation that concludes that although some of the fuel pins may exceed the 1766 grams, the average fuel pin mass average is less than 1766 gram The licensee submitted the evaluation to Region I for review and has stated that an amendment will be submitted to remove the referenc~ to individual fuel rod uranium weigh The resident inspector has no further questions at this time.
14 Thinning of the Thimble Tubes used in the Incore Detector System History The licensee has experienced thimble tube thinning in past operation on both units because of vibration in the region where the thimble tube enters the bottom of the fuel assembly nozzl The vibration causes wear and a wall thinning effect that has led to leaks that had to be isolated at the seal tabl This thinning effect is unique to Salem because a Westinghouse recommended design change was not incorporated prior to initial reactor startu Inspector Observations During this outage eddy current testing was performed on the thimble plugs yielding results of the thinning of the tube The tubes that exhibited greater than 50% wall thinning were isolated prior to startu The isolation of the tubes does not violate the Technical Specification requirements for the Incore Detection Syste The licensee also provided Safety Evaluations SER-86-042 and 043 which delineates the justification for operation of the system with tube thinnin With regard to the Safety Evaluations, experience and testing has shown that the tubEs can be degraded up to 90% without leakin The licensee is planning to perform another eddy current examination if a forced outage of sufficient duration occurs prior to the next refueling outag Also there is an analysis underway to pull the tubes back an inch in lieu of additional eddy current testin The licensee is also investigating the replacement of these tube This item will remain open pending further review by the licensee and inspector (50-272/86-11-03). Employee Concern On May 7, 1986, a concerned contractor employee called the resident inspector 1 s office with several concern The employee said there was no health physics coverage, and that he was afraid of being doused with contaminated water while working in the sump area of #21 Residual Heat Removal (RHR) Pum The inspector contacted licensee management and relayed the concer The licensee performed the followin conducted a walkdown and survey of the #21 RHR Pump sump area conducted additional discussions with the workers to reiterate the licensee's rules while working under an RWP conducted an additional valve line up of the system in question The inspector reviewed the RWP, and conducted a tour of the #21 RHR Pump sump area with the following finding *
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The RWP was correct; however, it had been changed to reflect the current conditions (less radiation and contamination) that allowed the continuous H.P. coverage to be discontinue A hose that had been newly installed in the sump area was not a drain hose as the contractor had expected but rather an additional means for removing water should it get into the sum The licensee was conducting surveys in accordance with approved procedure The inspector met with the concerned employee later io the day and the employee stated that he no longer had any concerns and was working in the are No violations were identifie Commissioner's Tour On April 18, 1986, Commissioner Bernthal and a member of his staff toured the Salem Statio The Commissioner accompanied the resident inspector and Section chief on a tour of the control room, turbine building, switchgear room and the controlled area of the auxiliary building.
No violations were identifie Unresolved Item Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations or deviation The unresolved item identified during this inspection is discussed in paragraph.
Exit Interview At periodic intervals during the course of the inspection, meetings were held with senior facility management to discuss the inspection scope and finding An exit interview was held with licensee management at the end of the reporting perio The licensee did not identify 2.790 material.
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APPENDIX A PSE&G SALEM/NRC REGION I MANAGEMENT MEETING DETAILS (5/6/86) Meeting Purpose:
A management meeting was conducted in the Region I Office on May 6, 1986, to discuss the licensee 1 s analysis comparing operating histories (reactor trips) of Salem Units 1 and This comparison was done in response to a recent SALP Board recommendatio The licensee 1 s preientation included the results of an earlier review, conducted in August 1985 at PSE&G 1 s initiative, to identify differences in performance at Units 1 and.
Attendees: PSE&G:
c. J. R. L. B. R. F. NRC:
McNeill; Vice President, Nuclear Zupko, Jr., General Manager, Salem Operations Burricelli, General Manager, Engineering and Plant Betterment Reiter, General Manager, Licensing and Reliability Hall, Manager, Reliability and Assessment Murray, Principal Engineer, Reliability Programs Thomson, Senior Engineer, Reliability T. E. Murley, Regional Administrator R. W. Starostecki, Director, Division of Reactor Projects (DRP)
S. J. Collins, Chief, Reactor Projects Branch No. 2, DRP L. H. Bettenhausen, Chief, Operations Branch, Division of Reactor Safety (DRS)
J. P. Durr, Chief, Engineering Branch, DRS L. J. Norrholm, Chief, Reactor Projects Section No. 2B, DRP T. J. Kenny, Senior Resident Inspector - Salem R. J. Summers, Project Engineer, Reactor Projects Section No. 28, DRP K. H. Gibson, Reactor Engineer, Reactor Projects Section No. 28, DRP Licensee Presentation and Discussion The licensee's presentation included the results of two separate studies conducted to identify operating differences, and the underlying causes, at Salem Units 1 and The first study was initiated by PSE&G Nuclear Department management in August 198 The areas reviewed were for overall lifetime performance and included capacity factors, reactor trips and forced outages, maintenance activities (corrective and preventative) and subjective comments from key plant personne The second study resulted from a SALP Board recommendation to review the Salem reactor trip data to identify operating differences and additional corrective actions, if necessar *
Appendix A
The details of the licensee's presentation are provided in Enclosure 1 to this repor The general conclusions of the two reviews are~ except for the most recent operating cycles, there are no fundamental differences in plant performance; and, that plant performance improvements are already underway which should enhance overall operation In addition, PSE&G management has set fairly high operating goals for Salem 1 and 2 for the long ter Improvements are needed in both human factors and equipment reliability to be able to achieve these goal The preventative maintenance effort is planned to increase and the cost benefit attitude for justifying plant improvements has changed from a cost basis to a reliability basi These factors should increase the equipment reliabilit The training instructors are now actively pursuing human factor impact throughout the plant so that these items can be incorporated into the various training programs and improvements made to the plant where necessar Region I staff questioned the potential need for improvements to balance of plant (BOP) equipment, especially items where one out of one *logic may exist, that cause reactor tr~p challenge Also, the staff asked whether the plant Technical Specifications needed to be changed to allow the achievement of Salem's long term goal PSE&G response to these items stated that key BOP equipment will be placed under a quality program which will result in new improved maintenance procedures for such; and, although some Technical Specification changes are planned to facilitate higher reliability, their goals are achievable within the current Tech-nical Specification requirement The staff had no further questions at this time and concluded that the licensee's approach was fundamentally sound and completes the recommended action as stated in the SALP.
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SALEM UNIT l/UNIT 2 PERFORMANCE COMPARISON o PURPOSE. OF THE STUDY o AREAS INVESTIGATED o FINDINGS o RECOMMENDATIONS/DISPOSITION SALEM REACTOR TRIP EVALUATION o SUMMARY OF REACTOR TRIP DATA (1981-FEB,, 1986)
o CORRECTIVE ACTIONS TAKEN/ONGOING o EFFECTIVENESS OF CORRECTIVE ACTIONS o CONCLUSIONS
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SALEM UNIT l/UNIT 2 PERFORMANCE COMPARISON PURPOSE OF THE STUDY o DETERMINE IF THERE IS A RE8L. DIFFERENCE IN PERFORMANCE b IDENTIFY FACTORS CONTRIBUTING TO DIFFERENCES IN PERFORMANCE o DEVELOP RECOMMENDATION.ON "METHODS FOR IMPROVING PLANT PERFORMANCE"
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AREAS REVIEWED o
CAPACITY FACTOR DATA VERSUS TIME o
REACTOR TRIP/FORCED OUTAGE DATA OVER PLANT LIFETIMES CAUSES <OPERATOR ERROR, MAINTENANCE ERROR, EQUIPMENT MALFUNCTION)
TIME OF OCCURRENCE <INITIAL START-UP, RESTART, POWER OPERATION>
o 11A/B 11 PRIORITY WORK ORDERS GENERATED DURING OPERATION CATEGORIZED BY SYSTEM AND COMPONENT FAILURE MODES VERSUS TIME FROM START-UP o
PREVENTIVE MAINTENANCE PROGRA o 11 GUT 11 FEELINGS FROM OPERATIONS, MAINTENANCE, AND PLANNING PERSONNEL-3-
FINDINGS PERFORMANCE DIFFERENCES BETWEEN UNITS 1 AND 2 o
LIFETIME CAPACITY FACTOR IS GREATER FOR UNIT 1
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51% VERSUS 45.5% CAS OF 2/28/86)
o UNIT 2 HAS MORE NON-TRIP FORCED OUTAGES FROM POWER OVER ITS LIFETIME FOR A SHORTER OPERATING PERIOD:
UNIT 2 - 14 OVER 730 EFPD
<LIFETIME AS OF* 2/28/86)
UNIT i -
8 OVER 1621 EFPD <LIFETIME AS OF 2/28/86)
MOST OF.THIS DIFFERENCE OCCURS IN THE CURRENT CYCLES CSIX FORCED OUTAGES FOR UNIT 2, ONE FORCED OUTAGE FOR UNIT 1)
MOST OF UNIT 2 FORCED OUTAGES IN THE CURRENT CYCLE DUE TO RCS LEAKS o
FOR CURRENT CYCLES, UNll 2 HAS 50% MORE PRIORITY A/B WORK ORDERS D.UR I NG OPERATION FOR THE F I RST F I VE MONTHS OF OPERATION
- o UNIT 1 HAS LONGER_CONTINUOUS RUN FOR CURRENT CYCLE CONTRIBUTING FACTORS* TO PERFORMANCE DIFFERENCES o
UNIT 2 CAPACITY FACTOR EXPECTED TO BE LOWER SINCE UNIT 2 IS A YOUNGER PLANT RANDOM INFANT EQUIPMENT FAILURES OCCUR EARLY IN PLANT LIFE THAT ARE DIFFICULT TO ANALYZE, EXPERIENCE HELPS TO CORRECT THESE LARGE VARIATIONS IN CAPACITY FACTOR EXPERIENCED FOR BOTH PLANTS A SIGNIFICANT CONTRIBUTOR TO THE LOW LIFETIME CAPACITY FACTORS FOR BOTH UNITS WAS THE TWO MAJOR EQUIPMENT PROBLEMS EXPERIENCED CGENERATOR AND REACTOR TRIP BREAK~R
- PROBLEMS) WH I CH ARE NOT EXPECTED TO RECUR o
MAINTENANCE WORK LOAD DURING THE LAST UNIT 2 REFUELING OUTAGE EXCEEDED THE RESOURCES AVAILABLE BOTH UNITS WERE DOWN AT THE SAME TIME FOR SEVERAL WEEKS
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UNIT 2 ENTERED THE OUTAGE THREE MONTHS EARLY WITHOUT PROPER PLANNING CDUE TO MAJOR GENERATOR PROBLEMS)
LIMITED MANPOWER AND SPARE PARTS RESOURCES AVAILABLE MORE OUTSIDE CONTRACTORS WERE USED WITH MINIMAL PSE&G OVERSIGHT-5-
PERFORMANCE SIMILARITIES BETWEEN UNITS 1 AND 2 FINDINGS o
PARTIAL OUTAGES DO NOT REPRESENT MAJOR CONTRIBUTORS TO CAPACITY FACTOR LOSS o
AVAILABILITY AND CAPACITY FACTORS FOR CURRENT CYCLES ARE RELATIVELY HIGH:
UNIT 1 UNIT 2 CAPACITY FACTOR AVAILABILITY CTHROUGH 2/86)
CTHROUGH 2/86)
68
73 o
REACTOR TRIP DATA IS SIMILAR FOR BOTH UNITS:
CURRENT CYCLE 1981 - FEB,, 1986 UNIT 1
55 UNIT 2 9 -
o BOTH UNITS HAVE HAD DIFFICULTY REACHING FULL POWER DURING INITIAL STARTUPS AND RESTARTS AFTER TRIPS o _ SEVERAL REACTOR TRIPS HAVE OCCURRED DURING TECH SPEC SURVEILLANCE ACTIVITIES o
NO OPERATIONS INDUCED TRIPS FROM POWER HAVE OCCURRED ON EITHER UNIT FOR LAST TWO CYCLES CONLY ONE Cl) TRIP IN FOUR (4) CYCLES FOR UNIT 1)
o OPERATOR ERROR, MAINTENANCE ERROR AND EQUIPMENT MALFUNCTION ARE ALL SIGNIFICANT CONTRIBUTORS TO TRIPS FOR BOTH UNITS-6-
PERFORMANCE SIMILARITIES BETWEEN UNITS 1 AND 2 FINDINGS o
THE TOP TEN CRITICAL SYSTEMS CBASED ON A/B WORK ORDERS DURING OPERATION FOR THE CURRENT CYCLES> ARE SIMILAR SIX (6) OF THE TOP 10 CRITICAL COMPONENTS DIFFER BUT FAILURE MODES ARE SIMILAR o
BOTH UNITS HAVE HAD SEVERAL REPEAT EQUIPMENT PROBLEMS o
THE MAJORITY OF THE EQUIPMENT PROBLEMS ARE VALVE RELATED CLEAKS, OPERABILITY, MECHANICAL FAILURES>
o THERE APPEARS TO BE NO APPARENT CORRELATION BETWEEN FAILURE MODE AND TIME AFTER START UP o
THERE ISN'T ANY SIGNIFICANT CORRELATION BETWEEN NUMBER OF WORK ORDERS PER MONTH VERSUS TIME FROM START UP.
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POSSIBLE CONTRIBUTORS TO PAST DEGRADED PERFORMANCE OF BOTH UNITS THERE WAS AN ENTIRE CHANGE OUT OF OPERATORS FOR BOTH UNITS OVER A 12-MONTH TIME PERIOD BY BEGINNING OF 1983 AND ALSO LOST APPROXIMATELY TEN EXPERIENCED OPERATORS AT THE SAME TIME IN 1980 TOTAL CHANGE OUT OF OPERATORS OVER SHORT PERIOD OF TIME NEGATED SOME LEARNING CURVE EFFECTS CURRENT SHIFT ROTATION fVIAY NOT BE OPTIMAL FROM A HUMAN FACTORS PERSPECTIVE SHIFT CHANGE EVERY SIX DAYS DIRECTION OF ROTATION Cl2x8, 4x12, 8X4) TRAINING PROGRAM HAD ONLY LIMITED BALANGE OF PLANT TRAINING OF SAFETY SYSTEMS AND MAJOR SECONDARY EQUIPMENT MANY OF OUR OUTAGES ARE BALANCE OF PLANT RELATED MAINTENANCE PERSONNEL NOT GIVEN DETAILED TRAINING ON INTERACTIONS OF SYSTEMS AND COMPONENTS
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. * LACK OF A COMPREHENSIVE PREVENTIVE MAINTENANCE PROGRAM IN THE PAST LIMITED MAINTENANCE/REPAIR PROCEDURES FOR BOP COMPONENTS LIMITED SUPERVISORY OVERSIGHT OF MAINTENANCE ACTIVITIES LACK OF A GOOD EQUIPMENT RELIABILITY DATA BASE ROOT CAUSE ANALYSIS HAS NOT ALWAYS BEEN FULLY EFFECTIVE WE HAVE NOT ALWAYS BEEN AGGRESSIVE IN PURSUING CORRECTIVE ACTION TO ROOT CAUSES
. * OVERLY RES TR I CT I VE TECHN I CAL SPEC IF I CAT IONS LACK OF BASIS FOR MANY LCO'S/AOT'S
EXCESSIVE TESTING/SURVEILLANCE REQUIREMENTS LACK OF DEFINITIVE BASES AND INCLUSION OF UNNECESSARY
-
.
INFORMATION MAKE TECH SPECS CONFUSING AND DIFFICULT TO INTERPRET
-9-
.SUMMARY OF RECOMMENDATIONS AND DISPOSITION RECOMMENDATION 1:
DEVELOP A PLANNED/STRUCTURED METHOD TO IMPROVE PLANT RELIABILITY DISPOSITION AN AVAILABILITY IMPROVEMENT TASK FORCE HAS BEEN FORMED WITH THE OBJECTIVE OF IDENTIFYING MEANS OF ACHIEVING INCREASED PLANT RELIABILITY AND AVAILABILIT RECOMMENDATtONS DUE TO VPN BY MID-198 RECOMMENDATION 2:
DEVELOP A MORE SYSTEMATIC APPROACH FOR PLANNING OF MAINTENANCE ACTIVITIES CMATCHING RESOURCES WITH WORKLOAD, SPARE PARTS AVAILABILITY, MORE PSE&G INVOLVEMENT, SYSTEMATIC EVALUATION PROCESS FOR CANCELLING OF WORK ORDERS>
DISPOSITION:
A NEW STATION CENTRAL PLANNING GROUP IS BEING FORMED WHICH WILL ADDRESS THESE CONCERN THE NEW GROUP WILL BE RESPONSIBLE FOR PLANNING ALL STATION MAINTENANCE ACTIVITIES, BOTH OUTAGE AND NON-OUTAGE, AND WILL BE MADE UP OF EXPERIENCED PERSONNEL-10-
- RECOMMENDATION 3:
THE CURRENT SIX DAY SHIFT ROTATION SHOULD BE RE-EXAMINED FROM A HUMAN FACTORS PERSPECTIVE DISPOSITION:
AN EMPLOYEE INVOLVEMENT PROGRAM CEIP> GROUP HAS BEEN FORMED TO EVALUATE THE CURRENT SHIFT ROTATIO THEIR FINDINGS.WILL BE PRESENTED TO PSE&G MANAGEMENT IN MID-1986 RECOMMENDATION 4:
INCREASED EFFORT SHOULD BE DIRECTED AT IMPROVING THE EFFECTIVENESS OF MAINTENANCE ACTIVITIES CBETTER ROOT CAUSE ANALYSIS, MORE SUPERVISORY OVERSIGHT, BOP PROCEDURES, BETTER UNDERSTANDING OF SYSTEMS INTERACTIONS)
D I SPOS I TI ON: *
ROOT CAUSE ANALYSIS WILL BE PERFORMED BY A NEW STATION SYSTEM ENGINEER GROUP <MORE EXPERIENCE WITH SPECIFIC SYSTEM/
COMPONENTS, OWNERSHIP OF SYSTEMS>
A BOP QUALITY PROGRAM IS UNDER DEVELOPMENT WHICH IS BASED ON THE INPO GOOD PRACTICE FOR BOP QUALITY PROGRAM THIS WILL RESULT IN NEW PROCEDURES BEING DEVELOPED FOR MAINTENANCE OF KEY BOP EQUIPMENT EFFORTS ARE UNDERWAY TO INCREASE SUPERVISORY OVERSIGHT OF-MA I NTENANCE ACT I v*1 TI ES NEW CENTRAL PLANNING PROCESS WILL HELP TO IDENTIFY CRITICAL SYSTEM INTERACTIONS EFFORTS ARE UNDERWAY TO DEVELOP A SYSTEM FOR RECOGNIZING
.
-
.
REWORK ON COMPONENTS-11-
- RECOMMENDATION 5:
DEVELOP A BETTER EQUIPMENT RELIABILITY DATA BASE
- DISPOSITION:
o A NEW MANAGED MAINTENANCE INFORMATION SYSTEM CMMIS) IS CURRENTLY UNDER. DEVELOPMENT WHICH WILL PROVIDE A MUCH IMPROVED DATA BASE RECOMMENDATION 6:
INCREASED EFFORT SHOULD BE DIRECTED AT TECHNICAL SPECIFICATiON OPTIMIZATION
. DI SPOS I TI ON:
o TECH SPEC OPTIMIZATION IS ONE OF THE AREAS BEING ADDRESSED BY THE AVAILABILITY IMPROVEMENT TASK FORCE-12-
CONCLUSIONS o
FOR THE CURRENT CYCLES, THERE ARE SOME NOTICEABLE DIFFERENCES IN PERFORMANCE BETWEEN UNITS 1 AND HOWEVER, THE REASON FOR THE DIFFERENCES CAN BE EXPLAINED AND CORRECTIVE ACTION IS BEING TAKEN WHERE NECESSAR o OVER THE PLANT LIFETIMES, THERE ARE NO MAJOR FUNDAMENTAL DIFFERENCES IN PERFORMANCE o
SEVERAL PROGRAMS ARE UNDERWAY WHICH WILL ENHANCE PERFORMANCE FOR BOTH UNITS-13-
SALEM REACTOR TRIP EVALUATION (1981 - FEBRUARY, 1986)
o SUMMARY OF TRIP DATA
.o CORRECTIVE ACTIONS TAKEN/ONGOING o EVALUATE EFFECTIVENESS OF CORRECTIVE ACTIONS o CONCLUSIONS-14-
- 0
SALEM REACTOR TRIP EVALUATION C 1981 - FEBRUARY, 1986 >
TOTAL NUMBER OF TRIPS IS HIGH-FOR BOTH UNITS C55/54> -
TRIPS/YEAR SHOW A DECLINING TREND A SIGNIFICANT PORTION OF TRIPS OCCURS DURING BOTH PLANT START-UP AND POWER OPERATION:
UNIT 1 UNIT 2 FROM START-UP
24 FROM POWER
30 -
o - MAJOR CAUSE OF TRIPS IS SPLIT BETWEEN EQUIPMENT MALFUNCTION AND HUMAN FACTOR ERRORS:
EQUIP. MALFUNCTION H. F. ERRORS EQU I p I DES I GN/_UNDETERM I NED UNIT 1 27 (49%)
. 23 (42%)
5 (9%)
UNIT 2 _
20 (37%)
28 (52%)
6 ( 11%)
o EQUIPMENT FAILURE AND HUMAN FACTOR TRIPS OCCUR FROM BOTH START-UP AND STEADY STATE POWER CONDITIONS
-
60~ OF HUMAN FACTOR TRIPS OCCUR FROM START-UP
- 60% OF EQUIPMENT FAILURE TRIPS OCCUR FROM POWER o
MOST HUMAN FACTOR TRIPS ARE EITHER OPERATOR OR MAINTENANCE RELATED THE_MAJORITY OF OPERATOR ERROR TRIPS OCCUR DURING THE STARTUP EVOLUTION THE MAJORITY OF THE MAINTENANCE RELATED TRIPS OCCUR DURING STEADY STATE POWER OPERATION-15-
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,..
TOTAL REACTOR PS VERSUS TIME CSALEM UNITS 1 AND 2)
25
- TOTAL
-
--
---
~ UNIT 2.
--- -
- -----.---*-. ---* -.,..----------.
1983 1984 0._____ ____
-.--*
1981 1982 1985 YEAR
25
~
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I
- STEAM GENERATOR LEVEL-RELATED TRIPS VERSUS TIME (SALEM UNITS 1 AND 2)
TOTAL
"' '*--- -
-
-
,_,__., ___,. ___... _~------...... *----....:..-.~***rw~*-----~--__.__,-... -... ~-----,-r-cL..a.o*~~-~-*...._-----.,.'-------"'t 1981 1982 1983 1984 1985.
YEAR
..
.
.
o
- EQUIPMENT FAILURES TEND TO BE RANDOM FOR BOTH UNITS WITH THE EXCEPTION OF THE MAIN FEEDWATER REGULATING VALVES NINE (9) TRIPS HAVE OCCURRED SINCE 1981 DUE TO MALFUNCTIONS OF FEED REGULATING VALVES CE.G,,
CONTROLLERS, LOOSE WIRES, ETC.)
o THE MAJORITY OF THE TRIPS ARE S.G. LEVEL-RELATED:
UNIT 1 UNIT 2 S.G. LEVEL RELATED 38 (69%)
38 C70%)
OTHER
16 o
STEAM GENERATOR LEVEL-RELATED TRIPS ARE SPLIT BETWEEN STEADY STATE POWER OPERATION C35 TRIPS) AND START-UP OPERATIONS C41 TRIPS)
o CAUSES OF STEAM GENERATOR LEVEL-RELATED TRIPS ARE SPLIT BETWEEN EQUIPMENT FAILURE (34 TRIPS> AND HUMAN FACTOR ERRORS C33 TRIPS)
o FOUR MAJOR CONTRIBUTORS TO S.G. LEVEL-RELATED TRIPS ARE:
OPERATOR DIFFICULTY IN CONTROLLING START-UP EVOLUTION (20 TRIPS)
STEAM GENERATOR FEED PUMP LOW SUCTION PRESSURE TRIPS C12 TRIPS)
oo FOUR C4) EQUIPMENT FAILURE RELATED oo TWO (2) HUMAN FACTOR RELATED oo SIX (6) PRESSURE TRANSIENTS-18-
..
. L
.
..,
STEAM GENERATOR FEED PUMP OVERSPEED TRIPS (8 TRIPS>
oo FIVE (5) EQUIPMENT FAILURE RELATED oo THREE (3) HUMAN FACTOR RELATED MISCELLANEOUS EQUIPMENT FAILURE C25 TRIPS>
OTHER LEVEL TRIPS oo EIGHT (8) HUMAN FACTOR RELATED oo THREE (3) UNDETERMINED o
SALEM REACTOR TRIP DATA GENERAL TRENDS ARE CONSISTENT WITH WESTINGHOUSE OWNERS GROUP TRIP REDUCTION PROGRAM INITIAL FINDINGS-19-
.
t...
.
.._
...
REACTOR TRIP CORRECTIVE ACTION STEAM GENERATOR LEVEL-RELATED TRIPS o
OPERATOR DIFFICULTY IN CONTROLLING START-UP INSTALLED AN AUTOMATIC CONTROL SYSTEM FOR THE START-UP BYPASS VALVES (1983)
UPGRADED STEAM GENERATOR LEVEL CONTROL SYSTEM oo INSTALLED E.Q. TRANSMITTER TO ALLOW LO-LO LEVEL SETPOINT TO BE REDUCED oo DEVELOPED ADJUSTMENTS TO THE LEVEL CONTROLLERS I REDUCE LEVEL CYCLING UPGRADED OPERATOR TRAINING oo NEW SIMULATOR OPERATIONAL (2/84)
oo OPERATORS TRAINING ON MANY TYPES OF LEVEL-CONTROL TRANSIENTS DURING PERIODIC REQUAL. TRAINING oo SIMULATOR MODEL ENHANCED TO MAKE TRANSIENTS LOOK MORE LIKE REAL PLANT oo RECENTLY HELD TEAM TRAINING SESSIONS WHICH EMPHAS I ZED COMMAND, CONTROL AND COMMUN I CAT I ON ;.
FUNCTIONS IN THE CONTROL ROOM DEVELOPING MODIFICATIONS FOR THE CONDENSER STEAM DUMP VALVES RESULTS:
S/G LEVEL TRIPS EXPERIENCED FOR CURRENT CYCLES DUE TO OPERATOR DIFFICULTY IN CONTROLLING START-UP SHOWS AN IMPROVING TREND UNIT 1 - ZERO (0) TRIPS IN NINE (9) START-UPS UNIT 2 - THREE (3) TRIPS IN NINE (9) START-UPS-20-
I
--.--
*
1_,
<
IL o
STEAM GENERATOR FEED PUMP LOW SUCTION PRESSURE TRIPS INSTALLED A TEMPORARY ALARM AND DEVELOPED AN OPERATION ALARM RESPONSE PROCEDURE TO IDENTIFY STEPS TO BE TAKEN TO AVOID FEED PUMP LOW SUCTION PRESSURE TRIPS
<TEMPORARY MEASURES UNTIL INSTALLATION OF NEW CONDENSATE PUMPS>
INSTALLED LARGER CAPACITY CONDENSATE PUMPS ON UNIT 2 oo LARGER PUMPS ALSO BEING INSTALLED ON UNIT 1 RESULTS:
NO S/G LEVEL TRIPS DUE TO LOW FEED PUMP SUCTION PRESSURE HAVE OCCURRED FOR THREE.YEARS ON UNIT 2 AND ONLY ONE <1>
TRIP HAS OCCURRED ON UNIT 1 IN LAST TWO YEAR o STEAM GENERATOR FEED PUMP OVERSPEED TRIPS PERFORM PERIODIC CHECKS ON S.G.F.P. OVERSPEED TRIP GEARS CHANGED PROCEDURE FOR PUTTING S.G.F.P. BACK TOGETHER AFTER MAINTENANCE RESULTS:
ONLY ONE S/G LEVEL TRIP DUE TO STEAM GENERATOR FEED PUMP OVERSPEED HAS OCCURRED IN THE LAST TWO YEAR THIS WAS DUE TO A MAJOR EQUIPMENT FAILUR REACTOR TRIP CORRECTIVE ACTIONS ALL TRIPS o
HUMAN FACTORS TRIPS ALL LER'S, BOTH TRIP RELATED AND NON-TRIP RELATED, ARE REVIEWED BY TRAINING DEPARTMENT AND FACTORED INTO REQUAL. TRAINING: IF APPROPRIATE UPGRADED OPERATOR AND MAINTENANCE PERSONNEL TRAINING UPGRADED OPERATING PROCEDURES IN THE PAST TWO YEARS IN RESPONSE TO SOME TRIPS THAT HAVE OCCURRED o
EQUIPMENT MALFUNCTION TRIPS S I GN IF I CANTL Y UPGRADED P. M.. PROGRAM MADE EQUIPMENT DESIGN CHANGES IN RESPONSE TO TRIPS THAT HAVE OCCURRED ON-LINE EQUIPMENT PERFORMANCE TRENDING o
GENERAL RESULTS:
GREATER OPERATOR AWARENESS AS A RESULT OF TEAMS WORKING TOGETHER FOR SEVERAL YEARS POST-TRIP REVIEW PROCESS HAS IMPROVED SIGNIFICANTLY EQUIPMENT MALFUNCTION AND HUMAN FACTOR TRIPS HAVE BEEN REDUCED BUT NOT YET DOWN TO AN ACCEPTABLE LEVEL-22-
ONGOING ACTIONS o
WESTINGHOUSE OWNERS GROUP TRIP REDUCTION ASSESSMENT PROGRAM <INVESTIGATING TRIP SETPOINT RELAXATIONS, METHODS OF BETTER S.G. LEVEL CONTROL)
TECH. SPEC. SUBCOMMITTEE ACTIVITY o
DEVELOPING AN IMPROVED STEAM G~NERATOR LEVEL CONTROL MODEL FOR INCORPORATION INTO TRAINING SIMULATOR o
INVESTIGATING AUTO-RUNBACK CIRCUITRY THAT WOULD MAINTAIN PLANT ON LINE, AT LOWER POWER LEVEL, FOR SOME EQUIPMENT MALFUNCTIONS <E.G., S.G.F.P. TRIP)
. o I NVEST I GAT I NG HOW TO I NCREASE THE REL I AB I LI TY OF THE FEEDWATER REGULATING VALVES CBF19'S)
ACTIONS RESULTING FROM THE AVAILABILITY IMPROVEMENT TASK FORCE FINDINGS.WILL*HELP FURTHER REDUCE REACTOR TRIP FREQUENCY l
NEW "PRO-ACTIVE" POSTURE-23-
- f.:,_; <.._
'
CONCLUSIONS o
SEVERAL CORRECTIVE ACTIONS.TAKEN HAVE BEEN EFFECTIVE IN
- REDUCING TR I PS o
SEVERAL PROGRAMS ALREADY UNDERWAY SHOULD HELP TO FURTHER REDUCE TRIPS IN THE FUTURE CE.G-., PM'S, CENTRAL PLANNING PROCESS, ETC I
)
'
o IMPROVEMENTS IN HUMAN FACTOR AND EQUIPMENT RELIABILITY ARE STILL NEEDED TO MEET LONG TERM AVAILABILITY/REACTOR TRIP GOALS AVAILABILITY IMPROVEMENT TASK FORCE IS ADDRESSING THIS-24-