IR 05000277/1986099
| ML20238E684 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 09/08/1987 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20238E676 | List: |
| References | |
| 50-277-86-99, 50-278-86-99, NUDOCS 8709150138 | |
| Download: ML20238E684 (79) | |
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ENCLOSURE I
U. S. NUCLEAR REGULATORY COMMISSION l
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REGION I
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SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE INSPECTION REPORT 50-277/86-99 AND 50-278/86-99 PHILADELPHIA ELECTRIC COMPANY i PEACH BOTTOM ATOMIC POWER STATION !
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ASSESSMENT PERIOD: FEBRUARY 1, 1986 TO MAY 31, 1987 BOARD MEETING DATE: JUNE 26, 1987
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j B709150138 870908 PDR ADOCK 05000277 G PDR
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SUMMARY OF RESULTS A. Overall Facility Evaluation During the assessment period, facility performance was determined to l be unacceptable in the areas of plant operations and assur..nce of I quality (AQ). A special NRC investigation ascertained that control '
room operators had been sleeping and otherwise inattentive to their !
licensed duties. As a result, NRC lacked reasonable assurance that '
the facility'would be operated in a manner to assure that the health and safety of the public would be protected and by NRC Order dated March 31,.1987, Unit 3 was placed in a cold condition and Unit 2 remained in the defueled condition until the end of the assessment perio The central reason for this unacceptable performance was that plant management was unable or unwilling to correct known deficiencies in operator conduct that had potentially significant safety consequence _
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Category Category Recent ;
Functional Area Pervious Period This Period Trend I
! (April 1, 1985 to (Februarv 1,1986 January 31,1986) to May .,1987) ; Plant Operations 2 *
l Radiological 2 2 Controls l Maintenance 2 2 Improving Surveillance 2 2 I Fire Protection and 2 3 Improving Housekeeping
.' Emergency 2 2 Preparedness Security and 3 2 Safeguards I Technical Support Not evaluated 2 Improving Training and 2 **
Qualification Effectiveness Assurance of Quality 3 * Licensing Activities 2 2 Refueling / Outages 1
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Performance was determined to be unacceptable as reflected in the issuance of a shutdown Order and, therefore, no SALP rating is appropriat **
'The extent to which apparent weaknesses in supervisor training contributed to the inattentive control room behavior leading to the March 31, 1987 shutdown Order was still under review at the close of the assessment period; therefore, no rating was assigned to this are ***
Refueling and outage activities were not evaluated this period as the planned Unit 2 outage began near the end of the perio _ _ - _ _ _ _ _ - _ _ _ _ - _
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I FUNCTIONAL AREA ASSESSMENTS Plant Operations (59.1%, 4142 hours0.0479 days <br />1.151 hours <br />0.00685 weeks <br />0.00158 months <br />)* Analysis The previous assessment period rating of plant operations was Category 2, declining. Problems were noted with respect to procedural compliance and an apparent complacent attitud Personnel errors by operators resulted in numerous reactor scrams. It was noted that the onsite review committee should become more self-critical . Plant operations were staffed with adequate licensed and nonlicensed personne During the current assessment period, resident and specialist inspectors routinely reviewed plant operation The functional area,of plant operations was also reviewed during the Region I diagnostic inspection. the probabilistic risk assessment, and special team inspection Near the end of the current assessment period and as delineated in the NRC Order, dated March 31, 1987, an ongoing NRC special safety investigation of licensed activities established that licensed control room personnel had at times been sleeping in the control room and otherwise inattentive to licensed dutie The inattentiveness was apparently most prevalent during the night shift (11:00 p.m. to 7:00 a.m.). Shift management (shift suparvisor and shift superintendent) either kr.ew and condoned the inattentiveness or should have known. In addition, plant management either knew or should have known and either took no actiun or inadeocate action to correct the situatio During the current assessment period, a total of 16 scram sig-nals occurred on both units. Unit 2 had two automatic scrams at power and Unit 3 had eight automatic scrams at power. In addition, Unit 3 had three manual scrams at power and three automatic scrams while shut down. This was a significant reduc-tion in scrams from the previous SALP perio Of the total scrams this period, four can be attributed to per-sonnel errors by plant operator A manual scram was initiated when it was discovered that a control rod was left full-in during a Unit 3 startup with the rod worth minimizer (RWM) out of service. The other three scrams were caused by a combination of personnel error and equipment malfunctions or design deficiencies. The root cause analysis for each trip is presented in Table *Precentage of inspection time, inspection hour _________________-__a
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The Plant Operations Review Committee (PORC) functioned wel Inspectors attending PORC meetings _ observed a questioning ,
attitude and frank discussions among nember During a Unit 3 startup early in the assessment period, an out-of-sequence control rod withdrawal occurre The startup was j being conducted with the RWM out of servic The event resulted in two major technical. specification (TS) violations:
inadequate operator checks with the RWM cut of service, and unauthorized bypassing of the rod sequence control . system (RSCS). A civil penalty was assessed for these TS violation The root cause was the failure of four licensed operators to follow procedures and an apparent complacent attitude by operations personnel. Licensee corrective actions included manually scramming Unit 3, disciplining the four operators, and
- analyzing the unit for the rod drop accident associated with this abnormal rod patter Plant operations are supported through administrative and technical operating procedures. Most procedures were technically. adequate to accomplish the intended activit During the probabilistic risk assessment team inspection, certain weaknesses and human factors concerns were identified with station blackout recovery procedures. These weaknesses could result in error or delay if operators conducting the procedure were not thoroughly familiar with off-normal system operations. In a number of cases, the licensee had previous ,
knowledge of the weaknesses and had plans for their correctio Sets of operating procedures, located on carts, allow operators to reposition them about the control room for ease of opera-tions. A g>>od practice was noted in having copies of some
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annunciator alarm cards posted in the plant. Cards containing the emergency diesel generator local control panel alarm {
response procedures are available at each engin j The licensee continues to pursue a formal' program to reduce the number of control room nuisance alarms. Periodic reviews by operators, shift technical advisors (STAS), quality assurance /
quality control (QA/QC) personnel, and plant management have resulted in a reduction in the number of continuously lit alarms to about 15 total with both units at powe l
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Control room and plant operators demonstrated the ability to successfully perform emergency operating procedures and accident recovery actions during an extensive NRC team inspection. In )
nearly 20 separate simulations, operators were asked to walk j through postulated emergency situations related to loss of I offsite and onsite electrical power, anticipated transients i
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without scram (ATWS), and containment venting. In all the i simulations, the operators were thoroughly familiar with the .
associated instrumentation and controls, systems, and proce-dures. In the cases assessed, the operators were able to com-pensate for the observed weaknesses in procedures because of their extensive experience ard trainin The plant has implemented the symptomatic emergency operating or transient re:ponse implementation plan (TRIP) procedures. These procedures are in the form of logic chart diagrams and are located conveniently near the center of each unit's control area. The logic diagrams were reviewed and found to be clear and easy to read. The TRIPS are available during simulator training classe Some emergency procedures remain event-oriented and the licensee needs to incorporate them into the TRIP Control room operator response to plant transients and reactor scrams is generally good, as evidenced by inspector observations during several transients. Operators effectively use the TRIP procedures. Timely and adequate operator response to several transients prevented unit scrams. These transients include unit runbacks caused by the EHC and recirculation systems, condensate and reactor feed pump trips, and feedwater level control system problem One significant instance of operator inattention to plant parameters was noted, however, prior to the EHC-induced, high-flux scram of Unit 3 on March 17, 1087. Neither the unit reactor operator nor the shift supervisor noted the perturba-tions on various control room indications that occurred for one and a half hours prior to the scram. Although these indications were not accompanied by control room alarms, che onshift opera-tors should have observed the meter and chart recorder perturba-tions caused by frequent turbine control vahe cyclin Licensed operators continued to perform well on NRC exams. Seven of eight reactor operators (R0s) and 13 of 14 senior reactor operators (SR0s) passed their licensing exams during this assessment period. In addition, six contractors passed a fuel-handling-only SRO exam. At the end of the assessment period, cperations had a staff of 17 SR0s and 19 R0s. The
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operating shift requires a TS minimum staff of five licensed l cperators (two SR0s, three R0s). The licensee uses a six-shift l rotation schedule.
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the shift superintendent during outage periods and during the conduct of daily meetings. The licensee has added a contract engineer to each shift and has assigned an SR0 for permits and
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i blocking coordination for the 1987 Unit 2 outage. An additional l nonlicensed operator was added to each shift. This operator is l
responsible for reactor water cleanup and condensate demineralized regeneration.
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non-operational personnel to use the north (Unit 3) door. This minimizes traffic in the shift superintendent's office are Recent utilization of chains across the north and south access L to the control room complex area (that is, control boards) has
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resulted in minimizing traffic and noise. Permission of an onshif t licensed operator is required to enter the control board area. The background noise in the control room associated with the paging system is a continuing concern. The licensee is implementing modifications to reduce background noise associated with the paging syste The licensee has implemented programs in an effort to improve overall communications and to increase employee awareness of l plant operations. Unit status boards are now located in the L security building entrance and exit areas. Message boards have
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also been added to inform personnel of unit operations, NRC inspections, current station schedules and other items of interes '
The licensee generally utilizes conservatism when making 10 CFR 50.72 reports to NRC. Usually, Licensee Event Reports (LERs) I are well written, suomitted in a timely fashion, and exhibit
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good root cause determination. There remsins, however, considerable room for improvement. A detailed review of a i sample of LERs by the NRC Office for Analysis and Evaluation of Operational Data (AE00) indicated that 7 of 15 inadequately addressed root cause determination and identification, 7 of 9 q involving personnel procedural error did not adequately discuss the details of the error, and 12 of 15 did not adequately address corrective actions taken or planned to reduce the probability of recurrence of the even The LERs are listed and described in Table 6. 10 CFR Part 21 reports are also well written and submitted in a timely manne i An inadvertent heatup and pressurization while in cold shutdown
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occurred on Unit 2 during the current assessment period. The !
licensee had removed shutdown cooling from service to inspect a l Limitorque motor operator, poor planning and inadequate ]
procedures for decay heat removal resulted in a reactor coolant temperature rise to 245 F. The event was identified and reported by the licensee, and corrective action was taken. The
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licensee's investigation, including the event and upset reports, was accurate and thorough, and identified the root cause.
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During.the NRC special team inspection, prior to and after the shutdown Order of March 31, 1987, instances of control room operator informality were noted. These instances included
. operators with their feet up on the computer console. The licenses is currently addressing this issue in response to the NRC Order. Also, a Nuclear Operations Monitoring Team has been established to monitor operator formality on a continual basi In summary, the declining trend in performance of operations, noted in the previous SALP report, was not arrested and resulted in unacceptable performance during this period. The' problems of operator complacency and procedural compliance continued and in 1 part led to the NRC shutdown Order.. The central reason for the "
unacceptable performance in this area was that plant management was unable or unwilling to correct known deficiencies in operator conduct that had potentially significant safety consequence ) Conclusion '
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Performance was determined to be unacceptable as reflected in the issuance of a shutdown Order and, therefore, no ;
SALP rating is appropriat . Recommendations 1
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B. Radiological Controls (12.2%, 853 hours0.00987 days <br />0.237 hours <br />0.00141 weeks <br />3.245665e-4 months <br />) Analysis Radiological Controls and Chemistry Analysis Performance in this area was rated category 2 during the previous assessment period. Weaknesses previously identified were in the radiation protection (RP) program and in the transportation are The radiological controls area was reviewed during specialist inspections, including radiation protection, radwaste and ef fluents, environmental monitoring, transportation, and chemistry controls. Radiological controls were reviewed during the di9 gnostic team inspection. The resident inspectors routinely reviewed selected area Licensee corrective actions, including RP program revisions committed to as part of the Enhancement Program and a radwaste reorgani:ation, appear to be substantia Recurrent problems in these areas indicate, however, that planned reforms have not yet affected operational performanc Radiation Protection The' current organizational structure and staffing levels in the !
Radiation Protection Department have been ir. adequate to improve performance since the previous period. Upper management does not provide aggressive leadership nor oversfght, as evidenced by a lack of long-range planning and program priorities. First-line supervision is overburdened by paperwork and the necessity to directly supervise 150 technicians. Consequently, they are unavailable to provide infield oversight. Control cf day-to-day activities within the RP organization is generally good because contractor personnel perform work under limited direction from an experienced permanent staff. There is no consistent indepth auditing by corporate or qualified site personnel to identify program weaknesse !
Station management has experienced difficulty in resolving technical problems in a thorough and timely manner. Weak radiological controls were established to support work in an onsite maintenance hot sho Shortcomings were observed in the radiological control of diving operations in the spent-fuel pool. The resolution of radiological deficiency reports often did not prevent recurrence and were not acted upon in a timely manner. Although management was aware of problems with the deficiency reporting system, a revised procedure that was available in June 1986 was not implemented until an NRC {
j inspection detailed the same problems in September 1986, j l
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l l- Certain program weaknesses are traceable to unclear procedures _j and policies. This has contributed to problems noted with j completion of routine surveys, alarming dosimeter use, air i sempler use and calibration, and confusion regarding fertile I female exposure controls. In an attempt to correct the I I
procedural inadequacies, the supervisors often issue informal l memorand There are no assurances, however, that key personnel )
have read or understand the requirements contained in these memorand Management involvement improved late in the period. Improvements in the radiological deficiency reporting system, routine survey program and a reorganized Station ALARA* Review Committee ;
resulted in new procedures and a well planned approac !
Specific goals, developed as part.of the Enhancement Program, '
include RP staffing increases, revision of position guides and all operating procedures, and increased corporate oversigh Reporting of these activities is very good, with frequent and detailed information provided to most managers onsite. The !
working relationship between the operations, maintenance, and radiation protection departments is poor, however, and often hostile. This negative environment resulted in an altercation between an RP technician and a maintenance worker during work in a contaminated high-radiation area. In addition, licensee RP staff and NRC inspectors perceive that operations personnel would not respond to health physics deficiencies. Management attention has recently been directed towards improving the RP 1riterface with other grcup Training programs for workers and RP technicians are very goo Lesson plans and facilities are generally good, and instructors appeared to be well prepared and enthusiastic. Minor weaknesses were found in that the continuing education program for technicians and the RP staff was poorly controlled and ineffective in providing current information on radiation safety matter The ALARA* program continues to be marginally effective, in part because of a lack of aggressive support by upper level site management and corporate management. Exposures have been consistently high. There are no long range goals or coordinated plans to improve performance. Short-range planning for outages is ineffectiv Inadequate planning information for the Spring 1987 Unit 2 outage prevented forecasting an ALARA goal for the outage. ALARA reviews of major plant modification work initiated at the corporate level are not eviden *As low as reasonably achievable, with reference to radiation exposure _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
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Radioactive Waste Mant.gement and Effluent Controls NRC review during the previous assessment period indicated concerns with the identification of radionuclides and the proper classification of waste shipments. Similar problems occurred during the current period. Specifically, technica'l data qenerated by a contractor that was to be used in waste classification was mislaid; this caused an apparently misclassified radwaste shipment. Subsequent review by the licensee located the missing documents, end the violation charge was withdrawn. Problems with the licensee's control of vendor-analyses and other technical data needed to support radwaste classification were shown by this sequence of event In addition, the licensee failed to identify a radionuclides present in a dry active-waste shipment. Licensee corrective actions instituted during the previous period failed to recognize and resolve underlying contributing causes concerning poorly defined interfaces between licensee groups and inattention to technical aetai The review of the licensee's quality assurance program (as it related to the radioactive waste management program) indicated that management attention had been directed to providing ade-quate quality control of ongoing radwaste activities. The licensee's audit program, however, provided limited review of waste generator requirements because of limited scope and technical expertise A Unit 3 c andensate storage tank (CST) spill and unplanned release to the river occurred on February 16-17, 1986. Although the release did not exceed regulatory limits for maximum permissible concentration in the river, NRC identified problems with the evaluation of the radiation hazards of the CST spill and release, as well as inadequate procedures for minimizing the release to the river. Licensee corrective actions included the development of e new special-event procedure for radioactive i liquid spills and an analysis and determination of the l radioactive material release !
Environmental Monitoring Review of the licensee's environmental monitoring program indicated that the licensee was conducting an effective progra A large number of environmental media were sampled and an effective QC program was in place to assure the quality of
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sample analysis. The meteorological monitcring system was !
l properly calibrated and functioning. Audits of the program by f the licensee's Quality Assurance Division were thorough and audit-identified deficiency followups were adequately resolved by the licensee.
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l p Transportation In response to problems noted in NRC and Agreement State reviews of radwaste shipments during the previous assessment period, the licensee took generally effective corrective action, including ;
modification of pallet lifting slings, revision of special procedures for control rod blade shipments and increased quality control participation in radionuclides verification. Corrective actions to address radwaste misclassification concerns were ineffective, however, in preventing recurrence during this perio Audits performed by the QA group identified that 21 operating personnel failed to attend the 1985 annual retraining progra In response, the licensee provided special makeup training
. sessions in 1986. Similar findings in previous assessment
. periods suggest that increased licensee attention to training and retraining programs in the radiological controls area is neede Water Chemist ry Control A good water chemistry control program was provided, including an s ggressive program to monitor and control intergranular strens corrosion cracking (IGSCC). The licensee is adequately staffod and has state-of-the-art equipment for the nonrac'iological water chemistry measurement area. Technical innovations, including continuous crack growth-rate monitoring, upgraded / augmented sample analyses and preparation for routine hydrogen water chemistry analyses suggested a strong management commitment to providing state-of-the-art IGSCC control. Although occasional lapses in maintenance allowed resin ingress and severe chemical transients to occur, the licensee maintained low steady-state reactor water conductivities.during full power operation, indicating a commitment to controlling contaminant ingress and corrosion of pressure boundary and heat transfer surfaces. The licensee was responsive to NRC initiatives pro-viding an interlaboratory standards program and improving laboratory measurement control. The licensee has revised and 1 upgraded some chemistry control procedures and techniques, and continuing effort is being applied to the measurement control area at the sit Summary Increased upper management attention toward resolving weaknesses in the radiological controls program at Peach Bottom was apparent late in the period. Correcting the deeply rooted {
cultural problem, along with the timely implementation of the i broad-based Enhancement Program, will require aggressive onsite leadership that had not been demonstrated in the past. Progress is not certain, and detailed milestones need to be developed and
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closely monitored. In contrast, the separate chemistry control program exhibited consistently good performance in all areas reviewed. This suggests that wide variations in performance among departments can exis . Conclusion Rating: Category 2 3. Recommendations Licensee
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Increase supervisory staffing to provide oversight of infield activitie Take strong and aggressive ,tctions to improve working g relationships and effective communications between the '
radiation protection and other work group N_R_C, None i
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C. Maintenance (9.3%, 649 hours0.00751 days <br />0.18 hours <br />0.00107 weeks <br />2.469445e-4 months <br />) Analysis Maintenance activities were eviewed by the resident inspectors ,
during each routine inspection. Specialist inspections examined #
maintenance and related activities during selected reviews. A Region I diagnostic team inspection reviewed the overall maintenance program of the licensee. Specific maintenance activities examined included: administrative and maintenance procedures, preventive and corrective maintenance, snubber replacement, emergency service water (ESW) pipe repair, Intermediate range monitor (IRM) replacements, Limitorque motor operated valve (MOV) disassembly and repair, switch adjustments, lubrication and wiring replacement, spent fuel storage rack replacement, condensate storage tank (CST) dike and heat trace work, and electrical circuit breakers and transformer l For the previous assessment period, the rating was Category Management was strongly involved in large maintenance activities, personnel were well trained, work was adequately [
planned for large jobs, procedures were detailed, and '
maintenance workers followed procedures. It was also noted that poor work practices resulted in 13 unplanned reactor scram signals while shut down, Smaller maintenance tasks appeared to suffer in many cases from lack of planning and plant management attention. Based on problems observed during the previous assessment period, it was recommended that the licensee improve control of vendor informatio The onsite Maintenance Division is now adequately staffed with crafts and engineering support personnel. A significant asset is that mobile craft personnel are available to respond on short notice. A shift maintenance assistant foreman is on site at all times to respond to shift needs. This foreman attends the preshif t briefing and works well with the Shift Superintenden Three scrams were associated with maintenance during the current l assessment period. Placing a condensate filter demineralized in service caused operator-initiated manual scrams in September and October 1986 on Unit 3. The licensee had started to use a new ,
type of filter element in the demineralizers, and the filter l was suspected to be the cause of the problem. The root cause of '
the first scram was inadequate postmaintenance inspection and testing. Had the root cause been identified following the September event, the October event mioht not have occurre I There have been no similar problems sirce the second event.
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l Coordination of activities between maintenance and operations was adequat Coordination greatly improved during the latter part of the current assessment period. The licensee had instituted daily planning meetings and a midshift status and coordination meeting. Control room personnel were well informed of the status of work performed by maintenance technicians. The ;
daily planning meetings also resulted in closer relationships !
among maintenance, health physics, and QC personnel in understanding the total scope of planned work activities, as evidenced during the question and answer portion of the meeting Maintenance activities were extensive during the assessment period. Over 10,800 corrective and preventive maintenance tasks were done. A computerized program is used for history and maintenance planning (CHAMPS) and for tracking maintenance, equipment history, failure trends, and scheduling of resource A computer generated maintenance request form (MRF) has proved to be an efficient means of interfacing between plant staff, quality control, maintenance planning, and operations. Also, the licensee implemented the equipment trouble tag (ETT) system for material deficiency identification, tracking and correctio Another maintenance enhancement has been the establishment of the maintenance root cause failure analysis and coordination engineer (MRFACE) functio The concept of the MRFACE function is to find the root cause of maintenance failures and take corrective action through failure analysis and investigation to identify and correct the root cause of failures. The HRFACE concept has been operational for a short time, but early results show indepth and timely investigation During this assessment period, two cases of poor equipment maintenance were identified. The lubrication program for operating equipment had lapsed two years ago because operating personnel were not implementing the program, and there was inadequate oversight by the shift superintendents. The licensee reinitiated the program after discussions with the inspectors; however, a recent inspection shows that the program is still not being fully implemented. The maintenance of the emergency load center transformers was found to be poor in that the gas pressures in the transformers were not being maintained as recommended by the manufacturer. When informed, the licensee immediately began to maintain transformer gas pressure. These are examples of poor plant management oversigh _ - _ - .
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Craft training programs were accredited by INP0 during the assessment period. A dedicated training facility at Barbados
. Island (former fossil station) is equipped with actual plant equipment for hands-on training. Mockup training at the site includes an under vessel control rod drive setup and a Limitorque MOV. These mockups have proved useful in training and radiation exposure reductio Maintenance on the emergency diesel generators (DG) .is performed with recognition that the DG have to provide overall system reliability and minimizes outage times. Maintenance is conducted efficiently and continuously on a three-shift basis until completed. Although the availability of the DGs remains high, several concerns were raised. First, a number of defects in DG support equipment were noted by the PRA team that left the impression that the level of DG maintenance may be trending downward. Secondly, the licensee has been slow in updating the DG vendor manuals, as required by Generic Letter 83-28. These concerns signal the need for the licensee to review efforts to assure continued high availabi?ity of the DG The battery maintenance activities are being upgraded to incorporate provisions of IEEE Stanc'ard 450 and were acceptabl The PRA team identified weaknesses in the maintenance and
. testing of support systems. Several examples were found that affect both preventive and corrective maintenance, including lack of recorded maintenance and testing for the low battery voltage annunciation relays and the underwater visual inspection of the emergency cooling wcter sluice gates. None of these items have resulted in equipnent failure; however, improvements can be made to strengthen the maintenance practices and progra The licensee started a project to cograde maintenance procedures to provide more accurate and specifr: guidance. Procedures and the ETT system implementation improved equipment problem identification and tagging. This acti m improved maintenance response to unanticipated problem During this assessment period, maintenance tradesmen were found to be knowledgeable, and they showed a strong sense of pride regarding the quality of their wor The craft and engineering involvement in the ESW 1eak repair was performed in an efficient and professional manner. However, conflicts between crafts and radiation protection personnel were noted (see Radiological Controls,Section IV.B).
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Quality assurance of maintenance is good. QA audits of signi-ficant maintenance activities were found to be complete and QA findings were given appropriate management attention. By design of the maintenance request form system, QC is an integral part of the maintenance activity. QC is extensively involved in maintenance activities. Findings are tracked and trends are evaluate The coverage and data analysis provided by QC for maintenance activities is considered a strength; however, one potential weakness concerns delays by the maintenance department in resolving QC comments on work packages before they are closed ou The licensee is aware of these delays and is working toward more timely resolutio In summary, the licensee has a good overall. maintenance progra Equipment is maintained by an adequate staff and procedure Maintenance problems are well-documente There have been some weaknesses in the maintenance area; however, the licensee's response has been adequate to resolve these problem . Conclusion Rating: Category 2 Trend: Improving 3. Recommendations None
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. Surveillance (10.2%, 715 hours0.00828 days <br />0.199 hours <br />0.00118 weeks <br />2.720575e-4 months <br />) Analysis In the current assessment period,. region-based inspectors reviewed the results of surveillance tests (ST) applicable to operating radiological controls, fire protection, maintenance, emergency preparedness, water chemistry control and environmental monitoring. Programmatic reviews of the ST program were conducted during the diagnostic team inspection and a resident inspection. Resident inspectors routinely reviewed i selected surveillance areas and observed testing during each mont In the previous assessment period, surveillance was rated Category The following problems regarding surveillance test activities were noted: missed tests, scram signals caused during instrumentation and controls (I&C) testing, and poor management controls to ensure ontime test completion. Also, the surveillance program did not address actions to be taken when a test became overdu Three scrams (one at power and two while shut down) occurred !
L during the current assessment period as a result of surveillance testing. An automatic scram on Unit 3 daring substation relay testing was caused by an instrument technician error. Two auto-L matic scrams on Unit 3 during prestartup surveillance testing were caused when a licensed operator failed to follow the test procedures. The licensee disciplined the persons involved in these personnel errors and modified the EHC syste In the previous assessment period, NRC noted weaknesses in the surveillance program with respect.to plant management controls and oversignt to ensure that STs are performed when scheduled and that compensatory actions are initiated when a test becomes overdue. Before the licensee responded to these concerns, the j diagnostic team inspection noted several overdue tests on both !
units. The overdue STs involved testing the operability of the reactor protection system and emergency core cooling system When informed of these overdue tests, the licensee performed the tests, and the results were satisfactor Based on these concerns and problems, the licensee revised the surveillance program. The revised program increases plant management oversight and involvemen Subsequent reviews by NRC indicated that the surveillance testing program is functioning '
adequately to ensure that required tests ire performed in a timely manner. Administrative controls were instituted to improve plant management awareness of tests that
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could become overdue. Before any test becomes overdue, plant management and PORC review are required, including corrective actions and/or compensatory measures. The Peach Bottom Station Enhancement Program includes actions to increase plant manage-ment involvement and oversight and.QA/QC involvement. These actions are complet QA/QC involvement in surveillance testing improved during the period. QC personnel use detailed monitoring checklists to assure that STs are performed when scheduled. They also perform independent verification reviews of instrument STs. QA audits of the ST activity are timely and recommendations are communicated to managemen The performance of STs, use of test procedures, shift oversight and test control, and test result reviews are good. Routine surveillance witnessing and observation by regional and resident inspectors indicates that the test program is functioning adequatel Surveillance records, including test results and documentation, were generally easily recoverable and readable. One exception was the availability of as-found data for safety relief valve (SRV) setpoints. The licensee initiated corrective actions to revise SRV surveillance to ensure as-found data compilatio In summary, surveillance testing has been successful in con-firming that safety-related equipment can perform as require Equipment problems have been uncovered during test performanc Surveillance procedures, test conduct, and results reviews are good, l.icensee actions to improve plant and line management oversight and control of surveillance activities have increased the assurance of timely test performanc . Conclusion Rating: Category 2 3. Recommendations None l
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E. Fire Protection and Housekeeping (4.0% 281 hours0.00325 days <br />0.0781 hours <br />4.646164e-4 weeks <br />1.069205e-4 months <br />)
1, Analysis During the previous assessment period, the licensee continued to perform fire protection modifications. Two major fires occurred during that period. The cause of the radwaste building fire in a non-safety related cable tray and divers' cage was unknow Improvements were noted in housekeeping and overall plant appearanc In the current assessment period, fire protection was reviewed during the 10 CFR 50,48 and Appendix R team inspection, and fire protection and housekeeping activities were reviewed during each resident inspection. Also, a regional specialist reviewed licensee fire compensatory measures, fire watches, and a Unit 3 fire that occurred on March 4, 198 The licensee conducted a self-initiated review to confirm compliance with Appendix R fire protection requirements. An NRC Confirmatory Action Letter (CAL 86-07) was issued to specify that the Appendix R review was to be completed by September 30, 1986. Any violations were to be reported to NR The licensee's review identified a number of conditions at Units 2 and 3 that were in violation of the fire protection requirements of 10 CFR 50.48 and Appendix A management meeting was held subsequently to discuss the nature of the violations, causes for the non-conforming conditions, a des-cription and status of corrective actions and schedule for completion, and interim compensatory measures. The interim ;
compensatory measures, which consisted primarily of increased fire watch patrols, upgraded training of fire watch personnel and improved administrative controls over combustibles, were reviewed by a regional specialist. A licensee Justification for Continued Operation (JCO) while the conditions of nonconformance are being permanently corrected, with the compensatory actions in place, was also reviewed, along with the TRIP-300 series procedures for fire protection developed by the licensee. Based on this eva'uation, NRC concluded that, pending completion of permanent corrective actions, the compensatory measures taken and the .1C0 prepared in response to the identified violations provide adequate assurance that both units have the ability to achieve safe shutdown following a fire. The violations are currently under review by NRC. An enforcement conference was held after the assessment period.
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A'special NRC Appendix R inspection was conducted on April 9, 1987, to review JD CFR 50, Appendix R nonconformances identified by the licensee. Five of the approximately 50 nonconformances were reviewed. It was verified that all five were violations i involving Section III.G.1 and III.G 2 of Ac;.endix Early in the assessment period, the' licensee was neither aggressive nor methodical in pursuing resolutions to Appendix R issues. Lack of engineering management attention in this area resulted in the selection of an individual to lead the Appendix R effort who was not given authority to direct all aspects of the projec As a result, a final. comprehensive safe shutdown analysis was delayed. An analysis of this type is required to provide assurance of system independenc The failure to meet scheduler requirements and complete the fire )
hazards analysis in a timely manner indicates poor engineering !
management attention to a major NRC initiative. The licensee ,
has not devoted adequate resources in the past to the resolution '
of Appendix R issues. The licensee appears to now realize this and has made a recommitment to implement the rule. This is evidenced by the completion of the reanalysis and the establishment of schedules for completion of modification During the current assessment period, a major fire occurre The fire was in a inaintenance cage in the Unit 3 turbine building at the 195-foot level on March 4, 1987. A review of the area and of the fire hazards analysis (FHA) determined that no safe-shutdown equipment was in the area, but an acetylene welder's cart was in the are Neither the FHA nor the prefire strategy plan procedures documented the presence of this car An off-site fire and explosion occurred at the North Substation (about one mile from the protected area) on April 13, 1986. The cause of the fire and explosion was the failure of the No. I transformer phase A because of an apparent internal fault. The command and control of the overall fire brigade and off-site fire company response were goo Overall plant housekeeping and appearance improved markedl The licensee's detailed plant cleanup program includes painting, decontamination, and general area cleanup. Many additional plant areas are now available for general access. These actions have improved the work environment and need to be continue In summary, lack of engineering management attention early in the assessment period resulted in poor implementation of fire protection requirement Increased attention has, however, resulted in completion of the fire protection report, identi-fication of Appendix R noncompliance, and actions to resolve these problem >
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. Conclusion Pating: Category 3
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~ Trend; Improving Recommendations Licensee Assure prompt completion of fire protection modification NRC Perform 10 CFR 50, Appendix R reinspectio I i
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. Emergency Preparedness (2.2%, 225 hours0.0026 days <br />0.0625 hours <br />3.720238e-4 weeks <br />8.56125e-5 months <br />) Analysis I During the current assessment period, one full participation exercise was observed and one routi% safety inspection was conducted. The routine inspection specifically related to follosup of licensee Cor,firmatory Action Letter (CAL) commit-ments and the miniraum staffing plan for onsite emergency response. Emergency preparedness was also reviewed during selected resident inspe:tion During the previous assessment period, licensee performance in this area was rated Category 2. Weaknesses were identified in the emergency preparedness program and the emergency exercis A CAL was issued that addressed deficiencies in the ifcensee's emergency plan training and organizatio The licensee's response to NRC initiatives, as outlined in tne CAL, were reviewed. The licensee's actions were found to be satisfactory and progress was made in most areas, with the exception of the onsite minimum staffing level. The licensee then added an onshift junior technical assistant to perform initial dose assessments. A drill was held in September 1986 to evaluate the initial emergency response organization using a larger staf In the full participation exercise in October 1986, the licensee demonstrated satisfactory response capabilitv. Actions taken by emtrgency response personnel were prompt, event classifications were accurate, and protective action recommendations to off-site authorities were timel Site training and staffing concerns were also identified during the full-scale emergency exercise held on October 9, 1986. A full complement of qualified, licensed operations personnel was not available throughout the afternoon and evening exercis Other training weaknesses were observed in implementation of the new dose assessment computer model. The NRC team determined that interpretation of results and information flow exhibited by dose assessment personnel needs improvement to ensure that appropriate protective action recommendations are mad Although an adequate level of corporate support exists to
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administer basic emergency preparedness program functions, the l
I implementation by on site management requires improvements. This was evidenced by inefficient use of key emergency staff within the Emergency Operations Facility (EOF) and Technica; Support ( Center (TSC) during the exercise. Other deficiencies were I identified in Emergency Plan Implementing Procedures that l
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had received review and approval from site management. The licensee has reorganized the corporate onsite support for emergency preparedness. A new position of emergency response facilities coordinator now has responsibility for the E0F, TSC, Operations Support Center (OSC), .and auxiliary OS The emergency preparedness program was improved somewhat by the efforts of'the corporate emergency preparedness staff. Further )
improvements are needed in the cose assessment function. The full-scale exercise was held without major deficiencie . _Concl usion Rating: Category 2 3. Recommendations Licensee Conduct exercise prior to restart due to major changes in plant staf NRC None
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L G. Security and Safeguards (2.0%,144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br />)
l Analysis A region-based inspector conducted one special physical protection inspection and three ioutine, unannounced physical protection inspections. Routine resident inspections were performed throughout the assessment period. In addition, security was reviewed during the diagnostic team inspectio During this assessment period, the licensee and the security-force contractor aggressively pursued a planned course of action to identify and correct the root cause of previously identified poor performance. To improve the overall performance of the security organization and the security program in general, the licensee developed and implemented several significant change Senior corporate officials affirmed their support for and intent to implement an effective security program. As a result, a reorganization of corporate responsibilities was initiate The position of Manager of Nuclear Services was created, and the incumbent was given the responsibility to establish an organization with the sole responsibility for nuclear securit That organization was established and is headed by the Director of Nuclear Security. The role of the Director is well defined and includes responsibilities for the management and oversight of the nuclear security programs at both of the licensee's operating nuclear stations. A technical analyst was assigned to assist the Directo In conjunction with these corporate changes, the licensee created eight security supervisory positions at the facility to provide 24-hour oversight of the contract security force. The responsibilities of these supervisors include assuring that the contractor properly implements the licensee's security program and that the security force performs at a high level. At the end of this assessment period, however, only three of these positions bad been fille In addition to the eight licensee supervisors, the licensee's senior onsite security representative, the Nuclear Security Specialist, was also assigned a technical assistan This technical assistant is responsible for monitoring key aspects of the security program on a day-to-day basi The development and implementation of this expanded licensee oversight organization, along with the corporate organizational changes and the new security computei discussed below, provide evidence that the licensee is attempting to implement a sound security program that goes beyond minimum compliance with NRC requirement _ _ __-_ __ -_____ - __-_
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These changes' are very recent, however, and their impact on the program has not yet been assessed by NRC. Additionally, a problem with the security contractors management of personnel resources during the current shut down period, which was broughts to NRC's attention through allegations at the end of the assessment period, indicated that these changes have not yet been effective. This is discussed further later in this assessmen Significant capital resources were expended to construct a new security building onsite and to purchase a new computer. The building provides efficiently designed work areas for licensee and contractor security management and for staff records and equipment. The building will also house a new state-of-the-art, integrated security computer. The new. computer should improve the reliability and effectiveness of the existing security systems, equipment, and associated hardware. It is scheduled for installation in August 1987. These actions are further evidence of the licensee's intent to correct prior program deficiencie The licensee's training program is carried out by individuals who are' experienced and assigned to security training on a full-time basis. Training facilities have adequate classroom space and instructional aids are utilize Lesson plans are well-developed, generally thorough, and kept current through feedback from supervisors. The licensee's expanded staff should enhance the training program by providing additional and meaningful feedback. Security procedures and instructions were being re rised, as necessary, to be more clear ar.d concis In an effort to ensure that the security program is effective, the licensee is conducting surveillance of the performance of the security forc Experienced and knowledgeable personnel perform these surveillance, and the findings are aggressively pursued to ensure prompt and effective corrective action and feedback to the training program. These surveillance are in addition to the annual security program audit required by NR Security management is actively involved in industry initiatives dealing with nuclear security programs. There is evidence of support for the security program at a high level in the licen-see's organizatio The licensee maintains the security-related facilities in a professional manner. The main guard house, security office
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L building, and other areas are presentable and clean. The Central and Secondary Alarm Stations are clean and well-organize The licensee submitted three security event reports pursuant to 1 10 CFR 73.71 during the current assessment period. One report concerned the failure of the security computer. The cause of
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the failure was determined to be a hardware malfunction that was
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promptly corrected. The second report concerned a potential tampering incident involving a motor control center. The licensee conducted an extensive investigation but was unable to reach a definitive conclusion regarding the matte During the latter part of the assessment period, the licensee submitted the third security event report, pursuant to 10 CFR j 73.71, concerning a security watch person found asleep on May 29, 1987 at a Unit 3 vital area access control poin Prompt licensee action, in accordance with the Contingency Plan, was implemented. In each event, the licensee's securit compensatory measures were timely and appropriate. In the-potential tampering event, however, the licensee failed to integrate the broader operational considerations associated with tamper.ing into their response and compensatory measures until prompted to do so by NRC management, as discussed in Section J,
" Assurance of Quality". The reports submitted to NRC were clear, concise, thorough, and prompt; they indicated appropriate security management revie The one reportable incident of equipment malfunction during this period is noteworthy in that it provides continuing evidence of increased licensee attention to preventive maintenance and surveillance testing of security-related systems and equipmen An event concerning a watch person found asleep on post is, however, of concern. Allegations of security personnel being forced to work excessive hours and under trying conditions were received by NRC at about the same tim NRC review (just after this assessment period) found that security force personnel were working long hours (in excess of 12, at times), were receiving late meal breaks, or no.ne at all, and were required to remain on post for extended periods without rotation. The apparent causes of the unfavorable working conditions were poor scheduling on the part of the security force contractor and the use of reserve security personnel by the contractor for other tasks (fire watches). It appeared that, at least since the beginning of the current shutdown period, that the licensee had failed to exercise appropriate oversight of its security contractors and, once again, had not adequately prepared for an outage. As a result of NRC findings and those of the plant manager, the licensee has made a commitment to take actions to correct the problem. The effectiveness of those actions will be closely monitored during the forthcoming assessment perio _ _ _ _ _ _ _ _ _ _ _ - _
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During'the assessment period, the licensee submitted one change to the Training and Qualification Plan and one change to the Security Plan under the provisions of 10 CFR 50.54(p). Also, the licensee responded to the Miscellaneous Amendment to 10 CFR 73.55, codified by NRC on August 4, 1986, that required changes to be submitted to NRR to define how the miscellaneous amend-ments to 10 CFR 73.55 would be met. The program is the responsibility of the licensee's corporate nuclear security staff. The staff is very effective ir carrying out this respon-sibility and ensures that plans are current and changes are properly coordinated when required. They often communicate and review plan changes with regional licensing personnel to ensure a clear understanding of the changes. When the plan changes are submitted to NRC, they are of good quality and indicative of thorough knowledge and understanding of NRC security objective In summary, the licensee has pursued several security program improvements during this assessment period. Increases in licensee personnel for program oversight and direction were implemented, and management involvement in and support of the program were evident. The increased emphasis on the security program demonstrates the licensee's desire to implement a high-ouality program with a well-qualified and professional forc There remains, however, the long-standing problem of the licensee's approach to contractor oversight.as indicated by the licensee's lack of awareness of and inattention to recent security program staffing requirements. Until the contractor oversight problem is resolved, it is doubtful that the licensee will meet its goa . Conclusion Rating: Category 2 3. Recommendations None
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.. Technical Support Activities (N/A) l Analysis During the current assessment' period, technical support is being considered a separate fcnctional area for the first time. The approach to the resolution of technical issues from a safety-standpoint continues to be an evaluation criterion for each functional are Technical support includes onsite engineering support from the plant staff, offsite engineering support from the corporate Engineering and Research Deuartment, onsite technical support from the maintenance en'.jineering group, and offsite support from the corporate Nuclear Support Departmen Corporate engineering and design support from the Engintaring and Research Department was noted previously to be, historically ,
stron In response tv siltation and corrosion problems'in the !
emergency service water' (ESW) system piping, a number of modifi- !
catiora were made and surveillance programs instituted. Piping ,
was replaced or cleaned, a chemical addition program was begun,' I periodic flow testing was performed, and periodic visual inspec-tions were conducted. Strong management interest and engineering support of these programs provide confidence that future ESW system corrosion or flow problems will be detected and corrected in a timely manne Corporate engineering support was also good in relation to the following problems that occurred during the assessment period *
(1) Limitcrque motor operated valve environmental qualificati 1 i internal wiring assessments and inspections of both units; (2)
spent fuel pool storage rack replacements on both unit Exceptions to strong Engineering and Research Departme'.t support j were lack of oversight in the fire protection area an.i lack of a current emergency AC or DC lead growth lis . Lack of engineering management attention resulted in a 10 CFR 50, Appendix R program that was neither aggressive nor methodica Major modifications are perforned by an er.-site Construction Division staff consisting of permanent stat. ion personnel supplemented with craft personnel to support outage activitie The group, which is a part of the Engineering and Research Department, is experienced in installing plant modification l The group takes advantage of extensive advanced planning and utilizes the concept of a Construction Job Memorandum to
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summarize work scope for field personnel. The group has been l
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successful in coordinating among the licensee's matrixed organizations with minimal fmpact on plant operations, i
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t l Walkdowns of systems, and effective communications among work groups, including participation in daily planning and coordina-tion meetings, have served to accomplish this goa A field engineering group is also available for electrical design and modifications. This group has consistently provided expertise to solve reactor safety problems. An example was the effort to identify the root cause of problems with the RPS power supply breakers and initiation of corrective action The onsite Independent Safety Engineering Group (ISEG) reports to the offsite Nuclear Safety Sectio ISEG provides technical support primarily in the area of event followup (Event Reports).
Event Reports are accurate and complete, and provide for a root cause determination. ISEG also provides technical support for the newly implemented human performance evaluation system (HPES).
Good corporate support exists to administer the emergency l preparedness program functions. The support is demonstrated through dose assessment computer models, emergency plan updates, and offsite emergency planning coordinatio The site technical engineering group provides technical support for the plant. System engineers assigned to this group are responsible for specific plant systems and subsystems. These engineers continue to be a valuable source of engineering knowledge in the operation, testing, and troubleshooting of plant systems. The licensee has enhanced the systen engineer concept according to INPO guidelines. This should aid in the determination of root causes of failures. Also, a plant performance monitoring program has bet:n initiate These activities have resulted in overall station performance improvements. One instance was noted, however, in which a scram was caused by poor troubleshooting activities by the onsite engineer During the assessment period, improvements were noted in the area of corporate security support for the Peach Bottom Statio A reorganization of corporate responsibilities occurred, and a new corporate manager was assigned to assist in the implementa-tion and oversight of security programs. In addit' n, a technical analyst was assigned to assist this new corporate security manage Apart from routine licensing activities, the licensee has been very supportive of NRC and EPRI efforts to resolve generic safety issues or to develop information to improve the safety of plant operations. Peach Bottom Unit 2 has been the reference boiling water reactor (BWR-4) with a Mark I containment system in several NRC sponsored probabilistic risk assessment (PRA)
studies dating back to the Reactor Safety Study, WASH-140 V w _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ - - - _
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During the past three years, the licensee devoted considerable resources to supporting NRC's recently published Reactor Risk Reference Document, NUREG-1150 (draft for comment). The 1984 Industry Degraded Core Rulemaking (IDCOR) Technical Summary Task 21 report and the 1986 Individual Plant Evaluation (IPE)
report were further PRA efforts for Peach Bottom. During the past year, the licensee has also supported NRC research efforts-for its Containment Safety Study and development of a Contain-ment Event Tree related to the capability of the Mark I contain-ment system to withstand severe accidents. As noted in NRC's I letter of September 19, 1986, the degree of cooperation and technical quality of the analyses and evaluations provided by the licensee were major factors in being able to complete a feasibility study of the effectiveness of venting BWR Mark I containment systems (NUREG/CR-4696, " Containment Venting Analy-sis for the Peach Bottom Nuclear Power Plant"). Both the licensee's engineering and plant staffs provided considerable operating data, cost estimates and design evaluations to support NRC's PRA program and efforts to resolve generic issue No. 105 on interfacing system loss-of-coolant accidents (LOCAs). Other NRC research programs that the licensee supported during the past year included the " Effects of Iodine Chemistry on Source Term Estimates", development of a PRA-based Interactive System for Peach Bottom, development of improved acoustic monitoring techniques, online water chemistry analysis methods for many impurities that were not previously measured, and online crack growth measurement system The initiative displayed in undertaking research and development programs, studies, and followup activities in working with the NRC staff, as well as
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the quality and timeliness of the effort, evidences management 1 sensitivity to and involvement in resolution of safety concern In summary, licensee management is strongly oriented toward technical support and engineering, and has integrally contained, within all disciplines, effective engineering support, in addition to the historically strong corporate design engineering 1 function. Notable exceptions early in the assessment period l were the oversight in the fire protection and site electrical load growth aran Engineering activities during the assessment period, were e malated, particularly in the second half of the period, as exteraivo planning for the refueling outages in 1987 was underwa . Conclusion Rating: Category 2 Trend: Improving 3. Recommendations None
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I. Training and Qualification Effectiveness (N/A) Analysis Technical training and qualification effectiveness, while being considered a separate functional area, continues to be an
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evaluation criter in for each functional area. This functional area was considered and discussed as an integral' part of othe functional areas and the respective inspection hours were included in each one. Consequently, this discussion is a synopsis 'of the assessments related to training conducted in other areas. Technical training effectiveness was measured primarily by the observed performance of personnel and, to a lesser degree, as a review of program adequacy. The discussion below addresses three prircipal areas: licensed operator training, nonlicensed staff training, and status of training accreditation by the Institute of Nuclear Power Operations (INPO).
During the current assessment period, resident and specialist inspectors routinely reviewed training. Three operator licensing exams were given by region-based examiners. Training was reviewed during programmatic reviews of operat. ions (licensed, nonlicensed, and requalification), maintenance, radiation protection, general employee training (GET), general respiratory training (GRT), fire protection, emergency preparedness, and securit Circumstances leading to the March 31, 1987 Shutdown Order indicated a lack of supervisory and leadership training which will be discussed in the " Assurance of Quality" functional area (Section J).
In May 1985 the licensee obtained INPD accreditation for the following five programs: Nonlicensed Operators, Reactor Operators, Senior Reactor Operators, Chemistry Technicians, and Radiological Protection Technicians. Self-evaluation reports for five additional training programs (Shift Technical Advisor, I&C Technician, Electrical Maintenance, Mechanical Maintenance, and Technical Staff and Managers) were submitted in November 1985. These five programs were accredited in October 1986. It is too early to see any operational performance chtnges resulting from INP0 accreditation. The accreditation process has, however, had some positive observable impacts on the training program, including improvements in the formulation of training goals and objectives; better, more detailed L Qualification Manuals; and improved overall planning and ( documentation of training activitie _ _ _ _ - _ _ _ _ - _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ - _ - -
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The licensed operator training and requalification training programs function wall, as evidenced by NRC_ exam performanc During the assessment period, 93% of those taking the NRC exams achieved passing grades. Scyen of eight of the reactor operator ,
candidates, 13 of 14 senior reactor operator candidates, six of !
six fuel-handling SR0s, and three of three instructors passed the NRC exams during this assessment perio Both R0 and SR0 candidates were familiar with plant equipment location and with reference material in the Control Room. It was noted during one oral exam that SRO candicates were hesitant to implement the Emergency Plan when presented with various significant transient !
Parallel gi ding' was conoucted on 20% of the exams given for requalifics on in 1986, and good correlation was netod between the facili:, and the NRC grading process. The facility is moving forward with its simulator plans and has currently planned to start using the simulator for training in 198 The Peach Bottom training muterial used for exam preparation has {
significantly improved since the Operator License Training 1 Programs were accredited by INP One problem was identified during the inspection of the condensate storage tank (CST) spill of February 198 Information and training on the CST heat trace system was lacking. This resulted in operators checking the system operability without adequate knowledge of how the heat trace system works. The operators mistakenly signed off that the systen was operable when it was inoperabl General employee training (GET) and general respiratory training (GRT) were reviewed both during resident and specialist inspections. The GET and GRT programs are adequate. The licensee has expanded GET and added new " Nuclear Professionalism
- Job Orientation" training to that progra The security training program is conducted by experienced full-time instructors. Training facilities and instructional aids ;
are adequat Lesson plans are well-developed, generally thorough, and kept current through feedback from supervisor Improved security performance has been realized through training improvement During the full-scale emergency exercise held on October 9, 1986, a weakness in site training was identified. The imple-mentation of the new dose assessment computer model was not I
ef fective in that judgement in interpretation of results and in l
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information flow exhibited by dose assessment personnel was weak.
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Two scrams were caused by personnel errors resulting from ' lack-of knowledge. A test engineer lifted a lead during troubleshooting activities and caused an automatic scram at low power. A resin injection resulted in a manual scra The Region I diagnostic and specialist team inspections previded in-depth reviews of the maintenance training program. The inspectors visited offsite training facilities, reviewed training materials, and discussed the program with licensee personnel. The maintenance training program is considered to be well-developed and implemented in an effective manne In summary, the licensee has achieved accreditation of all training programs. Licensed operators continue to perform well on NRC license exams and the Peach Bottom simulator is' currently being procured with delivery expected next year. However, the extent to which apparent weaknesses in the training of supervisors played a part in the performance that led to the NRC shutdown Order of May 31, 1987 had not been completed by the end of the assessment perio . Conclusion Rating: No rating was deemed appropriate by the Board because the extent to which apparent weaknesses in supervisor training contributed to the inattentive control room behavior leading to the March 31, 1987 NRC shutdown Order had not been completely assessed at the end of the perio . Recommendations See Assurance of Quality
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l Assurance of Quality (N/A)
1 Analysis Management involvement and control in assuring quality continues to be an evaluation criterion for each 'unctional area. During this assessment period, however, assurance of quality is also being considered as a separate functional are The various aspects of the programs for assuring quality were considered and discussed as integral part of each functional area, and the respective inspection hours are included in eac Consequently, this discussion is a synopsis of the assessments relating to the quality of work conducted in other area In the previous assessment, the focus of established onsite and offsite committees and of plant and corporate management did not appear to be the resolution of operational problems and the assurance of. operational quality. Multiple breakdowns in the
- defense indepth concept occurred in the area of control room
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operator demeanor oversight Neither plant nor corporate management, nor first- or second-line supervision, nor the site QA organization addressed the control room operator inattentiveness issue. Once identified to plant management, either no corrective action or inadequate corrective actions were taken. This inability to ensure formal control room demeanor led to the conclusion that the licensee could not ;
operate the facility safel Areas reviewed during the current assessment period were: Plant Operations Review Committee (PORC), Nuclear Review Board (NRB),
Independent Safety Engineering Group (ISEG), Operating Experience Assessment Committee (OEAC), Quality Assurance (QA),
and Quality Control (QC).
The control and oversight of the surveillance testing (ST) l program by plant management was a significant weakness. The improper focus of ST performance accountability and oversight resulted in numerous missed tests of safety systems. QA audits of the ST program did not note whether or not tests were performed on time. The licensee has revised the ST program controls and oversight, and QA and QC have become involved in ST program review _ _ - - _ _ - _ _ __ _
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In addition, plant management did not integrate broader operational considerations into the security response and compensatory measures in their response to a potential tamperi event. That is, when it was suspected that one means of shut-ting down the reactor had been tampered with, the licensee did not. inspect to assure the operability of the other systems until prompted to do so by NRC managemen The failure, of plant management to fully integrate all functional aspects of the response to such an event is indicative of a lack of involvement in assuring high quality operation The onsite review committee (PORC) and the offsite review committee (NRB) continue to adrainistratively function well, as demonstrated by a questioning attitude with regard to safety issues. Improvements were noted with respect to PORC and NRB involvement in post scram reviews, root cause determinations, personnel error reductions, and improved procedural adherenc The Peach Bottom Enhancement Program addresses these issues and area The ISEG and OEAC groups continue to provide strong te:hnical suppor ISEG event reports are timely, accurate and exhibit good root cause evaluation. ISEG is adequately staffed and per-forms independent daily control room walkthroughs. In addition, ISEG is involved in the overall scram reduction program and in implementing the INPO sponsored Human Performance Evaluation System (HPES). HPES is designed to identify and correct situations that cause or could have caused human errors. The OEAC is a technical review group established to orovide a broad interdisciplinary review of significant internal and external operating experience information pertinent to the safe operation of the Peach Bottom and Limerick Stations. The OEAC provides recommendations to management on improving operations, procedures, training and maintenance by mesns of verbal and written reports or meeting minute A second scram initiated by resin injection on Unit 3 was caused by inadequate corrective actions from a previous identi-cal occurrence. Had adequate actions been taken, this scram could have been prevente Weaknesses were noted in the depth of QA audits of the health physics and surveillance programs. Audits meet program requirements; however, the scope and findings of the audits a're oriented toward compliance with administrative requirement Improvements in QA audits were noted during the latter portion of the assessment period.
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4 In summary, the declining trend in performance noted in the previous SALP report was not arrested and unacceptable performance resulted during this assessment period. The problems noted, which were operator to.nplacency and procedural compliance, continued and, in part, led to the shutdown Order; however, the central reason for unacceptable performance in this area was that plant management was unable or unwilling to correct known deficiencies in op rator conduct that had potentially significant safety consequence . Conclusion Rating: Performance was determined to be unacceptable as reflected in the shutdown Order, and therefore no SALP rating is appropriat . Recommendations Specific recommendations were not made by the Board in light of the unigee circumstances of the shutdown Order which will necessitate an extensive review by the licensee to clearly identify the root causes that led to the order and demonstrate to the NRC that they have been corrected prior to any decision on restart of the plants.
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K. LicensingActivities(N/A) An .. ' y si s Euring the previous assessment period, the staff's assessment 5:a 3 that there had been a declining trend in licensing management involvement in and control of quality of licensing submittals. The trend manifested itself in a noticeable decline in the timely response to and resolution of technical issues and the need to give more attention to the no-significant-hazards consideration (NSHC) determinations for license amendment applications. The previous report noted that there was strong management involvement and attention, as well as engineering support for those issues having the greatest potential safety significance or the greatest potential for impact on plant operation (for example, the 1985 Unit 3 refueling outage, and the pipe inspection prcgram). On these latter issues, the licensee demonstrated capability for good planning, close cooperation and technical support by the Engineering and Research Department, thorough safety evaluations to support proposed actions, and high quality responses to NRC staff questions. The staff did not disagree with the licensee's priorities, but noted the disparity in attention being given to the priority issues vs NRC-initiated issues (for example, generic issues, NUREG-0737 items, Generic Letter 83-28, etc.).
The licensee responded to the previous assessment by letter dated August 12, 1986. Proposed corrective actions were 1) to develop a management system for tra: king the preparation of Technical Specification applications and 2) to improve the quality of significant hazards consideration determination During the current assessment period approximately 24 multiplant actions, 15 plant specific actions, including five Appendix R exemptions related to fire protection, and five amendments to the operating licenses and technical specifications were considered. There was substantial evidence of licensing management involvement and control in assuring quality, especially for the more prominent issues that affected plant operations, had high safety significance, or were required to be completed on a specified schedul Three examples are the Unit 2 reload application, the program to reduce the potential for intergranular stress corrosion cracking, and the standby liquid control system technical specifications in response to the ATWS rule, 10 CFR 50.62.
i
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There also is significant evidence that increased involvement by management in the Nuclear Support Department during this assessment period resulted in improvement in the quality af the licensee's NSHC determinations. Near the end of the period the staff utilized several of tne licensee's NSHC determinations without changes in the Federal Register notice of staff consideration of a license amendment application. In addition to the routine licensing activities specifically associated with the facility, the licensee provided substantial assistance to NRC efforts or, the resolution of generic safety issues. These efforts on probabilistic risk assessment, containment safety, interfacing system LOCA effects and other issues are aiscussed in further detail in the Technical Support section of this repor The licensee's cctions on these issues demonstrated a sensitivity to and involvement in the resolution of safety concern The licensee's responsiveness to NRC initiatives related directly to the plant has been mixe The licensee's response to those issues governed by NRC rule, such as the 10 CFR 50.62 ATWS review, or those issues governed by a Confirmatory Action Letter, such as fire protection, is generally timely, technically thorough, and well-supported by the licensee's licensing and engineering staf The licensee's responsiveness to generic letter and multiplant issues not included in the above category has in some cases been much less aggressive. Examples are the extended time required to get the diesel generator vendor manual update program going in l response to Generic Letter 83-28, the extended response time to l the staff's April 28, 1987 ietter on drywell vacuum breaker and the lack of a response to the staff's September 9,1986 letter on the procedures generation package. Although the performance in this regurd was not aggressive through much of the assessment period, the licensee was more responsive in the last several months of the period in supporting oral communications on about eight of these issues, such as NUREG 0737 Item II.E.4.2(7),
Generic Letter 83-28 Items 1.2 and 3.1.1/3.1.2, and overtime limits. Thus, a sustained higher level of performance in this area remains to be confirmed in the next assessment period.
l Considering the current licensing staff's resources and its broad-based responsibilities, many of which require substantial time onsite, the level of resources available to the licensee's Licensing Section for the plant appeared to be adequate to meet the majority of the licensing needs during the assessment perio There were, as noted above, a number of issues that were not pursued as aggressively as other issues. The Licensing Section received additional resources late in the assessment
_ _ ___
o L ..
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period, however, and there has been a recent increase in the level of attention given to these issues. Thus the effectiveness of these measures in ensuring that licensing i
'
issues are expedited remains to be confirmed in the next assessmen In summary, the technical quality of the licensee's' initial
'
submittals is generally adequate, the quality of the NSHC determinations has improved in the latter portion of the period, and the licensee has provided support to NRC efforts to resolve issues of generic safety significance. A slightly improving i trend in several areas remains to be confirmed .in the next assessment perio . Conclusion Rating: Category 2 Recommendations None I
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e
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V. SUPPORTING DATA AND SUMMARIES Investigations and Allegations Review Investigations i
!
The NRC Office of Investigations (01) completed an investigation (I-85-019) on May 6, 1986 in response to an allegation I concerning an employee who was threatened with dismissal if he i continued to pursue radiological concerns regarding a possible :
overexposure in March 1985. The OI report synopsis was issued 1 May 20, 1986, and a redacted version of the entire OI report was I issued September 22, 198 Enforcement conferences were held on I May 30 and November 18, 1986. A civil penalty was issued on February 8, 198 The NRC Office of Investigations is currently conducting a special investigation in response to an allegation concerning licensed operators being asleep or otherwise inattentive to their duties while on watch in the control room. The allegation was received on March 24, 198 . Allegations
! Ten allegations were received and reviewed by NRC Region I l during the assessment period. Five allegations were not substantiated, two allegation were substantiated, and three allegations are currently under review by Region I. One of the substantiated allegations concerns maintenance personnel being abusive to health physics technicians and an ensuing altercation that occurred in a radiologically controlled are The other substantiated allegation concerns the inattentive licensed control room operators, as stated abov Escalated Enforcement Actions Civil Penalties A Civil Penalty of $200,000.00 was associated with NRC Inspection 278/86-09 conducted March 18 - 21, 1986. The violations were concerned with an out-of-sequence control rod during a Unit 3 startu (Enforcement Action #86-59, dated June 9,1986.) ;
l A Civil Penalty of $50,000.00 was associated with the NRC OI i investigation (I-85-019) described in item 1 abov (Enforcement Action #87-05, dated February 8, 1987.) Orders
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Order Suspending Power Operation and Order to Show Cause (Effective Immediately), dated March 31, 1987. (EA 87-46)
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Confirmatory Order regarding actions to be taken for Unit 3 recirculation pipe weld defects, dated March 20, 198 Modified Order to remove.the required Unit 3 mid-cycle shutdown for weld examinations dated, December 31, 198 . Confirmatory Action Letters (CAL)
CAL #86-07, dated April 11, 1986, regarding actions to be taken with respect to fire protection and 10 CFR 50, Appendix . Enforcement Conferences
--
Unit 3 out-of-sequence control rod during startup on March 27, 198 Alleger fired for talking to NRC; discussed during meetings on May 30 and November 18, 198 Management Conferences Held During the Assessment Period
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Unit 3 CST spill status on February 21, 198 SALP management meeting on July 11, 198 Peach Bottom Enhancement Program status meetings on October 3 ,
and December 19, 1986, and February 20, 198 '
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Fire protection and CAL #86-07 status meeting on October 16, 198 Peach Bottom Order and plant status meetings on May 5, 1987 and May 15, 198 Licensee Event Reports (LERs)
Fifty-tnree LERs (Table 6) were submitted during the assessment period for Units 2 and Causally-linked event sets were identified as follows:
,
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Ten LERs concerning Reactor Protection System (RPS) and )
Engineering Safeguards Features (ESF) actuations caused by personnel erro Eleven LERs concerning RPS and ESF actuations caused by equipment failure Five LERs concerning containment leak rates exceeded because of equipment failure _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _
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Nine'LERs concerning RPS and ESF actuation caused by design or construction error Automatic Scrams and Unplanned Shutdowns During the assessment period, 26 automatic scrams and unplanned shutdowns occurred, as follows:
Unit 2
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Two. automatic scrams at power
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Five unplanned shutdowns Unit 3
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Eight automatic scrams at power
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Three automatic scram signals while shut down
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Three manual scrams at power
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Six unplanned shutdown Table 5 summarizes these scrams and shutdowns and indicates the root caus F. Licensing Activities NRC-Licensee Meetings and Site Visits Site Visits: February 24, March 17-21, May 20, July 3, and July ~14,1986 Meetings in 1986:
April 4 Quarterly meeting on licensing issues July 14 Quarterly meeting on licensing issues July 24 Discussion with Region I on safety issues management system (SIMS)
September 24 Containment venting analysis October 16 Followup on SALP review November 24 Relief from mid-cycle examination of welds December 3 Quarterly meeting on licensing issues
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2. Schedular Extensions Granted None 3. Relief Grar .ed April 8, 1986; certain inservice inspection requirements 4. Exemptions Granted November 14, 1986; certain requirements of Appendix R December 31, 1986; certain requirements of Appendix R 5. License Amendments Issued Amendments Nos. 117 and 121, issued March 14, 1986; revised TSs to permit bypassing of a scram signal for MSIV closure or main condenser low vacuam while not in the RUN mod Amendments Nos. 118 and 122, issued July 9, 1986; revised TSs approving changes in plant organizatio Amendments Nos. 119 and 123, issued July 30, 1986; revised TSs to add surveillance and operability requirements pertaining to Appendix R modifications on fire doors and penetration seal Amendments Nos. 120 and 124, issued August 11, 1986; revised TSs to clarify required water level in spent fuel poo Amendments Nos. 121 and 125, issued September 12, 1986; revised TSs to increase allowable hydrogen concentration limit downstream of the off gas recombiner to 4% (volume) and to decrease the number of hydrogen analyzers equired to be opercule during power operation from two to on I l
6. Emergency / Exigent Technical Specifications None 1
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T-1 l TABLE 1 History of SALP Reviews at Peach Bottom Atomic Power Station
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i Categories and Review Periods Functional Area 4/85- 1/84- 3/83- 3/82- 7/81- 7/80- 5/79-
. 1/86 3/85 12/83 2/83 6/82 6/81 6/80 1 Plant Operations 2 2 2 2 2 2 SAT Radiological 2 3 2 3 3 2 SAT Controls
' Maintenance SAT Surveillance 2 2 2 3 2 1 SAT Fire Protection & 2 2 2 3 3 3 SAT Housekeeping i Emergtacy 2 2 2 1 2 2 SAT I Preparedness'
1 Security and 3 a 1 1 2 2 SAT i Safeguards l
.,
~57'" Ref ueling/ Outage 1 1 2 2 2 1 SAT Activities
! Training and 2 N/A N/A N/A N/A 2 SAT Qualification Effectiveness '
>
10. Assurance of 3 N/A N/A N/A N/A 2 SAT :
Quality l
1 Licensing 2 1 1 2 1 N/A N/A Activities
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TABLE 2 INSPECTION HOUR SUMMARY Annualized Functional Area Hours % of Time Hours Operation 4142 5 Radcon/ Chemistry 853 1 Maintenance 649 Surveillance 715 1 Emergency Pre .2 169 Sec/ Safeguards 144 . Technical Support * 0 Training * 0 Licensing ** 0 Quality Assurance * 0 Fire Protection 281 TOTALS 7009 10 * Hours expended are included with other functional areas
- Hours expended are not included with direct inspection statistics
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T-3 TABLE 3 INSPECTION REPORT ACTIVITIES Report No./ Dates Inspector Hours Areas Inspected Unit 2 Unit 3 86-03 86-03 Resident 222 Resident operational safety 02/01/86 03/26/86 86-04 86-04 Specialist 0 Operator licensing examinatio1s 03/10/86 03/13/86 86-05 86-05 Resident 75 Condensate storage tank spill and 02/16/86 04/04/86 release 86-06 86-06 Specialist 60 Emergency preparedness 03/10/86 03/13/86 86-07 86-07 Resident 279 Resident operational safety 03/27/86 05/09/86 86-08 86-08 Specialist 176 Fire protection 03/17/86 03/21/86 86-09 Resident 55 Out-of-sequence control rod 03/18/86 03/21/86 86-09 86-12 Resident 286 Resident operational safety 05/10/86 06/17/86 86-10 86-10 Specialist 7 Security 04/01/86 04/02/86 86-11 86-11 Specialist 34 Quality assurance annual review 04/07/86 04/11/86 86-12 86-13 Resident and 640 Region I Diagnostic Team 06/18/86 07/03/86 Specialist Inspection 86-13 86-14 Resident 202 Resident operational safety 07/04/86 08/15/86 86-14 86-15 Specialist 42 Spent fuel racks and emergency 08/04/86 08/08/86 service water 86-15 86-16 Specialist 138 Annual emergency preparedness 10/08/86 10/10/87 exercise 86-16 86-17 Resident 190 Resident operational safety 08/16/86 09/26/86
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T-4
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TABLE 3 (continued)
l Report No./0ates Inspector Hours Areas Inspected Unit 2 Unit 3 86-17 86-18 . Specialist 8 Security 08/25/86 08/28/86 86-18 86-19 Specialist 89 Radiological controls 09/08/86 09/12/86 86-19 86-20- Resident 268 Resident operational safety 09/27/86 11/07/86 86-21 86-22 Specialist 39 Radiological controls 10/20/86 10/24/86 86-23 86-24 Specialist 32 Environmental monitoring 11/03/86' 11/07/86 86-24 86-25 Resident 158 Resident operational safety 11/08/86 01/02/87 86-25 Specialist 448 Probabilistic Risk Assessment 11/21/86 12/19/86 Team Inspection 86-26 86-i!6 Specialist 44 Radiological controls 12/15/86 12/17/86 87-01 87-01 Specialist 29 Nonradiological chemistry 01/12/87 01/1.5/87 program 87-02 87-02 Resident 182 Resident operational safety 01/03/87 01/31/87 i 87-03 87-03 Specialist 59 Water chemistry control '
01/26/87 01/30/87 -
87-04 87-04 Specialist- --
Operator examination 02/16/87 02/19/87 87-05 87-05 Specialist 74 Radiological measurements ,
02/02/87 02/06/87 {
87-06 87-06 Specialist --
Fitness for duty 02/02/87 02/06/87 87-07 87-07 Resident 585 Resident operational safety 02/01/87 03/13/87 f
87-08 87-08 Specialist i 246 Maintenance and surveillance J 02/23/87 03/06/87 program l
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TABLE 3(continued)
i Report No./ Dates Inspector Hours Areas Inspected Unit 2
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Unit 3 87-09 87-09 Resident 163 Resident operational safety 03/14/87 04/24/87 87-10 87-10 Resident / 530 Special safety team 03/24/87 04/09/87 Specialist inspection l
87-11 87-11 Specialist 5 Special fire protection 04/09/87 04/09/87 87-12 87-14 Specialist 30 Local leak rate testino 04/20/87 04/24/87 87-13 87-13 Specialist 118 Radiological controls 04/20/87 04/24/87 87-14 87-12 Specialist 20 Security 04/20/87 04/23/87 87-15 87-15 Resident 307 Resident operational safety 04/25/87 05/31/87
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TABLE 4 ENFORCEMENT SUMMARY SEVERITY LEVEL Functional Area 1 2 3 4 5 DEV TOTAL I: Operations 2 2 4 Radcon/ Chemistry 5 5 Maintenance 2 1 3 Serve 11ance 2 2 Emergency Pre Sec/ Safeguards Outages Training Licensing Quality Assurance Fire protection 1 1 TOTALS: 2 12 1 15 Functional Report No./ Gates Requirements Level Area Violation 277/86-09 10CFR 50 & 4 Maintenance Failure to promptly 05/10/86 06/17/86 QA Plan identify and correct
'
deficiency conditions i associated with the emergency load transformers 277/86-12 TS 4 Surveillance Failure to perform 06/18/86 07/03/86 Requirements surveillance on emergency core cooling systems 277/86-18 TS Radiation Failure to initiate an 09/08/86 09/12/86 control A-86 deficiency report
277/86-21 10CFR 4 Radiation Certification of l 10/20/86 10/24/86 20.311(c) control errors of shipment i 119-86 278/86-05 TS 6. Radiation Failure to have liquid 02/16/86 04/04/86 control relecse procedure for condensate storage tank (CST)
278/86-05 10CFR20.201, 4 Radiation Failure to sample and 02/16/86 04/04/86 control evaluate release from CST l
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T-7
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TABLE.4(continued)
Functional Report No./ Dates Requirements Level Area Violation 278/86-05 10CFR50.59 4 Maintenance Failure to perform a 02/16/86 04/04/86 safety evaluation for operation with dike
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modified and CST with l higher levels of radioactivity than evaluated 278/86-07 TS 4 Surveillance Failure to perform 03/27/86 05/09/86 4.6.E. jet pump operability checks while in single loop operation 278/86-09 TS 3 Operations Failure to follow the 03/18/86 03/21/86 3.3.8. correct rod pull program with rod worth minimizer bypassed 278/86-09 TS Operations Failure to ensure rod 03/18/86 03/21/86 & in correct position with rod sequence control system bypassed
.278/86-12 10CFR50, 4 Operations Failure to promptly 05/10/86 06/17/86 App B identify and correct deficiency conditions associated with the emergency load center transformers 277/86-25 FSAR DEV Maintenance Failure to test and 12/08/86 12/19/86 8.7. inspect OC under voltage relay annunciation 277/86-25 TS 4 Fire Failure to establish 12/08/86 12/19/86 3.14. protection adequate backup fire watch 277/87-07 10CFR20,202, 4 Radiation Failure to post high 02/01/87 03/13/87 203 control radiation area 277/87-10 TS 4 Operations Failure to follow 03/24/87 04/09/87 6. shutdown cooling procedure
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TABLE 5 AUTOMATIC SCRAMS AND 'JNPLANNED SHUTDOWNS The automatic reactor scram signals and unplanned shutdowns that occurred during this assessment period fall into two categories. These categories
~
. include personnel' error and equipment failure or malfunction. The following table assesses the root cause of each trip or shut down within each category from NRC's perspectiv The root cause was determined by the SALP Board and may not agree with LER analysis, if applicabl Personnel Errorr There were 10 of 26 trips or shutdowns attributed to personnel error. For the purposes of this table, personne' error has been broken into four groups: (1)
poor judgment, the individual should have known that the outcome of the action could cause a trip; (2) lacking (nowledge - the individual had not been instructed or did not have the lnowledge that an act performed would cause a tr:p; (3) inattention to detail - the individual took a haphazard approach to ar, unrelated task which subsequently led to a trip; and (4) equipment malfunction - an equipment failure or malfunction in conjunction with a personnel error, where both were necessary to cause the tri Eguipment Malfunction or Failure There were 16 of 26 trips or shutdowns attributed to equipment malfunction or failure. For the purposes of this table, equipment malfunction or failure has been broken into three groups; (1) Random failure - isolated failt.as not considered generic, (2) Design deficiencies - failures attributed to equipment design, and (3) Construction deficiencies - failures attributed to improper installation during constructio Unit 2 (2/1/86 thru 5/31/87)
Power SALP N Date Level Description Root Cause* _ Area 1 3/17/86 100% Plant shutdown because the in- Equipment take basin water level began malfunction - NA dropping due to fouling of the random outer screens with trash and debris. Heavy spring rains and snow melt caused higher than normal Susquehanna River flo !
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TABLE 5 (continued)
Power SALP No. Date Level Description Root Cause* Area 2 4/23/86 100% Automatic. scram due to turbine Equipment NA control valve (TCV) fast closure malfunction -
during testing of combined random intennediate valves (CIV)
caused by the turbine generator power / load unbalance circui The #1 CIV inadvertently closed when the #2 CIV was closed, causing a crossover steam pressure spike and power / load unbalance trip and TG trip. The closure of the
- 1 CIV was caused by air trapped in
-
the EHC hydraulic lines. Addition of restricting orifices under a General Electric Company (GE)
modification to prevent hydraulic transients was performed on #1 and
- 2 CIV (LER 2-86-13)
3 6/18/86 100% Plant shutdown because of a Equipment NA leak in the emergency service failure -
water (ESW) system supply to random to 2A RHR pump room coole ESW was isolated to all ECCS room coolers resulting in TS LCO requiring a shutdow (LER 2-86-14)
4 8/7/86 100% Plant shutdown because the Equipment NA air-operated testable check failure -
valve (AO-22) on the RCIC random pump discharge line developed a packing lea /29/86 100% Plant shutdown because of Equipment NA contamination of the electro failure -
hydraulic control (EHC) system random oil with water. A packing leak in one of the nine bypass valves l allowed water to drip into an EHC system oil catch pan. The water from the catch pan cavity drained back to the EHC oil supply tan ;
I I
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T-10
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TABLE 5 (continued)
Power SALP N Date Level Description Root Cause Area 6 10/13/86 70% Automatic scram on low main Equipment NA condenser vacuum because of failures -
inleakage from 28 condensate pump construction suction valve. The 2B condensate deficiency pump was out of service because of a failed thrust bearing, and the suction butterfly valve did not seat because there was a piece of wire mesh remaining from startup screen material on the suction line. The startup screen was not completely removed following initial startu (LER-2-86-22)
7 10/15/86 1% Plant shutdown for repairs to two Equipment NA condensate pumps, a traversing failures -
incore probe system ball valve, random and an intermediate range monito Unit 3 (2/1/86 thru 6/30/87)
1 3/6/86 3% Automatic scram on reactor low Personnel DPS water level when operator lowered error /
the "C" RFP speed with the motor inattention gear unit (MGU) to prevent a high to detail water level trip due to increasing level. The increased level resulted from a swell when two bypass valves were opened to lower reactor pressure as the vacuum started decreasing because of condensate pump maintenanc The "C" RFP turbine control was out of alignment. The MGU low-speed stop was determined to be 800 rpm in lieu of 2200 rp Thus, the RFP did not quickly respond to subsequent water level decrease and the reactor scrammed on low water level. (LER 3-86-05)
2 3/16/86 30% Plant shutdown because the in- Equipment NA l take basin water level began malfunction -
'
dropping because of fouling of random the outer screens with trash and debris. Heavy spring rains and snow melt caused a higher than normal Susquehanna River flow.
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T-11
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TABLE 5 (continued)
Power SALP N Date Level Description Root Cause Area 3 3/18/86 3% Manual scram inserted by the Personnel OPS-operator when it was discovered error /in-that a Group 1 control rod was attention out of sequence and inadvertently to detail left in a full-in position during startup. A scram was inserted because of concern.that the reactor might be outside'the bounds of the rod drop accident analysis. (LER 3-86-09)
4 4/11/86 80% Plant shutdown for minioutage Equipment NA sto repair miscellaneous valves failure -
and three minimum flow valves for random the three reactor feed pump /11/86 Shut Auto scram on reactor low water Personnel OPS Down level while in hot shutdown during error plant.cooldown with RCIC syste Automatic level control was out of service because of the failure of the RFP's minimum flow valve The operator's inattention to detail while controlling level with the RCIC system led to the scram. No rod movement occurred. (LER 3-86-10)
6 4/26/86 81% Automatic scram because of TCV Personnel SRV fast closure during substation error /in-breaker relay testing. Testing attention on the 65 breaker, which was open, to detail -
caused the 15 breaker to open, I&C as well as a generator lockout
.
Technicians 1 and turbine trip (load reject). j I&C technician error resulted
'
in the trip not being bypassed q for the 15 breaker. This caused the 15 breaker to ope {
(LER 3-86-12)
7 4/26/86 Shut IRM scram during neutron Personnel SRV ,
Down monitoring surveillance testin error / poor Two IRMs (one in each RPS trip judgment -
channel) were unbypassed with Licensed their mode selector switches "out Operator of operate." This caused an IRM failure to -
INOP scra No rod movement follow test I occurre (LER 3-86-13) procedure
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T-12 gr TABLE 5 (continued)
Power SALP NA Date Level Description Root Cause Area 8 4/26/86 Shut SDV high level scram when the Personnel SRV Down operator placed the reactor mode error / poor selector switch in startup with judgment -
an actual high level in the Licensed i SDV instrument volume. No rod Operator movement occurred. (LER3-86-14) failure to follow test procedure 9 7/19/86 87% Automatic scram due to high APRM Equipment NA flux caused by a reactor pressure failures -
spike when one MSIV closed. The multiple MSIV closure was caused by loss of (MSIV power to the AC coil with the DC DC coil and coil faile Loss of power was off site caused by en electrical fault of power)
one off-site power sourc (LER 3-86-16)
10 8/11/86 86% Automatic scram on reactor low Equipment NA water level when one RFP tripped failures -
'
because of actuation of the RFP multiple, protective instrumentation random (MCC (thrust-bearing detector). A electrical nonvital MCC tripped because of fault and an electrical fault of the RFP lube oil generator stater coolant pump PCV)
breaker. This caused a loss of the running RFPT lube oil pump. The standby pump started, but failure of the pressure control valve (PCV) caused a lobe oil surge. This caused a thrust bearing wear detector trip and an RFP trip. The two operating RFPs could not recover level prior j to the scra (LER 3-86-18) '
11 9/14/86 57% Manual scram from 58% when MSL Personnel MNT radiation monitor spiked due to error / I
'
3H condensate demineralized resin 1ccking {
injectio Scram initiated as knowledge - J required by procedure OT-10 Inadequate
'
The resin injection was caused post mainte-by residual resin left in the nance demineralized outlet plenum inspection and/or piping after element
.
replacemen (LER 3-86-19)
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TABLE 5 (continued)
Power 5 SALP No. Date- Level Description <
Root Cause Area
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,
12 9/16/86 60% Plant shutdown because the HPCI Equipment NA systemwasdeclaredinoperable failure -
I because of a packing leak in the random inboard steam supply isolation valve (M0-15).
131 10/07/86 77% Automatic scram on low reactor Equipment MNT water level. A single fault in malfunction the FWLCS caused "A" and "C" RFPs to runback because of hign indicated feed flow that resulted in RFP check valve slam. _0perator took manual control and tripped
"B" RFP because of high indicated feed flow, and other RFPs could not recover the water leve The fault was an RFP "B" corroced thermocouple that provides temperature compensatio (LER 3-86-20).
14 10/30/86 26% Manual scram when MSL radiation Personnel AQ monitor spiked due to 3E error /
concensate demineralized resin inadequate injectio Scram initiated as corrective part of shutdown caused by high action reactor conductivity (41 micrombos/
cm). The resin injection was caused by residual resin left in the demineralized outlet plenum and/or piping after element replacemen (LER 3-86-22)
15 11/4/86 5% Automatic scran during startup Personnel OPS when turbine shell warming was error /
in progress. Turbine first stage inattention pressure increased to the setpoint to detail at which the turbine stop valve closure scram was no longer I'
bypassed (i.e., 30% reactor power).
Tnis resulted in an automatic scra Inadequate pressure indication on the meter and pressure switch _
setpoint drift, combined with }
inattention by the reactor operator, I caused the scra (LER 3-86-23) {
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l T-14 E TABLE 5 (continued)
g^
Power SALP N,o . . Dat Level Description Root Cause Area l ,
- 16 12/28/86 100% Plant shutdown for repairs to Equipment NA steam leaks on the "C" moisture failure -
separator drain tank manway random and on the "D" main steam line inlet flange to the HP turbine.
l l 17- 1/2/87 40% Plant shutdown for additional Equipment NA l repair to a. steam leak on the failure -
"D" main steam line inlet flange random to the HP turbine.
l 18 3/5/67 100% Plant shutdown to repair a Equipment NA
. hydrogen leak in the main failure -
generator, random 19 3/17/87 '85% Automatic scram from 85% power Equipment NA because of APRM high ilux. The malfunction /
high flux condition resulted from random
,
pressure transients caused by EHC failures system induced oscillations of the turbine control valves. The EHC oscillations were caused by mechanical failure of panel '
,
internal cooling fans. The starting and stopping of these fans induced sufficient electrical noise to cause the turbine control valves to spik (LER 3-87-01)
20 3/25/87 1% Automatic scram from 1% power due Personnel MNT to low reactor water leve erro r/ lac king ,
During EHC troubleshooting an knowledge engineer lifted a lead and caused all the bypass valvres to open. The resultant 1cvel swell caused the operating RFP to tri The blowdown and loss of RFP resulted in low level and a scra J (LER 3-87-02)
I
_ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . . _
- _ - _ _ .
.
T-15
.-
TABLE 6 LISTING OF LERs BY FUNCTIONAL AREA AND CAUSE CODES AREA A B C D E X- TOTAL Operations 6 5 2 9 22 Radcon/ Chemistry Maintenance 3 4 2 1 10 Surveillance 6 1 1 9 3 20 Emergency ~ Pre Sec/ Safeguards Outages Training Licensing Quality Assurance Other Fire Protection 1 1 2 Totals: 16 10 0 5 19 4 54 Cause Code A - Personnel Error B - Design, Manufacturing, Construction, or Installation Error C - External Cause D - Defective Procedure E - Component Failure X - Other LE3 NUMBER EVENT DATE CAUSE CODE DESCRIPTION 2-86-06 2/01/86 B Primary containment isolation system (PSIC) group II/III outboard isolation 2-86-07 2/03/86 X Isolations caused by reactor protection system (RPS) motor generator 2-86-08 02/27/86 E Containment isolation caused by RPS, MG set trip 2-86-09 3/18/86 B Automatic scram caused by loss of power to "A" RPS with scram discharge volume (SDV) high level
,
2-86-10 4/13/86 E Containment isolations during
,
automatic switching caused by I
transformer failure 1 ,
J
- _ _ _ _ _ _ - _ .
l T-16 i .
TABLE 6 (continued)
LER NUMBER EVENT DATE CAUSE CODE DESCRIPTION
,
2-86-11 3/26/86 A Reactor water cleanup system l inboard isolation 2-86-12 4/13/86 E Loss of power to 4KV buses because of failed diesel speed device 2-86-13 4/23/86 B Scram during turbine conbined intermediate valve test because of air in the hydraulic system 2-86-14 6/18/86 E Shutdown required by technical specifications - emergency service water (ESW) pipe leak 2-86-15 6/20/86 E Exceeded allowable contain'nant leak rate limit 2-86-16 7/09/86 X High pressure coolant injection system inoperable because of control box failure 2-86-17 7/04/86 A Failure to trip the rod block logic of the reactor manual control system 2-86-18 08/26/86 A Reactor water level transmitter out of service 2-86-19 8/08/86 E Exceeded allowable containment leak rate limit 2-86-20 8/17/86 D Failure to maintain cold shutdown conditions and exceeding torus volume limit 2-86-21 8/13/86 A Work on core spray valves without safety blocking 2-86-22 10/13/86 B Low condenser vacuum scram caused by foreign material in valve seat 2-86-23 10/11/86 E Possible degradation of primary containment because of binding in !
3/8-inch ball valve
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
.
T-17
TABLE 6 (continued)
LER NUMBER EVENT DATE CAUSE CODE DESCRIPTION 2-86-24 12/9/86 E 0xygen analyzer isolation caused blown fuse 2-87-01 3/11/87 A Exceeding suppression pool water volume limit 2-87-02 3/13/87 E Exceeding total containment leakage rate limit 2-87-03 3/28/87 A Group II certainment isolations caused by fuse removal 2-87-04 4/7/87 E Group III containment isolations caused by partial loss of offsite power 2-87-05 4/7/87 E Exceeding total containment leakage rate limits 2-8/-06 4/23/87 E Group III containment isolations-caused by RPS MG set rheostat 3-86-01 2/03/86 B PCIS Group II/III outboard isolatien 3-86-02 2/17/86 A Reactor water cleanup isolation while moving leads 3-86-03 2/19/86 A Containment isolation caused when wrong fuse was removed 3-86-04 2/28/86 B Two inadvertent starts of 3D residual heat removal (RHR)
pump 3-86-05 3/06/86 B Low level scram involving reactor feedwater pump speed control 3-86-06 3/04/86 B Loss of 3A core spray system logic 3-86-07 3/05/86 D Inadvertent bypassing of the rod sequence control system 3-86-08 3/04/86 X Inoperable high pressure coolant j injection valve because of relay '
failure
i
-
\
I i
-- _ __ -_ -- _ -- _ - __ _ ----- D
_ _ - _ _ -__
....
T-18
-
TABLE 6 (continued)
LER NUMBE EVENT DATE CAUSE CODE DESCRIPTION 3-86-09 3/18/86 A Out-of-sequence control rod 3-86-10 '4/11/86 B Reactor. low water level scram while using reactor core isolation cooling (RCIC) for level control f while shutting down 1 3-86-11 4/25/85 E Group III isolation as a result of blown PCIS fuse 3-86-12 4/26/86 A Full scram while testing 500kv line fault detection circuit 3-86-13 4/26/86 A False neutron monitoring scram caused by personnel error 3-86-14 4/26/86 A Scram discharge volume high level scram caused by personnel error 3-86-15 6/24/86 B RWCU system isolation. caused by wiring error and blown fuse 3-86-16 7/19/86 E Scram and isolations caused by power supply transient and MSIV closure 3-86-17 8/07/86 X Missec: fire watch patrol because of personnel error 3-86-18 8/11/86 E Full scram on low reactor water levei j
3-86-19 9/14/86 D Manual scram after resin injection 3-86-20 10/07/86 E Reactor scram caused by feedwater transient
03-86-21 10/21/86 A Suppression pool water level above technical specification limit 03-86-22 10/30/86 D Manual scram after resin injection ;
f 03-86-23 11/04/86 A Full scram during turbine shell I l- warming j c 4 l
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ 1
_ - _ _ _ _ _ _ _ _ _ _ _ _
l
.o T-19 l
.
l Table 6 (continued)
LER NUMBER EVENT DATE CAUSE CODE DESCRIPTION 03-86-24 11/08/86 E RCIC system isolation caused by loose wire 3-87-01 3/17/87 E Automatic scram on high flux caused by turbine electro hydraulic control (EHC) problems 3-87-02 3/25/87 A Automatic scram on low reactor water level during EHC troubleshooting 3-87-03 3/31/87 0 PCIS group II shutdown cooling isolation 3-87-04 4/2/87 E MSIV 80D local leak rate test failure 3-87-05 4/6/87 A Containment group II isolations of reactor water cleanup during surveillance testing I
i
)
_ _ _ _ _ _ - - - - -_ -_-- -._._- __ ,
- _ _ _ _ _ _ _ _ _ _ . . _ _ - _ - _ _ - -_
!
.9 -
T-20
Table 7 List of Acronyms and~Initialisms Used in this Report i AC alternating current AEOD Office of Analysis and Evaluation of Operational Data ALARA as low as reasonably achievable A0 air operated APRM average power range monitor AQ assurance of quality ATWS anticipated transient without scram BWR boiling water reactor CAL confirmatory action letter CIV combined intermediate valves CS condensate storage tank'
DC direct current DG diesel generator l
L ECCS emergency core cooling system EHC electro hydraulic control EOF emergency operations center EPRI Electric Power Research Institute EQ environmental qualification ESF engineered safety feature ESW emergency service water ETT equipment trouble tag FHA fire hazard analysis FWLCS feedwater level control system !
\
GE General Electric Company GET general employee training GRT general respiratory training '
HP high pressure or health physics .
HPCI high pressure coolant injection !
HPES human performance evaluation system IDCOR Industty Degraded Core Rulemaking IEEE Institute of Electrical and Electronics Engineers
.IGSCC intergranular stress corrosion cracking '
INOP inoperative l INP0 Institute of Nuclear Power Operations l
"
IPE Independent Plant Evaluation j IRM intermediate range monitoring ]
ISEG Independent Safety Engineering Group '
I&C instrumentation and control JC0 justification for continued operation
)
- _ _ _ _ - -
l *
T-21
Table 7 (continued)
LCO- limiting condition LE license event report LOCA loss-of-coolant accident MCC motor control center MC motor generator MGU motor gear unit MNT maintenance MOV motor-operated valve MRF maintenance request form MRFACE maintenance root cause failure analysis and coordination engineer MSIV main steam isolation valve MSL main steam line NRB Nuclear Review Board NSHC no-significant-hazard consideration OEAC Operating Experience Assessment Committee 01 Office of Investigations OPS operations OSC operations support center PCIS primary containment isolation system PCV pressure control valve PORC Plant Operations Review Committee PRA probabilistic risk assessment QA quality assurance !
QC quality control RCIC reactor core isolation cooling RFP reactor feedwater pump ;
RFPT reactor feedwater pump turbine RHR residual heat removal RP radiation protection RPS reactor protection system RSCS rod sequence control system RWCU reactor water cleanup RWM rod worth minimizer SALP systematic assessment of licensee performance SDV scram discharge volume SIMS safety issues management system i SR0 senior reactor operator SRV safety relief valve or surveillance (in Table 5)
ST surveillance test STA shift to:hnical advisor TCV turbine control valve TG turbine generator TRIP transient response implementation procedure TS technical specifications TSC technical support center
_ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _
__ -._____ _
,
O
. F-1
FIGURE 1 NUMBER OF DAYS SHUT DOWN AT PEACH BOTTOM ATOMIC POWER STATION Unit 2 Unit 3 Feb. 86 5 Day l Feb. 86 28 Days l l l Mar. 86 1 Day l Mar. 86 3 Days l l 1 Apr. 86 2 Days l Apr. 86 12 Days l l l May 86 May 86 June 86 5 Days l June 86 I
July 86 July 8611 Days l l
Aug. 86 18 Days l Aug. 86 17 Days l l l Sep. 86 1 Day l Sep. 86 2 Days l 1 I Oct. 87 11 Days l Oct. 86 2 Daysl l I Nov. 86 Nov. 36 4 Days l l
Dec. 86 Dec. 86 3 Days I (
l I
,
Jan. 87 Jan. 87 4 Days l j l
l
1 l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ l
_ _ _ _ _ - - _ _ _ _ _ _ - - _ _ _
e
.
F-1(continued)
FIGURE 1 NUMBER OF DAYS SHUT DOWN AT PEACH BOTTOM ATOMIC POWER STATION Unit 2 Unit 3
,
Feb. 87 Feb. 87
- '.
Mar. 87 18 Days l Mar. 87 4 Days l l .I
.
Apr. 87 30 Days l Apr. 87 30 Days l
~
__! !- 1 May 87 31 Days l May 87 31 Days l l . I
1
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