IR 05000354/1985098
| ML20210C407 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 04/29/1987 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20210C302 | List: |
| References | |
| 50-354-85-98, NUDOCS 8705060179 | |
| Download: ML20210C407 (72) | |
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ENCLOSURE 2
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SALP BOARD REPORT
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE INSPECTION REPORT 50-354/85-98 PUBLIC SERVICE ELECTRIC AND GAS COMPANY HOPE CREEK NUCLEAR GENERATING STATION ASSESSMENT PERIOD: NOVEMBER 1, 1985 - NOVEMBER 30,-1986 BOARD MEETING DATE: JANUARY 28, 1987
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PRESENTATION TO LICENSEE: APRIL 7, 1987
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SUMMARY OF RESULTS 3.1 Overall Facility Evaluation l
The licensee completed the transition from a construction facility
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to an operating nuclear power plant during this assessment period.
The plant progressed from a 90*4 complete construction status to being only a few weeks away fnm commercial operation in a thirteen month period. A very ambitious schedule was established by manage-ment, and, although not met for most milestones, it did provide good direction throughout the period.
Despite the ambitious schedule, a good perspective on quality and nuclear safety was maintained.
Plant procedures and administrative programs are generally of high quality, due in part to the operating experience evaluation program.
Some aspects of the radiation protection program, however, warrant additional management attention.
Efforts to improve administrative activities without sacrificing quality are also needed.
The incident report program provides excellent feedback of operating experience to all departments.
Control room operations have been conducted in a consistently pro-fessional and safety conscious manner. Noise and access control, especially during power ascension testing, have been excellent.
Except for two operator-error-induced scrams early in the test pro-gram, the operators have performed well throughout the period.
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shift turnover meetings, work control group, and Technical Speciff-cation interpretations promote good performance in the operations
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area. Areas which warrant attention include: maintenance of control room logs, reducing the number of alarming annunciators and reducing the number of unplanned scrams and reportable events.
The organization is generally well staffed with qualified personnel.
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The radiological and chemistry department vacancies which have been recently created need to be filled promptly in order to provide the necessary supervisory oversight. Approximately one-third of all reportable events were attributable to personnel error (mostly during surveillance tests). The major contributor to these events has been spurious initiation signals of the engineered safety features (ESF).
The occurrence rate of these events has been significantly reduced by comprehensive corrective action programs.
Overall, a solid foundation has been established for the first cycle of plant operation. Management support is evident, particularly in the areas of emergency planning, security, and quality assurance.
The licensee recognizes the need for additional attention to support programs, in particular, radiological controls.
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3.2 Facility Performance Functional Category Category Recent Area Last Period This Period Trend (11/1/84-10/31/85) (11/1/85-11/30/86)
A.
Plant Operations 1*
B.
Radiological Controls
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C.
Maintenance
1 D.
Surveillance Not
Evaluated E.
1 F.
Security and Safeguards
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Outages Not No Rating Evaluated H.
Preoperational and Startup Testing
2 I.
Licensing Activities
1 J.
Training and Qualification Not
Improving Effectiveness Evaluated K.
Assurance of Not
Quality Evaluated This area was titled Operational Readiness in the previous SALP
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IV.
PERFORMANCE ANALYSIS A.
Plant Operations (33%, 3030 Hours)
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Analysis
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The functional area of " operational readiness" was evaluated
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to be Category I during the previous assessment period.
Some weaknesses were identified but the general conclusion was that the transition from construction to operations was well control-
led, staffing was adequate and experienced, training programs
were effective, and administrative controls under development I
appeareo generally adequate.
The SALP Board recommended that the applicant provide NRC an operational readiness presentation, based on a self appraisal, which was completed during April, 1986.
The operations area was under continual review by two resident inspectors for the entire assessment period and by a third
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resident for a total of four months.
In addition to the resident inspectors, this area was reviewed by preoperational and startup program inspectors, the augmented inspection and operational assessment teams, and senior NRC management during numerous site visits. Two sets of initial operator licensing examinations were given to a total of 25 candidates during February and July, 1986.
Training and qualification effec-tiveness is discussed in Section J of this report.
Plant operations have been conducted in a consistently conservative and safety conscious manner.
The transition of
project completion responsibility from the Vice President -
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Engineering to the Vice President - Nuclear on December 2, 1985 (5 months prior to fuel load), helped changa the focus from construction completion to plant operations. Assigning the assistant general manager for Hope Creek operations to the
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position of project completion manager and providing him with the resources necessary to do his job, significantly contrib-uted toward establishing a high standard of performance and emphasis on nuclear safety. A safety conscious attitude was apparent throughout the entire Hope Creek operations organization.
Senior plant management is intimately involved l
with the day-to-day operation of the plant.
The station's
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general manager and all department managers attend a daily l
management meeting to discuss current issues and establish priorities for future activities.
The Vice President - Nuclear
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occasionally attended these meetings. All work activities are
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scheduled by the planning department based upon the priorities l
established by management and input from the work group supervisors.
This method of planning and scheduling has worked i
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well throughout the power ascension program and has ensured that the " big picture" was maintained.
The station operations
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i review committee (SORC) has generally done a thorough job of
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overseeing plant operations.
The offsite safety review group l
performed a number of in-depth reviews including an independent
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investigation into the causes for the inoperability of the l
reactor building to suppression chamber pressure relief system.
In addition to an accurate assessment, their recommended cor-i rective actions were timely and effective.
The licensee has been responsive to NRC concerns both prior to, and since, plant licensing. Major NRC team inspections such
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as the As-Built, Technical Specification, Augmented Inspection Team, and the Operational Assessment Team inspections received l
timely and effective support during the assessment period.
Priortoplantlicensing,commitmenttrackingsystemhasensured l
all appropriate NRC open items were resolved.
The licensee s prompt resolution of outstanding inspector concerns. Numerous briefings were conducted for the NRC on spurious engineered safety feature (ESF) actuations, Bailey 862 solid state logic modules, the inoperable reactor building to suppression chamber pressure relief system, and the loss of offsite power tests.
Plant procedures and administrative controls are thorough and based upon a review of over 3000 documents such as IE bulle-tins, circulars, information notices, INPO documents and vendor
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recommendations. However, in an effort to incorporate these numerous requirements, recommendations, and good practicos, a large administrative burden has been created for the plant staff. Occasionally, this burden impacts negatively on the implementation of the overall program. A review of the equipment malfunction identification tagging (EMIT) system identified a large percentage of tags on equipment in the plant were no longer valid, and system walkdowns by the NRC i
have identified a number of discrepancies in the tagging request inquiry system (TRIS) valve lineups.
It appears that this administrative burden contributed to a month long delay in determining the inoperability of the reactor butiding to suppression chamber pressure relief system.
The Operations Department has a more than ample number of both licensed and non-licensed operators to meet staffing require-monts and man a 5 shift rotation with a minimum use of overtime.
The control room is consistently maintained in a professional manner with very good access and noise control. Noise control i
is especially aided by the plant page system design which pre-
vents routine pages from being heard in the control room.
The control room environment is also aided by the use of a work control group that processes all work orders, surveillance tests, and blocking permits outside of the control room with
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the exception of the senior nuclear shift supervisor's (SNSS)
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final approval.
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The narrative control room log needs to be improved. On occasion, the logs have been found to lack detail and are inconsistently maintained among different shifts.
In addition, there is significant duplication of information between the
$NSS, shif t supervisor, and control room operator's legs. Also, the control room alarm system needs to be improved. The large number of overhead annunciators that are in alarm at any given time, interferes with the ability to understand current plant conditions. During full power operations, over 50 annunciators in alarm have been observed.
With the exception of two violations identified shortly after initial licensing, no further Technical Specification adherence problems have been identified in the operations area.
The establishment of a formalized TS interpretation log has aided the operators in establishing a consistent and well thought out approach to TS compliance.
Shift briefings conducted by the SNSS in the operations support center are noted as a strength.
Pre-shift briefs are conducted for both the operators and all other support organizations.
Despite the pre-shift briefings, the interface between opera-tions and chemistry needs improvement. A number of TS action statement violations involving a failure to take a sample have occurred, partially as a result of inadequate communications between departments.
Control room operator errors directly caused, or may have con-tributed to, 3 of the 14 unplanned scrams during the power ascension program.
(All plant scrams are described in Table 6 of this report.) Although higher than desired, this number appears consistent with other recently licensed BWRs. There have been 89 reportable events since low power license issuance on April 11, 1986. Of these events, 47 can be categorized as personnel errors and/or new procedure problems.
The major contributors to the reportable events are:
7 loss of coolant accident (LOCA) signals, 9 engineered safety feature (ESF)
actuations, and 5 high pressure coolant injection (HPCI) system actuations. The majority of these spurious signals share the common root causes of valve misoperation, and inadvertent con-tact with sensing lines during drywell work.
The licensee formed a task force to investigate these events and completed the following corrective actions:
(1) installed quick discon-nects on instruments, (2) installed identification tags on sensing lines, (3) installed protective cages around instrument racks, and (4) blew back instrument lines to remove entrapped air.
Based upon recent performance, these correctivo actions have been effective.
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l The licensee has implemented a strong housekeeping program throughout the plant. An integral part of this program is the plant management tours made on a routine basis and the follow-up
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inspections to verify implementation of corrective action. A
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plant painting program is being implemented that should also l
improve the plant's appearance.
Considering the status of the l
plant during this assessment period, housekeeping and cleanit-j ness are adequate.
During the assessment period, two inspections were performed
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to review the licensee's fire protection program, the system's l
installation, and the FSAR and Technical Specifications for
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compliance with Generic Letter 86-10. During these inspec-tions, corporate and site management exhibited thorough and aggressive involvement with, and control of, fire protection program activities.
It was also evident that priority was given to problems requiring hardw.tre solutions.
The licensee requested deletion of the fire protection Technical Specifica-tions as recommended in Generic Letter 86-10. The NRC deter-mined that deletion of these Technical Specifications was in accordance with the guidance provided in the Generic Letter and that existing fire protection requirements have been incorporated into plant procedures and equivalent adminis-trative controls exist to control these activities.
It was concluded that adequate controls exist to evaluate fire pro-tection program changes and ensure the ability to achieve and maintain safe shutdown in the event of a fire.
Staffing for the fire protection program and training of personnel were judged to be adequate.
In summary, the proper perspective on safety has been estab-lished throughout the plant staff and station procedures.
Control room operator performance during plant transients and events has been a noteworthy strength.
Strong management attention is evident in the day-to-day operation of the facility.
2.
C_o_n.c l__u s i on Rating: Category 2 Trend None
Doard Recommendations Licensee: Evaluate methods to improve administrative activities consistent with safe plant operations.
NRC: None
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Radiological Controls and Chemistry (6L 592 Hours)
1.
Analysis During the previous assessment period, the licensee's perform-ance was evaluated as Category 2 in the area of Radiological Controls and Chemistry. Weaknesses identified during the period were:
a lack of adequate licensee oversight and atten-tion to detail in the development of the radiation protection program; a need-to improve coordination and communication between the operations, radiation protection, and chemistry groups; and a lack of adequate justification to support deferral of operability of certain process radiation monitors.
Inspection early in this period found a continuation of the radiation protection program development problems identified during the last period.
These included inadequacies in the radiation work permit program, high radiation area access control program, and airborne radioactivity sampling and anal-ysis program. A number of technical deficiencies in procedures were also identified and were attributed to inattention to detail during procedure reviews by the station and corporate radiation protection group.
Those problems were attributed to the lack of a thorough operational readiness review of the program by the licensee. Although QA audits of selected ele-ments of the radiation protection program were performed, they focused primarily on procedure compliance and not on program adequacy. While a limited operational readiness assessment of station radiation protection program adequacy was performed by the corporate radiation protection group, the assessment find-
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ings were not tracked to resolution or verified closed by the
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corporate group.
The licensee initiated aggressive action to I
resolve subsequent NRC findings.
The findings were priori-tized and contractor support was obtained to assist in their resolution.
Despite the number of findings, the licensee was able to adequately resolve them to the satisfaction of the NRC prior to issuance of the low power license.
In order to fur-ther upgrade the program, the licensee, af ter issuance of the low power license, initiated a contractor review of the entire program to identify other weaknesses. The findings are tracked by computer to resolution and monitored by management.
The effectiveness of this review has yet to be verified.
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A contributing factor to the lack of adequate program develop-
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ment was a reorganization of the station radiation protection group which resulted in the loss of some key supervisory personnel and the lack of a fully staffed corporate radiation l
protection group.
The losses adversely affected the corporate group's capability to provide normal program development
support. At the close of the assessment period some posi-tions remained vacant and administrative procedures had not
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been updated to reflect revised reporting chains and personnel responsibilities.
Experienced contractor personnel were effec-tively used to augment the organization.
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the radiation protection program was not sufficiently challenged l
to allow NRC to fully evaluate oversight and control of in plant
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radiological work activities. However, limited NRC review of radiation protection technician performance in the field, and review of an unplanned exposure to the hand of a technician indicated weaknesses in the supervisory oversight of initial program implementation and the training program for some technicians. Also, the assignment of a junior technician to handle radioactive sources was considered inconsistent with the goal of assuring that personnel are assigned to tasks commen-surate with their training and experience.
A need to increase supervisory oversight of activities in the radiation protection area was evidenced by the following:
some technicians using improper meters to perform radiation surveys, inadequate documentation of radiation surveys, lack of consis-tent performance of surveys, and use of inadequate radiation work permits to control work with radioactive sources. The licensee initiated appropriate action to review and resolve the deficiencies associated with the identified problems.
Technicians were reinstructed regarding proper meter use and documentation of surveys, source control was tightened, and reviews of program implementation were initiated.
The train-ing program was permanently revised to address the identified problems.
In addition, supervisors were counseled and directed not to assign individuals, including junior technicians, to tasks for which they had not been qualified.
The special inspection to review implementation of NUREG-0737 post-accident sampling and analysis recommendations identified a number of problems requiring licensee attention. Although appropriate sampling and analysis equipment was installed and operable, and procedures were in place where needed, NRC review and observation during walkthroughs identified a lack of ade-quate field testing of procedures, weaknesses in training and qualification of personnel, and weak intragroup communications.
l The weaknesses identified did not preclude collection of samples but did delay their collection.
The licensee initiated aggres-i
sive and timely corrective action to address these NRC identified l
problems.
Regarding effluent monitoring and control, NRC review determined that the licensee's recovery from delayed installa-
tion / testing of the process and effluent monitors, resulting from the vendor going out of business, was well planned and executed.
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Reviews of the ALARA Program found that a management commitment to ALARA was evident.
In addition, state-of-the-art techniques are evaluated and adopted as appropriate.
Radiation protection l
personnel have been placed in the planning and scheduling group to provide for effective group interface and understanding of i
planned work. Although a basic ALARA Program is in place, pro-
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gram elements needing up grade were the ALARA goals program and on going job reviews.
These areas are being reviewed and eval-l uated by the licensee in response to NRC concerns.
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Reviews of radiation protection facilities and equipment found l
them to be of acceptable quality.
Radiation protection equip-ment was considered state of the art with ample supplies available. The supplies were adequate to support plant opera-tion, demonstrating adequate management attention to this important area.
Resolution of effluent sample line loss issues associated with the north and south plant vent monitors was delayed due, in part, to the resignation of the Senior Radiation protection Supervisor-Radioactive Material Control and the subsequent elimination of the position.
Personnel were unable to locate contractor line loss test reports and line loss test results were not reviewed, evaluated and incorporated into plant effluent surveillance procedures, demonstrating poor control of records and inadequate evaluation and use of test results.
The water chemistry control program was reviewed and found to conform to generally-accepted industry standards for controlling contaminant ingress, activated product tra pport, and corrosion of pressure boundary and heat transfer surfaces.
Radiological capability test standard intercomparisons showed all measure-ments to be in agreement.
However, comparisons of chemistry measurements 'or metals and boron were in disagreement and weaknesses in controlling, charting and trending chemical measurements were noted. Resolution of these technical issues was delayed, in part, by the resignation of the Chemistry Engineer.
Reviews of preoperational/startup testing of radwaste systems and initial implementation of the radwaste management program indicated that management attention was directed to developing, implementing, and maintaining a generally effective radwaste management program.
The licensen requested and received approval for deferral of test completion for the gaseous and solid radwaste systems into the startup phase.
Preoperational testing of the liquid radwaste system showed that the system was able to perform its intended function. Tests were completed in a timely manner and met generally-accepted industry standards l
for such tests.
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The development of the packaging and shipping program was delayed by discussions between the Hope Creek Generating Station and the Salem Station regarding a unified packaging and shipping program.
No radwaste shipments from Hope Creek t
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period.
In summary, NRC reviews at the beginning of the period identi-fled numerous programmatic deficiencies, particularly in the area of radiation protection.
These deficiencies were attrib-
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uted to lack of a thorough review of program operational readiness, reorganizations, and some staff vacancies. However, the licensee was able to prioritize the NRC identified problems and resolve them in a timely manner.
The remaining problems indicate a need to strengthen the internal audit program, sta-bilize the organization, fill identified position vacancies and improve inter-and intra group communication.
2.
Conclusion Rating:
Category 2 Trend: None 3.
Board Recommendations Licensee: None NRC: None l
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C.
Maintenance (5%, 445 Hours)
1.
Analysis The previous SALP evaluated the maintenance functional area as a category 2.
Noted strengths included the maintenance training program and experienced supervisors and managers.
The majority of weaknesses identified were associated with the transition from construction to operations, and the shift of equipment responsibility from Bechtel to PSE&G. The SALP Board recommended that this interface problem be resolved in order to prevent problems during the operations phase. Early in this assessment period, the station maintenance group assumed full responsibility for the maintenance of all equipment.
During this assessment period, NRC inspectors conducted admin-istrative program and procedure reviews, and observed a limited number of corrective and preventive maintenance activities.
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enced personnel although the use of contractors is still required to complete the required staffing in the instrument and controls (I&C) area.
There are approximately 60 personnel in the mechanical and electrical maintenance sections, all of whom are permanent PSE&G employees. Approximately one half of the 80 I&C personnel are contractors.
The reliance on con-tractors is being reduced as new hires complete their required training.
These staffing levels appear to be adequate for the plant work load since the number of outstanding corrective maintenance work orders is maintained at approximately 800.
Less than 10% of the outstanding corrective maintenance work orders would be categorized as safety related high priority.
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800 because all preventive maintenance (PM), and surveillance tests (ST), are also given work order numbers by the inspection order (IO) program. The 10 system appears to be an effective
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l management tool for the scheduling and tracking of periodic PM l
and ST requirements.
l The majority of maintenance department activity has been in the areas of minor valve repair, gasket leaks, early life failure l
replacements, preventive maintenance, and surveillance tests.
l Surveillance tests are further discussed in Section D.
The major activities observed during this assessment period include control rod drive (CRD) seal replacement, repair and replace-ment of service water elbows at the safety auxiliary cooling system (SACS) heat exchangers, and replacement of the B resid-
ual heat removal (RHR) pump.
Although these activities were t
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generally well controlled, some problems were identified. A lack of procedural adherence and a failure to satisfy the appropriate prerequisites was observed during the CR0 seal
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replacement. Also, the service water system was declared operable following reassembly, even though a deficiency report documenting questionable wall thicknesses had not been dispositioned. The licensee has taken corrective action for these problems, however, there has not been sufficient basis to evaluate their long term effectiveness.
The plant management meetings, shift turnover meetings, and use-of a planning department to prioritize and schedule all work activities has been an effective method of placing management's plan into action. The maintenance planners are responsible for developing a complete work package including special instruc-tions, procedures, tool and parts requirements, and retest requirements. This significantly reduces the administrative burden on the worker in the field and ensures a consistency among work packages.
Based upon a limited review in this area, good practices that have been noted are a comprehensive preventive maintenance program, the use of M0 VATS on all safety-related, motor operated valves, the incorporation of the operational experience evalu-ation program findings into procedures, and the development of a master equipment list.
In summary, based upon a limited amount of review, it appears that a good foundation of procedures and programs has been established in the maintenance area. Corrective actions have been taken for procedure adherence and operability determination problems which occurred early in the assessment period.
There has been limited activity in program implementation during the period and the organization has not been fully challenged.
2.
Conclusion Rating: Category 1 Trend:
None 3.
Board Recommendations Licensee:
None NRC: Maintain normal inspection activity, i
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D.
Surveillance (9%, 823 Hours)
1.
Analysis
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The surveillance area was not evaluated during the previous assessment period.
Surveillance tests performed by the
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licensee are the responsibility of several. departments, depending on the surveillance. The operations, maintenance, chemistry,-and site protection departments participate in.
surveillance testing. This section addresses surveillance tests performed without reference to the particular department involved.
Surveillance activities were routinely witnessed by NRC inspectors.
Because of problems encountered with the review of preoperational test packages, an increased emphasis was placed on the technical adequacy and performance of sur-veillance tests during this assessment period. The surveillance
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program is a well defined, computer based system that utilizes technically adequate procedures.
The use of the computerized inspection order (IO) system for scheduling all periodic sur -
veillance tests allows for efficient and generally effective-management oversight of the approximately 5000 surveillance tests performed on an annual basis.
Prior to the initial entry into each reactor operational con-dition, the completion of mode change required surveillance tests was frequently the critical path. Test progress normally i
lagged the schedule for a number of reasons including:
- Not all surveillance procedures were fully written and-approved before needed.
- Technicians were not familiar with all procedures.
- Time delays for equipment failures were not factored into the schedule.
Of the 89 reportable events during this assessment period, 26 are associated with the performance of surveillance tests.
Deficient surveillance procedures resulted in, or contributed to, the July 25 scram on low water level and the November 14 scram on high pressure. Schedule pressure and technician-unfamiliarity with surveillance procedures contributed to many of the reportable events and to an NRC concern regarding the l
use of unauthorized temporary procedure changes which altered the intent of the procedure but had not been Station Operations Review Committee (SORC) approved.
Based upon recent perform-ance, these problems have been corrected.
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Six instances of a failure to perform required surveillance tests or take the action required by the appropriate technical specifications were identified.
The lack of effective communi-cation between the operations and chemistry departments-has caused failures to obtain and analyze a number of samples required by the technical specifications.
It is noted that many of these samples are situational in nature and cannot be placed into the normal scheduling program. The licensee recagnizes that a problem exists and has taken steps to improve the situation. There has not been sufficient basis to evaluate the effectiveness of the corrective actions.
On numerous occasions during this assessment period, a single channel loss of coolant accident (LOCA) signal was generated from a not always apparent cause.
It appears likely that some of the LOCA signals resulted from valve operations on or around the reactor pressure and level instrument racks which feed the reactor protection and emergency core cooling system logic schemes. However, because the exact cause for all of these LOCA signals could not always be positively determined, the licensee formed a task force to identify the root cause of these LOCA signals. The investigation included a review of all available data. Although no positive determination could be made of the cause for the signals, a comprehensive action plan was-carried out. These actions included: blowing back all instrument lines to remove entrapped air, installing identification tags on all LOCA/ECCS instruments and sensor lines, installing quick disconnects on LOCA/ECCS instruments, technician training, review of all LOCA/ECCS surveillance procedures, and installing cages around instrument racks.
These actions have apparently been effective since no spuri-ous LOCA signals have been generated during the last four months of the appraisal period.
Regarding effluent monitoring and control, our review found that
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the licensee's recovery from delayed installation / testing of the process and effluent monitors, resulting from the vendor going out of business, was well planned and executed. However, the simultaneous need for preoperational testing of the monitors and continuous surveillance of those monitors to support early operation led to occasional lapses in Technical Specification surveillance tests. On two occasions, effluent monitors were removed from service and necessary grab samples were not taken resulting in self-identified failures to meet Technical Speci-fication surveillance requirements. The failure to ensure adequate communications among the various testing, operations and technical support groups, and to clearly assign responsibil-ity for declarations of operability /inoperability, contributed to the problems noted.
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The licensee implemented adequate local _ leak rate test-(LLRT)
and containment integrated leak rate test (CILRT) programs.
l The tests were. conducted using acceptable procedures and equipment, and the test personnel were knowledgeable and well-qualified.
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In summary,- the majority of difficulties experienced in the sur-veillance area can be attributed to the self-imposed schedule pressures associated with the plant entering the startup' phase of testing. The procedures and administrative controls in place are adequate to implement an effective surveillance program.
Increased attention is needed to improve communications between-departments in order to reduce the number of missed nonroutine surveillance tests.
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Conclusion i
Rating: Category 2 i
Trend: None 3.
Board Recommendations-Licensee: None
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None NRC:
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Emergency Preparedness (5%, 454 Hoursj 1.
Analysis During the previous assessment period, the licensee was eval-uated as Category 2 in the area of Emergency. Preparedness.
That assessment was based on -the results of an Emergency Preparedness Implementation Appraisal (EPIA) conducted on August 12-16, 1985, observation of the annual exercise held on October 29, 1985, and two routine inspections.
Several critical emergency planning (EP) program areas were _ determined to be incomplete and indications were that management attention had been diverted from Hope Creek EP capabilities develo'pment to (i) upgrading the Salem EP program and (ii) corporate reorganization.
The licensee's performance during the October 29, 1985 exercise was good with only a few weaknesses noted.
During this assessment period, there were two inspections.
One inspection was a follow-up emergency preparedness inspection conducted February 3-6, 1986, of concerns identified during the August 1985 EPIA. All but two of the concerns identified during the EPIA had been resolved. One unresolved item related to incomplete emergency preparedness training since sufficient numbers of personnel had not been qualified to provide a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> emergency staffing capability. The licensee committed to com-plete key personnel training and provide qualified staff prior to exceeding 5% power. The licensee affirmed,-in writing, on April 8, 1986, that training had been completed and would be maintained.
Full staffing capability was satisfactorily demonstrated during the November. 12, 1986 full participation exercise. A second unresolved item involved-the radiation monitoring system (RMS). The installation, calibration, func-tional testing and operability of process and effluent monitors has now been confirmed by reactor health physics inspections.
The RMS computer links were completed and computer-capability demonstrated.
Functionality of the RMS during_ simulated emer-gency conditions was confirmed during the November 1986 exercise.
The second inspection included observation of the November exercise.
The licensee satisfactorily demonstrated the ability, within scenario limitations, to:
identify accident conditions; declare the correct emergency action level; notify governmental authorities; activate and staff emergency response facilities; take proper corrective actions; develop protective action recom-mendations; interface with governmental authorities including the NRC Director of site Operations; effectively plan recovery operations; and adequately provide measures to protect public
~l health and safety.
In addition, strong performance was noted in the areas of personnel exposure control ~and radiation surveys. No significant deficiencies were identified; and,-
overall licensee performance during the exercise was adequat.
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The Emergency Preparedness Manager resigned and has been replaced by a staff senior emergency planner promoted to fill the vacancy. The Artificial Island emergency preparedness staff which supports Hope Creek. consists of twelve professionals.
Management has provided appropriate support of.EP.
An Alert was declared.on May 2 when-offsite power was lost to the vital buses and again on July 6 when tampering was con-sidered a possible cause for a plant fire _ suppression system actuation.
In both cases, notifications were made promptly and the emergency plan effectively implemented.
The licensee has installed a state-of-the-art siren system to meet the requirement for an alert and notification system.
This system provides hard copy diagnostics of performance for any one or all sirens. Additionally, an advanced surface water clearing plan for Delaware River surface waters has been developed and was satisfactorily tested during the November 1986 exercise.
The licensee and the State of New Jersey have negotiated'an agreement whereby the State receives 10 CFR 50.72 notifications in the same time frame as'the NRC as well as the follow-up 10 CFR 50.73 Licensee Event Reports.
FEMA will complete its review of-the New Jersey State
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Radiological Emergency Response Plan for Artificial Island during 1987 to determine if approval is warranted per 44 CFR 350.12.
"350" approval has been given to the Delaware Plan, contingent upon a successful siren test.
In summary, the licensee has dedicated sufficient corporate management attention and resources to establish an effective emergency preparedness program.
Strong performance has been noted during events and drills.
2.
Conclusion Rating: Category 1 Trend: None 3.
Board Recommendations
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Licensee: None NRC:
None
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F.
Security and Safeguards'(4%, 348 Hours)
1.
Analysis During the previous assessment period, the licensee was evalu-ated as Category 1 in the area of Security.and Safeguards. The previous SAL. assessment of this area was based on reviews of pre-operational activity in the development of a site security program. The licensee was effective in:
integration with the Salem security program, resolution of outstanding issues, and training security personnel.
During this assessment period, the licensee completed both the integration of the Hope Creek facility security program with the Salem program and a majo'r upgrade to the security program that began several years ago.
That upgrade included a combined access-control facility, installation of an integrated security.
computer system and associated hardware, computerized access
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control devices, state-of-the-art assessment aids and new personnel search ~ equipment. Those extensive activities were-completed by developing and implementing plans in a comprehen-sive, well thought out and organized manner. Management attention and oversight of the program was evident throughout the period from the smooth transition and relatively trouble-
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free implementation of program changes. The licensee provided NRC with thorough ar.d clear progress reports and prompt noti-fications whenever changes.to the plans were necessary.
The licensee aggressively addressed previously issued NRC security related guidance during the development of the Hope Creek program. The licensee demonstrated a clear understanding of the. safety and safeguards issues and effectively applied Salem program experiences to the Hope Creek program.
Solutions to technical safeguards problems were sound, timely and conservative. Concerns ider.iified by NRC were. promptly and effectively resolved by the licensee in a competent manner.
The NRC Site Evaluation Team was able to review and certify the Hope Creek security program for implementation with minimal'
difficulty and delay due to adequate records and preparation.
Aggressive corporate management attention to the development and implementation of-the security program aided in NRC certification.
The licensee has been effective in fostering a highly professional attitude towards maintaining performance objectives of the NRC approved security plans by-continued and effective management. The performance of the-security systems and equipment has been sound and relatively trouble free since the initial startup period. This performance-results from the extensive design, procurement and engineering effort expended on program development.
To date, the impact of integrating Hope Creek into the Salem security program has been essentially unnoticeable when viewed from an NRC regulatory perspective.
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Corporate management's interest in establishing and maintaining a strong security program was further demonstrated by the high quality of security force performance indicated during a special NRC inspection of the security force training and quali-fication program. That inspection was conducted to determine the quality and effectiveness of the training program and to measure the ability of security personnel to carry out their assigned duties. The training is conducted by individuals who are experienced and competent in their field and who are assigned to security training only.
Training facilities have adequate classroom space and good training aids.
Lesson plans are well developed, thorough, and kept current through feedback from supervisory personnel who perform on-the-job surveillance of security personnel performance.
The results of the special inspection indicated that the security training program is broad in scope, of high quality, and administered in a highly profes-sional manner. The results indicated extensive corporate and onsite licensee management involvement in the training program as well as a strong positive influence on the part of the con-tractor's site management and supervisory personnel.
The licensee's security plans, procedures, and instructions are clear, concise and thorough.
Letters and reports submitted to NRC are also clear, promptly submitted, technically accurate, and seldom generate questions from the NRC.
The licensee's security management and contract security force supervisors display a very positive and conservative attitude towards plant security issues and compliance with regulatory requirements. These individuals are quick to understand issues that arise during simulated and actual security events and how those issues can impact on plant security.
The security program is strongly supported by the other plant operating divisions on site and frequent interface is evident.
The maintenance staff detects unacceptable conditions with security equipment, and then aggressively pursues corrective action before they develop into major problems. When minor problems were found during NRC inspections, security managers were most often already aware of them and were in the process of establishing corrective actions. This degree of cognizance is creditable to a strong internal audit and surveillance program and is further evidence of the licensee's desire to l
implement a high quality security program.
Security force personnel exhibit excellent morale because of their recognized and respected role onsite, the excellent support they are afforded by the management of all divisions and the quality of the equipment they have been provided. As a result, they carry out their assigned duties and responsi-bilities in a professional and dedicated manne.
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Corporate security management is actively involved in the Region.I Nuclear Security Organization and other nuclear
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industry groups engaged in security innovations and the development of security program standards. This is evidence.
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of' management support of the security program at a high-level in the licensee's organization.
To_ ensure continued effectiveness of the security program, the licensee conducts in-house surveillances to monitor the performance-of the security organization.
Experienced and
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knowledgeable personnel perform these surveillances and the findings are aggressively pursued to ensure prompt and effec-tive corrective action and feedback to the training program.
These surveillances are conducted in addition to the annual'
security program audit required by the NRC.
Housekeeping of the access control facility and other security areas is noteworthy. The general state of cleanliness demon-strates a high degreeaof pride and morale on the part of the
security force.
The licensee submitted two security event reports pursuant to 10 CFR 73.71(c)-during the assessment period. Both events were bomb threats that were adequately responded to_by the-licensee-and were subsequently determined to be hoaxes.
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During the assessment period, the~ licensee submitted a temporary change (TC) applicable to both the Salem and Hope Creek security plans.
This TC identified compensatory measures that would be implemented during modifications necessary to
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consolidate the Salem and Hope Creek protected area.
Prior to the submittal of this change, the licensee contacted Region I Safeguards personnel and requested a meeting onsite to review and discuss the modification plans. The resulting TC fully described the issues. The approach to and planning _for-this modification is another indicator of the licensee's commitment-to maintain an effective and high quality security program.
In summary, close licensee management attention to this area has resulted in an effective security program following a smooth transition period during which the Salem features were expanded to encompass the Hope Creek site.
2.
Conclusion
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Rating: Category 1 Trend: None
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Board Recommendations Licensee: None N_RC: Due to the hiring of a new security force contractor, maintain normal levels of inspectio.
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G.
Outages (NA)
1.
Analysis This. functional area was not evaluated during the previous assessment period. Additionally, because the plant was in the preoperational and startup testing phases of operation through-out the period, there were no typical outages. Significant maintenance activities were discussed in Section C.
For this reason, this assessment will focus mainly on outage management and organization and on the engineering support and planning organizations.
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The licensee has established a planning department that is responsible for the planning, prioritizing, and coordination of outage, as well as routine, work activities. The station's priorities are established at a daily plant management meeting.
This meeting is followed by a supervisor's plan of the day meeting where additional manpower, resources, and coordination details are discussed. All departments are represented at both meetings so that a complete picture of plant activities can be established.
The written plan of the day schedule is used as the basis for the day's activities and for leading the shift turnover meetings. This entire process fosters good inter-department communications throughout the nuclear department and has the potential of becoming an effective outage manage-ment process.
The licensee utilizes a station system engineer concept.
Each system engineer is assigned a limited number of systems and is responsible for remaining cognizant of all factors that could affect the system's operability. They have been found to be knowledgeable and were helpful to the various NRC team inspec-tions and the resident inspectors.
The systems engineers have done a noteworthy job in root cause determinations and in developing corrective actions for spurious LOCA/ESF actuations, feedwater control problems, and high drywell temperature problems.
The engineering staff has not consistently displayed the initiative to identify and resolve technical problems in the field.
For example, the Bailey 862 reliability program, increased vendor failure analysis requirements, and more capable test equipment were a result of the NRC's interest in this area rather than the nuclear engineering staff's initiative. Once problems are identified to the engineering staff, they have displayed the capability to provide adequate technical support. The use of an engineering department single point of contact, who attends the morning management meeting, has promoted effective communications between the Hope Creek operations and engineering staff.
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The Engineering Department has self-identified the need for further improvement and contracted for an independent evaluation of the department with the following results:
- The need to establish a performance measurement system
- The need for. simplification of management processes in the following areas:
Design Change Requests Management By Objective
Work prioritization Project tracking and control Procurement Decision making / communication
More effective resource utilization and, More engineering technology utilization The licensee has established a task force to address the above concerns and is currently taking trips to selected-utilities who have demonstrated proven performance in these areas. The task force members are conducting biweekly reviews with the Vice President - Nuclear.
Completion of the Engineering Department improvements is set for mid-February 1987.
In summary, the station's method of planning and prioritizing work has been effective throughout the power ascension program.
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The station system engineers have done a noteworthy job, however as_ identified by the licensee, the Engineering-Department needs to improve performance and responsiveness.
Due to the lack of normal outage activities, no rating has been issued in-this area.
I 2.
Conclusion Rating: No rating Trend:
None 3.
Board Recommendations Licensee: Meet with NRC to present the results of Engineering Department task force findings and plans for addressing-the findings.
NRC: None
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H.
Preoperational and Startup Testing (38%, 3478 Hours)
1.
Analysis
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The functional area of preoperational testing was evaluated to be Category 2 during the previous. assessment period. Weaknesses were identified in the area of preoperational test procedure scope, content, adherence and. level-of review and in the area of.
i system turnover from construction to the startup group. The
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SALP Board recommended that the applicant be especially thorough in the, review of preoperational test results and the overall-program and that the applicant ensure that modifications and
. shifting of test commitments do not negate the results of com-pleted preoperational tests.
During the current assessment period, the licensee achieved a -
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number of major milestones. These included the completion of system turnover from construction, the completion of the pre-
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l operational testing program, initial fuel loading, initial criticality and the completion of a significant fraction of the startup testing program resulting in the initial attain-
ment of rated power operation.
Inspections of preoperational
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and startup testing activities were conducted by region-based and resident inspectors. Concerns identified by the inspectors involved inadequate test procedures, failure to follow test s
i procedures, inadequate review of test results and failure to comply with a license condition when making changes to the l
test program.
The preoptrational testing program was judged to be technically and administratively adequate. Senior management personnel were I
appropriately involved in activities and pursued an aggressive
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schedule for completion of the program.
Staffing levels were i
maintained throughout the program and' test personnel'were judged
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to be qualified and adequately traintd.
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While improvement was noted from the last assessment period, weaknesses in the system turnover process continued to impact preoperational testing. To meet schedule milestones, systems were turned over from construction to the startup group with-
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numerous incomplete work items. The. performance of preoper-
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ational tests on these incomplete systems required extensive changes in test procedures and the documentation of many test
exceptions. One example of this process involved the station
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service water preoperational test.
The approved test results contained 225 test exceptions, of which 62 remained open at the time the test results were accepted.
The numbers of test
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changes, test exceptions, and retests made-the determination of the adequacy of the preoperational testing difficult.
In-response to these concerns, the licensee agreed to perform j
Technical Specification surveillance testing and forego credit-for completed preoperational tests.
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On August 8, 1986, a reactor shutdown was initiated and an unusual event declared after it was determined by the licensee's staff that the reactor building to suppression chamber pressure relief system was inoperable and the plant was operating in violation of Technical Specifications.' The subsequent investi-gation into this event determined that a design drawing error made in 1983 during plant construction caused this system to be inoperable.
In the event that a vacuum was created in the sup-pression chamber, the butterfly isolation valves in series with each vacuum breaker would have remained shut, which would have prevented the vacuum breakers from fulfilling their design safety function. This condition had existed since 1983, and remained undetected until August 8, 1986. The licensee's corrective action included a design change to correct the design drawing error, a review of plant systems for similar
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design errors, a verification of valve lineups on all reactor building systems and a review of all temporary modifications.
No similar problems were identified.
Administrative deficiencies related to the accelerated test schedule were noted by the NRC.
Instances of procedure noncon-formance, unauthorized deletion of quality assurance mandatory witness points, and inadequate review of test.results were identified during the NRC review of the test results of pre-operational tests SN-1, automatic depressurization system, and
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GT-1, drywell ventilation.
In response to these identified deficiencies, the licensee committed to improve their review
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process and added an additional level of review.
This addi-tional review was in the form of a test review board that performed an independent technical assessment of procedures and results. While improvement in this area was noted, adminis-trative control of the preoperational testing program continued to be an area of concern for the balance of the program.
i The startup testing program was well-defined, technically comprehensive and adequately managed with appropriate adminis-trative controls. The licensee assigned the daily management, planning and conduct of the startup testing program to the NSSS vendor (General Electric). An aggressive schedule for startup testing was established and vigorously pursued during the assessment period. Staffing levels were ample and test personnel were experienced and well qualified. Technical training for test personnel was extensive and included use of the plant simulator for performance training on selected procedures.
Lack of senior management involvement in the day-to-day conduct of the test program is viewed as a weakness.
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Early in the startup testing program, deficiencies similar to those encountered during the preoperational testing program were identified by NRC inspectors. An approved startup test procedure for initial criticality, Full Core Shutdown Margin, contained numerous deficiencies including incomplete pre-requisites and initial conditions, inadequate precautions and limitations, and a technically incorrect method of assessing the acceptance criterion for reactivity anomalies. Testing activ-ities during the initial criticality were poorly coordinated and resulted in the failure to follow approved procedures. The approved test results for. Full Core Shutdown Margin were found to contain calculational errors in the determination of reactiv--
ity limits for control rod withdrawal.
The licensee's response to these findings was swift, thorough and effective in preventing.the recurrence of these problems.
To ensure the adequacy of startup test procedures, a Technical
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Review Board (TRB) was instituted ~and assigned to. review all
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test procedures. The TRB proved to be an extremely effective tool and a significant improvement in the overall quality of test procedures was noted.
Coordination of testing activities
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during major evolutions was significantly improved by the i,
increased involvement of the senior nuclear shift supervisors in
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the planning and preparations for testing. Additional personnel
were assigned as independent reviewers of test results to ensure
that adequate time was available to perform thorough reviews.
On September 11, 1986 the Loss of Offsite Power (LOP) test was terminated early when a number of systems functioned improperly.
Among the problems noted were the failure of one' diesel gener-i ator output breaker to close automatically, the loss of power to the acoustic monitor system, the failure of the reactor building ventilation system and the loss of the reactor auxil-iaries cooling system and the emergency service air compressor.
Following the test, the licensee initiated an extensive investigation into the causes of the identified deficiencies
and took corrective actions to resolve these problems. On September 19, 1986 the licensee performed a noncritical LOP test to verify that the corrective actions taken had been successful in resolving the problems. All problems identified during the initial LOP were verified to have been corrected, however, additional problems were encountered that were not identified during the initial test due to short time span of that test.
Following the noncritical LOP test, the NRC dispatched an Augmented Inspection Team to review the licensee's actions and to follow the resolution of all identified problems. On October 2, 1986 the noncritical LOP test was repeated and all test acceptance criteria were verified to be satisfied. A final demonstration of satisfactory integrated plant response
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- was performed on October'11, 1986 when the critical LOP test was successfully repeated.
One concern identified by the Augmented Inspection Team involved the adequacy of the licensee's administrative controls for mak-ing changes in the startup testing program. A license condition requires that changes in the startup testing program be made in
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accordance to 10 CFR 50.59.
Follow-up investigation revealed that management had failed to establish adequate administrative controls in this area.
Licensee's corrective acticn on this finding included a detailed review and revaluation of all startup testing program changes and the issuance of an admin-istrative procedure to control any future changes.
In summary, senior station management involvement in the startup testing program was judged to be adequate. While they were always very responsive to NRC concerns and consistently insti-tuted effective corrective action, their failure to incorporate the lessons learned in the preoperational testing program into the startup testing program resulted in similar deficiencies and NRC findings early in the program. One significant strength noted was the performance of operations shift personnel during i
the startup testing program.
They consistently maintained a cautious, conservative attitude toward plant operations and testing activities. They were in clear control of all station activities and contributed significantly to the successful
coordination of major testing activities.
2.
Conclusion Rating: Category 2 Trend: None 3.
Board Recommendations Licensee:
None
NRC: None
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Licensing Activities (NA)
l 1.
Analysis During the previous assessment period, the. licensee was.evalu-ated as Category 2 in the area of Licensing Activities. The previous SALP assessment was based on the resolution of SER'open-issues, numerous NRC audits conducted at licensee facilities, and the licensee's responses to NRC initiatives. ~The assessment noted that the licensee should have been more aggressive in the resolution of certain SER issues and should have been more timely in the resolution of those issues in light of. the fuel load date.
This was the first complete SALP cycle in which the Vice President-Nuclear and his staff have been responsible for all site activities. Corporate management has exhibited strong involvement and control in Hope Creek licensing activities.
At every major onsite meeting during the SALP cycle attended by NRR, the Vice President-Nuclear actively participated.
Similarly, the Vice President-Nuclear and his staff travelled to NRR offices in Bethesda, Maryland on. numerous occasions to participate in meetings important to the licensing of Hope Creek.
Overall, management has exhibited strong involvement in licensing activities for both special events and day-to-day activities.
PSE&G appeared very motivated in producing quality responses to staff questions and exhibited evidence of prior planning in producing responses to NRC concerns.
In most cases, responses are sufficiently complete and timely.
Decision making is done at a level which assures adequate management review. Management involvement is evident in PSE&G's responses to staff concerns as most responses indicate awareness of policy, design and opera-tional considerations.
During the current rating period all outstanding SER issues were resolved, a number of exemptions to the regulations were processed and granted, a compressed power ascension test pro-gram was proposed and submitted to the NRC, and-the low and full power operating licenses were issued.
In addition, following licensing, a number of Technical Specification amendment-requests have been submitted.
In all cases, the licensee has exhibited a clear understanding of the issues involved as exemplified by the licensee's effort to " compress" the power ascension test program.
For each test, the Itcensee identified the purpose of the test, the proposed modifications, and pro-vided safety evaluations supporting the requested modifications.
During various conversations with the licensee regarding the proposed modifications, the licensee exhibited a very clear understanding of the issues involved. Similarly, the licensee exhibited clear understanding of the issues involved when it
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submitted various exemption requests.
Each request was accom-panied by a detailed safety evaluation in support of the request, and the necessary findings under 50.12a.
In each case, it was the licensee's responsibility to demonstrate to the staff's satisfaction the acceptability of the proposed action.
The licensee did so with clear knowledge and full understanding of the issues at hand and their implications on plant operations.
Conservatism is routinely exhibited by the licensee when the issue involves safety significance. Most of the licensee's submittals have exhibited careful forethought, consideration of the proposed action, and technically sound responses.
In most cases, technically sound resolutions are proposed initially; however, during this rating period, one example exists where this was not the case.
In this instance, involving the testing of Bailey 862 solid state logic modules (SSLMs), the licensee proposed removing a fixed number of module sample popu-lation on a regular basis for testing during power operations.
Following discussions with the staff, the staff and licensee both agreed that this was not an acceptable test method, and the proposal was superseded.
In this instance, the licensee appeared overeager to resolve the NRC concern without assuring 1.tself that a safety concern did not exist. Overall, however, sound resolutions are initially proposed.
In most cases, PSE&G was responsive to staff initiatives. With the exception of not submitting the detailed control room design review Summary Report II on the schedule required by a license condition, most submittals met the deadlines. The licensee has provided timely responses to a number of Generic Letters during -
this rating period.
PSE&G appears to make special~ effort:, in resolving issues in a timely fashion, and with full knowledge of the issues at hand. The licensee's responses are usually tech-nically sound and thoroughly presented and supported. -In the few cases where the licensee has not provided sufficiently detailed responses, upon notification of this, the licensee has been very responsive in supplying the needed additional information.
Usually this evaluation is provided within twenty-four hours. As noted earlier, acceptable resolutions to issues are initially proposed in most cases.
Positions in the Hope Creek organization,' including senior-level management, are well defined.
Positions and their associated responsibilities are accurately described in the FSAR and Emer-gency Plan and appear to be consistent with actual practice.
Since the last SALP cycle, PSE&G has filled the vacancies that existed in the organization. The staff has reviewed the quali-fications of the individuals filling the previously vacant
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positions and found them acceptably qualified. The licensee i
has maintained a substantial and knowledgeable licensing staff to assure timely and quality responses to NRC concerns.
In conclusion, corporate management is taking a very active role in licensing matters and. responses to NRC initiatives continue to be timely, thorough, complete and conscious of safety impacts.
2.
Conclusion Rating: Category 1 Trend: None 3.
Board Recommendations Licensee: None
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J.
Training and Qualification Effectiveness (NA)
1.
Analysis During this assessment period, Training and Qualification Effectiveness is being considered as a separate functional area for the first time. Training and qualification effectiveness continues to be an evaluation criterion for each functional area.
The various aspects of this functional area have been considered and discussed as an integral part of other functional areas and the respective inspection hours have been included in each one.
Consequently, this discussion is a synopsis of the assessments related to training conducted in other areas. Training effec-tiveness has been measured primarily by the observed performance
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of licensee personnel and, to a lesser degree, as a review of program adequacy.
The licensee operates and maintains well equipped training facilities which provide training for all of the nuclear departments including operations, I&C technicians, electricians, mechanics, chemists, health physics technicians, machinists, and welders. The Hope Creek training program is modeled after the Salem program which has been INPO accredited in all ten training areas.
The NRC administered two sets of initial operator licensing examinations at Hope Creek during this SALP assessment period (February 1986, and July 1986). A total of 25 Senior Reactor Operator candidates were examined with 22 passing.
Weaknesses identified during the oral examinations included an unfamiliarity with the flow signals to the APRM/RBM systems, ADS logic, and fire protection equipment.
It was also noted that several candidates had a fundamental misconception about the operation of the feed water control system (FWCS).
Two unplanned reactor scrams, early in the power ascension pro-gram, were a result of control room operator errors.
In both cases, an IRM range switch was incorrectly downranged resulting in an IRM-high trip. A difference between the simulator and the as-built feedwater system may have contributed to two other scrams. The simulator does not accurately reflect the as-built condition of the feedwater turbine reset logic and the actions required to reset the turbine from the control room. Because of these differences, the operators were slow to recover a tripped feed pump and the reactor scrammed on low level.
Prompt cor-rective action in the form of shift briefings was taken and simulator upgrades are planned.
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Strengths observed during the oral examinations included the candidates' familiarity with safety and major systems (with the exception of the FWCS). Also, most candidates displayed a responsible attitude toward their duties as licensed operators.
A weakness in the ability to interpret and apply the Technical Specifications was noted during the grading of many of the SR0 written examinations. Also identified were weaknesses involving the response of the FWCS (as mentioned above) and fire brigade manning exemptions.
The Hope Creek full scope simulator is performing well and is providing a valuable tool for licensed operator training. The simulator was also used to perform validations of all major power ascension tests prior to actual in plant performance.
This significantly improved the quality of power ascension test procedures and provided valuable training to both operators and test engineers.
The plant operators, in general, have positive attitudes towards the training program. They felt they have been ade-quately trained on plant systems and system operations. They also feel the lecture and simulator programs are excellent.
Although varying opinions were observed as to the technical adequacy of the written training material, it was agreed that the readability of these materials could be greatly improved.
Based upon discussions and direct observation, the performance of licensed operators in the control room has been observed by the NRC to be excellent.
The operators are proficient in recovering from plant transients and equipment malfunctions in a competent and professional manner and have demonstrated a consistently improving knowledge of Technical Specifications as evidenced by daily discussions with NRC inspectors.
Knowl-edge of system operational characteristics, familiarity with procedures, and actions on transient response were noted, and are indicative of effective and valid training for licensed operators.
The licensee's corporate and station management involvement in training is good. Training review groups evaluate training on a regular basis and provide feedback to the training program.
The training department is well staffed with experienced personnel.
Laboratory facilities are excellent and provide hands on train-ing on such things as rebuilding circuit breakers, Limitorque valve operators and motors. The 2 year assignment of licensed operators to the training department is also a positive feedback mechanism.
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The training program has been responsive to the requests of various departments on a timely basis. When it became apparent that valving errors by I&C technicians were causing spurious
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LOCA signals, the training department set up a training instru-
ment rack on site and orovided training to all technicians.
This training directly contributed to a reduction _in spurious i
signals.
In addition, a modified SR0 training program is pianned for personnel designated as system engineers.
It
appears that these changes will have a positive impact on the performance of the engineering support groups.
Regarding training and qualification of radiation protection personnel, a documented training and qualification program
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for radiation protection personnel has been established and implemented. The program consists of formal classroom and
on-the-job training. The program is not yet INp0 accredited and is based on a job-task-analysis for the Salem Station.
Some findings this. period (e.g., lack of adequate training for individuals handling sources and improper use.of radiation
survey _ instruments) suggest a need to perform a specific job
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task analysis for radiation protection personnel at Hope Creek
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and an upgrade of the program as appropriate. The licensee is j
planning to do this as part of efforts to become INP0 accredited in this area.
Management's interest in establishing and maintaining a quality security program was demonstrated by the high quality of secu-rity force performance indicated during a special NRC inspection of the security force training and qualification program. That
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inspection was conducted to determine the quality and effective-ness of the training program and to measure the ability of
security personnel to carry out-their assigned duties. The training is conducted by individuals who are experienced and
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competent in their field and who are assigned to security train-ing only. Training facilities have adequate classroom space and
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good training aids.
Lesson plans are well developed, thorough,
'
and kept current through feedback from supervisory personnel who perform on-the-job surveillance of security personnel performance. The results of the special inspection indicated that the security training program is broad in scope, of high quality, and administered in a highly professional manner, i
Also, the results indicated extensive corporate and onsite licensee management involvement in the training program as well as a strong positive influence on the part of the contractor's site management and supervisory personnel.
'
In summary, based upon the high examination pass rate and operators performance in the control room, the licensed i
operator training program is effective. Problems encountered
during plant operations were due to inexperienced personnel more than training inadequacies.
_
.
.
.
.
.
..
l 2.
Conclusion Rating: Category 2 Trend:
Improving 3.
Board Recommendations Licensee: None NRC: None
,
.
.
_______m
_.
_
._
.
.
.
.
K.
AssuranceofQuality(NA)
1.
Analysis Assurance of Quality is a new separate functional area for this SALP period and is a summary assessment of management oversight-and effectiveness in implementation of the quality assurance program and administrative controls affecting quality.
Activities affecting the assurance of quality as they apply specifically to a functional area are addressed under each of
,
the separate functional areas.
Further, this functional area
is not an assessment of the quality assurance department alone, but is an overall evaluation of management's initiatives, programs, and policies which affect or assure quality.
During the assessment period, four inspections were performed in the area of quality programs and administrative controls affecting quality. These inspections covered the following areas:
- Administrative procedures, records, design control and modification, review committees and staffing and
'
nonlicensing training for operations;
- Bulletins and Construction Deficiency Reports (CDRs); and
- Licensee actions concerning the Salem ATWS event.
In addition, the implementation of the Quality Assurance (QA) program was reviewed by the resident and region based inspectors in conjunction with other functional areas.
Overall, the licensee appears to have. developed a strong program for asstring quality during operations. The licensee established a generally effective program for ensuring the timely issuance of the plant administrative procedures.
These procedures are well written, complete and meet the FSAR commitments.
<
The operating experience evaluation program's review of over 3000 industry documents from the NRC, INPO, and vendors has had a positive impact on the quality of plant procedures.
In addition, the incident report program provides a rigorous mechanism to ensure that Hope Creek's own operational experi-ence is evaluated and changes made to procedures when required.
All occurrences meeting certain criteria, whether reportable to the NRC or not, are documented and investigated.
Each dispo-sition is performed by the appropriate work group and includes l
I the corrective action taken or planned.
Station management is required to review and approve the disposition of all incident reports.
-
-
-
-
-
-
-
- -
-
.
.
- -
.
.
-
[
/
-
.
In the design change and modification area, the licensee has ade major organizational changes with respect to engineering s port for plant operations. A new engineering manual has be developed that is a distinct improvement on previous pro dures.
In the rea of review committees, careful forethought and plannin by management in the establishment of the various committe is evident. The Station Operations Review Committee (SORC) has been extensively involved with the preparations for operations ince it became functional in July 1984.
Since then, the li p ge has made significant changes in the SORC review proces W enhance the quality and timeliness of com-mittee reviews.
er strengths include the Offsite Safety Review Group ini
've to be in the online review of pro-posed design chan s/ modifications which exceeds 10 CFR 50.59 requirements.
The ite Safety Review Group performed a timely and in-depth
~ ew of the reactor building to suppression chamber p re relief system inoperability.
The licensee has implem an effective program, with adequate staffing to fol NRC bulletins, circulars, w-information notices and C s.
The evaluation, analysis and resolution of problems and nitiatives have been effective and timely.
In the area of licensee actions con rning the Salem ATWS l
event, licensee management has b gressive in taking an n
active part to assure that the AT sue receives proper emphasis.
This aggressive approac i dicated by the Vice President - Nuclear's letter to stat'
ersonnel regarding
" Commitment Management," and by licen eh ocedures which have been implemented including the followl
eliability and Assessment Management, Response Coordin ion, Vendor Interface and Reliability Monitoring. The license has established a Response Coordination Team which is respon ible for review, approval and implementation of all vendor s plied information, regulatory bulletins, industry standards, en ineering recommen-
dations, and operational experiences as appli ble.
j
.
During this assessment period, the implementatio of the QA program was judged as generally very good.
Stron points observed during review of other functional areas i luded extensive QA review of preoperational test results d excel-lent surveillance coverage of the containment integra ed leak rate test (CILRT). One weakness concerns timeliness o addressing quality concerns. QA had previously identif da deficiency involving the use of unapproved temporary pro
-
dures for performance of surveillance tests, however, the practice was not corrected until an NRC inspector identifie the same concern.
Upon subsequent review, the NRC found tha
.
.
.
.
_ _ - _ _ _ -
l
-
46A
.
In the design change and modification area, the licensee has made major organizational changes with respect to engineering support for plant operations. A new engineering manual has been developed that is a distinct improvement on previous procedures.
In the area of review committees, careful forethought and planning by management in the establishment of the various committees is evident.
The Station Operations Review Committee (SORC) has been extensively involved with the preparations for operations since it became functional in July 1984.
Since then, the licensee has made significant changes in the SORC review process to enhance the quality and timeliness of com-mittee reviews. Other strengths include the Offsite Safety Review Group initiative to be in the online review of pro-posed design changes / modifications which exceeds 10 CFR 50.59 requirements. The Offsite Safety Review Group performed a timely and in-depth review of the reactor building to suppression chamber pressure relief system inoperability.
The licensee has implemented an effective program, with adequate staffing to follow-up NRC bulletins, circulars, information notices and CDRs.
The evaluation, analysis and resolution of problems and NRC initiatives have been effective and timely.
In the area of licensee actions concerning the Salem ATWS event, licensee management has been aggressive in taking an active part to assure that the ATWS issue receives proper emphasis.
This aggressive approach is indicated by the Vice President - Nuclear's letter to station personnel regarding
" Commitment Management," and by licensee procedures which have been implemented including the following:
Reliability and Assessment Management, Response Coordination, Vendor Interface and Reliability Monitoring. The licensee has established a Response Coordination Team which is responsible for review, approval and implementation of regulatory bulletins and operational experiences as applicable.
During this assessment period, the implementation of the QA program was judged as generally very good.
Strong points observed during review of other functional areas included extensive QA review of preoperational test results and excel-lent surveillance coverage of the containment integrated leak rate test (CILRT). One weakness concerns timeliness of addressing quality concerns. QA had previously identified a deficiency involving the use of unapproved temporary proce-dures for performance of surveillance tests, however, the practice was not corrected until an NRC inspector identified the same concern.
Upon subsequent review, the NRC found that
.
- - -.
-
-
.
- a large number of QA identified concerns were not responded to in a timely manner by various departments.
The licensee has since increased the visibility of QA concerns and improved the timeliness of corrective actions.
.
The itcensee's philosophy on assuring quality at Hope Creek keys on individual achievement of a high level of performance, emphasizing personnel responsibility, accountability, and pride of ownership.
In keeping with this philosophy, programs to promote quality awareness and employee involvement have been instituted during this SALP period and appear to be well received by station personnel.
Examples of these programs are:
'
- Plant Material Improvement Programs which include cleanup, painting, and labeling activities in the plant.
I
- Employee Involvement Program facilitates management / worker interfaces and awards for good performance.
'
- Quality Awareness Committee comprised of nuclear department volunteers who periodically issue a " Quality Gram" promoting improvements in quality performance.
- Quality Awareness Days are sponsored by individual depart-ments and inform other departments of quality-improvement activities in progress within the sponsor department.
- Quality Concerns Reporting Program enables plant personnel to confidentially express quality concerns to be investigated by licensee QA personnel, t
Due to the low radiation and radioactive material source term, the radiation protection program was not sufficiently challenged to allow NRC to fully evaluate oversight and control of in plant activities. However, a need to increase supervisory oversight of activities in this area was evidenced by:
technicians repeatedly using improper meters to perform radiation surveys, less than adequate documentation of radiation surveys, lack of i
"
consistent performance of surveys, and use of inadequate radi-ation work permits to control work with radioactive sources.
Although corrected by the licensee when brought to his atten-tion, these examples demonstrate a lack of aggressive oversight of in plant activities during initial program implementation.
A combination of these weaknesses resulted in a technician
'
receiving an unplanned Lxposure of 1.4 rads to his hands.
Reviews of the external and internal exposure controls program prior to plant licensing found examples of deficient procedures being established and implemented.
Examples include a less
)
than adequate:
radiation work permit (RWP) program, high radi-
,
ation area access control program and airborne radioactivity
'
l
.
- -
-
--
--
- -
-
- -
-
-
-
. - -
-
...
. -
.
.
sampling and analysis program. Although corrected in a timely manner, these examples are indicative of lack of adequate attention to detail during program development and a lack of acceptable reviews.
Quality Assurance review of the technical program development and implementation of the radiation protection program at Hope Creek was limited. Technical evaluation of program procedures was conducted solely by the Station Operations Review Committee.
Less than adequate procedures were generated due to insufficient technical review.
Reviews of preoperational/startup testing of radwaste systems and initial implementation of the radwaste management program indicated that management attention was directed to developing,
.
implementing, and maintaining a generally effective radwaste management program. Application of the Quality Assurance program to preoperational tests of the radwaste systems was
'
thorough and demonstrated an effective' identification, track-
ing and closure of test discrepancies. A contingency plan for
'
processing solid radwaste was developed using vendor-supplied i
solidification equipment temporarily attached to the solid radwaste system. Vendor procedures were reviewed and incor-porated as controlled plant procedures and included inspection hold points and other controls governing the vendor's process control programs.
In summary, the licensee has established a generally effective program for ensuring quality. The operating experience evalu-ation program has had a positive impact on the quality of plant
'
procedures and management has frequently reinforced the role of the individual in assuring quality. However, increased station and corporate management attention is warranted in the radio-logical controls area.
2.
Conclusion Rating: Category 2 Trend: None 3.
Board Recommendations Licensee: None NRC: None
-
-
-
-
-
- -
.
.
V.
Supporting Data and Summaries A.
Investigations and Allegation Review No investigations were conducted during the assessment period.
Five allegations were received during the assessment period.
Hiring impropriety
-
Crack or scratch in a main steam isolation valve (MSIV) poppet
-
assembly.
Member of Safety Analysis Group does not have a degree.
-
Improper drawing control, retests after maintenance,
-
performance of preoperational tests, setup and calibration of radiation monitors.
Inadequate training in Chemistry Department, unqualified
-
supervisors.
All of the allegations were investigated and no significant safety issues were identified.
B.
Escalated Enforcement Actions On September 24, 1986, a Confirmatory Action Letter (CAL No. 86-12)
was issued to the licensee to inform them that an Augmented Inspec-tion Team (AIT) was being dispatched to the Hope Creek site to assess the anomalies related to the Loss of Offsite Power (LOP)
tests.
The. CAL also confirmed that the licensee would take the following actions:
-
Defer any additional LOP integrated testing until the NRC AIT team leader determines that such testing can continue.
Provide any LOP test procedures to the NRC AIT for their review
-
prior to implementation.
Make available to the NRC AIT relevant written material related
-
to deficiencies identified during the LOP tests conducted on September 11 and 19, 1986, including:
preoperational test results surveillance test results
- component installation and function test records
- - -
.
-
-
_ m..
y_
..-
_ _
.
._
__
.
_.
.... _. ~ _ _. _ _
.. _ _ _ _ _ -.
h,
^
s
.
-
50.
o
'
' Provide a written report to the Regional. Administrator.. prior.to
-
restart that includes an analysis of theiLOP testing conducted i
on September.11 andjl9; 1986 Receive' Regional Administrator authorization for unit startup.
-
l'
On October 7,.1986, the CAL was modified' to allow a' plant startup in order to conduct a reactor critical LOP.-,The : CAL was further
'modified on 0ctober 16 to allow.lfmited continuation of the power
'-
ascension. test program. Based upon.the-AIT findings,. licensee-commitments made.in an October 15, 1986 meeting,~ relating to Bailey 862 modules,'and discussions between NRC Region I a'nd-PSE&G'
.
on October 17, 1986, a. letter terminating the~ CAL was-issued on
!
October 21.
!
.
-On. November.17, 1986, an' enforcement. conference was held to-discuss
'
t'
. design deficiencies. identified during the ; LOP test,~. Regulatory Guide-1.97 instrumentation, and the inoperability of the Reactor Building-
~
.
~to Suppression Chamber Pressure-Relief System.
Enforc'ement: action
.
was under review at the conclusion of?the assessment' period.
)
C.
Management Conferences February 27, 1986: SALP management meeting
,
-
March 10, 1986:
NRC/ Region I - PSE&G readiness for fuel
-
l load
j March 11, 1986:
NRC/NRR.- PSE&G readiness for fuel load
-
j June 5, 1986:
SpuriousLECCS actuations,' management'
-
i.
changes, -lessons learned at similar plants
!-
during startup, control of work practices July 21, 1986:
Commission meeting for' Hope Creek-full
-
-
power. license July 24, 1986:
Corrective action program to. prevent
-
,
j spurious ESF actuations i-i
-
September 19, 1986:' LOP Test results F
October 15, 1986:
LOP Test results and Bailey 862. modules
-
-
'
November 17, 1986:
Enforcement Conference, Design deficiencie's,
-
LOP, Vacuum breaker operability, RG 1.97
}
instrumentation c
!
!
,
l.
!
e L
_ _ _ _ _
-
D.
Licensee Event Reports (LERs)
1.
Causal Analysis Eighty nine LERs, numbered 86-01 through 86-89, were reported during this assessment period. These LERs are characterized in Table 1 by cause for each functional area. Three common causal chains were identified, a.
Emergency Core Cooling System (ECCS) Actuations
,
Nineteen LERs (354/86-2, 86-7, 86-10, 86-14, 86-19, 86-20,
!
86-21, 86-23, 86-24, 86-33, 86-39, 86-41, 86-42, 86-43, 86-46,86-53,86-54,86-59,86-61) describe actuations of the ECCS due to low reactor vessel water level signals.
!
Seven ECCS actuations occurred as a result of personnel l
l error while conducting surveillance tests and nine have unexplained causes.
Investigations eventually discovered that ECCS initiations could result when personnel in the drywell stepped on or bumped reactor vessel level instru-ment piping. While it could not be positively determined that this explanation applied to all unexplained ECCS actuations, the licensee has concluded that it is the most probable cause.
b.
Surveillance Testing Fourteen LERs (354/86-8, 86-9, 86-17, 86-20, 86-21, 86-33, 86-38,86-43,86-52,86-53,86-57,86-62,86-87,86-89)
describe I&C technician personnel errors.
Seven LERs (354/86-2,86-6,86-13,86-15,86-49,86-55,86-66)
describe events initiated due to I&C procedural errors.
Seven of these LERs initiated ECCS equipment and are identified in Section V.D.1.a. of this report.
Actuation of Control Room Emergency Filtration (CREF) System c.
Eight LERs (354/86-12, 86-16, 86-17, 86-25, 86-36, 86-47, 86-74,86-75) describe inadvertent actuations of CREF.
Six CREF actuations occurred due to drift of the high voltage power supply to the ventilation duct radiation monitors. One actuation resulted from an I&C technician error during a surveillance test and another actuation was a result of a design deficiency. The licensee has replaced all high voltage power supplies which have caused inadvertent CREF actuations with an upgraded model.
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....
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2.
AE0D Review The Office for Analysis and Evaluation of Operational Data (AEOD) assessed fifteen of the LERs submitted during the assessment period using a refinement of the basic methodology-presented in NUREG-1022, Supplement 2.
The results of this evaluation, which was sent to the licensee by letter dated January 9,1987, indicate that Hope Creek has an overall LER
. score approximately equal with the industry average.
The principal weaknesses identified in the LERs, in terms of safety significance, involve the requirement to provide identi-fication of failed components and the requirement to discuss the safety consequence of the event..The failure to adequately identify the manufacturer and model number of the components that fail prompts concern that others in the industry won't have immediate access to information involving possible generic problems. Deficiencies in the safety assessment discussions cause concerns about whether the potential safety consequences of each event are being identified and evaluated.
A strong point for the Hope Creek LERs evaluated is the discus-sion of the mode, mechanism, and effect of failed components.
t-
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_ _ _ _ _ _ _ _ _ - _ _ _
_
..
'
'
TABLE 1 TABULAR LISTING 0F LERs~ BY FUNCTIONAL AREA HOPE CREEK GENERATING STATION (November 1, 1985 - November 20,1986)
Area Cause Code A
B C
D E
X TOTAL A.
Plant Operations
6 2-
12
8.
-Radiological Controls
1
-- 5 C.
Maintenance
1
8 D.
Surveillance
1 9-
27 E.
Security and Safeguards G.
Outage H.
Preoperational and Startup Testing I.
Licensing Activities J.
Training and Qualification Effectiveness K.
Assurance of Quality
1
3 Other
1 Totals
10
18
89 Cause Codes:
A.
Personnel Error B.
Design, Manufacturing, Construction, or Installation Error C.
External Cause D.
Defective Procedure E.
Component-Failure X.
Other
-
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. _ _.
..-.
.
..
.
- TABLE 2 LER SYN 0PSIS Hope Creek Generating Station
'LER NUMBER EVENT DATE CAUSE CODE DESr' 'PTION 86-001 2/15/86 A
Damaging of "0" Diesel Generator 86-002 4/13/86 D
Inadvertent "B" Channel LOCA Signal During Surveillance Test-Performance 86-003 4/15/86 A
Inadvertent RPS Initiation During Performance of NMS Component Troubleshooting Activities86-004 4/16/86 E
Noncoincident Scram Signal
'
Resulting from Neutron-Monitoring System Component. Failure 86-005 4/16/86 A
FRVS Inoperability During Core Alterations86-006 4/17/86 D
Primary Containment Isolation Resulting From a Procedural
'
Inadequacy 86-007 4/20/86 X
B Channel Engineered' Safety Features Actuation
86-008 4/24/86 A
Missed Surveillance During Initial Core Loading Due to
!
Personnel Error-86-009 4/25/86 A
Inadvertent RPS "A" Trip System'
Initiation During Surveillance Testing 86-010 4/26/86 X
A Channel Engineered Safety Feature Actuation 86-011 5/02/86 A
Loss of Off-Site Power 86-012 5/4/86 E
Control Room Emergency Filtration
!
Actuation Resulting From i
Equipment Malfunction
,
-
. - - -.
-
,
,_..-.c,
...
,
~
e,,
-, - _
.,. -,,
-
.
-
- -.
..
.
Table 2'(Cont'd)
LER NUMBER EVENT DATE CAUSE CODE DESCRIPTION i
86-013 5/6/86 D
Inadvertent Isolation of RWCU System During Surveillance Test
'
Performance 86-014-5/6/86 X
D Channel Engineered Safety Feature Actuation 86-015 5/6/86 D
Spurious A' Channel LOCA Initiation i
86-016 5/8/86 E
A Control Room Emergency i
Filtration Initiation 86-017 5/9/86 A
Inadvertent Actuation of the "A'
Control Room Emergency Filter Unit During Troubleshooting 86-018 5/12/86 B
Failure of Service Water Strainers86-019 5/13/86 X
D Channel Engineered Safety Feature Actuation 86-020 5/15/86 A
D Channel Engineered Safety Feature Actuation 86-021 5/15/86 A
D Channel-En'ineered Safety g
Feature Actuation 86-022 5/16/86 A
Inadvertent Isolation of Reactor Water Cleanup System
86-023 5/19/86 A
B Channel Engineered Safety Feature Actuation 86-024 5/25/86 D
Inadvertent "D" Channel LOCA
'
Signal During Surveillance Test Performance I
86-025 5/30/86 A
Power Supply Trip Causes Control Room Emergency Filtration Chiller Activation 86-026 5/30/86 E
Automatic Start of a Control Room Chiller 86-027 6/2/86 B
Installation of Combustible Material
in the Traveling Screen Motor Room
)
>
_. - _
_ _ _., _ _ _ - -,, _.. _,. - _
.. -
_
.
+
Table 2 (Cont'd)
LER NUMBER EVENT DATE CAUSE CODE DESCRIPTION 86-028 6/7/86 X
Spurious Actuation of the "A" Channel of the Standby Liquid Control System
.
I 86-029 6/11/86 E.
Automatic Start of "B" Control Area Chiller 86-030 6/18/86 A
Automatic Start of "B" Control Area Ventilation Train
.86-031 6/29/86-A Reactor Scram Due to Personnel Error,in Ranging IRMS86-032 6/30/86 E
Initiation of Manual Scram for Troubleshooting of Reactor Manual Control System
- 86-033 7/3/86 A
Inadvertent "B" Channel LOCA Signals During Instrument Calibration Performance 86-034 7/12/86 E
Main Steam Isolation Valve Closure and Subsequent Manual Scram 86-035 7/4/86 E
Reactor _ Scram Signal Originating From The Neutron Monitoring L
System 86-036 7/7/86 E
Isolation of The "A" Control Room Ventilation Unit Due to Radiation Monitor Upscale Trip j
86-037 7/12/86 A
Failure to Comply With Technical j
Specifications Action Statement 86-038 7/13/86 A
Missed Channel Checks on Reactor I
Protection and Isolation
<
Actuation Instrumentation 86-039 7/14/86 X
"A" Channel LOCA Logic' Actuation 86-040 7/9/86 A
Inoperable RCIC Actuation Instrumentation
r
n
-
,-
-
.-n
. --
--~,-e
,.
-.
-
-, - - - -
.,
e-
.
.
._ _
_..
_.
-
- _ - - - _.. - _ - - - _ - - _ _ _ _ _ _.
Y
..;
-
,
Table 2 (Cont'd)
.
LER NUMBER EVENT DATE CAUSE CODE DESCRIPTION 86-041 7/15/86-X Inadvertent HPCI System Initiation
86-042 7/17/86 X'
Inadvertent HPCI-System Initiation
86-043 7/17/86 A
Inadvertent HPCI System Initiation.Due to an I&C Error
!86-044 7/25/86 E
Reactor Scram on Low Level Resulting from an EHC Transient 86-045 7/19/86 A
Reactor Scram Due to IRM Ranging i
Error 86-046 7/20/86 X
Inadvertent HPCI' System Initiation 86-047 7/29/86 E
Actuation of the Control Room Emergency Filtration System Due
.
To Radiation Monitor Spike j
86-048 7/30/86 E
Full Reactor Scram on Low Water Level
- 86-049 8/1/86 D
Missed Response Time Surveillance Due to Procedure Inadequacy 86-050 8/1/86 D
Reactor Water Cleanup System.
.
Isolation on High Differential i
Flow
,86-051 8/3/86 E
Reactor Water Cleanup Isolation
!
on Spurious High Temperature Trip 86-052 8/20/86 A
Violation of the Surveillance
'
' Requirements for the Suppression Pool Temperature. Monitoring System 86-053 8/4/86 A
"A" Channel LOCA Logic Actuation 86-054 8/4/86 A
"A" Channel.LOCA Logic Actuation
and Full Reactor Scram i
i
-
,
.., - -
,,
--,
,
..
,,,,
,,, - - -..,
,-
-- -
.-
.
-
..
.
-.
_.
. - -.
~
.
I
'
,
Table 2 (Cont'd)
.LER NUMBER EVENT DATE CAUSE CODE DESCRIPTION 86-055 8/5/86
Primary Containment Isolation Due
,
To Procedure Inadequacy 5'
f 86-056 8/8/86 A-Inoperable Reactor Building to
' Torus Vacuum Breakers i
86-057 8/8/86 A'
InadvertentActuationof'the"N" Channel NSSSS Isolation Logic-86-058 8/8/86 A
Failure to Sample Results in Technical Specification Violation-86-059 8/14/86 X
"B" Channel ESF Logic Actuation
-
86-060 8/16/86
Violation of Suppression Pool Level Technical Specification
.
l 86-061 8/22/86
Inadvertent HPCI System
Initiation 86-062 9/20/86 A
Failure to Satisfy TS j
Surveillance Requirement for Leakage Detection Monitors-
,86-063 8/28/86 B
ASCO Solenoid Valve' Air Supply-
Pressure Rating 86-064 8/31/86 A
Reactor Scram on Low Level
!86-065 9/6/86 X
Full Reactor Scram on Low Reactor
-
Water Level 3 86-066 9/7/86 D
Missed Surveillance: - Turbine Bypass Valve Testing 86-067 9/15/86 B
SRV Acoustic Monitors Inop:
'
Seals Missing-i
.
.86-068 9/17/86 A
Missed Surveillance: North Plant-
Vent 86-069 9/24/86
Reactor Scram - IRM/APRM
'86-070 10/22/86 A
"C" Core Spray Pump Discharge
Pressure Transmitter Isolated-
4
.. _. _.. - - -. - _ _
-
, _ _
-.-., _,
. _... _. -. -
-
, -.
--
.
- Table 2 (Cont'd)
LER NUMBER EVENT DATE CAUSE CODE DESCRIPTION 86-071 10/4/86
.B PASa Sample Valves Installed in Less Favorable Orientation 86-072 10/3/86 X
Inoperable Reactor Building Exhaust Radiation Monitoring Instrument 86-073 10/3/86 B
Electrical Penetration' Assembly Installation Error 86-074 10/2/86 A
Inadvertent Actuation of "B" Control Room Emergency Filtration Unit when Connecting a Recorder 86-075 10/5/86 B
Inadvertent Actuation of "B" Control Room Emergency Filtration Unit During Troubleshooting 86-076 10/5/86 X
Inadvertent Automatic Start of
"B" Emergency Diesel Generator 86-077 10/10/86 E
Inadvertent Isolation of Reactor Water Cleanup System 86-078 11/11/86 E
RWCU Isolation 86-079 10/19/86 E
RWCU Isolation on High Differential Flow 86-080 10/18/86 B
Full Reactor Scram on Low Reactor Water Level 3 86-081 10/19/86 E
Isolation of Reactor Cleanup 86-082 10/28/86 A
High Pressure Coolant Injection System Inoperative 86-083 10/30/86 E
ESF Actuation 86-084 10/30/86 A
North / South Plant Vent Monitors Inoperable 86-085 11/14/86 A
Reactor Scram on High Pressure 86-086 11/14/86 X
Reactor Building Ventilation Isolation 86-087 11/17/86 A
ESF-A Channel NSSSS Isolation 86-088 11/18/86
Loss of RHR Room Cooling 86-089 11/19/86 B
RWCU Isolation Due to Loose Wire
-
-
-
- -.
-
-
.
.
._
_
-.. _.
._
-
--
t'
.
.
!
TABLE 3-INSPECTION HOURS SUMMA'RY (11/1/85 - 11/30/86).
HOPE CREEK GENERATING STATION HOURS
% OF TIME
- L A.
Plant Operations................
3030-
-33
'
B.
Radiological Controls and Chemistry..
592
<
.....
C.
Maintenance...................
445
'
0.
Surveillance..................
823-
6
!
E.
Emergency ~ Preparedness.............
454
i F.
Security and Safeguards.............
348
'4-
G.
Outages.....................
N/A i
H.
Preoperational and Startup Testing..
3478
.....
l I.
Licensing Activities..............
N/A
!
J.
Training and Qualification Effectiveness....
N/A i
]
K.
Assurance of Quality...
N/A
...........
'
.(
i Total 9170 100
.
r 8,
-
.
!
-
.
.
-
.
-
.
.
.
.
.
,
,
-
-
.
-.
.... -
.
.
..
.-
,
<
,.
^
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,
'
..
-61'
-
.:
P I
!
TABLE-4
.
. ENFORCEMENT SUMMARY (11/1/85-11/30/86)'
~
Hope Creek Generating Station
'
1-T SEVERITY LEVEL 1-l AREA
2
4
DEV TOTAL
!
OPERATIONS
4 i
RAD PROTECTION
,
MAINTENANCE
i SURVEILLANCE.
4
'
EMERGENCY PREP,
!
SEC/ SAFEGUARDS l
OUTAGES TRAINING EFFECTIVENESS
'
LICENSING ASSURANCE OF QUALITY PRE 0P/STARTUP.
8
11-
<
i
- -
l TOTALS:
12
19
..
l
,.
i
t
1
,
i
.
i
,
'
J
.
.
t
-
..
-
.
. - - -
.
- - - -
- -
- -
-
-
-
-
- -
-
-
-
-
.
__
_
. _.
._
.
- f d
s TABLE 4 (Cont'd)
,
ENFORCEMENT SUMMARY I-INSPECTION V10L.
FUNCTIONAL j
REPORT REQUIREMENT LEVEL AREA VIOLATION
!
354/85-61 APPENDIX B
PRE 0P/
MANDATORY WITNESS POINT BYPASS j
12/01/85 01/12/86 STARTUP DURING PREOP TEST
354/85-61 APPENDIX B
PRE 0P/
INADEQUATE QUALITY CONTROL 12/01/85 01/12/86 STARTUP INSPECTION
!
I 354/85-65 APPENDIX J
PREOP /
VALVE IMPROPERLY OPERATED IN j
2/23/85 01/03/86 STARTUP PREPARATION FOR INTEGRATED LEAK
RATE TEST.
2 354/86-03 APPENDIX B
PRE 0P/
INITIAL CRITICALITY PROCEDURES 01/06/86 01/17/86 STARTUP 354/86-06 APPENDIX B
PRE 0P/
FAILURE TO FULLY TEST CORE SPRAY l
01/13/86 02/09/86 STARTUP LOGIC
'
i
)
354/86-10 APPENDIX B
PRE 0P/
BYPASSING OF MANDATORY WITNESS j
01/27/86 02/07/86 STARTUP POINTS.
i 354/86-10 APPENDIX B
PREOP /
INADEQUATE REVIEW OF TEST RESULTS.
!
01/27/86 02/07/86 STARTUP
i
!
354/86-20 TECH SPECS
OPERATIONS FRVS INOPERABLE DURING CORE 03/17/86 04/30/86 ALTERATIONS.
354/86-20 ' TECH SPECS
OPERATIONS MISSED SBLC SURVEILLANCE TEST.
i 03/17/86 04/30/86 f
-
,
l 354/86-30 LC0 3.0.4 &
OPERATIONS TECHNICAL SPECIFICATION VIOLATION:
i 3.7.4 RCIC INOPERABLE DUE TO NO AUTO-SWAP
]
06/10/86 07/14/86 0F SUCTION TO SUPPRESSION POOL.
.
!
354/86-32 APPENDIX B 5 SURVEILLANCE FAILURE TO PERFORM POWER ASCENSION
!
06/23/86 07/03/86 TEST IN ACCORDANCE WITH APPROVED j
PROCEDURES i
354/86-35 APPENDIX B 5 SURVEILLANCE USE OF A POWER ASCENSION PROCEDURE 1-07/07/86 07/24/86 INAPPROPRIATE TO THE CIRCUMSTANCES I
i i
t i
.
-
- -
-
-
- -
... -.
_..
-
-.
..-
...
.
.
_
..
--
- - -.
s.
I l
a i
.
f
,
'
TABLE 4 (Cont'd)
i
'
INSPECTION VIOL.
FUNCTIONAL REPORT REQUIREMENT LEVEL AREA VIOLATION 354/86-35 10 CFR 50
SURVEILLANCE FAILURE TO FOLLOW PROCEDURE AND i
07/07/86 07/24/86 FAILURE TO ADEQUATELY REVIEW TEST j
RESULT
!
354/86-41 TECH SPEC
PRE 0P/
REACTOR BUILDING / TORUS VACUUM
3.6.4.2 STARTUP BREAKER ASSEMBLIES INOPERABLE
'
08/13/86 09/02/86
,
,
354/86-41 TECH SPEC
PREOP /
ACOUSTIC MONITORS NOT POWERED
3.3.7.5 STARTUP FROM UNINTERRUPTIBLE SOURCE t
j 08/13/86 09/02/86 354/86-40 TECH SPEC
SURVEILLANCE UNAUTHORIZED OPERATOR AIDS j
08/12/86 9b8/86 354/86-48 TECH SPEC
OPERATIONS CORE SPRAY PRESSURE TRANSMITTER
,
l 6.8.1 ISOLATED 10/14/86 11/17/86
354/86-49 10 CFR 50
PREOP /
FAILURE TO FOLLOW PROCEDURE FOR
10/11/86 10/16/86 STARTUP TORQUING ROSEMOUNT TRANSMITTER
!
l 354/86-53 LICENSE
PREOP /
FAILURE TO COMPLY WITH LICENSE i
NPF-57 STARTUP CONDITION C10 - DID NOT PERFORM l
10/27/86 10/31/86 TIMELY 50.59 REVIEW
!
!
i
~
I
- I
!
i i
i
!,
i
.
,
TABLE 5 INSPECTION REPORT ACTIVITIES (11/1/85-11/30/86)
Hope Creek Generating Station REPORT / DATES INSPECTOR HOURS AREAS INSPECTED 354/85-55 SPECIALIST 40 PREOP TEST PROGRAM 11/04/85 11/15/85 354/85-56 RESIDENT 251 ROUTINE RESIDENT INSPECTION 10/28/85 12/01/85 354/85-57 SPECIALIST 76 PRE 0PERATIONAL SECURITY PROGRAM REVIEW 11/12/85 11/15/85 354/85-58 SPECIALIST 829 AS-BUILT TEAM INSPECTION IN AREAS OF 12/02/85 12/13/85 MECHANICAL, ELECTRICAL, INSTRUMENTATION AND CONTROL AND STRUCTURAL SYSTEMS 354/85-59 SPECIALIST 72 PRE 0PERATIONAL INSPECTION OF CHEMICAL AND
11/18/85 11/22/85 RADI0 CHEMICAL MEASUREMENT PROGRAM.
354/85-60 SPECIALIST 44 PRESERVICE INSPECTION PROGRAM 11/18/85 11/22/85
-
354/85-61 RESIDENT 265 ROUTINE RESIDENT REPORT. MAJOR FOCUS ON 12/01/85 01/12/86 PRE 0P TESTING.
354/85-62 SPECIALIST 105 STAFFING, TRAINING, QUALIFICATION OF 12/09/85 12/18/85 PERSONNEL AND LOCAL LEAK RATE TESTING.
354/85-63 SPECIALIST 34 CONSTRUCTION PROGRAM 12/16/85 12/23/85
354/85-64 RESIDENT 300 TECHNICAL SPECIFICATION REVIEW CONDUCTED BY 12/02/85 12/13/85 PARAMETER INC.
354/85-65 SPECIALIST 115 CILRT INSPECTION 12/23/85 01/03/86 354/85-66 SPECIALIST 63 FOLLOWUP ON GENERIC LETTER 83-28, QA 12/30/85 01/03/86 RECORDS AND MEASURING AND TEST EQUIPMENT.
354/86-01 SPECIALIST 142 FIRE PROTECTION AND FOLLOWUP ON 01/07/86 01/11/86 CONSTRUCTION PROGRAM OPEN ITEMS.
,
354/86-02 SPECIALIST 146 PLANT PROCEDURES AND FOLLOWUP ON PREVIOUSLY
'
01/27/86 02/14/86 IDENTIFIED ITEMS.
!
.
..
.
..
.- -
..
. -.
.
.
-.
..
-
-_
m-
-
Table _5 (Cont'd)
REPORT / DATES INSPECTOR HOURS AREAS INSPECTED
'
354/86-03 SPECIALIST 151 PRE 0P AND POWER ASCENSION PROGRAMS 01/06/86 01/17/86 354/86-04 SPECIALIST 74 QA PROGRAM OVERVIEW 01/06/86 01/16/86 354/86-05 SPECIALIST 95 PREOPERATIONAL WATER CHEMISTRY CONTROL'
01/13/86 01/24/86 PROGRAM AND FOLLOWUP ON PREVIOUSLY
<
IDENTIFIED ITEMS.
.
i 354/86-06 RESIDENT 410 ROUTINE RESIDENT REPORT WITH EMPHASIS ON 01/13/86 02/09/86 PREOP TESTING.
,
<
354/86-07 SPECIALIST 50 RADIOLOGICAL CONTROLS INSPECTION
j 01/21/86 02/14/86 354/86-08 SPECIALIST 32 PRE 0PERATIONAL SECURITY PROGRAM REVIEW.
01/27/86 01/31/86 354/86-09 SPECIALIST 140 FOLLOWUP OF EMERGENCY PREPAREDNESS 02/03/86 02/03/86 IMPLEMENTATION APPRAISAL.
354/86-10 SPECIALIST 145 PRE 0PERATIONAL TEST PROGRAM IMPLEMENTATION.
-
01/27/86 02/07/86
'
i
354/86-11 SPECIALIST 44 RPV INTERNALS RECORD REVIEW i
01/27/86 01/31/86 PRESERVICE INSPECTION PROGRAM.
!
!
354/86-12 SPECIALIST 103 PREOPERATIONAL AND STARTUP PROGRAM 02/10/86 02/21/86 IMPLEMENTATION, 354/86-13 SPECIALIST 82 FOLLOWUP ON OUTSTANDING ITEMS AND t
j 02/10/86 02/14/86 MECHANICAL SNUBBER INSPECTION.
!
'
354/86-14 SPECIALIST 73 SAFETEAM INSPECTION
02/03/86 02/07/86 i
354/86-15 RESIDENT 501 ROUTINE RESIDENT INSPECTION WITH EMPHASIS 02/10/86 03/16/86 ON OUTSTANDING ITEMS FOLLOWUP AND
'
!
PREOPERATIONAL TESTING.
'
354/86-16 SPECIALIST 0 OPERATOR LICENSING EXAM.
)
02/24/86 03/24/86 i
354/86-17 SPECIALIST 36 MAINTENANCE AND I&C SURVEILLANCE l
02/24/86 02/28/86 PROCEDURES.
l
-
.
-
--
-
-
-
-.
- -
O
'
TABLE 5 (Cont'd)
REPORT / DATES INSPECTOR HOURS AREAS INSPECTED 354/86-18 SPECIALIST 75 PREOP, STARTUP, CILRT, AND SURVEILLANCE 03/03/86 03/14/86 TEST INSPECTION.
354/86-19 SPECIALIST 35 FOLLOWUP ON OPEN ITE 5.
03/03/86 03/06/86 354/86-20 RESIDENT 530 ROUTINE RESIDENT 03/17/86 04/30/86 354/86-21 SPECIALIST 108 PRE 0P AND STARTUP PROGRAM REVIEW.
03/12/86 03/21/86 354/86-22 SPECIALIST 128 INSPECTION BY 3 REGION-BASED INSPECTORS OF 03/31/86 04/11/86 PREVIOUS INSPECTION FINDINGS.
354/86-23 SPECIALIST 156 ROUTINE INSPECTION BY 5 REGION-BASED 04/14/86 04/25/86 INSPECTORS OF PREVIOUS INSPECTION FINDINGS 354/86-24 SPECIALIST
INSPECTION BY 2 REGION ?ASED INSPECTORS OF 04/28/86 05/09/86 PRE 0PERATIONAL TESTINL.
354/86-26 RESIDENT 271 ROUTINE RESIDENT INSPECTION 05/01/86 06/09/86 354/86-27 SPECIALIST 100 INSPECTION FINDINGS ON PREVIOUS 5/19/86 5/30/86 INSPECTIONS.
354/86-28 SPECIALIST 65 SECURITY INSPECTION OF TRAINING PROGRAM FOR 5/27/86 5/30/86 SECURITY PERSONNEL.
354/86-29 SPECIALIST 38 SPECIAL INSPECTION IN SUPPORT OF LICENSING 5/27/86 5/30/86 ACTION RELATED TO LICENSEE REQUEST DATED 5/13/86 TO DELETE FIRE PROTECTION TECH.
SPEC.
354/86-30 RESIDENT 353 ROUTINE FOLLOWUP INSPECTION.
6/10/86 7/14/86 354/86-31 SPECIALIST 75 INSPECTION OF PREVIOUS INSPECTION FINDINGS, 6/9 /86 6/20/86 POWER ASCENSION TEST PROGRAM.
354/86-32 SPECIALIST 94 INSPECTION OF POWER ASCENSION TEST PROGRAM 6/23/86 7/3 /86 COVERING INITIAL CRITICALITY 354/86-33 SPECIALIST 28 UNANNOUNCED INSPECTION OF RADI0 ACTIVE WASTE 6/16/86 6/18/86 (RADWASTE) PROGRAM DURING INITIAL FUEL LOAD ACTIVITIES.
.
..
..
_ _ _ _ _ _ _.
.
..
.
.
.
.
l
!
l l
- TABLE 5 (Cont'd)
REPORT / DATES INSPECTOR HOURS AREAS INSPECTED 354/86-34 SPECIALIST 0 OPERATOR LICENSING EXAMINATIONS 7/7/86 7/11/86 354/86-35 SPECIALIST 90 INSPECTION OF OVERALL POWER ASCENSION TEST 7/7/86 7/24/86 PROGRAM, QA/QC INTERFACES AND TOURS OF THE FACILITY
-
354/86-36 RESIDENT 204 ROUTINE RESIDENT INSPECTION 7/15/86 8/11/86 354/86-37 SPECIALIST 26 INSPECTION OF PREVIOUS FINDINGS IN 7/30/86 8/1 /86 RADIATION AREAS 354/86-38 SPECIALIST 45 POWER ASCENSION TEST PROGRAM, PROCEDURE 8/11/86 8/22/86 REVIEWS, QA/QC INTERFACES AND TOURS OF THE FACILITY.
354/86-39 SPECIALIST 36 INSPECTION OF RADI0 ACTIVE WASTE PROGRAM 8/12/86 8/15/86 354/86-40 RESIDENT 92 ROUTINE RESIDENT INSPECTION 8/12/86 9/8/86 354/86-41 RESIDENT 93 SPECIAL INSPECTION OF THE CAUSES FOR 8/13/86 9/02/86 INOPERABILITY OF REACTOR BUILDING TO SUPPRESSION CHAMBER PRESSURE RELIEF SYSTEM.
354/86-42 CANCELLED 354/86-43 SPECIALIST 31 INSPECTION OF OVERALL POWER ASCENSION TEST 9/2/86 9/5/86 PROGRAM.
354/86-44 SPECIALIST 72 ROUTINE INSPECTION OF SOLID RADI0 ACTIVE 9/08/86 9/12/86 WASTES (RADWASTE) PROGRAM DURING STARTUP ACTIVITIES.
354/86-45 SPECIALIST 154 INSPECTION OF THE LICENSEE'S IMPLEMENTATION 9/22/86 9/26/86 AND STATUS OF NUREG-0737 354/86-46 SPECIALIST 60 INSPECTION OF OVERALL POWER ASCENSION TEST 9/11/86 9/19/86 PROGRAM 354/86-47 RESIDENT 283 ROUTINE RESIDENT INSPECTION i
9/9/86 10/13/86 354/86-48 RESIDENT 195 ROUTINE RESIDENT INSPECTION 10/14/86 11/17/86
..
. -.
.
-
_
.. -.
.- -
'
a;
- --
4:
TABLE 5 (Cont'd)
.
. REPORT / DATES INSPECTOR HOURS AREAS INSPECTED 354/86-49 SPECIALIST 50 INSPECTION OF OVERALL' POWER ASCENSION TEST 10/11/86 10/16/86 PROGRAM
'
354/86-50 TEAM INSP 538 INSPECTION OF THE LOSS OF 0FFSITE POWER
'
9/25/86 10/3/86 TEST'0N SEPTEMBER 11, 1986 354/86-51 SPECIALIST 227 INSPECTION OF EMERGENCY PREPAREDNESS
-
11/10/86 12/1/86 PROGRAM AND IMPl.EHENTATION
i 354/86-52 TEAM INSP 344 OPERATIONAL READINESS TEAM INSPECTION
10/20/86 10/31/86 354/86-53 SPECIALIST 37 INSPECTION OF OVERALL POWER ASCENSION TEST 10/27/86 10/31/86 PROGRAM l
354/86-54 CANCELLED 354/86-55 SPECIALIST 48 INSPECTION OF OVERALL POWER ASCENSION TEST 11/10/86 11/19/86 PROGRAM
.
V l
'!
!
!
!
j
.
t
!
!
i
!
'
_ -,. - _
.- _._
_-
_,. _
-. _ _, - - _. _ _ _ _.. _
. - -. _.
. __
_
-
--
o C
TABLE 6 UNPLANNED AUTOMATIC SCRAMS AND SHUTDOWNS (11/1/85 - 11/30-86)
HOPE CREEK GENERATING STATION Root Functional Date Power Level Description Cause Area 1.
4/15/86 Shutdown IRM high scram due to Personnel Surveillance bumping IRM cable.
error 2.
4/16/86 Shutdown High APRM scram due to Equipment failure of 1 LPRM input.
failure -
random 3.
4/25/86 Shutdown IRM high scram due to Personnel Surveillance bumping IRM cable.
error 4.
6/29/86 less than 1% IRM high scram, caused by Personnel Operations downranging the wrong IRM error (non-coincident RPS mode).
6/29/86 Restart 5.
6/30/86 1%
Manual scram to trouble-Equipment shoot reactor manual failure -
control system.
random 7/1/86 Restart 6.
7/4/86 2%
High APRM trip due to a Equipment momentary upscale spike of failure -
an LPRM. A half scram was random already present due to inoperable instrumentation.
7/5/86 Restart 7.
7/12/86 1%
Manual scram after the Equipment MSIVs were automatically failure -
closed due to steam flow random transmitter drif t.
7/13/86 Restart
8.
7/19/86 less than 1% IRM high scram caused by Personnel Operations downranging vice upranging error 2 separate IRMs.
7/20/86 Restart
..-
---
o
TABLE 6 (Cont'd)
Root Functional Date Power Level Description Cause Area 9.
7/25/86 3%
Reactor vessel low level Inadequate Surveillance Scram caused by the loss of procedure feed flow after RFP trip on swell induced high level.
7/26/86 Restart
10. 7/30/86 6%
The EHC power supply failure Equipment caused the bypass valves to failure -
open.
The resulting swell random tripped the feed pumps and level could not be restored prior to the low level scram.
7/31/86 Restart 11. 8/31/86 5%
Reactor feed pumps tripped Equipment Preop /startup due to level control dif-failure ficulties as a result of inadequate minimum flow valve tuning. Operators were unable to reset the trip before low level scram.
9/1/86 Restart 12. 9/6/86 38%
Low level scram while Personnel Operations swapping feed pumps as a error result of unstable control of "C" RFP prior to completion of tuning.
9/6/86 Restart 13. 10/18/86 50%
Feedwater control test box Faulty test Preop /startup had internal wiring errors box wiring that caused RFP runback and low level scram.
10/19/86 Restart 14. 11/14/86 98%
High pressure scram due to Procedure Preop /Startup control valve closure test deficiency surveillance exceeding maximum combined flow limit.
11/28/86 Restart