IR 05000324/1987019
| ML20235G435 | |
| Person / Time | |
|---|---|
| Site: | Brunswick, 05000000 |
| Issue date: | 09/23/1987 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20235G429 | List: |
| References | |
| 50-324-87-19, 50-325-87-19, NUDOCS 8709300043 | |
| Download: ML20235G435 (43) | |
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l ENCLOSURE 1 SALP BOARD REPORT U. S. NUCLEAR REGULATORY COMMISSION
REGION II
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I SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE INSPECTION REPORT NUMBERS 50-325/87-19 50-324/87-19 CAROLINA POWER AND LIGHT COMPANY BRUNSWICK UNITS 1 & 2 NOVEMBER 1, 1985 THROUGH JUNE 30, 1987 ffg73OOO43syo993 O
ADDCK 05000324 PDR
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1 SUMMARY OF RESULTS A.
Overall Facility Evaluation During this assessment period, the licensee continued to maintain and improve on safe plant operations. Major strengths were identified in the area of security.
Major weaknesses were not identified in any functional area.
Operation activities were satisfactory; however, the units sustained a
number of scrams with complications which resulted in safety-related equipment failures.
Valve failures, during scram recovery, resulted in the loss of HPCI and degraded RCIC performance.
A valve maintenance and diagnostic program instituted toward the end of the assessment period should improve valve reliability.
Management has instituted several programs to address this problem.
Along with the valve maintenance program, a snubber improvement program has also been initiated.
Several preexisting equipment qualification problems were identified by the licensee.
Radiation
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protection activities continue to improve. Although Brunswick has reduced its total exposure significantly since 1985, it is still one of the highest exposure sites in the U.S.
Major programs are ongoing in the area of maintenance.
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Management provides " good support to the emergency preparedness program. Some weaknesses were noted in the annual exercise, however, the overall performance was satisfactory.
Safeguards measures were I
effectively planned and implemented. Outage scheduling and planning
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improved from the last assessment period. A Site Work Force Control j
Group was established to coordinate work and minimize system outage i
time.
Brunswick has improved the use of QA personnel resources by increasing the percentage of personnel perf ormi rig in plant inspections and QA surveillance.
CA/QC increased plant management
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attention to NCRs by issuing them to the plant manager or manager of construction and engineering, instead of the individual subgroup i
managers.
Licensing actions were generally satisfactory, with an
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improvement in LER quality. The licensee is developing expertise in using probabilistic risk assessment techniques to evaluate safety margins for plant systems.
Training improvements continued in maintenance.
All ten Institute of Nuclear Power Operation (INPO)
training areas completed accreditation at Brunswick.
Brunswick has i
been very responsive to NRC initiatives overall.
The licensee focuses on reactor safety and there were no violations that warranted escalated enforcement action during the assessment period.
The licensee has also been very cooperative in coordinating with the NRC the closecut inspection of outstanding inspection findings at j
Brunswick.
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FaciHty Performance Summary
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Th'e performance categories for the current'and previous SALP period in each functional area are as follows:
May-1, 1984-November 1, 1985-Salp Period Functional Area October 31, 1985 June 30, 1987 Ending Trend A.- Plant Operations
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B. Radiological Controls
2 Improving C. Maintenance
2 D Surveillance 2-
E. Fire Protection
2 F. Emergency Preparedness-
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G. Security.
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H. Outages
2 I. Quality Programs and Administrative Controls Affecting Quality
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J. - Licensing Activities
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K. Training
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g IV.
PERFORMANCE ANALYSIS A.
Plant Operations l
1.
Analysis During this assessment period, inspections were performed by the resident and regional inspection staffs.
Management initiated several new prcgrams to improve performance.
The operations decisions were virtually always-conservative and made with safety being paramount, The operations staff and technical support group were always responsive to NRC ~ initiatives.
No major Limiting Condition of Operation- (LCO) violations occurred during the report period.
Reactor scram frequency was near the industry average, but typically ' occurred with complications.
Procedural compliance and operator discipline showed some minor weaknesses. Technical support for operations was responsive when called upon.
The number of Licensee Event Reports (LERs) attributed to personnel errors in operating activities increased over the last SALP, period.
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Management demonstrated their involvement in assuring quality through several new initiatives.
New programs requiring corporate approval were initiated.
These new initiatives included:
operations staf fing plan, including the addition of several shift technical advisor positions; establishment of a Quality Team in operations (part of the corporate Total Quality Program); and full implementation of the Automated Maintenance Management System.
Procedures were implemented that required timely reviews by technical support of operability questions.
Initial reviews were timely, thorough and technically sound. On occasion, secondary followup on certain technical issues was slow and did not address all the issues, for example; the licensee identified problems with motor control auxiliary contacts in a timely manner, but was slow to completely resolve the issue. A trending program tracked LCOs and annunciators, among other parameters.
Procedures and policies were well defined in operating l
procedures and standing instructions.
However, the occasional lack of attention to detail or operations discipline continued from the last assessment period.
The number of LERs
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attributable to personnel errors in operations activities increased from 5 last assessment to 18 this assessment.
However, only some of those events had safety significance.
Certain program and staffing improvements were outgrowths of i
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management's attempt to resolve the personnel error issue.
- Included in their response was a plan to implement INPO's Human Performance. Evaluation System to better identify and - correct-human errors.
Oral and written communications to the NRC via the Emergency Notification System (ENS) improved toward the end of the assessment period.
However, weakness-was noted - in LER -
and ENS reporting during the first. half of the report' period.
Management was very responsive to NRC questions after events.
The' plant restart decision making process was improved using the.
Scram Incident Investigation Team (SIIT) and detailed review procedures.
The SIIT, comprised of about 10 senior staff'
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members from different plant groups, _has as its only responsibility, reviewing scram data to resolve any safety questions prior to s ta rtup '.
That process was effective, especially in resolving partial safety system initiations.
Operations staffing was ample.
Six shift rotations on 12-hour l
days provided time for training and relief when required. Shift j
staffing exceeded regulatory requirements, with eight licensed j
personnel assigned to each shift.
Only one licensed reactor
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operator and one licensed senior reactor operator left the site
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during'the' assessment period.
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Operations and technical support group performance demonstrated that the training programs had made positive contributions to the conduct of operations. A modest number of personnel errors demonstrated that the training program had not been completely effective in minimizing personnel errors.
The Real-Time Training program had been effective when given in a timely manner. On a few occasions, training had not been received on shift soon enough to prevent repeat events.
Plant housekeeping remained a strength and continued to receive significant management attention.
Post-outage cleanup was aggressive and systematic.
On infrequent occasions, debris and trash accumulated in locations not often frequented, especially contaminated areas.
Drywell close-out inspections and cleanup left the area clean for restart.
j Control room demeanor was professional. On-shift personnel were
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identified by name tags.
Pre-shift briefings and formal
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turnovers were conducted.
Shift foremen regularly conducted j
routine tours outside the control room. Some weakness was noted i
on board walkdowns and was documented in several resident inspection reports throughout the assessment period. Access to the at-the-controls area was well controlled.
Control room
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human factors modifications began at the end of the assessment
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Reactor scram data indicated that no widespread generic problem existed regarding plant performance.
The number of unplanned, j
automatic scrams per 1000 critical hours over the period of January 1986, through June 1987, was 0.72 and 0.49 for Units 1 and 2, respectively.
This compares to a national average of 0.53 over the same period for similar vintage plants.
Of the 20 scrams or reactor protection system actuations, five were due to equipment failure, six were due to personnel maintenance, four were due to personnel operations, one was due to personnel testing, one was due to design, and three were due l
to other causes.
On several occasions, plant safety equipment
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failed when called upon during a scram.
Operator action mitigated the equipment failures during several post-scram transients.
The staffing plan was well developed and implemented.
The number of licensed shift personnel was increased to eight people. A job rotation program was established.
As a result, five experienced SR0s were rotated from operating shift to staff positions to increase staff experience and provide additional career growth. STA staffing has. increased to eight, with five holding SR0 licenses.
Plant capacity factors increased from the last assessment-period.
Capacity factor' for Unit 1 increased from 36.7% to 65.7%.
Capacity factor for Unit 2 increased from 47.2% to 53.5%.
Management continued to work on the system engineer concept.
The system engineers suffered from understaffing and an unclear definition of their role. In plant walkdowns and training time had to be reduced to respond to the other demands placed on them. Management has formulated, but not yet implemented, plans to correct the problem.
Eight violations were identified:
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Severity Level IV violation for failure to implement j
annunciator procedure and failure to maintain corrective I
action associated with freeze protection of condensate storage tank level switches.
(325/85-40-01)
b.
Severity Level IV violation for failure to take adequate corrective action to maintain control of jumpers.
(325/86-15-01 and 324/86-16-02)
c.
Severity Level V violation for failure to maintain the diesel generator service water supply valve breaker energized in accordance with procedure. (324/86-15-01)
d.
Severity Level V violation for failure to utilize a procedure when manipulating a service water valve, thereby l
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leaving the. valve' in :a position other-than required by operating: procedure. (324/86-25-01)
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Severity Level V violation for failure to perform a -
required review or approval for a temporary change.to an l
operating procedure, thereby resulting in an inadvertent-Reactor.
Protection system bus
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(325/86-29-01)
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Severity Level V violation for failure to incorporate a design change in an annunciator procedure which resulted in the procedure not providing ' adequate -instructions to l
isolate the Control Room Ventilation system if called upon.
.(325/87-02-02 and 324/87-02-02)
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Severity Level V violation:for failure to make a report.per 10 CFR 50.72 within the required time. (324/87-11-01)
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Severity Level V violation for failure to complete restart authorization and documentation prior to unit ' restart f rom scram-.(325/87-17-01)
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Conclusion
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Category:
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Board Recommendations NONE
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B.
. Radiological Controls 1.
Analysis During this assessment period, inspections were performed by the resident and regional inspection staffs in the areas of worker dose control,' gaseous and liquid effluent treatment and monitoring, and water chemistry.
Inspections included confirmatory measurements using the Region II Mobile Laboratory.-
The licensee continued to improve the radiation protection program. Resolution of technical issues was a program strength.
An example of this was the facility staff handling of the radiological impact of hydrogen water chemistry (HWC) testing on Unit 2.
The licensee's evaluation of projected radiation levels and monitoring of levels during the test and control of the affected areas, helped identify an optimum injection rate to achieve HWC while'still maintaining acceptable exposure rates.
The licensee's evaluation, monitoring and control were thorough and well documented.
Licensee line management involvement in radiological controls.-
appeared adequate and the Radiation Protection Manager (RPM)
received the support of other plant managers in implementing he facility radiation protection program.
Various members o.
management were involved sufficiently early in outage preparation to permit adequate planning of the work and consideration of ALARA related issues.
f Audits performed by the corporate office and onsite audit organization were adequate and of sufficient scope and depth to identify problems in the health physics area.
The audits were conducted using staff personnel with appropriate technical
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background in radiological controls.
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The licensee's health physics staffing level was adequate and j
compared favorably with other utilities having a facility of j
similar size.
An adequate number of ANSI qualified licensee
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health physics technicians was available to support routine
operations.
During outage operations, additional contract technicians were used to supplement the permanent plant staff.
The staff turnover rate was low during the evaluation period and the quality and experience level of the health physics personnel was a program strength.
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radiation protection training programs were adequate.
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licensee's HPT training program was accredited by INP0 during
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the assessment period.
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Radioactive releases of fission gas activity to the atmosphere at the two unit Brunswick site during 1986 were 2.26 E+4 curies per unit. By comparison, approximately 8.75 E+3 curies per unit of fission gases were released during 1985.
Unit 1 had undergone an extended maintenance outage in 1985.
The comparable value for the average of 21 operating U.S. BWRs of greater than 500 MWE in 1982 (last year for which summary data were available) was 4.85 E+4 curies per unit.
Liquid effluent releases of mixed fission and activation products during 1985 and 1986 were 0.08 curies per unit and 0.07 curies per unit respectively. Tritium releases in liquids during 1985 and 1986 were 4.94 curies per unit and 2.89 curies per unit respectively.
By comparison, the annual liquid releases from 21 operating U.S. BWRs of greater than 500 MWE for calendar year 1982, were 3.56 curies of mixed fission and activation products. Effluent releases for the past three years are summarized in the Supporting Data and SummariesSection V.J..
No significant trends were noted.
The licensee's calculated offsite doses for 1986 from radioactive effluents were 5.9 mrem to the whole body.
These values place the licensee well within the limits of 40 CFR 190.10, 25 mrem to the whole body over any 12 consecutive months.
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The quality control program for radiological measurements met the criteria outlined in Regulatory Guide 4.15.
The gamma spectroscopy systems were calibrated properly. Licensee results for gamma-ray measurements of inplant samples, split with the NRC during 1986, were in agreement for all sample types.
The licensee's cumulative exposure for 1986 was 954 man-rem per unit as measured by TLDs.
Although the 954 man-rem figure was weil above the national average of 622 man-rem for GWRs, it represented a decrease from the 1985 total of 1402 man-rem per unit and a continued reduction from the previous assessment periods. A major contributor to this dose was the recirculation pipe inspection and repair (weld overlay).
The initial projected collective exposure for 1987 is 850 man-rem per unit.
Through June 1987, the total accumulated exposure was 565 man-rem per unit.
The high cumulative exposure for the facility in 1536 resulted from major maintenance and modification ac6ivities conducted during :'25 outage days.
The collective dose projections for 1987 indicated a continued reduction in exposures received.
However, if the 1987 cummulative exposure reaches the licensee's projected value, they will again exceed the national average by a significant amount.
The facility remains as one of the n.lants with the highest cumulative radiation dose in the U.S.
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During calendar year 1986, the licensee rJde 92 solid radioactive waste shipments totalling 16,900 cubic feet per unit
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p and containing 9,831 curies of activity.
This. tota 1' compares favorably with the! 19861 BWR national average which was 17,400 cubic-feet per unit.' Through June 1987, the licensee had made 51: shipments totalling 9578_ cubic feet per unit, containing
~.4,309 curies of 'olid radioactive waste. Based on the projected s
work load for the rem 61nder of 1987, the licensee's total should again be below the national. average: of waste shipped.
The radioactive waste ' precessing program resulted in one violation for failure to ensure that waste shipment containedless than.1%
. free-standing itquid and no unabsorbed oil.
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'InL the area _ of water chemistry, the licensee was co3sistently-maintaining ' chemistry control within the limits reco'mmended by-the.'BWR.0wners Group, whose criteria the licensee' had endorsed
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'through corporate and plant policy statements and administrative instructions.
The licensee maintained approximately 94,000 square feet of the radiological control area (RCA) under contamination controls at the end of'1986,'which represents 12% of the floor space _of the facility (including the: reactor i buildings, the auxiliary
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' building, the turbine building'and the.radwaste building). That total 'had been reduced to 91,000 square feut in early 1987;
tj however, due to the Unit 2 ' outage, 'approximately 9,000 squan,
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feet were added,to that total. Through June 1987, that area was again. reduced to 94,000. square feet under contamination controls.
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One violation was identified:
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Severity' Level IV ~ violation _ for. failure to ensure that a 1.
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waste _ shipment contained less't.han 1% free standing liquid J
and no unabsorbed oil as required by NQC regulation and the burial site license conditions.
(q25/86-20-01 and 324/86-21-01)
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Conclusion Category:
Improving i
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Board Recommendations-
Licensee performance in the > area of Radiological Controls continues to be Category 2 but with,'an improving trend during the SALP period.
The Board acknowledges reduced exposure from previous SALP perioils and encourages management to continue programs which achieved this reduction; NRC attention should be i,
maintained at normal levels.
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C.
Maintenance 1.
Analysis During this assessment period, inspections were performed by resident and regional inspectors. The Division of Human Facters Technology performed a maintenance program survey at the site in January 1986.
During the previous SALP period, maintenance management undertook z self-initiated Maintenance Improvement Program (MIP)
l to enhance work controls and scheduling, upgrade knowledge level of craftsmen and provide more management oversight and involvement.
The MIP was completed during the present SALP.
The MIP and other management initiatives resulted in several noteworthy contributions in the day-to-day functioning of the maintenance organization and corresponding increase in quality.
Examples of these include:
1) the automated maintenance management system which is a computerized work request generation, planning, scheduling and retrieval system; 2)
completion of phase one of the preventive maintenance improvement program which resulted in preventive maintenance,
tasks being restructured into system oriented work packages; 3)
formation of a site work force control group which preplanned system outages to allow for preventive and corrective maintenance; 4) Real Time Training program to provide feedback on maintenance related operational events at the site as we'l as industry experience; 5) assignment of a cognizant engineer to maintenance of major components such as the diesel generators, thus providing focus for all major work and troubleshooting efforts; 6) implementation of post-maintenance work inspections and maintenance program internal audits to supplement the existing work force audit program, thereby improving supervisor invvivement and oversight in all phases of work activities; 7)
incorporation of Environmental Qualification (EQ) requirements into maintenance instructions; and 8) formation of a dedicated procedure generation and revision group to upgrade existing maintenance instructions and to maintain a consistent quality of procedures.
Training activities have consistently received management support.
Technical knowledge and skill were expanded by utilizing cognizant vendors.
For example, prior to work on the main steamline isolation valves, the' valve vendor brought a valve to the site and provided instructions while craftsmen performed lapping of the seat.
The overall general knowledge level and skill increased; however, management identified a need to provide specialization and additional depth in specific areas.
For example, I&C personnel demonstrated only marginal familiarity with the turbine generator voltage regulator
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circuitry while developing corrective action to prevent malfunctions of the circuits which resulted in reactor scrams in September 1986 and January 1987.
Staffing levels were adequate to support all non-outage work and training activities. Staffing was supplemented during refueling
outages by corporate controlled traveling maintenance crews and by outside contractors for specialty work such as turbine overhaul.
The work force was stable, with a turnover rate of approximately one percent during the entire assessement period.
The total backlog time of corrective maintenance items was less than three weeks.
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The number of licensee identified minor violations and
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occasional weaknesses in work controls, as demonstrated by maintenance related Licensee Event Reports (LERs), internal review and audit programs and the workmanship issue described in paragraph IV.H, demonstrated that lack of attention to detail and poor work practices continue to contribute to personnel errors.
Technical issues were generally understood with resolutions usually viable and sound.
For example, concerning a problem with the main steamline isolation valve solenoids, management" assembled a teant to determine the cause and corrective action.
They demonstrated a clear understanding of the issue, promptly reported the event, accurately analyzed the deficiency and developed a technically sound resolution.
On occasion, technical resolutions of issues were resolved slowly due to a very conservative approach.
Two major efforts initiated by management to upgrade plant equipment reliability were the snubber improvement program and l
I the motor-operated valve maintenance and diagnostic programs.
To date, these efforts have been only partially successful. The
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licensee has established additional cleanliness controls in a l
dedicated snubber rebuild room.
An automated snubber tester with computer generated documentation is used during snubber testing. The licensee has also purchased a freon cleaning unit for the snubbers and has dedicated personnel assigned to the rebuilding task.
However, 100*!, functional testing has still been performed during every refueling outage as specified in Technical Specifications, primarily due to low bleed rate induced failures.
The additional cleanliness controls should reduce the minute debris that may be causing the low bleed rate failures.
Diagnostic testing by a Mechanical Actuator Characterized (developed by Babcock and Wilcox in conjunction with Limitorque Corp.), was performed on High Pressure Coolant Injection (HPCI),
Reactor Core Isolation Cooling (RCIC), and major containment
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4 isolation valves 'during the Unit I refueling outage.
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program experienced a number of equipment and technical problems which reduced the program's effectiveness Management demonstrated sound technical judgment in resolving the valve program's technical issues..Previously undetec*.ed problems, such as needed torque switch setting changes and replacement of spring packs, were identified and corrected.
Work was in progress at the end of the assessment period to resolve the remaining problems in the diagnostic test program. Also, during the Unit I refuel modification outage, completed in June 1987, safety related valves, selected for their safety importance, were completely overhauled.and rebuilt.
On
'mir own initiative, some safety-related valves were also pgraded to Category 1 EQ, exceeding DOR Guidelines.
Notwithstanding the above effort, the site.has experienced a number of
. safety related valve failures including failures which resulted in the loss of HPCI and degraded RCIC performance during scram recovery. Valve problems have been attributed to torque switch problems, incorrect and/or degraded spring packs, failed auxi'11ary contacts, motor failures and personnel errors.
However, as implementation progresses, the valve maintenance and diagnostic program should result in increased long term reliability of motor actuators and valves.
After extensive investigation, neither the licensee nor NRC has identified a -
single root cause for the valve failures.
The predictive maintenance program currently involves oil sampling analysis and vibration data trending on several major components such as HPCI and the diesel generator. Expansion of this program to include other items, especially balance of plant equipment, was being formulated. Other improvements to enhance equipment performance and reliability are planned, such as Phase 2 of the preventive maintenance improvement program and potential expansion of the predictive maintenance program.
Corporate support for maintenance improvement was reflected in the commitment to INPO to perform a self-assessment at all three nuclear plants in 1987.
The assessment is to be conducted in accordance with INPO 85-038, Guidelines for the " Conduct of Maintenance at Nuclear Power Stations."
The effort has been initiated at the site with the results expected this fall.
The Division of Human Factors Technology performed a maintenance program survey at the site in January 1986. The survey is part of the multi year Maintenance and Surveillance Program Plan to identify the technical and regulatory issues to be addressed, as well as planning of NRC's activities to accomplish these objectives.
The survey revealed that, of the 15 significant observations, most were positive with only portions of three being negative.
Management responsiveness to the NRC
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initiatives was demonstrated by initiation of action to enhance j
these areas.
j The licensee has implemented a new Maintenance Trending Analysis Program.
The program aids maintenance in identifying potential problem areas by allowing maintenance supervision to quickly note an increase in the use of a particular part or material.
Two violations were identified:
a.
Severity Level IV violation for failure to have an adequate freeze protection procedure and failure to install instrumentation bolting in accordance with drawings.
(325/85-40-02 and 324/85-40-02)
b.
Severity Level IV violation for failure to install anti-rotation devices in accordance with prescribed instructions. (325/87-02-05 and 324/87-02-05)
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Conclusion Category:
3.
Board Recommendations
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Licensee performance in the area of Maintenance continues to be Category 2 as rated during the previous SALP period. The Board recognized that the licensee has initiated and continued several aggressive maintenance programs.
The licensee should continue
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increased management attention in this area and the NRC
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inspection effort should be maintained at normal levels.
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D.
Surveillance 1.
Analysis During the assessment period, inspections were performed by resident and regional inspection staffs.
Procedure upgrade programs were completed and in place, implementing the operations and maintenance surveillance requirements.
Maintenance management retained 22 dedicated procedure writers to maintain the existing procedures.
Maintenance surveillance procedure development was in accordance with a writer's guide.
Each procedure included: detailed, unambiguous a:ceptance criteria, sign-offs for each step, separate sections for each instrument channel, and a guide for operations cor cerning the procedure, and other features to reduce human error. A dedicated staff of technicians performed all maintenance surveillance procedures to further reduce errors. Staffing was ample to conduct all required surveillance testing.
The scheduling and surveillance tracking systems provided the'
necessary controls to schedule surveillance testing. An updated comprehensive cross reference list is maintained comparing technical specification surveillance requirements and the implementing procedures.
The tracking system was being transferred to the plant mainframe computer at the end of the assessment period.
No missed surveillance tests were attributable to this system; however, several Licensee Event Reports showed that personnel errors resulted in a few late surveillance tests.
A repeat problem has been fire watches not being conducted as required.
The licensee performs approximately five thousand surveillance tests per unit per year.
During this assessment period two integrated leak rate tests (Type A) were observed:
Unit 2 in May 1986, and Unit 1 in May 1987.
In these tests, the licensee demonstrated management involvement as indicated by the program reviews, plant modifications and the use of consultants in the leak rate tests.
The licensee's staff demonstrated increased technical understanding, responsiveness to NRC concerns and the ability to plan, coordinate and control the many details involved in a complex leak rate test.
Also, procedures have been revised to be more specific.
As a result of licensee improvement in the above areas, improvement in the quality of leak rate test performance was observed. The numerous miscellaneous leakage sources throughout the plant, prevalent in leak rate tests prior to 1984, have been eliminated. Although both of the above tests
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- l failed to meet the integrated leak rate limits, the leakage was limited to a few specific sources, which were subsequently corrected, with the test rerun prior to restart of the unit.
The licensee was able to identify the leakage paths and to determine the cause of the leakage failure.
It is expected that trending and evaluation of this f ailure data will identify necessary corrective actions which should lead to further c-improvement in test results.
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One specific inspection was conducted on the snubber surveil-lance program. The snubber surveillance program showed evidence of prior planning with thorough, well-defined procedures.
Records were complete, well maintained, retrievable and legible.
The approach to resolution of problems encountered due to functional test failures was timely, technically sound, and thorough.
Staffing and training of personnel involved in the snubber surveillance program was adequate.
Decision making was usually at a level which ensured adequate management review.
Reviews of test results were timely, thorough and technically sound.
The corrective action system
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recognized and addressed non-reportable concerns.
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Two violations were identified:
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a.
Severity Level IV violation for failure to determine test data properly and inadequate test procedure for standby liquid control discharge flow test. (324/86-12-02)
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Severity Level IV violation for failure to implement integrated containment leak rate test procedure in that a drywell pressure instrument was not removed from service as required. (325/87-13-02)
2.
Conclusion Category:
3.
Board Recommendations NONE l
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E.
Fire protection 1.
Analysis During this assessment period, inspections of the licensee's
fire protection and fire prevention program were conducted by
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the regional and resident inspection staffs.
The licensee issued procedures for the administrative control of fire hazards within the plant, surveillance and maintenance of
,
'
the fire protection systems and equipment, and organization and training of a plant fire brigade.
These procedures were reviewed and found to meet the NRC requirements and guidelines except for the licensee's failure to properly identify and correct adverse plant housekeeping practices as identified in a violation (a.
below).
Otherwise, the program was being satisfactorily implemented.
The organization and staffing of the plant fire brigade met NRC guidelines. The training and drills for the brigade members met the frequency specified by the procedures and the NRC guidelines; however early in the period, some members of the fire brigade had not completed all quarterly fire brigade
training and the required number of drills.
The annual fire protection / prevention audit, the 24 month QA fire protection program audit by offsite organizations and the triennial audit by an outside fire protection organization required by the Technical Specifications, were reviewed.
These
audits were conducted within the specified frequency and
!
appeared to cover all of the essential elements of the fire protection program.
The licensee had implemented corrective action on discrepancies identified by these audits.
In general, the management involvement and control in assuring quality in the fire protection program was adequate due to the issuance and implementation of fire protection procedures that met the NRC requirements and guidelines.
The licensee's approach to resolution of technical fire protection issues indicated an understanding of issues. The responsiveness to NRC initiatives was above average.
Fi re protection related violations did not indicate a programmatic weakness. Corrective action was timely and effective.
Fire protection staff positions were identified and authorities and responsibilities were defined. These people were qualified
i for their assigned duties.
'
l Three violations were identified-
'
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a.
Severity. Level IV violation,for failure to properly-implement fire protection procedures to.. identify and ' take corrective actions on adverse housekeeping practices.
(325/86-03-02 and 324/86-04-02)
b.
Severity Level IV violation for failure to implement fire
~
protection procedure requirements for all. fire brigade members to participate in at least two drills'per year and quarterly training. in order to maintain fire brigade qualification. (325/86-03-01 and 324/86-04-01)
c.
Severity Level IV. violation for..f ailure to implement
. procedure requiring no smoking in the diesel generator building in that evidence indicated smoking had occurred-
-there. (325/86-17-02 and 324/86-18-02)
2.
Conclusion Category:
3.
Board Recommendations.
-NONE
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l F.
Analysis During the assessment period, inspections were performed by the'
resident and regional inspection staffs.
The. inspections included one routine emergency planning inspection and the observation of one full participation emergency exercise.
The routine inspection determined that: the licensee maintained a capability for notifying and communicating with required agencies and authorities during an emergency; made. changes to the emergency preparedness program in accordance with requirements, maintained an adequate emergency response training program (with one exception); disseminated emergency planning information to the public on an annual basis; and conducted an annual independent review of its emergency preparedness program.
The one exception to an adequate training program,'(referenced above), resulted in a violation for failure to provide first aid training to those emergency response personnel designated to provide this service.
The licensee committed to develop an adequate training ' program to provide first aid training to ' *
appropriate pers.onnel.
During -the annual emergency exercise, the licensee demonstrated the ability to identify promptly and classify correctly emergency events consistent with the Radiological Emergency Preparedness Plan and implementing procedures.
The licensee's emergency response personnel also demonstrated the ability to staff the emergency response facilities and to take required corrective actions to mitigate the simulated plant casualty.
Timely notifications were made to the State and local officials and proper protective action recommendations were made.
However, two exercise weaknesses and several other observations made during the exercise indicated that some improvements were required.
One exercise weakness was the failure of the
" Emergency Communicator" procedure to identify all parties on the automatic ringdown (ARD) telephone circuit. During initial l
use of ARD, an unlisted party began asking questions of the Control Room Communicator. As a result, the communicator became confused and momentarily delayed the Alert Notification.
The licensee acknowledged the weakness and is taking corrective actions to develop a different notification system called Selective Signal Systems.
The system will allow the communicator to select the parties for the notification circuit.
i Procedure revisions are under development to support this change. The other exercise weakness was the TSC's downgrading j
of the emergency classification without prior consultation with the EOF, and consideration of the effect of downgrading on the
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protective action recommendations that were being implemented by the State.
Other items needing licensee attention were a
- followup' message to the State and Local Governments being provided an incorrect' emergency classification, and the lack of adequate player input into the initial critiques of the exercise.
These areas will be reviewed. during the next
exercise.
However, the overall successful exercise performance was indicative of corporate management continuing to support an effective emergency response program.
One violation was identified
!
-j Severity Level V violation for failure to provide first aid j
training to those emergency response personnel designated to provide this service in Section 3.2.5 of the Emergency Plan.
(325/86-25-02 and 324/86-26-02)
2.
Conclusion Category:
3.
Board Recommendations NONE
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G.
Security 1.
Analysis During the evaluation period, routine inspections of security l
and safeguard activities were performed by regional and resident inspectors.
'
The licensee added two security specialists to the site security management staff to improve program management and oversight'of the contract security force. The staffing level of the contract force was adequate to accomplish their assigned responsibilities.. An effective audit program had been established as evidenced by documented audit findings and corrective actions by the security organization.
The licensee demonstrated an ability to plan and implement safeguards measures effectively.
Security plans and implementing procedures were current and provided ample guidance for effecting operational requirements. The plant maintained an exemplary law enforcement orientation program and was supported.
by an aggressive corporate program for auditing contractor personnel screening programs.
During the period, the licensee '
improved the site security program through the installation of new x-ray search equipment and security computers.
The licensee demonstrated responsiveness to NRC initiatives by enhancement of vital barriers with timely and appropriate actions taken in response to IE Notices. Actions in response to these items were documented and maintained in security organization files.
The operational capability of the security organization was enhanced by an effective training program. The effectiveness of the training and qualification program was evident in personnel performance and positive morale.
One violation was identified:
Severity Level IV violation for failure to alarm a portion of the Protected Area barrier.
(325/87-10-01 and 324/87-10-01)
2.
Conclusion Category:
3.
Board Recommendations NONE
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H.
Outages 1.
Analysis During the assessment period, Units 1 and 2 underwent lengthy refueling-modification outages.
Resident and regional based inspectors observed refueling and outage activities.
Management commitment to improved scheduling resulted in the 1986 Unit 2 outage exceeding the 196 day. critical path schedule by less than 2 days. The 1987 Unit 1 outage was achieved in 120 days, seven days less~than the critical path schedule.
I For Unit 2, cycle 7, pre-operational and startup tests were witnessed and/or reviewed by regional inspectors. Test methods, timeliness, and documentation were all acceptable.
Also, inspections included: review of procedures; observations of work activities; review of records and evaluation of data in the areas of inservice inspection' of safety-related components, associated piping and supports; inservice testing of pumps and
.
valves; nondestructive examinations; weld overlay repair,-
!
welding; induction heat stress improvement of the ~ recirculation and residual heat removal systems; and the replacement of the reactor water cleanup 304 stainless steel (SS) piping with nuclear grade 316 SS, which has low carbon content and high resistance to intergranular stress corrosion cracking (IGSCC).
Brunswick Unit 2 refueling outage seven started on December 1, 1985, and was completed on June 15, 1986.
Unit I refueling outage six started on February 14, 1987, and was completed on June 13, 1987.
Major activities conducted during these refueling outages consisted of:
100 percent inspection of Class 1, 2, and 3 safety-related
-
support on Unit 2 100 percent inspection of Class 1, 2, and 3 safety-related
-
bolting on Unit 2 115 Class 1 safety-related examinations on Unit 2
-
Generic Letter 84-11, Inspections of BWR Stainless Steel
-
Piping, activities on Units 1 and 2 Examination of the reactor vessel nozzle inconel butter
-
weld in accordance with IE Notice 84-41, IGSCC in BWR Plants, on Units 1 and 2 l
_ - _. - _ - _ -
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Local leak rate testing and containment integrated leak
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rate testing on Units 1 and 2 The licensee's overall control and planning for refueling outages resulted in a well planned and controlled. evolution.
Most work was planned with regard to scope, repair parts, and work procedures. At the conclusion of an outage, planning for the next refueling began.
The licensee's understanding and resolution of technical issues from a safety standpoint was demonstrated in their conservative approach to Generic Letter 84-11 and Information Notice 84-41 activities.
In addition to using the " state of the art" equipment and techniques to detect, size, repair, and mitigate IGSCC for Units 1 and 2, the licensee had weld reinforcements removed or re-surfaced when their presence prevented or limited high-tech automated ultrasonic examinations. CP&L corporate ultrasonic examiners were highly visible, performing surveillance inspections of vendors'
examination activities during both refueling outages. New calibration blocks that were more representative of the welds being examined were procured for the Inconel butter welds and the thermal sleeve to safe-end welds on the reactor vessel nozzles.
Contingency actions appeared to be excellent.
An example of this was the onsite staging of vendors and equipment, to install metal pipe clamps if IGSCC had been found in the
!
reactor vessel pozzle Inconel butter welds.
The licensee's interface and control of contractors had become stronger during refueling outages.
Procedures were reviewed more thoroughly, and vendors with advanced equipment and more depth in personnel qualifications were used, regardless of the cost factor.
This
!
was reflected in CP&L's vendor choice for the examinations j
performed in accordance with Generic Letter 84-11.
Surveillance, coordination and supervision of vendors had
improved notably since the previous assessment period, l
Completed reports submitted by vendors improved and the l
licensee's review of this material had, on occasion, discovered l
errors that demonstrated real growth in their technical i
understanding of the Code requirements and their concern for safety.
However, procedure and policies were occasionally violated as evidenced by the violations listed below.
l Violation (e)
involved procedure development; welding parameters used by the licensee to develop a generic welding procedure for overlay repair welding did not meet the design basis specification for the recirculation system.
Violation (f) involved failure of the licensee to develop adequate outage inspection plans for Class 1 and Class 2 safety-related supports and bolting on Unit 2.
This item also indicated how vendors had controlled CP&L's ISI activities in the past and how the licensee did not have sufficient technical expertise in the ISI area to realize ASME Code requirements were l
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not met during that period.
Discrepancies in this area were corrected during this assessment period by management strengthening of the ISI program with additional supervision, engineering, and surveillance personnel.
The ISI program is presently in the process of being transferred to a computer program for better control.
.
The plant modification process generaly showed evidence of prior planning with strict procedural control of the modifications and I
modification turnover processes. These were minor weaknesses in the modification development process occurred as indicated by violations a.
and c.
Personnel error due to inattention to detail contributed to violations b., d., and g..
A number of LERs were issued due to safety systems actuations caused by j
shorting of electrical equipment. An excessive number of minor fastener problems were found during the assessment period and documented in inspection reports. Most of the above problems involved no operability concerns but indicated a weakness in work quality.
Eight violations were identified:
a.
Severity Level IV violation for failure to perform an adequate modification acceptance test pertaining to the.
'
core spray. injection valve (324/86-12-01)
b.
Severity Level IV violation for failure to declare a support inoperable when a component was found missing.
(324/86-16-01)
c.
Severity Level IV violation for failure to ensure that a pressure switch installed in an environmental qualification application contained a conduit seal. (324/86-34-02)
d.
Severity Level IV violation for failure to have control rod drive pipe clamps installed in accordance with specifications. (324/87-02-01)
,
e.
Severity Level V violation for failure to maintain welding parameters within specification.
(325/86-07-01 and 324/86-08-01)
f.
Severity Level V violation for failure to complete Class 1 and 2 component inspections in the periods required by ASME Section XI. (324/86-10-01)
g.
Severity Level V violation for failure to control scaffolding, resulting in attaching scaffolding to a safety related snubber. (325/86-17-01)
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h.
Severity Level V violation for failure to operate an automatic welding machine within specifications.
(325/87-04-01).
2.
Conclusion Category: 2 3.
Board Recommendations _
NONE
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I.
Quality Programs and Administrative Controls Affecting Quality 1.
Analysis During the assessment period inspections were performed by the resident and regional inspection staff.
For the purposes of this assessment, this area is defined _as the ability of the licensee Jo identify and correct their own problems.
As such, it enwmpasses all plant activities, all plant personnel, as well as those corporate functions -and personnel that provide services to the plant.
The plant and corporate QA staff are part of the entity, and as such, they have responsibility for verifying quality.
The rating in this area specifically denotes results for various groups in achieving quality as well as the QA staff in verifying that quality is achieved.
Selected elements of the Brunswick' Improvement Plan (BIP) were inspected and found to be satisfactory. These BIP improvements resulted in a stronger QA program, more effective corrective
action and' establishment of a nonconforming trending program,-
l The licensee used onsite QA/QC personnel to perform inservice visual inspections during the most recent outage.
These inspections were thorough and professionally performed.
Management involvement in more effective and efficient use of Quality Assurance (QA) organization resources was evident during the period.
Beginning in the last assessment period and
continuing into the present period, QA management has been seeking more effective ways to utilize QA resources. Ther-was
a noticeable shift in QA policy and philosophy, which res ted
'
in an increased emphasis on surveillance type activities versus
documentation reviews.
For example, implementation of a sampling program versus a 100's review of completed documentation at the end of the last assessment period resulted in a transfer of two individuals from the document review group to the surveillance group.
Management reviews of the impact of the change during this assessment period has revealed a potential q
for additional refinement which should free additional resources for surveillance-type activities.
The review identified areas in which few or no significant findings have been issued.
QA management in conjunction with maintenance management is experimenting with revised sampling program acceptance criteria which may result in increased efficiency in these areas by both QA and the plant staffs.
Increased surveillance-type activities were exemplified by implementation of a Quality Verification Program.
This is an i
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f
.32-l in-depth inspection-of all attributes associated with. one.
construction / modification work-activity.
The inspection continues until'an acceptable level of performance is obtained.
Engineering assistance is utilized in selection of ~ attributes, thereby. ensuring inspection effort is directed to the more important safety aspects of the work activity..An inspection of electrical terminations using this process was completed.
A similar inspection'is in progress for conduit supports.
In the first quarter of 1987, QA management disseminated a modified policy on issuance of nonconformance reports (NCRs)..
Essentially the staff was directed to issue NCRs when. reasonable.
evidence was available that a deficiency existed.
The past practice had been to research the item until' certain that a deficiency existed... Thi s more aggressive policy re sul ted -' i n issues being resolved more quickly and corrective actions being identif.ied earlier.
Management also demonstrated a positive response to issues in the industry. Based upon the NRC's Safety System Functional Inspection (SSFI)' program at aaother facility in Region II, QA management undertook an SSFI of. the Standby Liquid Control system.
The inspection is scheduled to be complete by the end of September 1987.
The majority of the QA
!
staff effort is complete.
Completion of the inspection was l
,
'
pending technical assistance from; the - corporate - engineering -
staff.
.
Positive responsiveness to NRC -initiatives were demonstrated by QA management.
For.. example, the length of time ' involved i n issuing QA staff concerns on items whose regulatory basis i s unclear, appeared to be lengthy.
The li:ensee responded by establishing guidelines on how long such an item can exist before it must be dispositioned. Also, the length of time some NCRs are outstanding, i.e.,
time between issuance and full resolution was found to be excessive. A joint effort by QA and the plant staff was undertaken to reduce the number of NCRs
,
which were outstanding for long periods, (some in excess of three years).
Management added a program at Brunswick to further assure that the facility was modified and operated safely.
The Quality Check program, after successful implementation at the Harris Plant, was initiated at Brunswick in early 1987. This program provided employees the opportunity to express concerns and receive answers on any issue related to plant operation.
Management's response to the drug allegations revealed oy this
,
program demonstrated management's sensitivity to the NRC's Fitness for Duty Policy.
f
[
Management has attempted to identify the safety importance of modifications and place issues in their proper safety perspective by funding a Probabilistic Risk Assessment (pRA) for I
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both units. The licensee has computerized the data to keep the i
PRA current. The licensee discovered a common mode f ailure of i
.
the diesel generator building Heating Ventilating Air
.{
Conditioning (HVAC) system during development of the PRA.-
l Technical Support continued with several programs to enhance
quality. at Brunswick.
The FSAR commitment program was established and almost completed during the assessment period.
This program reviewed the updated FSAR for commitments, verified compliance and established a database.
The technical manual verification program, started in response to the Salem ATWS, is still ongoing. Development of the Engineering Data Base (EDBS)
system continues. EDBS will have safety class, drawing numbers, parts lists and other references to assist the plant in maintainence of equipment.
The Environmental Qualification (EQ)
program identified approximately ~ seven items that were not fully qualified.
The problems demonstrated the lack of a complete review prior to the EQ implementation date; however, aggressive pursuit of the issue once identified, resulted in positive corrective actions.
Enforcement actions are still pending on these issues.
A review was performed on all sections of the SALP report in an *
attempt to capture strengths and weaknesses related to management controls affecting quality.
The following are some perceived strengths in
.anagement
'
controls affecting quality.
Management initiated several programs which enhanced operational based decisions.
New management initiatives included: operations staffing plan, establishment of a Quality Team in operations and implementation of the automated maintenance management system.
Management was responsive to NRC initiatives and questions after events.
Maintenance management completed a self-initiated Maintenance Improvement Program (MIP) to enhance work i
controls and scheduling, upgrade knowledge level of craftsmen and provide more management oversight and involvement.
Training Activities have consistently received management support.
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2 Man.agement-takes.an active role in-providing policies which assure supervision in the field.
'
i-Maintenance Management retained 22 dedicated procedure writers to maintain the existing procedures.
l
Management was involved and approached a problem concerning ASCO solenoid value failures in a safe, conservative manner in that as soon as the problem was identified a team was formed to determine the cause and corrective action.
Two security specialists were added to the site security j
management staff to improve program management and oversight of the contract security force.
The following are some perceived weaknesses in management I
controls affecting quality.
f Increased management attention has reduced the number of reactor scrams however, increased attention is need to reduce the resulting complications.
Management attention is needed to correct the occasional.
lack of attention to detail.
Increased management attention is need to reduce the number of personnel errors.
A weakness was noted in the coordination between.on sito management and licensing management with regard to translating the schedular needs of the plant into a schedule for resolution of the associated licensing activity.
The licensing management had not consistently assured that the evaluations provided in accordance with 10 CFR 50.59 were adequate to clearly justify the No Significant Hazards (NSH) considerations determinations accompanying amendment requests.
One violation was identified:
Severity Level IV Violation for failure to ensure FSAR commitment concerning standby liquid control relief valve restoration to service was implemented. (325/86-11-03 and 324/86-12-03)
!
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f 2.
Conclusion I
Category:
]
)
3.
Board Recommendations l
l NONE l
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J.
Licensing Activities 1.
Analysis
-The basis for the appraisal in this area was the licensee's performance in support of licensing actions (amendment requests, responses to generic letters and other actions) which were reviewed and evaluated by the staff during the rating period.
]
The licensee's management demonstrated active participation in licensing activities and kept abreast of current and anticipated
>
licensing ' actions.
_ Management control and oversight of licensing ' activities were generally satisfactory.
The l
licensee's management has adopted a computerized. scheduling system thac provides excellent management control-over the
'
scheduling and prioritization. of licensing activities.
This=
l awareness of scheduling control was evident in the bi-monthly
"
licensing action review meetings held with the ~ NRC staff.
Consequently, commitments to NRC. requirements and responses. to requests ' for information -were implemented on time, and when
,
'
conditions preclude prompt implementation or response, justifi-cation-_for_the delay was provided.
'
>
The licensee's management is committed to a program to clarify and simplify the Brunswick facility Technical Specifications to i
minimize operator error and increase equipment reliability.
!
However, a weakness was noted in the coordination between
on-site management and. licensing management with regard to t
translating the schedular needs of the plant into a schedule for resolution of the associated licensing activity.
In addition, j
the licensing management had not consistently assured that the evaluations provided in accordance with 10 CFR 50.91 were adequate to clearly justify the No Significant Hazards (NSH)
consideration determinations accompanying amendment requests.
The licensee's management and staff generally demonstrated sound
technical understanding of issues involving licensing actions, j
The licensee had developed on-site expertise to approach
licensing actions from a probablistic risk assessment (PRA)
perspective in an attempt to increase margins of safety in the
!
functioning of plant systems.
The plant-initiated PRA studies identified deficiencies in the design of a safety-related ventilation system and prompted further licensee studies to upgrade the system.
While the licensee brought considerable l'
technical expertise to bear in resolving NRC staff concerns, some of the licensee submittals lacked the necessary information for review by the NRC.
That is, while the licensee staff
,
usually showed a good understanding of the technical issues and j
I applied a conservative approach to their resolution, the
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technical arguments used to justify the licensee's position were frequently not' communicated in a thorough and clear manner.
The licensee kept abreast of industry approaches to the resolution of generic plant safety issues and demonstrated an i
awareness of programs at other facilities.
This was accomplished through membership in all. major utility advisory and owners' groups.
Licensee management was represented. in
{
leadership positions in many of these groups. The licensee had
!
taken the initiative in the resolution of many multiplant safety issues. Where appropriate the-licensee either adopted methods
,
of resolution promulgated by industry-supported groups or proposed alternate methods of resolution.
Where a
.
plant-specific alternate method of compliance had been proposed,
l the licensee has occasionally not provided thorough technical arguments to justify the alternate approach, as indicated, for
!
example, by the difficulties involved in the resolution of the hydrogen recombiner issue.
l The licensee was generally responsive to NRC initiatives as evidenced by programs for improvement of the facility Technical Specifications and the use of probabilistic risk assessment to
!
evaluate the impact on safety of proposed amendments.
The use, i
of hydrogen water chemistry was tested by the licensee in
<
response to the NRC program to reduce intergranular stress
, corrosion cracking at boiling water reactor facilities.
The-licensee has demonstrated a readiness to support NRC review of licensing issues by providing additional information on an expedited basis.
The licensee has developed an integrated schedule (long-range plan) to assure that plant modifications required by NRC regulations were implemented on a high priority
,
basis, and that other modifications to improve safety were l
scheduled appropriately.
In general, the licensee worked
'
expeditiously to resolve safety issues in a timely manner.
The size of the licensing staff was more than adequate to
!
support licensing activities. The licensee assigned a principal j
engineer to supervise Brunswick licensing activities and i
continued to provide a licensing staff member at the site.
The creation of an onsite licensing engineer position has not eliminated concerns related to coordination of site licensing requirements with corporate office licensing activities. While the licensing staff displayed a wide knowledge in various technical disciplines, a weakness was evident in the training provided to the licensing staff relative to an understanding of the regulatory process, particularly with regard to the application of the standards in 10 CFR 50.92 for the determination of NSH considerations in amendment requests.
LER s t.5mi tta l s were made on a timely basis and contained detailed information on the event description and event
,
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evaluation.
A recent AE00 report dated July 22, 1987, compared j
LER's from November 1985 through June 1987 with the reporting requirements of 10 CFR 50.73(b) and the guideline contained in NUREG-1022.
The report concluded that the documentation, reporting quality of the LER's generally improved significantly in all major areas since the previous analysis of Brunswick LER's.
One violation was identified:
Severity Level V violation for f ailure to maintain the spent fuel pool PWR fuel storage capacity to no more than technical specification design description. (324/87-03-01)
!
l 2.
Conclusion Category:
3.
Board Recommendations NONE
.-
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Training l
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1.
Analysis J
!
During this assessment period, routine inspections of plant i
training programs were performed by the regional and resident inspection staffs. Management continued to be responsive to NRC initiatives and concerns and sought improvements to the plant training programs.
l The reactor operator replacement training program was reviewed i
and found acceptable.
Operator licensing replacement examinations were administered by the NRC during this period.
Twenty reactor operator replacement examinations were i
f administered with fourteen candidates passing, and sixteen senior reactor operator (SRO) replacement examinations were l
administrated with fourteen candidates passing.
These results I
are comparable to the industry average for replacement examinations.
No requalification examinations were administered by the NRC to j
Brunswick licensed operators and senior operators during this-
'
evaluation period.
All ten training areas have been accredited by the Institute of Nuclear Power Operations (INPO).
The reactor operator, senior reactor operator, and nonlicensed operator training programs were accredited in May 1985.
The remaining areas were accredited this SALP period. These include
)
instrumentation and control, electrical and mechanical technician training programs (in December 1985), radiation control, environmental and chemical, technical staff and management, and shift technical advisor training programs in September 1986.
The simulator is in the process of being modified to better reflect the status of the plant.
The modifications have been installed and final acceptance testing was in progress at the end of the SALP period.
The licensee's general employee training was not assessed during this period, but other unlicensed employee training was considered as not being completely effective as previously noted
,
in the assessments for Plant Operations, Fire Protection, i
l Emergency Preparedness, and Licensing Activities. The licensee implemented a Craft and Technical Development Program in the area of maintenance training.
General maintenance craft training was conducted by the training department.
Specialized
]
training on specific pieces of important equipment or tasks was l
identified by maintenance and conducted by the applicable l
vendors. This program provides management policies to assure
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that foremen and supervisors are involved in direct supervision of maintenance activities in the field and also provides a feedback mechanism for procedure deficiencies or work related problems to be corrected and developed into special training sessions.
No violations or deviations were identified.
2.
Conclusion Category:
3.
Board Recommendations I
NONE
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l V.
SUPPORTING DATA AND SUMMARIES.
A.
Licensee Activities
,
During the assessment period, major activities included normal power operations, refueling of both. units with barrier fuel (60'. barrier fuel in each unit) and extensive modifications and repairs as follows:
,
Unit 1 Appendix R' modifications.
Environmental Qualification modifications.
Inspection of large diameter single phase piping for
. erosion / corrosion.
. Emergency Response' Facility.Information System modifications.
Chemical decontamination of the reactor recirculation system.
Main turbine low pressure rotor work.
The unit ended.the assessment period conducting normal power
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operations.
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Unit 2 Chemical decontamination of the reactor recirculation system.
Recirculation pipe inspection, weld overlays, and induction heat stress.
Service water pipe replacement.
Environmental Qualification modifications.
Appendix R modifications.
10 year Inservice Inspection.
Emergency Response Facility Information System partial installation.
HFA relay replacement.
Unit 2 shutdown on November 30, 1985, to comply with 10 CFR 50.49 requirements after the Commission did not approve a schedular exemption.
The unit was restarted on June 15, 1986.
The unit ended the assessment period conducting normal power operations.
l One declaration of an unusual event was made during the assessment period.
On February 24, 1986, a contaminated injured man was transported to a local hospital.
The contaminated material was returned to the site the same day.
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The routine scheduled INPO evaluation was performed in April 1986. A special INP0 survey was conducted of maintenance activities during March 1987, as part of an industry effort to determine the factors which were deemed responsible for ehhanced programs at sites with improving maintenance performance, l
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B.
Inspection Activities During the assessment period, routine inspections were performed at the Brunswick facility by the two resident inspectors and the
regional inspection staff.
Special inspections were conducted to j
augment the routine inspection program as follows:
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October 2 - 4 and (!ovember 13 - 15, 1985, review of main steam isolation valve solenoid problem which resulted in a steamline j
not capable of being isolated while Unit 2 was in cold l
shutdown.
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January 13 - 17, 1986, a special site survey of the maintenance program and practices was conducted by the Division of Human Factors Technology, Office of Nuclear Reactor Regulation, July 8 - 10, 1986, allegation followup, posting of high radiation areas, storage of high level waste and health physics practices.
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January 5 - 8, 1987, assessment of the hydrogen water chemistry test.
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May 4 - 8, 1987, review of operational events, control room activities, maintenance, training and Licensee Event Reports.
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