IR 05000309/1985015
| ML20136G584 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 07/22/1985 |
| From: | Callan L, Dyer J, Foley T, Kearney J, Lamastra M, Martin T, Mckee P, Paulus R, James Smith, Danielle Sullivan, Larry Wheeler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE), Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20136G570 | List: |
| References | |
| 50-309-85-15, 850725, NUDOCS 8508190559 | |
| Download: ML20136G584 (41) | |
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0FFICE OF INSPECTION AND ENFORCEMENT DIVISION OF INSPECTION PROGRAMS PERFORMANCE APFRAISAL SECTION (PAS)
Report:
50-309/85-15 Docket:
50-309 License: DPR-36 Licensee: Maine Yankee Atomic Power Company 83 Edison Drive Augusta, Maine 04336 Facility: Maine Yankee Atomic Power Plant Inspection at: Maine Yankee Atomic Power Plant, Maine Yankee Atomic Power Com-pany Offices, and Yankee Atomic Electric Company Offices
Dates of Inspection:
May 13-24 and June 3-7, 1985 Inspectors:
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~~/[:GL/6 T. D. Martin, Inspection Specialist, IE - Team Date Leade
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?lul6 L.J.Ca9Jan, Chief,PerformanceAppraisalSection, Date '
IEhey 7/2T/BC J.E. Dyet, Inspection Specialist, IE Date T O.Md~
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T. Folep, Senior Residdnt Inspector, Region I Date 44al-A d&
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J.P.Kprney,ReactorOpe{ationsEngineer,IE mun Jha/n'
N. J. Lamastra, Health Physicist, NRR Date
.Sh/c..Q~b's 7/nh e R. C.
aulus, Senior health Physicist, IE Date
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7dadr J. D. Srpiph, Inspectiorl Specialist, IE Date /
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V.J.fullivan,Jr.,InspedtiotSpecialist,IE D6te MGV k
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L.L.'W$feler,InspecfionSpecialist,IE
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8508190559 850806 PDR ADOCK 05000309 i
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l-Accompanying Personnel:
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H. Vollmer, IE
- E. C. Wenzinger, RI
- C.
F. Holden, Jr., RI
- Present during the exit interview on June 7, 1985.
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' Approved by:
A bI Phillip F./McKee, Chief Ddte Operating Reactor Programs Branch, IE s.
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i INSPECTION SUMMARY The inspection was conducted May 13-24 and June 3-7, 1985 (Report 50-309/85-15).
I Areas Inspected: A special, announced inspection was performed of the licensee's management controls over selected licensed activities. The inspection was con-ducted by 10 NRC inspectors and involved 871 inspector-hours on site and at the corporate offices.
Results: The licensee's management controls for eight functional areas were re-viewed and conclusions were drawn in each area based on observations presented i
in this report. The licensee's performance in each area was categorized in ac-cordance with NRC's latest guidance for evaluating licensees under the Systematic Assessment of Licensee Performance (SALP) Program.
For the areas inspected, the conclusions are presented as Category One, Category Two, or Category Three.
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i Functional Area Category
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Plant Operations Two Surveillance Two Maintenance Two j
Training One
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Design Changes and Modifications Two Quality Programs Two Procurement Two Radiological Controls Two t
l Additionally, 13 potential enforcement findings were presented to the NRC Region I Office as Unresolved Items for followup.
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TABLE OF CONTENTS Topic Page Inspection 0bjective..............................................
Plant 0perations..................................................
Surveillance......................................................
Maintenance.......................................................
Training.........................................................
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Design Changes and Modifications.................................
Quality Programs.................................................
18; Procurement......................................................
Radiological Controls............................................
Unresolved Items................................................. 30 Management Exit Meeting..........................................
APPENDICES A
Persons Contacted and Documents Examined B
Abbreviations iv
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INSPECTION OBJECTIVE The objective of the inspection was to evaluate the management control systems that had been established in support of licensed activities.
The results provide input to the NRC evaluation of licensees from a national perspective.
The inspection effort covered licensed activities in selected functional areas.
In each of the functional areas, the inspectors interviewed responsible person-nel, observed activities, and reviewed selected records and documents to deter-mine whether:
1.
The licensee had written policies, procedures, or instructions to provide management controls in the subject area.
2.
The policies, procedures, and instructions were adequate to ensure compli-
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ance with the regulatory and internal requirements.
3.
The licensee personnel who had responsibilities in the subject areas under-stood their responsibilities and were adequately qualified, trained, and retrained to perform their responsibilities.
4.
The requirements of the subject area had been implemented and appropriately
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documented in accordance with management policy.
The specific findings in each area are presented as observations that the in-spectors believe to be of sufficient importance to be considered in a subsequent evaluation of the licensee's performance.
The observations were the perceived strengths and weaknesses that were used as the basis for determining the team's evaluation and categorization of each area in accordance with the following per-formance categories.
Category One - Reduced NRC attention may be appropriate.
Licensee management attention and involvement are aggressive and oriented toward nuclear safety; licensee resources are ample and effectively used so that a high level of per-formance with respect to operational safety or construction is being achieved.
Category Two - NRC attention should be maintained at normal levels.
Licensee management attention and involvement are evident and are concerned with nuclear safety; licensee resources are adequate and are reasonably effective so that satisfactory performance with respect to operational safety or construction is being achieved.
Category Three - Both NRC and licensee attention should be increased.
Licensee management attention or involvement is acceptable and considers nuclear safety,
but weaknesses are evident; licensee resources appear to be strained or not ef-
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fectively used so that minimally satisfactory performance with respect to opera-tional safety or construction is being achieved.
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The performance categories defined above have been developed to meet the NRC's latest guidelines for evaluating each licensee under the Systematic Assessment of Licensee Performance (SALP) Program. These categories have been published in the Federal Register.
Some observations may be potential enforcement findings. These observations, referred to as Unresolved Items, were discussed with the licensee and were presented to the NRC Region I Office for followup.
PLANT OPERATIONS OBSERVATIONS 1.
The management of the operations staff and the amount of plant-specific knowledge in support organizations were considered strengths. Maine Yankee had a six-shift rotation for the operations staff. The requirement for two licensed senior reactor operators on each shift had been met.
The number of licensed operators was sufficiently large to permit operating well with-in the overtime limits specified in NUREG-0737, Item I.A.l.3.
There was strong management emphasis on ensuring full participation in scheduled train-ing activities during the week in the shift cycle devoted to training. The spare week was used for most administrative duties and assisting the on-shift staff as necessary.
There was an abundance of operating experience and knowledge of systems among the personnel in support organizations.
2.
The Shift Technical Advisor (STA) program was considered a strength. The STA function was performed by Nuclear Safety Engineers (NSEs) who met the STA requirements of NUREG-0737, Item I.A.1.1.
The NSE Section was independ-ent of the plant operations chain-of-command.
The section supervisor repor-ted offsite to corporate headquarters. The NSEs were on an 8-hour shift and provided continuous support to the operations crew.
They routinely per-form a critical review of operations activities for compliance with Technical Specifications and other requirements. NSEs prepared 10 CFR 50.73 Licensee Event Reports and made 10 CFR 50.72 reports.
NSEs performed quality assur-ance surveillances of operations activities, reviewed temporary plant modi-fications (jumpers and lifted leads), and supported the corporate engineer-ing staff when requested. The NSE Section long-range career development program included obtaining a senior reactor operator (SRO) license (the NSE section supervisor and one NSE had an SR0 license; two NSEs had a reactor
operator (RO) license). There was a comprehensive program for initial and continuing training.
It also was noted that the NSEs had the professional respect of the operations staff and plant management.
3.
A weakness was noted in the program for the independent verification of operating activities.
The Maine Yankee Tagging Rules, revised March 21, 1985, did not require the independent verification of the temporary clear-ing and subsequent hanging of equipment control tags when the tagging ac-tions were performed to support equipment testing.
Independent verification only was provided for the initial hanging and final clearance of tags.
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Management oversight of temporary plant modifications was weak in that some temporary modifications had been installed for long periods of time without being incorporated as plant modifications. For example, temporary modifi-cations 80-13 (incore detector drives) and 78-25 (coolant loop isolation valve limit switch bypass) had been installed for approximately 5 and 7 years, respectively.
Procedure 16-1, " Maine Yankee Operation Safeguard, Yellow Tag-Control Log," stated that any temporary modification not closed out for a 6-month period shall be reviewed by the Plant Operation Review Committee (PORC) for a possible engineering design change recommendation. A Maine Yankee QA audit revealed that the PORC reviews had not been performed.
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PORC subsequently reviewed the log and documented their concurrence with the 10 CFR 50.59 safety analyses, but there was no record of recommendations for
or against making any possible design changes.
5.
Several main process valves were not labeled, but the licensee had a pro-gram in place that called for all valves in the plant to be labeled by late 1986.
The failure to have all major system valves labeled and the long period of time the licensee planned to take to correct the problem were considered weaknesses.
6.
Plant cleanliness was satisfactory. A few minor accumulations of dirt
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and debris were noted in areas not frequently travelled. Control room spaces were clean and orderly and noise levels were generally low. Opera-tors were alert to plant conditions and displayed a professional demeanor.
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CONCLUSIONS
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Strengths included the number of licensed operators, the operations knowledge and experience of the support staff, and the shift technical advisor program.
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Weaknesses were noted in the independent verification of operating activities, the management review of temporary plant modifications, and efforts to ensure all appropriate valves were labeled.
This area was rated Category Two.
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SURVEILLANCE OBSERVATIONS 1.
Maine Yankee plant management demonstrated a strong commitment to identify
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and correct equipment problems before they could lead to safety system challenges or inoperable safety systems.
This commitment was evidenced by the following:
An aggressive inservice testing (IST) program had been implemented.
a.
Review of IST records indicated that test data trending was accomplish-ed in a comprehensive manner, with the trending results routinely used as a basis for shortening surveillance test intervals and e?rly iden-tification of equipment problems.
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b.
The plant had implemented an equipment reliability program applicable to major non-safety-related plant equipment not covered by the IST, Technical Specifications surveillance requirements, or preventive main-tenance programs. This reliability program provided a means to detect equipment degradation and initiate corrective action before an equip-ment failure occured. The effectiveness of this program was illustrated during the inspection by the identification of degradation of a main feedwater pump which, if gone undetected, could have led to pump fail-ure and a subsequent reactor trip.
c.
As discussed in Observation 2 of the Maintenance section of this re-port, the plant had implemented a highly effective preventive main-tenance program.
The degree of involvement by plant management in these programs and the t;mely action taken on program recommendations was considered a strength.
2.
Acceptance criteria often were not incorporated into surveillance test pro-cedures in a format consistent with the requirements of Maine Yankee Pro-cedures 0-10-2, " Surveillance Tests and Records," and 0-6-1, " Procedure Preparation, Classification and Format." Review of surveillance test pro-cedures revealed that each of the several departments (operations, main-tenance, plant engineering, instrumentation and controls, etc.) establish-ed and administered their own surveillance test procedures and that each department incorporated acceptance criteria into these procedures differ-ently. The significant inconsistencies are discussed below.
Surveillance procedures typically did not have an " acceptance criteria" a.
section, as specified by Procedure 0-6-1.
This section was intended to be at the end of the surveillance procedure and to summarize either qualitatively or quantitatively the results of the completed surveil-lance test. Operations Department surveillance procedures appeared to
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be the only ones that consistently implemented this format requirement.
The other surveillance procedures incorporated the various acceptance criteria throughout the body of the procedure without summarizing the results, b.
Acceptance criteria in surveillance procedures typically did not have the applicable Technical Specifications requirements referenced, as specified by Procedure 0-10-2.
Instrumentation and Controls Depart-ment procedures appeared to be the only ones that consistently refer-enced Technical Specification requirements next to acceptance criteria.
These inconsistencies in procedure format were of particular concern be-cause acceptance criteria that directly determine the operability of safety-related equipment may not be clearly identified and highlighted in the surveillance records so that appropriate and tinely corrective measures can be taken.
3.
The method by which Maine Yankee implemented the ANSI N18.7-1976 require-ment to establish an integrated master surveillance schedule appeared to be weak.
Each department having surveillance responsibilities established
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and implemented its own master schedule.
There was no central oversight for planning and scheduling of surveillance testing.
The significance of this weakness was somewhat diminished after a review of several years of Licensee Event Reports and NRC enforcement actions revealed that Maine Yankee does not have a history of failing to perform surveillance tests within the required time interval.
4.
A weakness was noted regarding the means used to provide assurance that all Technical Specification surveillance requirements were incorporated into the surveillance program. Specifically, a.
The 1983 and 1984 Quality Assurance audits of Technical Specifications were ineffective, particularly with respect to ensuring that surveil-lance requirements were implemented (see Quality Programs, Observa-tion 2).
b.
A means had not been established to cross-reference Technical Specifi-cations surveillance requirements with corresponding implementing proce-dures to ensure that all requirements were being met.
Interviews with personnel responsible for implementing the surveillance test program re-vealed confusion, in some instances, regarding which implementing proce-dures applied to which specific Technical Specifications requirements.
This was particularly true regarding instrumentation channel calibration surveillance requirements where implementation of each requirement often involved two or three individual test procedures.
During the course of this inspection, the licensee drafted an apparently comprehensive cross-reference between Technical Specifications surveillance requirements and corresponding implementing procedures.
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Review of records of completed surveillance tests performed by Operations Department personnel revealed that failed as-found test data were not consistently being included as part of the test record.
Instead, only the successful as-left test results were being documented in some instances.
Recent examples where failed as-found test results were not documented in the surveillance record include a.
The March 12, 1985, performance of Surveillance Procedure 3.1.2, "ECCS Routine Testing," did not record the inoperable as-found condition of valve LSI-M-31.
This valve was found to be inoperable as a result of the failure of its associated automatic bus transfer to shift.
b.
The March 1985 performance of Surveillance Procedure 3.1.20, " Safe-guards Valve Testing," did not record the inoperable as-found condition of valve LM-A-57.
This valve was found to be inoperable because of its failure to meet the stroke-time acceptance criteria.
Interviews with Operations Department shift personnel who perform surveil-lances indicated that there was uncertainty regarding the need to record as-found test data if the test was unsuccessful. The fact that Operations
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Department surveillance procedures typically did not provide for recording both as-found and as-left data contributed to this uncertainty.
This policy of not always recording as-found data appears to be consistent with Step 7.1.3 of Procedure 0-10-2, " Surveillance Tests and Records," which requires recording as-found data only for "...those surveillance items which require calibration, adjustment, etc...."
Since most of Operations Depart-ment surveillances do not require calibration or adjustment, recording of as-found data was not specifically required in most cases by Maine Yankee surveillance procedures.
The inspector verified that inoperable equipment discovered during surveillance testing was reported to the control room in a timely manner and recorded in the control room log.
However, the failure of the surveillance procedures, discussed above, to require recording as-found conditions is contrary to Section 5.3.10 of ANSI N18.7-1976.
This matter was discussed with the licensee, and will remain unresolved pending followup by the Region I Office (309/85-15-01).
CONCLUSIONS Aggressive inservice testing and equipment reliability monitoring programs were a strength. Weaknesses included not always incorporating acceptance criteria into procedures consistent with licensee procedural format requirements, the lack of an integrated master schedule, the lack of a cross-reference of Techni-
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cal Specifications surveillance requirements with corresponding implementing p% cedures, and not consistently including as-found test data as part of the surveillance test record.
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This area was rated Category Two.
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MAINTENANCE OBSERVATIONS 1.
The qualifications and training of mechanics, electricians, and instrument-ation and control (I&C) technicians was considered a strength (see Training, Observation 11).
2.
The licensee had established a comprehensive and effective preventive main-tenance program that included both safety-related and non-safety-related systems and components.
The level of management commitment and scope of this program was considered a strength. Management performed quarterly reviews of Discrepancy Reports (DRs) and Repair Orders for repetitive equip-ment failures.
On the basis of these reviews and the results of inservice testing and equipment reliability programs, numerous additions and deletions
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had been made in the preventive maintenance program (see Surveillance, Ob-servation 1).
3.
High pressure injection (HPI) system Limitorque valve operability has im-proved since the implementation of a valve maintenance program based on Institute of Nuclear Power Operations (INPO) Report 83-037, " Assessment of obtor-0perated Valve Failures." A review of HPI system Limitorque valve maintenance history indicated significant improvement in valve operability since the 1983 outage.
Interviews with maintenance personnel revealed that the licensee had implemented the INP0 program during the outage. The program has since been incorporated into detailed maintenance procedures.
In addi-tion, extensive Limitorque valve training for mechanics, electricians, and I&C technicians was conducted in March 1984.
4.
The licensee failed to adequately implement their measuring and test equip-ment (M&TE) program. A review of program implementation for 45 pieces of test equipment revealed the following:
a.
Calibration procedures for outside micrometers and torque wrench-es and calibration cross-check procedures for dead weight test-ers did not designate the calibration points to be checked.
b.
Evaluations were not conducted to verify the validity of tests made with M&TE that were later discovered to be out of tolerance.
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Test equipment pieces61-007, 61-113,61-038, and 62-151 were found to be out of tolerance on April 28, 1983, July 21, 1983, August 20, 1984, and September 12, 1984, respectively. Evaluations
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were not conducted, as required by the Maine Yankee Quality As-
surance Plan, to verify the validity of previous tests performed using this M&TE.
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Evaluations were not conducted, as required by the Maine Yankee Quality Assurance Plan, to document the basis of acceptance when using calibration devices having accuracies less than four times the accuracy of the equipment being calibrated.
Interviews of maintenance supervisors revealed that they were unaware of this requirement.
In addition, no guidance was provided in the licen-see's administrative procedures for conducting these evaluations.
The following are examples where a required evaluation was not conducted:
(1) The digital voltmeter for the reactor protective system cal-ibration and indicating panel (RPSCIP) digital voltmeter was calibrated each refueling outage to an accuracy of 0.002 Vdc using a Fluke calibrator with an accuracy of 0.0012 Vdc at 20 Vdc. The acceptability of using the Fluke calibrator having an accuracy of only 1.67 times greater than the RPSCIP digital voltmeter had not been evaluated.
(2) The monthly calibration checks of the thermal margin cal-
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culator reference voltage device with an accuracy of 0.005 Vdc were conducted using the RPSCIP digital voltmeter having an accuracy of 0.002 Vdc.
The acceptability of using the RPSCIP digital voltmeter having an accuracy of only 2.5 times greater than the reference voltage device had not been eval-uated.
(3) The monthly calibration checks of the RPSCIP digital volt-meter having an accuracy of i0.002 Vdc were performed using a Fluke digital voltmeter having an accuracy of 0.007 Vdc.
The acceptability of using the Fluke digital voltmeter having an accuracy 3.5 times less than the RPSCIP digital voltmeter had not been evaluated.
d.
The I&C M&TE usage log sheets did not always identify where the test equipment was used, as required by Procedure 0-06-5, "Measur-ing and Test Equipment," revision 2.
(1) The usage sheet for test equipment 62-124 did not show any equipment usage between January 8 and May 22, 1985. A brief review of instrument calibration records revealed that the instrument had been used a minimum of six times during this period to calibrate other M&TE.
(2) The usage sheets for test equipment pieces62-103, 62-124,62-134, and 62-138 did not indicate that they had been used to calibrate M&TE belonging to groups other than the I&C section.
The failure to adequately implement a M&TE program was discussed with the licensee and will remain unresolved pending followup by the NRC Region I Office (309/85-15-02).
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A program to control vendor technical manuals was not being developed and
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implemented consistent with commitments made to the NRC. The licensee orig-inally committed to the NRC to develop and implement such a program by January 1, 1985.
On December 12, 1984, the licensee formed a Vendor Equip-ment Technical Information Program (VETIP) task force to review the current methods of controlling vendor technical manuals and to make recommendations for implementing a program based on the INP0 Nuclear Utility Task Action Committee (NUTAC) report.
On December 14, 1984, the licensee made a request to extend the commitment date for implementing their program from January 1 to September 1, 1985. Review of the implementation schedule of the VETIP task force recommendations revealed that full program implementation was scheduled for June 1986.
CONCLUSIONS Strengths included a highly qualified staff, a strong training program, a compre-hensive and effective preventive maintenance program, and an effective motor-operated valve (Limitorque) maintenance program that has improved high pressure injection system reliability. Weaknesses were noted in the implementation of the measuring and test equipment program and in the timeliness of developing and implementing a program to control vendor technical manuals in accordance with
commitments to the NRC.
This area was rated Category Two.
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s TRAINING OBSERVATIONS 1.
The qualifications of the operator training staff was a strength.
Nine of the eleven staff members had senior reactor operator (SRO) licenses.
This enhanced their effectiveness as instructors and increased their credibility with the Operations Department.
The Training Qualifications Review Board (TQRB) was considered a stre 2.
Members of this board were selected from senior management positions,ngth.
in the Departments of Operations and Training.
The TQRB reviewed major aspects of licensed operator training including a review of current programs and pro-gress of individual participants, a review of special or remedial training programs to correct individual and generic program deficiencies, the selec-tion of personnel for R0 and SR0 training, and the final approval of person-
nel for the NRC licensing examination. The effectiveness of the screening process for operator candidates was demonstrated by a greater than 90% pass-ing rate for R0 and SRO examinations given by the NRC over the last 5 years.
3.
Plant modifications training was considered a strength. A comprehensive refueling modifications training package and associated lectures were prov-ided for all licensed personnel before the outage. An updated modifications training package was issued and lectures were held for those modifications that most affected operations before plant startup.
Training for the remain-ing modifications was provided during the first requalification session after plant startup.
4.
Training was effective in minimizing personnel errors as evidenced by a re-view -- for potential training deficiencies -- of Unusual Occurrence Reports, Plant Information Reports, and Licensee Event Reports generated during the last 2 years.
This review revealed few occurrences that could be attributed to a training weakness.
5.
The quality of the annual licensed operator requalification examinations was considered a strength. Requalification examinations and reexaminations were well balanced in content and difficulty. Regrading of examinations was done thoroughly, and often resulted in lower grades, including an occasional failure of a section or the entire examination.
Very few repeat or similar questions were identified between the initial examination and reexaminations.
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6.
A weakness was noted regarding the participation in licensed operator requal-
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ification training sessions. About 11% of the licensed personnel attended less than 50% of the 1903/84 annual requalification training sessions. The majority of those who missed the requalification training sessions were non-shift personnel. Additionally, 19 of 172 annual requalification lecture series examinations for 1984/85 were overdue at the time of the inspection.
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Two individuals that scored less than 80% on the lecture series examinations did not receive additional training and re-examination. Maine Yankee Pro-cedure 18-20-1, " Licensed Operator and Operations Instructor Training Pro-grams," revision 2, states that all licensed personnel will demonstrate proficiency by examinations or evaluations of material covered in the requal-ification lecture series.
It also specified that a score of at least 80%
constitutes an acceptable grade on the examinations.
The licensee issued a training policy statement during the inspection en-titled " Licensed Operator Requalification Course Participation." This policy statement addressed requalification program attendance, examinations, and the review of assigned documents in a timely manner.
This policy state-ment also allowed the exemption from lecture attendance by pretesting. The licensee was cautioned that this exemption, if abused, might prevent them from meeting the regulatory requirements in 10 CFR 55, Appendix A, for li-censed operator participation in a continuing requalification program.
7.
A weakness was noted in the timeliness of the review of the Document Review Notebook. This notebook was used to keep operations personnel current on Abnormal Operating Procedures (A0Ps), Emergency Operating Procedures (EOPs),
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Technical Specifications (TS), and their associated changes. Records showed that five licensed operators were late, from 2 to 11 months, in their re-quired reading of AOPs, E0Ps, and TS in the Document Review Notebook.
This was considered more significant because this notebook was used as a primary means to keep the operations staff current on changes to AOPs, E0Ps, and TS, all of which had changes made since March 1,1985. Three of the five li-
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censed operators who were behind from 2 to 7 months were operating shift personnel.
8.
Tha Auxiliary Operator (AO) initial training program was a strength.
This training program was comprehensive, covering a broad range of plant hard-ware and other theoretical topics such as thermodynamics. However, the A0 requalification training was not preplanned and adequately supervised.
The licensee recognized this weakness and implemented corrective action during this inspection.
9.
Interviews and records review revealed that no initial or continuing chemistry training had been formally conducted and documented since 1983 for chemists or Radiation Control Technicians (RCTs).
Training was started during the inspection to correct this problem.
10. The fire brigade training was considered a strength.
The training consisted of frequently scheduled onsite training and offsite training with the local fire department. Unannounced simulated fire drills were conducted onsite
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using smoke machines, a modified transformer to simulate a transformer fire, and other devices for realism.
Each of the team members was evaluated by the Fire Protection Supervisor and Fire Brigade leaders after each onsite and offsite drill.
These evaluation results were used to critique the team and to provide additional training in weak areas.
The teams were larger than required, and a radiation control technician was assigned to each toan.
The fire brigade teams' qualifications and degree of participation in this collateral duty was a strength.
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J 11. Fhintenance staff training was generally considered a strength due to the
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frequency of training sessions, subjects taught, simulator demonstrations, and the selection of instructors who were experienced technicians in the areas being taught. Classroom training sessions were generally conducted weekly with excellent attendance. These sessions covered plant systems, plant modifications, administrative procedures, and professional develop-ment subjects. The simulator was used effectively to demonstrate system operation.
12. A review of schedules and training program materials revealed that prep-eration for INPO accreditation of the 10 training programs listed in SECY-85-1, " Policy Statement on Training and Qualifications of Nuclear Power Plant Personnel," was being actively pursued. A full time accreditation project manager and a dedicated staff of three instructors had been assigned to this task. Although evidence of this effort could be seen in the program materials reviewed, a significant amount of work remains to meet the licen-see's goal for submitting the self-evaluation report for the 10 programs to INP0 by December 1986.
CONCLUSIONS
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Strengths in the training area included a highly qualified instructional staff, effective management involvement in the training process through the TQRB, and positive results as evidenced by both a high passing rate on NRC licensing exam-inations and a lack of incidents that could be related to training. Additional strengths were noted in plant modifications training, licensed operator requali-fication examination quality, auxiliary operator training, fire brigade training, and maintenance staff training. Weaknesses were noted with the level of partici-pation by licensed personnel in the licensed operator requalification program, the review of procedure revisions by licensed operators, and chemistry tech-nician training. The licensee was considered to be making satisfactory pro-gress toward INP0 accreditation.
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This area was rated Category One.
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s DESIGN CHANGES AND MODIFICATIONS OBSERVATIONS 1.
Maine Yankee had implemented a new system of procedures in September 1983 to control engineering design change requests (EDCRs).
This system con-solidated the previous Maine Yankee Atomic Power Company and Yankee Atomic Electric Company design change programs into one prcgram that was managed at the station. This improved control of the design change process.
The program for closecut of EDCRs and return of the modified systems to opera-tion was particularly effective. For example, the following requirements had to be completed before placing a modified system into operation:
a.
All installation and test procedures had to be completed and reviewed.
b.
All affected operating procedures had to be revised and issued.
c.
Operators had to receive training on system changes (see Training, Observation 3).
d.
Control room drawings used for plant operations had to be revised to show new plant configurations. All other drawings had to be stamped to identify the outstanding EDCR changes to be entered.
e.
Commitments had to be obtained from cognizant departments for the ac-complishment of the other EDCR logistic requirements such as FSAR, procurement specification, and non-operator training updates.
The improvements created by this program were considered to be a strength.
2.
The flow of design change information between the various organizations re-sponsible for development and implementation of the modifications was not adequately controlled.
Procedure 17-21-2, " Engineering Design Change Re-quest (EDCR) - Maine Yankee," revision 0, identified the project and field engineers as being responsible for coordinating the flow of design change information.
However, minimal guidance was provided as to how this coor-dination of design change information flow would occur or whether the re-sponsibility for releasing design information could be delegated. As a re-sult, unqualified personnel were releasing design information, and inconsist-
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ent inputs were being provided to supporting organizations as evidenced by the following:
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Temporarily employed college students had reviewed, approved, and is-sued implementing instructions for construction of safety-related plant modifications.
One example was EDCR 84-15, " Steam Generator Wet Layup Recirculation System Taps."
b.
In one instance discovered by the inspector, the initial pressure for placing a modified system in service was specified differently in the operating procedures, design inputs, and design calculations.
EDCR 83-509, "Non-Return Valve (NRV) Vacuum Assist System," connected a line between the condenser and the NRVs to hold the valves open during plant cooldown.
The piping size calculations were based on placing the sys-tem in service at 150 psia. The design inputs in the EDCR package specified 165 psia (150 psig), and Procedure 1-7, " Plant Cooldown," re-vision 26, specified placing the system in service at 180 psia (165 psig).
In this case, the lack of consistent specifications could have resulted in the system being placed in service at a higher pressure and resultant steam flows greater than the system was desi port. (See Observation 5 for more details on EDCR 83-509.)gned to sup-The failure to adequately control design information flow between internal
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and external organizations is contrary to ANSI N45.2.ll-1974, Section 5.
This issue was discussed with the licensee and will remain unresolved pend-ing followup by the NRC Region I Office (309/85-15-03).
3.
The Maine Yankee system for control of design inputs had the following weaknesses:
a.
Design inputs were not always documented in a timely manner to support the remainder of the design change process. After the design analyses were completed, design inputs were documented in an introductory para-graph of the EDCR review package.
In many cases, this summary para-graph was incomplete and did not list all the inputs used for the EDCR analyses.
Instead, many of the inputs were identified by methods out-side the EDCR process such as conceptual project authorizations (CPA),
internal memoranda, letters, meeting minutes or verbal communications.
These inputs were not governed by quality assurance procedures and were not always part of the EDCR package being reviewed for design verifica-tion.
b.
Design inputs transmitted verbally were not always followed up by writ-ten documentation.
EDCR 84-63, " Eliminate Seal-in Feature for Valves CS-M-66 and CS-M-71," modified the motor-operated isolation valves for the chemical tank in the containment spray system to allow the valve to reverse direction in mid-stroke.
The potential for either motor,
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operator, or valve stem damage during this operation had not been ana-lyzed. Consequently, the design verification review identified an open item requiring vendor approval for the change.
Interviews revealed
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that verbal concurrence was obtained from the vendor by the project engineer to close the open item, but there was no written documentation of this concurrence from the vendor.
EDCR 84-63 was subsequently ap-
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proved for installation during the upcoming 1985 refueling outage with-
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out a documented analysis or vendor concurrence to demonstrate that these valves would not be damaged by this modification.
The failure to adequately document design inputs in a timely manner appears to be contrary to the requirements of ANSI N45.2.11-1974, Sections 3.1 and 1.4.
This issue was discussed with the licensee and will remain unresolved pending, followup by the NRC Region I Office (309/85-15-03).
4.
Load calculations conducted by Yankee Atomic Electric Company (YAEC) for EDCR E3-07, " Modifications of Feedwater Line Supports H-13 and H-15," did not reflect the actual installation of the anchor bolts.
One 3/4-inch anchor bolt was damaged during its installation on H-15 and was purposely sheared off to be even with the concrete surface.
Another anchor bolt was subsequently installed 2 inches from the damaged bolt, but the load calcu-lations for this anchor bolt were not derated because of the close prox-imity of the two bolts.
This appears contrary to the anchor bolt vendor manual which requires 10-diameter separation between anchor bolts to re-tain their full load capacities. After the inspector raised concern about the adequacy of this design, YAEC contacted the vendor and was told that a 15-20% load derating would be an acceptable estimate of the described sit-
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uation. When the derating was applied to the existing calculations, the results indicated an overloaded and unacceptable situation for H-15.
YAEC then reevaluated other calculation assumptions, taking a less conservative approach, and concluded that the present installation was satisfactory.
This problem of inadequate anchor bolt separation went undetected through the review and approval process.
Interviews with YAEC mechanical engineer-ing supervisory personnel revealed that the YAEC-Maine Yankee Project had no procedures providing guidance for anchor bolt installation and analyses in such situations.
Additionally, this particular calculation was the sub-ject of a previous NRC Notice of Violation (83-02-01). As described in this Notice of Violation, load calculations on the H-15 installation were not made until after an NRC inspector had noted the deficiency.
The failure to make an adequate design calculation was discussed with the licensee and will remain unresolved pending followup by the NRC Region 1 Office (309/85-15-04).
5.
No operational test was conducted of the main steam non-return valve (NRV)
vacuum assist system before releasing the system for operation. This modi-fication was accomplished by EDCR 83-509 during the last outage and inter-views revealed that the licensee intended to test the system by putting it into operation during the next cooldown.
However, Procedure 1-7, " Plant Cooldown," revision 26, was issued to allow this system to operate during a normal cooldown and there were no administrative controls in place to ensure that the initial operation was conducted in a controlled manner.
Additionally, no procedure had been prepared to control the testing of this system.
This was considered particularly significant because the valve manufacturer had approved this modification but warned the licensee that system startup or operation with a large pressure drop across the NRV disc could damage the valve. Adding to the significance of this issue was
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s the procedural inconsistency identified in Observation 2.b that could have resulted in this system being placed into operation at a pressure 20% higher than the system design value. The failure to test the NRV vacuum assist system before releasing it for operation is contrary to the requirements of ANSI N18.7-1976, Section 5.2.19.2.
This issue was discussed with the li-censee and will remain unresolved pending followup by NRC Region I Office (309/85-15-05).
6.
The scheduling of Plant Operations Review Committee (PORC) reviews and Plant Manager approval in the EDCR development process was weak. For safety-related EDCRs, developed by Maine Yankee, the PORC review and Plant Manager's approval occurred before the design verification review was conducted by YAEC. This preventet station management from reviewing the complete EDCR package and essentially eliminated their concurrence for the resolution of open items identified in the verification review.
In the case of EDCR 84-63, discussed in Observation 3.b., the requirement for contacting the vendor to obtain permission to modify the motor-operated valves occurred after the PORC review and Plant Manager's review and approval. Management review of this open item resolution may have resulted in further investi-gation and documentation.
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7.
The following deficiencies were identified in the drawing control program:
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a.
Two additional sets of uncontrolled drawings were kept with the con-trolled drawings in the control room. This increased the potential for the use of uncontrolled drawings for plant operations and was con-trary to Procedure 0-01-2, " Drawing Control," revision 3.
b.
Controlled drawings at various locations did not reflect the same sta-tus of outstanding EDCRs.
The inspectors reviewed 15 controlled draw-ings in four locations and found inconsistencies with 7 of the drawings.
This was contrary to Procedure 17-22-3, " Drawing Update," revision 0, which required that controlled drawings reflect the outstanding design changes.
During the inspection, the licensee audited the controlled i
drawings and corrected all the identified problems with EDCR updates, c.
Procedure 0-01-2 allowed the same person to identify a drawing discrep-ency, field verify it, and make necessary changes to control room draw-ings under the drawing change request (DCR) process.
This allows the potential for one person's mistake to go unchecked through the drawing
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change process and be implemented into the controlled drawings.
The inspectors found three examples where the same person had identified, field verified, and changed a control room drawing.
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The failure to follow procedures for the control of drawings was discussed with the licensee and will remain unresolved pending followup by the NRC Region I Office (309/84-15-06).
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8.
Maine Yankee had no procedure for the control of completed EDCRs after they were designated as quality assurance records. EDCR packages that should i
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t have been controlled under the single facility storage requirements of ANSI N45.2.9-1974 were routinely checked out to licensee personnel by the Plant Engineering Department (PED). A review of the checkout sheets indicated that some EDCRs were checked out to individuals who were no longer employed by the utility and others were removed from the files for several months at a time.
This is contrary to Procedure 0-05-3, "QA Records Management Sys-tem," revision 0, which required that removal of QA records from the tech-nical file center or vault be controlled by procedures. During the inspec-tion, the licensee recalled all outstanding EDCRs and developed Procedure 17-208, " Storage and Maintenance of Original Design Change Packages and Original Drawings." This procedure required stricter control and account-ability of EDCR quality records. The failure to adequately store and con-trol design quality documents was discussed with licensee management and will remain unresolved pending followup by the NRC Region I Office (309/85-15-07).
CONCLUSIONS Strengths identified included the establishment of a consolidated design change program between the site and the Yankee Atomic Electric Company organization and
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the programs for closecut of Engineering Design Change Requests (EDCRs) and re-turning the system to operation.
Weaknesses identified in the design change pro-gram included poor control of design information, inconsistencies among sets of controlled drawings, lack of adequate control over design change quality assur-ance records and Plant Operation Review Committee reviews occurring too early in the design change process.
Problems also were identified with the testing and design calculations of individual EDCRs.
This area was rated Category Two.
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QUALITY PROGRAMS OBSERVATIONS 1.
The inspection of quality programs included an examination of the Correc-tive Action Program, the Quality Control Program, and the Operational Qual-ity Assurance (QA) Program. The following provides a general overview of these three programs:
a.
The corrective action program was based on a strong and effective instruction, Maine Yankee Procedure 0-08-2, " Corrective Action Proce-dure," which governed corrective actions for the plant and corporate offices.
This document classified all corrective actions, assigned them to departments for action, put them into one of the Maine Yankee tracking systems, and generally provided a framework for determining root causes and for taking action to prevent recurrence.
One program-matic weakness in the corrective action system was identified and is discussed in Observation 6, below, b.
The Maine Yankee Quality Control section reported to the on-site QA organization and was staffed by technically qualified, experienced personnel. They performed independent QC inspections, material re-ceipt inspections and follow-up on inspection discrepancies and Non-conformance Reports (NCRs).
No weaknesses were identified in the Quality Control Program.
c.
The Maine Yankee QA organization performed in-plant surveillances; provided review of design changes, purchase requests, discrepancy reports and procedures, and provided coordination, input, oversight, and followup to the internal audits conducted by the Yankee Atomic Electric Company (YAEC). These tasks were performed in a generally satisfactory manner. However, several internal audits conducted by YAEC in 1983 and 1984 were considered unsatisfactory in that they did not provide an adequate assessment of the quality of programs audited. Specific examples of unsatisfactory audits are discussed in Observations 2 and 3.
2.
The 1983 and 1984 audits of Technical Specifications and Plant Changes were considered inadequate.
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a.
Audits of Technical Specifications (TS) compliance were provided in Maine Yankee Audit Reports 84-15, October 1984, and 83-15, October
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1983. These audits had the folicwing deficiencies:
(1) Deficiencies identified in checklists were not always identified as deficiencies in the report. For example, item I.3 of the 1983 i
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and 1984 checklist asks, "Are proposed changes to Technical Speci-fications... incorporated into operating and administrative proce-dures?" Both the 1983 and 1984 audit checklists identified that no procedure existed to control this activity, but this issue was not identified as a deficiency in either the 1983 or 1984 audit reports. Additionally, neither the 1983 nor 1984 checklists iden-tified any proposed changes to TS; therefore, whether any such changes had in fact been incorporated into operating and adminis-trative procedures was not known.
(2) The depth and sample size of both the 1983 and 1984 audits were considered insufficient to provide a valid basis for determining acceptability. For example, in the 1984 checklist, item III.8.a asks whether, " Surveillance has been performed to satisfy the re-quirements of TS 4.2 regarding Equipment and Sampling Tests." TS 4.2 requires approximately 35 equipment and sampling tests to be conducted with various frequencies.
The audit checked only two tests, pressurizer level and containment leakage, by review of the control room log for the period of June 1984 to September 1984.
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This sample size is considered too small to determine with confi-
dence that the prescribed tests of TS 4.2 were indeed accomplished during the previous year. Similar examples of shallow audits and inadequate sample sizes were identified in audit checklist 83-15 in verifying compliance with TS 4.6, " Periodic Testing of Safety Injection and Containment Spray Systems, Steam Generator Auxiliary Feed Pumps, and Main Steam Excess Flow Check Valves"; and in both i
checklists 83-15 and 84-15 in verifying compliance with TS 4.12,
" Fire Protection System Surveillance Testing." The sample size inadequacy in the case of TS 4.12 was particularly significant because both checklists (for 1983 and 1984) selected the identi-cal sample:
the diesel fire pump battery surveillance require-ments.
b.
Audits of Plant Changes were provided in Maine Yankee Audit Reports 84-7, October 1984, and 83-78, January 1984. These audits had the fol-lowing weaknesses:
(1) The checklists for audits 83-78 and 84-7 did not audit all the at-tributes required by ANSI N45.2.11-1974. Specifically, the re-quirements of ANSI N45.2.11 to audit test procedure development, paragraph 4.5, and design interface control, paragraph 5, were not included.
(2) Maine Yankee PED arocedures relating to design changes were not audited against t1e requirements of ANSI N45.2.11 to ensure that all required quality attributes were included. Audit 83-78 was conducted for the sole purpose of reviewing the implementation of a newly established Design Change Program.
However, only 5 of 13 new Maine Yankee design change procedures were included in the audit checklist.
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(3) The checklist for audit 84-7 was inadequate because it did not check compliance with the Maine Yankee procedures that relate to design change control.
These procedures, listed below, were not used as references, nor were they included in the checklist.
Procedure No.
Title 00-01-1 Control of Activities Affecting Design 17-21-3 Yankee Nuclear Services Division (YNSD) EDCR 17-22-4 Design Drawing Control Interface 17-23-3 Submittal of Completed EDCR 17-24-4 Qualification Tests (4) Sample size was considered to be too small to provide a reliable assessment.
In audit 84-7, only two EDCRs were reviewed against the complete checklist. Five EDCRs received spot checks against portions of the checklists, and these checks were not recorded on the checklist.
This apparent failure to perform adequate audits as described in a and
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b. above, is contrary to 10 CFR 50, Appendix B, Criterion 18, and Tech-nical Specifications requirements for performing audits. This issue was discussed with the licensee and will remain unresolved pending fol-lowup by the NRC Region I Office (309/85-15-08).
3.
Additional audit weaknesses were identified during the inspection.
Audit coverage of the control of measuring and test equipment was pro-a.
vided in Maine Yankee Audit Report 84-11, September 4, 1984.
This audit was considered generally weak because it failed to identify any of the programmatic and implementation problems associated with the measuring and test equipment program discussed in detail in Observa-tion 4 of the Maintenance section of this report.
b.
Audit coverage of training was provided in Maine Yankee Audit Report 84-5, August 24, 1984. Licensed operator requalification was not au-dited except for a question regarding operator enrollment in the pro-gram. The checklist did not examine the schedule, lecture content, on-the-job training attributes, the evaluation methods, or records required by 10 CFR 55, Appendix A.
4.
The YAEC QA organization performs audits for Maine Yankee under the cogni-zance of the Nuclear Safety Audit and Review Committee (NSARC) as required by Maine Yankee TS 5.5.B.9.
A review of audit reports and NSARC meeting minutes revealed an apparent lack of NSARC oversight and involvement in the audit process. For example, there was no record of actions by the NSARC to ensure the correctness and completeness of the YAEC audit checklists.
In general, the NSARC appeared to perform a passive role with regard to Maine Yankee audits, limiting involvement to a review of audit findings during the semiannual meetings.
The lack of NSARC involvement in the audit process was considered a weakness.
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5.
A programmatic weakness in the corrective action system was that it did not include certain material and procedural nonconformances that occurred under the cognizance of the Operations Department. The most significant erations problems were reported as Plant Information Reports (PIRs) plant op-Unusual
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occurrences and problems of a lessor nature that could require tracking, trending, reporting to management, and action to prevent recurrence were not documented within the corrective action system. For example, valve HSI-M-50 in the high pressure safety injection (HPSI) suction line from the refueling water storage tank was found to be in the open position, rather than in the required shut position, on three separate occasions over a period of several months.
This was apparently caused by inadvertent operation of the valve
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operating switch because of its proximity to another frequently operated
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switch on the control panel. These events were documented in Unusual Oc-
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currence Reports (UORs), but did not become part of the corrective action system, and were not reviewed by PORC.
Maine Yankee uses UORs to document a wide range of abnormal events and fail-ures that occur in plant operations. They are non-proceduralized reports prepared by the on-shift Nuclear Safety Engineer to record the pertinent facts and ramifications of operational problems soon after they occur. The system for preparing UORs contained substantial flexibility.
It was largely based on the professional judgement of the involved engineers, operators, and supervisors.
It was apparent from the quality of the UORs that the ex-isting system was working well.
It was noted, however, that some UORs con-(
tained items (such as the example of the HPSI valve noted above) that should have been entered into the corrective action system to ensure that positive management controls were exercised for resolution and prevention of recur-rence.
The apparent shortcoming of the corrective action system to encom-pass certain problems identified in the plant operations area was consid-ered a weakness.
6.
A weakness was identified regarding the adequacy of the corrective actions taken to prevent stem leakage on a motor-operated reactor coolant system loop isolation valve, HCV-125.
In September 1978, limit switch 3 was jump-ered and limit switch 5 was removed from its circuit to permit the valve to be fully backseated to prevent stem leakage. At the time of the inspection, no other documented action had been taken to determine and correct the root cause of the stem leakage problem.
This valve has been routinely operated with the limit switches bypassed.
IE Information Notice (IN) 82-10 warned that a symptomatic repair such as this might cause damage to the valve or operator, which could impair the safety function of the system.
Despite this IN and the long period during which the limit switches have been dis-abled, no permanent corrective action was taken to correct the valve stem leakage.
The insufficient consideration given to IN 82-10 appeared to be an isolated case. A review of the actions taken as a result of several other ins re-vealed comprehensive followup documented in accordance with Procedure 20-2-1,
" Operational Assessment System."
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A weakness was noted in implementing corrective actions in the plant chemistry
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area. Repeated past deficiencies have been reported regarding the use and L
presence of solutions with outdated shelf lives in the primary and secondary
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chemistry laboratories (Maine Yankee Surveillance Reports 84S-462, October 23, L
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1984 and 845-404, July 1984, and Maine Yankeee Audit 84-2, August 1984). Dur-i ing this inspection an NRC inspector determined that the following chemicals
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with outdated shelf lives were currently in use:
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Lithium standard, expiration date April 1985
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Silicon standard, expiration date November 1984 j
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Potassium phosphate buffer, expiration date April 1985 l
The actions required by Procedure 7-02-1, " Quality Assurance / Quality Control i
for Chemistry Technical Specification Surveillance Tests," revision 6, to
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identify these as expired solutions, to flag their analytic results, and to replace them expeditiously had not been taken.
The use of chemicals with
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outdated shelf lives was discussed with the licensee and will remain unre-solved pending followup by the NRC Region I Office (309/85-15-09).
CONCLUSION The corrective action system was effective because there was a strong procedure that coordinated plant and corporate activities. However, the corrective action system apparently did not encompass certain operations-related problems.
The i
following weaknesses in the implementation of the corrective action system were identified:
(1) out-of-date chemicals were found in use in the chemistry labora-l
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tory, and this issue had been previously identified by the licensee on numerous occasions and (2) bypasses were installed that eliminated the backseat limit
switches on a reactor coolant system loop stop valve despite a warning given in
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an IE Information Notice.
The operational quality assurance program was gener-
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ally satisfactory. However, significant weaknesses were identified in several
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QA internal audits.
Oversight of the QA audit process by the Nuclear Safety Audit and Review Committee was also weak.
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This area was rated Category Two.
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PROCUREMENT OBSERVATIONS
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1.
The Maine Yankee Administrative Procedures (0 series) for controlling the i
procurement, receipt, handling and storage of nuclear safety-related materials and services were an overall strength.
The procurement procedures succinctly detailed the controls and applicable requirements for the administration of the requisitioning process. The procedure for receipt, handling and storage of procured material provided effective guidance for the management of ware-house activities. Additionally, the licensee was in the process of revising these procedures to further streamline the procurement process, facilitate improvements in warehouse areas, and implement a computer enhanced inventory system. Weaknesses were noted in the implementation of the procurement pro-
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l gram procedures, as discussed below.
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2.
Purchase orders (P0s) did not adequately specify packaging and storage re-quirements for safety-related components. A review of 25 P0s revealed that the only requirement identified on these documents for the handling, receipt,
storage, or packaging of the purchased item was the standard phrase, " ANSI l
N45.2.2 requirements apply." Special instructions that provided the details
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for the care of sensitive components were not present.
Interviews of re-
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ceipt inspectors and material handlers revealed that they were unable to de-
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termine which ANSI N45.2.2 requirements were applicable.
l As a matter of practice, components were packaged in plastic bags before being stored.
Desiccants were not utilized unless supplied by the vendor, i
and specific temperature and humidity controls or other special maintenance, j
j storage, or handling controls were not implemented. This resulted in the
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j following items being carried as ready for issue despite their storage prob-l i
lems:
a.
Electronic pressure transmitters were packaged without desiccants being monitored or periodically replaced (P0 34355).
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Electronic terminal blocks and electronic equipment such as an oscilloscope were received and stored without any special con-
trols (P0 32568/39785).
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An emergency diesel generator turbocharger was in storage for several l
l years with no maintenance being conducted (P0 26322).
e Procedure 0-02-01, " Material Equipment and Service Purchase," revision 3, Section 5.2.2 and Procedure 0-02-2, " Maine Yankee Purchase Specifications,"
revision 1, Attachment 1, Section 10.0, both require the Plant Engineering I
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i Department (PED) to specify the detailed requirements for the handling and storage of the components in the Material Purchase Request.
Interviews
with PED personnel revealed that they were not aware of these procedural l
requirements.
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The failure to follow procedures for completing material purchase requisi-
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tions and purchase orders was discussed with the licensee and will remain unresolved pending followup by the NRC Region I Office (309/85-15-10).
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Engineering services for the analyses of safety class systems design changes
were not always procured from qualified contractors.
Interviews with PED j
management revealed that it was common practice to procure specialized en-
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gineering services from contractors not on the approved vendors list.
In one instance, the licensee contracted to a local vendor that was not on the
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approved vendors list to assist with the development of Engineering Design I
i Change Request (EDCR)83-509, "NRV Vacuum Assist System." These services l
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included piping layout reconunendations, stress calculations, piping size
determinations, and inputs for system operating procedures.
The licensee
reviewed the final product, but there was no review or surveillance of the
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contractors program conducted before the contract award.
This appears con-
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i trary to the Maine Yankee QA Plan,Section VII.C., and Procedure 0-02-1, Section 5.2.3, which requires approval of a vendor's QA program before a service contract is issued. This issue was discussed with the licensee and will remain unresolved pending followup by the NRC Region I Office (309/85-l 15-11).
l Interviews with PED personnel revealed that their involvement with the pur-chasing process was limited to the requirements of Procedure 17-24-3,
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" Technical Bid Evaluations," revision 0, which did not address the evalu-ation of the vendor's QA program.
This finding coupled with the lack of
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procedural awareness identified in Observation 2 indicates that there may
be a lack of procedural coordination between PED and station administrative procedures (0 series). The licensee acknowledged that this was a potential j
problem and that they were in the process of reviewing this issue as part i
of their procedure review program.
4.
The inspector toured the warehouse storage facilities and found these areas to be generally well maintained.
Isolated deficiencies regarding the con-
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trol of housekeeping, adhesives, and preservatives were identified during the tour. The licensee acknowledged these findings, immediately corrected the specific deficiencies, and stated that these concerns would be incorpo-rated into the preventive maintenance program being developed to correct the deficiency identified in Violation 85-05-01, issued by NRC Notice of Violation enclosed in Inspection Report 50-309/85-05.
CONCLUSIONS
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A strength was identified in the station administrative procedures for control l
of procurement, receipt, handling and storage of safety-related material
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and services. Weaknesses identified included the inadequate specification
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of storage and handling requirements on purchase orders and the failure to procure engineering services for safety-related systems from a qualified vendor.
This area was rated Category Two.
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RADIOLOGICAL CONTROLS OBSERVATIONS 1.
The program for control of external radiation doses was considered a strength.
For the 10-year period 1974-1983, the collective occupational dose at Maine Yankee averaged 331 man-rems per year, about 32 percent less than the average of all operating pressurized water reactors.
An effective ALARA (As Low As Reasonably Achievable) program was considered to be an im-portant factor which contributed to the dose control program.
It also was observed during inspections and reviews of Radiation Work Permits (RWPs)
that plant management was actively reducing the contaminated areas within the plant.
Plant management considered this action important in order to reduce the volume of low-level radioactive waste and RWP requirements (protective clothing, respirators, etc.) within radiation control areas.
2.
The survey program to support special operations was considered adequate.
However, weaknesses were identified in certain aspects of the routine survey program, a.
Procedure 9.1.8, " Monitoring for Personnel Contamination," established a limit for personnel skin contamination of 300 counts per minute (cpm)
above background as measured by an Eberline RM-14 radiation monitor with an HP-210 probe.
The limit was applicable to personnel exiting the radiation control area.
This limit is significantly higher than the limit of 100 cpm above background generally followed by power reactor licensees using the same or similar equipment. The likelihood that an individual would receive a skin dose in excess of regulatory limits was considered to be small.
Nevertheless, the licensee's pro-cedures would have permitted personnel to leave the plant with contam-ination greater than that which could be detected by the equipment in use.
b.
In order to reduce the volume of waste shipped to burial sites, the licensee sorts low-level solid wastes consisting of paper, cloth, and plastic. Procedure 9.1.22, " Sorting of Primary Side Debris," speci-fied that material that indicated less than 100 cpm above background, as measured with an Eberline RM-14 Radiation Monitor with a pancake probe, was to be disposed of as normal noncontaminated trash. Further,
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the procedure specified that trash sorting shall be conducted in an area where the background dose rate is less than 0.1 mr/hr. During the inspection, it was found, using the licensee's survey instruments,
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that the background in the trash sorting area was less than 0.1 mr/hr, but this corresponded to a level of about 350 cpm. At this background level, it appeared that the licensee would not be able to measure, with
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a high degree of confidence, radiation levels of 100 cpm above back-ground.
Thus, the potential existed for the release of radioactively contaminated materials to normal trash channels.
c.
The licensee's routine program to measure concentrations of radioac-tivity in air in areas other than the containment building consisted of 30-minute air samples collected weekly at single locations in the
'mel building, the spray building, and the letdown area of the primary auxiliary building. A 30-minute air sample was collected monthly in the vent room of the primary auxiliary building. Routine air samples were not collected in other areas of the radiation control area or in any areas outside the radiation control area. No continuous air mon-itors or continuous air sampling equipment were regularly used either in the radiation control area or other noncontaminated parts of the plant. With such a limited airborne survey program, it appeared that the licensee would be unable to detect changes in plant conditions in areas not surveyed and in those areas that were surveyed, during the intervals between the surveys.
10 CFR 20.201(b) requires each licensee to make such surveys as "... are reasonable under the circumstances to evaluate the extent of radiation hazards that may be present." The limited scope of the routine airoorne survey program was discussed with the licensee and will remain unresolved pending followup by the NRC Region I Office (309/85-15-12).
3.
The use of signs and labels within the radiation control area was examined during tours of the area. High radiation areas and contaminated areas ap-peared to be posted properly.
The signs contained additional information regarding radiation work permit and protective equipment requirements.
Ex-amples of inadequate and poor posting and labeling also were noted.
a.
It was observed that waste containers in the radiation control area holding radioactive trash were not labeled with a " Caution - Radio-active Material" sign.
It was not apparent that this constituted a violation of 10 CFR 20.203(f); however, this practice was considered a weakness in that radioactive waste may be mixed with nonradicactive waste and workers could become contaminated.
b.
" Hot Spot" signs were used to identify small areas having radiation levels significantly higher than the surrounding areas.
The follow-ing items were observed:
(1)
Information on some " Hot Spot" signs was as much as 6 months old.
The radiation level indicated on a sign was much higher than that measured by a licensee representative who accompanied
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the inspector.
(2) Some " Hot Spot" signs contained no information regarding the c
radiation level.
(3) Most " Hot Spot" signs showed the radiation level in units of mr/hr. Some signs, however, showed the radiation level in units of R/hr, a possible source of confusion to workers.
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c.
Some equipment was marked with signs stating " Radioactive" or "Contam-inated Material." The use of such signs is not covered in NRC regula-tions or in the licensee's procedures.
d.
A chair within the radiation control area was marked with a " Radio-active" sign.
A licensee health physics technician was observed sitting on the chair. When asked if the chair was radioactive, the employee stated that he did not know.
The employee stated that equip-ment was sometimes marked as radioactive to ensure that it was not removed from the controlled area.
The use of signs denoting the pres-ence of radioactivity for personal or administrative purposes is con-sidered a very poor practice that could result in undesirable effects such as worker disregard of cautionary signs.
The foregoing posting and labeling practices are considered weaknesses.
4.
Weaknesses were noted with the calibration of laboratory counting equip-ment.
a.
No procedure existed to calibrate a Nuclear Measurements Corporation
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gas flow proportional counter used in the radiochemistry laboratory.
This instrument is used to determine the concentrations of liquid rad-waste in tanks before release as effluents. The records documenting the last calibration of this instrument also were not available.
In-terviews revealed that this detector had been calibrated in October 1984 by performance of a plateau curve and 10 1-minute counts with a source of known intensity. A control chart was developed based on the 10 1-minute counts and 3-sigma tolerance limits were determined that cor-responded to approximately 210 percent of the mean.
It was considered a weakness that there was no acceptable tolerance limit for how high the 3-sigma value could be, particularly since no statistical checks (such as the chi square test) were done to verify the repeatability of measurements. A review of recent liquid effluent data revealed that the licensee has been significantly below their regulatory limits.
The failure to provide a procedure to calibrate this instrument was discussed with the licensee and will remain unresolved pending follow-up by the NRC Region I Office (309/85-15-13).
b.
A Tennelec gas flow proportional counter used by HP technicians was not recalibrated after each gas change as required by HP Procedure 9.343, " Operation and Calibration of the Tennelec LB5100, Series II, Automatic Proportional Counter," Revision 0.
This instrument was in use at the time of the inspection and had last been calibrated in January 1985.
Interviews revealed that there was a lack of under-
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standing by the technicians of this recalibration requirement and at least one gas replacement had occurred since the last calibration. A review of the daily source check log revealed that this instrument had
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been operating within its normal range since the last calibration.
The apparent lack of control over the calibration of this instrument is considered a weakness.
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CONCLUSIONS
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Strengths were found in the ALARA program and the control of external radiation exposures. Weaknesses were identified in the routine survey program in the areas of monitoring personnel for contamination, sorting of low-level waste, and air-borne monitoring of plant areas. Weaknesses also were noted in the calibration of laboratory counting equipment. Poor practices were identified in posting and labeling.
This area was rated Category Two.
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UNRESOLVED ITEMS An unresolved item is a potential enforcement finding which requires additional consideration by the NRC regional office.
Observation Area Number Subject Surveillance
Failure of surveillance procedures to require the recording of as-found condi-tions (309/85-15-01).
Maintenance
Failure to adequately implement a measur-ing and test equipment program (309/85-15-02).
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Design Changes and 2 and 3 Failure to adequately control design in-Modifications puts and design information flow between internal and external organizations (309/85-15-03).
Failure to make an adequate design cal-culation (309/85-15-04).
Failure to test a newly installed system before releasing it for operation (309/85-15-05).
Failure to follow procedures for the control of drawings (309/85-15-06).
Failure to provide adequate controls for design change quality assurance records (309/85-15-07).
Quality Programs
Failure to conduct adequate audits in 1983 and 1984 of Technical Specifica-tions and Plant Changes (309/85-15-08).
- Quality Programs
Use of expired chemicals in the chemis-try laboratory (309/85-15-09).
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Procurement
Failure to adequately specify receipt, handling, storage, and packaging require-ments on material purchase requisitions and purchase orders (309/85-15-10).
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Observation Area Number Subject
Failure to provide adequate controls for the purchase of safety-related en-gineering services from an unapproved contractor (309/85-15-11).
Radiological Controls 2.c Limited scope of the routine airborne survey program (309/85-15-12).
4.a Failure to provide a procedure to cali
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brate an instrument used to measure con-centrations of liquid radwaste (309/85-15-13).
MANAGEMENT EXIT MEETING s.
An exit meeting was conducted on June 7, 1985, at the Maine Yankee Atomic Power Station. The licensee's representatives are identified in Appendix A.
The scope of the inspection was discussed, and the licensee was informed that the inspec-tion would continue with further in-office data review and analysis by team mem-bers.
The Team Leader discussed the issuance of an inspection report and advised that the team would draw a conclusion for each functional area inspected and rate the management controls for each area in accordance with the Systematic Assess-ment of Licensee Performance (SALP) Categories.
The licensee was informed that some of the observations could become potential enforcement findings. These would be presented to the NRC Region I Office for followup.
The team members presented their observations for each area inspected and responded to questions from licensee's representatives.
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APPENDIX A Persons Contacted The following is a list of most of the persons contacted during this inspection.
There may be some other technical and administrative personnel who also were con-tacted. All personnel are affiliated with Maine Yankee unless indicated other-wise.
f Corporate Office President
- Executive Vice President
- Assistant Vice President / Manager of Operations
- Plant Manager
- Manager Nuclear Engineering and Licensing
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- Manager Quality Assurance Manager Operational Support Manager Plant Engineering
- Manager Training Manager Operations Department
- Manager Technical Support Manager Haintenance Lead Engineering Assistant, PED Mechanical Engineers, PED Plant Engineering Supervisor, PED Lead Electrical Engineer, PED Lead Performance Engineer, PED Senior Engineer - Mechanical, PED Senior Engineer - Structural Analysis, PED Engineer, PED
- Manager, Maine Yankee Project, YNSD Vice President, YNSD Engineering Supervisor, YNSD-MYP QA Auditor, Yankee Atomic Outage Coordinator Assistant to the Manager, Technical Support Radiological Controls Section Head Chemistry Section Head Radiation Protection Supervisor Lead Radiation Control Specialist Radiation Control Specialists Chemistry and Radiation Control Technicians Chemistry and Radiation Control Trainees Attended exit meeting on June 7, 1985.
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Radiation Control Engineering Aides Radiation Control Foremen Director of Quality Assurance, YAEC
Quality Audit and Engineering Supervisor, YAEC Quality Assurance Section Head Lead Quality Assurance Engineer Lead Quality Control Inspector Quality Assurance Engineers Quality Control Inspectors i
Manager, Operational Support
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Operations Department Manager Assistant to the Operations Department Manager
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Plant Shift Superintendents Shift Operations Supervisors Control Room Operators Auxiliary Operators
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Nuclear Safety Engineering Supervisor Nuclear Safety Engineers Operations Department Administrative Assistant Maintenance Manager
Maintenance Section Head Maintenance Staff Assistant Electrical Maintenance Supervisor Mechanical Maintenance Supervisor Lead Plant Mechanic
Lead Plant Electrician I&C Section Head l
I&C Staff Assistant I&C Supervisors I&C Technicians Outage Coordinator Assistant Outage Coordinator Assistant to the Director, Quality Assurance Operations Training Section Head
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Speciality Training Section Head Operations Training Group Supervisor Simulator Group Supervisor Senior Operations Training Instructors Training Instructors Plant Services Supervisor Material Control Section Head Facility and Construction Coordinator Stores Supervisor Receipt Inspectors
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Storekeepers i
Attended exit meeting on June 7, 1985.
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DOCUMENTS EXAMINED
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The following is a list of the categories of documents examined.
Those specific i
documents referenced in the report are listed by title and the most recent re-i vision, if applicable, where they first appear.
Technical Specifications FSAR Quality Assurance Manual Administrative Procedures (0 series)
Plant Engineering Department Procedures (17 series)
Yankee Atomic Electric Company (YAEC) Quality Assurance Manual Yankee Nuclear Service Division (YNSD) Engineering Procedures (WE series)
PED Goals and Objectives Job Descriptions
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Organizational Charters and Subcharters Engineering Design Change Request Packages Controlled Drawings PED Work Control System Drawing Change Requests i
YNSD Design Calculations
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Approved Vendor List Modifications Training Packages Engineering Service Requests Vendor Technical Manuals Internal Audits NSARC Meeting Minutes t
QA Audit Schedules Nonconformance Reports QA Department Procedures Combined Utilities Assessment Report of the YAEC QA Program
Surveillance Reports Independent QC Inspections QA Quarterly Reports, YAEC Auditor Qualifications, Maine Yankeee and YAEC t
QC Training and Qualification Records Quality Improvement Program Operational Assessment System Program l
Comitment Management System Program
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Maintenance Procedures I&C Surveillance Procedures Plant Engineering Manual Procedures j
Discrepancy Reports
Repair Orders j
Radiation Protection Manual
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Job Orders
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Independent Inspection Checklists l
Measuring and Test Equipment Records i
Machinery History Records VETIP Correspondence s
l Health Physics Procedures
Instrument Calibration Procedures t
Radiation Work Permits Radiation Survey Records Dosimetry Records Operations Training Section Lesson Materials Speciality Training Section Lesson Materials
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INP0 Accreditation Plan 1984/85 Master Training Schedule Training Program Procedures Personnel Training Records Initial and Requalification Examinations Unusual Occurrence Reports Plant Information Reports Licensee Event Reports
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Training Qualification Review Board Minutes Control Room Document Review Books Station Operator Logs Shift Turnover Sheets Bypass Log Surveillance Test Records Equipment Tagging Records Purchase Orders Material Purchase Requests Non-Destructive Examination Procedures Welding Procedures
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APPENDIX B i
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ABBREVIATIONS ANSI American National Standards Institute A0 auxiliary operator l
j A0P Abnormal Operating Procedure ALARA as low as reasonably achievable
CFR Code of Federal Regulations j
CPA conceptual project authorizations DCR drawing change request
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DR discrepancy report ECCS emergency core cooling system l
EDCR engineering design change request i
E0P Emergency Operating Procedure
HPI high pressure injection j
I&C instrumentation and control
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IE Office of Inspection and Enforcement l
IN information notice IST inservice testing l
INP0 Institute of Nuclear Power Operations l
H&TE measuring and test equipment i
MYP Maine Yankee Project i
j NRC Nuclear Regulatory Commission NRV non-return valve i
NSARC Nuclear Safety Audit and Review Committee NSE Nuclear Safety Engineer
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NUREG Nuclear Regulatory Commission Report l
j NUTAC Nuclear Utility Task Action Committee (INPO)
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PAT Performance Appraisal Team PED Plant Engineering Department PIR Plant Information Report
l PO purchase order j
PORC Plant Operations Review Committee i
QA quality assurance
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RCT radiation control technician RO reactor operator RPSCIP reactor pSystem calibration and indicating panel RWP radiation work permit SALP Systematic Assessment of Licensee Performance
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SECY Office of the Secretary (NRC)
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SR0 Senior Reactor Operator
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STA Shift Technical Advisor TQRB Training Qualifications Review Board
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TS Technical Specifications UOR unusual occurrence report VETIP Vendor Technical Information Program f
YAEC Yankee Atomic Electric Company YNSD Yankee Nuclear Services Division
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