IR 05000271/1985098

Forwards SALP Rept 50-271/85-98 for 851019-861231.Facility Performance Remained Generally Consistent Throughout Assessment Period,W/Strong Orientation Toward Safe Plant Operations

ML20215H784
Person / Time
Site:	Vermont Yankee
Issue date:	06/16/1987
From:	Russell W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	Murphy W VERMONT YANKEE NUCLEAR POWER CORP.
References
NUDOCS 8706240212
Download: ML20215H784 (4)

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JUN 161987 Docket No. 50-271 Vermont-Yankee Nuclear Power Corporation ATTN: Mr. Warren P. Murphy Vice President and Manager

.of Operations RD 5, Box 169- 1 Ferry Road-Brattleboro, Vermont 05301 j i

Gentlemen:

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Subject: Systematic Assessment of Licensee Performance (SALP) Report No. l 50-271/85-98 This refers to the evaluation we conducted of the activities at the Vermont Yankee Nuclear Power Station for the period of October 19, 1985 - December 31, 198h, and discussed with members of your staff on March 27, 1987 in a meeting in Vernon,

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Vermont.' The list of meeting attendees is attached.as Enclosure 1. The NRC Region I SALP. Report is provided as Enclosure 2. Our letter of March.12, 1987 (Enclosure-3) forwarded the SALP Board Report and requested comments within 30 days of our meeting. As discussed during the March'27 meeting and subsequently documented in your April 24,'1987 letter (Enclosure 4), you provided additional comments on our j report. i

Our overall assessment of your facility concludes that performance remained gener-ally consistent throughout the assessment period, with a strong orientation toward ,

safe plant operations. We note in particular the organizational and performance !

improvements you achieved in the area of Security and Safeguards. Although per-formance in each functional area is at least at the Category 2 level, we know that you share our goal to strengthen overall operations and thereby achieve even higher

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performance levels. We note in particular your determination to correct problems noted in the Radiological Controls, Maintenance and Modifications, and Emergency

' Preparedness areas. We continue to urge Vermont Yankee management to be aggressive in self-evaluation in all activities required to assure quality and safety in ;

operations, and to vigorously pursue identified opportunities for improvement.

We welcome your comments concerning our evaluation of the individual functional areas. Your comments on the entire SALP Report have provided clarifying informa-tion relative to our assessment and are considered most responsive in identifying )

your intended actions to address noted weaknesses. None of the comments or addi-tional information provided by you will change the rating in the individual func-tional areas; however, we feel certain issues raised by you warrant our comment.

These are discussed below.

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Radiological Controls (Section IV.B, Page 12, Paragraphs 2 and 3). We agree I that based on information obtained after the assessment was completed, the comments concerning the control rod drive modifications do not accurately re-flect the work activity. We will delete reference to that job in our discus-sion of the ALARA issue, as reflected in Page 12 of the amended report.

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Vermont Yankee Nuclear Power 2 JUN 16 W i-Corporation l

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Maintenance (Section IV.C, Page 16, Paragraph 4). -We feel the comments re- !

garding post maintenance testing of IRM detectors are accurate, and the text I will stand as written. The proposed revisions to Procedures OP 4301 and 5307 that will use detector breakdown voltage measurements and a revised functional test method to provide for precritical (preoperational) identification of the ;

types of detector problems noted in July 1986 are responsive to our concerns 1 on this issue.

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Board Recommendations for Maintenance.(Section IV.C.3, Page 18). We provide the following clarification with respect to the need for your review of NRC bulletins and notices: " Vermont Yankee management should assure that present and future actions in response to IE bulletins and notices are fully respon-sive to the NRC requests and that the scope of your reviews fully address the

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stated concerns." Your intentions to audit this activity to assure effective-ness is appropriate and is considered responsive to our concerns.

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-Surveillance area (Section IV.D, Page 20, Paragraph 2). We concur that no further' actions are necessary at this time with_ respect to adding Type B and C leak rate test results into the total containment leakage determination.

However, as you have indicated, procedural modification may'be later necessary as a result of pending NRC actions to revise 10 CFR 50, Appendix J. l l

We consider our meeting and subsequent interchange of information to be beneficial i and have improved the mutual understanding of your activities and our regulatory program. No reply to this letter is required. Your actions in response to'the NRC_ Systematic Assessment of Licensee Performance will be reviewed during' future inspections of your licensed facility.

Sincerely, Original' Signed By WILLIMI T. RUSSELI, William T. Russell Regional Administrator

Enclosures:

1. SALP Management Meeting Attendees 2. NRC Region I SALP Report 50-271/85-98 3. NRC Region I Letter, T. Murley to W. Murphy, March 12, 1987 4. Vermont Yankee Letter, W. Murphy to T. Murley, April 24, 1987

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f _ Vermont Yankee Nuclear Power 3 Corporation M 16 W

REGION I i I

I SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE l

INSPECTION REPORT NUMBER 50-271/85-98 VERMONT YANKEE NUCLEAR POWER STATION ASSESSMENT PERIOD: OCTOBER 19, 1985 - DECEMBER 31, 1986 BOARD MEETING DATES: FEBRUARY 23-24, 1987 i

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SUMMARY OF RESULTS A. Facility Performance Last Period This Period

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(11/1/84- (10/19/85-Functional Area 10/18/85) 12/31/86) Trend 1. Plant Operations 1 1 2. Radiological Controls 2 2 3. Maintenance and Modifications 1 2 4. Surveillance 1 1 5. Emergency Preparedness 2 2 6. Security and Safeguards 2 1 7. Refueling and Outage Management 1 1 8. Assurance of Quality 2# 2 9. Training and Qualification Effectiveness ## 1 10. Licensing Activities 1 1 The SALP performance history since July 1981 is summarized in Table 6.

1. Previously assessed as " Quality Assurance"

1. 1. Not previously assessed as a separate area l

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B. Overall Facility Evaluation i

i The SALP Board assessment confirmed a strong orientation toward' safe y plant operations, with management, staffing and performance' strengths

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most notable in the functional areas of plant operations; outage manage-

ment; security; training; surveillance; and. licensing activities.

The successful and timely completion of the recirculation pipe replace- ,

ment project with minimal unforseen problems during this report period 1 is due, in part, to the extensive planning and preparation noted in the last SALP period. Management strengths at all levels were noteworthy in the completion of the pipe replacement in a manner that assured high.

quality. work and restoration of the plant from a modification construc-tion status. We acknowledge the improvements made in the security: area to correct an adverse performance trend. We further acknowledge efforts to address safety improvements in the Mark I containment for events that are beyond the plant design basis. This demonstrates a strong commitment .I to safety. We encourage-the continuation of the engineering evaluation of the Containment Safety Study recommendations, and the implementation of further improvements, consistent with other safety priorities.

Radiological controls'for the recirculation pipe replacement outage were strong and effective in minimizing exposures and maximizing worker pro-tection while completing the project. The radiological program is also effective in assuring adequate health physi _cs controls for routine plant operations. However, additional management initiatives are needed in the radiation protection area, particularly to improve and complete-for-malization of the ALARA program. Improvements are needed in the facili-ties and function of the Technical Support Center to improve your emer- .i gency. response capabilities. While we recognize the efforts to improve l the quality assurance program, additional attention is needed to address i weaknesses in the effectiveness of program implementation, and to assure

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deficiencies in procured. components are aggressively identified and cor-rected.

While overall performance in the operations, surveillance and maintenance areas remains high, our assessment noted several apparent weaknesses that could result in a reduced level of performance unless they are corrected.

These matters include actions to: (1) address potentially degraded system and/or equipment issues in a timely manner; (2) assure preservice and !

periodic surveillance procedures are adequate to provide timely identi- i fication of deficiencies; and, (3) ensure routine maintenance and sur- !

veillance activities provide a high degree of assurance of continued i satisfactory performance of the primary containment.

As a result of this assessment, NRC activities in Category 1 functional areas are eligible for reduced inspection effort. We will consider the level of performance and initiatives to address identified shortcomings in our prioritization of the inspection program for the~ facility. ~

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IV. PERFORMANCE ANALYSIS i

A. Plant Operations (914 hours0.0106 days <br />0.254 hours <br />0.00151 weeks <br />3.47777e-4 months <br />; 21%)

1. Analysis This area was rated as Category 1 during the previous assessment period, with a conclusion that the licensee continued to demonstrate l a strong commitment to safe plant operations. Areas identified L where improvements could be realized included communications within the operations department and actions to assure that procedure bi-ennial review dates are met.

Plant activities during routine power operation and shutdown periods were reviewed during this assessment period. The licensee continued to demonstrate strong and effective management controls that assured safe facility operations, as evidenced by a good plant performance  !

record, adherence to license conditions and safe operating practices, and a demonstrated commitment to safety by operators and supervisory personnel. Plant housekeeping conditions were excellent. Operators consistently demonstrated a good overall understanding of plant systems and status, and responded conservatively to equipment prob-lems that involve technical specification limiting conditions for operations. Plant management involvement in operating activities .

was evident during log reviews and tours of the facility. Corporate l management was involved in site activities during the refueling outage and in response to operating problems, as evidenced by site visits, facility tours, and meeting attendance.

Plant actions were successful prior to the 1985 outage in eliminat-ing the backlog of plant procedures that were overdue in the bien-nial review cycle. No action was taken to use subcommittees based on the success of the existing review methods. To address concerns regarding the conduct of plant operations review committee (PORC)

reviews of procedures within two years, the PORC review process was revised to assure that standing procedures were acceptable for con- 1 tinued use pending issuance of the new revi ions. However, the i number of overdue procedures increased during the 10-month outage i and reached a total of 70 by August, 1986. Licensee efforts were '

successful in reducing the total procedures past due for review to 58 as of January 1987. The continued review of routine procedures at the corporate level is notable in maintaining the overall quality of plant procedures.

Communications within the operations department impro'.d due to ,

effective management and coordination by operations supervision.

The operations support staff was used effectively to address tech-nical and procedural issues, to provide an interface with other l groups and activities, and to address personnel and administrative I matters. Staffing within the operations department remained accept-able with sufficient numbers of operators to maintain six shifts

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for normal operations. The use of overtime was well controlled, resulting in only occasional use of overtime during routine opera-tions'and moderate use of overtime during the extended refueling and maintenance outage. The completion of a senior operator's cer-tification for the incumbent operations supervisor remained an out-l standing qualification item for this assessment period. Licensee l management should assure that the present plans to obtain-this certification by mid-1987 are completed.

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No generic or overall programmatic weaknesse's were identified during the single set of license examinations given during the :ssessment period. The licensed operator and non-licensed operator training programs have obtained INP0 accreditation. Seven of seven SR0 up-grade licenses and three of three instructor certifications were issued as a result of examinations given in June,-1986. The licen-see completed the simulator facility during the period and began using it (1) in the operator requalification program, (2) in opera-tor refresher training in routine plant operations prior to restart from'the refueling outage, and (3) in the 1986 emer 'ncy_ prepared-ness exercise. NRC inspections identified weaknesses in the train-

'ing of operators on modifications completed during the outage.- The  ;

licensee revised and implemented an improved program that covered i a broader scope of outage modifications and included input from operations supervision. Delays in the startup of the simulator l

' deferred its use for emergency operating procedures (E0Ps) training. *

The licensee accelerated E0P training on the simulator in response to NRC initiatives. Operator acceptance ~of and capabilities for ,

using the E0Ps is good. j One significant item discussed in Section IV.D resulted in a viola- I tion of a licensed condition. The violation involved the inoper- l ability of the standby liquid control system for all of operating cycle XI (August 1984 to September 1985) and was caused by the in-ability.of the redundant squib valves to fire upon demand. The squib valve failure was caused by a manufacturing error in the trigger assemblies. The circuit configuration provided an erroneous indication that squib valve firing continuity existed. The NRC staff concluded that the event was caused primarily by the unin- .;

tended primer wiring change by the vendor, and secondarily by the '

licensee's procurement and post-installation testing program that did not detect that change. The licensee's aggressive response to resolve the technical issue, complete comprehensive corrective ac-tions, and to promptly recognize and report the potential generic l concerns, were indicative of a strong management commitment to j safety.  !

Control room protocol was maintained conducive to safe operation through the control of access, limiting business to plant operating activities, and through plant management controls that restrict un-related reading material or other potentially disruptive activities. )

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A control room. radio,.that had management approval, was removed in 1 response to-initiatives by the NRC.' There is good morale among the operators, and a good sense of cooperation will all plant depart- j ments. Operator. performance during:startup and routine' operations

' was noteworthy and reflects the success of the operator training program and, in particular, the simulator training provided.to the

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l operators prior to startup from the 10-month outage. The startup '

from the outage was completed in a cautious and orderly manner.

There was.a notable improvement in th'e quality.of licensee event.

reports,-which demonstrated the licensee's responsiveness to staff comments'from the last assessment period. A detailed evaluation-of LER quality using a sample of 15 LERs was made by'the NRC's Office for Analysis and Evaluation of Operational Data (AE0D) using the basic methodology presented in NUREG 1022, Supplement 2. In-general, the LERs.were found to be above av'erage in quality, based i on the: requirements contained in 10 CFR 50.73. LER quality improved-toward the end of the assessment period following the use of new procedures developed in response to the first set of AEOD comments regarding the previous assessment period. No adverse trends-were noted in this functional area from a review of the LER history.

A generally conservative approach:is taken for reports made under i 10 CFR 50.72 and 50.73.

There.were two LERs submitted during.the assessment' period involving reactor scrams during plant operation. One scram was caused by de-fective equipment, and<the other event was caused by a combination ,

of inadequate procedures and operator performance in an operational '

mode that is rarely used - reactor critical and at operating' tem.

perature with the MSIVs closed for plant maintenance. The overall scram rate for the limited period of-plant operation during this-assessment'(from June 30-December 31,.1986) was low at n.46 per 1000 critical hours.- There were no scrams with the reactor at power.

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In summary, the strength of the licensee's management controls'in the operations area is demonstrated by the overall good plant per-formance record which included only two reactor trips, the minimal number of personnel errors, the commitment to training, the con-servative approach taken to equipment operability issues, and the overall good performance in adhering to license conditions.

2. Conclusion Category 1 i

'3. Board Recommendations l

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B. . Radiological Controls (411 hours0.00476 days <br />0.114 hours <br />6.795635e-4 weeks <br />1.563855e-4 months <br />, 10%)

1. Analysis The radiological controls area was rated as Category 2 during the previous assessment period. Program weaknesses included the lack I'

of adequate radiation safety instruction to a worker that resulted-in an unplanned radiation exposure of the worker in the TIP room, and the lack of a formalized "as-low-as reasonably-achievable" 4 (ALARA) program. l

Seven routine inspections were conducted during this period in the d area of radiological controls. These inspections focused on the ]

implementation of radiological controls during the recirculation j piping replacement project (RPRP) outage and during routine, non-

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outage conditions. Radiological controls were also reviewed during the restart team inspection.

One outstanding regulatory issue, relating to the placement of the ,

containment high range monitors (NUREG 0737 Item II.F.1.3), was resolved during this period by NRC acceptance of the detector loca-tions, as discussed in an NRC safety evaluation dated March 25, 1986.

Radiation Protection The external exposure controls program was acceptable, with suffi- l cient quantities of radiation protection equipment provided in good condition to support the outage as well as normal operations.

Radiation safety records including surveys and personnel exposure records were generally complete, well maintained and available.

There was good interfacing of health physics (HP) personnel with l other plant operating groups, and good integration of HP controls with routine work activities. The station HP group has sufficient stable and experienced staff for routine activities.

The radiation protection organization was augmented for the outage !

and provided positive control over work activities and minimized the number of unplanned exposures. The training and qualification program for the outage provided an adequate understanding of work and adherence to procedures, resulting in few personnel errors.

The effective implementation of radiological controls during the pipe replacement outage resulted from strong management support for the routine daily aspects of the program and sufficient staffing to support the activity. '

Minor lapses in achieving industry standards and in technical over- j sight of the internal exposure control program were noted during 1 NRC inspections. For example: (1) whole body counts, indicative j of small uptakes of radionuclides, were erroneously judged to be I skin contamination; (2) numerous instances of anomalous data were l

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o d in the 101 body counter logs, which were not investigated i o viewed by he 'censee; (3) control charts for the whole body {

co t were im roperly established, and the individual responsible for h whole b dy counter did not understand the use and interpre- {

tatio o 01 chart information; and (4) review of the respira-tory p tection program found practices, equipment use and storage that we ot d and not in keeping with acceptable current in- ;

dustry s nc r and practices. These examples are indicative of l a lack of 1 1 anagement attention to detail and a weak commitment i to industry ;a rds for program upgrade. ]

The licensee's a ity to resolve the long-standing ALARA issue continues to be N C concern. The item remains unresolved long after deficienci were hrnunh+ to the licensee's attention during previous SALPs an in ections da back to 1983. During the current period, ALA A implementing pro ures remained in draft form. ;

The licensee create a ew "ALARA Coordi tor" position, but failed {

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to fill the position, planning, work package l preparation, and the d e- racking system for the RPRP outage was l provided by the prime co tr W , the 1800 person-rem expended for l the pipe replacement occu re v in part, because of the partially I successful primary system comtgination (DF of 3.1 instead of 5.0),

and due to the increased wo hours needed to complete work on the N2 nozzles. However, the lac of-Qtive and timely plant staff review, understanding, and eva uatioh of the contractor's work packages and dose-saving techni esplsocontributedtothein- 'l creased expenditure.

Contractor activities associated wi t trol rod drive insert / '

withdrawal line hanger modifications i benefit from historical radiation dose rate data. As a resul i ' easing radiation fields were experienced as the plant restarte w ulted in unexpected additional exposure to contractor person 1, d which, similar to l the RPRP outage, made the original dose e imates ina urate. Pre-planning for significant radiological work ctivi ies f the up-coming 1987 outage and for the planned spent fuel '- 1 e 3ansion project continues to be fragmented without cl L A go pls, ob-jectives and procedures.

With the exception of the internal exposure contr Ts ea, the lic-ensee's audits of the radiation protection program er echnically sound and thorough. Plant responses to audit defi c were generally timely and corrective actions were accepta he lic-ensee has identified the need to improve the timelines o lant responses to Manager of Operations implementation direc es for audit " observations" in the radiation protection area. is matter is discussed further in Section IV.H, Assurance of Qualit,

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12a noted in the whole body counter logs, which were not investigated :

or reviewed by the licensee; (3) control charts for the whole body l counter were improperly established, and the individual responsible i for the whole body counter did not understand the use and interpre-tation of control chart information; and (4) review of the respira-tory protection program found practices, equipment use and storage that were outdated and not in keeping with acceptable current in-dustry standards and practices. These examples are indicative of a lack of plant management attention to detail and a weak commitment to' industry standards for program upgrade.

The licensee's inability to resolve the'long-standing ALARA issue continues to be an NRC concern. The item remains unresolved long after deficiencies were brought to the licensee's attention during previous SALPs and inspections dating back to 1983. During the .

current period, ALARA implementing procedures remained in draft form.

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The licensee created a new "ALARA Coordinator" position, but failed to fill the position. Even though pre-job planning, work package preparation, and the dose-tracking system for the RPRP outage was provided by the prime contractor, the 1800 person rem expended for l the pipe replacement occurred, in part, because of the partially I successful primary system decontamination (DF of 3.1 instead of 5.0), I and due to the increased work hours needed to complete work on the j N2 nozzles. However, the lack of active and timely plant staff ;

review, understanding, and evaluation of the contractor's work l packages and dose-saving techniques, also contributed to the in-creased expenditure.

With the exception of the internal exposure controls area, the lic- i ensee's audits of the radiation protection program were technically j sound and. thorough. Plant resporses to audit deficiencies were :

generally timely and corrective ections were acceptable. The lic- l ensee has identified the need to improve the timeliness of plant responses to Manager of Operations implementation directives for audit " observations" in the radiation protection area. This matter is discussed further in Section IV.H, Assurance of Quality.

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Radioactive Waste Management / Effluent Controls The licensee showed initiative by developing special procedures and implemented an effective radioactive waste program to support rad- <

waste operations associated with the recirculation pipe replacement outage. Special procedures were developed to ensure solidification

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quality control for resins used in the pipe decontamination, analy-ses were conducted to properly classify the.special waste streams associated with these resins, and a special process control program was instituted to ensure proper solidification. These efforts re-sulted in the successful completion of disposal operations.

Weaknesses were noted in the radiochemical measurements program.

Although NRC analyses of actual effluent samples verified the lic- l ensee's ability to measure radioactivity in effluents with respect to Technical Specifications and other regulatory requirements, guid-ance for laboratory quality assurance / control was not proceduralized. I Weaknesses noted in radioanalytical procedures were indicative of ;

occasional inattention to technical detail in procedure review, ;

licensee emphasis on conservatism at the expense of radioanalytical '

accuracy, and a failure to implement commitments to NRC Regulatory Guide 4.15.  ;

Chemistry The program to monitor chemical parameters in various plant systems with respect to Technical Specifications and other regulatory re- ,

quirements was reviewed. Although the results of the standard !

measurements comparison showed the licensee was in general agreement !

with NRC standards, continuing weakness in the measurement control program was noted, which suggests an incomplete resolution of tech- t nical issues identified in earlier NRC reviews. The license had initiated a measurement control program for the laboratory, but had not included control charts for boron, chloride, iron, nickel, cop-per and chromium measurements. This finding indicated an incomplete technical review of the previous NRC position and a continuing weakness in laboratory analytical quality control. i Transportation i

Review of ongoing transportation activities indicated the licensee 1 maintained the effective radwaste packaging and shipment program noted in previous assessment periods. Personnel adherence to pro- 3 cedures and shipping requirements was good. Program improvements j in the quality assurance area were continued with the development <

and implementation of procedures to support the shipping of unique l radioactive wastes associated with the pipe replacement. Close i management attention to planning and implementing the program was j noted with strong peer review of the technical aspects of prepara- ,

tion, packaging and shipping activities. However, one incident at j i

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the end of the assessment period involved a resin. shipment received ;

at the Barnwell site with contamination levels on the outside of'- .

the inner liner that exceeded the burial site requirements. NRC '

followup of this item, documented in the subsequent assessment period, showed that management's corrective actions were extensive and thorough to prevent recurrence.

Summary The licensee implemented an effective radiological controls program to support the recirculation pipe replacement outage work activities. ,

The routine radiological controls program is generally well imple-mented with an adequate staff, that is integrated with plant work control activities. NRC staff concerns relative to the establish-ment of a formalized ALARA program, with a dedicated staff, approved ALARA procedures, and systems in place for goal setting, pre-job planning, and man-rem tracking are still not resolved as of the end of this assessment period. There are some indications of a lack of plant management oversight and attention to technical detail in <

the areas of internal exposure control, chemistry, and effluent measurements.

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Category 2

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15 l C. Maintenance and Modifications (441 hours0.0051 days <br />0.123 hours <br />7.291667e-4 weeks <br />1.678005e-4 months <br />, 10%)

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This area was rated as Category 1 during the last assessment period. ;

The licensee completed actions during the current assessment period .

to assure resolution of pipe support issues per IE Bulletins 79-02 i and 79-14, in accordance with the previous SALP Board recommenda- i tions. Licensee actions on vendor interface controls is. discussed further below, i

The routine preventive and corrective maintenance programs were re-viewed during the present period, and there were several inspections ;

of the modification program and plant modifications.

The routine preventive and corrective maintenance programs remain a licensee strength, as evidenced by no NRC-identified programmatic deficiencies or trends of maintenance problems, and good plant re-liability. There were no instances where component operability was lost as a result of maintenance activities. Routine maintenance procedures were well implemented by a qualified staff at both the craft and supervisory levels. Maintenance of safety-related equip-ment is given proper priority and problems are trended to assure early detection of problems. Evaluations regarding inoperable or degraded components were consistently proper. Safety related sys-tems were properly controlled prior to and after removal from ser- ;

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The licensee was aggressive in the technical review of equipment problems to establish and correct root causes. Thorough review and followup of identified deficiencies resulted in the identification of potentially generic deficiencies that were filed as Part 21 re-ports for Conax squib primers (LER 86-04), GE penetration assemblies (LER 85-10), NAMC0 contact blocks, Limitorque motor operators (LER 86-12), and ASCO rebuild kits. Licensee efforts to improve motor-operated vaP reliability by incorporating M0 VATS testing into the routine preventive maintenance program starting in the 1985 outage were noteworthy.

i Several potentially significant equipment operability issues were identified during operating periods and the outage. The issues I included (1) potentially cracked wear rings in the residual heat ]

removal (RHR) pumps, (2) potentially inadequate minimum flow for the residual heat removal and core spray (CS) pumps, (3) the de-graded floor of the condensate storage tank, (4) crack indications !

in both core spray (CS) nozzles, and (5) the turbine building block j wall discrepancies. While acceptable interim measures and correc- !

tive actions were implemented where necessary, and plans are in !

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! progress to address each item, these matters require additional licensee attention to assure acceptable, long-term resolutions on a timely basis.

Two instances were noted, discussed below as LER 86-12 and in Sec-tion IV.D as LER 85-07, where maintenance practices contributed to operability problems. As noted in Section IV.F below, maintenance practices also resulted in two reportable security events. The Limitorque deficiency concerned the hydraulic lockup of torque switches due to lubricating grease in the spring pack area, and was identified from the followup evaluation of the failure of the B re-circulation pump suction valve (LER 86-12). The lockup condition occurred due to a combination of the spring pack internals design and a new, more viscous, vendor-approved grease added to valves during the outage. Actions were taken to modify valves with the new grease, and additional actions are planned to address other valves.

However, the Limitorque fix installed in 1986 was a change that was part of a component modification included by the vendor in his pro-duct since 1975. In another issue concerning the Bingham RHR and CS pumps, the licensee became aware in 1986 that the vendor had recommended, in the 1979-1980 time frame, new pump minimum flow values. The fact that the licensee was unaware of these vendor suggested product changes appears as examples of deficiencies in the program for the control of vendor information. This matter remains open from the last SALP assessment.

Post-maintenance testing was effective to assure operability of systems prior to return to service, with three notable exceptions.

The first involved the failure to " bench test" the control rod air solenoid valves after rebuilding them during the outage. Faulty ASCO solenoid parts were subsequently identified by scram timing tests performed during the startup. However, appropriate component testing could have provided earlier detection of the defects prior to return of the reactor protection system to service. A second example concerned the failure of post-installation tests to identify three failed IRM detectors and the cross connection between two IRM channels following replacement of all six detectors during the out-age. The IRM problems caused the June 30, 1986 reactor scram. The last example concerned the failure to complete a local leak rate i test on a reactor water cleanup system valve following maintenance  !

that disturbed the valve seat (LER 86-16). No testing was conducted I because the completed work exceeded the designated scope of repairs, and the job supervisor was not cognizant of the full extent of the worker's actions. Although no discernible trends were identified, the above cited examples of post-maintenance testing discrepancies are significant, and indicate a need for more thorough reviews by worker and supervisory personnel to assure testing will verify sys-tem operability following repairs. i I

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There were no violations cited in the maintenance area during the assessment period. A related violation discussed in Section IV.H, l that was indicative of poor corrective action followup concerned' i the failure to preclude the installation of potentially defective '

NAMC0 contact blocks in main steam isolation valve position circuitry following maintenance on the switches. .The three LERs submitted for this area showed no adverse trends. The relatively low number of personnel' errors is noteworthy.

Two issues in the maintenance area are indicative of a lack of man-agement responsiveness to NRC initiatives. The first concerns the !

response to Generic Letter 86-06 for which no actions were taken ,

to review quality assurance measures applied to nonsafety-related GE AK 2-25 field breakers used in the ATWS mitigation circuitry.

Further actions are necessary to verify QA measures applied to the breakers to assure reliability of the trip system per 10 CFR 50.62 requirements. The second concerns the potential for wear ring j cracking in the Bingham RHR pumps. Considerable NRC initiatives i were required to solicit a commitment to inspect'the pumps. A com- ]

mitment was made to start the inspections by the.end of 1986, pend- .)

ing the processing of a technical specification change to allow two *

week RHR pump' outages, and pending the development of a long-term fix. The start of work was deferred due to vendor production prob- 1 lems on producing acceptable replacement impellers. Licensee man- l agement initiatives were' evident to assure quality components would i be provided by the vendor.

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The design change control program remained a significant licensee-strength, as evidenced by the implementation of the following major. I design changes inspected during the outage: recirculation pipe re-placement; core spray nozzle overlay repair; and, seismic pipe ~sup-ports. The planning, scheduling, special projects organization, a and management and quality assurance controls instituted for the i recirculation pipe replacement project were effective in assuring a quality output. Management involvement in the pipe replacement was consistently applied at the right level, particularly in early 1986 when problems were experienced with the N2 nozzles welds. The recirculation hanger restoration program was effective with three independent levels of review and a good system for prompt identifi-cation and correction of deficiencies.

Two violations and one deviation were identified in the modifica-tions area regarding the response to masonry wall issues raised in IE Bulletin 80-11. The bulletin requirements were not appropriately addressed during the 1980-81 time frame, in that technical bases for analytical assumptions were not completely verified, block wall surveys were completed without the benefit of written instructions and criteria, and the informal 1980 survey results were not retained as a permanent record. Masonry walls were resurveyed using detailed

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written instructions and criteria'. The current activities displayed-a more rigorous approach'and were performed acceptably. The resur-

'vey validated the-basic adequacy of the 1980 work, but identified four additional walls in the heating and ventilation corridor that-had not been identified previously in the IEB 80-11 response. The-seismically induced failure of the walls.could affect controls'for safety-related electrical equipment located in'the turbine building, a non-seismic structure. Actions were in progress ~at the end of ;

the assessment period to develop a corrective action plan and.

schedule. The findings for the 1980 work do not appear indicative of programmatic weaknesses for present-day activities, based on other inspections of modification activities during the assessment period.

i In summary, the maintenance and modification program remained a significant strength as evidenced by overall plant reliability, the adequacy.of completed design change work, and the thoroughness of work by an experienced and qualified staff to identify and correct problems as they occur. However, improvements are needed in the areas of: (1) post-maintenance testing, (2) updated-vendor informa-tion, and (3) followup of NRC-identified concerns.

2. Conclusion Ca+egory 2. l 3. Board Recommendations Licensee Resolve the masonry wall and RHR pump inspection issues ev.peditiously.

Review program for control of vendor information and change as necessary to assure that vendor information is current.

,

Review the status of actions taken in response to IE Bulletins and Notices to identify areas requiring additional work or attention.

NRC ,

Assure that an adverse trend does not develop in the licensee's i response to NRC initiatives.

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l D. Surveillance (242 hours0.0028 days <br />0.0672 hours <br />4.001323e-4 weeks <br />9.2081e-5 months <br />, 6%) I

1. Analysis i

This area was rated as Category 1 during the previous assessment I period. j i

The operational and outage surveillance testing programs were re- '

viewed during the present assessment period. Activities inspected included the surveillance testing and calibration control program, ,

the containment integrated leak rate testing (CILRT) for the 1986 outage, the Cycle XII startup test program, and the power ascension- ]a test program.

The surveillance program is effective and well-controlled as evi-denced by the master surveillance schedule which is maintained up-to-date and reflects the latest changes. An effective program is in place to ensure the latest Technical Specifications amendments are incorporated into the surveillance test schedule and test pro-cedures. The records were well maintained, accessible, and complete. ;

Quality control review of test procedures and witnessing of test i activities was well documented in quality assurance inspection re- q ports. Surveillance tests were noted to be completed on time.

q l

Surveillance procedures were technically adequate and provided the '

right amount of detail to assure correct performance. Weaknesses identified in the CILRT and standby liquid control system (SLC) pro-cedures are discussed further below. The licensee efforts to up-grade procedure quality using the INPO guides for format and content are notable. Surveillance activities were completed in accordance with established controls and the success of management initiatives to encourage procedure adherence is evident. Planning and staffing for surveillance activities were adequate and test personnel are experienced, knowledgeable of the facility and equipment under test and show good regard for administrative policies.

The two Level V violations in this area resulted from personnel

>

performance errors and/or procedural inadequacies. None of the violations involved significant safety concerns; they were not con-sidered indicative of training or programmatic deficiencies. Cor-rective actions in both cases were prompt, effective, and went be-yond minimum acceptable actions. The Level IV violation resulted from a December 1986 inspection and is presently under review by the licensee and the NRC. The issue involves a matter of interpre-tation of the Section XI requirements on how to use the 96-hour period to analyze test data and when to declare components inopera-tive when test results are in the " required action" range. While resolution of the item will have programmatic implication for in-service testing, there was no significant safety impact on plant operations.

y F.

The Level III violation for this area was' described in LER 86-04 and concerned the discovery during annual surveillance testing that both SLC squib valves were inoperable. The discovery was made by the insitu firing of.the primers that had been installed for operat-Ling Cycle XI and were scheduled to be replaced. The problem was caused by a manufacturing error in the trigger assemblies that made them incompatible with the facility firing-circuitry. The surveil-lance procedure was inadequate because neither the preservice con-tinuity. checks nor the post-operating cycle firing could identify mis-wired primer charges prior to use in the SLC squib valves. The procedure instructions to check continuity. lacked sufficient detail i to assure.that continuity across the appropriate terminals was j measured. The procedure had been used successfully for several 3 years.for situations that did.not involve unintended product changes j or loss of configuration control. Licensee actions were comprehen- g sive and effective to correct the deficiency, and to assure that {

no other surveillance procedure that provided an after-the-fact ~

operability demonstration contained similar weaknesses. This event is considered to be an isolated example of a surveillance procedure inadequacy that resulted in an LC0 violation, and is not indicative

) i of programmatic weaknesses in the surveillance area. The opera-tional surveillance test program is thorough and technically sound, and remains a significant strength in the licensee's administrative controls to assure safe plant operations.

NRC inspection of the containment integrated leak rate test com- l pleted in June 1986 identified procedure and personnel training ,

weaknesses. None of the discrepancies created a problem in the ;

. performance of an adequate Appendix J leak rate test. Licensee )

management was responsive to address the NRC concerns and to correct the identified discrepancies. One procedural matter still open from the last SALP period concerns the need'to add Type B and C leak rate test results in the total containment leakage determination.

Two LERs concerned ccatainment leak rate test inadequacies: LER 86-01 concerned the discovery that two demineralizer water system valves were never included in the Type C leak rate test; and, LER 85-07 concerned the identification during Type C testing of eight 1 valves with excessive seat leakage. The demineralized water valves were administrative 1y controlled closed and thus had negligible impact on containment leakages. A failure analysis for eight of the nine Type C test failures identified no repetitive failures or program inadequacies. The failure of the outboard check valve in the B feedwater line was recurrent (LERs 84-11 and 83-10), which showed that the previous failure reviews were inadequate to identify the root cause. The technical evaluation completed for the 1985 test failure was thorough to correct the root cause poor mainten-ance practices when installing the val a.

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i LER 86-07 concerned the' failure of.the June 1986 Type A c'ontainment integrated. leak rate test. -The failure resulted principally from leakage in the drywell . head manway cover, and identified the need

.

to revise maintenance procedures to tighten the cover bolts after- l moving the head in place prior to the Type A test. The cover had

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not been' removed during the outage and previously had passed a Type B leak rate test. There were~no safety concerns for containment leakage in the as-found condition for the last operating cycle. l The LERs reported in this area revealed no adverse trends or common ;

causes.- The overall performance-of routine duties by. plant workers. l was good with low numbers of personnel. errors. Previous actions !

by plant management to address personnel errors have been effective. j In summary, the inservice and operational surveillance programs re--

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mained effective in demonstrating continued operability of-safety systems. The single exception concerning the failure ~to' identify defective squib valves prior to operating Cycle XI is noteworthy j but an isolated exception to an otherwise excellent program. Plan- 1 ning and staffing for the surveillance programs were adequate, and i personnel were' experienced and. knowledgeable of the facility and ]

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program requirements. Improvements required in the containment. -,

surveillance procedures are not indicative of programmatic weak- '

nesses. ,

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- 2. Conclusion

Category 1. l

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3. Board Recommendations  !

None. j

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22 1 E. Emergency Preparedness (200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />, 5%)

1. Analysis 4

!

Licensee performance in this area was rated as Category 2 (improving)

during the previous assessment period. Marginally acceptable per-formance was documented during the 1985 annual exercise as a result of weaknesses in command and control, and in the flow of information ;

between various licensee response facilities and between the lic- i ensee and offsite agencies. An improved response to NRC-identified deficiencies was demonstrated during the last few months of the period.

During the current assessment period, one partial-scale exercise i was observed, two routine safety inspections specifically related '

.to follow-up of previous deficiencies were conducted, and changes j to emergency plans and procedures were. reviewed. 1

,

The two inspections performed in April and July,1986, generally related to the follow-up of corrective actions taken as a result f

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of the 1985 emergency exercise. The licensee's actions were deter- i mined to be consistent and progressing according to schedule. Dur- )

ing the July inspection,'NRC concerns were raised about basing Pro- l tective Action Recommendations (PAR's) primarily on dose, instead W of NRC guidance which emphasizes consideration of rapidly degrading plant conditions. This had been previously brought to the licen-see's attention in a 1983 Information Notice. The licensee acknowl- 1 edged the concerns and has initiated appropriate corrective actions 1 to revise the affected EP implementing procedures. I

A partial participation exercise was conducted on December 3, 1986.

The licensee demonstrated a satisfactory emergency response cap-ability. Actions by plant operators were prompt and effective. ,

Event classification was accurate and timely. Personnel were l generally well trained and qualified for their positions. No sig- l nificant deficiencies were identified. There were, however, recur- !

ring weaknesses noted in the operations of the Technical Support Center (TSC) including: (1) lack of coordinated actions between the ;

control room and TSC personnel; (2) failure of TSC personnel to aggressively follow and coordinate plant activities; and (3) exces-sive noise levels in the TSC.

In summary, Vermont Yankee's performance indicates that training has been effective. Management involvement has been generally i effective as evidenced by timely completion of corrective actions l and improved responsiveness to NRC concerns. However, improvements are needed for the Technical Support Center facility and emergency response functions. A dedicated emergency preparedness coordinator (EPC) has been detailed from the Yankee Atomic Electric Company effective December 1986 and a dedicated on-site EPC position is

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l expected to be filled in March 1987. The lack of a dedicated on-site EPC with appropriate staff has hampered the licensee's' effec-tiveness in efficiently maintaining and implementing its emergency preparedness program.

2. Conclusion i

Category 2.

l 3. Board Recommendations l

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None.

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F. Security and Safeguards (79 hours9.143519e-4 days <br />0.0219 hours <br />1.306217e-4 weeks <br />3.00595e-5 months <br /> ,2%)

1. Analysis This area was rated as Ca.tegory 2 during the previous SALP assess-ment period. Two areas requiring improvement were identified: in-sufficient management oversight of the security program, and the need to upgrade security systems and equipment. As a result of violations during that period, which included one Level III viola-l tion (escalated enforcement - no civil penalty), the licensee in-itiated aggressive actions during this assessment period to address those concerns and to improve the program overall.

Orie routine physical security program inspection was conducted dur-ing the assessment period. Resident inspector review of program implementation continued throughout the period. No violations of program requirements were identified. Actions were completed during this period to satisfactorily address previous assessment concerns regarding responses to alarms and control of isolation zones. Short-term measures were established and maintained, and long-term actions are in progress per established commitments to fully replicate cen-tral alarm station (CAS) functions in the secondary alarm station (SAS).

Plant and corporate management attention to and support of the pro-gram were significantly increased during this period, resulting in improved oversight, more attention to maintenance and upgrading of systems and equipment, establishment of more effective supervision, and the establishment of a technical working group, composed of various disciplines, to assess the performance of the security sys-tems and to maintain them in a state of high reliability. Addi-tionally, the importance of strict compliance by all plant employees to security procedures was strongly emphasized by management.

Measures were taken to improve interface and communications between the security contractor and licensee security (proprietary) super-vision, and between the security group and other plant functional groups. These measures included the formation of a security task group to make recommendations on and/or resolve specific security problems when they arise and by requiring security supervisors to attend plant status and planning meetings and other significant meetings that could impact the security program. This increase in management involvement is evidence of the licensee's desire to im-plement an effective and high quality security program.

At the licensee's initiative, management representatives met with l the NRC Region I safeguards staff on two occasions during this l period to discuss improvements to the security program and to obtain '

a clear understanding of pertinent NRC performance objectives. A comprehensive summary of program improvements and their status was provided to NRC midway through the assessment period. Both the !

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meetings and the summary proved beneficial in providing a clear 4 understanding of management's short- and long-term plans for the '

security program. The summary also provided clear evidence of the licensee's aggressive pursuit of planned goals and its intent to i meet commitments.

i Shortly before the beginning of this assessment period, the licensee assigned a new proprietary supervisor to -direct and oversee, on a routine, daily basis, the activities of the contract security force.

The incumbent rapidly assumed the duties and responsibilities of the position and, in large part, is personally responsible for in- ,

iti ling cnd/or pursuing many of the program improvements discussed in th s section.

Staffing of the security group and program implementing procedures are consistent with the commitments in the NRC - approved security -

plan and appear to be adequate for the workload. Authority and responsibilities for members of the security group are clearly de-fined and well disseminated. Security force members were found to be knowledgeaV e of their assigned duties and to exhibit a good appearance and a professional demeanor. Also, morale was observed <

to be high. i

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Facilities were found to be well maintained with sufficient space allocated for the operational needs of the program. Records were also well maintained and readily' retrievable and record repositories were found to be secured in accordance with Safeguards Information requirements. The resources allocated for the administrative, technical and logistical program support appeared to be adequate, but should be monitored by management to avoid program degradation in the future.

The security force training and requalification program is well developed and administered by three qualified instructors. The j program was enhanced during this period by routinely evaluating and l feeding information gained from program performance and program j changes back into the training program. The same technique was used l to increase the effectiveness of security training for other plant 1 personnel. The licensee also increased the number of contingency plan drills during this period. These drills serve as a reinforce-e nt of the training program and sharpen performance of the security force members.

The security audit program was modified during this period. A new corporate audit plan was developed utilizing generic NRC Physical Protection Program inspection criteria, and was implemented during this period. Security management also used these criteria to ac- i complish surveillance of security program implementation on a rou- I tine basis. It appears that the combined effects of the new audit plan, surveillance of program implementation and increased drills have enhanced performance of the security force, as no performance-l i

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26 l

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related problems were identified during this period. These initi- 1

- atives are further indication of the licensee's desire to maintain '

an effective and high quality program.

A total.of six security event reports, required by NRC regulation, were submitted to the NRC during this period. Two events resulted from a computer software problem for which short term corrective ,

action has been taken and.long term resolution is planned; two re- I sulted from maintenance oversights;.and the other.two resulted from {

a long standing condition for which corrective action is now being !

pursued. The reports were submitted in a timely manner and all indicated that appropriate compensatory measures were promptly and correctly implemented. .The reports were clear and comprehensive,-

which permitted rapid review and evaluation by NRC of the events.

The reports clearly represented an improvement over reports sub-mitted in.the previous assessment period, in that all event reports clearly identified the cause of the event. The licensee's program for identifying and reporting security events is considered effec-tive and improving.

During the assessment period, the licensee submitted six changes to-the Security Plan and one change to the Contingency Plan under the provisions of 10 CFR 50.54(p). During November, 1986, the

,

i licensee also submitted a proposed change to the Security Plan under-the provisions of 10 CFR 50.90 to address findings identified during a Regulatory Effectiveness Review and an NRC Region I inspection.

Plan changes were discussed with regional licensing personnel prior to submittal to ensure a clear understanding of the intent of the changes. All changes were of high quality. .The licensee's efforts during the period were concentrated on upgrading the security; pro-gram, eliminating weaknesses, and making plans and procedures clearer and consistent with NRC performance objectives, as evidenced by these licensing activities. Plant security personnel were ac-tively involved in'these activities and are knowledgeable of NRC requirements.

In summary, the licensee has aggressively pursued many program im-provements during this assessment' period. Program oversight and direction were improved by increases in management involvement and support and by the assignment of a new proprietary supervisor.

Enhancements in the training and audit / surveillance programs im-proved personnel performance. Improved interface and communications within the security group and between security and other plant functional' groups have resulted in better coordination and systems performance. The submittal of program changes to upgrade the plans and make them clearly responsive to NRC performance objectives l demonstrated the licensee's increased attention to security. Con-

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tinuation of this level of effort and attention should result in the implementation of a high quality security program.

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2. Conclusion Category 1.

3. Board Recommendations l

None.

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I G. Refueling and Outage Management (1594 hours0.0184 days <br />0.443 hours <br />0.00264 weeks <br />6.06517e-4 months <br />, 38%)

1. Analysis <

This area was rated as Category 1 during the previous assessment j period. The supervision and control of refueling and maintenance l outages was considered a significant management strength.

) ,

All phases of outage activities were reviewed during the present assessment period, including activities to: (1) mobilize and prepare the piant for replacement of the recirculation system piping;

!

(2) replace the piping and complete nondestructive examination (NDE) I on the finished welds; (3) restore the drywell and plant systems; i and, (4) restart the plant. An NRC team inspection at the end of j the outage reviewed the completion of outage activities and assessed )

the readiness of the plant to restart. 1 Management of the Recirculation Pipe Replacement i

Prior to this SALP period, the licensee had performed extensive l review and planning for the outage work necessary to replace the l recirculation piping system. The hardware changes began in October i 1985 and were extensively inspected by the NRC.  !

Preparations for the project included selection of the contractor j at the preceding refuel outage. Videotapes and measurements of the ;

drywell internals were made at that time and were used early in the project planning process. Numerous mockups were prepared for train-

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ing and qualification purposes. Prior to actual pipe removal, work i packages were substantially complete, including full review by both 1 the contractor and the Vermont Yankee project team. 1

{

Planning and Staffing 1

Planning, staffing, welding and NDE aspects of the pipe replacement !

program were found to be performed in accordance with prudent man-agement practice, NRC regulations and applicable codes and standards. ]'

Management involvement and control in assuring quality were noted to be favorable for both the piping contractor and the licensee.

The outage and modification operations were well planned and the contractor's quality assurance program was effective. Existing materials and equipment were removed in a manner such as to facili- )

tate the installation of new items, in that old items were sized l and dismantled the same way the new items were fabricated and moved in the drywell. An extensive hands-on craft training and mock-up program to simulate drywell conditions, includino full scale mock-ups,

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7 rigging, welding equipment and anti "C" clothing and respirators,

was in use. These factors minimized unnecessary radiation exposure 1 and essured a quality installation.

l L The licensee and contractor organizations were well staffed with

'

a designated project team to accomplish the project. The licensee provided continuity of personnel from planning until completion of the replacement work and restoration of the drywell components.

Welding The licensee had an effective program in place, including quality assurance, to plan, engineer, ma'tage and inspect pipe replacement activities. When project problems were encountered such as with pipe whip restraints being insufficiently welded (during original construction) and failure of:an embed plate in the torus room, en-gineering participated effectively in problem scope definition, inspection, evaluation and disposition. Production signoff and welding startup documentation problems were found, but these were minor problems that were corrected by the contractor and monitored by the licensee.

The most significant welding problem was with making high quality '

pipe root pass welds. The licensee recognized this problem early in the planning process and attempted to prepare the contractor by training and qualification of welders to produce quality root welds without significant in process weld repair. The piping contractor had difficulty in producing acceptable root pass welds without significant rework; however, the pipe replacement program plan re-quired root pass welding to be inspected visually and by radiography prict to completion of the weld, and the radiographic inspections  ;

by the pipe contractor effectively identified root pass weld defects.

' This allowed early removal and repair of root. defects preventing major weld repairs after weld joints were completely welded.

Nondestructive Testing An NRC inspection using the 100 mobile nondestructive examination van was conducted during January 21-31, 1986. This inspection re-  !

verified the quality welding and nondestructive examination of the licensee's program, by performing nondestructive tests and visual examinations of welds and components. NRC results were compared to,the results of the licensee and the piping contractor. The out-age NDE activity was adequate and in accordance with the ASME code and NRC requirements.

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l30 W' ._

Piping and Pipe Support Systems O' * [ Several' outstanding items in the area of piping and pipe support

. _

y'. '" 4 systems were' inspected. Major' items included licensee activities. j f' '

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.' in response to Confirmatory Action Letter 85-06 regarding IEB 79-02 requirements .for safety related dead weight supports, and testing

.y !cf small bore pipe supports. '

s- .

y "The licensee's management exhibited good planning for-~ activities

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performed by.the engineering organization and by contractors. .The'

3 , approach to resolution of technical issues was generally conserva-1 ~e tive. However, three items that required.further review were iden-

• '7 t'ified, which included: (1) overview of contractor visual weld ex-aminations for weld repairs on whip restraints; (2) evaluation of whip . restraints using original GE criteria for pipe rupture. loads rather than the more conservative criteria of NRC Standard Review

. Plan (SRP) 3.6.2;.and, (3) the disposition of support-mounted base- 3 plate gaps was not specific'about areas inspected,-number of sup- '

ports with gaps versus. total number of installations, and'the ap- ,

, , -plicability of the problem to cable tray'and heating, ventilation,' l and air. conditioning.(HVAC) supports.

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' Understanding of the.above issues by engineering organization per- 1

sonnel was evident.even when disagreements'on the course of action i existed. The licensee was found to be responsive to NRC initiatives, l and in some cases, corrective actions exceeded minimum requirements. I This was evident in the Piping Seismic Reanalysis Program using /

Regulatory Guides _1.60 and 1.61 in the replacem et of pipe support base plates and shell anchors for all large bore supports and the majority of small bore supports outside primary containment. ,

Management of Routine Outage Activities

Routine outage maintenance, modification and operating activities {

were planned and well controlled. Plant staffing was adequate and i effectPr.1 e d to coordinate and supervise outage and contractor j activities. Plant housekeeping was maintained in good condition '

in spite of the extensive modification activities and the duration l of the outage. Daily staff meetings for both the plant staff and "

the recirculation project staff were effective in maintaining good communications between outage groups for planned activities. One instance was noted where the sequence of vessel restoration was not adequately coordinated with core spray nozzle work (early installa-tion of the steam separator / shroud head assembly), which indicated the need to pay better attention to detail in the planning efforts.

The licensee formed a group dedicated to plant startup and restora-tion efforts to recover from the outage. This group did an excel-lent job in planning and coordinating the activities of the various groups required to bring the plant from a construction status to operational startup readiness. Two outage restart issues, involving

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(1) the proper calibration of the stack high range monitor and (2) the environmental qualification of solenoid valves in the resi- J dual heat removal service water cooling circuit, required additional ]

licensee and NRC management attention to achieve acceptable resolu- 1 tion. The licensee was responsive to NRC initiatives to resolve 3 the issues.

The Cycle XII Startup Physics Test was performed in accordance with approved procedures by highly qualified personnel. Records were i well prepared, complete and readily retrievable. Staffing was ample and reactor engineers involved in these tests were knowledgeable in assigned areas. The Operational Quality group implemented a detailed surveillance plan for the outage, including procedure re-view, test witnessing, and test results review for selected startup tests. QA's coverage on these tests was thorough and comprehensive.

Refueling, in-vessel maintenance, and new fuel inspection procedures were technically adequate and implemented by an experienced, well qualified staff. Personnel displayed a good regard for safety.

Two instances were noted where greater attention to detail by re-fueling personnel could have avoided fuel loading errors and damage to the refueling mast. Licensee management response to the incidents were appropriate. Problems with refueling equipment were minimal, and deficiencies noted during in-vessel inspections were followed through with the appropriate level of detail to assure proper re-solution. Extensive management effort was expended to complete a reactor coolant system decontamination.

The single violation identified in this area concerned the documen-tation by contractors of completed structural weld visual examina-tions on the wrong reports. Licensee followup actions were exten-sive to assure no deficiencies in the required NDE examinations occurred. The LERs submitted for this area involving either engi- i neered safeguards features (ESF) actuation signals, or deficiencies i identified as a result of extensive examinations and work completed ]

during the outage were not indicative of an adverse trend and were J not causally linked. Two events, LERs 85-10 and 85-13, concerned the identification of original construction deficiencies involving i drywell cable penetrations and embedded base plates, respectively. 1 A third defect, not meeting 10 CFR 50.73 reporting criteria, con-cerned inadequate original construction welding on recirculation piping whip restraints. Corrective actions for the items were technically sound and thorough. The programmatic weaknesses that led to the original construction defects are not considered applic-able to present day controls.

In summary, the control of outage activities remains a significant licensee management strength, based on the successful supervision of numerous and complex outage tasks, and the quality evident in the completed work.

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2. Conclusion Category 1.

3. Board Recommendations None.

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H. Assurance of Quality (226 hours0.00262 days <br />0.0628 hours <br />3.736772e-4 weeks <br />8.5993e-5 months <br />, 5%)

1. Analysis i

Management involvement and control in assuring quality is being j considered as a separate functional area in addition to being one j of the evaluation criteria in the other functional areas. The i various aspects of quality assurance (QA) program requirements have l been discussed as an integral part of each functional area and the respective inspection hours for quality assurance activities are j included in each one. The hours listed above reflect specific in-spections of the QA program. This discussion is a synopsis of the assessments relating to the assurance of quality for activities in other functional areas.

The related area of quality assurance was rated as Category 2 during )

the last assessment period due to significant problems that were j identified early in the assessment period, but which were signifi- !

cantly improved during the period. The deficiencies included prob- )

lems in four areas, including: (1) receipt inspection, (2) preven- 1 tive maintenance of stored electrical materials, (3) implementation of the " peer" inspection program, and (4) ineffectiveness of the j corrective action program. Although licensee management had iden-tified programmatic problems in the receipt inspection area, that j information was not quickly identified to upper management levels i for action, and corrective actions were not initiated in a timely manner which was indicative of a lack of sensitivity to QA issues.

Licensee commitments to complete an overall QA program implementa-tion review was an outstanding item at the end of the last assess-ment.

During this SALP period, the licensee QA programs for procurement )

control, receipt, storage and handling, and QA Audits were found to be adequate. The QA technicians were found to be adequately '

trained, qualified and certified to the level of their responsibili-ties and involvement. Quality in plant operating activities is assured by an overall attitude by personnel in the various func-tional areas to perform assigned work correctly the first time.

Quality assurance measures were effectively implemented in plant j activities inspected by incorporation of controls into routine pro- l cedures, audits and/or surveillance. Procedures and administrative '

requirements were well established and implemented by qualified

. staff. Plant personnel at the working level generally exhibit a good attitude towards QA, procedure adherence and completeness.

The QA/ quality control (QC) functions are adequately staffed with personnel with appropriate technical capabilities and training.

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~The licensee's audit program indicated that all facets.of audits

.related to plant operational activities were properly planned, ,

scheduled and conducted. The audits were technically ~ thorough and in-depth. The audit checklists were well organized and'comprehen ,

L sive. Recent quality assurance group findings identified the fail-ure of plant responses to meet the Manager of Operations (M00) im-plementation directive for audits- About half (71) of the 1986 audit finding response actions were not completed per the M00 direc-tive. Most of these.were in the radiation protection area, which -]

also had a large number of findings recently due to'high quality, I in-depth audits. Most deficient response items concern' audit."ob-servations," which are auditor-suggested changes or recommendations )

"

for improvement. The large number of overdue items is_due,'in part, >

L to the failure of the present program to differentiate between de-ficiencies and observations, and assign prioritized M00 implementa- )

!

P tion deadlines accordingly. Licensee management has'taken the in . j itiative to address the program deficiency and initiate corrective 1 actions.

The licensee demonstrated a strong commitment during the, assessment' I period to the' identification and correction of. problems in the J quality assurance department and area. This commitment was evi-'

denced by: (1) completion of an in-depth audit-of the QA program ~

R implementation by an outside contractor and the development of a I corrective action plan; (2) more active involvement-in site activi- i ties, including dedicated support for_the recirculation pipe re- !

placement project; and (3) implementation of an improved peer in- '

spection program, supplemented by onsite QC groups for the mainten-ance and' instrument and controls areas during the 1985-1986 outage w Organizational changes in the quality assurance department, announced at the end of the assessment period, will provide a full-time man- j ager onsite for the operational quality group and the operational surveillance function. ,

While current QA activities were found to be properly directed and '

involved in ongoing work, problems were found during NRC inspection of the licensee actions for IEB 80-11. These problems (violations and deviation discussed in Section IV.C. above) resulted from de- ;

ficiencies in the YAEC QA and VY .nlant operation's QA programs that i existed prior to 1984. Specifically, the QA audits of engineering did not focus on specific program activities, and as a consequence, deficiencies in the follow-up activities related to IEB 80-11 were not self-identified and corrected. NRC followup review of licensee actions to ensure that the QA program deficiencies are corrected were in progress at the end of the assessment period. Further lic- !

ensee actions are warranted to assure deficiencies in engineering i activities are identified and corrected.

l i

35 Several significant potentially generic equipment issues were iden- l tified and reported by'the licensee during the present assessment- '

period, as discussed in Section IV.C. above. Three of these issues (Conax squib primers, NAMC0 contact blocks, and ASCO rebuild kits)

resulted in actual or potential equipment operability problems which stemmed from defective parts supplied by the manufacturer that were 3 not detected by the licensee's preservice inspection and testing ;

program prior to introduction of the materials into plant systems-Overall, licensee management responsiveness to the issues was evi-dent, with generally prompt and thorough followup actions. Licensee management showed initiative: (1) to identify root causes, (2) to ,

identify manufacturer's weaknesses, and (3) to assure corrective '

actions for the specific discrepancies were implemented, including working with the vendors as necessary to assure that quality re-placement parts were ultimately obtained and placed into service.

The single violation for this area concerned corrective actions i following the initial discovery of the NAMC0 block defects that were i limited in' scope and incomplete. Licensee followup actions after i NRC involvement were thorough to achieve adequate corrective actions. j Although no LERs were issued for this functional area during the assessment period, the events described in LERs 86-04 and 86-13 describe related issues. The material deficiencies associated with the squib valves were discussed above. The material differences associated with the failed IRM detectors further highlight recent problems in the procurement and preservice inspection programs for assuring quality parts are used in the plant.

These matters were the subject of a management meeting held at the NRC's initiative. The licensee concluded from the evaluation of the problems that the defects were not discovered by the established plant controls, in part, due to the vendor's previous record of '

i supplying quality products. Continued management actions are war-ranted to assure implementation of rigorous and comprehensive vendor surveillance and preservice inspection programs, that will ensure early detection of degraded vendor performance, and identify defec-tive materials prior to introduction into the plant.

In summary, the licensee has taken actions during the assessment period to identify and correct deficiencies in the QA program through the completion of an in-depth audit and development of a '

corrective action plan. There is generally good regard for quality assurance during the conduct of routine activities. Improvement' ,

'

have been noted in the areas of audits and receipt, storage and handling. The " peer" inspection program was properly implemented

,

'

for the outage and effective. Several equipment issues were iden- l tified where defective parts supplied from the manufacturer were not detected by the preservice inspection and test program prior to installation into the plant. Licensee followup actions to de-ficiencies were aggressive, thorough, and adequate to identify root i

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I causes, and to assure that acceptable corrective actions were com- !

pleted; Additional licensee actions are needed to assure completion ;

of program improvement items and to complete recent initiatives to improve QA program effectiveness. I 2. Conclusion

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Category 2.

3. Board Recommendation .

None.

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I. Training and Qualification Effectiveness (114 hours0.00132 days <br />0.0317 hours <br />1.884921e-4 weeks <br />4.3377e-5 months <br />, 3%) ,

1. Analysis  !

l

Training and qualification effectiveness were reviewed as a separate j functional area for the first time. These performance attributes '

continue to be an evaluation criterion in each functional area. (

One inspection completed during this assessment period included a i review of the effectiveness of the non-licensed training programs. I The licensed operator training program was reviewed as part of the restart team inspection and during the operator license examinations given in June, 1986.

Since various aspects of this functional area are discussed as an l integral part of other functional areas (which also include the re- '

spective inspection hours), this discussion is a synopsis of the training assessments in the other areas. Training effectiveness i was measured primarily by the observed performance of licensee per- i sonnel, and to a lesser degree, by a review of program adequacy. l The three principal areas addressed below are licensed operator !

training, non-licensed staff training, and the status of INP0 ac- '

creditation.

The licensee has developed a licensed operator training staff that ;

reduces the reliance on contractors. The department includes opera- '

tor instructor and simulator groups, the former of which includes i contractor personnel used for classroom training in the licensed operator programs. About half of the licensee operator instructors have senior reactor operator licenses. The simulator program is administered by SRO-licensed instructors. The effectiveness of the licensed training program was demonstrated by the success rate of the ten candidates pretented for the senior operator upgrade and instructor certificate NRC examinations in June, 1986, and due to the fact that no overall programmatic concerns or weaknesses were identified during the examination process.

As discussed in Section IV.A above, weaknesses were discovered in the operator training programs for outage modifications, which was limited in scope and depth. Licensee corrective actions were timely and comprehensive. The licensee commitment to use the simulator to train the operators on the new E0Ps prior to startup from the outage was jeopardized by delays in the simulator startup schedule.

The training was conducted on an accelerated schedule in response to NRC initiatives, anct completed during the six-week requalifica-tion training period following the July 1 startup. Prior to the startup, the licensee also completed an evaluation and certification of the senior shift operators on their capabilities to use the new procedures. The senior operators demonstrated proficiency in the use of the E0Ps during simulator trials prior to startup. Licensed i

-,

operators demonstrated proficiency in use of the E0Ps during the 1986 emergency exercise, and have shown an acceptance of the new procedures.

The licensee uses departmental training programs for non-licensed personnel, which has proven successful and effective as demonstrated by the few personnel errors during the assessment period. Plans are in progress to consolidate non-licensed training into the training department to provide better coordination of technical training. The six non-licensed training programs are being upgraded per the schedule discussed below for INP0 accreditation. Routine-NRC inspections and the assessment of personnel performance in the radiological controls, security, maintenance, surveillance, and emergency preparedness areas have shown the effectiveness of the i licensee's training programs for these groups. I

!

One inspection reviewed the training programs for instrument and I control (I&C) technicians, maintenance personnel, auxiliary opera-tors and peer inspectors. Weaknesses were identified in the program for peer inspection, based on interviews with plant workers, in that additional training was needed in acceptance criteria, communica-tions and report writing. Additionally, a weakness was identified in the training provided I&C personnel regarding the proper use of '

a sampling table for material inspections. Overall, however, the inspection confirmed that licensee management had committed re- ;

sources both to improve the non-licensed training programs and to l the conduct of the courses.

]

The licensed operator and non-licensed operator training programs received INP0 accreditation in April, 1986, which included accredi-tation for the non-licensed auxiliary operator and shift engineers l training programs. The non-licensed training programs for the l health physics, electrical maintenance, chemistry, mechanical main- l tenance, instrument and control and technical staff positions are l under development for INP0 accreditation. A self-evaluation report I was submitted to INP0 in November, 1986 and actions are in progress to increase instructor staffing and implement the non-licensed pro- !

grams during 1987. An INP0 evaluation team visit is scheduled for the Fall of 1987. Laboratories are available in the new training facility for I&C, chemistry, health physics, and maintenance train-ing programs, but are not fully equipped or used. Plans are in progress to equip and use the labs in the non-licensed programs.

In summary, licensee management has demonstrated a continued com- l mitment to training, as evidenced by: (1) the implementation of the l simulator training program; (2) the improvements to and accredita-tion of the licensed operator training program; (3) staffing of the training department which includes program specialists to assist in program development and effectiveness reviews; and, (4) the im-provements in the non-licensed training programs, along with efforts i

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to obtain accreditation. The effectiveness of this commitment is reflected in the plant performance record, the correct performance of operating activities and.the low numbers of. personnel errors.

While training program improvements are progressing per industry commitments, the overall effectiveness of the training and qualifi-cation programs is evidenced in the overall success in plant and personnel performance.

2. Conclusion Category 1.

3. Board Recommendations None.

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J. Licensing Activities 1. Analysis This area was rated as Category 1 during the previous assessment period. Performance in the area of licensing activities was evalu-ated based on the following attributes: (1) management involvement and control in assuring quality; (2) approach to resolution of tech-nical issues from a safety standpoint; and, (3) responsiveness to NRC initiatives. This evaluation represents the combination of assessments by the NRC Office of Nuclear Reactor Regulation (NRR)

operating reactor project manager, technical reviewers and the resident inspector.

Vermont Yankee Nuclear Power Corporation has demonstrated a con-tinued awareness of the various licensing issues by virtue of ex-tensive experience in the industry, technical expertise, and active participation in industry and professional organization activities.

Management takes actions in a timely manner to ensure safety issues are properly addressed.

The fact that no emergency technical specification (TS) changes have been requested during the report period evidences consistent plan-ning by management to take into account license requirements. The recent request (December 30, 1986) for cask lifting device TS ap-proval within a month of planned cask use is an exception to usual practice.

In mid-1986 Vermont Yankee undertook a Containment Safety Study addressing concerns related to the capability of Vermont Yankee's Mark I containment to withstand severe accidents. The conduct of this study was, in part, guided by Vermont State and the staff's initiative to. improve the severe accident performance of BWR Mark I containments, and has been coordinated with the State of Vermont through the state's Vermont State Nuclear Advisory Panel (VSNAP).

On several occasions Vermont Yankee management met with and formally responded to questions on the study from the staff and VSNAP. The initiative displayed in undertaking this study and the followup activities related to it, as well as the quality and timeliness of the effort, evidences management sensitivity to, and involvement in, resolution of safety concerns.

Candid discussions between the NRR project manager and licensee management have satisfactorily served to integrate safety and opera-tional interests from the licensing point of view. Integrated scheduling is an option that is available, if the present less for-mal process becomes unsatisfactory. Since the last SALP, the lic-ensee has initiated a practice of regularly informing the project l manager of the licensee's prioritization of pending licensing ac-

,

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tions. At-the same time there has been some effort to withdraw from the "pending" list requested actions that no longer are required, or that require revision in order to be acted upon.

Vermont Yankee's engineering staff, in concert with support from Yankee Atomic Electric Company.(YAEC), assures that most engineering work, either done inhouse or performed under its direction by con- .

tractors, adequately addresses complex technical issues. An example t of the licensee's initiative and technical capability is the Con-tainment Safety Study referred to above. Vermont Yankee also has under review a unique technical effort in the qualification of the RELAP-5YA code for BWR analysis. The licensee frequently forms ,

technical judgements independently from the industry. These judge-ments are well-thought out with adequate technical bases. For ex- 4 ample, the exemptions granted on December 1, 1986 to Vermont Yankee for Appendix R are the result of the licensee's presentations to the staff detailing the validity of certain special technical con- 3 siderations pertaining to Vermont Yankee. Safety evaluations sub- i mitted by the licensee in support of proposed technical specifica- I tion changes, or to resolve technical issues, have been clear and substantive.

J Open and effective communication channels exist between the NRC and YAEC Vermont Yankee licensing staff in Framingham, Massachusetts with involvement of licensee m'inagement in Brattleboro, Vermont as appropriate. The effectiveness of this communication has improved since the last SALP period. The licensee meets established commit-ment dates or provides a written submittal explaining the circum-stances and establishing a new target date. Conference calls with the staff are promptly established and include appropriate engineer-ing and plant personnel. An example of licensee responsiveness to NRC. initiatives is the recent resolution of TMI Action Plan Item II.K.3.18 pertaining to the automatic depressurization system (ADS)

logic. Additional management attention is needed to assure timely closure of unresolved items from Generic Letter 83-02 pertaining to technical specification requirements for NUREG-0737 items. l

!

No changes in the YAEC Vermont Yankee licensing staff have occurred i during this assessment period. The licensing engineer has gained !

licensing experience thereby increasing effectiveness. Both the j Framingham and Brattleboro offices have supported licensing discus- J sions in a professional manner. The involvement of Brattleboro staff in licensing discussions have been somewhat greater than in l the previous assessment period.

In summary, for the current assessment period, there was consistent l evidence of prior planning, and assignment of priorities that demon- j strates continued close management involvement and control in as- j i

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suring quality. Licensee initiatives in completing the Containment-Safety Study lead industry actions in addressing Mark I containment severe accident issues.

.2. Conclusion  ;

Category 1.

3. Board Recommendations  ;

None.  ;

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V .' SUPPORTING DATA AND SUMMARIES P

A. Investigation and Allegation Review 1. Inspection 85-40, Worker Drywell Exposure Concerns

, 2. Inspection 86-08, Worker ~ Fitness For Duty Concern

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Worker Exposure Concern B. . Escalated Enforcement Actions I 1. Civil penalties None.

2. Actions Pending/ Resolved

.None. -l

3. . Orders.

None, l

j 4. ' Confirmatory Action Letters .

None.

C. Management Conferences 1. Inspection 85-37 concernedLa management enforcement conference held on October.21, 1986 involving a security incident that was consid-ered in the last SALP assessment period.

~

2. Inspection 86-21 documented the results of a management' conference held at NRC request to discuss material deficiencies.

D. Licensee Event Reports 1. Tabular' Licensing Type of Events i

A. Personnel Error . .............. 2 B. Design / Mfg / Construction / Install Error . . . . 8 j C. External Cause. . . . . . . . . . . . . . . . O !

D. Defective Procedure . . . . . . . . . . . . . 7 I E. Component Failure . . . . . . . . . . . . . . 0 X. Other . . . . . . . . . . . . . . . . . . . . 6 TOTAL 23 )

i A tabulation of Licensee Event Reports (LERs) by functional area, j and an LER synopsis, is attached as Table 4. y

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44 4 Licensee ~ Event Reports Reviewed )

I LER Nos. 85-07 through 85-12 and 86-01 through 86-16, and S4-23 i through 84-24 and 85-01 through 85-06. l 2. Causal Analysis Seven sets of causally linked or similar events were identified, a. LERs 86-16, 86-08, 85-05, 85-03, 85-02, 84-24 and 84-23 concern events caused by personnel errors. Five of the seven events J are from the last assessment period. The overall trend shows '

the effectiveness of management efforts to reduce personnel errors. .

b. LER 85-07 concerned the failure of feedwater check valve 96A to pass a Type C leak rate test, which was a repeat occurrence from LERs 84-11 and 83-10. Three failures occurred before the root cause was identified, which was indicative of marginally acceptable failure evaluations.

c. LERs 86-15., 86-14, 85-07, 86-06, 86-01, and 85-09 were events attributed to procedural deficiencies. While the total number for one assessment period appears high, consideration of the circumstances for each one shows most events involve special ;

'

circumstances attendant with the pipe replacement outage or unusual operational conditions. Thus, the number of events is not necessarily indicative of adverse trend in procedure )

quality. ,

d. LERs 86-13, 86-12, 86-11, 86-05, 86-04, 86-02, 85-13, and 85-10 concern events attributable to desi0n, manufacturing or con- !

struction problems. The total number of events are not signi-ficant in view of the depth and scope of activities and exam- i

'

inations completed during the 10-month recirculation pipe re-placement outage. The number of identified original construc-tion defects reflects the thoroughness of the licensee's reviews to identify and correct deficiencies.

e. LERs 86-16, 86-07, 86-01, 85-07 and 85-06 are events concerning the primary containment boundary and involve either inadequate test procedures, missed surveillances, degraded isolation valve or penetration boundaries, or physical loss of the barrier.

None of the events individually constitutes a significant safety concern or impacts primary containment integrity during power operations outside the leakage requirements of Technical !

Specification 3.7.A. This matter warrants further licensee I attention to ensure procedures, surveillance and maintenance activities are conducted in such a manner as to provide a high assurance of containment reliability and performance.

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f. LERs 86-15, 86-13, 86-11, 86-10, 86-08, 86-06, 85-09 and 85-04 are events involving reactor scram signals or ESF.actuations.

Five of the events had no safety significance since they oc-curred with the reactor either shut down and/or defueled, and becaJse they resulted from circumstances unique to recircula-tion pipe replacement outage. Two of the remaining three events occurred-as a result of equipment malfunctions, which are not causally linked.

g. LERs 85-11, 85-05 and 84-23 concerned inadvertent group 3 isolations caused by personnel performance errors or equipment malfunctions. The events have no safety significance and no adverse trends are apparent.

E. Licensing Activities 1. NRR Site Visits November'7-8, 1985 January 9-10, 1986 July 28 - August 11, 1986 September 29-30, 1986 October 9-10, 1986 2. Commission Briefings None.

3. Schedular Extensions Granted None.

4. Reliefs Granted December 19, 1986; Certain inservice inspection requirements 5. Exemptions Granted December 1, 1986; Certain requirements of Appendix R 6. License Amendments Issued Amendment No. 91, issued October 24, 1985; revises TS regarding iodine spiking.

Amendment No. 92, issued March 27, 1986; revises TS to reflect changes in recirculation system piping.

Amendment No. 93, issued June 24, 1986; revises TS regarding Nil Ductility Transition Temperature. j

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Amendment No. 94,' issued August 8, 1986 revises TS regarding Single-

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Loop Operation.

Amendment'No. 95, issued' August 11, 1986; revises TS concerning.re-porting (50.72 and 50.73).

Amendment No. 96, issued August li, 1986; revises TS pertaining to i

.NUREG-0737 modifications-(Generic Letter'83-36).  ;

Amendment No. 97, issued December 3, 1986; revises TS to permit RHR wear ring replacement during fuel Cycle 13.

]

7. Emergency / Exigent Technical Specifications ';

None.

8. Orders Issued None.

9. NRR/ Licensee Management Conferences None. 4 i

10. Operating Reactor Licensing Reviews.  !

Review of Control Room Carpet *

Review of Alternate Inspection of Feedwater Nozzles *

Review of Core Spray Safe End Repairs *

Review of Use of PVRC Damping Analysis *

Review of Containment Safety Study . I Review of Analysis Using RELAP SY-A

• Indicates action completed

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TABLE 1 INSPECTION REPORT ACTIVITIES 10/19/85 - 12/31/86 VERMONT YANKEE NUCLEAR POWER STATION Report Number /

Period Inspector Hours Areas Inspected 85-33 Specialist 38 Preparation for recirculation pipe 10/21/85-10/25/85 replacement 85-34 Specialist 58 Calibration and surveillance programs 10/28/85-11/01/85 85-35 Specialist 119 Recirculation pipe design change 11/04/85-11/08/85 85-36 Resident 189 Routine, resident 10/21/85-11/21/85 85-37 Specialist 2 Meeting (enforcement conference) - security 10/21/85 85-38 Specialist 27 Nonradiological chemistry 12/02/85-12/06/85 85-39 Specialist 40 Radiation controls / outage 12/02/85-12/06/85 85-40 Resident 112 Routine, resident 11/26/85-12/31/85 85-41 Specialist 39 Recirculation pipe replacement 12/16/85-12/20/85 86-01 Resident 202 Routine, resident 01/02/86-02/03/86 86-02 Specialist 486 NRC Mobile NDE Van 01/21/86-01/31/86 86-03 Specialist 74 Radiological controls / outage 02/03/86-02/07/86 86-04 Resident 134 Routine, resident 02/04/86-03/17/86 T1-1

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Table 1 Continued Report Number /

Period Inspector Hours Areas Inspected 86-05 Resident 61 Special/ inoperable SLC system 02/09/86-02/27/86 86-06 Specialist 29 Security-03/10/86-03/14/86 86-07 Specialist 36 Radiological controls / outage 03/24/86-03/28/86 86-08 Resident 213 Routine, resident 03/18/86-05/05/86 86-09 Specialist 33 Emergency preparedness L-04/22/86-04/24/86 86-10 Resident 567 Resident / plant restart inspection 05/06/86-06/30/86 86-11 Specialist 0 Operator licensing / examination report 05/28/86-06/06/86 86-12 Specialist 53 Piping, supports and restraints ;

05/12/86-05/16/86 86-13 Specialist 316 Outage / restart team inspection 06/02/86-06/06/86 86-14 Specialist 35 Containment ILRT 06/21/86-06/25/86 j 86-15 Resident 198 Routine, resident j 07/01/86-08/04/86 {

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86-16 Specialist 18 Emergency preparedness 07/23/86-07/24/86 l

86-17 Specialist 104 IE Bulletin, masonary walls '

09/09/86-09/26-86 86-18 Resident 147 Routine, resident 08/05/86-09/03/86 86-19 Specialist 32 Startup physics testing '

08/11/86-08/15/86

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Table 1' Continued Report Number / .

Period Inspector Hours Areas Inspected '

86-20 Specialist 76 Quality assurance 08/25/86-08/29/86 86-21 Specialist 12 Meeting / material deficiencies '

07/29/86 I

86-22 Resident 291 Routine, resident i 09/04/86-11/03/86 i 86-23 Specialist 68 Independent Measurements Van 10/06/86-10/10/86 86-24 Specialist 50 Radiological controls 11/03/86-11/07/86

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86-25 Resident 243 Routine, resident 11/04/86-12/31/86 86-26 Specialist .119 Emergency preparedness partial 12/02/86-12/04/86 exercise i

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TABLE 2 INSPECTION HOURS SUMMARY i l

10/19/85 -'12/31/86 l

VERMONT YANKEE NUCLEAR POWER STATION l

HOURS l

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, Actual Annualized Percent A. P1 ant Operations 914 767 21 B. Radiological Controls 411 345 10 C. Maintenance'and Modifications 441 370 10 ,

I 0. Surveillance 242 203 6 !

E. Emergency Preparedness 200 168 5 F. Security and Safeguards 79 66 2 i G. Refueling and Outage Management 1594 1337 38 H. Assurance of Quality 226 190 5 I. Training and Qualification Effectiveness 114 96 3 1 J. Licensing Activities NA NA NA i TOTAL 4221 3542 100%

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TABLE 3:

ENFORCEMENT SUMMARY!

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'10/19/85 - 12/31/86

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VERMONT. YANKEE-NUCLtAR POWER STATION

I. Violations Versus Functional Area by Severity Level-Severity Level No.-of Violations in Each Severity Level Functional Area DEV. V IV' III II I Total.

, .A. _ Plant Operations 0-

' B. Radiological Controls 0 C. Maintenance and Modifications 1- 1- 1 3

D. . Surveillance 2 1 1 4 E. Emergency. Preparedness .O F. Security and Safeguards 0

.-. G . Refueling and Outage. Management 1 1 H. Assurance of Quality 1 1 I. Training and' Qualification

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.0 Effectiveness 2

'J. -Licensing Activities 0 TOTALS 1- 4 3 1 0 0 9

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i Table 3 Continued II. Summary

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Inspection Severity Functional Number Requirement Level Area Brief Description 85-36 TS 6.5.A V Surveillance Failure to calibrate process radiation monitors per OP 4511.

86-05 TS 3.4.A III Surveillance SLC system inoperable for operating Cycle XI 86-08 App B Crit IV Qual.Assur. Inadequate corrective ac-XV & XVI tions for defective NAMC0 blocks.

86-12 TS 6.5.A V Outages Failure to document re-straint visual examination reports.

86-15 TS 4.4.C V Surveillance Failure to sample SLC tank after water addition.

86-17 App B Crit V IV Maintenance IEB 80-11 block wall surveys completed without a written procedure.

86-17 App B Crit V Maintenance Failure to retain results ,

XVIII of IEB 80-11 block wall surveys.

86-17 IEB 80-11 D Maintenance Failure to verify block wall Commitment mortar properties per IEB requirements.

86-25 TS 4.6.E IV Surveillance Failure to perform inservice per Section XI and TS 4.6.E l

T3-2

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TABLE 4 LICENSEE EVENT REPORT-10/19/85 - 12/31/86 t. VERMONT YANKEE NUCLEAR POWER STATION I. LER by Functional Area

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Number by Cause Code Functional Area A B C D E X Total A. Plant Operations 1 1 2 B. Radiological Controls 1 1 C. Maintenance and Modifications 1 1 1 3-D. Surveillance l' 3 3 7 E. Emergency Preparedness F. Security and Safeguards G. Refueling and Outage Management 1 5 2 2 10 H. Assurance of Quality I. Training and Qualification Effectiveness J. Licensing Activities _ _ _ _ _ _ __

TOTAL 2 8 0 7 0 6 23 Cause Codes: A. Personnel Error B. Design, Manufacturing, Construction or Installation Error C. External Cause D. Defective Procedure E. Component Failure X. Other 1 I,

T4-1

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l' . Table 4~ Continued a

II. LER Synopsis

'LER' Number 1 Summary 85-007: While performing type C. rate testing, nine valves were found to have seat leakage.above that permitted by TS 3.7 A.4.

,

'85-008 Persennel discovered an error in the offsite dose rate calculations procedure, which did not correspond to TS Section 3.8.E.85-009 A full. scram signal was received from the scram circuit while im-plementing a design change in the RPS: system.85-010' Conductors. installed in contact with the sharp edge of.the end'of the electrical penetration assembly sleeve.

85-011- Spurious signals'on refuel floor zone radiation monitor resulted ,

in isolation of the'RB ventilation system and activation of the SGTS. !85-012 Freon was released.in the vicinity of the CR airLintake duct while servicing the air conditioners which activated the CR habitability; system.85-013 Em'b edded plate failed because the embedments had been removed prior to installation.

'86-001 Containment. valves missing from Appendix J 1eak rate test program.

.86-002 Calculational error discovered in the Environmental: Qualification Program which underpredicted the postulated radiation exposure.for

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parts in the H2/02 analyzer system.86-003 Two main ~ steam relief valves actuation pressures were above the setpoints.86-004 The "A" SLC squib valve failed to fire during performance of' annual surveillance.86-005 Ultrasonic indications of intergranular stress corrosion c}acking (IGSCC) were detected in Inconel 182 weld butter of both spray nozzles.86-006 Unanticipated scram signal during LPRM cable testing / troubleshooting.86-007 1985/86 Appendix J Type A test failure due to penetration leakage.86-008 Unanticipated scram during mode switch movement T4-2

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Table 4 Continued i LER Number Summary 86-009 "C" SRV accumulator failed leak test due to check valve seat leak. I

!86-010 Inadvertent reactor scram signal on loss of RPS caused by broken !

relay armature. I 86-011 Inadvertent scram signal on loss of RPS "B" alternate power and neutron monitoring system A trip.

j 86-012 Motor operator failure due to torque switch hydraulic lockup.86-013 Reactor scram during startup due to inoperable intermediate range monitors.

!86-014 Failure to sample SLC tank after water addition. l 86-015 High flux reactor scram in hot standby due to cold feedwater injection.86-016 Missed Appendix J leak test due to personnel error.

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TABLE 5 1 UNPLANNED REACTOR TRIPS AND SHUTDOWNS 10/19/85 - 12/31/86 VERMONT YANKEE NUCLEAR POWER STATION I. Assessment There were no reactor scrams with the reactor operating at power during the assessment period. Two scrams occurred with the reactor in the startup/ hot standby condition. Five inadvertent reactor trip signals occurred with the j reactor in cold shutdown or defueled. Inadvertent PCIS Group III isolations occurred due to spurious trips from a refueling zone radiation monitor. The causes for these ESF actuations involved equipment malfunctions, personnel error,-inadequate design, and procedure deficiencies. This section assesses the root cause of each event from the NRC's perspective.

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Procedure Deficiencies l

Three events occurred wholly, or in part, because procedures did not correctly address either plant conditions or operational constraints associated with the activity. The events occurred during circumstances unique to the outage modification activities or operation in a mode rarely used.

Equipment Malfunction

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Two events resulted from random equipment failures and another event was caused from a combination of equipment failures and installation errors.

Inadequate Design One event occurred because of instabilities in the RPS alternate power supply, which is attributable to a power protection panel design that is too sensitive to voltage fluctuations on the plant power buses.

Personnel Error One event was caused by an operator failing to properly operate the reactor mode switch. A second event occurred, in part, as a result of operator errors in an operational mode rarely used - reactor critical in hot standby with the main steam lines isolated. H

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Table 5 Continued A

h, .II. Tabulation 3 Power Root Functional Date Level " Description Cause Area 1. 10/7/85 Defueled Scram signal occurred while Inadequate Outages installing a design change Procedure in RPS system.

2. 10/8/85 Defueled PCIS Group III Isolations Equipment Outages due to spurious trips from Malfunction '

refuel floor zone radiation monitor 3. 5/21/86 Cold Scram signal during LPRM Inadequate Outages l Shutdown cable testing and trouble- Procedure l

,

shooting. i 4. 6/4/86 Cold Inadvertent scram signal Personnel Outages Shutdown during mode switch movement. Error 5. 6/9/86 Cold Inadvertent scram signal due Equipment Outages l Shutdown to loss of RPS power caused Failure / Random by broken contactor armature.

6. 6/10/86 Cold Inadvertent scram signal due Inadequate Outages /

Shutdown to instabilities in RPS Design Design alternate power supply.

7. 6/30/86 Startup Reactor scram during startup Equipment Operations IRM "6" due to inoperable intermedi- Failure /In-ate range monitors. stallation Error 8. 10/4/86 Hot High flux on IRM Range 1 due Inadequate Operations-Standby to cold feedwater addition. Procedure /

IRM "1" Personnel Error i

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TABLE 6'

SALP HISTORY.

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, l7/1/81 - 12/31/86.

VERMONT YANKEE NUCLEAR POWER STATION

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~ Asses'sment Report > ' Period ~ OPS JRADCON MAINT SURV- EP. FP SEC OUTG< QP.~LIC TRG'

- 03/83- . 7/1/81-- 1 1- 1 1 2' :1 :1 1 N- 1~ N; i

.,_ 6/30/82- 3

11/83 5/1/82 . ~2- l' .1

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l' N 1 1 2 N L 1- N --

4/30/83 d 12/84~ -

5/1/83- "1 2 2 1 1 2 1- 1 '2 1 N-10/31/84

.12/85: '11/1/84 - 11 2 1 'l- 2 N 2 l' 2- 1 N 10/18/85-

. 2/87- '

-10/19/85- '11 2' 2 1 2 N 1 l' 2 1 l'

12/31/86'

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N = Not Evaluated During Assessment' Period.'

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FIGURE 1 i NUMBER OF DAYS SHUTDOWN VERMONT YANKEE NUCLEAR POWER STATION October 1985 XXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXI 31 Days -Pipe Replacement Outage November 1985 XXXXXXXXXXXXXXXXXXXXXXXXXXXXXXl 30 Days - Pipe Replacement Outage December 1985 XXXXXXXXXXXXXXXXXXXXXXXXXXXXXXI 30 Days - Pipe Replacement Outage January 1986 XXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXI 31 Days - Pipe Replacement Outage February 1986 XXXXXXXXXXXXXXXXXXXXXXXXXXXXI 28 Days - Pipe Replacement Outage March 1986 XXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXI 31 Days - Pipe Replacement Outage April 1986 XXXXXXXXXXXXXXXXXXXXXXXXXXXXXXl 30 Days - Pipe Replacement Outage May 1986 XXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXI 31 Days - Pipe Replacement Outage June 1986 XXXXXXXXXXXXXXXXXXXXXXXXXXXXXXI 30 Days - Pipe Replacement Outage i July 1986 'YRI 2 Days - Scram During Startup August 1986 l 1 I

. September 1986l October 1986 XXXRI 4 Days - Turbine Boot Repair and Scram l i November 1986 l 1 l  !

December 1986 l l

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, . jps mac o UNITED STATES

' ,4 41 NUCLEAR REGULATORY COMMISSION Enclosure 3

,( 3 REGION I

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In iE 631 PARK AVENUE

. KING OF PRU55tA. PENN5Yi.VANI A 19405

.....

Docket'No. 50-271 Vermont Yankee Nuclear Power Corporation i

ATTN: Mr. Warren P.' Murphy '

Vice President and Manager of Operations RD 5, Box 169 Ferry Road Brattleboro, Vermont 05301 Gentlemen:

Subject: Systematic Assessment of Licensee Performance (SALP) Report No. l 50-271/85-98 On February 23, 1987, the NRC Region I SALP Board reviewed and evaluated the per-formance of activities associated with the Vermont Yankee Nuclear Power Station.

This assessment is documented in the enclosed SALP Board Report. A meeting has been scheduled for March 27, 1987 at 10:00 a.m. at the site to discuss the assess-ment. That meeting is intended to provide a forum for candid discussions relating to the performance evaluation.

At the meeting, you should be prepared to discuss our assessment and your plans to ensure improved or continued emphasis upon those activities which would have a positive effect upon performance, Any comments you may have regarding our report may be discussed. Additionally, you may provide written comments within 30 days after the meeting. j Following our meeting and receipt of your response, the enclosed report, your written response (if deemed necessary) and a summary of our findings and planned actions will be placed in the NRC Public Document Room.

Your cooperation is appreciated.

Sincerely,

/

Thomas E. Murley Regional Administrator !

Enclosure: NRC Region I SALP Report No. 50-271/85-98 {

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. Vermont Yankee Nucl88 ower. . 2 *-

MAR 1219g7

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.1 cc w/ encl:

y Mr. R. W. Capstick. Licensing Engineer

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Mr..J. Gary Weigand, President and Chief Executive Officer-Mr. J.~ P..Pel.letier, Plant Manager

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Mr. Donald Hunter, Vice President .-

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q Mr. Cort' Richardson, Vermont Public Interest Research Group, Inc.

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.j Gerald Tarrant, Commissioner, Department of Public Service Chairman Zech Commissioner.' Roberts

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Commissioner Asselstine

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Commissioner Bernthal Commissioner Carr Public Document Room (PDR)'

Local Public Document-Room (LPDR)

Nuclear Safety Information Center (NSIC)

, Record' Center, INP0 i E

NRC Resident Inspector. .f '

State of New-Hampshire l

, . State of Vermont

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' Enclosure 4 L" " VERMONT VANKEE .

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' NUCLEAR POWER CORPORATION

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RD 5 Box 169, Ferry Road, Bratt'eboro, VT 05301 Y 07-44 y ENGINEERING OFFICE 1671 WORCESTER ROAD I

• FRAMINGHAM, MASS ACHUSETTS 01701

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• TELEPHONE 617 672 8100 April 24, 1987

!

U.S. Nuclear Regulatory Commission ,

Region I  ;

631 Park Avenue i King of Prussia, PA 190406 j Attn: Regional Administrator References a) License No. DPR-28 (Docket No. 50-271)  ;

b) Letter, USNRC to VYNPC, Systematic Appraisal of Licensee Performance, dated 3/12/87 )'

Dear Sir:

Subject: Vermont Yankee Response to Systematic Appraisal j of Licensee Performance This letter is submitted in response to the March 12, 1987 SALP Board Report of. Vermont Yankee's performance during the recent assessment [ Reference i b)]. Although many of the comments provided within this letter were articulated I at the March 27, 1987 meeting between the NRC and Vermont Yankee, we would like to take the opportunity afforded by your letter to formally document our )

response. 1 In general, we are pleased with the results of the Board's evaluation. We are particularly happy with your recognition of the major organizational and program improvements made in the area of plant security.

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Not only do we view the SALP Report to be a fair and comprehensive eva- l luation, but we consider it a diagnostic tool which allows Vermont Yankee mana- l gement to take action on perceived weaknesses before they develop into l operational deficiencies. While we agree with the majority of the conclusions, we do offer the comments identified in Appendix A.

As we stated at the Vernon SALP meeting, we continue to pursue oppor-tunities to strengthen our overall operation and therefore, strive to achieve Category I ratings in all functional areas.

Very truly yours, j l

Am _ _ - g W.P. Hur hy -

j g ,.; j ,,,, _/, , s . - Vice President e '

v iy oyYp/6 J Manager of Operat I

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h APPEN0!X A IV. PERFORMANCE' ANALYSIS

]

I o . REFERENCE: A. ' Plant Operation, Analysis - page 9, paragraph 1

" Completion of a Senior operator's certification for the incumbent '

Operations Supervisor remained an outstanding qualification item for this assessment, period".. Licensee management should tssure that the present plans to P obtain this certification by mid-1987 are completed."

Comments-The Operations Supervisor, although not presently licensed, previously held .;

.an SR0' license at the-Vermont Yankee' plant (1972-1974).- Vermont Yankee 1 Procedure AP 0720 specifies that anyone who has held an SRO. license at the plant

[ is considered SRO certified regardless of license status. As such, the Operations Supervisor is considered SRO certified.

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The Operations Supervisor is presently undergoing licensing, training in preparation for_a July license examination. It should be noted that prior to returning to Vermont Yankee, the. Operations Supervisor held a valid SRO license at a sioilar BWR facility.

REFERENCE: B. Radiological Controls, Analysis - page 12,' paragraph 2 i

"The. Licensee's inability to resolve the long-standing ALARA issue con-tinues to be an NRC concern. The item remains unresolved long after deficien-cies were brought to the Licensee's attention during previous SALPs and  !

inspections dating back to 1983. During the current period, ALARA implementing >

procedures remain in draft form. The Licensee created a new "ALARA Coordinator" position, but failed to fill the position."

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Comments i t

During the 1985 recirc pipe replacement / refueling outage, a Vermont Yankee employee was temporarily assigned the responsibility of ensuring that all work expected to involve significant radiation exposure was eveluated to ensure con-sistency with ALARA principals. During this time, the individual tasked with  ;

the ALARA program was assisted by no fener than ten full-time contractor support

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I personnel. In July.of 1986, as outage work diminished, ALARA staffing was i decreased to a level of two (one full-time and one part-time). .

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We contend that we have been effective in limiting personnel radiation exposures over the years but agree that we should formalize our practices in 5 policies and additional procedures. We filled the ALARA Coordinator position j-and. issued a Radiation Protection Policy in December 1986, and in March 1987 i issued the ALARA procedures in final approved form. We are confident that these j actions and continued diligence to minimizing radiation exposures will result in i a very effective ALARA program. j i

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' REFERENCE:' B. Radiological Controls, Analysis, paragraphs 2 and 3

"However, the'1ack of active and timely plant staff review, understanding,

<and evaluation of the contractor's work packages and dose-saving techniques, also contributed to the increased expenditure." j

" Contractor activities associated with the control rod drive insert / l withdrawal line. hanger modifications did not benefit from historical radiation ;

dose rate data. As a result, increasing radiation fields were experienced as L the plant restarted which resulted in' unexpected additional exposure and which, j similar to the RPRP outage, made the original dose estimates inaccurate."

Comments i We believe that your statements concerning the control rod drive modifica-tion do not accurately portray the circumstances of the work activity. Dose rate projections were in fact made based on historic information, however, an unexpected change in plant conditions resulted in somewhat higher radiation j levels;than had been predicted. The' details of this job have subsequently been ;

discussed with the NRC inspector who originally reviewed tnis job and we believe I has resulted in a'better understanding of the events and circumstances related i to these activities.

I REFERENCE: B. Radiological Controls, Analysis - page 12, paragraph 3

" Preplanning for significant radiological work activities for the upcoming 1987 outage and for the planned spent fuel pool expansion project continues to be fragmented without clear ALARA goals, objectives and procedures."'

Comments During the assessment period (October 1985 - December 1986), the scope of the 1987 outage work in general and fuel pool work specifically was not defined

, to the point that would allow meaningful ALARA review. Beginning in January 1987, the ALARA Coordinator began formally reviewing outage work packages.

Particular attention has been given to and will continue to be given to any work

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in the spent fuel pool. The ALARA Coordinator has been in close contact with all departments involved with the fuel pool rerack to ensure the work scope is ;

fully understood and that the radiological dose is minimized.

REFERENCE: C. Maintenance and Modification, Analysis - page 16, paragraph 3

"The fact that the Licensee was unaware of these vendor suggested product change appears as examples of deficiencies in the program for the control of ,

vendor information." '

' Comments Vermont _ Yankee and the industry as a whole has been involved in directed efforts'to better assimilate vendor information into plant rtintenance and

' operating practices. Consistent with this objective, Vermont Yankee implemented i

m a 5-3,

I the industry sanctioned V-TIP program. V-TIP procedurally required all incoming vendor technical.information to undergo plant review. We recognize the need for additional improvement in this area and are therefore continuing to work toward meaningful program enhancement.

. REFERENCE: C. Ma'intenanc( and Modification, Analysis, page 16, paragraph 4

"A second example concerns the failure of post-installation tests to iden-tify three failed IRM detectors and the cross connection between two IRM chan-nels following replacement of all six detectors during the outage."

Comments We remain unaware of any post maintenance test techniques that would have i enabled us to identify the type of problem associated with failed IRM detectors l prior to reactor er.cicality. We, therefore, request that this-example not be l cited as indicative of inadequate post maintenance testing. I Aside from this, we agree the examples mentioned are indicative of post- ,

maintenance shortcomings. It should, however, be recognized that the generally i smooth startup and trouble free operation following the extensive work performed i during the 1985/86 outage is indicative of sound maintenance and post-maintenance test practices.

Test procedure revisions have been undertaken to better allow the iden-tification of post maintenance deficiencies in the cases cited.

REFERENCEr C. Maintenance and Modifications, Analysis - page 18 paragraph 1

"The resurvey validated the basic adequacy of the 1980 work, but identified j four additional walls in the heating and ventilation corridor that had not been )

identified previously in the IEB 80-11 response." '

l Comments l

We believe that the walls were not identified in the 1980 Bulletin response j because the turbine building which is classified "non-seismic" was not con-sidered within the scope of the original Bulletin. This was specifically stated in the Vermont Yankee letter to the NRC dated September 19, 1986. The walls were identified and are now being addressed solely as a result of the conser-vatism inherent in the re-survey initiated by Vermont Yankee.

Response tgt Board Recommendations Resolve the Masonry Wall and RHR Inspection issues expeditiously, o Resolution of both issues is well under way. RHR inspection work was started April 20th and is scheduled for completion in May 1987.

Modification relating to the Turbine Building Block Walls is expected to be completed by October 1987.

Review program for control of vendor information and change as necessary to assure that vendor information is current.

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s o We will review our existing program in an effort to increase its effec-tiveness. We continue to participate in industry efforts for improvement in this area.

Review the status of actions taken in response to IE Bulletins and Notices to identify areas requiring additional work or attention.

o We are awaiting additional information from the Commission allowing us to focus our review efforts. Once the scope has been more clearly defined, we will begin a timely reassessment of the identified documents. Following completion, an audit will be performed to validate the effectiveness.

REFERENCE: D. Surveillance, Analysis - page 20, paragraph 2

"One procedural matter still open from the last SALP period concerns the need to add Type B and C leak rate test results in the total containment leakage determination."

Comments -

While total leakage is informally calculated by comparison of A test ;

results with the sum of B and C results, we are unable to identify any specific '

10CFR50 Appendix J requirements which requires adding Type B & C identified leakage to the Type A test results.

We understand that the Appendix J requirements are presently undergoing Commission review to, among other things, provide clarification of this point.

Once the revised regulation is issued, procedure modification will be made as necessary.

REFERENCE: E. Emergency Planning, Analysis - page 22, paragraph 4

"There were, however, recurring weaknesses noted in the operations of the Technical Support Center (TSC) including: 1) lack of coordinated actions between the Center (TSC) and TSC personnel; 2) failure of TSC personnel to aggressively follow and coordinate plant activities; and 3) excessive noise levels in the TSC."

Comments Vermont Yankee is pursuing actions to correct the causes of the noted weaknesses in TSC performance. The facility is being enlarged, organized, and rearranged to improve efficiency and effectiveness. Team drills will be con- ;

ducted, involving the TSC staff and the facilities with which the TSC inter-faces, to improve the performance of specific TSC functions as well as coordination among facilities. Applicable implementing procedures will be revised, if required, to provide assurance of consistently effective performance by the TSC staff.

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Additional steps taken to improve the Vermont Yankee emergency preparedness program include selection of an experienced employee with excellent qualifica-tions to serve as the full-time Emergency Planning Coordinator. He will provide the point of contact for all emergency preparedness matters. This position has been upgraded and the reporting responsibility has been changed from the Plant Health Physics Supervisor to the Assistant to the President, who is highly 1 qualified in emergency planning.

REFERENCE: Licensing Activities, Analysis - page 41, paragraph 3

" Additional management attention is needed to assure timely closure of unre- l solved items from Generic Letter 83-02 pertaining to Technical Specifications l requirements for NUREG 0737 items." l l

Comments The Generic Letter 83-02 identified 13 items requiring Technical Specification modification. Of these, we consider 9 to be resolved. Of the items that remain unresolved (I.A.1.3; II.K.3.3; II.K.3.13 and II.K.3.22),

Vermont Yankee has submitted position statements on each. In an effort to expe-dite resolution, we will be contacting the VY Project Manager in NRR to ensure that the necessary information has been provided to allow further evaluation of ;

our requests.  !

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