IR 05000271/1984025

Revised SALP Rept 50-271/84-25 for May 1983 - Oct 1984. Detailed Evaluation of Util 850301 Comments Encl

ML20128P245
Person / Time
Site:	Vermont Yankee
Issue date:	05/28/1985
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20128P236	List: IR 05000271/1984025 ML20128P231
References
50-271-84-25, NUDOCS 8506030493
Download: ML20128P245 (67)

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ENCLOSURE 2 U.S. NUCLEAR REGULATORY _ COMISSION REGION I-SY'STEMATIC ASSESSMENT OF LICENSEE PERFORMANCE INSPECTION REPORT 50-271/84-25 VERMONT-YANKEE NUCLEAR POWER-CORPORATION VERMONT YANKEE NUCLEAR POWER STATION ASSESSMENT PERIOD: MAY 1, 1983 - OCTOBER 31, 1984 BOARD MEETING DATE: DECEMBER 11,.1984

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' SUMMARY ~. . . . . . . . . . . . . 42 TABLE 2~- VIOLATION SUMMARY . . . . . . . . . . . . . . . . 43 TABLE 3 - INSPECTION REPORT ACTIVITIES .......... 47 y. 1: TABLE _.4 -~ LISTING OF LERS BY FUNCTIONAL AREA ....... 50

. TABLE 5 - LER SYN 0PSIS .................. 51

TABLE 6~ SUMMARY OF LICENSING' ACTIVITIES.. . . . . . . . . 55

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I. INTRODUCTION

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.A. Purpose'and Overview-The Systematic Assessment of Licensee Performance (SALP) is an. integrated NRC staff effort to collect the:available observations and data on-a,

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-periodic basis'and to evaluate licensee performance based.on this infor-mation. SALP is supplemental to normal regulatory processes used to.

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, ensure compliance with NRC rules and regulations. SALP is intended to-

be.sufficiently.. diagnostic to provide a rational basis for allocating-NRC; resources and to; provide meaningfull guidance to the. licensee's

.. management to promote quality and safe! plant construction and operation.

.A NRC SALP Board,. composed of--the staff members-listed below, met on-December 11, 1984-to review the collection of performance observations and data to assess the licensee's~ performance in'accordance with the.

guidmce in NRC Manual Chapter 0516,1" Systematic Assessment.of; Licensee Perfa.mance". A summary of the guidance and performance criteria is

provided in Section II of'this report.

This report is the SALP Board's assessment of the licensee's safety per-formance at"the Vermont' Yankee Nuclear Power Station for the eighteen - .

.. month period of May 1, 1983-through October.31,L1984. The length of the s

review period is reflected in'the number of inspection hours and in the scope lof NRC observations and findings.

B. SALP Board Members R..W.- Starostecki, Director,~ Division of Projects and Resident

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Programs (DPRP) ~ .

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E..C. Wenzinger,-Chief, Projects Branch ~No. 3, DPRP.

O" L. E.- Tripp,. Chief,-Projects Section No. 3A'-DPRP

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s V. L. Rooney, Licensing Project Manager, ORB No. 2, Office of Nuclear Reactor Regulation -(NRR)

S.' D.- Ebneter, Chief, Engineering Programs Branch, Division of Engineering and-Technical Programs-(DETP)

B. Sheron,iChief,. Reactor Systems Branch, DSI, NRR

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W.:J. Raymond, Senior Resident-Inspector-

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OtherIttendees G. W. Meyer, Project Engineer, RPS 3A, DPRP

.a ( C." ; Background

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-Licensee Activities

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The facility was involved in a major refueling and maintenance out-age at the start of the assessment period. The scheduled eight week-outage was extended by about 6 weeks to evaluate and repair cracks

"in-the-recirculation system piping. Major modifications completed

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during the outage included. torus modifications from the Mark I long term program, including resupport of the torus attached piping; installation of alternate shutdown systems per Appendix R require-ments; and, installation of a recirculation seal injection system.

The facility started up from the outage on June 17, 1983. Two scrams occurred during the escalation to full power. The first occurred on June 20, 1983 when the condenser low vacuum switches were bypassed.too soon. The reactor was manually scrammed on June 29, 1983 when the reactor recirculation pumps were tripped inad-vertently during a logic test. The reactor reached full power operation on July 5, 1983.

The facility operated at full power from July 5, 1983 until June 15, 1984, with the exception of the following unscheduled outages:

a reactor shutdown was initiated on August 26, 1983 to repair a steam --leak in the main turbine bypass valve steam sealing supply line; the reactor scrammed automatically cn August 27, 1983 on high vessel-level due to an operator error when switching from auto to manual feedwater control; the reactor scrammed automatically on high pressure on January 5, 1984 due to a malfunction in the turbine electrical pressure regulator; the plant was shutdown on January 20, 1984 to replace the main turbine expansion joints; and, the reactor automatically scrammed on April 16, 1984 due to an MSIV isolation caused by failure of MSIV.80C during routine testing.

The power coastdown to the 1984 refueling outage began on May 20, 1984.

An Alert emergency was declared on June 15, 1984 when a TIP detector failed to stop in the shielded position upon withdrawal from the core during routine surveillance. The unshielded detector created contact dose rates of 100 R/hr on the drive. housing, and general area dose rates of 5 R/hr in the Northwest corner of the Reactor Building-252 foot elevation. The licensee activated the emergency

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response centers and responded well to protect plant workers and control the unshielded probe.

The facility was involved in a scheduled refueling and maintenance outage from June 15 until August 6, 1984. Major activities during the outage included the completion of modifications to provide z envirenmental qualification (EQ) for electrical equipment, and the examinaticn'and repair of recirculation system welds. Following

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plant startup on August 6th, the reactor was shutdown to replace connectors on the MSIVs on August 9, 1984 The connectors were installed during the outage as part of the EQ upgrade and failed after several days of plant operation. The plant was shutdown from August 13-15, 1984 to identify and repair tube leaks in the main condenser. Subsequent plant operation was limited to about 80%

full power due to the failure on August 12, 1984 of one of the three condensate pumps.

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Following repair of the condensate pump, plant power was increased-to rated conditions for the first time since the outage. .An anomaly

.in the core power to flow ratio was first noted on September,11,

.1984. Actions were-taken to study the core conditions and identify

.the possible.causes for'the anomaly. A suspected problem with the-

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steam separator assembly was' confirmed on September 16, 1984 when lthe~ anomaly was verified to occur only at core flows above a cer-

-tain value during routine surveillance testing.

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The plant was-shutdown on September 18, 1984_following a series of discussions with-the NRC staff to examine ~the vessel internals. The examina-

tions confirmed that the separator was not securely bolted to.the.

core shroud-assembly'by Maintenance personnel during vessel assembly-in August,:1984.' The. plant was restarted on September 29, 1984~

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following a ten' day outage.

During power ascension on September 29, 1984, MSIV 800 failed during routine testing,-and the plant was' shutdown to effect repairs.

Power. operation resumed on October 1,,1984 and continued until Oc-itober 23, 1984~. _ Both diesel generators _ failed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> while in a standby conditions during. steady _ state plant operations on'Oc-

-tober 22-23,- 1984. ,The diesels:became inoperable due to a generator

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' lockout condition caused by-spurious operation-of the; generator-differential _ relays in both diesels. A plant shutdown was initi-

~a ted. The differential relays were repaired and returned to ser-vice.

The_ plant was operating at rated power at the conclusion of:the

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. assessment ~ period.

C L2. tInspection Activities

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in One NRC resident! inspector was' assigned to the site during the en-tire assessment period. The total NRC inspection. hours'for the

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• period ~was:3903 hours (resident and' region based) with a distribu-tion in the appraisal functional areas-as shown in Table.1.

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A special team inspection of the' facility for' compliance with the l requirements of 10 CFR 50, Appendix R, Section III.G was. conducted

, from. August 29, - Septemberi2, 1983.

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NRC Emergency Preparedness Inspection Teams observed the EOF-IN

'if emergency exercise on' August 11, 1983, and the EOF-0VT emergency 2 exercise on September 21, 1983.

-Tabulations of Violations and Inspection Activities are attached as Tables ~2 and_3, respectively.

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~II. CRITERIA Licensee performance is assessed in selected functional areas, depending on whether the facility is in-a construction, preoperational, or operating phase.

Each functional area normally represents areas significant to nuclear safety and the. environment, and are normal programmatic areas. Special areas may be added to highlight significant observations.

One or more of-the following evaluation criteria were used to assess each functional area.

1. -Management involv'ement and control in assuring quality 2. Approach to resolution of technical issues from a safety standpoint 3. Responsiveness to NRC initiatives 4. Enforcement history 5. Reporting and analysis of reportable events 6. Staffing (including management)

-7. Training effectiveness-and qualification-

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However, the SALP Board is not limited to these criteria and others may have been used where appropriate.

Based upon the SALP Board assessment, each functional. area evaluated is classified into one of three performance cateoories. The definitions of these performance categories are:

Category 1. Reduced NRC attention may be appropriate. Licensee management attention and involvement are aggressive and oriented toward nuclear safety; licensee resources are ample and effectively used so that a high level of performance with. respect to operational. safety or construction is being-

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achieved.

.- . Category 2. NRC attention should be maintained at normal levels. Licensee management attention and involvement are evident and are concerned with nuc-

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lear safety; licensee resources are adequate and reasonably effective so that satisfactory performance with respect to operational: safety or construction

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is being achieved.

Category 3. Both NRC and-licensee attention should be increased. Licensee

' management attention or involvement is acceptable and considers nuclear safety, but weaknesses are evident; licensee resources appear to be strained or not effectively used so that minimally satisfactory performance with re-spect to' operational safety-or construction is being achieved.

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- The SALP. Board has' also categorized the performance trend over the course of

.the SALP' assessment period. The.' categorization describes.the general or pre-vailing tendency-(the performance gradient) during the SALP period. The performance trends are defined as follows:

' Improving: . Licensee. performance has. generally. improved over the course of the:SALP assessm'nt e period.

Consistent: Licensee performance has. remained essentially constant-over the course of the:SALP assessment period.

-Declining:

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Licensee. performance has generally declined ov'er the. ' ,

. course of the SALP: assessment period.

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.III.-SUMARY OF RESULTS-

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A. Overall Facility Evaluation

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'During the previous assessment period, increased licensee managementi i attention was identified as being ~ necessary in the functioaal areas of Plant l Operations, and Refueling and Outage Management to achieve im-: _ ,

provements. LSpecifically,< management attention was necessary to assure.

personnelfadhered to established. procedures and policies, and.to maintain an aggressive approach to resolve operational problems. Improvements were noted in both functional areas during the current assessment period.

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This assessment.noted numerous personnel errors during the performance

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of routine duties in the surveillance, radiological controls,Eoperations and refueling functional areas. The_ errors resulted from either a lack

of: attention to details during .the_ performance of: routine ~ dutes or an over-reliance on' experience as a substitute for strict adherence to es-

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Ltabl.ished procedures.- Increased management attentien is warranted to assure an adverse trend does .not develop and: lower performance. Licensee

- management has-been responsive to NRC concerns regarding personnel er-'

-rors by evaluating available performance data-and addressing the issue with the plant staff. '

More' aggressive management' involvement is needed in the radiation pro-1

! 'tection area to correct programmatic weaknesses by formalizing _the ALARA g' program, assure that anomalous conditions are aggressively pursued and

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resolved, and enforcing a rigorous frisking. policy to assure _ licensed'-

. material is adequately _-controlled. Modifications and corrective actions

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Lto meet Appendix-R requirements should be completed in accordance with

. commitments made to the NRC staff. ' Additional management attention is required in.the' Maintenance area.to strengthen supervisory oversight and -.

QA/QC controls. 1

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The strength._of theLlicensee's_ management controls is most' notable in the' generally conservative approach taken to assure safety in plant-0 '

_ operations, the planning'and control _of outage activities and design changes,_:the effective housekeeping program, the security area,=the com-

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b pletion of licensing actions, and in the preventive maintenance and

. operational surveillance programs. The pol. icy for operation _under. ap- '

? Kparent anomalous operational conditions should:betreviewed and strength-ened as necessary to assure conservatism ~ in plant operations is main-

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'tained:at previously observed levels. . Significant improvements.were- ,

lnoted_in.the emergency preparedniss area.

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Category . Category Functional Area ~ Last Periodi -This Period Trend

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(May 1, 1982 . -(May 1, 1983 -

April 30, 1983)

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October 31, 1984)

^A. Plant Operations ~ 2 1 Declining-

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Controls 1 2 Consistent'

C. Maintenance 1 2 Declining

- D .- Surveillancei 1 1 Consistent E. Fire Protection and Housekeeping 1 2 . Consistent-F... Emergency Preparedness No Basis 1 . Consistent-G. ' Security and Safeguards: 1 1- Consistent-1H. Refueling and Outage: Management. 2 1 Consistent i; 21.g ~ Quality. Assurance Not Assessed 2 Improving Separately

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/J. Licensing Activities' 1 l' Consistent-

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IV. PERFORMANCE ANALYSIS- ,

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~ Operations (28%)

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1L Analysis During the' previous assessment period, problems were identified in

.the areas of offsite review committee activities; completion of-

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50.59 evaluations for electrical jumpers; control _of_ containment

> valve _ lineups; licensee response to NRC initiatives on resolving containment isolation valve operability issues; and, violating sec-

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,' ondary containment while moving fuel, which also involved failure to report-the _ incident. in a timely manner. : The failure of senior licensed operators to strictly adhere to administrative policies during the incident was a significant concern. The need to take a more aggressive approach in'the; evaluation and resolution of -i

- equipment problems was deemed an area requiring greater. emphasis to improve performance. The licensee was generally responsive in .

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correcting deficiencies in these areas in a timely manner.

<Thislarea was under continuous. review by the' resident and regional inspectors during the current assessment period. Specialist in-spectors reviewed the'non-licensed and licensed operator requali- ,

fication training programs.

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, : Site and corporate-management have generally demonstrated a strong and effective commitment to safety in plant operations. Plant man--

agement and. supervisor reviews of daily plant status and periodic

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-tours of the facility are evident and demonstrate management:in-L ,

. , , .volvement and control of routine operations.- Corporate management routinely visits the plant and.their' involvement in routine plant

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.  : activities and in response to problems has been evident. Questions-regarding interpretation of technical specification LCO requirements are usually discussed with the NRC staff before the items become- '

-an~ issue.

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Operators are knowledgeable of plant status, technical specifica-tions, and procedural requirements. Licensee management satisfac- e

, torily resolved NRC issues such as TMI. Action Plan training items,

which have been' incorporated into the progfam. Two SRO and two R0

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licenses.were issued durinq the period. Plans to construct and

,. operate a site specific simulator are progressing and should be completed on schedule in~1985. Full staffing for the control room

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. per-NUREG 0737 Item I. A.I.3. was achieved early in the-assessment c" ' period. Operations staffing and organization remained adequate,

,' - but personnel changes did occut' in the Operations Superintendent E position, and a pending change in the Operations Supervisor position :

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has been announced. The number of changes in this assessment .

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.- period,)togetherwith.thoseinpreviousperiods,indicat'ethere

is'a need to achieve stability in management and leadership for the department.

Non-licensed. training programs.were well defined and. implemented.

E Instructor positions were understaffed, however,.while management attempted to fill positions vacated by promotions and resignations.

Minor programmatic improvements were found necessary, such as better

- documentation of attendance sheets and the. subjects covered during specific ~ training sessions-needed to be more' specific. Portions

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of,the maintenance. retraining were not completed in a timely manner.

However, subsequent inspection found that the' maintenance retraining-

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- was satisfactory and the requalification attendance records had been

- corrected.

The licensee has demonstrated-a proper regard to regulatory re-quirements and safety, as evidenced by. actions taken in response -

- to several operational problems. Examples included the follow--

ing: .the seismic upgrade currently in progress for the hydrau-

- li.c control' units; the compensatory measures taken in 1983-while supports:were upgraded on small diameter service water cooling-l lines;'.. installation, operation and maintenance of recirculation

- weld _ leakage detectors;-voluntary operation under reactor coolant

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~1eakage limits more restrictive .than those required by _the technical

- specifications;'the detection of defects in the uninterruptible

- power system batteries and replacement'of both battery banks; and, the detection'and replacement of defective cells-in~the A main

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station batteries. The. licensee was responsive in resolving tech--

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nical questions-raised by the NRC. staff regarding service water system performance', and concerns.regarding updated drawings-for the control room. Overall, licensee actions during this assessment

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period have demonstrated a greater degree of aggressiveness in-resolving' equipment problems.

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However, the management decisions and actions to continue plant operation'from September 16-18, 1984 with an anomalous core power-

. to-flow relationship'and in spite of. clear indications that the plant was operating in an unanalyzed condition, appeared as a sig-

n - nificant deviation from the normally conservative approach taken

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-k " to assure safe plant operations. NRC considered that the licensee

< had an insufficient basis to continue operation with the anomaly,

-.and the-licensee's decision to do so was neither prudent nor con-servative. While subsequent reviews and after-the-fact analyses confirmed that no unsafe conditions were created by operation with an improperly secured-steam separator assembly, the NRC raised questions regarding the licensee's philosophy related to operation under apparent unanalyzed conditions and the criteria used to as-

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sure adequate safety margins are maintained.

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seven violations of minor safety significance were identified during

, the current assessment period. .. Three of the violations involve improper valve or electrical breakers configurations,in systems importantito safety. The system valve lineup problems had minimal i~,4 '

isafety significance-in that no major flow path valves have been found: incorrectly positioned on safety-systems.

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One violation concerns'an administrative procedure for. valve lineup

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controls--that was inconsistent with technical specification re-quirements. NRC concerns associated with the item were first iden-

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tified to the licensee in May, 1984 and resolution of.the concerns was still in progress at the conclusion of the assessment period.

While further-discussions are required ~to resolve differences of-opinion regarding the issue,-the licensee has been less than fully

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. responsive to correct perceived weaknesses in the administrative p~  : controls under- review, which if left uncorrected,. could result in v a significant safety concern.

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a RAviews'of' plant operations by the resident inspector. and licensee personnel identified several instances during the assessment period'

where personnel' errors or performance either resulted in degraded performance for an; activity, contributed to an operational event, l' 1 or affected compliance with regulatory requirements. Examples in-cluded the following: several minor leaks were not noted by oper-

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ators during the 1983 reactor _ hydrostatic test; inadvertent scram in' August,11983 caused by'an operator not following procedures while

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shutting down the plant; tagging order not properly cleared from-instrument air system in-June,1983; tagging order not properly cleared on advanced offgas system in August, 1984 and resulted.in

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a glycol spill and intrusion into the radwaste system; tagging order not properly cleared from the core spray system in August, 1984;

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_ failure of senior licensed operator to follow startup procedure in April,' 1984 resulted in inoperable.HPCI system; and, failure to

. -' maintain cooling water supply valves to RHR' service water pumps aligned per procedures, and. failure of auxiliary ' operators to note

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the mispositioned valves during routine' rounds. ~Each incident above can be characterized as deficiencies in personnel performance rather

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than a disregard for administrative controls and procedures.-

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The number of events noted above (8) is of concern to the NRC, al-

, though no one incident had a measurable effect on plant safety.

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attention to detail during the conduct of daily activities. The number of personnel errors associated with tagging operations was smallrin comparison with the large number of component manipulations

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completed under tagging requests. However, personnel performance

-' problems with tagging requests were not limited to operations per-sonnel-(see Section H below). The. licensee was responsive to NRC

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-concerns about the occurrence of personnel errors on a plant wide n .

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basis and addressed this issue with.the plant staff. Additional licensee reviews regarding the implementation of tagging controls were in progress at the conclusion of the assessment period.

The number.of reportable events submitted for_this area is not con-sidered excessive. Equipment failures and the resultant LERs are expected in this area. No excessive failures occurred on any one system or component.

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Some of the equipment problems that have been noted (e.g. binding of MSIV actuator plates with bushings and guide rods) concern problems for which an engineering fix has been estab-lished and modifications for redundant components are being com-pleted in a phased manner. Event reports submitted by the licensee are generally complete and accurate. An evaluation of LERs for the period was completed by the NRC's Office for Analysis and Evaluation of Operational Data (AEOD). This review found that the reports were technically accurate,. complete, and intelligible to a knowledgeable safety engineer not intimately familiar with the plant. None of the events involved what AEOD would consider to be a significant event or plant safety issue. Planned corrective action taken by the licensee were considered to be commensurate with the nature, seriousness and. frequency of the problems found.

A conservative approach is taken in potentially reportable matters, particularly as regards 10 CFR 50.72 notifications. The licensee has been aggressive in identifying problems and reporting them to the NRC. The Plant Operations Review Committee has been effective in its' technical review of problems and proposed license amendments and has made recommendations appropriate to resolve identified issues.

2. Conclusion:

Category 1, declining. Improvements in the area of personnel.ad-herence to procedures and policies has occurred. The licensee re .

sponse to the anomalous core conditions near the end of the assess-ment period'is indicative of a non-conservative operational philo-sophy of concern to NRC which led to the declining trend categori-

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3. Board Recomendation

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-Licensee Additional improvements can be realized by a more aggressive re-sponse to NRC initiatives on procedure improvements, encouraging greater attention to details in the conduct of routine activities to reduce personnel errors, and, taking a more conservative approach in response to apparent operation with unanalyzed conditions.

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NRC Conduct routine inspection program and monitor performance for trends.

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1 B; ~ Radiological Controls (23%)

l 1.- Anal sis

.The licensee'.s Radiological. Controls Program during the previous '

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assessment period was categorized as Category 1 and no major prob--

.less were identified during that period.

' During the current assessment period, weakness.es in the licensee'.s s- . Radiological-Controls-Program were identified. Eleven inspections

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..were performed in the Radiological Controls Program resulting in nine identified violations.

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.1.1 Radiation ~ Protection

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Five inspections of this program area were conducted by Region

p' . I Radiation Specialists. These inspections included-reviews-during normal-and' outage operations and a special review for-possible transuranic contamination. -The Resident-Inspector

, reviewed ongoing radiation protection activities.

Procedures are generally well controlled and documented. 'One violation. concerned the failure to properly review and approve

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procedures for the operation of a van-mounted whole body .

counting system. The item was an isolated instance due to the temporary nature of the operation involved.

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Reviews of routine operations, planning and preparation for-the outages'and outage' activities indicated that a generally

effective radiation protection program was maintained. How-ever, management needs to formalize and improve the-"As Low v 'As Reasonably Achievable" (ALARA) program to support piping replacement in 1985. The ALARA program lacked.an adequately 4 stated and understood management policy statement providing

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~ a commitment to ALARA. The charter for the ALARA Committee

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failed to define the term "high radiation exposure-jobs" with-in.that committee's purview. Procedures for ALARA-instruc-

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-tions, pre-operational briefings,-use of engineering controls, practice in low radiation' exposure areas and scheduling tasks

,' to reduce radiation exposures were not in place. Records were not available of n.an-rem estimates for several outage activi-ties involving up to ten man-rem projected exposures.

The licensee's personnel contamination survey program had _.

1 programmatic weaknesses identified by the NRC. These weak-nesseswereduetoapoorlyunderstoodandfrequentlyignored policy for personnel ' frisking." These weaknesses were ad-

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dressed and corrected by the licensee in response to NRC in-L 't l-

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itiatives, including Confirmatory Action Letter 83-10. The

- discovery by the licensee of areas of contamination and un-controlled material outside the radiologically controlled areas

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(RCA) of the_ plant, both inside and-outside,the protected area,-

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indicates that additional controls are needed for the movement.

and storage:of licensed material,outside the RCA. The licensee-adequately. responded to these instances with technically sound.

corrective actions.

- An NRC-sponsored survey of-possible transuranic contamination in commercial nuclear plants.showed the possible presence of.

alpha-emitting radionuclides in the licensee's facility. The

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licensee failed to promptly investigate and resolve this po-S tential; safety issue when notified of the survey results. In-

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response to NRC initiatives,.the: licensee determined that

- alpha-emitting radionuclides were not present in sufficient'

quantities .to significantly contribute to worker exposures.

The radiation protection organization.and staffing level.were

- generally adequate to support normal operations and the.out--

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- ages. Selection, training and qualification programs for.re-placement personnel in radiation protection were also gener-

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. ally adequate and contributed to generally acceptable personnel performance.and adherence to procedures during the outages.

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Documentation of radiation protection activities was generally '

complete, adequately maintained and available. However, supervision-approved changes to requirements in radiation work permits'were not1 documented in'at least two. instances.. Dosi-metry records were well-organized and available.

The total personnel exposure reported for the facility for 1983l

- was 1528 MAN-REMS, which includes exposure for the~recircula-tion pipe repairs,. and -is about. average for.BWR plants of the same age.

l 1.2 Radioactive Waste Management / Effluent Controls a

Four inspections of.this program area were conducted by.Re-

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gional Radiation Specialists during this assessment period.

These' inspections-reviewed normal operations and the discovery

, , by the licensee of radioactive contamination outside the

radiologically controlled areas of the plant. The Resident-Inspector also reviewed ongoing activities in this program.

area.

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One-violation concerned the failure to properly record test results for Standby Gas Treatment System Train B. This viola-tion was an isolated event and suggested a lack of attention to detail in re~ viewing the results of one particular test.

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The;11censee's response to this item was timely and. adequate.

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to assu_re that supervisory personnel would review test results.

A_second violation concerned the: failure to adequately imple-

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ment the. requirements of NUREG 0737 Item II.F.1.2 (effluent

i monitors for particulates and noble gases) and Item II.F.1.3 .

(containment high range monitor) in accordance with the March 1983 Confirmatory Order. The items concerned.a failure to-meet-the.specified design requirements for the containment

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monitor, and'a failure to provide a technically' adequate de-sign for the effluent monitors, and suggest a lack of manage-

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ment involvement to assure' quality-in the. design engineering.

'However, the licensee.has disagreed with the second. violation Jand this matter was still under review at the conclusion of.

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.the assessment period.

An effective radioactive waste management and effluent controls program was maintained.- Planned' releases of liquid radwastes were minimized as result of prior planning for contro1:of activities. . There were no unplanned releases during the as-

.sessment. period.

Reviews of staffing and organization structure showed all

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. positions are' adequately identified, authorities and respon--

. ,

-sibilities are defined,-and a generally adequate staff avail-

'able. : Annual retraining.in applicable procedures, Technical Specifications ~and related areas is required by the~ licensee.

. This retraining; program is generally complete and contributes to a generally acceptable level of personnel performance with few' personnel errors.

Radioactive waste ma'nagement and effluent ~ control procedure's are generally complete, adequately maintained and available.

However, several minor technical-inadequacies in procedures were. identified. Plant procedures failed to provide instruc-tions for. converting stack gas monitor readings to gaseous

. release concentrations. .A' calibration procedure for the main v. steam line monitors did not specify that the monitors must be calibrated when found out of tolerance during functional tests, cand did not adequately address how to determine background levels to establish the trip setpoints. Four procedures for iodine chemical separation of the reactor coolant sample did-

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not require calibtation of the iodine carrier. The routine

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environmental program failed to detect the small but measurable of cobalt-60 in the Connecticut River at levels up to'

buildup 750 pCi /kg at the discharge of the site North storm sewer, due to.a source originating from the turbine building roof vents.

Increased attention to technical detail in procedures during revision and review is needed.

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e n. '

Quality assurance audits of radioactive ~ waste management.and effluent control areas were timely, generally thorough and performed in accordance with the licensee's Technical Speci-fications.' Actions takenLin response to audit findings were

~

- generally timely and thorough.'

Four of the seven LERs for this functional area involved missed surveillances which resulted from personnel. errors committed by failing to pay attention to details. Plant _ management re-view of personnel performance deficiencies is. warranted to assure adverse trends do not occur.

.

1.3ITransportation -

One onsite inspection of this area was-made by_ regional-radi-ation specialists. Additionally, receipt inspections by a representative of the State of Nevada and by Boston' Edison

' Company employees of two licensee shipments were reviewed

'during the assessment period. Three' violations were identi-fied: failure _to transport licensed low specific activity

'(LSA) material in a strong tight package;. delivery of a spent.

resin shipment exceeding 200_ millirem per hour on the external'

surface of the'. package to the Beatty,-Nevada burialisite and-failure to provide training to licensee employees performing-inspection activities'affecting quality of licensed shipments.

An' enforcement conference and a management meeting _were held to discuss'these violations with the licensee. The State of Nevada temporarily suspended the licensee's burial privileges.

The licensee has documented the specific responsibilities assigned to the' Operations, Maintenance and Chemistry and Health Physics Departments in plant procedures. Procedures

, a affecting shipping activities of these departments were revised Lto reflect changes in 10 CFR 71 and DOT regulations effective during the assessment period. Records for material shipments were complete, well maintained and kept for a period in excess of_the two year requirement.

4 Quality assurance audits of the transportation area'were con-L ducted by-technically qualified personnel, and were generally-timely and addressed most aspects of the program. However,-

, the audits failed to identify the weakness in the training of quality control inspectors.

2.- Conclusion Category 2, consistent.

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3. Board Recommendation

' Licensee

'FormalizeLALARA program policies and procedures. .Take a more ag-gressive approach to investigate and resolve indications of anoma-

'lous radiological conditions. Maintain and enforce a strong-frisk-

-ing program to. reduce the probability for inadvertent release of material from the radiological controlled area. Strengthen controls for the preparation and QC review of radwaste shipments. Emphasize attention to detail in the conduct of routine duties and procedure reviews.

NRC

The inspection frequency in the transportation area should be in-creased. The inspection frequency in the remaining areas should remain as prescribed by the routine program.

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C. . Maintenance (3%)

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1 TL Analysis' ,

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s- .There were no significant deficiencias. weaknesses or violations identified in this area .during~ the -ins. assessment period and.per-

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formance for the area was rated as Category 1. The preventive- b maint'enance program was. identified.as a-notable strength in 'the

licensee's system of management controls.-

~

Duringthis.assessmentperiod'thearea~wasundercontindulreview:

by the resident inspector. 0ne;in'spection conducted by a Region

~ based inspector reviewed the maintenance program to determine.the

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extent to which-maintenance _ practices.mhy contribute to_ system un-

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availability.- Specialists conducted tso inspections in this area.

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m cThe second inspection was conducted to followup deficiencies iden-tifiedLduring the first inspection.

. . .

Reviews of items' requiring'mainte' nance-found that safety related

' items are given priority attention. -The Instrument & Control and=

~ * ~ Maintenance Departments are staffed by ' experienced personnel and

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> personnel turnover has been low. Supervisor monitoring'and in-volvement in-daily work activities is evident and remains an element

,

' V _of. strength. Craft and supervisory personne! demohstrated a good

,, working. knowledge _ of plant systems and components within;their realni

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. of expertise. -Completion of an SR0 certification program by the~ ,

l tional Maintenance-area. . Supervisor.-is There:is a generally an asset thatregard good will strengthen;the func y?

for administrative

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r and proceduralfrequirements. The special inspection of maintenance practices.; identified no programmatic elements that would adverse impact.on equipment availability.- The licensee'scause

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am e progrAa .,

assures that equipment. failures'are evaluated for frequency of-occurrence and root.cause, and that maintenance errors are detected,-

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evaluated and corrected. A recent' example of the latter involved J . the determination by the I&C Supervision of an improper powey supply that wasiinstalled in the ECCS vessel level instruerentation channel during routine corrective ~ maintenance.

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, The' preventive' maintenance program continues to be well documented'

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in both the Maintenance and Instrument & Controls-areas. The wel1~ \ ,

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' maintained Visirecord system is an asset in the' licensee's review

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of equipment failure histories and trends. Deficiencies-in equip-4  ; ment performance were identified and corrected..~A notable example

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. includes the degraded performance identified and corrected on the-

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- 125'VDC station batteries and the 480 VDC uninterrupt ble power .

Lsupply batteries. Notwithstanding the above, a sign 11 cant equip- l E, ment problem occurred at the end of this assessment period, when '

}T - two -independent,1 redundant relays failed nearly simulianeously and

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-rendered both emergency diesel ' generators inoperable. Licensee actions;to evaluate and eliminate potential common mode-failure

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-mechanisms.are being'followed by the NRC staff.

_

The lisensee program for maintenance and surveillance of' pipe sup-

" ports,and_ restraints remained effective. However,-an item regarding-upgraded snubber technical specifications has been a long standing

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issue'with the licensee ~that has'.just been recently resolved. Fol-

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lowing the NRR request in November, 1980 that_the licensee revise the snubber technical. specifications, and after two meetings (May,

1983 and February, 1984) initiated by the NRC staff to resolve the

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(technical-issues, the licensee. submitted a proposal.to. change the

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itechnical. specifications in Octobe" '1984. The licensee's reluct- .

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..ance to respond to the Staff's request caused the. unnecessary ex--

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'penditure of' additional NRC resources to resolve the issue in a N ~

' timely manner.

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Two of the violations identified during this period concern problems-

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in-the documentation of work activities in the' corrective-mainten-

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-ance program, which appear as a programmatic weakness in assuring

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the quality 'of. the completed work.' Examples of these problems in-

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'cluded: incomplete or inaccurate information on maintenance work

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requests regarding the scope ~of work required;' incomplete procedural

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-instructions regarding.the requirements for independent inspections; a lack'of documentation of independent inspections; and,-'a failure

~m - to' define-operational: testing and acceptance criteria in all cases.

The exact work done:and replacement' parts used during maintenance was not determinable'in.all cases ~after the work was done.- The

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incomplete-information recorded on the maintenance requests. suggests a lack of attention 1to detail in maintaining records. Workers were

- 'J not; knowledgeable of:the intent and meaning of~ administrative re-

.' .quirements regarding documentation'~of work scope and: independent Linspection. ;The' post maintenanceLreview of work requests was in- _

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adequate since it failed to identify the lack of documentation N of: parts used, testing completed,'and inspections conducted. QA

, ireviews:of these areas also appears inadequate since'the above

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. problems were not noted during QA audits. T.he' licensee was r.espon-

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.sive to the NRC concerns since a followup inspection in this area

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indicated increased management attention and general improvement g, ,

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in the~ documentation of maintenance activities.

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The third. violation in-this area stems from the failure to properly

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bolt the steam separator to the core shroud during vessel reassembly

_ following.the 1984 refueling outage. Although the procedure used

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for the evolution was satisfactorily _used perhaps a dozen times in

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the past to set'the separator, it was inadequate to assure that the

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separator.was bolted tightly in place. The licensee reviewed the

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_ event and determined that-the maintenance crew performing the work

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in July,1984 did not have a full understanding of the bolting

.i * process and required further training. This event demonstrated

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that the. licensee's-QC and QA controls were' insufficient to detect

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.the improper. procedure, training and bolting prior to subsequent-

' plant operation in that condition.

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iThe' licensee ~took the. initiative during this. assessment period to

,  : provide SR0 certification for managers in the department. This-

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action created a~ temporary' vacancy in the Maintenance Supervisor

_ position from about February to April, 1984. There was also a vacancy in'the Maintenance-Superintendent position from about Feb-

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"ruary :to October,1984,. that was' created at' first when the incumbent

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attended the SRO certification program,.and was extended later when'

s that individual was promoted to a new position. The Maintenance-tand I&C-departments were temporarily realigned-under the Operations'

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.and Technical Services Superintendents during'.the interim period.

LA new appointee for the Maintenance Superintendent position began

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work ~in_0ctober, 1984, and is being gradually phased into the po-sition. The~ temporary vacancies created in these. key management 1 positions and.the extended time taken to. fill the vacancy in the iSuperintendent position' reduced the level of management. attention-for the; area and created a lack of continuity in management over-sight and control.

. = Two licensee event reports (LERs) are listed for this functional e _ area. .One event concerned the

installation by'I&C personnel of

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-the wrong type (gamma sensitive), detector for a stack gas-monitor-sduring routine preventive maintenance. The second LER concerned the failure by maintenance personnel to properly bolt the steam separator to the core shroud'in July, 1984. The. failure of main-

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tenance personnel to strictly follow procedures during vesse1< dis-

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assembly in September, 1984 resulted in a security event and was-L.a licensee identified violation-of the technical specifications.

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.These events are indicative of personne1~ performance problems and-

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ssuggest a lack of attention to detail:during the performance of-routine" duties. 'The' licensee has been. responsive to NRC concerns

'

by reviewing and responding to' apparent trends in-this area ~and by addressing these concerns in a memorandum to all_ plant personnel.

.

2. -

Conclusion

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Category 2,. declining.

J ~ 3.- Board Recommendations

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Licensee Strengthen management oversight and QA controls to assure previ-a , . ous;1evels of performance are re-established and' maintained.

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- NRC Monitor licensee performance through the routine inspection program.

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- D. . Surveillance (5%)

11. -Analysis There were no significant concerns identified in this area during the previous. assessment period. Two violations of minor safety significance were identified during the current assessment period.

-The implementation of an operational surveillance program by an experienced staff remains a significant strength.

Surveillance activities during normal operations and refueling out-ages were reviewed-by the resident inspector during routine'inspec-tions. There were two specialist inspections in the areas of con-

tainment leak rate testing and snubber testing. Two additional inspections by Regional personnel reviewed the in-service inspection program and nondestructive examination for repair of recirculation system welds.

'The licensee performed recirculation weld inspections per IE Bul-

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letin 83-02 during the 1983 outage, and per NRC Generic Letter 84-11duringthe1984 outage. No problems were noted during the NRC review'of the licensee s program for augmented ultrasonic examination of recirculation piping; the ISI program, data, and results; and, personnel certifications and qualifications. The licensee maintained adequate control over the recirculation weld

. examinations by direct surveillance of ISI vendor activities. The results of the evaluations were promptly evaluated by the licensee and appropriate corrective actions were implemented to correct unacceptable conditions. The licensee cooperated with NRC initi-atives to-independently measure and evaluate cracked welds in the recirculation piping.

Surveillance activities were completed in accordance with the established procedures. Planning and staffing are adequate. Per-sonnel are well experienced in test activities and associated procedures, and are knowledgeable of the facility, its operation and the equipment under test. There is a generally good regard for administrative policies and procedural controls. One excep .

tion' concerned the violation identified in the I&C area, where technicians exercised poor' judgement by terminating the torus level instrument calibration prior to formal completion of the procedure.

-Surveillance records are well organized and readily retrievable.

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Surveillance results are trended to identify and trend equipment problems. There are relatively few (5) licensee event reports (LERs) in this~ area due to instrument setpoints found out of technical specification limits. This record demonstrates that

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'the~ practice of-trending setpoint drift and adjusting the sur-veillance interval accordingly is working well. Surveillance test results are consistently reviewed by supervisory personnel.

One notable exception concerned the violation regarding the in-operable seismic monitoring instrumentation. However, this item was an isolated incident and does not significantly detract from otherwise good performance.

-Surveillance and testing procedures are generally well-written and technically accurate. A notable exception concerned two procedural problems that resulted in improper ' instrument setpoints for the SLC system (LER 84-13) and the main steam line radiation monitors (LER 83-25).- Two instances involving inadequate procedures also caused plant transients, as follows: inadvertent ECCS actu-

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.ation and depressurization during hydrostatic test with the plant shutdown due to manipulation of instrument valves for PT 56D; and, inadvertent ECCS actuation while shutdown caused by a valving manipulations during a level instrument calibration. Resolutions to technical issues in all areas reviewed were generally sound.

However,~one item for improvement concerns the licensee's failure to implement the NRC position on repairs and adjustments to con-tainment boundaries as part of the containment leak rate testing.

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Another item that warrants continued management attention is the occurrence of. personnel errors during the performance of routine duties. Errors committed by technicians resulted in the follow-ing during the inspection period: five (5) instances where_the required surveillance was either not done, or the results were lost, due to personnel error; and, plant trip (manual scram by.

operators) following loss of both recirculation pumps due to inadvertent ECCS actuation caused by technicians while securing

.from a logic test. Each instance above involved a failure to pay attention to details.

The number of reportable events for this area (17) is not con-sidered exceptional. Equipment failure remains the predominant cause for reports submitted in this area. The identification of equipment problems during testing and the resultant LERs are expected.

2. Conclusion Category 1, consistent.

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._Er

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r l 3. JBoard Recommendations

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Licensee

Increase attent'on.to~ detail during the- performance of. routine-

. duties to assure level of'beirformance does not. decrease.

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E. Fire Protection and-Housekeeping (9%)

1. Analysis One minor violation (level V) was identified during the previous assessment period, concerning the failure to meet the conditions of a fire control permit during welding operations. Performance in this area was assessed as category.1'due to the strength of the licensee's controls in both the fire protection and house-keeping areas, which were evident by the physical plant condi-tions and a history.of extensive work activities without any majorincidents.

Routine implementation of the fire protection program and plant housekeeping conditions during normal operations and outages was reviewed continually by the resident inspector during the as-

. sessment period. A review of the licensee's implementation of the 10 CFR 50, Appendix R requirements was completed during the assessment period.

Plant cleanliness and housekeeping remained an element of strength during this assessment period, based on routine reviews by NRC inspection personnel. Good housekeeping and maintenance practices during normal operations were evident throughout the facility.

The' routine fire protection program has been implemented consistent with previous observations. Fire detection and suppression systems are well maintained and controlled. Fire equipment was'in good working condition and adequate spares were available. There were

.

no major incidents. There was only one observation of a minor inadequacy in the fire watch controls established for the drywell-in July, 1984. There have been no major changes in supervisory-personnel or training programs for the areas. Overall, management controls have remained effective in maintaining good performance in the routine programs.

The specialist team inspection conducted to review the licens-ee's actions to comply with the safe shutdown requirements of Section III of Appendix R,'10 CFR Part 50 was completed early in the assessment period. The one violation identified during this inspection is being considered for escalated enforcement action.

The violation involves failure by the licensee to reanalyze and provide adequate fire protection for the Reactor Building.

While the licensee did well in implementing the requirements of the rule for those areas outside the reactor building, the li-censee did not take the initiative to assure that his assump-tions for the reactor building were consistent with the NRC L

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26 ORIGINAL PAGE l'

-f s ff's positions. Licensee exceptions to the requirements were no roperly identified to the NRC staff.

The<1 ensee took considerable' time to respond to the issues-identi 'ed by the inspection team. The NRC positions regarding the Appe ix R requirements were clearly presented to the li-censee by he NRC. review team in August, 1983, but the licensee-did not bec e fully committed to address the identified defi-ciencies and ifferences between his and the NRC staff's posi-tions until Ma ch, 1984. ' Considerable NRC effort was required to get the lice ee to agree to perform the reanalysis and im-plement the acti s necessary to correct the_ violation. The licensee's interim ompensatory measures for the inadequate fire protection of the r ctor building were not promptly implemented-and required that the RC staff take the lead to prescribe addi-tional compensatory me ures during meetings and telephone conver-sations with the license in May, 1984.

Modifications to install r ote shutdown equipment that'was

,

electrically independent fro ~ the normal shutdown equipment were completed during 1983. The 1 ensee developed procedures to-operate the systems and declare the shutdown panels operational for.the startup from the 1984 re eling outage in accordance with the 10 CFR 50.48 schedular r uirements. However, no. dry run of the remote shutdown emergenc procedure by a shift crew was completed prior to NRC inspectio of the area in October, 1984. NRC and licensee reviews of th arocedure determined that the procedure would - 1bably have worke if needed to shut.down the plant, but only ith great difficult Significant improve-ments were required to better integrate an coordinate'the ac-tions by the shift crew. The licensee shou have performed the ,

dry run.with a shift crew prior to accepting he procedure and -,

'

declaring the shutdown systems operational.

2. Conclusion Category 2, consistent. The lower rating this asses ment period is due.to the licensee's incorrect implementation of e Appendix R rule and the licensee's slowness in responding to NR initiatives once the deficient areas were identified.-

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REVISED PAGE

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. staff's positions. ~ Licensee exceptions to the requirements were I

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not~ properly identified to the NRC staff.

The licensee todk considerable' time to respond to the inues

identified by the inspection team. The NRC positions.regarding

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-thef Appendix R requirements were clearly presented to the li-Lcensee by the'NRC review team in August, 1983, but the licensee

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did not become fully committed to address the identified defi-

iciencies and differences between his and the NRC staff's post-

,tions until April, 1984. Considerable NRC effort was required

to.get the. licensee to agree to perform the reanalysis andsim-plement the actions necessary to correct.the violation. However,

.the NRC. staff recognizes the extensive corrective actions subse-quently implemented by the licensee to correct the deficiencies-identified:in the violation. The licensee's interim ~ compensatory measures for~the inadequate fire protection ofithe reactor building

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were not promptly implemented.and required that the NRC staff take

.

the lead to prescribe additional compensatory measures during meetings and-telephone conversations with the licensee in May, 1984.

E

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- , Modifications to install remote shutdown equipment that was-electrically independent from the normal shutdown equipment were-completed during 1983.- The licensee developed procedures to-o

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Loperate'the systems and. declared the shutdown panels operational

& for the startup from the 1984 refueling outage in accordance

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w , with'the'10 CFR 50.48' schedular. requirements. However, no dry-crun of-the remote shutdown emergency procedure by a shift crew was' completed prior to NRC: inspection of the-area in October,

- 1984.' NRC and-licensee reviews of the procedure determined that

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f the. procedure would probably have worked if f needed to shut down

the plant,1 but only with great difficulty. Significant improve-ments were required to better integrate and coordinate the ac-tions by the' shift crew. .The licensee should have performed the

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. dry run with a shift crew prior to accepting the procedure and declaring the shutdown systems operational.

2. Conclusion-Category 2,-consistent. The lower rating this assessment period i- '

is'due.to the licensee's incorrect implementation'of'the Appendix-R rule and.the-licensee's slowness in responding to NRC g- initiatives once the deficient areas were identified.

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q - 3. Board. Recomendation Licensee-Finish implementation of the Appendix R analyses and modifica-

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tions in accordance with established comitments.

NRC.

t Review the licensee's actions to fully comply with the fire ~ pro-

tection rule during the next assessment period.

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?F.: Emergency: Preparedness (16%) .

11., Analysis x

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There were no' inspections in this area, and thus, there was no basis for-an evaluation during the previous assessment' period.

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Performance in the area was rated as Category 2 for the May 1,- '

1981 to: April 30, 1982 assessment period, based primarily on concerns associated with the emergency response actions taken

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during the April 24, 1982 loss of feedwater event.

W During this assessment period, there were five-(four announced-

-and one unannounced) routine inspections of'the-emergency pre-paredness program.' There were no' violations noted during the asses'sment period which related to the-licensee's state of emer-

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Lgency preparedness. One reportable event occurred on June 15,

,

1984 where the emergency organization was activated to correct a malfunction)of-a transverse incore probe (TIP), which bypassed its~

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removai stop and produced readings in the Reactor Building at least'1000 times normal. Plant operators recognized the condi-

-tion, correctly classified the event and declared an Alert emer-

.gency? -The licensee performed well in responding to the condition, '

activating.the emergency response centers, controlling the probe

.and protecting plant workers. ~A one-time exemption from conducting-

.the scheduled November 14, 1984 exercise was granted based on a

. review'and evaluation of the response to.the event including the-resident--inspector's observations of the response to the event.

The inspections during this period identified only minor. areas

, for improvement and generally indicated an adequate state of emergency preparedness in the areas examined. The close-out of

l previously identified items-indicated that the licensee has com-mitted attention and resources'to the emergency preparedness  !

program.

Two exercises were conducted during the appraisal period.- An-EOF-injexercise on August 11, 1983 tested the licensee's inter-nal emergency response capability, and an EOF-out exercise on September 21, 1983 tested the licensee's offsite capabilities

.and coordination with the offsite emergency response organiza- -

tions. Both: exercises were observed by teams of NRC and NRC-contractor personnel. Although areas for improvement were iden-tified, the NRC~ inspection teams determined that the licensee had demonstrated that they could implement their emergency plan ,

and implementing procedures in a manner which would adequately-protect'the health and safety of the public.

L L The licensee has committed an adequate amount of resources to i -

the emergency preparedness program which is comparable to the

. industry. Management attention and involvement in the emergency (,v

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preparedness program was demonstrated by evidence of prior plan- !

ning and assignment of prior _ities; adequately stated and under- !

stood policies and procedures with reviews generally timely,

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-thorough and technically sound. Understanding of technical is-sues was generally apparent with conservatism generally exhibit-ed. Timely, viable and generally sound resolutions to technical issues and NRC initiatives were provided. Staffing was adequate with positions identified, authorities and responsibilities well defined and key positions filled in a timely manner. The train-

.ing program was defined and implemented for a large portion of the staff. The training program makes a positive contribution to the emergency preparedness program. During the June 15, 1984 incident, the_ licensee's reporting of the event was prompt and complete; the event was properly identified, analyzed, and ef-fective corrective action was taken. The licensee has also in-cluded the NRC in.the planning for the renovation of its corporate headquarters'to include an E0F/ training facility in Brattleboro,

, Verment, which should be completed on schedule in 1985.

The licensee has been responsive to NRC initiatives and accept-able resolutions were proposed and implemented on a timely ba-sis. Significant improvement in the licensee's performance was noted in comparison with prior performance.

2. Conclusion Category 1, consistent.

3. Board Recommendations-NRC Reduce the priority of emergency planning inspections at Vermont Yankee.

,

%

t.1

>w'

..

30-

,

G. Security and Safeguards (2%)-

. 1. Analysis During the previous assessment' period, there were no significant .

-concerns or deficiencies identified.in the security and safe-guards area. One minor violation occurred during an outage be-

'

-

cause maintenance personnel failed to notify the security organization prior.to changing the. status of a vital area barri-

= er. . Licensee performance in the' area was rated as Category 1.

.During the current assessment period, one routine, unannounced '

_

physical ~ protection inspection was performed by a region based inspector. Routine resident inspections continued throughout '

the assessment period and one severity level IV violation was-identified. ' Corrective measures-were prompt and effective. A

' licensee identified' level IV violation occurred during an outage in September, 1984 when maintenance personnel failec to notify ithe: security organization prior to changing the status of a vi- '

.tal area barrier. Actions to correct and report the occurrence

'

were prompt and appropriate. The actions by the security

' organization in both incidents were proper. Additional licensee j management attention is warranted to assure that all plant-personnelfare sensitive to security controls.

Management effectiveness was evidenced by proper system and

,g = equipment operation and maintenance'and performance of security-personnel. -Administrative practices were well organized and records were neat, accessible.and correctly maintained. .The Site Supervisor interfaced effectively with the contract securi-

- .ty force management' staff and communications within the contrac-tor organization reflected a thorough and professiona1' understand-

.ing of the Safeguards Plans and implementing procedures for physical- 4 protection'of the site. Also evident was support for the security-

~

program by other members-of the-site management staff. ,

-

An increase in security ev'ents associated primarily wi.th securi- :

ty. system problems was:noted during-the physical protection in-spection in August 1984. It is notable that the security

-

. organization,-on its own initiative, had a statistical study of security systems' failures performed and'had presented the.re-

-

sults to' licensee management in order to effect technical im-

~

provements. Management attention to this study was evident by the scheduling of vendor. visits to the site during September 1984 and by the funds that have been made available to hire a contractor to study the present security equipment and recom-mend system' enhancements. Corporate involvement was also evi-

' dent by its support of this activity. '

,

i j ".

..

.

. -n

- -

<

, ,

,

.., y

<

- ; L311

,

~

iStaffing~of'.both the~11censee's an~d the contractor!s security forganization was consistent with program manning requirements.

,

, LSupervisorswere-knowledgeableof-theirfunctionaldutiesand responsibilities. ; Security officer training,~ qualification and performance standards were professionally developed and execut-ed .~ Facilities were well maintained and uniformed guards;re-flected good appearance. ,

'

The: licensee submitted three' revisions to the Security Plan un-

<

' der the provisions of 10 CFR 50.54(p) during the assessment pe-

. . -

riod. .The revisions were acceptable. The licensee was '

. responsive to NRC concerns and questions regarding the

. revisions.

,

2. Conclusion

'

,

Category.1, consistent.

• --

- 3.

.

. Board Recommendations Licensee

..

. Emphasize-and encourage security awareness to' station staff to

-

- - cassure previous performance. levels are maintained.

NRC

'None.

1. r I

i 4 1 L. ,

r-b

_ . . . - .-. - . . - . .- . -. - . - ~ ~ -

N

%

~

'

< .

s

- 4 y..

~

l  : H. Refueling and Outage-Management (8%)

,

,

~

'

A significant concern was~ identified in this area during the previous

-

-

(assessment period regarding the-failure of licensee personnel to fol- i

_

Llow administrative controls.and the area was rated as Category 2.

The NRC assessed a penalty.for the failure to maintain secondary

- '

-

-

. containment and then mitigated the. fine to a zero dollar amount based

'

~

< on the licensee's corrective actions. Performance in this area was

_ considered degraded due to the . lack of regard for administrative .

-

policies-and controls exhibited.by several senior employees, including two senior licensed operators, which resulted in:the secondary

' containment violation.

All phases of refueling and outage activities were reviewed by the resident and region-based inspectors during the 1983 and 1984 refuel-ing outages. . Inspections were conducted in the areas of fuel' receipt

'

inspections, refueling' activities, design changes and modifications,

. recirculation system non-destructive examinations and piping repairs,- ,

' radiological controls and the cycle 10'startup physics ' test program.

'

1The planning and control of refueling outage activities remains a-licensee management strength. Detailed planning.for outage activities is followed through with the proper level of supervision. Station

personnel are effectively used;to coordinate-contractor work activi-

'. ties during. outage modifications. Communication.and coordination

~

between outage groups was generally' effective.' One exception con-

.cerned the lack of coordination between the maintenance ~and security

= groups-during the opening of.an access to a vital area.

~

-Refueling-and spent fuel pool activities, and other in-vessel sur-

• _

lveillance and maintenance work were conducted.by qualified personnel-in accordance_with established procedures. A notable exception con-

> lcerning.the installation of the steam separator by maintenance per-sorinel is discussed in Section'C above. The' licensee maintains'

adequate control over contractors performing vessel'related work.

'

Mechanical problems with refueling. equipment were minimal. Personnel responsible forL new fuel receipt, inspection and_' installation. activi-

' ties were knowledgeable ~of the operations and conducted the activity (with ample. regard for nuclear safety.

,

For the startup physics. testing program, there was adequate planning ,

-

for the testing and the tests were conducted in accordance with

. approved procedures and accomplished by'an adequate and qualified staff. Records were complete, well maintained and readily available.

,

P >

.

l-

. . . . - - . . . . . _ . . - _ . _ . , - . _ _ . _ _ . , . - _ , _ . _ _ . . . . - . , _ ~ , - _ , _ . . . . . - - ~ . - . . - ,

-

-

.:

.c

Coordination with the offsite engineering groups is generally effec-

tive for the= timely completion of design changes. The identification and repair of recirculation pipe cracks was completed in a manner acceptable to the staff, and without extending the scheduled length of the'1984 refueling outage. Major modifications to meet the-Envi-ronmental Qualification rule for electrical equipment were also com-pleted. During the 1983 outage, the schedule.for completing the torus attached piping modifications was extended by two months because addi-tional supports were necessary to meet the load requirements. Addi-tionally, errors in the development and application of a support spacing table provided by the engineering organization resulted in the need-for compensatory measures while the supports for equipment cooling lines were redesigned and modified. Modifications to provide.the pro-

per stiffness for the hydraulic control units and to complete support base plate modifications per IE Bulletin 79-02 were still in progress at the end of the assessment period. Additional management attention is warranted to-assure quality work is received from the engineering

< organization in a_ timely manner, and to maintain a high level of per-formance in this area.

The first violation' identified in this area concerned the failure to properly review and approve a wiring change made during the installa-tion of the new scram instrument volume instrumentation during the 1983 outage. The violation occurred during the post installation testing when I&C technicians identified and corrected an apparent wiring error they thought was made when the design change was installed.

The second violation concerns the loss of secondary containment.iden-tified by the licensee while fuel movement was in progress in July, 1984. The item was categorized as a Level IV violation due to the lack of-safety significance, based on the circumstances attendant to

.the event, and based on the extensive corrective actions taken upon discovery of the condition. There were significant differences be-tween the 1983 event which significantly affected the last assessment and the 1984 event. The event in 1984 occurred primarily as the result of a single contractor individual's failure to follow the established tagging controls prior to opening a service water line in the Reactor Building to install a mechanical bypass. The contractor's actions constituted an error in judgement, rather than a disregard for the administrative controls. More importantly, the--

condition was detected during routine surveillance by Operations per-

, sonnel, who quickly assessed the situation, suspended fuel movement,

. corrected the breach in containment, and reported the incident to

'

plant management and the NRC. The actions by Operations personnel

v.

,..

.", :

during this event, and generally during other less significant inci-dents during the assessment period, demonstrated a high regard for-administrative policies and controls.

The five LERs for this area consisted of three events involving per-

sonnel error, the most significant one was discussed above, and two events involving equipment failures. The equipment. failures involve the alternate RPS power supplies (GE power protection panels). in-

-stalled by the licensee in 1983 in response to NRC concerns. Resolu-tion of the problem is-in progress.

2. -Conclusion Category 1, consistent.

3. . Board Recommendations None.

F

pqe{P ~ l I

~

rg - '

j

'

- 4 -

'

_ 35 -

m- _ 7,

.

.

i

~'

. . . . . . ..  !

-

I. - Quality Assurance (6%) i

. 1. Analysi s'.

This functional area was not addressed separately during the-previous assessment. period since no problems were noted in the. .

implementation of the. Quality Assurance (QA) program as:part of:

the licensee's system of management controls.

"

"

All: specialist. inspections address the QA/QC interfaces in the' '

e areas inspected during the current assessment period.- Two-spe--

m - '

cialist inspections specifically addressed the licensee's QA .

. Department and:its inspection, surveillance and audit overview ,

, . activities. .One of these inspections also reviewed modifica-4y '

-tions and procurement QA program controls. j m '

+

'

@t

' Several problems were identified during inspections early in the assessment period. The licensee failed to establish a program to assure that a meaningful level of independent inspections

'

-

were performed on safety-related work other than design changes '

@ ' '

- and modifications.. Additionally, QA surveillance (random moni-'-

,

-

Ltoring) was not conducted in five major QA program areas. There

-

were no specific procedures or instructions established to con-

'

duct monitoring of the areas since the surveillances~were con--

3 sidered a random, informal activity. The onsite QA group was >

<

budgeted at one and one-half QA personnel, who did audits at.

1other sites' in addition to 'other QA/QC overview of onsite activ-ities'. Corporate QA management involvement onsite was minimal.

-

' The' review for plant ~ modifications was not adequately delineat -

ed. Engineering procedures did not require a review for appro-O ' '

priate safety classification in certain design changes, QA did

.  : not have-a procedure, to detail modification package reviews,- and

'OQA procedures for modification reviews did not require a review-

,, of safety classification. Concerns in the procurerent program

'

'

. included a lack of independent review or audit of purchase or-  :

,

ders classified as nonsafety-related, and equipment reclassifi- '

k cations were not always updated in the computer-based spare parts: inventory. These problems were indicative of a QA program

_

_

, 'that was inadequately staffed and too narrow in its scope and~

coverage.

The followup specialist inspection in the QA area identified that the licensee had allocated five positions to the onsite QA

.

staff and transferred a manager to the site on a part time ba-

-

,

.

.

sis. Improvement was noted-in all' areas because of the in--

( . creased staffing and re-assignment of in plant ~ audits to the

-

'

= corporate QA organization. The licensee was responsive to NRC concerns raised during the first specialist inspection. Howev-er, the second inspection-identifled a concern in the procure-m

'

. ment area in that-the licensee failed to include appropriate items purchased between 1975-83 in the recently established ia

' f

-

-3 '

,

'f t- ~N T e ++'y +v gy w e r- 'g e- ow,st,.y,-y== vo-w rtpvyy-wiv-f- -vvvvMg-+.g.-.-y-.,g.ww..-v%-ye- y eveme =- -y m rm -w-sium.-me=,,iy.y*-m--=- w p w w- : y7ww-9pi vws v y-p ww w,9, =g-*y-egee==eg 19v eq ==r we ed - w w-w g-myygp-

,

y--

.

y . . . ,

,

. 2

. s:,,

' '

3'6

-

,

shelf-life' controliprogram. .Theseiltens included material such

,

as' replacement diaphrages,Jgaskets, seals and 0-rings. Compo-nents containing internal parts ~ of'like nature such as ASCO and

,NANCO: switches, have. not been included?in the current shelf-life

program. These later findings. indicate that while improvements have.been made; further. plant management attention to QA is

warranted.

~

NRC reviews found that training for QA personnel met the re-quirements of the. standards. However, examples were identified

. where QA personnel were assigned overview responsibility without receiving any formal training for the area being reviewed (e.g.,

surveillance and audit of radwaste shipments).

The' violations and concerns-discussed above indicated-that there E >

was. inadequate management. overview and QA involvement in some safety-related activities.

' 2. - Conclusion

,

Category 2, improving.

3. Board Recommendations

'

Licensee

,

_ Continue efforts to fully-staff th'e onsite QA group and to fully implement surveillance and inspection functions. Resolve concerns identified in the procurement program regarding control'of equipment

,

'

,

shelf life.

'

-NRC .

Review implementation of QA/QC improvements during the next 7 assessment period.

~

. Schedule a management meeting with the licensee to discuss QA program implementation.-

I

< ,

F

,

-.

.

-

=

o

..

J. Licensing Activities 1.. Analysis This evaluation represents the integrated inputs of the Operat-ing Reactor Project Manager and those technical reviewers who expended significant amounts of effort on the Vermont Yankee

. licensing actions during the current rating period. The rating also reflects the comments of the NRR Senior Executive assigned to the assessment.

The basis for the appraisal was the licensee's performance in support of licensing actions that were either completed or had a significant level .of activity during the current rating period.

These~ actions, consisting of amendment requests, exemption re-quests, responses to generic letters, TMI items, and other ac-tions, are summarized in Table 6 along with other licensing activity data.

During the present rating period, Vermont Yankee senior manage-ment personnel involvement and apparent attention to quality in issues of major safety significance, such as recirculation sys-tem IGSCC inspection and repair, was apparent. The quality of licensing submittals generally evidenced management attention.

Appropriate allocation of technical manpower in support of li-censing activities and, on occasion, reprioritization of activi-ties to meet changing safety priorities also indicates good corporate management involvement. Licensing Management involve-ment has further been apparent in candid, constructive discus-sions with NRC staff on complex licensing issues. The licensee has maintained an adequate licensing staff to assure timely re-sponses to NRC needs.

One area where management attention could be increased is in the planning of amendment requests to assure that submittals are sufficiently timely to realistically accommodate NRC processing and Federal Register noticing time requirements.

The licensee's management and its staff have demonstrated clear technical understanding of issues involving licensing actions.

Its approach to resolution of technical issues has demonstrated extensive technical expertise in all technical areas involving

.

licensing actions. The decisions related to licensing issues I

have exhibited conservatism in relation to significant safety matters such as early commitment to a scheduled replacement of recirculation system piping.

The licensee frequently forms technical judgements independent of the industry, and these judgements are usually well thought out and well supported. An example of the licensee's indepen-dent technical capability is development of its own fuel

,

^

.o:

.

performance code (FROSSTEY). In this effort, the licensee dem-onstrated a level of technical competence and self-sufficiency

- that is difficult to find elsewhere in the industry outside of

'the major fuel. vendors.

An example of the licensee's industry-wide technical awareness and initiative was its recent proposal to remove the feature that automatically transfers high pressure coolant injection-(HPCI) suction from the condensate storage tank (CST) to the torus on high torus water level. This proposal was based on the licensee's review of the Browns Ferry report on Station Blackout which revealed that elevated suppression pool water temperatures during blackout can damage the HPCI pumps which use the water pumped from the pool for cooling the lubricating oil of the HPCI pump turbines.

The licensee has been generally responsive to NRC initiatives.

During the rating period, it generally made reasonable efforts-to meet or exceed commitments. Responsiveness by the licensee facilitated closing out several complex and historically tangled multi plant issues, such as Feedwater Nozzle Cracking and Appendix J. Responsiveness in accommodating independent staff consultants in the 1984 piping inspection outage resulted in development of a more sound technical basis for understanding the IGSCC phenomena at Vermont Yankee than would otherwise have been developed. Licensing actions associated with the Environmental Qualification rule for electrical equipment were completed during the period. The number of items requiring a schedular extension were few and adequate justification for continued operation was provided to the staff in a timely manner.

An isolated instance of licensee nonresponsiveness occurred in the licensee's provision of post-accident sampling information which required about two years to obtain, despite two letters, a number of telephone calls, and an NRC site visit.

2. Conclusion-

' Category 1, consistent.

3. Board Recommendations None.

7 ,

.

+ .

,

..

m _

< ,

'

IV.. ' SUPPORTING DATA AND SlM4 ARIES i AL Investigations, Petitions and Allegations

~

'

In June,.1984, the NRC received an allegation.from a contractor en-ployee that adverse actions were taken against'him and several other.

contract workers based on unsatisfactory medical screening results,

but termination of employment as a radiation worker was _ delayed until

~

a high exposure job was' finished. This matter was reviewed by the-resident and regi on based inspectors.- These inspections determined

-

Lthat there were no violations of radiation protection requirements and there was no attempt by the licensee to knowingly expose workers

to radiation who were not-medically qualified for radiation work.

The NRC received a petition under 10 CFR 2.206 on October 25, 1983 from the Vermont Yankee Decommissioning Alliance and the Vermont Pub-

-lic Interest Research Group requesting that the NRC issue'an~ order toi the licensee to show cause why the license'should not be suspended pending resolution of certain pipe crack related issues. The direc-tor of NRR issued a-decision dated April 16, 1984, after considering the information filed by the petitioners, and concluded that the re-quest.was not warranted. The request was denied.

-

-

.The' Office of Investigations conducted no investigations related'to '

_

Vermont Yankee during the-current assessment period.

' '

-

B. Escalated Enforcement Actions

, 1.. Civil Penalties None

.

2.. Actions Pending/ Resolved 4

+ -IR 83-26 violation.for failure to provide protection per

>

. Appendix. II, III.G.2 for. equipment in the Reactor Building.

+- IR'83-30 Level III violation for transport package dose t rates in excess of limits. No penalty issued since burial site suspended 1icense.

'3. Orders a. Extended Mark I completion Date, June 17, 1983

'

-b! Confirmed Pipe Crack Leak Detection, June 27,'1983 c. Revised' Simulator Examination Requirements, December.12, 1983 d .~ Confirmed Emergency Response Capability, June 12,1984 e. Confirmed Commitments on Pipe Crack Issues, August 28, 1984

-

ha

W ,,9 , . . -y .-

,,.- .

7; : ../

, ...

l -, ,

-40.

L

'

4. Confirmatory Action Letters (CAL)

.a. ' CAL 83-04, Confirmed' Actions on Service Water Pipe-Supports, June 117, 1983

_

~

'[ '

^ b. CAL 83-10, Confirmed Actions on Plant Frisking Policy,

. August 10, 1983:

C. ' Management Conferences-

1. ' a. SALP Management Meeting at the Vermont Yankee Site, June 28,

"

1983 b.- Management Meeting to Review Transportation Violations,

' '

November 22,' 1983 c. Management Meeting to Review Appendix R Actions, November 22,

>

. 1983

, < d. Management Meeting to Review Appendix-R Actions, January 10,

'

1984-

'

e. .

Management Meeting to Review Appendix R Actions, May 24, 1984'

o

, f. Management Meeting Requested by Licensee to Review Violations in 1 Inspection Reports 84-02 and 84-06, September 24, 1984

- D. Licensee Event-Reports y

Type.of Events:

A. Personnel Error . . . . . . .-. . . . . . . . . . . . 14

% 3 B. Design / Mfg /Const/ Install Error .. . . . . . . . . . . 3 C.- External.Cause' . . . . ..; . . . . . . . . . . . . . O D. Defective Procedure . . . . . . . . ... . . . . . . . 4

! E. . Component Failure . . . . . . . . . . . . . . . . . . 20

' ~

X. Other-.:. . . . . . .-. , . . . . . . . . . . . . . . 2'

"' Total- 43 Licensee Event Reports Reviewed:

-Reports 83-14'to 83-34 and 84-01 to 84-22 i- .

. .).

!

( :.

a p ..

[ .-

(_ .

.

. ;;

,; ,

4 .

'

Causal Analysis-Six sets of causally linked events were. identified.

~

a. LERs 83-14,183-15, 83-19, 83-24, 83-29, 84-03, 84-05, 84-06, 84-08, 84-09, 84-10, 84-12, 84-15, and 84-17-are events-(14 to-

,

.tal) due to personnel error. -Reportable events. involving errors were about equally distributed _in the plant operations, radio -

logical controls, surveillance and refueling functional areas.

b. 'LERs~84-16 and 84-18 occurred in part due to the alternate RPS

-

power supply.either-tripping or being in a tripped condition.

c. -LERs'83-16, 83-30 and 84-04-involved mechanical failures of main steam isolation valves.

'

d. LERs 83-17,.83-20, 84-20 and 84-22 involve events in which relay-

'

failures. resulted in' operation in a degraded mode. -LER 84-22 concerns the near simultaneous failure of both diesel generators

.due to failed zener diodes in redundant generator differential-relays.

e. -LERs 83-21, 83-31, 83-33, 83-34, 84-02 and 84-13 are events that occurred due to setpoint drift on instruments of relief valves, f. LLERs 83-25, 83-26, 841 13 and 84-21 are events that' occurred as a

-

L result of inadequate procedures.-

,

i

,

!.

! ..

p

,

i tv

q _

= 6! +

~:JU..

'

42'

'

-

, TABLE 1

. ,

'

, INSPECTION HOURS SUMMARY (5/1/83 - 10/31/84)

.

VERMONT YANKEE NUCLEAR POWER STATION

_

HOURS % OF TIME

.

iA. ' Plant Operations . . . ... . . . . . . . . . . 1087 28.

'

i B. - Radiological Controls :. . . ... . . . . . . . 912 23 C; Maintenance . . . ... . . . ... . . . . . . . 111 3

D. JSurveillance .z.. ................ '191 5

,

'E.. Fire Protection and Housekeeping . . . . . . . .

339 9 F. Emergency Preparedness .. _........... .

654 16

.

G. Securi.ty and' Safeguards ..... ..... 65 2

~ H. Refueling and Outage Management . . . . . .. . 305 8 I'. : Quality : Assurance '

............... 239 6

'J . - Licensing'ActivitiesL. . . . .-. .=. . . . . .

.

• *

fo Total 3903 100%.

.

• Hours expended in facility licensing activities and operator license '

. activities are not included with direct inspection hour statistics.

t T

l' '

..

o I s

t

] i

• -

,

J

TABLE 2 VIOLATIONS (5/1/83 - 10/31/84)

VERMONT YANKEE NUCLEAR POWER STATION A. Number and Severity Level of Violations Severity Level I O Severity Level II 0 Severity Level III 1 Severity Level IV 16 Severity Level V 8-Deviation 0 Under Review 1 Total 26 B. Violation Vs. Functional Area Severity Levels FUNCTIONAL AREAS I II III IV V DEV A. Plant Operations 6 1 B. Radiological Controls 1 5 3 C. Maintenance 3 D. Surveillance 2 E. Fire Protection and Housekeeping * 1- TBD F. Emergency Preparedness G. Security and Safeguards 1 H. Refueling and Outage Management 1 1 I. Quality Assurance 1 J. Licensing Activities Totals 1 16 8

• 1 violation - category to be determined

{' c b

e (TABLE 2 Continued)

C. ~ Summary

..

Inspection ~ Inspection- Severity Functional Report No. Date Level Area Violation 83-13' 5/2-5/83- V H Failure to review a field change to EDCR per'. approved procedures

~

83-22 7/11-20/83 IV' C Failure 'to provide indeperi*

. dent QA and supervisory inspection of maintenance IV C Failure to completely doc e ument maintenance requests IV I Failure to conduct 0QA sur-veillance of all areas 83-26 _8/29-9/2/83 *TBD E Failure to provide fire pro-tection per Appendix R II.G'

in Reactor Building 83-27 9/6-10/3/83 IV A Failure to maintain main steam line trip setpoints less than 3X background IV A Failure to maintain proper SLC system valve lineup V D . Failure to adequately re -

view seismic instrument test results-83-29 10/4-31/83 V G . Failure to control. access-to an access controlled area 83-30 8/23/83 III B Nevada Burial Site received package with 250 mrem /hr~

surface dose rate IV B Failure to provide strong tight package for blade guide shipment 83-33 12/6-9/83 V B Failure to use approved pro-cedure for temporary whole body counting system

_

~

..

6'

45 i

.

qs (TABLE 2 Continued) /

'84-02. 1/24-27/84 V- B Failure to train QA per- '

% sonnel in transportation j ' regulations

.

84-OS 2/28-4/2/84 IV ~ A -Procedure for-valve controls inconsistent with technical specification requirements 84-07 3/26-30/84 V- B Failure to review SBGTS test data 84-08 ~4 /3-5/7/84 IV A- .HPCI actuation channels partially inoperable IV A Failure to follow procedure to reset ECCS logic ,

V A Failure to maintain 125 VDC breakers positioned per procedure 84-10

'

5/8-6/4/84 V D Failure of I&C personnel to properly secure from testing.

84-11 5/21-25/84 IV B Failure to meet March 1983 Order on NUREG'0737-Items II.F.1.2 and-II.F.1.3 84-18 7/31-9/18/84 IV A Failure to' maintain core spray and RHR service water valve lineup per procedure

.

84-20 7/17-31/84 IV H Failure to maintain secon-dary containment during-refueling 84-21 9/19-10/31/84 IV C Failure to bolt separator.

to core shroud IV B Failure to survey material '

prior to release from the restricted area

'

IV B Failure to post and label radioactive material.in the unrestricted area

'

'

L

y .- . . . - . _ . .

-

. .

. . . .

'

y ,

li' . :,

, :.;.

a

• ' (TABLE 2 Continued)

IV Failure to secure radio-

~

B

,

' active material in.'the un .

restricted area from unauth-orized removal

_

'

,

'

• TBDL- To-Be Determined: the appropriate escalated enforcement action for this item is.under' review by NRC management.

.

i. .l t-E,'

..

..

b

'

. "

L-

,

,

t

.

,s- >

b.~ ,

t

,_. - ~ _ . - . , - . , . . _ _ . - . . . - , , . _. , , , , . . - .- _

r. . ,, . ,--- .

. _ -_ - - . _.

.

,

'

TABLE 3

' INSPECTION REPORT ACTIVITIES (5/1/84 - 10/31/84)

VERMONT YANKEE NUCLEAR POWER STATION

!

Inspection Report No. '/nspectionHours I Areas Inspected 83-11 30 . Radiological ControlsL- Outage 83-12 15 Emergency Preparedness - PNS 83-l'3 26 Outage Activities and Modifications 83-14 97 Routine, Resident Refueling-L83-15 13 ISI - Recirculation Weld Overlay NDE

'

83-16' 56 Environmental Monitoring 7 83-17 127 Routine, Resident 83-18 48 Containment Leak Rate Test'

_ 83-19 37 Fire Protection Appendix R 83-20 47 Radiological Controls 83-21 124 Routine, Resident-83-22 144 -Quality Assurance Salem Issues

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83-23 256 EOF-IN Emergency Drill 83-24~ 61 Radiological Controls 83-25 96 Emergency Pre-paredness

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E

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- (TABLE 3 Continued) l Inspection Report No. Inspection Hours ~ Areas Inspected I s t I t 83-26 257 Fire Protection Appendix R-83-27 126 Routine, Resident 83-286 200 EOF-0UT Emergency-Drill 83-29 '87 Routine, Resident

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83-30 16 Transportation-

' Activities

83-31 100- Routine, Resident

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83-32 37 . Licensed and Non-licensed Training

'

Programs

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.83-33 58 Radiological Controls

.84-01 210 Routine, Resident

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'

84-02- '44 Transportation Activities-

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84-03; 35- Snubbers and Pro-posed Technical-Specifications

=84-04- 9 'Special-Review Incident Radio-logical Controls 84-05 132 Routine, Resident

84-06 26 Radiological Controls

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'84-07- 102 Radiological Con-trols-Independent Measurements

'84-08 103 Routine, Resident L84-09' '28 Refueling Outage Startup Testing

_ 1 84-10 61 Routine, Resident

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(TABLE 3 Continued)-

Inspection Report No. Inspection Hours Areas Inspected 84-11 169 Special Review of

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NUREG 0737 Items 84-12 229 Routine, Resident 84-13 37 Inservice Inspec-

'

tion Recirculation Weld Repair 84-14 ---

Number not;used.

84-15 --

License Examina-tions 84-16 50 Emergency-Pre-paredness j- 84-17 48- Radiological Controls 84-18 195 Routine, Resident 84-19 -27 Security and Safeguards.

84-20 43 Special Review Loss of Secondary Containment In-cident 84-21~ 202 Routine, Resident 84-23 95 Quality Assurance Maintenance i.,

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TABLE 4

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TABULAR LISTING 0F LERS BY FUNCTIONAL AREA VERMONT YANKEE NUCLEAR POWER STATION Area- Number /Cause Code Total A. Plant Operations 3A =1B 20 SE 1X 12 B. Radiological Controls 4A 2E 1X 7

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C. Maintenance :U4 1D 2 D. Surveillance 3A 10 13E 17

E. Fire Protection and Housekeeping None 0 IF. Emergency Preparedness None 0 G. Security and Safeguards None 0 H. -Refueling and Outage Management 3A 2B 5 I. Quality Assurance None 0 J. Licensing Activities None 0 TOTAL 43

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.Cause Codes: A. Personnel Error B. . Design / Mfg /Const/ Install Error C. External Cause D. Defective Procedures E. Component Failure X. Other

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TABLE 5 LER SYNOPSIS (5/1/83 - 10/31/84)

-VERMONT YANKEE NUCLEAR POWER STATION LER Number Type Summary Description 83-14 30 day Surveillance test for service water system radiation monitor missed due to scheduling error 83-15 14 day Supports for service water copper tubing and torus attached piping inadequate due to engineering errors 83-16 30 day MSIV 86A closed in 5.5 seconds during testing due to-broken spring in hydraulic dashpot assembly 83-17- 30 day- Time delay relay for LPCI injection throttle permissive failed to actuate during testing .

83-18 30 day -HPCI tripped during testing due to vibration induced loosening of screws in overspeed trip unit 83-19 30 day- Workers inadvertently opened breaker for RHR torus spray valve RHR 38A 83-20 14 day During surveillance testing, B diesel generator tripped due to faulty shutdown relay, and B core spray suction valve failed closed on thermal overload 83-21 30 day Drywell pressure transmitter PT 101D actuated at 2.57 psig during testing due to setpoint drift 83-22 '30 day RWCU suction valve CU 18 failed as a throttle valve during testing due to faulty closing torque switch 83-23 30 day Routine environmental monitoring' program failed to detect Co-60 buildup in the Connecticut River at dis-charge of the site North storm sewer due to a source originating from-the turbine building roof vents 83-24 30 day Shift supervisor removed AEOG radiation monitor from service without following administrative controls g

-

83-25 14 day During routine operations, all four main steam line radiation monitors had trip setpoints greater than 3 times background due to an inadequate procedure 83-26 30 day Leakage test valve SLC-36 found open during routine operations due to inadequate procedure used during outage for recirculation loop decontamination

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(TABLE 5 Continued)

.LER Number Type Summary Description

83-27 14 day During' surveillance testing, A diesel generator failed due to faulty air start. check valve, and B core spray suction valve failed closed on thermal overload

~, ..

~83-28 L30 day ^RHR pump 10 was made inoperable by removing all

'

control fuses while investigating smoke caused by.

dirt and oil on motor heater 83-29: 30 day -Stack gas quarterly strontium analysis incomplete due to administrative errors in processing samples

'83-30. 30 day- During surveillance testing, MSIV 86A closed in 2.6 seconds due to'a missing plug on speed controller

,

83-31 30. day Safety. relief valve lifted at 1122 psig versus 1100

.psig during testing due to setpoint drift 83-32 14 day HPCI inoperable.due to lost position indication on pump discharge valve,-RCIC inoperable due to failed motor windings on pump discharge valve

-..

'

83-33 30 day Main steam 'line radiation monitor 251C trip setpoint-found greater than 3 times background during testing due to zero bias' voltage shift 83-34 30" day' HPCI suction transfer on low CST level found out of'

specification during testing due to setpoint' drift

.

of level-instrument LT 107-5B

~

~84-01 30 day- Reactor trip from 100% FP'on January 5,1984 due to

, high' pressure' caused by-turbine-control system EPR

' oscillations 84-02 -30 day Both, reactor building-torus vacuum breakers found -

'

inoperable-during. testing due to setpoint drift of

. delta-P switches to 0.55 and 0.51 psid

~.

a ,

. 84-03 30 day 21 samples from the weekly surveillance of the

' #E environmental stations were mistakenly-discarded as rubbish -

- '84-04 -_30 day' Reactor scram from 100% FP on April 16, 1984 due to. '

MSIV isolation caused by MSIV 80C failing closed

'

during testing due to-a faulty air pilot valve assembly l

74 -

<

i:.._-

.

.

(TABLE 5 Continued)

LER Number TyJLe -Summary Description 84-05' 30 day HPCI inoperable during 5 days of power operations due to blocked high drywell pressure initiation logic, which was not reset by operator following April 16, 1984 scram 84-06 30 day Both post accident torus level instruments found inoperable following return to service after routine calibration 84-07 30 day Alert declared on June 15, 1984 due to excessive high-radiation levels in the reactor building due to an unshielded TIP detector 84-08 30 day Weekly data-at one environmental air station lost because technician failed to restart sample pump following collection of sample cartridges-84-09 30 d v While performing control rod friction testing with L: reactor shutdown on June 17, 1984, a reactor-scram occurred due to.a valving error committed during a level instrument calibration 84-10 30 day- Both stack monitors inoperable d'uring power opera--

tions: one monitor had instrument drift by 1/2 decade; a gamma sensitive detector was mistakenly installed for the second monitor during maintenance

'84-11 30 day Type C leak rate testing identified components with

,

leakages in excess of technical specification limits 84-12 .30 day Secondary containment violated during fuel handling due to a contractor's failure to follow procedural controls during installation of a mechanical bypass 84-13- 30 day. Redundant SLC relief valves had lift setpoints found less than technical specification _ limits during

' testing; system function not compromised

.84-14 30 day Weekly air sample data lost at environmental station due to blown fuse on sample pump, possibly caused by electrical storm 84-15. 30 day Inadvertent scram signal generated on July 24, 1984 while shutdown for refueling when operator inadver-tently tripped RPS power during "second" verification checks a _

. -.

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s

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, --

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L

~-  ;

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54'

'

.

t g -(TABLE 5' Continued) 1 LER' Number _ Type ' Summary Description

'

84-16 '_ 130 day Inadvertent scram signal generated on July 28, 1984 with-plant in refueling shutdown due to spurious loss. ,

of alternate RPS power during bus switching operations

84-17' , 30 day Service water system sampling not performed for two

'

days per technical-specifications while' radiation

-

,

.

-

. monitor out of service due to improper shift turnover 84-18 30' day Inadvertent scram signal generated on August 1, 1984 with plant in refueling shutdown _due to~ spurious loss

>

of alternate RPS power during bus switching operations.

< '

84-19 30. day Rosemont 1152~1evel transmitters potentially inop--

.erable'due to loose circuit board mounting screws '

supplied by vendor 84-20! 30 day RCIC inoperable when failed annunciator relay.in control circuit caused loss of power to inboard steam supply valve L84-21. 30 day Reactor power to flow anomaly.due to: steam separator being not tightly. secured to core' shroud, caused by,

-inadequate procedure and personnel training

' '

l84-22 30 day. . Lockouts tripped on'both diesel generators-due to

-

y failed zener diodes in generator differential relays;

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)

f k

\ l s

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r

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c

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. k'

TABLE 6 SUMMARY OF LICENSING ACTIVITIES P

VERMONT YANKEE NUCLEAR POWER STATION 1. :NRR/ Licensee Meetings-

-

May 26, 1983 Mark _I'(Framingham,.MA)

- September 20, 1983 Licensing Schedules'(Framingham, MA)

' October 21, 1983, Pipe Cracks.(Bethesda, MD)

JJanuary 9,1984 Moisture Sensitive Tape;(Bethesda, MD)

April 18, 1984 Environmental Qualification (Bethesda, MD).

July 26, 1984 Pipe Cracks (Bethesda, MD)

2. NRR Site Visits-June 29, 1983 Post Accident Sampling September 22, 1983- : Emergency Facilities

.

,

- 3. Schedular Extensions Granted August 12, 1984. -Environmental-Qualification June 27, 1984~ NUREG 0737 Supplement 1

.

. ; - 4. Exemptions Granted h August 19,"1983. Testing of Certain~ Type C Valves in Accordance with Appendix J-

, August.21, 1984- _FSAR Update Submittal per 10 CFR 50.71(e) from L July 20, 1984 to November 30,:1984

!

'5. . License Amendments' Issued'

, , Amendment No.~79, Shift' Technical Advisor, May 2, 1983 Amendment No. 80,11nual Emergency Exercise, November _10, 1983 Amendment No. 81,-Reactor Vessel P-T' Curves, March 31,.1984-Amendment-No.-'.82,:SDV Air Dump System, August 1, 1984

.' Amendment No. 83, RETS, 0ctober 9, 1984 L'

- 6. Summary of Activities.-

~

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a. '16 Multi-Plant Actions-(8 completed):

RPS Power Supply (C-11) . completed .

High Energy Line Break (D-15).- completed

_

h la

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(TABLE 6 Continued)

Long Term Purge and Vent (B-24) - completed Feedwater Nozzle Cracking (B-25) - completed Implementation'of NUREG 0313 (B-05) - completed Control of Heavy Loads Phase I (C-10) - completed Appendix J (A-04) - completed RETS (A-02) - completed b. 39-Plant-Specific Actions (31 completed, including):

' Review of 1983 Pipe Inspection and Repair - completed Review of 1984 Pipe Inspection and Repair - completed

. Approval of EOF Location - completed Review and Denial of 2.206 Petition - completed Fire Protection Modifications - completed.

Reactor Vessel Pressure-Temperature Curves - completed FSAR Schedule Exemption - completed Modification of 0rder-for Simulator Examinations'- completed Changes in Emergency Drill Schedules Review of Fuel Analysis Code c. 14_TMI (0737) Actions (9 completed)

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ENCLOSURE 3 -

g g,, UNITED STATES

• l' $ NUCLEAR REGULATORY COMMisslON R EGION I

-$ 'g 431 PARK AVENUE

• f KING OF PRUSSIA. PENN5Yl.VANIA 19404

\ *

/

Do'cket No. 50-271- DEC 211984 Vermont Yankee Nuclear Power Corporation

, ATTN: Mr. W. 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.

50-271/84-25 The NRC Region I SALP Board has reviewed and evaluated the_ performance of activi-ties at the Vermont Yankee Nuclear Power Station for the period of May 1,1983 through October 31, 1984. The results of this assessment are documented in the enclosed SALP Board Report dated December 11, 1984. A meeting to discuss the

- assessment will be scheduled at a later-date.

'At the SALP meeting, you should be prepared to discuss our assessment and your plans to improve performance. .The meeting is intended to be a dialogue wherein any comments you may have regarding our report may be discussed. Additionally, you may provide written comments within 20 days after the meeting.

Your cooperation is appreciated.

Sincerely,

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m 3 r

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Richard W. Starostecki, SALP Board Chairman Director, Division of Project

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and Resident Programs Enclosure: SALP Report No. 50-271/84-25

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Vermont-Yankee Nucicar Power 2 DEC 211984

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Corporation -

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cc w/ enc 1:

R. W. Capstick, Licensing Engineer -

W.~ F. Conway, President and Chief Executive Officer J. P. Pelletier, Plant Manager Donald Hunter,_Vice President Cort Richardson, Vermont Public Interest Research., Group, Inc.

Public -Document Room (PO't)

Local Public Document Roon. (LPDR)

Nuclear Safety Information Center (NSIC)

NRC Resident Inspector State of New. Hampshire -

State of Vermont

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/* ENCLOSURE 4 #

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' VERMONT YANKEE NUCLEAR POWER CORPORATION

.

RD 5, Box 169, Ferry Road, Brattleboro, VT 05301 ,,,L, 7, p ENGINEERING OFFICE 1671 WORCESTER ROAD

• FRAMINGHAM, MASSACHUSETTS 01701

• TELEPHONE 6174s*2 4100 March 1, 1985 U.S. Nuclear Regulatory Commission

.. Office of Inspection and Enforcement L Region I 631 Park Avenue King of Prussia, PA 19406

'

Attention: Mr. Richard W. Starostecki SALP Board Chairman Division of Project and Resident Programs References: a) License No. DPR-28 (Docket No. 50-271)

5 b) Letter, USNRC to VYNPC, SALP Report No. 50-271/84-25, dated 12/21/84 c) Letter, USNRC to VYNPC, I&E Inspection Report No. 50-271/84-21, dated 12/6/84 d) Letter, VYNPC to USNRC, FVY 85-02, dated 1/14/85 e) letter, USNRC to VYNPC, I&E Inspection Report No. 50-271/83-26, Enforcement Conference Findings, dated 3/13/84 f) Letter, USNRC to VYNPC, I&E Inspection Report No. 50-271/83-26, dated 11/2/83 g) Letter, VYNPC to USNRC, FVY 84-24, dated 3/14/84 h) Letter, VYNPC to USNRC, FVY 84-53, dated 5/21/84

,

Dear Sir:

Subject: Systematic Assessment of Licensee Performance (SALP) Report Comments The purpose of this letter is to provide you with comments regarding the

'

most recent Systematic Assessment of Licensee Performance (SALP) Report which was issued by letter dated. December 24, 1984 [ Reference b)]. We appreciated the opportunity to discuss the findings of the report at the January 24, 1985 meeting in King of Prussia. In general, we believe that the report is a fair appraisal of our activities during the May 1983 through October 1984 reporting period; however, as discussed at our meeting, there are certain areas within the report that warrant clarification and/or correction.

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NCM p.

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! 'U.S. Nuclear Rsgulatory Commission

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March 1, 1985

Page 2 VElulONT YANKEE NUCLEAR POWER CORPORATION

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Please note that your December 24, 1984 letter requested that any comments on the report be formally submitted within twenty (20) days of the January 24, 1985 meeting. However, due to the complex nature of some of the areas discussed in the report, we required an additional two (2) weeks to prepare our comments.

The need for additional time was discussed with and agreed to by Mr. William Raymond of your staff.

We respectfully submit the following comments for your consideration:

(1)Section IV.A.1, Operations (Page H The SALP Report states that "the management decisions and actions to con-tinue plant operation from September 16-18, 1984 with an anomalous core power-to-flow relationship and in spite of clear indications that the plant was operating ~in an unanalyzed condition, appeared as a significant

. deviation from the normally conservative approach taken to assure safe plant operations. NRC considered that the licensee had an insufficient

,

basis to continue operation with the anomaly, and the licensee's decision was neither prudent nor conservative."

We strongly disagree with the NRC's assertion that our decision to continue operation was neither prudent nor conservative. As discussed in our January 14, 1985 [ Reference c)] response to I&E Insnection Report 84-21

[ Reference d)], we believe that we effectively and prudently analyzed the situation during the entire course of this occurrence (from indication that an anomaly existed through rejection of hypothetical causes to iden-tification of a tentatively identified cause). While we evaluated possible causes of the anomaly, we simultaneously assured to our satisfaction that continued operation was prudent and posed no safety concern.

The details of our engineering assessment of the anomaly is discussed in detail in our January 14, 1985 submittal. We believe the actions taken were consistent with our conservative philosophy of assuring safe plant operations. Thus, we believe that your stated conclusion that Operations declined over the Report period due to a "non-conservative operational philosophy" is unfounded.

(2)Section IV.E.1', Fire Protection and Housekeeping (Pages 25,and 26)

o The SALP Report states that as a result of the August / September 1983 inspection of Vermont Yankee's compliance to Section III of Appendix R to 10CFR50, Fire Protection Requirements, "one violation is being con-sidered for escalated enforcement action."

A-

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+ l ' U.S. Nucicar Regulatory Commission

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March 1, 1985

~Page 3 VEllMONT YANKEE NUCLEAll POWEft COllPOllATION As described in various submittals made to the NRC subsequent to the 1983 Appendix R inspection, Vermont Yankee has performed extensive engineering analyses to demonstrate the adequacy of our fire protec-tion capabilities and is continuing with the installation of addi-tional modifications to enhance our overall fire protection design.

We believe our response to the findings of the 1983 inspection have been timely, responsive to the NRC's concerns, and reflect a strong commitment on the part of Vermont Yankee and its management to achieve compliance with the provisions of Appendix R. We recognize the time it has taken to address all of the NRC's concerns; however, the con-tinual issuance of guidance criteria by NRR regarding Appendix R compliance (i.e., Generic Letter 83-33, Generic Letter 85-01, various Information Notices, and information resulting from Regional Workshops) has made the task of achieving literal compliance that much more difficult. Enclosure 1 provides a chronological history of events associated with the Appendix R issue at Vermont Yankee. These

.- events include formal submittals made by us _as well as numerous meetings held with various NRC Staff members to resolve this issue.

Based on these efforts, we believe that the pending enforcement action should be formally dispositioned with a finding that escalated enfor-cement action is not warranted.

o' The SALP Report also states that, "the licensee did not take the ini-tiative to assure that his assumptions for the Reactor Building were consistent with the NRC staff's positions. Licensee exceptions to the requirements were not properly identified to the NRC staff."

This statement reflects the basic misunderstanding between the NRC and Vermont Yankee with respect to compliance with Section III.G to

Appendix R. As discussed at the January 10, 1984 Enforcement Conference held in King of Prussia, we had received various correspon-dence from the NRC which indicated. to us that to comply with Section

~III.G, we need only provide alternate safe shutdown capability for the Control Room, Cable- Vault and Switchgear Room. This correspondence, discussed in detail in the March 13, 1984 Enforcement Conference Meeting Minutes [see Reference e)], was the basis for our conclusion that our overall Fire Protection Program satisfied the intent of Section III.G with respect to our Reactor Building and that the NRC was cognizant of our assumptions and positions which were documented in our 1978 Fire Hazards Analysis Report.

o The SALP Report states that "The licensee took considerable time to respond to the issues identified by the inspection team. The NRC o

positions regarding the Appendix R requirements were clearly presented

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V' [ I V.S. Nuclear Rtgulatory Commission.

' March 1, 1985

Page .4 VEllMONT YANKEE NCCLEAll POWEll Colt!'OllATION to th'e licensee by the NRC review team in August 1983, but the licen-see did not become fully committed to address the identified deficien-cies and differences between his and the NRC staff's position until March 1984 Considerable NRC effort was required to get the licensee to perform the reanalysis and implement the actions necessary to correct the violation."

In addition, the Report states that "the lower rating this assessment period is due to the licensee's incorrect implementation of the Appendix R rule and the licensee's slowness in responding to the NRC initiatives once deficient areas were identified."

As discussed above, we believe our response to the findings were timely and appropriate. Immediately following the 1983 inspection, we initiated a re-review of the requirements of Appendix R to 10CFR Part 50' and initiated a re-survey of our Reactor Building to ensure we met

'the separation criteria of Section III.G.2. It should be noted that our re-survey was performed assuming a fire induced loss of off-site -

~ power which was consistent with our original interpretation of the loss of off-site power provisions of Appendix R. We also initiated a breaker coordination study and an engineering analysis of plant safe shutdown systems against the separation and fire protection criteria E of Section III.G.2 of Appendix R.

Once the formal findings of the inspection were issued in the

~

~

November 2, 1983 Inspection Report [see Reference _f)], we performed an analysis of the safety significance of the inspection's findings. The results of our analysis.were presented to the NRC at a November 22,

'1983 meeting at King of Prussia where it was concluded that there were-no findings that would preclude our ability to safely shut down the plant in the event of a fire. At that meeting, we also presented a

! _ draft report entitled " Analysis to Demonstrate Safe Shutdown

• Capability During and After Fires," which documented our analysis.

Following the November 22, 1983 meeting, we continued with our engi-neering efforts to address the deficiencies cited in the November 2,-

1983 Inspection Report. We reported the scope and status of our efforts to the NRC at an Enforcement Conference held in King of Prussia on January 10, 1984. We also discussed differences between Vermont Yankee and the NRC in interpreting certain provisions of Appendix R. Specifically, the NRC stated that Appendix R implies the need -for safe shutdown capability for the plant, assuming a loss of off-site power for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> concurrent with a fire in any area of the c __  ;

L

• i ' U.S. Nuclear Rigulatory Commission

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March 1, 1985 Page 5 VEllMONT YANKEE NUCLEAlt POWEft COItPOllATION L . plant. Vermont Yankee's assumption, both in the pre-Appendix R Fire Hazards Survey and the post-inspection re-survey was that a non-mechanistic loss of off-site power need not be considered. The NRC also stated that Vermont Yankee's use of " fire zone boundaries" in the Reactor Building did not meet Appendix R requirements for physical

" fire area boundaries." Vermont Yankee's position was that the inherent spatial separation between areas of the Reactor Building, coupled with physical barriers and extensive fire protection features installed in the Reactor Building met the intent of the " fire area boundaries."

At the January 10 Confere'nce, Vermont Yankee committed to submit the results of the Section III.G 2 re-survey (including an associated cir-

~

cuits study), docket formal requests for exemption from the specific separation criteria of Section III.G.2, and submit the details of fire protection system enhancements and/or corrective actions deemed necessary as a result of the re-survey. At the time, we still

,

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believed our loss of off-site power and fire area boundary positions were justifiable.

On March 14, 1984 [see Reference g)], we submitted requests for exemp-tion for certain areas of the Reactor Building that did not meet the

-

specific separation criteria of Section III.G.2. The exemption requests included the technical basis for justification as well as

'

. descriptions of specific fire protection system enhancements /modifi-

- cations which we deemed were necessary as a result of our III.G.2 re-survey.

-In April 1984, our design engineers attended an NRC sponsored Regional Workshop held to discuss and clarify NRC positions regarding compliance with the provisions of Appendix R. Following the Regional Workshop, a meeting was held between Vermont Yankee and Yankee Atomic Electric Company engineers and management to discuss the results of the workshop and the status of our Appendix R compliance efforts.

c Based on an assessment from engineers who attended the Regional

^ Workshop that the NRC was adament on the need to assume a random loss of.off-site power coincident with a fire, Vermont Yankee management directed the engineering staff to expand the scope of the III.G.2 re-curvey (including the associated circuits study) to include a random loss of off-site power. In addition, based on the results of the workshop and discussions with NRR fire protection engineers, it was recognized that additional detailed engineering would be necessary to support Vermont Yankee's position that " fire zone boundaries" in the

, Reactor Building were an acceptable alternative to the NRC's " fire area boundaries." It was agreed that the report documenting the ex-L. ~ -panded III.G.2 re-survey would include detailed justification of the acceptability of " fire zone boundaries."

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. ( U.S. Nucicar Regulatory Commission

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March 1, 1985

- Page 6 VEltMONT YANKEE NUCLEAR POWER CORPOllATION Vermont Yankee then requested that a meeting be held with the NRC (NRR and I&E) to discuss the status of our engineering efforts and establish a means to close out Appendix R for Vermont Yankee. On May 21, 1984 [see Reference h)], we issued a comprehensive response to the findings detailed in the November 2,1983 Inspection Report. This letter also . included a proposed schedule for the completion of the expanded III.G.2 re-survey (including the associated circuits study),

provided schedules for completing enhancements /modificiations known to be necessary as a result of our initial re-survey, and committed to certain interim compensatory measures until the enhancements /

modifications were completed.

We met with the NRC in King of Prussia on May.24, 1984 to discuss and clarify the contents of our May 21, 1984 Inspection Report response, as well as other correspondence recently submitted by Vermont Yankee

'(i.e., a request for exemption from the 72-hour cold shutdown requirement). At that time, we restated our intention to conduct a complete III.G.2 re-survey of the plant including a circuits separation and associated circuits study. At the meeting we also . stated our intention to keep the NRC informed as to our progress and findings as the study progressed, focusing on any additional compensatory measures

which may be deemed necessary. Finally, at the request of the staff, we agreed to consider additional compensatory measures beyond those committed to in our May 21, 1984 Inspection Report response.

The expanded III.G.2 re-survey and associated circuits study was per-formed during June and July 1984 culminating in a draft report which detailed the findings. During August 1984, we performed an engi-neering evaluation of the findings to scope the necessary modifica-tions and corrective measures to address additional areas that did not meet the specific separation criteria of Section III.G.2 of Appendix R. We also began drafting a comprehensive report which would be sub-mitted to the NRC. In early September, we' informed the NRC of the ,

status of our efforts and committed to implement additional interim compensatory measures based on the findings of the expanded re-survey.

On November 26, 1984, we submitted the results of the expanded re-survey in a report entitled " Safe Shutdown Capability Analysis". This

. report was submitted in draft. form at the request of the NRC so as to ensure that the format and technical content were sufficient for NRC review purposes. This report contains the re-survey results, asso-ciated circuits analysis, basis for acceptability of " fire zone bounardies," and identifies all corrective actions required for ulti-

-mate compliance with Appendix R.

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1. m _ 5 SU)S.LNuclear Regulatory Commission

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EMarch 1,'1985'

• Pagi 7J

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VElulONT YANKEE NUCLEAR POWER CORPORATION

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We ha'd i scheduled a meeting with the NRC for December 21, 1984 to-di.scuss their comments on the report, but the NRC was unable to sup-port the meeting. Because of the subsequent difficulty in trying to

~

> reschedul.e the meeting, it. is our intent to. finalize and formally l docket the~' report in the near future. We will _ also be submitting additional _ requests for exemption from the provisions- of Section III.G

~o f Appendix R-(as described in Appendix A of the " Safe Shutdow1

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Capability Analysis Report"), and will continue with the completion of M

'

.necessary fire protection enhancements / modifications and other correc-

tive1 actions _(i.e., procedural revisions).

~

- Asidiscussed.above,' we. believe Vermont Yankee has been responsive to

{the concerns. identified by the NRC and is committed to closing out the.

"

Appendix R . issue at .the earliest possible time'. Although we had phi-g losophicalidifferences with the-staff with respect to the need to y .' assume a' random loss of off-site power coincident with a fire as part ok .of ~our reanalysis and the acceptability of fire zones in lieu of fire V- - -

areas, we believe these differences in interpreting the requirements 4 - ' of; Appendix. R should not be characterized as failure to respond to the

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- NRC's concerns in a timely manner. - We also believe that _ based on the

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4 extensive ' efforts we have made to achieve full compliance with W >

Appendix R, a strong case can be'made for.a better performance rating than the one received:for.the Fire Protection and Housekeeping func-

- tional area. ~ At a minimum, we expect' the text of the~ final SALP Report to more accurately reflect the high degree of management atten-tion:and true level of responsiveness to this issue that Vermont-

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Yankee ~has displayed.

~ Vermont" Yankee took and continues to take an ~ aggressive leadership:

position in completing compliance with . Appendix R and looks forward to

1a prompt review by NRC of the " Safe ~ Shutdown Capability Analysis "

Report" such that the steps we are now taking based ~on that report can be concluded as acceptable in meeting the provisions of Section III.G

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of Appendix'R.-

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(3)'SectionIII.A,OverallFacilityEvaluation(Page61 '

~The_SALP Report states that, "this assessment noted numerous personnel Lerrors during the performance of routine duties in the-surveillance

' radiological controls, operating and refueling functional areas. The

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"

errors resulted from either a lack of attention to details during perfor-mance of routine duties or an over-reliance on experience as a substitute-for strict adherence to established procedures."

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- ? ' U,S. Nucicar -Regulatory Commission

. ' March 1, 1985

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Pagm 8 We clearly recognize.the need for continued management attention and involvement to minimize the instances of human error. Events resulting fro :' human error are and will continue to be evaluated to determine if pro-

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cedural, policy, programmatic, or design changes are warranted. We will continue to focus our attention at minimizing the instances and consequen-ces of human error.

Again, we appreciate meeting.with you and the Board to discuss the subject report and to reaffirm Vermont Yankee management's continued commitment to the safe and efficient operation of the Vermont Yankee Nuclear Power Station. Be assured that although we do not always agree with your assessment findings, we will' always be responsive to your concerns of safety and/or compliance. We view

. your assessment of our perfo;mance as positive input to enable us to carry out our commitments and responsibilities. We hope you consider our feedback as positive input into your evaluation / assessment process.

Very truly yours, VERMONT YANKEE NUCLEAR POWER CORPORATION W y Warren P Murph'y V I Vice President and Manager of Operations WPM /dm-t-

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if ' ' ENCLOSURE 1 l

VERMONT YANKEE /NRC CHRON0 LOGY OF EVENTS REGARDING' APPENDIX R TO 10CFR50

~Since the inspection of August 29/ September 2, 1983, Vermont Yankee has taken nany steps in its attenpt to comply with NRC requirements. These steps include:

1. September 1983 - Began a review of Appendix R and the licensing correspon-dence in order to understand why part III.G had not been applied to-the Reactor Building.

2. October 1983 - Concluded that, per Appendix- R, the entire plant should .have been re-examined for compliance with III.G. Began re-survey. Generic letter 83-33 issued.

3. . November 1983 - Received written results of inspection. Studies and reviewed inspection results in detail with management.

Completed first draft of " Analysis to Demonstrate Safe Shutdown Capability During and After Fires".

2 Held internal meetings to consider safety implications of the inspection findings.

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' Attend'ed meeting with Region I staff to demonstrate that continued safe

< ' operation of the plant was possible_ and justified, based on the low proba-bility of the adverse safety effects from the.non-conforming areas and the redundant safe shutdown capabilities of the plant. The plant was allowed

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to continue operating. The NRC Region I Report confirming continued safe operation was received in~ January 1984.

4 . December 1983 - Continued to examine're-survey.- Scoped design changes and exemptions to correct inspection deficiencies. Held internal management and engineering meeting to discuss status of Appendix R.-

.5. January 1984 - Attended Enforcement Conference at Region I. Committed to completing design changes to correct inspection deficiencies by December 1984 6. February 1984 - Plant and NSD engineers attended the Fire Protection Seminar in Washington, D.C. These engineers met informally with NRC staff the day before the seminar to review draft exemption requests relative to inspecticn items.

7. March 1984 - Formally submitted exemption requests relative to Inspection Report 83-26 commitments.

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- -2- ENCLOSURE 1

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8. - April 1984 - Engineers attended Region I Regional Workshop and also met

. informally with NRR staff for scoping comments, as the workshop indicated that drafts of studies on III.G compliance were to be informally submitted to the staff and discussed before formal submittal.

9. April la84 - Meeting on Appendix R was held during which management-directed the engineering staff to expand the scope of the III.G.2 re-survey to include random loss of off-site power.

10. May 1984 - Held scheduled and follow-up conference with Region I staff.

The NRC minutes of this meeting note that the meeting was at Vermont Yankee's request. Vermont . Yankee was only required to commit to a schedule within thirty days of the meeting with NRR staff to review the draft work on III.G. The NRR meeting has not yet taken place.

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Vermont Yankee committed to Interim Compensatory Measures at this time.

The need for these measures had only been clarified at the Regional Workshop.

11. Mid-July 1984 - Conducted formal re-survey, circuit separation, and asso-ciated circuit analyses.

% -12. _ July 1984 - Completed minor changes to Alternate Shutdown System and a _ declared it operational per required schedule.

Made changes to two plant circuits so that plant would conform to the-inspection and associated circuits study. One change addressed an inspec-tion finding. One change resulted from the associated circuits study.

13. August 1984 - Completed design scoping of corrective measures. based on re-survey and associated circuits study. Began drafting report and optimizing solutions .

14 September 1984 - Informed NRC Region I of progress. Implemented additional Interim Compensatory Measures.

15. November 1984 - Submitted draft " Safe Shutdown Capability Analysis" to NRR and Region I staffs, per process indicated at regional workshops.

Extensive color photographs' included.

16. January 1985 Completing installation of design changes to address Inspection Report commitments. Began design change on additional emergency-lighting, made necessary by findings of draft " Safe Shutdown Capability Anaysis Report".

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-.V -3- ENCLOSURE 1

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. 17. Fe' o ruary 1985 - Completed installation of all modifications detailed in

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tiarch 14,1984.. Inspection Report response letter.

18. .In Progress - Completing installation of additional modifications detailed in :the " Safe Shutdown Capability Analysis Raport". Preparing formal exar.p-tion requests for submittal to NRR. Completing necessary procedural changes as detailed in the Analysis Report. Continuing with interim ccm-pensatory measures until all actions complete.

19. September 1983 through Present - Scope and status of engineering efforts to

' address Appendix R inspection ' findings were discussed at routine monthly meetings- held between Vermont Yankee and Yankee Atomic Electric Company -

engineers and management. These monthly meetings.are held to discuss the scope and ' status of ongoing engineering activities.

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ENCLOSURE 5

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NRC-REVIEW 0F SALP REPORT 50-271/84-25 COMENTS PROVIDED IN VERMONT YANKEE-NUCLEAR POWER CORPORATION LEIIER FVY 85-22, MARCH 1, 1985 A. Operations (SALP Report Section IV'.A.1)

lThe' Region I' staff h'as- reviewed the comments in your letter .regarding (1).your disagreement with the NRC assertion that your continued operation from Sep-

?temberc16-18,'1984 was neither prudent nor conservative, and (2).that the NRC

conclusion.that Operations declined over the report period due to a "noncon-

. servative operational philosophy" was unfounded.-

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' :The Region.I staff agrees that-your response to the event was " conservative and prudent" to a-degree. .However, at the. time of.the event on September 16-17,-1984,~you were unable to respond to NRC staff questions regarding the

. potential outcome of analyzed transients. Your explanation of-the event lacked;a quantitative assessment that showed there was reasonable. assurance g

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'within~the spectrum of analyzed transients and accidents that safety margins would be preserved. -Your operations proceeded during that period without a previously accepted engineering basis to provide the. assurance that the plant was' operating without undue risk. Given these' uncertainties, the Region I

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' staff-did not find sufficient justification for. continued operation in that-anomalous state and our position was and remains that your approach was not conservativ'e enough to assure safety margins had not been significantly eroded.

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Your plans to continue operation to further study the'possible causes'for-the

anomaly were apparently curtailed because of the increasing level of NRC con-

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cern. The post event engineering evaluation concluding that no unacceptable-conditions would have-occurred as a result of the-loose reactor internals did

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not in any way1 justify your continued operation in that anomalous unanalyzed condition.

For these reasons your decision to continue operation on September 16,'1984 and your overall response' to this event are characterized as less prudent and

conservative than has been previously observed.. The staff conclusion in Paragraph IV. A.1 of the SALP Re' port remains Category 1, declining, for-the -

Functional. Area. The Board recommendation also remains unaffected.

B. Fire Protection and Housekeeping (SALP Report Section IV.E.1)

-The Region. I staff has reviewed the comments in your letter on the statements

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in Section IV.E of-the SALP. Following is our detailed evaluation of your

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comments.

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lli SALPTStatement LThe SALP Rep'rt o states that as a result of the August / September 1983

' inspection'of Vermont Yankee's compliance to Section III of. Appendix

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1R to 10 CFR.50, Fire Protection Requirements, "one violation is being

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. considered for escalated enforcement action."

c 11[1 .VY Comment-p. .

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As, described.in various submittals made to the NRC subsequent to.- ,

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the 1983 Appendix R inspection, Vermont. Yankee has performed exten-

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D sive engineering a'nalyses to' demonstrate'the' adequacy of our fire-protection capabilities and is continuing with.the installation of- l additional-modifications to enhance our overall fire protection de-

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. sign.

We believe,our response to;the findings:of the 1983 inspec-

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tion have been timely,; responsive to the NRC_s concerns, and re- ,

flect aistrong commitment on the.part of Vermont Yankee and its "

management to achieve compliance with the provisions of' Appendix

, R. We recognize the time it has taken to address all of the NRC's

, concerns; however, the' continual issuance of guidance: criteria by NRR-regarding Appendix R compliance:(i.e., Generic Letter-83-33, 1 Generic. Letter;85-01, various'Information Notices, and information resulting from Regional Workshops) has made the task of achieving-literal compliance that much more difficult. Enclosure 1-provides-a chronological _ history of events associated with the. Appendix R-issue:at Vermont Yankee. These events include formal'submittals-

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'made by us as well--a's numerous meetings; held with various NRC staff members to resolve this issue. Based on these efforts, we.believe  :

-that the pending. enforcement action should be. formally dispositioned t with a: finding that escalated enforcement action is not warranted.

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1.2- NRC Staff Evaluation The staff agrees that VY performed extensive engineering analysis and additional modifications as corrective actions for the-inspec-tion findings identified in'the violation. However,.the~ staff dis- ;

agrees with the VY comment that VY responses to the findings have 1

< been timely. The findings were identified during the inspection . '

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in August-September 1983, but VY-did not become fully committed to- H address the identified deficiencies and differences between his and.

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the NRC staff's position until April 1984. Considerable ~NRC effort

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was' required.to get VY to perform the reanalyses and implement the-actions necessary to correct the~ violation.

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'2. SALP Statement- I The SALP: Report states that, "the licensee did not take the initiative l to assure that his assumptions for the Reactor Building were consistent l

with the NRC staff's position. Licensee exceptions to-the requirements  !

were not properly identified to the NRC staff."  :

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Encl'osure:5

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2.1' VY Comment

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, cThis statement. reflects-the basic misunderstanding between the NRC tand Vermont Yankee with respect to compliance with Section III.G

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to Appendix R. .As discussed at the January 10, 1984 Enforcement Conference held in King of Prussia, we have received various cor-

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respondence from the NRC which indicated to us that to comply with Section III.G, we need only provide alternate safe shutdown cap-

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ability.for the_ Control Room, Cable-Vault and Switchgear Room.

Th_is correspondence,-' discussed in' detail in.the March 13, 1984.En-

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forcement Conference Meeting Minutes (see Reference e),-was the .

basis for our- conclu'sion that our overall Fire Protection Program satisfied the intent of Section III.'G with respect to our Reactor Building and that.the NRC was cognizant of our assumptions and

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-positions which were documented in our 1978 Fire Hazards Analysis'

. Report.

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- 2. 2 : NRC Staff Evaluation lWhenAppendix_R.cameintoeffect.in1981,VYshould.haveaskedfor

clarification from the.NRC staff with respect-to the need to do'a post-Appendix R _ reanalysis _ including-the Reactor building and proper-assumptions to use therein'i: 'Instead, VY incorrectly _ ass _umed that their pre-Appendix R analysis and its assumptions were good enough.

- At the time Appendix R came.into effect in 1981', the licensee's technical staff reviewed Appendix R, theLpreamble in the Federal-

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tRegister and Generic Letter 81-12.- Their conclusion was that a-

-resurvey was required. The licensee's management had.been deeply involved in Fire Protection concerns for several years and disagreed twith their technical staff. There is no evidence of;the licensee

, requesting ~ clarification for their need to conduct Appendix R re--

analyses from NRR.

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< 3. :SALP Statement

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' The.SALP' Report states that, _"The licensee took considerable--time to. re-

-spond to the issues identified by the inspection team. ~ The NRC positions regarding the Appendix R requirements were clearly presented'to the lic-ensee by the NRC review team in August 1983, but the licensee _did not become fully: committed to addres's the identified deficiencies and dif- '

. - Lferences between his and the:NRC staff's. position until April 1984.

-Considerable NRC effort.was required to get the licensee to perform the

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. reanalysis.and implement the actions necessary to correct the violation."

The Report also' states that, "the lower rating this assessment period-is due to the licensee's incorrect implementation of the Appendix R rule

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w - and the licensee's' slowness in responding to the NRC initiatives once-deficient areas were identified."

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3.1' VY Comment ~

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As discussed above; we believe our response to the findings were.

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timely and appropriate.' ~Immediately following the 1983 inspection,

-we. initiated a re-review of the requirements of Appendix'R to 10

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-CFR:Part 50 and' initiated a re-survey of our.. Reactor,8uilding_to ensureiwe met the separation criteria of Section~III.G.2. It should

be noted that our re-survey'was performed assuming a fire' induced

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loss'of off-site power which was consistent with our original in-terpretation of the loss of off-site power provisions-of Appendix

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R. :We'also initiated a breaker coordination study and an engineer-

~-ing analysis of plantisafe shutdown systems against the separation:

1 and fire protection criteria ~of Section-III.G.2.of Appendix R.

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- f Once the formal ~ findings of'the inspection were' issued in'the No'-

vember 2, 1983 Inspection Report (see Reference.f), we performed

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an analysis of the safety significance of the inspection's findings.

The results of our analysis were presented to the NRC'at a November

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-22, Ic83 meeting at King of Prussia where it was concluded that '

therewere no findings that would preclude our l ability to' safely shut down the plant in the event of~a fire. 'At that: meeting,-we

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also presented a draft report. entitled, " Analysis to Demonstrate

, . Safe Shutdown Capability During and After' Fires,"=which documented our analysis.

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Following the November 22, 1983. meeting', we continued with-our en-

_ gineering efforts- to' address the deficiencies cited in the November

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-2,-1983' Inspection' Report.' -We reported the scope and status'of our-

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efforts to the NRC at an. Enforcement Conference held in King of-o Prussia on January .10,- 1984. We also discussed differences between Vermont' Yankee and the.NRC in interpreting-certain provisions of Appendix R. -Specifically, the NRC-stated that Appendix R implies

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the.'need for? safe: shutdown capability of the plant, assuming _a loss of off-site power for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> concurrent with a fire in any area

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.of the plant. Vermont. Yankee's assumption, both in the pre-Appendix

'R Fire Hazards: Survey and the post-inspection re-survey was that-

. a non-mechanistic;1oss of off-site power need not'be considered.

The NRC.also ' stated that Vermont Yankee's'use of " fire zone bounda-C ries"Lin the Reactor Building did not meet Appendix R requirements -

! for physical " fire- area: boundaries.'" Vermont Yankee's position was

. that the inherent spatial separation between areas of the Reactor-

Building,' coupled with physica1Jbarriers and extensive fire protec-

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" tion features installed.in,the-Reactor Building were the intent of the " fire area boundaries.

'At-the~ January 10 Conference, Vermont Yankee committed to submit

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the results 'of the'Section III.G.2 re-survey (including an associ-r

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ated circuits' study), docket'. formal requests for exemption from the specific separation criteria of-Section III.G.2, and submit the de-

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tails of fire protection-system enhancements and/or corrective ac-

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Enclosure-5 5'

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-tions deemed necessary as a result of the re-survey. At the time,-

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'we still- believed our loss of off-site power and fire area boundary

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positions were justifiable. ,

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On March 14,-1985 (see Reference g), we submitted requests for ex- i a emption for certain areas of the Reactor Building that did not meet

, thefspecific separation criteria for Section III.G.2. The exemption

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. requests. included the technical basis for justification as well as M:

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descriptions of specific fire protection system enhancementshodi-

>fications which we deemed were necessary as a result of our III.G.2 re-survey.

'In April 1984,'our design engineers attended an NRC sponsored-Re-

.gional Workshop-held to. discuss and clarify NRC positions regarding compliance with the provisions of Appendix R. Following the Re- i

, gional. Workshop, a meeting was held between Vermont Yankee and Yan-kee Atomic Electric Company engineers and management to discuss the

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results of the workshop and the status.of our Appendix R compliance

-efforts. . Based on an assessment from engineers who attended the.

Regional Workshop'that the NRC was adamant on the need to assume

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a random loss of off-site power coincident with a fire, Vermont'

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1Yankeemanagementdirectedtheengineeringstafftoexpandthescope- '

of.the III.G.2 re-survey (including the associated circuits study)

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to' include a random loss of off-site power. In; addition, based on r the results of the workshop and discussions with NRR fire-protection.

L engineers, it was recognized that additional detailed engineering would be necessary to support Vermont Yankee's position that " fire zone boundaries" in the. Reactor building were an acceptable. alter-fi, native to the NRC's " fire area-boundaries." -It was agreed that the, report documenting the expanded III.G.2 re-survey would include de -

tailed justification of the acceptability of " fire zone boundaries."

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Vermont Yankee then requested that a meeting be held with the NRC (NRR and I&E) to discuss the status of our' engineering efforts and

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establish a means to close out Apnendix R for Vermont Yankee. .On .

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.May 21, 1984 (see. Reference h),3.t i_ssued a comprehensive response

. . to the findings detailed in the Avember 2, 1983 Inspection Report.

This letter also included a prapoteci schedule for the completion

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.of the expanded III.G.2 re-survey (including the associated circuits-study), provided' schedules for completing enhancements / modifications;

'known to be'necessary as a resu;t of our initial re-survey, and

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committed to certain-interim compensatory measures until the en-

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hancements/ modifications were complete.

,' ~ , We met with the NRC in King of Prussia on May 24, 1984 to discuts

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.and clarify the contents of our May 21, 1984 Inspection Report re-sponse, as well'as other correspondence recently submitted by Ver-

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mont Yankee (i.e., a request for exemption from the 72-hour cold shutdown: requirement). At that time, we restated our intention to.

conduct a complete III.G.2 re-survey of the: plant including'a cir-

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Enclosure 5 6 cuits separation and associated circuits study. At the meeting we also stated our intention to keep the NRC informed as to our pro-gress and findings as the study progressed, focussing on any addi-tional compensatory measures which may be deemed necessary. Finally, at the request of the staff, we agreed to consider additional com-  %

pensatory measures beyond those committed to in our May 21, 1984 Inspection Report response.

i The expanded III.G.2 re-survey and associated circuits study was  !

performed during June and July 1984 culminating in a draft report i which detailed the findings. During August 1984, we performed an engineering evaluation of the findings to scope the necessary modi- 3 fications and corrective measures to address additional areas that did not meet the specific separation criteria of Section III.G.2 ]<

of Appendix R. We also began drafting a comprehensive report which would be submitted to the NRC. In early September, we informed the _

NRC of the status of our efforts and committed to implement addi-

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tional interim compensatory measures based on the findings of the 7 expanded re-survey. On November 26, 1984, we submitted the results r of the expanded re-survey in a report entitled, " Safe Shutdown Cap- (

ability Analysis." This report was submitted in draft form at the request of the NRC so as to ensure that the format and technical -

content were sufficient for NRC review purposes. This report con-tains the re-survey results, associated circuits analysis, basis for acceptability of " fire zone boundaries," and identified all cor-rective actions required for ultimate compliance with Appendix R.

We had scheduled a meeting with the NRC for December 21, 1984 to discuss their comments on the report, but the NRC was unable to M support the meeting. Because of the subsequent difficulty in trying -

to reschedule the meeting, it is our intent to finalize and formally docket the report in the near future. We will also be submitting additional requests for exemption from the provisions of Section '

III.G of Appendix R (as described in Appendix A of the " Safe Shut- ]

down Capability Analysis Report"), and will continue with the com- i pletion of necessary fire protection enhancements / modifications and i other corrective actions (i.e., procedural revisions). Y As discussed above, we believe Vermont Yankee has been responsive to the concerns identified by the NRC and is committed to closing  :

out the Appendix R issue at the earliest possible time. Although e.

we had philosophical differences with the staff with respect to the d need to assume a random loss of off site power coincident with a i fire as part of our reanalysis and the acceptability of fire zones _

in lieu of fire areas, we believe these differences in interpreting -

the requirements of Appendix R should not be characterized as fail- -

ure to respond to the NRC's concerns in a timely manner. We also

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believe that based on the extensive efforts we have made to achieve -

j s full compliance with Appendix R, a strong case can be made for a 4 better performance rating than the one received for the Fire Pro-  ;

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Enclosure 5- 7

'tection and Housekeeping functional area. At a minimum, we expect the text _of the final SALP Report to more accurately reflect the high degree of management attention and true level of responsiveness-to this issue that Vermont-Yankee has displayed. l Vermont Yankee took and continues to take an aggressive leadership position in compMting compliance with Appendix R and looks forward to a prompt review by NRC of the " Safe Shutdown Capability Analysis-Report" such that the steps we are now taking based on that report can be' concluded as acceptable in meeting the provisions of Section III.G of Appendix R.

3.2 NRC Staff Evaluation.

The staff maintains the position stated in the SALP and as quoted in Section 3 above. As stated in the NRC Staff Evaluation, Section- l 1.2 above, the staff disagrees with the VY comment that the VY re-sponses to the findings have been timely. Specifically, during the August 1983 Appendix R inspection of Vermont Yankee, the licensee was told by the inspection team that their Appendix R analysis was substantially defit:ient in that it included several major flaws.

These included analyses based on fire zone boundaries when they should have been based on fire areas. Furthermore, the analyses did not make the required assumption of -loss of offsite power con-current with a fire in any area of the plant. The licensee contended that this was a Region I position as opposed to an NRR position.

It is noted that two NRR members participated in the region-based inspection. The licensee was encouraged to request formal NRR clarification of any perceived differences in Regional versus NRR position regarding required analyses assumptions. Neither the in-spection team or NRR provided the licensee _ any bases for their view -

that the inspection team imposed requirements beyond these sanc-tioned by NRR. The licensee did not request clarification from NRR-regarding these issues.

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NRC Regional staff again pointed out these deficiencies to the lic-ensee during the November 22, 1983 meeting and the January 10, 1984 meeting in the Region and in several phone conversations between Region I and a memhis of the licensee's staff prior to the April 1984 Regional Workshop. In each of these instances, the Regional staff discussed the licensee's need to either correct the deficient analyses or request forma 1' clarification of the subject issues from ,

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NRR.

-In addition to the above, the NRC issued Generic Letter 83-33 in l December 1983 to all licensees (including Vermont Yankee). This j letter provided additional clarification of the issues supporting the Regional view of the licensees need to correct the subject de- ,

ficiencies. i

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. < On March 23,~1984,'the NRC presented the first of five Regional Workshops regarding clarification of Appendix R-requirements. The .

%  : first Workshop was held .in Chicago,- Illinois. A draft document- 1

. clarifying-Appendix:R issues:was distributed to all participants.

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-The. Workshop and the document again supported the Region position-ion the subject. issues.

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Despite:the- above actions by NRC, it was not until the. Appendix R '

LWorkshop'in Region.I on April ~18,-1984 that the licensee. agreed to correct their analyses and determine the need for related modifica- -

tions. This-represented a'seven mon _th period where the licensee

did little to correct;significant analytical deficiencies identified

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by.the NRC, despite continued staff comments to the licensee to-encourage resolution of the. issues.

-The staff does recognize that the: licensee did conduct analyses,.

albeit deficient analyses,.and they submitted a draft copy'of the'

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' analyses to the-Region on a timely basis following the' -inspection.

-They made extensive and prompt modifications following the inspec-tion to correct specific. hardware deficiencies. identified by the:

iNRC. .In. addition, following the'. April 18, 1984' Regional Workshop, .

.they corrected the' deficient analyses.on a.. timely basis.

4i ~ Conclusion-

The staff disagrees with the VY comment th'at the VY responses to the in-spection findings identified in the violation 'have been timely.' However, the staff agrees that VY made' extensive corrective actions:for the de-

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ficiencies identified in the violation.

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.The' staff conclusi_on in.the SALP Report, Paragraph IV.E.2, remains Cate -

. gory 2, consistent for this Functional Area. The Board recommendation

_ l documented.in the SALP Report,-Paragraph IV.E.3,=also remains unaffected.

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