ML20056G973

Insp Rept 50-271/93-14 on 930627-0807.Violations Noted.Major Areas Inspected:Operations,Maint,Engineering & Plant Support

ML20056G973
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	08/27/1993
From:	Eugene Kelly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20056G969	List: ML20056G968 ML20056G970 ML20056G973
References
50-271-93-14, NUDOCS 9309080019
Download: ML20056G973 (15)

See also: IR 05000271/1993014

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U.S. NUCLEAR REGULATORY COMMISSION

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REGION I

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Report No.

93-14

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Docket No.

50-271

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Licensee No.

DPR-28

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Licensee:

Vermont Yankee Nuclear Power Corporation

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RD 5, Box 169

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Ferry Road

Brattleboro, VT 05301

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Facility:

Vermont Yankee Nuclear Power Station

Vernon, Vermont

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Inspection Period:

June 27 - August 7,1993

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Inspectors:

Harold Eichenholz, Senior Resident Inspector

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Paul W. Harris, Resident inspector

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Approved by:

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JPAt/rs

Egene M. Kelly, Chief

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Keactor Projects Section

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Scope:

Station activities inspected by the resident staff this period included: operations,

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maintenance, engineering and plant support. An initiative selected for 'this

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inspection was the conduct of operator rounds. Backshift and " deep" backshift

including weekend activities amounting to 27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br /> were performed on June 28-

30; July 1,2,14,16,19,20 and 31. Interviews and discussions were conducted-

with members of Vermont Yankee management and staff as necessary to support

this inspection.

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Findings:

An overall assessment of performance during this period is summarized in the

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Executive Summary.

A violation of administrative controls involving two'

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examples of failure to follow procedures by contract personnel was identified -

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(Section 3.1.1).

Unresolved items were opened regarding VY engineering _

evaluation of seismic considerations for scaffolding installed near plant equipment

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(Section 4.1), design controls for a diesel generator fuel oil piping failure (Section

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3.1.6), and overtime controls during plant outage periods (Section 5.5).

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93090B0019 930827'

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PDR

ADOCK 05000271

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EXECUTIVE SUMMARY

Vermont Yankee Inspection Report 93-14

Operations

The operating organization responded well to plant equipment damage caused by an electrical

storm.

Operating personnel were knowledgeable of equipment and licensed conditions,

particularly limiting condition (LCO) maintenance for ECCS, reactor building temperature,

cooling tower fan degradation, and recirculation pump seal performance. Inconsistencies were

identified in auxiliary operator round documentation.

Maintenance

Failure to control the activities of a maintenance contractor in the areas of configuration and

material control, resulted in a violation of administrative procedures. Maintenance on an

emergency core cooling system was well planned and controlled. The review of a diesel

generator fuel oil leak identified concems involving design cor.Dguration control and initiation

of a Nonconformance Report.

Engineering

Engineering evaluation of scaffolding for seismic consideration was identified as an unresolved

item. Overall quality of the LERs reviewed was good. The review by the Plant Operations

Review Committee of an event involving the "A" emergency diesel generator was cursory.

Plant Support

Inadvertent siren activations resulted in appropriate corrective actions and a Vermont Yankee

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assessment of the adequacy of their siren maintenance program. The security organization

assessed industry information feedback to enhance testing and ensure that a security vulnerability

did not exist. The acceptability of Vermont Yankee administrative controls for overtime during

outages is unresolved pending NRC staff review.

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TABLE OF CONTENTS

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EXECUTIVE SUMMARY

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TABLE OF CONTENTS .......................................iii

1.0

SUMMARY OF FACILITY ACTIVITIES

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2.0

OPERATIONS (71707, 93702, 92700) . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

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2.1

Operational Safety Verification . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

2.2

Operator Awareness of Plant Maintenance . . . . . . . . . . . . . . . . . . . .

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2.3

Conduct of Operator Rounds

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2.4

Electrical Storm . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2

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3.0

MAINTENANCE (62703, 61726, 92700, 90712, 93702)

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3.1

Maintenance

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3.1.1 Configuration and Material Control . . . . . . . . . . . . . . . . . . . . 3

3.1.2 Power Loss to Reactor Core Isolation Cooling Suction Isolation

Val v e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4

3.1.3 Thermography of Cooling Tower Fan 2-1 . . . . . . . . . . . . . . . . 5

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3.1.4 Emergency Core Cooling System Maintenance . . . . . . . . . . . . . 6

3.1.5 Vernon Tie Line Inoperability . . . . . . . . . . . . . . . . . . . . . . . 6

3.1.6 Diesel Generator Fuel Oil Leak . . . . . . . . . . . . . . . . . . . . . . 7

3.2

Surveillance - Core Spray Sparger Break Detection System Differential

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Pressure Indicating Switch (DPIS) . . . . . . . . . . . . . . . . . . . . . . . . . 8

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4.0

ENGINEERING (71707,90712,90713,92700) . . . . .

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4.1

Seismic Evaluation of Installed Scaffolding . . . . . . . . . . . . . . . . . . . . 8

4.2

Review of Written Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8

5.0

PLANT SUPPORT (71707, 40500, 90712, 90713, 92700)

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5.1

Radiological Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9

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5.2

Inadvertent Siren Activation . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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5.3

S ecu ri ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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Overtime Controls During Outages . . . . . . . . . . . . . . . . . . . . . . . .

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5.5

Plant Operations Review Committee . . . . . . . . . . . . . . . . . . . . . . .

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A d ministrative . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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6.1

Preliminary Inspection Findings

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Note: Procedures from NRC Inspection Manual Chapter 2515, " Operating Reactor Inspection

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Program" which were used as inspection guidance are parenthetically listed for each applicable

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report section.

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DETAILS

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1.0

SU515fARY OF FACILITY ACTIVITIES

Vermont Yankee Nuclear Power Station was operated at or near full power during this

inspection period. On July 25, the licensee (or VY) reduced reactor power to 65 percent to

conduct a final control rod pattern adjustment prior to reactor power coastdown due to end-of-

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cycle fuel depletion. Coastdown was begun on July 26 and will continue at approximately 0.25

percent of rated power per day until the planned shutdown on August 27 to begin Refueling

Outage XVII.

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2.0

OPERATIONS (71707,93702,92700)

2.1

Operational Safety Verification

Daily, the inspectors verified adequate staffing, adherence to procedures and Technical

Specification (TS) limiting conditions for operation (LCO), operability of protective systems,

status of control room annunciators, and availability of emergency core cooling systems. Plant

tours confirmed that control panel indications accurately represented safety system line-ups.

Safety tagouts properly isolated equipment for maintenance.

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2.2

Operator Awareness of Plant Maintenance

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During this period, ihe inspector observed that control room operators (CROs) were cognizant

of maintenance in the plant. For example: (1) shift personnel referred to the details contained

within the LCO Maintenance Plan for the "B" low pressure coolant injection (LPCI) system and

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articulated the scope and schedule for the maintenance; and (2) following a July 30th electrical

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storm (Section 2.4), shifts were knowledgeable of the work orders initiated to repair damaged

equipment and information gathered from the event. Similarly, the inspector observed that an

appropriate level of detail was communicated during the shift turnovers regarding the status of

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maintenance. This included increased awareness of reactor recirculation pump seal performance,

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the status of disabled computer points, and temperature conditions within the reactor building,

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Operating shifts were knowledgeable of LCO action statements entered, status of redundant

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systems, and documentation controlling maintenance.

2.3

Conduct of Operator Rounds

On July 14 and 22, the inspector assessed the performance of reactor building log keeping using

NRC Information Notice (IN) 92-30, " Falsification of Plant Records" and plant Procedure AP

0150, Rev. 29, " Conduct of Operations and Operator Rounds" as inspection guides. On both

dates, the inspector concluded that the conduct of operator rounds in the reactor building (RB)

were in accordance with plant procedures and TS. However, some equipment checks were

inconsistent and log keeping was incomplete.

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Specifically, Procedure AP 0150 requires that general equipment checks be performed on

selected pieces of equipment located within the log keeping round. These qualitative checks

include the assessment of motor and pump housing temperatures, oils levels, fluid leakages,

ventilation flow, noise and vibrations. During the July 14 RB tour, equipment checks performed

by the auxiliary operator (AO) on the uninterruptible power supplies (UPS) and reactor

rceirculation pump (RRP) motor generator sets were performed differently for each machine.

For example, the status of brush arcing was not assessed on the RRPs and ventilation flow and

bearing temperatures were not consistently verified. Despite this, the AO properly assessed the

performance of the UPSs and RRPs in accordance with the specific log requirements.~ The

second observation occurred when three log keeping requirements ("B" UPS, residual heat

removal system, and the reactor water cleanup pump "B") were not documented following the

completion of the round, although the rounds sheet had yet to be reviewed by the Shift

Supervisor. In this case, the AO acknowledged the inspector's observation, was knowledgeable

of results of the surveillances, and was observed by the inspector performing the required

rounds. The log was subsequently updated and presented for Shift Supervisor review. The

Operations Manager (OM) acknowledged this concern and discussed this issue with the subject

operating crew and during a subsequent Shift Supervisor meeting.

2.4

Electrical Storm

On July 30, VY sustained multiple lightning strikes during a brief but severe electrical storm.

The lightning strikes occurred between 3:21 p.m. and 3:23 p.m. when a control room panel

mounted indicator light for the stack offgas filter (non-nuclear safety) caught fire; 12 annunciator

cards on control rov.: panels (CRP) 9-6, 9-7, 9-8 (feed, condensate, electrical distribution

panels) alarmed and would not clear; 36 emergency response information system (ERFIS)

computer points were inoperable; and a number of other alarms annunciated. Control room

operators verified that the plant remained at steady state power, confirmed actual status for alarm

conditions, and commenced actions to restore the am.unciators and lost panel indications. Eye

witnesses confirmed lightning strikes in the vicinity of the onsite 300 ft. meteorological tower

and 345 kV switchyard.

The burning indicator light was immediately extinguished by the CRO using a halon gas

extinguisher. Inspections of the control panels confirmed that the source of the 4-inch long

candle-like flame was the plastic lens cap for the indicator light. No other damage was sustained

due to the fire. The fire brigade commander (the Shift Engineer) was in the control room during

the event, coordinated the fire response, and determined that the severity of the fire was

minimal. Based on the immediate actions taken and fire assessment, the Shift Supervisor

determined that activation of the fire brigade was unnecessary. The inspector determined that

the actions of the operating crew were appropriate and in accordance with plant Procedure OP

3020 Rev.17, " Fire Brigade and Fire Fighting Procedure."

The assessment of plant equipment damage was timely. A coordinated effort was implemented

to restore damaged control panel indications and, within one hour, repairs were completed. The

Shift Supervisor effectively delegated control panel responsibilities to the senior CRO so as to

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maintain an appropriate " big picture" perspective during the plant recovery phase.

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atmosphere within the control room remained professional and the CROs were attentive to duty,

as demonstrated by the identification of the loss of the Vernon hydroelectric tie line (Section

3.1.5). Department managers assisted in the coordination of plant and equipment inspections

to further identify storm related damage. The assessment of damaged equipment and the effect

on plant operations were effectively communicated to the Shift Supervisor. All alarming

conditions were properly responded to. Prior to this event, the operating crew implemented

compensatory measures in accordance with Procedure OP 3127, Rev. 4, " Natural Phenomena"

upon being informed of the storm in the area.

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MAINTENANCE (62703, 61726, 92700, 90712, 93702)

The inspectors observed selected maintenance and surveillance on safety-related equipment to

determine whether these activities were effectively conducted in accordance with TS and

administrative controls (Procedure AP-0021 and AP-4000) using approved procedures, safe

tagout practices and appropriate industry codes ad standards. Interviews were conducted with

the cognizant engineers and maintenance personnel and ,endor equipment manuals were

reviewed.

3.1

Maintenance

3.1.1 Configuration nnd Material Control

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Plexiglass Covering

On July 23, the inspector observed that a temporary plexiglass I anel was installed over the

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controls of the alternate shutdown panel (CP-82-1) located in the reactor core isolation cooling

(RCIC) room inside the reactor building. The plexiglass prevented access to all panel controls

and was securely fastened to panel support brackets. The Shift Supen'isor was immediately

informed and the plexiglass was removed. Panel CP-82-1 is used in accordance with plant

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Procedure OP 3126, Rev. 9, " Shutdown Using Alternate Shutdown Methods" to allow the

remote, safe shutdown of the reactor from outside the control room during eveats involving

control room evacuation. The functions performed on this panel include core and containment

cooling and reactor vessel water level control.

The plexiglass had been installed July 22 to prevent damage to the controls on the alternate

shutdown panel during maintenance on a ventilation damper. Inadvertent breaker operations had

occurred previously during maintenance, and the installation of the plexiglass was considered

by the contractor performing work (Mercury Co.) to prevent recurrence (NRC Inspection Report

93-02). Plant Procedure AP 0020, Rev.15, " Temporary Modifications" requires review and

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evaluation of modifications to plant equipment. The inspector concluded that the installation of

the plexiglass restricted access to the shutdown panel and represented a temporary modification,

and therefore should have been evaluated for any adverse impact on operator accessibility to the

controls.

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Housekeeping

On July 27, during a RB tour, the inspector observed that housekeeping conditions in the "B"

residual heat removal (RHR) toom potentially affected the proper performance of the "B"

emergency core cooling system (ECCS).

For example, a 2 X 4 inch piece of wood,

approximately five feet long, rested unsupported on top of the RHR "B" motor; tools and

materials were stored on top of the RHR suction and discharge piping and pump impeller

casings; and rags, plastic and anti-contamination clothing were scattered on the floor in the

general vicinity of the floor sump pumps. The Operations Manager and Shift Supervisor were

immediately informed and cleanup of the room commenced.

Operations personnel and

department managers inspected the room and concluded that the operability of the ECCS

equipment was not affected, however, the conditions were not in accordance with VY

administrative controls for housekeeping requirements. The subject materials and conditions

were the result of asbestos abatement by contractors of RHR piping, and had been present for

a period of time in excess of a shift but less than one day.

Plant Procedure AP 6024 Rev. 8, " Plant Housekeeping" requires that, during work efforts, the

work party shall maintain good housekeeping practices at all times. The inspector concluded

that the above described conditions found on July 27 reflected poor housekeeping practices, and

were in violation of Procedure AP6024

Findings

Technical Specification 6.5 requires that written procedures be adhered to. The two examples

enumerated above represent a failure to adhere to procedural requirements and are considered

a violation (VIO 93-14-01).

3.1.2 Power Loss to Reactor Core Isolation Cooling Suction Isolation Valve

During the restoration valve lineup from a Reactor Core Isolation Cooling (RCIC) pump suction

transfer surveillance, the molded case circuit breaker for RCIC-18 tripped on August 2nd,

immediately after the valve reached its full open position. Control panel indication lights

deenergized and operators were unable to remotely operate the valve (RCIC-18). The RCIC

system was declared inoperable, a 4-hour report in accordance with 10 CFR 50.72 was promptly

made, and corrective maintenance was commenced.

Motor operator and torque switch

inspections, meggering and resistance checks, verification of running currents, and cycling of;

the molded case circuit breaker to verify freedom of motion, were inconclusive in determining

the failure mechanism. Following three hours ofinvestigation, VY was unable to determine the

cause of the breaker trip, but nonetheless declared the RCIC system operable.

On August 4, following evaluation of the design operation of the RCIC system suction transfer,

review of Procedure OP 5210, Rev.1, "MCC Inspections," and discussions with the cognizant

engineer, the inspector questioned the basis for not performing time-current characteristic testing

of the molded case circuit breaker. This test would verify the proper trip setting of the breaker

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and provide further assurance that the breaker itself was not the cause of the " spurious" trip.

In addition, the manager stated that VY would review the surveillance to assure that the

condition was not caused by operator actions while implementing the procedure.

Later that same day, VY re-entered the RCIC system 7-day LCO (back-dating the LCO entry

date to August 2) to perform the time-current test and to simulate the actions taken during the

initial restoration valve lineup. Vermont Yankee verified that: (1) the molded case circuit

breaker was operating within proper time-current characteristics, and (2) that an automatic

closure signal to RCIC-18 was present when RCIC-18 reached the full open limit switch. The

closure signal was present because the suppression chamber isolation valves (RCIC-39 and 41)

were full open when the condensate storage tank (CST) isolation valve (RCIC-18) reached its

full open position. This particular lineup satisfies RCIC suction logic to initiate closure of the

RCIC-18 valve, in part, to prevent inadvertent draining of the CST to the suppressien chamber.

The inspector independently verified that the logic circuitry supported this conclusion and that

the high pressure coolart injection (HPCI) system was similarly vulnerable. Vermont Yankee

also verified that the motor current generated by the immediate reversal (plugging current) of

the motor operator was sufficient to trip the molded case circuit breaker. Three days into the

7-day LCO, the action statement was exited.

Vermont Yankee (a) reviewed the RCIC and HPCI surveillance procedures to clarify valve

restoration, (b) added RCIC-18 valve to the VY molded case circuit breaker testing program

(NRC Inspection Repc-t 92-81), and (c) assessed operator training in this area.

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Maintenance Manager .m a discussed management expectations at the weekly department head

meeting regarding backshift equipment repairs, inter-departmental communications, and

independent reviews of equipment failures.

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3.1.3 Thermography of Cooling Tower Fan 2-1

During semi-annual thermographic imaging of the alternate cooling tower fan 2-1 motor

controller, VY identified that the temperature of the "C" phase contact on the motor controller

was approximately 100 degrees hotter than the average contact temperature. Maintenance

engineering evaluated the condition and concluded that the fan was operable and that failure was

not imminent. A Priority 1 work order was written and within four days, parts availability was

verified and planning was completed to coordinate the removal of the fan from service. The

timeliness of this repair was consistent with the ability to prepare for the activity and ensure

readiness to conduct the work without excessive out-of-service time.

The inspector performed field inspections to assess the condition of the motor controller and

quality of maintenance. Good supervisory oversight was provided by the electrical foreman and

cognizant engineer, and field notes accurately described the as-found conditions.

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maintenance staff evaluated these conditions and identified discoloration, pitting, and residue

formation on the contact indicating excessive and non-uniform heating of the "C" phase contacts.

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The motor contactor was replaced and post-maintenance testing (PMT) verified the adequacy of

the maintenance performed. Thermography was conducted to base-line the new motor contactor.

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3.1.4 Emergency Core Cooling System Maintenance

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During this inspection period, VY voluntarily entered the TS LCO action statements for the "B

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LPCI and the "B" residual heat removal service water (RHRSW) sub-systems. Maintenance on

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the emergency core cooling system (ECCS) sub-systems was performed concurrently and lasted

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for approximately four days with the most limiting LCO period being seven days. Two

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dedicated 12-hour work shifts were assigned. lead supervisors were assigned to the project and

were observed in the field inspecting work in progress.

The surveillances verified that

instruments and controls were in calibration and reflected good attention to detail, as evidenced

by pinched wires identified within an instrument cabinet.

The scope of the maintenance was both corrective and preventive. Packing leaks were repaired;

motor control centers were inspected; and, lubricating oil for the ECCS pump motors was

changed.

In addition, limitorque operator inspections and testing were performed.

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inspection was also conducted by VY of the RHRSW-898 motor operator to assess the condition

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of the drive motor pinion gear key. This VY inspection identified no concerns and was

performed as part of corrective actions implemented from a key failure on RHRSW-89A (NRC

Inspection Report 93-13). Post-maintenance testing was effective, as demonstrated by the

subsequent good performance of equipment and the identification of an improper installation of

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a chamfered bushing in the mechanical seal on the RHRSW "B" pump. The original seal was

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replaced due to end-of-life performance concerns and the replacement seal leaked excessively

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during the PMT. Vermont Yankee determined that the leak was caused by poor workmanship

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during seal installation. Overall, the maintenance was well planned and documentation was

detailed.

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3.1.5 Vernon Tie Line Inoperability

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As a result of the electrical storm on July 30 (Section 2.4), the 4160 kV AC power line from

the Vernon hydroelectric power station became inoperable. The power line runs above gtcund

from the hydro station to the circulation water discharge structure where it enters underground

pipe and conduit chases prior to entering the plant. The power line is non-seismic, non-safety

class, and supplies safety-related bus nos. 3 and 4 through circuit breakers. The line is required

by TS to be operable for reactor startup. Vermont Yankee takes credit for the Vernon tie line

in their station blackout (SBO) analysis, citing adequate reliability and load carrying capability

(3 MW), and is currently being upgraded to meet the NRC SBO rule,10 CFR 50.63.

Vermont Yankee confirmed that the inoperability of the Vernon tie line was caused by a

lightning strike to a transmission pole, based upon the identification of electrical arc marks,

charring of the 4160 kV cable, and insulator damage.

Maintenance personnel promptly

commenced troubleshooting, performed a "one-for-one" evaluation for the insulator replacement,

and completed the repair withi, six days. Detailed inspections of line components confirmed

no other damage. Vermont Yankee effectively coordinated with the operators of the Vernon

hydroelectric station and within its own organization to restore the system to operation.

Availability of the Vernon tie line has been the subject of SBO correspondence with the NRC,

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and reported by VY to be highly reliable dudng the 20-year operating history of the plant. In

any case, the portion of line struck will be replaced by a new buried supply scheduled to be

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energized and tied into the plant in late August 1993.

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3.1.6 Diesel Generator Fuel Oil Irak

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During the monthly surveillance of the "A" emergency diesel generator (EDG), VY identified

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a fuel oil (FO) leak in a 1/2-inch pipe nipple on the suction side of the FO pump, and entered

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the 7-day TS LCO. An inspection of the "B" EDG FO piping identified no similar deficiency,

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but only because a similar problem had developed on that engine in 1989 and was repaired at .

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that time. Following evaluation, repair, and testing, the "A" EDG was declared operable. The

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"A" EDG was out-of-service for approximately a 12-hour period, and the event was considered

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to be a valid test failure.

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Two concerns were identified regarding this event. The first involved the identification by VY

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that, prior to this event, design configuration control for this piping had not been maintained.

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Based on their review of maintenance records, consultation with the vendor, and performance

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of ultrasonic testing, VY identified that incorrect schedule piping was installed at the suction of

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the FO pump and at the discharge of the FO priming pump for both EDGs. One-for-one

evaluations assessed and justified the use of piping nipples different than design; replacement

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nipples were installed accordingly. The preliminary root cause developed by VY engineering

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based on metallurgical evaluations was attributed to fatigue failure with pipe loading, surface

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quality, and wall thickness as contributing factors.

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The second concern involves VY's methods to disposition nonconforming conditions. Vermont

Yankee's method is the implementation of AP 6021, Rev.15, "Nonconformance Reports,"

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wherein an immediate assessment of equipment operability required by the cognizant department

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manager, and subsequent control room notification (if required) when a nonconforming condition -

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is identified.

For deficiencies identified between design drawings and field conditions,

engineering judgement can be applied to determine if the system is indeterminate. Following

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the above two actions, a Potential Reportable Occurrence (PRO) may be generated to assess

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reportability.

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In regards to the nonconforming pipe nipples, a PRO was initiated to assess the reportability of

the event, and a Nonconformance Report - classified as significant - was initiated two days after

the condition was identified. However, the loss of design configuration was a concern, as are

the replacement of the piping under a one-for-one evaluation (vice design change) and the

prompt documentation of conditions adverse to quality. The resolution of the NCR associated

with this diesel fuel oil failure is therefore unresolved (URI 93-14-02).

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3.2

Surveillance - Core Spray Sparger Break Detection System D:iferential Pressure

Indicating Switch (DPIS)

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On July 16, VY entered a 24-hour LCO action statement due to the concurrent inoperability of

the "B" LPCI system and the "B" core spray (CS) systems. During the planned maintenance -

on the "B" LPCI system (Section 3.1.4), the inspector identified that the calibration acceptance

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criteria for the CS sparger DPISs could exceed the administrative limit established by VY as a

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corrective action for a previous surveillance concern (NRC Inspection Report 93-13). I&C -

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personnel had concurrently evaluated this, and were in the process of resolving the issue at the

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time of the inspector's review. Vermont Yankee properly evaluated the condition and changed

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the Surveillance Procedure, OP 4347, Rev.- 14, " Core Spray Header Differential Pressure

Functional / Calibration." The DPIS14-43B instrument was recalibrated to a setpoint ofless than

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4.0 psid. The 24-hour TS action statement was exited fifty minutes after it was entered. No

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reduction in plant power occurred as a result of this event. - Appropriate reports were

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subsequently made.

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4.0

ENGINEERING (71707, 90712, 90713, 92700)

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4.1

Seismic Evaluation of Installed Scaffolding

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During a review of plant Procedure AP 0019, Rev. 6, " Control of Temporary and/or Portable

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Material," the inspector questioned whether the individuals performing scaffolding inspections

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were qualified (or provided appropriate technical guidance) to assess installed scaffolding with

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respect to its impact on plant equipment during a seismic event. Currently, scaffolding

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inspections are conducted, usually by contract personnel, using AP 0019 and the VY Safety

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Manual as guides. These procedures specify installation requirements to assure personnel safety

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and the structural integrity of the scaffolding. Explicit seismic requirements, Mch as seismic

"two-over-one," operator accessibility to safety equipment, and freedom of. motion

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considerations, are not addressed. Based on recent inspections of scaffolding in the vicinity of

safety-related equipment and systems, no safety concerns have been identified. Pending VY's

review of the adequacy of the scaffolding inspections, including appropriate engineering input

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and evaluation, this issue remains unresolved (URI 93-14-03).

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4.2

Review of Written Reports

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The inspectors reviewed Licensee Event Reports (LERs) submitted to the NRC to verify that the

details of the events were clearly reported, including accuracy of the description of cause and

adequacy of corrective action. The inspectors determined whether further information was

required from the licensee, whether generic implications were indicated, and whether the event

warranted further onsite followup. The following LERs were reviewed;

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LER 93-05: Control Rod Scram times Greater Than That Required By Technical

Specifications Due to Scram Solenoid Pilot Valve Components

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Vermont Yankee identi6ed that the core-wide average scram time and the scram times of seven

of 'he 2X2 control rod arrays for the notch 46 insertion limit were not in compliance with TS

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requirements. The apparent root cause was determined to be the component parts of the scram

pilot solenoids. Vermont Yankee also committed to evaluate the root cause for using the "as

left" instead of "as found" data; reviewing TS surveillances to ensure margins to limits are

monitored and maintained; performing cold hydrostatic tests of all control rods; and, conducting

single rod scram testing following reactor startup. The inspectors had no additional comments

on this event. NRC Inspection Report 93-09 documented the NRC review and assessment of

this event.

LER 93-01, Supplement 2: Degraded Fire Barriers Due to Inadequate Documentation

of Assumptions and Inadequate Procedures

Supplement 2 documents the completion oflong term corrective actions (LTCA), which included

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the implementation of a design change to provide acceptable seal designs and an enhanced

surveillance procedure and inspection. In addition, VY extended the estimated completion date

for revising the design procedures (LTCA #2) from 8/1/93 to 9/93 and provided amplifying

information regarding their determination that all fire barriers, that had not yet been inspected,

were indeterminate and therefore inoperable. The LER adequately described the completion of

LTCAs and appropriately addressed the reporting criteria. The extension of the estimated

completion date was reasonable and did not represent a safety concern. NRC Inspection Report

93-05 documented the NRC review and assessment of the event.

The above LERs were reviewed with respect to the requirements of 10 CFR 50.73 and the

guidance provided in NUREG 1022. The LERs had good documentation of event analyses, root

cause determinations, and corrective actions. Overall quality of the LERs was good.

Periodic and Special Reports

Vermont Yankee submitted the following periodic and special reports which were reviewed for

accuracy and found to be acceptable:

Report of Failed Fuel Status and Parameter Trends for 6/30 - 7/12/1993

Report of Failed Fuel Status and Parameter Trends for 7/13 - 7/27/1993

Monthly Status of Feedwater Nozzle Temperature Monitoring, dated July 2,1993

Monthly Statistical Report for June 1993

5.0

PLANT SUPPORT (71707, 40500, 90712, 90713, 92700)

5.1

Radiological Controls

Inspectors routinely observed and reviewed radiological controls and practices during plant tours.

The inspectors observed that posting of contaminated, high airborne radiation, radiation and high

radiation areas were in accordance with administrative controls and procedures (AP-0500 series).

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High radiation doors were properly maintained and equipment and personnel were properly

surveyed exiting the radiation control area (RCA). Plant workers were observed to be cognizant

of posting requirements and maintained good housekeeping.

5.2

Inadvertent Siren Activation

Within the last five months, three inadvert :nt siren activations occurred that resulted in various

levels of public and media attention. The most recent occurred on July 16 at 12:56 a.m. and

resulted in approximately sixty calls fron the public to the plant, VY the corporate of6ce, and

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local law enforcement. Some media coverage occurred. The siren was subsequently cleared

at 2:40 a.m. Notifications were made in accordance with VY procedures and found acceptable.

The sirens are part of the VY Public Notification System (PNS).

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Vermont Yankee's evaluation and corrective actions were good. Based on the PNS maintenance

contractor's review, VY learned that the inadvertent activation and the temporary inability to

silence the alarm were caused by the premature end-of-life failure of a metal oxide varistor

(MOV) within the siren suppression circuitry (Whelen Engineering Company, model MS-864).

The investigation determined that the aging was caused by excessive voltage drop across the

MOV and a circuit modification was implemented to prevent recurrence. The modification was

also implemented in another affected siren, located in Hinsdale NH. Additionally, VY assessed

the adequacy of the siren maintenance program, the training of the individuals who disable an

activated siren, and generic implications.

A Yankee Nuclear Service: Division Quality

Assurance Surveillance (Report No. 93-05) verified that the siren vendor met VY service

requirements, that E-Plan siren maintenance and test procedures were adequate, and that

personnel have appropriate technical qualifications to perform maintenance and surveillance.

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5.3

Security

The inspector verified that security conditions met re :atory requirements and the VY Physical

Security Plan. Physical security was inspected during regular and backshift hours to verify that

controls were in accordance with the security plan and approved procedures. The inspectors

verified that access controls were effectively implemented in Gate House 2 and vehicle access

gates. Post-orders were followed and appropriate Security Shift Supervisor oversight was

observed.

During corrective maintenance on the security process computer, personnel

accountability was maintained and compensatory measures were implemented.

On June 30, VY implemented appropriate compensatory measures in response to NRC security

advisory information regarding the potential for an external threat following the bombing of the

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World Trade Center. Plant management met to discuss the plant response to the seriousness of

this advisory and promptly documented instructions to plant security personnel. The inspector

verified that these instructions were implemented. The inspector toured the facility to assure that

perimeter lighting was adequate and that the Protected Area features were in good condition.

Central and secondary alarm stations were operable, and officers were knowledgeable of the

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advisory and compensatory measures.

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Vermont Yankee performed immediate testing of the Protected Area fence following a review

of security events listed on a nuclear industry information network. Security management

concluded that the potential for a similar security vulnerability may have existed at Vermont

Yankee Nuclear Station. The applicable surveillance procedure was revised to assure the

periodic implementation of an enhanced test method. Results of the test verified that a security

vulnerability did not exist.

5.4

Overtime Controls During Outages

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A review of VY outage overtime controls was conducted. The NRC's working hour guidelines

are documented in Generic Letter (GL) 82-12, " Nuclear Power Plant Staff Working Hours,"

dated June 15, 1982. Vermont Yankee guidelines and controls for the administration of

overtime are contained in Procedure AP 0894, Rev. 2, " Shift Staffing / Overtime Limits." A

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notable exception was found to exist between the NRC's guidelines and the VY administrative

controls. Specifically, Procedure AP 0894 does not consider the NRC's working hour guidelines

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to apply (including for contract personnel) during outage conditions, unless the individual is

assigned duty in the control room. This practice would permit workers to exceed the NRC

guideline limit of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in a one week period without the expected exemptions approved by

plant management. The exemptions are available to recognize that unusual circumstances may

arise requiring deviation from the guidelines, and that granting such exemptions would be based

on continued effectiveness of operating personnel.

A review of working hour records of VY's major maintenance and modification support

contractor (Mercury Co.) during the September 1990 refueling outage indicated instances where

workers exceeded the NRC's guideline of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in a one week period. It should be noted

that the guideline limits per GL 82-12 have not been embodied in the VY Technical

Specifications.

On August 4, the Plant Manager issued an interim outage overtime policy following a VY

evaluation of industry practices, which concluded that changes in VY outage controls may be

warranted to preclude excessive overtime.

Interim guidance for Refueling Outage XVII

scheduled to begin on August 28th was included in the Outage Manual.

Additional

considerations involving overtime use and compensatory measures were contained in the policy

change memorandum. These included: (1) training that is planned for outage supervisors will

emphasize fatigue attributes to be alert for; (2) consideration for tasks that have pre-approved

authority to exceed working hour limits the use of compensatory actions, such as, increased

supervisory oversight, more frequent breaks, and assigning days off; (3) routinely discuss during

management outage meetings the topic of worker fatigue, fitness-for-duty and overtime; (4)

expand the application of overtime controls to apply to all safety-related work and not just

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safety-related, in-service equipment; and (5) applying overtime limits to the radiation protection

and chemistry technicians that are assigned accident assessment and fire brigade responsibilities.

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Finally, a VY analysis of planned RFO XVII work hours for both plant and contract personnel

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indicates that the majority of work on safety-related equipment will be performed within NRC

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guidelines. Each currently identified exception to the limits has been reviewed and authorized

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by the Plant Manager. Subsequent discussions were held between NRC Region I and VY

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management representatives about the NRC concern involving VY outage overtime policies.

The NRC Office of Nuclear Reactor Regulation is reviewing VY's overtime practices. The issue

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of VY conformance to NRC guidelines of overtime controls will remain open pending further

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NRC staff review (URI 93-14-04).

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5.5

Plant Operations Review Committee

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Plant Operations Review Committee (PORC) meetings on July 21 and 30 were observed by the

inspector to assess the quality of presentations and discussions. Topics reviewed included a set

point change for the feedwater heater level control sy stem, a proposed TS change involving EDG

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fast starts, and a number of procedure changes. The inspector concluded that these reviews and

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presentations were good. However, similar performance was not observed during the review

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of plant events, in that, PORC's review on July 21 of the EDG fuel oil nipple failure (Section

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3.1.6) was superficial. A discussion regarding the potential for common cause failure on the

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opposite diesel, the preliminary root cause determination, the staff's preliminary identification

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of nonconforming conditions did not occur. This represented a missed opportunity on pan of

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PORC to independently assess a safety-related equipment failure.

This observation was

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acknowledged by the PORC chairperson and discussed by plant management at a department

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head meeting. The PORC on July 30 was effective during the review of plant events, self-

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identified concerns regarding EDG room fire protection, and a modification for the temporary

installation of a diesel powered station air compressor.

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6.0

Administrative

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6.1

Preliminary Inspection Findings

Meetings were held periodically with VY management during this inspection to discuss

inspection findings. A summary of preliminary findings was also discussed at the conclusion

of the inspection in an exit meeting held on August 6. No proprietary information was identified

as being included in this report.

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