ML20056G973
| 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: |
| References | |
| 50-271-93-14, NUDOCS 9309080019 | |
| Download: ML20056G973 (15) | |
See also: IR 05000271/1993014
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U.S. NUCLEAR REGULATORY COMMISSION !
REGION I l
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Report No. 93-14 l
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Docket No. 50-271 i
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Licensee No. DPR-28 l
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Licensee: Vermont Yankee Nuclear Power Corporation i
RD 5, Box 169 I
Ferry Road l
Brattleboro, VT 05301 :
Facility: Vermont Yankee Nuclear Power Station f
Vernon, Vermont j
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Inspection Period: June 27 - August 7,1993 l
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Inspectors: Harold Eichenholz, Senior Resident Inspector j
Paul W. Harris, Resident inspector j
Approved by: M/ M ~4 '
JPAt/rs
Egene M. Kelly, Chief ' Da'te ' l
Keactor Projects Section )
Scope: Station activities inspected by the resident staff this period included: operations, l
maintenance, engineering and plant support. An initiative selected for 'this i
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 !
Executive Summary. A violation of administrative controls involving two' i
examples of failure to follow procedures by contract personnel was identified - l
(Section 3.1.1). Unresolved items were opened regarding VY engineering _
evaluation of seismic considerations for scaffolding installed near plant equipment I
(Section 4.1), design controls for a diesel generator fuel oil piping failure (Section i
3.1.6), and overtime controls during plant outage periods (Section 5.5). ;
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93090B0019 930827' I
PDR ADOCK 05000271 i
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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 j
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EXECUTIVE SUMMARY ...................................... ii
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TABLE OF CONTENTS .......................................iii ;
1.0 SUMMARY OF FACILITY ACTIVITIES ........................ 1 l
2.0 OPERATIONS (71707, 93702, 92700) . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 ,
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2.1 Operational Safety Verification . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
2.2 Operator Awareness of Plant Maintenance . . . . . . . . . . . . . . . . . . . . I !
2.3 Conduct of Operator Rounds ............................ I !
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2.4 Electrical Storm . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
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3.0 MAINTENANCE (62703, 61726, 92700, 90712, 93702) ............... 3 i
3.1 Maintenance ...................................... 3
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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 !
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) . . . . . ............... 8
4.1 Seismic Evaluation of Installed Scaffolding . . . . . . . . . . . . . . . . . . . . 8
4.2 Review of Written Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
5.0 PLANT SUPPORT (71707, 40500, 90712, 90713, 92700) ............., 9
5.1 Radiological Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 I
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5.2 Inadvertent Siren Activation . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
5.3 S ecu ri ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 4
5.4 Overtime Controls During Outages . . . . . . . . . . . . . . . . . . . . . . . . 11
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5.5 Plant Operations Review Committee . . . . . . . . . . . . . . . . . . . . . . . 12 >
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6.0 A d ministrative . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
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6.1 Preliminary Inspection Findings ......................... 12
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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
report section. l
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! DETAILS ,
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. i
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. j
2.2 Operator Awareness of Plant Maintenance l
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l 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
l articulated the scope and schedule for the maintenance; and (2) following a July 30th electrical
! 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,
l 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
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. The
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.
3.0 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 ,
Plexiglass Covering
On July 23, the inspector observed that a temporary plexiglass I anel was installed over the
l 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 !
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
l 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. The
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. The
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. i
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 -
LPCI and the "B" residual heat removal service water (RHRSW) sub-systems. Maintenance on i
the emergency core cooling system (ECCS) sub-systems was performed concurrently and lasted
j 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. An
inspection was also conducted by VY of the RHRSW-898 motor operator to assess the condition !
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
! a chamfered bushing in the mechanical seal on the RHRSW "B" pump. The original seal was
l replaced due to end-of-life performance concerns and the replacement seal leaked excessively
l 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.
! 3.1.5 Vernon Tie Line Inoperability
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l 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 l
lightning strike to a transmission pole, based upon the identification of electrical arc marks, I
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. l
3.1.6 Diesel Generator Fuel Oil Irak i
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During the monthly surveillance of the "A" emergency diesel generator (EDG), VY identified l
a fuel oil (FO) leak in a 1/2-inch pipe nipple on the suction side of the FO pump, and entered !
the 7-day TS LCO. An inspection of the "B" EDG FO piping identified no similar deficiency, i
but only because a similar problem had developed on that engine in 1989 and was repaired at .
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 l
to be a valid test failure. l
Two concerns were identified regarding this event. The first involved the identification by VY f
that, prior to this event, design configuration control for this piping had not been maintained. i
Based on their review of maintenance records, consultation with the vendor, and performance l
of ultrasonic testing, VY identified that incorrect schedule piping was installed at the suction of l
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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 i
nipples were installed accordingly. The preliminary root cause developed by VY engineering i
based on metallurgical evaluations was attributed to fatigue failure with pipe loading, surface l
quality, and wall thickness as contributing factors. i
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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," j
wherein an immediate assessment of equipment operability required by the cognizant department !
manager, and subsequent control room notification (if required) when a nonconforming condition - ,
is identified. For deficiencies identified between design drawings and field conditions, !
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engineering judgement can be applied to determine if the system is indeterminate. Following j
the above two actions, a Potential Reportable Occurrence (PRO) may be generated to assess {
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reportability.
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) '
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 i
criteria for the CS sparger DPISs could exceed the administrative limit established by VY as a j
corrective action for a previous surveillance concern (NRC Inspection Report 93-13). I&C - l
personnel had concurrently evaluated this, and were in the process of resolving the issue at the j
time of the inspector's review. Vermont Yankee properly evaluated the condition and changed j
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 .I
4.0 psid. The 24-hour TS action statement was exited fifty minutes after it was entered. No i
reduction in plant power occurred as a result of this event. - Appropriate reports were .]
subsequently made. !
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4.0 ENGINEERING (71707, 90712, 90713, 92700)
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l 4.1 Seismic Evaluation of Installed Scaffolding l
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l During a review of plant Procedure AP 0019, Rev. 6, " Control of Temporary and/or Portable I
! 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 !
l respect to its impact on plant equipment during a seismic event. Currently, scaffolding ;
l inspections are conducted, usually by contract personnel, using AP 0019 and the VY Safety i
Manual as guides. These procedures specify installation requirements to assure personnel safety l
and the structural integrity of the scaffolding. Explicit seismic requirements, Mch as seismic l
"two-over-one," operator accessibility to safety equipment, and freedom of. motion i
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 i
and evaluation, this issue remains unresolved (URI 93-14-03). I
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4.2 Review of Written Reports
i 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;
e 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 ,
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 i
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
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resulted in approximately sixty calls fron the public to the plant, VY the corporate of6ce, and
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
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personnel have appropriate technical qualifications to perform maintenance and surveillance.
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 l
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
l 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 '
notable exception was found to exist between the NRC's guidelines and the VY administrative
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controls. Specifically, Procedure AP 0894 does not consider the NRC's working hour guidelines
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 i
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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 j
indicates that the majority of work on safety-related equipment will be performed within NRC l
guidelines. Each currently identified exception to the limits has been reviewed and authorized l
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. by the Plant Manager. Subsequent discussions were held between NRC Region I and VY
- 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 l
NRC staff review (URI 93-14-04). !
) 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 j
fast starts, and a number of procedure changes. The inspector concluded that these reviews and !
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
- 3.1.6) was superficial. A discussion regarding the potential for common cause failure on the
opposite diesel, the preliminary root cause determination, the staff's preliminary identification ,
of nonconforming conditions did not occur. This represented a missed opportunity on pan of ;
i PORC to independently assess a safety-related equipment failure. This observation was j
1 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- j
identified concerns regarding EDG room fire protection, and a modification for the temporary ;
, installation of a diesel powered station air compressor.
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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