ML20203G048

Insp Rept 50-482/86-16 on 860601-0705.Violation Noted: Failure to Adequately Isolate safety-related Equipment Prior to Maint

ML20203G048
Person / Time
Site:	Wolf Creek
Issue date:	07/24/1986
From:	Bruce Bartlett, Cummins J, Hunnicutt D, Hunter D, Greg Pick NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20203G039	List: ML20203G035 ML20203G042 ML20203G048
References
50-482-86-16, NUDOCS 8607310247
Download: ML20203G048 (16)

See also: IR 05000482/1986016

Text

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APPENDIX B

US NUCLEAR REGULATORY COMMISSION

REGION IV

NRC Inspection Report: 50-482/86-16 LP: NPF-42

Docket: 50-482

Licensce: Kansas Gas and Electric Company (KG&E)

Post Office Box 208

Wichita, Kansas 67201

Facility Name: Wolf Creek Generating Station (WCGS)

Inspection At: Wolf Creek Site, Coffey County, Burlington, Kansas

Inspection Conducted: June 1 to July 5, 1986

Inspectors: ,

J. E. Cummins, Senior Resident Inspector,

7////BI

Date

l Operations, (pars. 2, 3, 4, 5, 6, 7, and 8)

oW A M//!7

~ B. L. Bartlett, Resident Reactor Inspector, ' Dd te

Operations, (pars. 2, 3, 4, 5, 6, 7, and 8)

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bN nY

D. M. Hunnicutt, Chief, Operations Section 7 Dat(

9/8$>

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Reactor Safety Branch (pars. 9 and 10)

G. A. P ck,

A%

sactor Inspector

Ld D!U

'Dath

b

(pars. 9 1 10)

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Approved: I

tM

D. R. ifunter, Chief, Project Section B, Date

Reactor Projects Branch 2

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8607310247 860725 2

PDR ADOCK 0500

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Inspection Summary

Inspection Conducted June 1 to July 5, 1986 (Report 50-482/86-16)

Areas Inspected: Routine, unannounced inspection including plant status;

followup on previously identified NRC items; operational safety verification;

engineered safety features system walkdown; onsite followup of events; monthly

surveillance observation; monthly maintenance observation; plant safety review

' committee; and nuclear safety review committee.

Results: Within the nine areas inspected, one violation was identified

(failure to adequately isolate safety-related equipment prior to maintenance,

paragraph 8).

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DETAILS

, 1. Persons Contacted

Principal Licensee Personnel

j *G. L. Koester, Vice President-Nuclear

! *J. A. Bailey, Interim Site Director

F. T. Rhodes, Plant Manager

• G. D. Boyer, Deputy Plant Manager

• M. Estes, Superintendent of Operations

• M. D. Rich, Superintendent of Maintenance

• M. G. Williams, Superintendent of Regulatory, Quality, and

1 Administrative Services

0. L. Maynard, Manager Licensing

• K. Peterson, Licensing

i. *G. Pendergrass, Licensing

i *W. M. Lindsay, Supervisor Quality Systems

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• C. J. Hoch, QA Technologist

• W. J. Rudolph, QA Manager-WCGS

! *A. A. Freitag, Manager Nuclear Plant Engineering-WCGS

M. Megehee, Compliance Engineer

. *R. Smith, Nuclear Information Supervisor

• D. Walsh, Maintenance Services Supervisor

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• B. McKinney, Superintendent of Technical Support

,

The NRC inspectors also contacted other members of the licensee's staff

during the inspection period to discuss identified issues.

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• Denotes those personnel in attendance at the exit meeting held on

July 8, 1986.

2. Plant Status

The plant operated in Mode 1 during this inspection period except during

the time periods described below:

o June 4 to 11, 1986 - Plant in Mode 3 (hot standby) for investigation

,

and maintenance related to limitorque electrical termination lugs.

o On June 30, 1986, the reactor tripped from 100 percent power because

of 10-10 steam generator water level. During the subsequent startup

on July 1,1986, the reactor tripp d at approximately 20 percent

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power because of another 10-10 Steam Generator Level. The plant

}

was returned to Mode 1 on July 1, 1986,

i 3. Followup on Previoutl y Identified NRC Items

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(Closed) Violation (482/8601-01): Failure To Comply With Licensee's

i Temporary Modification Procedure

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The licensee discussed the temporary modification requirements with plant

supervisory personnel and issued a Quality Systems Engineering

Directive (86-01) alerting cognizant personnel to the possible need for

adding instructions to work request quality control inspection

instructions for restoration of work areas. The NRC inspectors have not

identified any additional occurrences of this type during subsequent plant

tours. This violation is closed.

(Closed) Violation (482/8604-01): Failure To Comply With Surveillance

Procedure Instructions

The NRC inspector reviewed Surveillance Test Procedure STS SF-001,

Revision 4, " Control and Shutdown Rod Operability Verification," and

verified that Section 2.2 had been changed to relax the precaution from no

change in reactor coolant system (RCS) baron concentration during

perfonaance of the test, to minimizing changes in RCS boron concentration

during the performance of the test. The NRC inspector verified by review

of required reading logs that operations personnel had initialed that they

had read Wolf Creek Event Report 86-17, which discussed this event and the

necessity for observing " precautions and limitations" during procedure

performance. This violation is closed.

(Closed) Violation (482/8604-02): Failure to Maintain Control Room

HVAC in Required State of Operability

The NRC inspector verified that signs requiring control room notification

prior to operating had been installed adjacent to the control room HVAC

control switches. The NRC inspector verified by review of required

reading logs that operations personnel had initialed that they had read

Licensee Event Report 86-006-00, which discussed this event. This

violation is closed.

(Closed) Violation (482/8604-03): Fireproofing Removed From Support Beam

The NRC inspector verified that a knowledgeable individual has been

assigned the responsibility of overseeing the removal and replacement of

fireproofing material, and verified by random sampling that no other

violations of this type have been identified. This violation is closed.

(Closed) Violation (482/8604-04): Out-of-Date Procedures

The NRC inspector performed surveillances of randomly selected alarm

response procedures located at remote alarm panels and no out-of-date

procedures were identified. This violation is closed.

4. Operational Safety Verification

The NRC inspectors verified that the facility is being operated safely and

in conformance with regulatory requirements by direct observation of

licensee facilities, tours of the facility, interviews and discussions

with licensee personnel, independent verification of safety system status

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and limiting conditions for operations, and by reviewing facility records.

The NRC inspectors, by observation, interview of personnel, and review of

documents, verified the physical security plan was being

implemented in accordance with the security plan and that radiation

protection activities were controlled.

By observing valve position, electrical breaker position, and control room

indication, the NRC inspectors confirmed the operability of the Safety

Injection System, the Coolant Charging System, and the 4160 Volt AC Class

1E. System. The NRC inspectors also visually inspected safety components

for leakage, physical damage, and other impairments that could prevent

them from performing their designed function.

No violations or deviations were identified.

5. Engineered Safety Features (ESF) System Walkdown

"The NRC inspectors verified the operability of ESF systems by walking down

selected accessible portions of the systems. The NRC inspectors verified

valves and electrical circuit breakers were in the required position,

power was available, and valves were locked where required. The NRC

inspectors also inspected system components for damage or other conditions

that could degrade system performance.

The ESF systems walked down during this inspection period and the

documents utilized by the NRC inspectors during the walkdown are listed

below:

System Documents

Auxiliary Feedwater System (AL) Drawing M-12AL01, Revision 0,

Auxiliary Feedwater System P&ID(Q),

Drawing M-12FC02, Revision 1, Auxiliary

Feedwater Pump Turbine P&ID(Q),

Checklist CKL AL-120, Revision 7,

Auxiliary Feedwater Normal Lineup,

Surveillance Tests, STS IC-706,

Revision 1, " Steam Generator N/R Level

Transmitter Response Time

Test-Protection Set I,"

STS IC-260, Revision 3, " Analog Channel

Operational Test Auxiliary Feedwater

Pump Suction Pressure Low Transfer To

ESW,"

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STS IC-560, Revision 2, " Channel

Calibration-Auxiliary Feedwater Pump

Low Suction Pressure (Transfer To ESW)"

STS IC-505B, Revision 1, " Channel

Calibration Steam Generator Narrow

Range Level,"

STS IC-505A, Revision 1, " Calibration

Of Steam Generator Narrow-Range Level

Transmitters,"

Class 1E 125 V DC System (NK) Drawing E-01 NK01(Q), Revision 15,

" Class 1E 125 V DC System Meter &

Relay Diagram"

Drawing E-01 NK02(Q) Revision 17,

" Class 1E 125 V DC System Meter &

Relay System"

Checklist CKL NK-131, Revision 1, "NK

Distribution Switchboard Switch

Lineup Checklist"

Class 1E 120 V AC System (NN) Drawing E-03 NN01(Q), Revision 12,

" Class 1E Instrument AC Schematic"

Checklist CKL NN-131, Revision 4,

" Instrument AC Power (Class 1E)

Switchboard Breaker Checklist"

Selected NRC inspector observations are discussed below:

a. The following discrepancies related to Checklist CKL NK-131, Revision

1, "NK Distribution Switchboard Switch Lineup Checklist" were

identified:

o On page 2 of 9 Switch NK 4103 was incorrectly described as an

active circuit; however, it was a spare as identified on

Switchboard NK41 and in note 6 of Drawing E-ul NK02,

Revision 17, " Class 1E 125 V DC System Meter and Relay Diagram."

o On page 8 of 9 Switches NK5402 and NK5408 were incorrectly

described as active circuits; however, they were spares as

identified on Switchboard NK54 and in note 9 of

Drawing E-01 NK02, Revision 17, " Class 1E 125 V DC System Meter

and Relay Diagram."

o On page 7 of 9 the description of Switch NK4410 was incorrect.

It should have read "MCB CONT PNL RLO25 and RLO26," rather than

"MCB CONT PNL RLO25 and RLO24."

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o On page 2 of 9 the description of Switch NK4112 was incorrect.

It should have read "MCB CONT PNL RL017 and RL018" rather than

"MCB CONT PNL RL017 and RL013."

! o The information plate for Switch NK4408 on Switchboard NK44

should have read " Fused DC Dist PNL RP316," rather than " Fused

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DC Dist PNL RP3116."

b. The following discrepancies related to Checklist CKL NN-131, Revision

4, " Instrument AC Power (Class 1E) Switchboard Breaker Checklist were

<

identified."

o On Switchboard NN01 the label plate for Switch No.11 was

labeled " Panel SB03B," it should be labeled " Panel SB038" as

shown on Drawing E-03 NN01(Q), Revision 12 " Class 1E Instrument

AC Schematic."

o On page 4 of 7 procedure identified Switch 13 on

Switchboard NN02 as " Panel SB054B," it should be " Panel SE054B"

as shown on Drawing E-03 NN01(Q), Revision 12, " Class 1E

Instrument AC Schematic."

o Switch 6 on Switchboard NN04 was being used to feed the " Neutron

Flux Monitoring Signal Amplifier SENY61A;" however, it was not

listed in Procedure CKL NN-131. Drawing E-03 NN01(Q),

i Revision 12, " Class 1E Instrument AC Schematic" shows that the

! switch was put in use by Design Change Package DCP-CS-63.

o Switch 8 on Switchboard NN04 was being used to feed " Neutron

Flux Monitor Processor SENY61B;" however, it was not listed in

Procedure CKL NN-131. Drawing E-03 NN01(Q), Revision 12, " Class

IE Instrument AC Schematic" showed that the switch was put in

use by Design Change Package DCP-CS-63.

The deficiencies described in a. and b. above will be an unresolved

item (482/8616-02) pending further review by the NRC inspector during

j a subsequent inspection to determine the reason for the deficiencies

and if corrective action is required.

No violations or deviations were identified.

6. Onsite Followup of Events

a. events

The NRC inspector

that occurred duringperformed

this report onsite

period.followup

The NRCof inspector

non-emergency (when

available) observed control room personnel response, observed

instrumentation indicators of reactor plant parameters, reviewed logs

and computer printouts, and discussed the event with cognizant

personnel. The NRC inspector verified the licensee had responded to

the event in accordance with procedures and had notified the NRC and

other agencies as required in a timely fashion.

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Engineered safety feature actuations that occurred during the report

period are listed in the table below. Where applicable, the NRC

inspector will review the LER for each of these events and will

report any findings in subsequent NRC inspection reports.

Date Event * Plant Status Cause

6-4-86 FWIA Mode 2 being High Water Level

shut down SG C

6-18-86 ESF Components Mode 1 **ESF Circuit Wire

Actuated (100%) Shorted During

Maintenance On Limi-

torque Valve EN HV-15

6-26-86 ESF Actuation Mode 1 Loss Of Power To ESF

(100%) Buss NB01 Due To Dis-

connect Opening in

Switchyard

6-30-86 Rx Trip Mode 1 Lo-Lo SG Water

(100%) Level-Condensate

Pumps Tripped Off

7-1-86 Rx Trip Mode 1 Lo-Lo SG Water

(100%) Level

7-2-86 CRVIS Mode 1 Loss Of Control Room

(100%) Radiation Monitors

• Event

CRVIS - control room ventilation isolation signal

FWIA - feedwater isolation actuation

Rx Trip - reactor trip

• • See paragraph 8 for details

Selected NRC inspector observations are discussed below:

o The ESF actuation on June 26, 1986, was caused by a loss of

power to ESF 4160 V AC electrical buss, NB01, due to a

disconnect in the switchyard opening. The emergency diesel

generator automatically started and picked up safety loads that

sequenced onto the bus per design. The disconnect in the

switchyard was reclosed and the affected electrical system

returned to a normal line up,

o The reactor trip that occurred on June 30, 1986, was due to a

10-10 steam generator water level. The 10-10 steam generator

water level condition was the result of all three condensate

pumps tripping off due to maintenance work being performed on

the condenser water level instrumentation.

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o The reactor trip that occurred on July 1, 1986, was due to a

10-10 steam generator water level condition. The event occurred

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.during plant startup at approximately 20 percent power, while

plant personnel were transferring steam generator feedwater flow

from the main feedwater regulator bypass valves to the main

feedwater regulator valves. The subsequent startup was

accomplished with no significant problems.

b. During this report period, the NRC inspectors reviewed licensee event

reports and defect / deficiency reports (DDR).

A DDR is an internal licensee report on defects and/or deficiencies

found in design or construction. These reports are issued to

appropriate personnel and agencies.

Wolf Creek event reports are internal documentation of events that

are not specifically reportable as a defect / deficiency report, but

are events which plant management should be made aware of.

Wolf Creek event reports reviewed this report period:

No. Title

86-039 " Improper Safety Class Marked on Work Request (WR)"86-047 " Overfilling the HF Evaporator"86-050 "No Flow At S/G Blowdown Effluent Radiation Monitor"

Defect / deficiency reports reviewed this report period:

No. Title

85-100 " Jacket Water Heaters For Standby Diesel Generators

Each Had One Cable Burnt. . . ."86-038 "GH RE-10A Automatically Isolates During Testing Of

"

GH RE-108 and Does Not Automatically Reset. ...86-046 ". . . Valve EF HV-97 Was Found Shut With Valve

Power Removed."86-052 ". . . Impairment 86-075 Was Not Checked By the

Responsible Officer. . . ."

The NRC inspector had the following comments:

o DDR 86-038 stated "On May 15, 1986, based on a conversation with

Callaway personnel, it was discovered that GH RE-10A

automatically isolates during testing of GH RE-108 and does not

automatically reset upon completion of GH RE-10B testing.

GH RE-10A is receiving flow from the building atmosphere while

it is isolated and therefore, no RM-11 indication of a problem

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exists,-although, a local alarm is present." GH RE-10A was the

radwaste building vent particulate and iodine radiation monitor

and GH RE-10B was the radwaste building vent noble gas monitor.

,

This condition could result in venting of a waste gas decay tank

with GH RE-10A out-of-service. The licensee subsequently

received and_ reviewed Callaway Plant LER 50-483/86-011-00 for

applicability ;to Wolf Creek Generating Station. The licensee's

review of this LER showed that instrumentation and control (I&C)

calibration procedures ensured that GH RE-10A was properly

restored to service after testing of GH RE-10B was completed and

a review of the past releases did not find any instances of GH

RE-10A being found isolated. A temporary procedure change was

made to ensure personnel understood the need to reset GH RE-10A

after testing of GH RE-10B and an information tag was hung in

the control room to ensure operators were aware of this

situation. In response to NRC inspector concerns on this

subject, the licensee has submitted an engineering evaluation

request (EER) to nuclear plant engineering (NPE) to see if

adding a separate annunciator in the control room for GH RE-10A

or adding an automatic reset to GH RE-10A would be desirable.

The completion of the EER and any related modifications was not

required by regulations and thus will not be tracked as an

inspector followup item or a commitment.

No violations or deviations were identified.

7. Monthly Surveillance Observation

The NRC inspectors observed selected portions of the performance of

surveillance testing and/or reviewed completed surveillance test

procedures to verify that surveillance activities were performed in

accordance with TS requirements and administrative procedures. The NRC

inspectors considered the following elements while inspecting surveillance

activities:

s o Testing was being accomplished by qualified personnel in accordance-

with an approved procedure.

o The surveillance procedure conformed to TS requirements.

o Required test instrumentation was calibrated.

o Technical Specification limiting conditions for operation (LCO) were

satisfied.

o Test data was accurate and complete. Where appropriate, the NRC

inspectors performed independent calculations of selected test data

to verify their accuracy.

o The performance of the surveillance procedure conformed to applicable

administrative procedures.

o The surveillance was performed within the required frequency and the

test results met the required limits.

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Surveillances witnessed and/or reviewed by the NRC inspectors are listed

below:

o STS RE-002, Revision 3, " Determination Of Estimated Critical

Position" performed June 11, 1986

o STS RE-002, Revision 3, " Determination Of Estimated Critical

Position" performed July 1, 1986

o STS RE-004, Revision 4, " Shutdown Margin Determination" performed

July 1, 1986

o STS PE-013, Revision 5, " Personnel Air'ock Seal Test," performed

June 18, 1986

o STS AL-102, Revision 3, " Motor Driven Aux FW Pump iB' Inservice

Pump Test," performed June 18, 1986

o STS RE-012, Revision 1, " Quadrant Power Tilt Ratio," performed

June 18, 1986

Selected NRC inspector observations are discussed below:

On June 11, 1986, the licensee notified the NRC inspectors that a

violation of TS Surveillance 4.6.3.1 had occurred in that this TS requires

that a post maintenance cycling test be performed on any containment

isolation valve listed in TS Table 3.6.1. Maintenance was performed on

the valves listed below, and the licensee determined a continuity test was

to be performed to verify operability prior to returning the valves to

operable status. The licensee subsequently identified the TS oversight

and performed cycling tests on the valves. The table below notes the

dates when r.aintenance was completed on each valve and when the cycling

test was performed. All the valves passed the subsequent cycling test.

Valve Date/ Time Date/ Time

No. Description Work Completed Stroke Tested

LF FV 095 RW Bldg. Demin Vault Floor

Dr. Lower Iso. 6-05-86/2358 6-09-86/2050

GS HV 020 Hydrogen Purge Inner

Containment Iso. 6-06-86/0430 6-06-86/0430

GS HV 021 Hydrogen Purge Outer

Containment Iso. 6-12-86/0810 6-12-86/1038

KC HV 253 Turb. Bldg. Hose Station Iso.

Elevatf ori 2051' 6-08-86/0813 6-08-86/1700

EG HV 058 CCW to RCS Iso. 6-07-86/2200 6-09-86/2142

EG HV 059 CCW Return From RCS Iso. 6-07-86/2200 6-09-86/2142

EG HV 060 CCW Return From RCS Iso. 6-07-86/1600 6-09-86/2050

EG HV 061 CCW Return From RCS Iso. 6-07-86/2322 6-08-86/0215

EG HV 062 CCW Return From RCS Iso. 6-07-86/2115 6-09-86/2050

EG HV 127 CCW To RCS HV-58 & HV-71

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Bypass Iso. 6-08-86/0455 6-09-86/2142

EG HV 130 RCS CCW Return HV-60

Bypass Iso. 6-06-86/2150 6-09-86/2050

EG HV 131 RCS CCW Return HV-59

Bypass Iso. 6-09-86/2103 6-09-86/2142

EG HV 132 . Thermal Barrier CCW Return

HV-62 Bypass Iso. 6-06-86/2150 6-09-86/2050

EG HV 133 Thermal Barrier CCW Return

HV-61 Bypass Iso. 6-08-86/1300 6-08-86/1325

EF HV 031 ESW A to CTMT Air Coolers 6-17-86/1100 6-17-86/1125

EF HV 032 ESW B to CTMT Air Coolers 6-13-86/2513 6-13-86/2315

EF HV 033 ESW A to CTMT Air Coolers 6-05-86/0236 6-05-86/1729

EF HV 034 ESW B to CTMT Air Coolers 6-07-86/1840 6-07-86/1945

EF HV 045 ESW A From CTMT Air Coolers 6-07-86/0445 6-07-86/0445

EF HV 046 ESW B From CTMT Air Coolers 6-07-86/1840 6-07-86/2040

EF HV 047 ESW A From CTMT Air Coolers

Bypass 6-13-86/0650 6-13-86/0650

EF HV 048 ESW A From CTMT Air Coolers

Bypass 6-13-86/2200 6-13-86/2222

EF HV 049 ESW A From CTMT Air Coolers 6-16-86/1700 6-16-86/1733

EF HV 050 ESW B From CTMT Air Coolers 6-13-86/2300 6-13-86/2306

EN HV 006 CTMT Spray Train A Disch.

Iso. Valve 6-09-86/0514 6-09-86/2050

EN HV 012 CTMT Spray Train B Disch.

Iso. Valve 6-08-86/1510 6-09-86/2050

EM HV 8801A BIT Outlet Iso. 6-08-86/0238 6-09-86/2050

EM HV 8801B BIT Outlet Iso. 6-08-86/0545 6-09-86/2050

BG HV 8100 Seal Water Return CTMT

Iso. Valve 6-09-86/1900 6-10-86/2045

BG HV 8112 Seal Water Return CTMT

Iso. Valve 6-06-86/0738 6-19-86/2045

BG HV 8105 Charging Pumps to Regen.

HV/CTMT Iso. 6-06-86/2155 6-10-86/0457

BB HV 8351A RCP A Seal Water Supply 6-08-86/1708 6-10-86/0012

BB HV 8351B RCP B Seal Water Supply 6-08-86/1750 6-10-86/0012

BB HV 8351C RCP C Seal Water Supply 6-08-86/1732 6-10-86/0012

BB HV 83510 RCP D Seal Water Supply 6-08-86/1850 6-10-86/0012

BB PV 8702A RCS Hot Leg 1 to RHR

Pump A Suction 6-07-86/1500 6-11-86/0141

BB PV 8702B RCS Hot Leg 4 to RHR

Pump B Suction 6-07-86/1500 6-11-86/0213

EJ HB 8701A RCS Hot Leg 1 to RHR

Pump A Suction 6-05-86/1725 6-11-86/0141

EJ HV B701B RCS Hot Leg 4 to RHR

Pump B Suction 6-06-86/1500 6-11-86/0213

EJ HB 8809A RHR to ACCUM Inj.

Loop 1 & 2 Iso. Valve 6-06-86/2055 6-09-86/2215

EJ HV 8809B RHR to ACCUM Inj

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Loops 3 & 4 Iso. Valve 6-09-86/0904 6-09-86/2215

EJ HB 8840 RHR/SI Hot Leg Recirc

Iso. Valve 6-08-86/2312 6-09-86/2215

EM HV 8802A SI Pump A Discharge Hot Leg

Iso. Valve 6-08-86/0049 6-09-86/2115

EM HV 8802B SI Pump B Discharge Hot Leg

Iso. Valve 6-09-86/1219 6-09-86/2115

EM HV 8835 Safety Inj Cold Leg

Iso. Valve 6-09-86/1052 6-09-86/2115

We have no further questions of this matter at this time. Further review

of this matter will be performed during the subsequent review of the

licensee corrective action report and the event report.

No violations or deviations were identified.

8. Monthly Maintenance Observation

The NRC inspector observed maintenance activities performed on

safety-related systems and components to verify that these activities were

conducted in accordance with approved procedures, Technical

Specifications, and applicable industry codes and standards. The

following elements were considered by the NRC inspector during the

observation and/or review of the maintenance activities:

o LCOs were met and, where applicable, redundant components were

operable,

o Activities complied with adequate administrative controls.

o Where required, adequate, approved, and up-to-date procedures were

used.

o Craftsmen were qualified to accomplish the designated task and

technical expertise (i.e., engineering, health physics, operations)

was made available when appropriate.

o Replacement parts and materials being used were properly certified.

o Required radiological controls were implemented.

o Fire prevention controls were implemented where appropriate.

o Required alignments and surveillances to verify post maintenance

operability were performed.

o Quality control hold points and/or checklists were used when

appropriate and quality control personnel observed designated work

activities.

, Selected portions of the maintenance activities accomplished on the WR

listed below were observed and related documentation reviewed by the NRC

inspector:

No. Title

i WR 02601-86 " Rework Limitorque Operator To Valve EM HV-8814B"

!

WR 02737-86 "EJ FCV-0611 Actuator"

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Rework Limitorque Operator To Valve AL HV-009"

WR 02513-86

WR 02514-85 "

Rework Limitorque Operator To Valve AL HV-011"

"

WR 00192-86 DCGM01B Breakee Trips Frequently"

"

WR 02778-86 KJ HV-002"

"

WR 02777-86 EF PDV-20"

"

WR 02776-86 EF HV-98"

WR 02399-86 " Alternator NE02 Failed Oil Analysis"

WR 02544-86 "EF HV-025, Vendor Installed Lugs and Wire Quality

Are Indeterminate"

WR 07073-85 "EN HV-015, Maintenance On Limitorque Operator"

Selected NRC inspector observations are discussed below:

While performing maintenance on Limitorque Valve EN HV-015 in accordance

with WR 07073-85, the electrician shorted electrical wires to ground

causing the actuation of the following engineered safety features

components:

o LOCA sequencer actuated.

o Containment spray pump "A'" started,

o Essential service water / service water crossconnect valves closed.

The wires that were shorted were part of the solid state protection system

slave relay test circuit. The electrician was not aware that the leads

had not been de-energized. TS 6.8.1 requires that, " Written procedures

shall be established, implemented, and maintained covering . . . a. The

applicable procedures recommended in Appendix A of Regulatory Guide

(RG) 1.33, Revision 2, February 1978." Section 9.a of RG 1.33, Revision

2, February 1978, states that, ". . . maintenance that can affect the

performance of safety-related equipment should be properly preplanned and

performed in accordance with written procedures, docunented instructions, '

or drawings appropriate to the circumstances." This unanticipated

actuation of ESF components due to the failure to adequately preplan

(isolate) a maintenance activity as required by the above documents is an

apparent violation (482/8616-01).

9. Plant Safety Review Committee (PSRC) j

!

The NRC inspector reviewed Section 6.5.1 of the TS regarding the onsite

review committee, PSRC. This review determined that the PSRC met

regulatory guidelines and commitments.

,

-~ ~- ~-- n--, , , , _ _ _ _ ,,____ , _ _ , , _ _ _ _ _ _ , _ , _ _ , _

,.

-15-

The PSRC reviews and analyzes defect / deficiency reports to detemine

whether a licensee event report should be submitted and what course of

action should be initiated to correct TS violations.

The PSRC reviewed the safety evaluations related to design changes, major

procedure revisions, or plant tests / experiments that reflected a change to

the TS or a condition not analyzed in the Final Safety Analysis Report

(FSAR) that resulted from 10 CFR Part 50.59 reviews.

The PSRC reviewed current unit operations through their considerations for

approval of " Jump-up" items such as temporary change forms regarding shif t

logs, surveillances, tests, and so forth.

Plant procedure revisions were also reviewed by the PSRC.

No violations or deviations were identified.

10. Nuclear Safety Review Comittee (NSRC)

The NRC inspector reviewed Section 6.5.2 of the licensee's TS regarding

the offsite review comittee, NSRC. The NRC inspector determined that the

NSRC met their comitments to regulatory guidance, in the areas of

composition, duties, and responsibilities.

The NSRC audit subcomittee reviewed selected audits that related to the

requirements listed in the TS. The review consisted of examining

completed reports of audits that were conducted by the quality assurance

(QA) department. The audits were scheduled such that the essential

elements - those elements necessary to meet regulatory requirements -

were met and reviewed at the required frequency. The NSRC committee

remained cognizant of their required audit activities by assigning a

comittee member with overview responsibility. The audit subcomittee

reported to the full NSRC.

The NRC inspector reviewed the NSRC's meeting minutes from March 1985

through April 1986. From the review of the NSRC meeting minutes, the

inspector determined that the NSRC comittee retained proper cognizance

through detailed subcomittee reviews in the areas of plant modifications

and violations and reportable events. The NSRC also reviewed the onsite

review comittee minutes and any significant operatirg abnormalities or

system structure and/or component design deficiencies.

No violations or deviations were identified.

11. Exit Meeting

The NRC inspectors met with licensee personnel to discuss the scope and

findings of this inspection on July 8, 1986. The NRC inspectors also

attended entrance / exit meetings of other NRC region based inspectors

identified below:

c

,. .

-16-

Inspection Lead Area Inspection

Period Inspector Inspected Report No.

b'

6-2/5-86 -

J. Whittemore Operator Licensing

Exams OL 86-01

,

6-16/20-86 J. Boardman Maintenance 50-482/86-15

_

6-16/20-86 D. Hunnicutt Review Committee 50-482/86-16

, , (pars. 9 & 10)

During the week of. June 16 through 20, 1986, the NRC Project Manager for

, WCGS, P. O'Connor, visited the site and held discussions with licensee

personnel.

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