Text

Insp Rept 50-482/85-26 on 850601-0731.Violations Noted: Failure to Control Measuring & Test Equipment W/Procedure & Temporary Changes Not Incorporated in Procedure
	ML20132E421
Person / Time
Site:	Wolf Creek
Issue date:	09/13/1985
From:	Bruce Bartlett, Bundy H, Cummins J, Martin L; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	;
Shared Package
ML20132E374	List: ML20132E370; ML20132E381; ML20132E421;
References
	50-482-85-26, NUDOCS 8510010082
	Download: ML20132E421 (15)
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APPENDIX B

US NUCLEAR REGULATORY COMMISSION

REGION IV

NRC. Inspection Report: 50-482/85-26

LP: NPF-42

Docket: 50-482

Licensee: 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 31, 1985

Inspectors:

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J. E. Cummins, Senior Reactor Inspector,

Date

Operations

(pars. 3, 4, 5, 6, 7, 8, 9, 10, 11,

12, and 13)

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B. L. Bartlett, Resident Reactor Inspector,

Date

Operations .

(pars. 4, 5, 6, 7, 8, 9, 12, and 13)

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H. F. Bundy' Resident Reactor Inspector,

Date

Operations (pars. 4 and 13)

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Inspection Summary

Inspection Conducted June 1 to July 31, 1985 (Report 50-482/85-26)

Areas Inspected:

Routine, unannounced inspection including plant status;

followup on previously identified items; operational safety verification,

engineered safety features system walkdown; startup test witnessing, startup

test data review; onsite followup of events; site emergency drill;

enforcement conference (security); allegation followup; security; and plant

tours. The inspection involved 450 inspector-hours onsite by three NRC

inspectors including 67 inspector-hours onsite during offshifts.

Results: Within the 13 areas inspected, two violations were identified,

(failure ty.ontrol measuring and test equipment in accordance with

procedure, paragraph 4, and. temporary changes not incorporated in procedure as

required, paragraph 6). One unresolved item is identified in paragraph 12.

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DETAILS

1.

, Persons Contacted

Principal Licensee Personnel

1

G. L. Koester, Vice President-Nuclear

+o#C. C. Mason, Director-Nuclear Operations

+o#F. T. Rhodes, Plant Superintendent

J. A. Zell, Operations Superintendent

+oH. K. Chernoff, Licensing

oM. G. Williams, Supt. of Regulatory, Quality, and Administrative

Services

+oK. Peterson, Licensing

~

+#0. L. Maynard, Licensing Supervisor

W. B. Norton, Reactor Engineering Supervisor

oR. M. Grant, Director-Quality

- o#J. W. Johnson, Chief of Security

+C. J. Hoch, QA Technologist

0+W. M. Lindsay, Quality Systems Supervisor

+od. Hoyt, Emergency Plan Supervisor

J. Houghton, Operations Coordinator

A. Freitag, Nuclear Plant Engineering Site

o+R. Flannigan, Site Representative, Kansas City Power & Light

+F. D. McLaurin, Startup Manager

G. D. Boyer, Superintendent of Technical Support

M. Nichols, Health Physics Supervisor

D. R. Smith, Superintendent of Plant Support

i j.

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

.

during the inspection period to discuss identified issues.

.

+ Denotes those personnel in attendance at the exit ineeting held on

July 1, 1985.

oDenotes those personnel in attendance at the exit meeting held on

August 5, 1985.

Denotes those personnel attending the Enforcement' Conference held in

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Region IV on June 27, 1985.

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2.

' Plant Status.

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On June 6, 1985, the reactor plant-initially entered Mode 1 (power

' operation greater than 5%) and on June 12, 1985, the turbine generator

was synchronized to the grid for the first time. During this inspection

period, power ascension testing at the 20, 30, 50, and 75 percent power

.

I,lateaus was completed.

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3.

Followup on Previously Identified Items

(Closed) Infraction (50-482/78-13): Failure to Meet Concrete Acceptance

Criteria for Containment Base Mat

^

This item was transferred to NRR for evaluation. The NRR evaluation and

conclusion is contained in the Final Safety Analysis Report (FSAR) for

the Wolf Creek Station (NUREG-0881) in paragraph 3.8.4.

This item was

incorrectly identified in NRC Inspection Report 50-482/84-22 as

50-482/78-04-B.

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(Closed) Open Item (50-482/8427-03): Calibration Status of Installed

Instruments

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This open item tracked licensee actions to resolve inspector concerns

over the controls associated with normally installed plant

instrumentation used to support Technical Specification required

,

surveillance testing.

In response to these concerns, the ifcensee has

pr epared a list of all such instruments and cross referenced the

instruments to the surveillances they support. The list has been

incorporated into Procedure ADM 02-300 along with requirements for the

instrument and control department.to notify operations any time one of

the instruments is found out of service or out of calibration.

Operations is tasked with the responsibility for assessing the impact of

- the instrument problem on the validity of the surveillance tests that

instrument supports. This item was closed in NRC Inspection Report

50-482/84-57 and 50-482/85-11 but was incorrectly identified as open

-item 50-482/8427-02.

(Closed) Open Item (50-482/8459-07): Corrections to 50% Pseudo Rod Drop

Test Procedure, SU7-SF09.2-

The 50% power pseudo rod cluster control assembly (RCCA) rod drop test

was deleted from the Wolf Creek power ascension test program. By letter

dated July 3,1985, the Office of Nuclear Reactor Regulation notified

' the licensee of its acceptance of the deletion of this test. Deletion

of the test was based on successful performance of the test on other

Westinghouse plants.

-(Closed) Open Item (50-482/8509-02):

Inclusion of Loop Resistance

Temperature Detector Response Times in Preoperational Test Procedure

SU3-SA01

Test Discrepancy Report No. 30 documented the inclusion of the required

response times into Preoperational Test Procedure, SU3-SA01.

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(Closed) Licensee Condition Attachment 2, Item 2.C(7) Qualification of

Persennel :

This license condition required KG&E to certify the individuals who will

be standing watch as shift advisors. . A list of the certified shift

advisors was.provided to the NRC via KG&E letter KMLNRC 85-128, dated

May 29, 1985, satisfying this license condition.

(Closed) SER Item (50-482/84-00-158): Alternative Shutdown Capability

for the Control Room

This item tracked the installation of five new isolation switches and

the modification of four existing isolation switches. These switch

changes were required to provide isolation of equipment from a control

room fire. To verify that this item was completed satisfactorily, the

NRC inspector reviewed licensee documentation of the required switch

changes, observed installed switches in the field, and reviewed

Procedure 0FN-00-017, Revision 3, Control Room Evacuation.

(Closed) SER Item (50-482/84-00-150): Seismic Qualification of

Thermccouple/ Core Cooling Monitor System

This item, tracked the onsite completion of Westinghouse Field Change

Notice (FCN) SAPM-10627. To ensure the field modifications were

.

completed, the NRC. inspector reviewed a signed off copy of FCN

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SAPM-10627, discussed the modifications with licensee personnel, and

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inspected portions of the completed modifications in the field.

(Closed) SER Item (50-482/84-00-52): Post Implementation Review of

Emergency Support Facilities

This item was closed in NRC Inspection Report 50-482/85-11, but was

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incorrectly identified as SER-Item 50-482/84-00-50.

(Closed) SER Item (50-482/84-00-151): Setpoint Adjustment of Barton

Differential Pressure Switches

This item tracked the change in setpoint for Barton differential

pressure indicating switches model numbers 288A and 581A. From

discussions with licensee personnel and review of the WCGS Total Plant

Setpoint Document, the NRC inspector determined that the required switch

setpoints had been changed.

(Closed) Open Item (50-482/8427-04): Availability of Information in

Auxiliary Shutdown Room

The NRC inspector verified that plant operating procedures, that provide

instructions for required operations from the auxiliary shutdown room,

have been placed in the room. The NRC inspector discussed the shutdown

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accomplished from the auxiliary shutdown room during Power Ascension

Test Procedure SU7-0014, " External to Control Room Shutdown," with the

cognizant shift supervisor (SS), and the SS stated that adequate

procedures and material were available in the auxiliary shutdown room.

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

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with licensee personnel, independent verification of safety system

status and limiting conditions for operations, and reviewing facility

records.

The NRC inspectors, by observation and direct interview,

verified the physical security plan was being implemented in accordance

with the security plan.

During a tour of_ the plant on June 26, 1985, the NRC inspector observed

Pressure Gauge WC 9967 connected to BN-FI-968.

This pressure gauge was

out of calibration, not installed in accordance with a temporary

modification order or approved procedure, and did not have a

10 CFR 50.59 applicability review.

Failure to control this modification

to plant equipment in accordance with procedures is an apparent

violar.on.

(50-482/8526-01)

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.

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The NRC inspectors also inspected system components for damage or other

conditions that might degrade system performance. The ESF systems

listed below were walked down during this inspection report period:

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Auxiliary feedwater system

.

Emergency diesel generators

.

Reactor coolant charging system

.

Safety injection system

.

Residual heat removal system

.

Containment spray, system

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Safety Class 1E 4.16KV AC system

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No violations or deviations were identified.

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6.

Startup Test Witnessing

Selected portions of the startup tests listed below were witnessed to

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ascertain conformance of the licensee to license and procedural

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requirements, to observe the performance of the staff, and to ascertain

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the adequacy of test program records, including preliminary evaluations

of test results.

507-8804 - RCS Flow Coastdown Measurement Test

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SU7-BB05 - Pressurizer Continuous Spray Flow Setting

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SU7-0009.1 - Load Swing at 30%

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SU7-0014 - External to Control Room Shutdown

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SU7-0020.2 - Turbine Generator Test at 20%

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5U7-0907.2 - Plant Performance

.

SU7-SC03.2 - Thermal Power and Setpoint Data at 30%

.

SU7-SE02.1 - Operational Alignment of Nuclear Instrumentation

.

SU7-SE02.4 - Operational Alignment of Nuclear Instrumentation

.

SU7-SE03.1 - Preliminary Axial Flux Difference Instrumentation at-

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50% Power

SU7-SE03.2 - Axial Flux Difference Instrumentation Calibration at

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75% Power

SU7-SF03.3 - Hot Full Flow Rod Drops

.

SU7-SF03.4 - Hot No Flow Rod Drops

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SU7-SF09.1 - RCCA or Bank Worth, 30% Pseudo Rod Ejection

.

SU7-SR01 - Incore Movable Detector

.

SU7-SR04 - Incore Instrumentation Operability

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SU7-SR02 - Incore Movable Detector and Thermocouple Mapping at

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Power

SU7-AB01.1 - Automatic Steam Generator Level Control Test

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SU7-AB01.4 - Automatic Steam Generator Level Control Test

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SU7-0008.2 - Power Coefficient Determination

.

SU7-0012 - Rod Drop and Plant Trip

.

SU7-SF06.4 - Operational Alignment of Process Temperature

.

Instrumentation at 75% Power

SU7-0010.1 - Large Load Reduction - 75% Power

.

NRC inspector findings are discussed below:

While observing the adjustment of power range nuclear instruments per

Wolf Creek Work Request (WR) 91611-85, the NRC inspector determined that

the calorimetric had been run with a copy of Surveillance Procedure

STS SE-001, Revision 2, " Power Range Adjustment to Calorimetric," that

did not have applicable temporary changes incorporated in it. This

resulted in the nuclear instrument channels being adjusted to a power

level of 23.6.% when they should have been adjusted to 24.6% A followup

calorimetric was performed and the nuclear instruments were adjusted

accordingly.

Reactor power was maintained constant dur;ng this

evolution.

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Step 3.1.1 of the licensee's Administrative Procedure ADM 02-021, "Use

of Procedures in Operations," requires that, prior to use, a procedure

will be verified to insure.that it has all changes incorporated in it.

Perfomance of this surveillance using a copy 'of STS SE-001 that did not

have the latest changes incorporated is an apparent violation.

(50-482/8526-02)

7.

Startup Test Data Review

a.

The following test data packages were reviewed by the NRC

inspectors for:

Verification that all test changes, including deletions, were

.

approved, reviewed, and incorporated properly.

Verification that all test deficiencies were resolved in

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accordance with the appropriate procedures.

Verification that deficiencies which constitute a reportable

.

occurrence as defined by Technical specifications (TS) have

been properly recorded.

Verification that the as run copy of the completed test data

.

package was properly completed.

Verification that the test suninary and evaluation were

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completed in accordance with procedure.

Verification-that the test results were properly approved.

.

SU7-SF03.1, " Cold, No Flow Control Rod System Testing,"

Revision 3, dated March 29, 1985.

SU7-SF03.2, " Cold, Full Flow Control Rod System Testing,"

Revision 2, dated April 6, 1985.

SU7-SF03.3, " Hot, No Flow Control Rod System Testing,"

Revision 2, dated April 26, 1985.

SU7-VF03.4, " Hot, Full Flow Control Rod System Testing,"

Revision 2, dated April 26, 1985.

SU7-SF09.1, "RCCA or Bank Worth Measurement at Power (30%

Power Pseudo Rod Ejection)," Revision 1, dated

June 23, 1985.

SU7-5011, " Initial Criticality and Low Power Test Sequence,"

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Revision 3, dated May 17,_1985.

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SU7-5012, " Initial Synchronization and 20% Power Test

Sequence," Revision 2, dated April 28, 1985.

SU7-S013, " Power Ascension and 50% Power Test Sequence,"

Revision 3, dated June 19, 1985.

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SU7-BB02, " Pressurizer Heater and Spray Capability Test,"

Revision 1, dated April 10, 1985.

SU/-BB04, "Rx Coolant System Flow Coastdown Test, Revision 1,

dated May 19, 1985.

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SU3-BB13, "Special Test Procedure for Pressurizer Relief

Valve," Revision 0,

b.

The following tests were reviewed by the NRC inspectors for:

Verification that the cognizant engineering function has

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evaluated the test results and has signified that the testing

demonstrated that the system met design requirements.

Verification that the licensee specifically compared test

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results with established acceptance criteria.

Verification that those personnel responsible for review and

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acceptance of test results have documented their review and

acceptance of the data package and the evaluation.

Verification of quality assurance / safety group t r other

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independent review of test results as prescribe.d in FSAR or

other commitments.

Verification that those personnel charged with responsibility

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for review and acceptance of test results have documented

their review and acceptance of the data package and the

evaluation.

SU7-BB03, " Reactor Coolant System Flow Measurement,"

Revision 1, dated March 24, 1985.

SU7-SR04, "Incore Instrumentation Operability Test,"

Revision 2, dated April 26, 1985.

SU7-S010, " Post Core Loading Precritical Test Sequence,"

Revision 3, dated May 19, 1985.

SU7-SE02.2, " Operational Alignment of Nuclear

Instrumentation," Revision 2, dated May 2, 1985.

SU7-SE03.1, " Preliminary Axial Flux Difference Instrumentation

Calibration," Revision 1, dated June 28, 1985.

No violations or deviations were identified.

8.

Onsite Followup of Events

The NRC inspector performed onsite followup of the nonemergency events

listed below. The NRC inspector 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. The NRC inspector will review the

license event report (LER) ~for each of these events and will report any

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findings in future NRC inspection reports.

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When availability of the NRC inspector allowed observation of these

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activities.

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Summary of all ESF Actuations during June and July are listed below:

Date

Event **

Plant Status

Cause

6-2-85

CRVIS

Mode 3

Spike on radiation monitor

6-4-85

AFAS

Mode 3

Loss of condenser vacuum

6-5-85

CRVIS

Mode 3

Spike on radiation monitor

6-6-85

Rx Trip

Mode 1

Lo-Lo S/G level

6-7-85

CRVIS

Mode 3

Spike on radiation monitor

6-9-85

AFAS

Mode 2

Main feedwater pump trip

6-9-85

AFAS

Mode 2

Hi-Hi S/G level

6-11-85

AFAS

Mode 1

Hi-Hi S/G level

6-13-85

Rx Trip /AFAS

Mode 1

Lo-Lo S/G level

6-14-85

AFAS

Mode 1

Hi S/G level

6-23-85

Rx Trip

Mode 1

RTB "A" accidentally opened

6-24-85

Rx Trip /AFAS

Mode 2

Lo-Lo S/G level

7-2-85

CRVIS

Mode 3

Spike on radiation monitor

7-5-85

AFAS

Mode 1

Spike on radiation monitor

7-5-85

CRVIS

Mode 1

Hi-Hi S/G level

7-7-85

CRVIS

Mode 1

Spike on radiation monitor

7-8-85

CRVIS

Mode 1

Spike on radiation monitor

7-8-85

CRVIS

Mode 1

Spike on radiation monitor

7-9-85

CRVIS

Mode 1

Spike on radiation monitor

7-9-85

Rx Trip

Mode 1

Lo-Lo S/G level

7-10-85

Rx Trip

Mode 1

Lo-Lo S/G level

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7-10-85

MFIVA

Mode 2

Hi S/G level

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7-11-85

AFAS

Mode 2

Hi S/G level

7-11-85

Rx Trip

Mode 1

Hi flux on IR

7-12-85

CRVIS

Mode 3

Broken detector tape

7-12-85

CRVIS

Mode 3

Spike on radiation monitor

7-12-85

CRVIS

Mode 3

Spike on radiation monitor

7-17-85

CRVIS

Mode 1

Faulty bypass switch

7-20-85

D/G Start

Mode 1

Relay accidentally bumped

7-22-85

CRVIS

Mode 1

Broken detector tape

7-23-85

Rx Trip /AFAS

Mode 1

Loss of control power

7-28-85

CRVIS

Mode 1

Spike on radiation monitor

7-31-85

Rx Trip /AFAS

Mode 1

PR HI negative rate

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CRVIS.- Control room ventialtion isolation

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AFAS - Auxiliary feedwater actuation signal

Rx Trip - Reactor trip

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MFIYA'- Main feedwater isolation valve actuation-

S/G - Steam generator-

RTB - Reactor trip breaker

IR - Intermediate range

PR - Power range

Specifics of the reactor trip and related events that occurred on

June 13, 1985, are discussed below:

At 03:45 CDT on June 13, 1985, with reactor plant power at approximately

15%, the main turbine generator was manually tripped due to excessive

vibration. Subsequent to the turbine trip at 03:53 CDT the operating

main feedwater pump 'A' tripped. The operator manually started the -

motor driven auxiliary feedwater pumps and reestablished feedwater flow

to the steam generators. When steam generator water levels continued to

decrease, the operator attemped to start the turbine driven auxiliary

feedwater pump, but it tripped on overspeed when the operator opened the

steam supply valves to the turbine out of sequence (i.e., the

trip /trottle valve was oper.ed prior to opening the main steam supply

val ve) . At 04:00 CDT a low-low level condition in steam' generator 'B'

caused a reactor trip, an auxiliary feedwater actuation, and a steam

generator blowdown and sample isolation. A feedwater isolation also

occurred due to low reactor coolant system average temperature in

conjuction with the reactor trip. At the time, the auxiliary feedwater

actuation occurred the motor driven auxiliary feedwater pumps were

already running having been manually started as described above. All

engineered safety features and reactor protection system equipment

operated per design requirements. When the operator closed the main

staam line isolation valves (MSLIV) manually in the slow-close mode to

help reduce reactor coolant syste:a cooldown, the ' A' steam generator

MSLIV failed to close. The valve closed when the operator manually went

-to the fast-close mode.

Evaluation and applicable corrective actions that were taken by the

licensee to correct the problems encountered during this event were as

follows:

The cause of the main feedwater pump ' A' trip that initiated the

.

event was not identified. Prior to -restart following the trip, the

licensee tested the main feed pump and its related control circuits

for proper operation and calibration. The main feed pump was

operated for several hours while key parameters were monitored with

test instrumentation. During this operation of the main feed pucp,

no indication of a malfunction was detected. Additional test

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instrumentation for monitoring key parameters was left installed on

.the pump during subsequent operatians to trend pump performance and

help identify any malfunctions.

All' operators were instructed on the proper sequence of valve

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operations for starting the turbine driven auxiliary feedwater

pumps.

The failure of the main steam isolation valve in the slow-close

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mode was due to low hydraulic oil reservoir level caused by a

. leaking 'O' ring. The 'O' ring was replaced correcting the -

probl em.

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Th) violations or deviations were identified.

9.

Emergency Drill Observation

On June 20, 1985, the NRC resident inspectors observed an emergency

preparedness' field exercise that was conducted by the licensee. The

-purpose of the drill was to train appropriate personnel to respond to a

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radiological emergency at WCGS. The players in the drill performed the

actions they would.be required to perform in an actual event and used

emergency procedures .where appropriate to respond to the simulated

events 1that took ~ place. Appropriate Coffey County personnel also

participated in the drill and the Coffey County Response Center was

. activated. The WCGS training' simulator was used to simulate reactor

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plant conditions and responses during the accident scenario.

-The NRC inspectors observed drill activities in the technical support-

center, the emergency operations facility, and the simulator control

room.' The NRC inspector 'also attended the licensee's critique of the

'

drill at its conclusion.

No violations'or deviations were identified.

10. Enforcement Conference

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On June 27,'1985, an enforcement conference with KG&E management was

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held in the NRC Region IV offices to discuss NRC concerns in the area of

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. security at WCGS. The events that generated the NRC concerns had been

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identified during an onsite visit by a Region IV security inspector and

Lby the resident: inspector. Details of. the NRC concerns were reported in

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NRC Inspection Report 50-482/85-27. Licensee representatives at the

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enforcement conference described corrective actions that were~being

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taken in the areas of concern.

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11. Allegation Followup

On June 3,1985, an anonymous alleger, via a letter, related the

following concern to.the NRC inspector:

" Rainwater flows from a manhole through conduit into the health physics

lab and if rain can flow into the lab there is.a possibility that-

radioactive substances could leak out."

The NRC inspector in following up on the above concern determined the

following from discussions with ifcensee personnel and a review of

ML20132E421

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