Text

Insp Rept 50-482/85-30 on 850801-31.Violations Noted:Test Port Not Reclosed in Accordance W/Procedure,Inadequate Procedure & Flow Orifices Installed Backwards.One Open Item Identified Re Cutler-Hammer Type E-30 Switches
	ML20138B669
Person / Time
Site:	Wolf Creek
Issue date:	10/02/1985
From:	Bruce Bartlett, Cummins J, Martin L; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	;
Shared Package
ML20138B646	List: ML20138B642; ML20138B654; ML20138B669;
References
	50-482-85-30, NUDOCS 8510210298
	Download: ML20138B669 (16)
	v • d • e

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APPENDIX B

US NUCLEAR REGULATORY COMMISSION

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NRC Inspection Report: 50-482/85-30 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: August 1 to 31, 1985

, Inspectors: '

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J. E. Cummins, Senior Reactor Inspector, Date

Operations, (pars. 2,3,4,5,7,8, ;

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B. L. Bartlett, Resident Reactor Inspector, Date ,

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

9, 10, 12, 13, and 15) i

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Approved:

L.Y Martin, CKef, Project Section B

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, Reactor Projycts Branch Da(e/ '

Inspection Summary

Inspection Conducted August I to 31,-1985 (Report 50-482/85-30) i

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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; security; plant tours; monthly ,

maintenance observation; surveillance witnessing; independent inspection; and

Ruskin fire dampers. The inspection involved 210 inspector-hours onsite by two '

NRC inspectors including 32 inspector-hours onsite during offshifts. '

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8510210298 851009 I

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PDR ADOCK 05000482

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Results: Within the 14 areas inspected, three violations were identified (test

port not reclosed in accordance with procedure, para. 4; inadequate procedure,

para. 8; and flow orifices installed backwards, para. 10). One open item was j

identified (Cutler-Hammer Type E-30 switches, para.13).

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DETAILS

1. Persons Contacted

Principal Licensee Personnel

G. L. Koester, Vice President-Nuclear

C. C. Mason, Director-Nuclear Operations

F. T. Rhodes, Plant Superintendent

R. M. Grant, Director-Quality

J. A. Zell, Operations Superintendent

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

Services

0. L. Maynard, Licensing Supervisor

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H. K. Chernoff, Licensing

K. Peterson, Licensing

W. B. Norton, Reactor Engineering Supervisor

i J. J. Johnson, Chief of Security

*W. M. Lindsay, Quality Systems Supervisor

R. Hoyt, Emergency Plan Supervisor

*R. Flannigan, Site Representative, Kansas City Power & Light

*W. J. Rudolph, Site QA Manager

i *L. E. Borders, Shift Supervisor

A. A. Freitag, Site NPE Manager

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C. M. Herbst, Assistant Project Engineer-Bechtel

D. L. Reed, Lead Engineer-ISEG

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 '

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September 4,1985.

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2. Plant Status

l On August 8, 1985, the plant initially reached 100 percent of full power

and commenced the 250-hour warranty run. On August 12, 1985, the 250-hour

run was interrupted fc:' approxiaiately 7 days when main feed pump (MFP) 'B'

was removed frcm service for maintenance which was completed on

August 20, 1985. The 250-hour wseranty run at 100 per'.ent power was

resumed on August 21, 1985, and completed on August e/, 1965. During this

inspection period, power ascension testing at the 90 and 100 percent

levels was completed.

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l 3. Licensee Actions on Previously identified Identified items

(Closed) Violation (50-482/EA05-27-IIC): failure to Operate the Plant to

Prpcodural Requirements.

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This' item involved the failure to tag the removal of the blind flanges to

four flow elements in accordance with Administrative Procedure (ADM)02-101. The NRC inspector reviewed the licensee's response to the Notice

of Violation, verified the change to ADM 02-101 for clarification had been

incorporated, reviewed the letter from the plant manager to all section

superintendents involved in test procedure implementation, and verified

through. field inspections of selected temporary modifications that the

requirements of ADM 02-101 are being met. This item is closed.

(Closed) Violation (50-482/8511-03): Failure to Operate the Plant to

Procedural Requirements.

, This is a duplicate documentation of enforcement action violation

(50-482/EA8527-IIC) which is discussed and closed above. This item is

closed.

(Closed) Open Item (50-482/8508-05): RD #28 Spent Fuel Pool Leak

Monitoring System.

This open item involved preoperational test deficiency deferral RD #28

concerning testing on the spent fuel pool leak monitoring system. The NRC

inspector reviewed Test Discrepancy Report (TDR)-004.to test SU3-EC02,

Revision 0, and verified the required leak rate readings were obtained.

This item is closed.

(Closed) Open Item (50-482/8511-10): RD #42 Rod Control Cluster Assembly

Change Fixture Engage Light.

The NRC inspector reviewed completed copies of Wolf Creek Work Requests

(WR) 2985-85 and 7606-85 which documented, respectively, the installation

and the functional testing of the engaged light for the rod control

cluster assembly change fixture. This item is closed.

(Closed) Violation (50-482/8519-02): Failure To Respond To Control Room

Annunciator.

This item involved a failure to respond to Annunciator F-79. The NRC

inspector has questioned on-duty personnel on their awareness of

annunciator panels and observed personnel follow the alarm procedures when

required. This item 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

Itcensee facilities, tours of the facility, interviews and discussions

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,

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verified the physical security plan was being. implemented in accordance

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with the security plan. '

NRC inspector observations noted'during this inspection period are !

discussed below:

J During a tour of the plant on August 5, 1985, the NRC inspector observed

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the plug missing to a test port located in the ductwork on the discharge

] of containment purge filter Adsorber Unit FGT01. At the time of the

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observation, the containment purge system was not in operation and there

j was no testing involving this system in progress. This failure to control !

activity affecting quality in accordance with documented procedures is a *

violation. (50-482/8530-01)

j_ - S. . Engineered Safety Features (ESF) System Walkdown

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

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selected accessible portions of the systems. The NRC inspectors verified

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] valses and electrical circuit breakers were in the required position,

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

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inspectors also inspected system components for damage or other conditions

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that might degrade system performance. The ESF systems listed below were

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walked down during this inspection report period: j

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l . Emergency diesel generators i

. Safety injection system

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. Residual heat removai system

i . Reactor coolant charging system

)' . Fire protection sprinkler system

. Essential service water system ,

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Control building heating, ventilating, and air conditioning l

. Containment purge system

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No violations or deviations were identified. !

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l 6. Startup Test Data Review i

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! The NRC inspector reviewed completed power ascension test' procedures. The '

following items were considered in this review

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. Verification that all test changes, including deletions,-were f

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Verification that all test deficiencies were resolved in accordance I

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. 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 summary and evaluation were completed in

accordance with procedure.

. Verification that the test results were properly approved.

The following test data packages were reviewed by the NRC inspectors for:

. SU7-008.4 " Power Coefficient Determination-90 Percent Power,"

Revision 1, dated July 22, 1985

. SU7-0010.1 "Large Load Reduction at 75 Percent Power," Revi-

sion 1, dated July 27, 1985

. 5U7-0020.6 " Turbine generator tests (90' Percent Reactor Power),"

Revision 0, dated July 29, 1985

. SU7-5014 " Test Sequence at 75 Percent Power," Revision 3, dated

July 16, 1985

. SU7-5015 " Test Sequence at 90 Percent Power," Revision 3, dated

July 16, 1985

. SU7-SC03.6 " Thermal Power and Statepoint Data Collection at

90 Percent Power," Revision 1, dated July 30, 1985

. SU7-SE02.8 " Operational Alignment of Nuclear Instrumentation,"

Revision 1, dated July 8, 1985

. SU7-SE03.2 " Axial Flux Difference (AFD) Instrumentation

Calibration at 75 Percent Power," Revision 1, dated

July 25, 1985

SU7-SF06.5 " Operational Alignment of Process Temperature

Instrumentation," Revision 1, dated July 25, 1985

SU7-SR02 "Incore Moveable Detector and Thermocouple Mapping at

Power," Revision 1, dated January 23, 1985

STS RE-013 "Incore-Excore Detector Calibration," Revision 0,

dated July 3, 1985

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STS SE-001 " Power Range Adjustment to Calorimetric," Revision 3,

dated July 27, 1985

SYS SR-200 " Moveable Incore Detector Operation," Revision 1,

dated November 1, 1984

No violations or deviations were identified.

7. Startup Tcst Witnessing

Selected portions of the startup. tests listed below were witnessed to

ascertain conformance of the licensee to license and procedural

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

the adequacy of test program records, including preliminary evaluations of

test results.

SU7-SE03.3 " Axial Flux Difference Instrumentation Calibration,"

Revision 1

SU7-009.3 " Load Swing Tests-100 Percent Power," Revision 1

SU7-0010.2 "Large Load Reduction-100 Percent Power," Revision 1

SU7-0011 " Plant Trip From 100 Percent Power," Revision 0

SU7-0013 "NSSS Acceptance Test," Revision 1

Selected observations made by the NRC inspectors are discussed below:

. SU7-SE03.3 - The difference between the indicated axial flux .

difference (AFD) and the measured flux difference exceeded the

1.5 percent acceptance criteria specified in Step 9.3 of the test

procedure. The indicated axial flux differences were approximately

2 percent greater than the measured flux difference. To correct this

situation, the nuclear instruments were realigned using an AFD !

instrumentation fine tuning procedure which was furnished by the

nuclear instrument vendor, Westinghouse, and after this realignment,

the axial flux difference indications were verified to be within

specifications. A followup incore flux map was run and the new axial *

flux difference indications were verified to meet the acceptance

criteria when compared to the data obtained from this incore flux ,

map.

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. SU7-009.3 - Step 6.11 of the test procedure specified a power increase

of 10' percent power, but the maximum power attained was approximately

7 percent when the step was performed. The controlling rod bank

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reached its upper limit restricting further powerfincrease. .The -

licensee's evaluation of this test deficiency determined that the

! Final Safety Analysis Report (FSAR) requirements (Section 14.2.3.9.3)

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was to achieve an approximate 10 percent load decrease / increase and I

the 7 percent increase met the intent of the FSAR requirements. The !

j response of changes in plant parameters during this test was compared !

to test data taken for the same plant parameter changes at the

75 percent power plateau and the licensee determined from this ]

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I No violations or deviations were identified. !

! 8. .Onsite Followup of Events l

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The NRC inspector performed onsite followup of the nonemergency events

i listed below. -The NRC inspector observed control room personnel response,

observed instrumentation indicators of reactor plant parameters, reviewed

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logs and computer printouts, and discussed the event with cognizant ,

personnel. The NRC inspector verified the licensee had responded to the

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

agencies as required in a timely fashion. :

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1 Engineered safety feature actuations that occurred during the report I

j period are listed in the table below. The NRC inspector will review the !

4- license event report (LER) for each of these events and will report any i

findings in future NRC inspection reports. !

Date Event Plant Status Cause +

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8/6/85 CRVIS Mode 1 Lightning strike on

transmission line

8/7/85 Rx Trip Mode 1 MSR Hi-Hi level t

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Specifics of-the reactor trip and related events that occurred on !

August 7, 1985, are discussed below: !

. At 06:26 CDT on August 7 1985, with reactor plant power at-

approximately 92 percent, a turbine trip occurred due to moisture

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separator reheater (MSR) Hi-Hi level. Immediately after the turbine !

trip, the reactor tripped on P-9 (turbine trip with reactor power

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greater than 50 percent). During the trip an auxiliary feedwater j

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actuation signal (AFAS), main feedwater isolation signal (FWIS) and

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steam generator blowdown isolation signal were received as expected f

on low steam generator level. All equipment actuated properly except j

fi for two blowdown valves which did not indicate closed as requir'ed. ;

N Blowdown isolation valve BM HV-1 indicated closed at the radwaste

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control panel but did not indicate closed on the ESF status panel but ]

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the closed indicator light on its control room handswitch failed to ;

bA 3 filuminate. ', Limit switches for the two valves were adjusted and.the {

valves were tested satisfactorily. The cause for the MSR high level j

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that initiated the trip could not be identified. Minor adjustments

to approximately 24 secondary system controllers were performed and

the plant was restarted. ,

In addition to the reactor trip and CRVIS listed above, the NRC inspectors

also performed followup on the defect / deficiency reports listed below.

The NRC inspectors reviewed the reports and verified:

. The reportability determination was accurate and complete.

. The plant safety review committee reviewed the report when required.

. The proper corrective action was taken when required.

The following defect / deficiency reports were reviewed:

.85-105 " Air Flow From Containment to Environment Without a

Release Permit"

.85-106 " Reactor Trip"

.85-107 " Noncompliance with Technical Specifications-Missed Hourly

Fire Watch"

.85-109 " Reactor Trip From Turbine Trip-MSR Hi-Hi Level"

NRC inspector comments are noted below:

. The review of Defect / Deficiency Report 85-105 " Air Flow From

Containment to Environment Without a Release Permit," identified

deficiencies in Licensee Procedure STS IC-275B which allowed the

release of containment air to the outside atmosphere without first

obtaining and analyzing a grab sample as required by Technical

Specification surveillance requirement 4.11.2.1.2. When the STS was

performed, Step 5.1.8 required that' isolation dampers GT HZ-4, 5, 6,

7, 8, 9, 11, and 12 be opened or checked open, with these isolation

dampers open there were only closed exhaust fan discharge dampers to

prevent the release of conttinment air to atmosphere. These

discharge dampers leaked allowing a release of containment air to

occur 'for 40 minutes before an operator observed the drop in

containment pressure and terminated the discharge. When STS IC-275B

was written and reviewed, there was a failure to consider that all of

the isolation dampers did not need to open simultaneously and that a

grab sample should be taken in the event of leakage. This failure to

prescribe a procedure appropriate to the circumstances is a

violation. (50-482/8530-02)

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9. Security

The NRC inspectors verified the physical security plan was being

implemented by selected observation of the following items:

] . The security organization is properly manned and the security

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personnel are capable of performing their assigned functions.

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. Persons,within the protected area (PA) display their identification

badges, when in vital areas are properly authorized and, when-

required, are properly escorted.

- . Vehicles are properly authorized, searched, and escorted or

controlled within the PA.

. Persons and packages are properly cleared and checked before entry

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into the PA is permitted.

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l . The effectiveness of the security program is maintained when security

equipment failure or impairment requires compensatory measures to be

employed.

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procedures and the physical security plan.

j No violations or deviations were identified.

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10. Plant Tours

-At various times during the course of the inspection period the NRC

inspectors conducted general tours of the reactor building, auxiliary

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building, radwaste building, fuel handling building, control building,

turbine building, essential service water building, and the secured area

, surrounding the buildings.. During the tours, the NRC inspector observed

housekeeping practices, fire protection barriers and equipment,

maintenance on equipment, and discussed various subjects with licensee

personnel.

NRC inspector findings are discussed below:

As a part of a plant tour,.the NRC inspectors inspected the installation

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of flow orifice plates installed in safety-related systems ~ to verify the

plates were installed in accordance with design drawings. The orifice

plates are stamped with the work " Inlet"'which identifies the ' side of the

plate that should be installed facing upstream to the flow.

Drawing 556-32170, Revision 4, " Outline Drawing Concentric Orifice

. Plates," shows the correct orientation for installing the orifice plates.

The NRC inspector determined that Orifice Plates-EG-FE-83, " Centrifugal

l Charging Pump 'A' 011 Cooler," and EG-FE-89, " Residual Heat Removal Pump

l Seal' Coolers," were installed backwards from the orientation shown on

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Drawing 556-32170. Failure to install these orifice plates in accordance i

with the drawing is a violation. (50-482/8530-03) *

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11. Monthly Maintenance Observation [

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The NRC inspectors observed maintenance activities performed on ,

safety-related system, and components to verify that these activities were

conducted in accordance with approved procedures, Technical

Specificatfor.s, and applicable industry codes and standards. The

folicwing elcments were considered by the NRC inspectors during the

observation and/or review of the maintenance activities:

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. Limiting conditions for operation (LCOs) were met and, where 6

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applicable, redundant components were operable.

. Activities complied with adequate administrative controls. I

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

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used.

. Craftsmen were qualified to accomplish the designated task and i

technical expertise (i.e., engineering, health physics, operati is)

, were made available when appropriate. ;

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. Replacement parts and materials being used were properly certified.

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. Required radiological controls were implemented. #

. Fire prevention controls were implemented where appropriate.

. Required alignments and surveillances to verify post maintenance

operability were performed.

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. 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 listed below were observed -

and/or related documentation reviewed:

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. WR No. 3292-85, Revision 1, Centrifugal Charging Pump 'A' (CCP-A) !

rotating assembly replacement. This WR documented the replacement of

the rotating assembly for CCP-A. The pump shaft was gouged during an ,

earlier removal of the shaft sleeve. .

. WR Nos. 09603-85, 09601-85, 090602-85: These work requests ,

documented the replacement of terminal blocks in the terminal box

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No violations or deviations were identified.

. 12. Surveiltance Witnessing

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The NRC inspector observed selected portions of the performance of

, surveillance procedures and verified the following items:

. Testing was being accomplished by qualified personnel in accordance

with an approved procedure.

The surveillance procedure conforms to TS requirements.

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, . Required test instrumentation was calibrated.

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. Technical Specification limiting conditions for operation were

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. Test data was accurate and complete. The NRC inspectors performed I

independent calculations of selected test data to verify their

accuracy.

. The performance of the surveillance procedure conformed to applicable

administrative procedures.

. The surveillance was performed within the . required f requency and the

, test results met the required limits.

Surveillances witnessed by the NRC inspectors are listed below:

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l . STS IC-5048 - Channel Calibration of Reactor Coolant System Flow

Transmitters

. STS BG 100A - Centrifugal Charging System 'A' Train Inservice Pump

Test, Revision 0

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. STS RE-004 - Shutdown Margin Determination, Revision 2

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No violations or deviations were identified.

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13. Independent Inspection

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a. Unqualified Terminal Blocks Installed in Valve Terminal Boxes

On August 8, 1985, the licensee informed the NRC resident inspector

and Region IV that unqualified (lower temperature rating) terminal

blocks had been used to replace damaged terminal blocks in electrical ;

terminal boxes on the main steam isolation valves (MSIV) and the main

feedwater isolation valves (MFIV). <

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Fourteen of the unqualified terminal blocks were installed during the

period from December 1984 through July 1985. The following table

shows the number of unqualified terminal blocks installed in each

affected valve and the approximate date they were installed.

Number of Unqualified Approximate

Valve No. Blocks Installed Date Installed

MFIV AE-FV-42 6 December 7,1984

MFIV AE-FV-39 4 December 8,1984

MSIV AB-HV-14 4 July 31,1985

The unqualified terminal blocks are Type NU-2 blocks manufactured by

Connectron Incorporated and supplied to WCGS by Anchor-Darling Valve

Company. These unqualified blocks were identical in design

configueation to the higher temperature qualified blocks but were

made out of polyamide (nylon) material whereas the qualified blocks

were made out of a polysulfone material. Anchor-Darling supplies

both the low temperature polyamide terminal blocks (Part No. W 30164)

and the nign temperaidre polysulfone terminal blocks (Part

No. W 32676). On July 9,1984, WCGS received a lot of 16 of the low

temperature blocks which were incorrectly identified by

Anchor-Darling on all~ the accompanying documentation as Part

No. W 32676, the part number for the high temperature qualified

terminal block. The licensee has determined that the 14 unqualified

terminal blocks that were installed in the MSIV and the MFIV terminal

boxes came from this lot of 16.

The licensee has replaced the unqualified terminal blocks with

qualified terminal blocks.

b. Cutler-Hammer Type E-30 Switches

On August 9, 1985, the. licensee informed the.NRC' resident inspector

of a potential problem with Cutler-Hammer Type E-30 push button

switches. The effects of the potential problem with the Type E-30

switches were first observed when three fire dampers failed to

automatically close as required when a control room ventilation

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isolation signal actuation occurred on July 19, 1985 (LER 85-057-00).

Followup investigation by the licensee determined that the dampers

had failed to close because a Type E-30 Cutler-Hammer handswitch

operating mechanism had hung up (failed to return to its normal

position). This was a spring operated return-to-normal momentary

contact switch. This type failure cannot be detected by observation

of the portion of the switch that is visible to the operator on the

front of the control panel but can be detected by observing the

switch plunger position at the back of the switch behind the control

panel.

[ Actions taken by the licensee to evaluate / correct the potential

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. Fifteen switches have been removed and sent to Cutler-Hammer for

evaluation. Five of these switches had exhibited some form of

failure whereas the.other ten were randomly selected to provide

additional data for evaluation.

. A total of 457 Cutler-Hammer Type E-30 switches were identified

to be installed in safety-related circuits. These switches were

inspected and five of them were found in a stuck position during

this initial inspection. Each of the 457 switches was evaluated

for its impact on safety and it was determined tnat 294 of the

switches could defeat an automatic safeguards actuation if it

failed. The licensee has identified each of the 294 switches

located on the main control board in the control room with a

small label, and implemented administrative controls which

require verification that each switch operated has functioned as

required. A weekly visual inspection of all identified switches

is being performed to verify that all switch plungers are in

their normal position.

. The licensee conducted a review of WCGS emergency operating

procedures (EMGs) and the activities related to each reference '

in the EMGs to operation of one of the E-30 type switches were

evaluated, and it was determined that there were no situations

where there would not be time to verify normal switch operation

prior to a required subsequent operation.

. The licensee also performed an engineering analysis and.

determined that the primary use of the switches is for valve or

ventilation damper control. Each switch circuit was analyzed

and all situations were identified where switch binding could

inhibit an engineered safety features operation, where a bound

switch would not be obvious (i.e., an associated alarm

energized), and where subsequent operation of the affected

component could be inhibited.

. The licensee, Cutler-Hammer, and Bechtel are analyzing the

potential switch problem to determine if the observed switch

failures were isolated cases or if there is a generic problem.

This is an open item pending.the results of this analysis.

(50-482/8530-04)

No violations or deviations were identified.

14. -Ruskin Fire Dampers

Three concerns related to the' ability of louvered fire dampers installed

at WCGS and manufactured by Ruskin Manufacturing Company to function per

design'under required conditions have been addressed by the licensee.

Each of these concerns and the corrective action is discussed below:

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a. Failure of Dampers To Fully Close Against Ventilation System Air

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The licensee, in conjunction with Ruskin and Bechtel (architect /

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engineer), made modifications to the damper closure spring and

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associated hardware and then performed an operational test of all

"

dampers (vertical and horizontal).

'

-b; Mullion Welds on Multisection Horizontal-Dampers.

.

By letter dated April 22, 1985, Ruskin Manufacturing Company, in

";,- accordance with the requirements of 10 CFR Part 21, notified the NRC

'- that welds used to join damper sections and mullion plates were not

in accordance with design drawings, Ruskin Drawing 5415.

'

Underwriters Laboratory's evaluation determined that the existing

. welds 1-inch on 9-inch centers in lieu of 1-inch on 6-inch centers

l'- were acceptable and would.not affect the 3-hour _ fire rating of the

,

dampers. Design drawings were changed to show the existing weld

conditions at WCGS. The license inspected all affected dampers and

rewelded one damper where the weld spacing exceeded the 9-inch

criteria.

,

~

c. -Inadequate Clearance Between Damper' Frames and Embedded Penetration

Sleeves.

i

By letter dated May 1, 1985, KG&E, in accordance with the-

requirements of 10 CFR Part 21, notified the NRC that some of the

Ruskin fire dampers installed at WCGS had less than the designed

i clearance between the fire damper frame and the penetration sleeve in

i which the damper was installed. This created a condition that might

! prevent the affected dampers from funtioning as designed due to

l buckling caused by-thermal expansion during a fire. The licensee has

implemented a program to inspect all affected fire dampers and to

replace or modify the dampers so that each damper installation has

the required gap between the damper and its associated penetration

sleeve.

! The above concerns with Ruskin fire dampers were reported and discussed in

[ WCGS LER 85-017-00. The modifications,and retesting have been

- satisfactorily completed on all the installed dampers-except for eight

i dampers which will require cold shutdown plant conditions due to their

location in safety-related ventilation systems. NRC Open Item

50-482/8515-01 was initiated to track concerns with the Ruskin fire

dampers. This open item was discussed in NRC Inspection Report

50-482/85-19. This item will remain open pending completion of all work

t

and testing of Ruskin' fire dampers related to the concerns discussed

above.

!

.

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-16-

15. Open Items.

Open items are matters which have been discussed with the licensee, which

will be reviewed further by the inspector, and which involve some action

on the part of the NRC or licensee or both. An open item disclosed

during the inspection is discussed in paragraph 13.

16. Exit Meetings

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

findings of this inspection on September 4, 1985. The NRC inspectors also

attended entrance / exit meetings of other NRC region-based inspectors

identified below:

Inspection Lead. Area Inspection

Period Inspector Inspected Report No.

8/12-16/85 R. Baer Radwaste 85-32

8/19-23/85 C. Hackney. Emergency

Preparedness 85-33

8/26-30/85 J. Kelly -Security' 85-34

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