IR 05000482/1986018

Insp Rept 50-482/86-18 on 860803-31.Violations Noted:Failure to Comply W/Licensee Temporary Mod Procedure & Failure to Lock Valve

ML20215E273
Person / Time
Site:	Wolf Creek
Issue date:	09/29/1986
From:	Bruce Bartlett, Cummins J, Hunter D, Mullikin R, Sharkey J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20215E248	List: IR 05000482/1986018 ML20215E245 ML20215E263
References
50-482-86-18, NUDOCS 8610150186
Download: ML20215E273 (17)

Text

,' i APPENDIX B US NUCLEAR REGULATORY COMMISSION

REGION IV

Nb Inspection Report: 50-482/86-18 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 3 - 31, 1986 Inspectors:['

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E. Cummins , Senior Resident Inspector,

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

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B. L. Bartl'ett, Resident Reactor Inspector, 9- F* T$

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

d . MLALL R. P. Mullikin, Proj ect Engineer 9kla Date (par. 8)

lbMt f/ h g . M. Sharkey, Inspection Specialist Date (par. 3)

Approved: s d,O ,, W) ' h/2-9/Eb D. R. Hunter, Chief, Project Section B, Date Reactor Projects Branch 2 8610150186 861006'

PDR ADOCK 05000482 O PDR

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',' 4-2-Inspection Summary Inspection Conducted August 3-31, 1986 (Report 50-482/86-18)

Areas Inspected: Routine, unannounced inspection including plant status; operational safety verification; engineered safety features system walkdown; monthly surveillance observation; followup on Regional request, environmental qualification of electrical equipment, and fire protectio Results: Within the seven areas inspected, two violations were identified (failure to comply with licensee's temporary modification procedure, paragraph 3; and failure to lock valve as required by licensee's procedure, paragraph 3). One unresolved item is identified in paragraph 7.

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.I-3-DETAILS

- Persons Contacted

' Principal Licensee Personnel G. L.'Koester, Vice President-Nuclear

• J. A. Bailey, Interim Site Director

• F. T. Rhodes, Plant Manager

• G. D. Boyer, Deputy Plant Manager

• M. Grant, Director-Quality M. Estes, Superintendent of Operations M. D. Rich, Superintendent of Maintenance

+M. G. Williams, Superintendent of Regulatory, Quality, and Administrative Services 0. L. Maynard, Manager, Licensing

• K. Peterson, Licensing

+*G. Pendergrass, Licensing

+*W. M. Lindsay, Supervisor, Quality Systems

+*C. J. Hoch, QA Technologist

+*W. J. Rudolph, QA Manager-WCGS

• C. E. Parry, Superintendent of Quality Engineering

• A. A. Freitag, Manager, Nuclear Plant Engineering-WCGS M. Megehee, Compliance Engineer B. McKinney, Superintendent of Technical Support

• K. Parks, Training Coordinator

• J. L. Houghton, Operations Coordinator-0ps

• R. L. Logston, Site Chemist

• R. D. Flannigan, Supervisor of Compliance

+^ Turinetti, Surveillance Coordinator

^ Austin, Operations Coordinator

+* Hoyt, Sr. Engineer Specialist-0ps

• E. Gimple, Technical Staff-Materials Quality

+J. L. Blackwell, Fire Protection Coordinator

+M. M. Nicholas, Superintendent of Plant Support The NRC inspectors also contacted other members of the licensee's staff during the inspection period to discuss identified issue * Denotes those personnel in attendance at the exit meeting held on September 4, 198 + Denotes those personnel in attendance at the exit meeting held on August 14, 1986 Plant Status The plant operated in Mode 1 during this inspection perio . _ - - . _ _ _ - - . - . . _ .

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_4 3 .~ 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 and limiting conditions for operations, and reviewing facility record The NRC inspectors, by observation and direct interview, verified the physical security plan was being implemented in accordance with the security plan and that radiation protection act,ivities were controlle By observing valve position, electrical breaker position, and control room indication, the NRC inspectors confirmed the operability of the component cooling water system (EG). The following documents were utilized by the NRC inspector when verifying operability of the EG system:

o Drawing M-12EG01, 02(Q), Revision 1, " Piping and Instrumentation Diagram Component Cooling Water System" o CKL EG-120, Revision 7, " Component Cooling Water System Valve, Switch and Breaker Lineup" The NRC inspectors also visually inspected safety components for leakage, .

physical damcge, and other impairments that could prevent them from performing their designed functio Selected NRC inspector observations tre discussed below: Locking Of Safety-Relsted Valves On August 20, 1986, the NRC inspectors, during a walkdown of the auxiliary feedwater system (AFW) observed that 3 out of the 4 turbine driven AFW pumps discharge valves were not properly lockwired. All four valve handwheels were in the required neutral position; however, AL HV-006 had its lockwire broken and twisted back together, AL HV-008 was not lockwired and the lockwire to AL HV-010 was lockwired so loosely as to defeat the purpose of the lockwire. This is an apparent violation (482/8618-02). The shift supervisor (SS)

and supervising operator (50) were informed and took immediate corrective action to replace the lockwire During an inspection conducted on August 5, 1986, of the licensee's temporary modifications program, the NRC inspector identified the following items that were contrary to licensee procedures:

o The required information for a lifted lead had not been recorded in Section D of Temporary Modification Order 86-81K Sr.ction 6.1.1.1.(1) of Procedure ADM 02-101, Revision 15,

" Temporary Modifications," required that for a lifted lead temporary modification the cabinet name/ location, terminal block

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-5-number and terminal number, wire number and tag number shall be entered in Section D of the temporary modification orde o Temporary Modification Order 86-72KC had not be'en initialed for the installation and verification of Tag No. 86-72KC-01 as required by Section 6.1.6.1 of Procedure ADM 02-101, Revision 15, " Temporary Modifications."

o The current procedure was not used in preparing Temporary Modification Order 86-72KC as required by Section 4.2.4 of Proceduro ADM 01-001, Revision 12, " Introduction of Wolf Creck Generating Procedures," in that this modification order prepared on July 10, 1986, used out-of-date Revision 14 of ADM 02-101, versus the current Revision 15, dated May 27, 198 The above failures to follow procedures is an apparent violation (482/8618-01).

4. Engineered Safety Features (ESF) System Walkdown The NRC inspectors verified the operability of an ESF system by walking dcwn selected accessible portions of the system. The NRC inspectors verified valves and electrical circuit breakers were in the required position, power was available, and valves were locked where require The NRC inspectors 61so inspected system components for damage or other conditions that could degrade system performanc The ESF system walked down during this inspection period and the documents utilized by the NRC inspectors during the walkdown are listed below:

System Documents High Pressure Coolant Injection (EM) Drawing M-12BN01, Revision 1, Borated Refueling Water Storage System P&ID(Q)

Drawing M-12EJ01, Revision 0, Residual Heat Removal System P&ID(Q)

Drawing M-12EM01, Revision 0, High Pressure Coolant Injection System P&ID(Q)

Drawing M-12EM02, Revision 1, High Pressure Coolant Injection System P&ID(Q)

Checklist CKL EM-120, Revision 5, " Safety Injection Systen Lineup Checklist"

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STS BG-002, Revision 4, "ECCS Valve Check and System Vent" STS EM-001, Revision 2, "ECCS Throttle Valve Verification" STS EM-003, Revision 2, "ECCS

. Flow Balance"

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STS EM-201, Revision 3, " Safety Injection System Inservice Valve Test" STS EM-202, Revision 0, " Safety Infection System Inservice Valve Test" Selected NRC inspector observations: The following discrepancy related to STS BG-002, Revision 4 was identified:

o Section 3.0, initial conditions, required the chemical and ,

volume control system (BG) and the residual heat removal (RHR)

system (EJ) to be lined up but did not require the high pressure coolant injection system (EM) to be lined u The following discrepancies related to Checklist CKL EM-120, Revision 5, were identified:

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o Breaker NG01BER2 for Valve EM HV-8801A, " Boron Injection TK Discharge Iso Viv,"~was labeled as EJ HV-8801 o Contrary to normal practice, equalize valve EM-C001 to FIS-949 and the calibration and vent valves, EM-C002 and EM-C003 to PT-947 were listed in CKL EM-12 o For the following valves the checklist and the P&ID did not *

agree on open/ closed status: .

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o EM-V015 o EM-V018 o EMHV-8870A o EMHV-88708 ,

o EMHV-8883 The following discrepancies related to STS EM-003, Revision 2, were identified:

o STS'EM-003 would only be performed, as required by TS,

"Following completion of modifications to the ECCS subsystems i

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-7-i that altea the subsystem flow characteristics . . . ." Yet Steps 5.1.18, 5.2.12, 5.3.10, and 5.3.22 only have the test performer " set and lock branch line throttle valves" and the steps fail to require that the test performer count turns so that the positions of the throttle valves could be verified later without performing another flow tes o Step 5.2.6 stated, " Start the CCP pump as determined in Step 3.6," when it should have stated to start the safety injection pum o The surveillance procedure failed to have the test performer determine the lowest capacity safety injection pump to start in Step 5. o Step 5.3.1 required the test performer to " install a 1550-3000 inch H2 O differential pressure gauge . . . .," when Step 4.1.3.1 stated to use "four 1500-3000 inch H2O 0/P gauges."

o Step 5.3.27 stated, "Close the flow element H.P. and L.P. test connection isolation valves," when it should have stated to close and lock the isolation valve o Step 5.4.14 has the test performer to close EJHV-87168 and isolate the cross-connect between "A" train cold leg discharge and "B" train cold leg discharg In addition to causing all of RHR to be inoperable, this would cause RHR cold leg injection to be tested in a configuration-different than what it is normally lined up. Step 5.4.27. required the same thing for "B" train pump testin o Steps 5.4.20, 5.4.31, 5.4.36, and 5.4.45 has the test performer record flow rate by using installed flow instruments _which were located upstream of the individual injection lines. Since this STS would'only be required to be performed by TS 4.5. following modifications that alter subsystem flow characteristics and TS 4.5.2.1 states to verify "the sum of the injection line flow rates . . .," then the individual injection line flows should be use o Step 5.4.35 stated to " shift RHR to the hot leg recirc. mode by performing the following valve lineup . . . ." If the test performer performed the valve lineup in the sequence written in the STS, then both RHR pumps discharge paths would be isolated from the reactor vessel and entry into a TS LCO would be require o Step 5.4.48 has a typographical error in that it has the test performer stop pump PGB05A instead of stopping pump PBG05 ; t^

-8-o ' Step 5.4.44 has the test performer to close EJHCV-607 and Step 5.4.45 to throttle EJHCV-606 until flow of 3200 i 100 gpm was reached; however, the procedure did not require the restoration of these valves to their normal open positions during the restoration of RHR to cold leg injection lineu O Step 5.4.53 tells the test performer to open BNHV-8313 when it should say open BNHV-881 The deficiencies described in c above have been discussed with the licensee and the licensee has agreed to correct the surveillance procedur No violations or deviations were identifie . 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 Technical Specification (TS) requirements and administrative procedures. The NRC inspectors considered the following elements while inspecting surveillance activities:

o Testing was being accomplished by qualified personnel in accordance with an approved procedur o The surveillance procedure conformed to TS requirement o Required test instrumentation was calibrate o Technical Specification limiting conditions for operation (LCO) were satisfie o Test data was accurate and complet Where appropriate, the NRC inspectors performed independent calculations of selected test data to verify their accurac o The performance of the surveillance procedure conformed to applicable administrative procedure o The surveillance was performed within the required frequency and the test results met the required limit Surveillances witnessed and/or reviewed by the NRC inspectors are listed below:

o STS IC-201, Revision 3, " Analog Channel Operational Test 7300 Process Instrumentation Protection Set I (Red)," performed on August 21, 198 . .- .--

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9-o STS EG-100A, Revision 0, " Component Cooling Water Pump A/C Inservice Pump Test," performed on August 19, 198 No violations or deviations were identifie . Followup On Regional Requests Class IE Battery Racks In response to a NRC Region IV request, the NRC inspectors followed up to determine if a battery rack problem, which had been identified at another nuclear plant, also existed at WCG The problem was that the spacing between the Class 1E battery end cells and the battery rack end support rails was greater than the allowable spacing used to seismically qualify the battery installatio The NRC inspectors determined by a review of documents and inspection of 1E battery installations NK11, NK12, NK13, and NK14, that corrective action had been taken at WCGS in May 1985, to adjust the battery rack rails ;o that seismic qualification tolerances were me The NRC inspector reviewed thc following documents:

o Wolf Creek Work Request No. 07958-85 o Plant Modification Request No. 00997, Revision 0, "NK System Battery Racks Must Be Inspected and Revorked, If Necessary, To Maintain Seismic Qualifications."

Emergency Diesel Generator Fuel Storage On August 14, 1986, the NRC inspectors received a request for information concerning questions raised as a result of emergency

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diesel generator (EDG) fuel oil contamination event at another nuclear plant. The problem was that, on June 26, 1986, following major preventive maintenance on EDG 2K4B, a 24-hour endurance load test was performed. Approximately 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> into the performance of this test, the EDG had to be secured due to high cylinder

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differential temperatures. Subsequent investigation determined the high cylinder differential temperature condition was caused by fuel starvation. The fuel starvation was caused by the day tank fuel supply outlet line y-strainer wire mesh cylinder being severely fouled / plugged and that the day tank bottom contat.1ed sludge deposits

- around the tank internal fuel suction line fuel " foot" valve / scree Following discussions with WCGS personnel, a review of the 15 requirements and of the surveillance procedures, the NRC inspector determined

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o New fuel oil was sampled in accordance with American Society For

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Testing and Materials (ASTM) Specification D4057-81 and D975-81

and that water and sediment requirements were satisfie o At.least once every 31 days a representative sample of fuel oil was analyzed for total particulate contaminatio o At least once every 31 days the licensee checked for and removed accumulated water from the' fuel oil storage tank botto o At least once every 31~ days the day tank was checked for, and had removed, accumulated water from the bottom.

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The licensee's program to ensure that accumulated sediment did not cause EDG fuel starvation appeared adequat No violations or deviations were identifie , Environmental Qualification (EQ) Of Electrical Equipment

! During an NRC inspection conducted May 12-16, 1986 (Inspection

. Report 50-482/86-14), the NRC inspector determined that a limitorque valve

was internally wired with unqualified wire and therefore did not meet

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specified requirements for equipment qualification in a harsh environmen This valve was one of 44 valves supplied to a specific specification by one vendor. -The licensee had verified that limitorque valves received on other specifications had met equipment qualification requirements by onsite walkdown inspections; however, for these 44 limitorque valve i

operators the licensee relied on the vendor's qualification program and

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therefore did not perform on-site field walkdown inspections. When the questionable wiring was identified, the licensee performed onsite

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inspections-(conducted May 16-18, 1986) of the 44 limitorque valves. The licensee also assembled a task force to scope the limitorque operator i inspection and rework. The' licensee identified and corrected the discrepancies- discussed below during this walkdown:

o Unqualified jumper wire in four operator _

o Unqualified terminal boards in two operators.

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o Sixteen AWG wire used rather than the specified 14 AWG wire in 12

operators.

I o Lugs crimped backwards on some of the removed jumper wires.

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, .During maintenance and inspection activities on essential service water

non-EQ operators, the licensee determined that undersized wire had been

used for vendor installed jumpers and some of the lugs had been crimped

.using a single crimped nonratchet type lugging tool.

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-11-Based on the above findings, the licensee expanded the scope of this environmental qualification of equipment activity, shut the plant down from Mode 1 to Mode 3, and replaced the internal wiring in 152 of 156 safety-related limitorque operators in the plant. The four encapsulated containment recirculation sump isolation valves which were not readily accessible were not inspected. A justification for continued operation (KG&E Letter No. KQWLKD 86-004, dated June 20,1986) was written to provide justification for not inspecting these four valves until the refueling outage scheduled to commence in October 198 The licensee performed an engineering evaluation of the internal wires removed from the 152 operators to determine if any discrepancies found in the removed wires could have potentially prevented the operators from functioning as designed during a post-accident environmen Wolf C~ reek Work Requests WR 70004-86, "Limitorque Valve Operators (Inside Containment)"andWR 70005-86, "Limitorque Valve Operators (Outside Containment)," were written to document the engineering evaluation of the internal jumper wires removed durino the complete rewiring (replacement of vendor supplied internal jumpers) of the limitorque operators. The types of problems identified and the licensee's engineering disposition of these problems are discussed below:

o Approximately 34(21)* lugs were installed using a "sta-kon" single crimp tool in lieu of the ratchet double crimping tool used onsite

! for installing similar lugs. Engineering evaluated the lugs on a case-by-ca,e basis and found them to be mechanically sound and to have had good electrical contact. Therefore, these lugs were determined to be acceptabl o Approximately 170(152)* lugs had been installed in the crimping tool backwards when they were crimped. These lugs were evaluated on a case-by-case basis and found to be mechanically sound and to have had good electrical contact. Therefore, these lugs were determined to be acceptable, o Approximately 290(179)* of the jumpers were 16 AWG wire rather than the specified 14 AWG wire. The smaller gauge wire was determined to be acceptable and would not have affected operation of the operator o Approximately 15(0)* Spade lugs installed on some of the jumpers were determined to be acceptable, o Approximately 41(40)* of the lugs had been trimmed changing them from ring type lugs to split tongue (spade) type lugs. Engineering evaluation stated that there was not enough information available to evaluate the performance of these lugs during conditions of tensile stress or acceleration (e.g., a seismic event). However, it was engineering's judgement that the lugs would have functioned properly during a seismic event, providing they had been installed using good workmanship practice <

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o Approximately 190(137)* of the jumper wires were made of unqualified (indeterminate) wire and therefore engineering could not endorse these jumpers for us *The number in parentheses is the approximate number of these items that were in Electrical Category A or B. The electrical equipment categories are:

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Category A - Item is required to function post acciden Category B - Item is not required to function post accident; however, it must not adversely fail post-acciden Category C - Item's status post accident has no effect on plant safet Category D - Item is in a mild environment post acciden The licensee's evaluation of the removed internal wires determined that the 27 valves listed below had indeterminate wiring in them and therefore the condition of the operators prior to their having been rewired was such that they may not have functioned as required in a post-accident environment. The primary concern being that some of the internal wires had teflon insulation which would deteriorate in the postulated post-accident high radiation fields:

o BG LCV-1128, Volume Control Tank To Coolant Charging Pump Suction

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o BG LCV-112C, Volt.me Control Tank To Coolant Charging Pump Suction

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o BG HV-8105, Charging To Regenerative Heat Exchan0er o BG HV-8106, Charging To Regenerative Heat Exchanger o BN HV-8812A, Refueling Water Storage Tank To Containment Spray

, Pump "A" o BN HV-003, Refueling Water Storage Tank To Containment Spray Pump "B" o BN HV-004, Refueling Water Storage Tank To Containment Spray Pump "A"

o EG HV-058, Component Cooling Water To Reactor Cooling Pumps o EJ HV-8716A, Residual Heat Removal Cross-Connect

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o EJ HV-8716B, Residual Heat Removal Cross-Connect o EJ HV-8804A, Residual Heat Removal To Safety Injection Pump Suction "A"

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,' i-13-o EJ HV-8804B, Residual Heat Removal To Safety Injection Pump Suction "B" o EJ HV-8809A, Residual Heat Removal Discharge To Cold Leg o EJ HV-8809B, Residual Heat Removal. Discharge To Cold Leg o EJ HV-8840, Residual Heat Removal Discharge To Hot Leg o EM HV-8801A, Boron Injection Tank Discharge o EM HV-8801B, Boron Injection Tank Discharge o EM HV-8802A, Safety Injection Discharge To Hot Legs o EM HV-88028, Safety Injection Discharge To Hot Legs o EM HV-8803B, Coolant Charging Pump "B" to Boron Injection Tank .

o EM HV-8821A, Safety Injection Cross-Connect o EM HV-88218, Safety Injection Cross-Connect o EM HV-8835, Safety Injection Discharge To Cold Legs o EN HV-006, Containment Spray Pump "A" Discharge o EN HV-012, Containment Spray Pump "B" Discharge o EN HV-015, Spray Additive To Containment Spray Pump "A" o EN HV-016, Spray Additive To Containment Spray Pump "B" The licensee was evaluating each of these 27 valves to determine if the post accident environment postulated in the original equipment qualification was more restrictive than the anticipated post accident environment for each operator locatio All the limitorque operator wiring deficiencies identified to date have been confined to vendor supplied internal jumpers. No deficiencies have been identified with field installed wirin Pending completion of the licensee's engineering evaluation and further NRC review, this is an unresolved item (482/8618-03).

8. Fire Protection / Prevention Program This inspection was conducted to determine whether the licensee has implemented a program for fire protection and prevention in conformance with regulatory requirements and industry guides and standard .

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-14-The NRC inspector reviewed the documentation constituting the licensee's approved fire protection program. The licensee's program provides for administrative controls of combustible materials and housekeeping for the reduction of fire hazards; handles disarmed or inoperable fire detection or suppression systems; provides for maintenance and surveillances on fire suppression and detection equipment; establishes personnel fire fighting qualification, training and fire protection staff responsibilities; provides fire emergency personnel designations as well as plans and actions; and establishes controls for welding, cutting, grinding, and other ignition sources. The following WCGS administrative procedures were reviewed:

o ADM 13-101, Revision 3, " Control of Ignition Sources," September 13, 1985 o ADM 13-102' Revision 5, " Control of Combustible Material," May 14,

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1986 o ADM 13-103, Revision 3, " Fire Protection: Impairment Control,"

October 24, 1985 o ADM 13-104, Revision 1, " Development and Use Of Fire Preplans,"

-January 6, 1986 o ADM 13-200, Revision 1, " Fire Protection Training Program,"

December 3, 1985 o ADM 13-901, Revision 0, " Fire Preplans Preparation Guide," April 15, 1984 o ADM 13-902, Revision 2, " Fire Drills," December 3, 1985 o ADM 13-903, Revision 0, " Fire and Life Safety Inspections," June 27, 1984 The NRC inspector found that Administrative Procedures ADM 13-101, ADM 13-102, and ADM 13-103 require all plant employees to report unsafe fire conditions, poor housekeeping practices, and fire protection impairments. The only way all employees would be aware of this requirement would be through the General Employee Training (GET) cours The NRC inspector reviewed both the GET and GET requalification student handouts. The GET course did instruct employees to report fire hazards, but the GET requalification course only instructed employees to report impaired fire barriers. This was mentioned during the exit interview on July 17, and the licensee stated that they would include the instruction in the GET requalification course for fire hazards reporting that is in the GET cours This adequately resolved this NRC concer There were no violations or deviations identified in this are _ - -

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, -15-The NRC inspector reviewed the following surveillance tests for completeness, required frequency, and commitments to TS requirements:

o STS FP-001, " Water Supplied Fire Protection Valve Position -

Verification," (September 1985 - June 1986)

o STS FP-002, " Yard Loop and Hydrant Flush and Hydrant Inspection,"

(December 1985 and May 1986)

o STS FP-004, " Fire System, Flow Test, Pump Sequential Start, and Annual Fire Pump Test," (December 1985)

o -STS FP-005, " Monthly Sprays / Sprinkler Valve Position Verification,"

(September 1985 - May 1986)

o STS FP-006, " Sprays and Sprinklers-18 Month Operational Test,"

(May 1986)

o STS FP-008, " Monthly Hose Rack Inspection," (September 1985 -

June 1986)

o STS FP-013, " Cycle Procedure for Testable Valves," (September 1985)

o STS FP-019, " Fire Door Visual Inspection," (January 1986)

o STS FP-020, " Trip Actuating Device Operational Test for Electrically Supervised Fire Doors," (September 1985 - June 1986)

o STS FP-021, " Fire Door Position Verification - Closed, Locked, Unalarmed," (June 1986)

o STS FP-024, " Fire Door Position Verification - Closed, Locked, Unalarmed," (May 1986)

o STS FP-601, " Diesel Fire Pump 1FP01PB - Monthly Operation and Fuel Level Check," (September 1985 - June 1986)

o STS FP-602, " Electric Motor Driven Fire Pump 1FP01PA - Monthly Operation," (October 1985 - June 1986)

o STS CH-007, " Fire Pump Diesel Fuel Storage Tank," (October 1985 -

April 1986)

o STS MT-013, " Fire Pump (Diesel) Battery Electrolyte and Voltage Inspection," (May 1986)

o STS MT-014, " Fire Pump (Diesel) Battery Specific Gravity,"

(November 1985 - May 1986)

o STS MT-031, "Halon System Tank Weight and Pressure," (February 1986)

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-16-The inspection of the above surveillance tests resulted in the following observation:

Surveillance Procedure STS FP-004 required that the calibration due date for the pitot tube and gauge be entered in the appropriate space in Section 4.1.3. However, for the fire system water flow test performed on

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December 5, 1985, the next calibration due date was entered as April 8, 1985. This appeared to indicate that an uncalibrated pitot tube and gauge was used for the test. The licensee investigated this finding and verified that the instrument used was calibrated at the time of the test and that the 1985 date was entered instead of 1986. This explanation was acceptable to the NRC inspector. However, it was mentioned to the licensee at the exit meeting that even though the instrument was calibrated the error was not detected in the review of the completed test documentatio The licensee noted this concern which was acceptable to the NRC inspecto There were no violations or deviations identified in this are The NRC inspector conducted a walkdown of the fire suppression water system for the circulating water house, east diesel generator room, and engineered safety features transformer and verified that they were operable as required by T A tour of accessible areas of the plant was conducted to verify that standpipe and hose stations were operable; adequate portable fire extinguishers were provided at designated places; and access to fire suppression equipment was not being restricted by any materials or equipmen Inspections and maintenance on all fire suppression equipment or devices were verified as being satisfactorily performed, and the general condition was satisfactory. During this tour the NRC inspectors discovered three hose stations with trash stuffed inside of the hose covers. The trash included masking tape and a rag. The licensee committed to determining whether trash has been discovered during their monthly hose rack inspections and to take corrective action as neede This adequately satisfies the NRC inspector's concer The NRC inspector also observed the condition of fire barrier penetrations during the plant tour. The wall and floor penetration seals were found to be functional, fire doors were found closed, as required, and the closing and locking mechanisms were functiona There were no violations or deviations identified in this are A review of the licensee's annual QA Audit (TE: 50140-K082, dated December 4, 1985) of their fire protection program was performed. The audit was found to meet CMEB 9.5.1 requirements and to be thorough in both scope and dept There were no violations or deviations identified in this are r _ _

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-17- Exit Meetina The NRC inspectors met with licensee personnel to discuss the scope and findings of this inspection on August 4, 1986. The NRC inspectors also attended entrance / exit meetings of other NRC region based inspectors identified below:

Inspection Lead . Area Inspection Period Inspector Inspected Report N /11-15/86 R. Mullikin Fire Protection 86-18 Followup (par. 8)

8/11-15/86 R. Caldwell Security 85-19 8/25-29/86 C. Hackney Emergency 86-20 Preparedness