IR 05000298/1986015

Insp Rept 50-298/86-15 on 860421-26.Violations Noted Failure to Provide Tech Spec Required Fire Watches & Failure to Have Procedure That Properly Implements Requirements of Fire Protection Tech Spec

ML20203F782
Person / Time
Site:	Cooper
Issue date:	07/16/1986
From:	Hunter D, Jaudon J, Mullikin R, Murphy M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20203F737	List: ML20203F734 ML20203F753 ML20203F782
References
50-298-86-15, TAC-61117, NUDOCS 8607310178
Download: ML20203F782 (25)

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APPENDIX B U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

NRC Inspection Report: 50-298/86-15 License: DPR-46 Docket: 50-298 Licensee: Nebraska Public Power District (NPPD)

P. O. Box 499 Columbus, Nebraska 68601 Facility Name: Cooper Nuclear Station (CNS)

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Inspection At: CNS Site, Brownville, Nebraska Inspection Conducted: April 21-26, 1986 Inspectors:

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[ 7It!r 'R.' P. Mul fikin,/Pr@ct Inspector, Project Dste'

Section B, Reactor Projects Branch

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/ ~7bf N K. E. Nurphy,/ Project #Tnspector, Project Date '

Section B, Reactor Projects Branch Participating -

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inspection: D..Notley, Office of Nuclear Reactor Regulation J. .Kudrick, Office of Nuclear Reactor Regulation f

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A. Coppola, Brook en Nat onal Laboratory

} K. Parki so , Br haven tional Laboratory i

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./ .Waudo , tiiief, Project Section A, Date l [Ractor ojects Branch

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D. R. Hunter, Chief, Project Section B, Date Reactor Projects Branch h

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Inspection Summary Inspection Conducted April 21-25, 1986 (Report 50-298/86-15)

Areas Inspected: Nonroutine, announced inspection for implementation of and l compliance to the safe shutdown requirements of 10 CFR 50, Appendix Results: Within the areas inspected, three violations were identified (failure to provide TS required fire watches, paragraph 3; failure to have a procedure that properly implements the requirements of the TS, paragraph 3; and failure to have a procedure that properly identifies, installs, and provides acceptance l criteria, paragraph 3.) Five unresolved items are identified in paragraphs 6.b, 6.c, .e, and 8.

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DETAILS

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NPPD

• G. R. Horn, Division Manager, Nuclear Operations .

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• W. Swantz, Senior Engineer

• K. Walden, Electrical /I&C Supervisor

• D. Danielson, Electrical Engineer

• J. Hackney, Lead Electrical Engineer

• P. Burrows, Fire Protection Coordinator

• E. M. Hace, Plant Engineering Supervisor

• J. M. Meacham, Technical Manager .

• V. L. Wolstenholm, Quality Assurance Manager, CNS -

• J. V. Sayer, Technical Staff Manager

• H. T. Hitch, Acting Administrative Services Manager

• C. R. Goings, Regulatory Compliance Specialist

• Crawford, Maintenance Supervisor M. Ward, Shift Supervisor ,

W. Schrader, Operations Engineer L. Bednar, Senior Staff Engineer M. Span, Assistant to Operations Manager R. Alexander, Lead Electrician

• J. Willis, Draftsman

• D. Fitzgerald, Draftsman

• F. Alderman, Fire Protection Specialist

• T. A. Wilson, Mechanical Engineer ..

• R. Brungardt, Operations Manager ,

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Other Licensee Personnel

• S. Burke, Project Engineer, Engineering Planning and Management (EPM)

• K. Cloran, Electrical Engineer, EPM

• R. Lemos, Appendix R Project Engineer, EPM , :

A. Morisi, Electrical Engineer, EPH {

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• Denotes those attending the exit interview cor. ducted on April 25, 198 The NRC inspectors also interviewed other CNS personnel during the inspectio !

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' List of Documents Reviewed Letters, Reports, and Procedures Title Date CNS/NRC Record of telephone conversation for clarification of 04/30/84 10 CFR 50, Appendix R Safety Evaluation Report Cooper Nuclear Station Critical AC Bus Coordination Study 10/85 Volumes I and II

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Cooper Nuclear Station Critical DC Bus coordination Study 02/86 Volume III Maintenance Procedure 7.3.1, Revision 9, Protective Relays 11/14/85 Setting and Testing Maintenance Procedure 7.3.2, Revision 9, Low Voltage Circuit 09/15/85 Breakers, Setting, Testing, and Maintenance MDC No. 84-7, Appendix R - Fire Protection for the Cable 05/17/84 Expansion Room MDC No.84-180 Diesel Generators - Addition of Isolation 04/22/85 Switches to Engine Panels MDC No. 84-93 Fire Dampers: DC Switchgear Rooms 05/23/84 MDC No. 84-5, Appendix R - Fire Protection for the Cable 06/07/84 Spreading Room MDC No. 84-8 Control Building Basement Fire Barriers to 03/21/84 Protect 125 VDC and 4160 VAC Cables MDC No.84-006 Cable Expansion Room - Fire Barrier 05/10/84 f MDC No.84-004 Cable Spreading Room Fire Barriers to Protect 04/26/84 125/250 VDC and 4160 VAC Cables

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MDC No. 85-01 Halon'1301 Fire Suppression System for Service 03/19/85 Water Pump Room and Fire Door Addition MDC No. 85-01, Revision 1, Installation of Fire Doors, Da.mper 04/18/85 and Breathing Sets Associated with the Halon Fire Suppression

. System.in S.W. Pump Room j Appendix R Associated Circuits of Concern 03/86 Selection of Cables Associated with Appendix R Safe Shutdown 11/85 Components, Volumes 1 and 2

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Fire Hazards Analysis 09/85 Response to 10 CFR 50, Appendix R, " Fire Protection of Safe 12/02/83 Shutdown Capability - Volume III" Appendix "R" Alternate Shutdown System Basis of Design 12/85 Document, NED BODD No. 85-02, Revision 0 NRC letter to CNS, Safety Evaluation for Appendix R to 04/16/84 10 CFR Part 50, Items II.G.3 and III.L, Alternate or Dedicated Shutdown Capability CNS letter to NRC, Appendix R - Schedular Exemptions; Request 06/07/85 for NRC letter to CMS,' Outstanding Fire Protection Modifications 08/21/85 CNS letter to NRC, Appendix R - Analysis of Cooper Nuciear 05/09/85 Station ,

Procedure 5.4.1, Revision 19, Diesel Fuel Oil Transfer Pump 10/03/85 Repair Procedure for Battery Charger and Exhaust Fan Repair Current Revision

" Fire Protection of Safe Shutdown Capability" CNS response to 06/28/82 10 CFR 50, Appendix R, Volumes I and II CNS Emergency Procedure 5.4.1 " General Fire Procedures," 09/30/85 Revision 19 CNS Emergency Procedure 5.8 " Emergency Operating Procedures" Current E.0.P. Sections 1 through 12 Revision NRC letter to NPPD re Exemption Requests 09/21/83

" Report on Core Uncovery due to Depressurization" - for CNS 07/15/85 by EPM

" Report on emergency lighting, alternate shutdown equipment 03/10/86 accessibility and portable communications systems" - for CNS

- by EPM Maintenance Procedure 7.3.12, Revision 3, Emergency Lighting 03/11/85 Units Inspection

. CNS Procedure 0.16, Revision 1, Control of Fire Doors 12/19/84

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CNS Procedure 0.23, Revision 0, Fire Protection Plan 08/08/85 CNS Procedure 2.2.30, Revision 24, Fire Protection System 10/17/85 CNS Procedure 2.2.72, Revision 4, Smoke, Temperature, and 02/29/84 Flame Detection CNS Procedure 2.3.2.37, Revision 8, Fire Protection - 02/20/84 Annunciator 1 CNS Procedure 2.3.2.38, Revision 5, Fire Protection (Manual 05/02/85 Pull Alarms) - Annunciator 2 CNS Procedure 2.3.2.39, Revision 5, Fire Protection 01/16/86 (Sprinkler System Actuation and CO2 ) - Annunciator 3 CNS Procedure 2.3.2.40, Revision 10, Fire Protection - 10/30/85 Annunciator 4 CNS Procedure 2.3.2.40, Revision 5, Fire Protection - 03/27/86 Annunciator 5 CNS Procedure 2.3.2.54, Revision 0, Pump House Fire Detection 04/17/84 Panel FP-PNL-5 CNS Procedure 2.3.2.55, Revision 0, Pump House Local Control 04/17/84 Panel FP-PNL-4 CNS Procedure 3.6.1, Revision 2, Fire Barrier Seal Activities 04/10/86 Control CNS Procedure 5.~4.2, Special Fire Procedures (5.4.2.1 thru 5.4.2.28) (Reviewed Selected Procedures)

l CNS Procedure 6.4.5.1, Revision 38, Fire Protection System 11/07/85 Annual Inspection CNS Procedure 7.10, Revision 4, Flame Process Control 10/18/85 Drawings l Title Date

! Flow Diagram - Residual Heat Removal System No. 2040, 02/11/74 l Revision 13 Flow Diagram - Reactor Core Isolation Coolant System 04/15/74 No. 2043, Revision 14 l

l Flow Diagram - Reactor Building - Main Steam System 03/05/78 No. 2041, Revision 21

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Flow Diagram - High Pressure Coolant Injection System 04/15/85 No. 2044, Revision 16 Flow Diagram - Core Spray System No. 2045, Revision 18 01/24/75

' Flow Diagram - Reactor Building - Service Water System, -- -- 74 Revision 20 125-DG2-1,-125 Volt DC Panel DG2-1 Fuel Oil Booster Pump 11/15/85 125-DG2-7, 125 Volt DC Panel DG2-7 DG2 Exciter Panel 11/15/85 125-PNL-AA2-10, 125 V. DC PNL AA2-10 11/20/85 250-SWGR-18-1 (Rec.), 250 V. DC SWGR 1B-1 (Rec.) 03/05/86 250-SWGR-1B-4 (Rec.), 250 V. DC SWGR 1B-4 (Rec.) 03/05/86 4160-1G-SWP1B(1), 4160 SWGR 1G - Breaker SWP-1B 10/04/85 4160-1G-SWP1B(2), 4160 SWGR 1G - Breaker SWP-1B 10/04/85 4160-1G-SWPIB(3), 4160 SWGR 1G - Breaker SWP-1B 10/04/85 I 4160-1G-RHRP10(1), 4160 SWGR 1G - Breaker RHRP-10 10/04/85 4160-1G-RHRP1D(2), 4160 SWGR 1G - Breaker RHRP-1D 10/04/85 4160-1G-RHRPID(3), 4160 SWGR 1G - Breaker RHRP-10 10/04/85 4160-1G-SS1G(1) (Rec.), 4160 Volt Bus 1(G) (Rec.) 09/20/85 4160-1F-SS1F(1) (Rec.), 4160 Volt Bus IF (Rec.) 09/20/85 480-1F-MCC-L, 480 V. Bus IF - Feeder MCC-L 07/24/85 480-1F-MCC-L (Rec. ), 480 V. Bus IF - Feeder MCC-L(Rec. ) 09/20/85 240-DP15-1A-1, DP15-1A Circuit 1 08/27/85 3A, Revision 3, Cooper Nuclear Station Appendix "R" 04/01/86 Circuit Separation 3B, Revision 3, Cooper Nuclear Station, Appendix "R" 04/01/86 Circuit Separation 4A, Revision 2, Cooper Nuclear Station, Appendix "R" 04/01/86 Circuit Separation 4A, Revision 2, Cooper Nuclear Station Appendix "R" 04/01/86 Circuit Separation

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48,. Revision 2, Cooper Nuclear Station, Appendix "R" 06/17/86 Circuit Separation 7A, Revision 3, Cooper Nuclear Station, Appendix "R" 04/01/86 Circuit Separation 78, Revision 3, Cooper Nuclear Station, Appendix "R" 04/01/86 Circuit Separation 8A, Revision 3, Cooper Nuclear Station, Appendix "R" 04/01/86 Circuit Separation 88, Revision 3, Cooper Nuclear Station, Appendix "R" 04/01/86 Circuit Separation CNS-1000, Revision 1, One Line Diagram 125 VDC Alternate 03/06/86 Shutdown CNS-1001, Revision 1, Loop Diagram HPCI - Pump Discharge 03/06/86 Pressure Indication CNS-1002, Revision 1, Loop Diagram HPCI - Turbine Steam 03/06/86 Inlet Pressure CNS-1003, Revision 1, Loop Diagram HPCI - Pump Suction 03/06/86 Pressure Indication CNS-1004, Revision 1, Loop Diagram HPCI - Turbine Speed 03/06/86 Indication CNS-1005, Revision 1, Loop Diagram HPCI Turbine 125V DC to 03/06/86 Register Box CNS-1006, Revision 1, Loop Diagram HPCI Flow Cont. & In /06/86 CNS-1007, Revision 1, Loop Diagram RHR Flow Indication & Loop 03/06/86 CNS-1008, Revision 1, Loop Diagram Reactor Vessel Level 03/06/86

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-150" - +60" CNS-1009, Revision 1, Loop Diagram Reactor Vessel Level 03/06/86-100" to +200" H 2O CNS-1010, Revision 1, Loop Diagram Suppression Chamber Water 03/06/86 Level

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CNS-1011, Revision 1, Emerg. Condensation Storage Tank 1B 03/06/86 Level Indication CNS-1012, Revision 1, Loop Diagram Torus Temperature 03/06/86 Indication

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Dwg. 3144, Revision 3, Reactor Building - El. 931'-6" Current Lighting Plan Revision 3. Fire Protection / Prevention Program This inspection was conducted to determine that the licensee was implementing a program for fire protection and prevention in conformance with regulatory requirements and industry guides and standard The NRC inspector reviewed the documentation constituting the licensee's approved fire protection program. These documents are referenced in paragraph 2 of this report. The licensee's program provides for the control of combustible materials and housekeeping for reduction of fire hazards. Administrative controls have been established to handle disarmed or inoperable fire detection or suppression systems; provide for maintenance and surveillances on fire suppression, detection, and emergency communications equipment; establishes personnel fire fighting qualifications, 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 source The NRC inspector conducted a walkdown of the fire suppression water system and verified that it was operable as required by technical specification 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 in each fire zon Access to fire suppression devices is not being restricted by any materials or equipment. Inspections and maintenance on all fire suppression equipment or devices were verified as being satisfactorily performed, and the general condition is satisfactor The NRC" inspector also observed the condition of fire barrier penetrations during-this tour. The closing mechanism for the fire door separating the turbine building from the control room access corridor was found to be weak.and would not always provide for positive latching of the door. This was brought to the attention of a licensee representative and action to l fix this door was initiate Locking mechanisms were found removed from

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the access doors to the auxiliary relay room, reactor protection system room 1B and room 1A in the control building, elevation 903'-6". The NRC inspector asked the licensee representatives if a continuous fire watch had been established, since the removal of these mechanisms compromises the fire rating of the door The NRC inspector was informed that the licensee considered the roving fire watch adequate. This fire watch had

! been established during re-work in this area to complete modification commitments under the Appendix R exemption request The NRC inspector

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informed the licensee that the applicable technical specifications, l paragraph 3.19.8 does not recognize any other condition and only a

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continuous fire watch satisfies the technical specification LC The I

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licensee posted a continuous fire watch. Failure to establish a continuous fire watch is an apparent violation of technical specifications. (298/8615-06)

A review of the licensee's Procedure 0.16, " Control of Fire Doors" revealed that the procedure defines various categories of doors and further describes the requirements for posting or not posting a fire watch or security guard on the various door categories. CNS Technical Specifications LCO 3,19 states that it applies to the integrity of all fire barrier and fire wall penetration fire seals and that " Fire barrier and fire wall penetration fire seals integrity shall be maintained. ", and "B. If the requirement of 3.19.A cannot be met, a continuous fire watch shall be established on at least 1 side of the penetration within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />." The bases for 3.19 states that, " Fire barrier penetration seals include cable penetration barriers, fire doors, and fire dampers." Failure to have a procedure that properly implements the requirements of the technical specifications is an apparent violatio (298/8616-07)

In the same corridor on the same elevation, fire door H109 was found to have a door to floor gap in excess of the 3/4" allowed by NFPA-80. The inactive side of this double door was installed by work item No. 86-0692 dated February 12, 1986. A review of this work item disclosed that it contained no acceptance criteria, did not define an installation tolerance, was not identified as a technical specification item, was not identified as a fire penetration and was identified as non-essentia This is a 3-hour rated fire door for DC switchgear room 1 Failure to have a procedure that properly identifies, installs, and provides acceptance criteria is an apparent violation of 10 CFR 50, Appendix B, Criterion V. (298/8615-08)

A review of surveillance records verified that the fire detection and suppression systems currently meet the technical specification operability testing requirements and that they are being conducted at the required frequencie Fire brigade training and drill records were reviewed as well as selected personnel record Individual qualifications and training were found to meet CMEB 9.5.1 requirements. A review of the current roster of qualified fire brigade members verified that brigade composition is in accordance with technical specification requirement . Emergency Lighting System The NRC inspector examined.the emergency lighting system required for safe shutdown. Section J of. Appendix R requires that emergency lighting units with at least an 8-hour battery power supply be provided in all areas needed for operation of safe shutdown equipment and in access and egress routes thereto. The licensee had not installed, at the time of this

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-inspection, all of the emergency lighting units required for the operation

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of safe shutdown equipment in the event of a fire that destroys and forces evacuation of the control room. The NRC inspector did not inspect for adequacy of emergency lighting in these area The NRC inspector reviewed the maint'enance procedure for checking emergency lighting units. This procedure (7.3.12, " Emergency Lighting Units Inspection," Revision 3, March 11,1985) outlines the requirement to check the operation of the units with the test switch and check the terminals for corrosion. It was found that there did not exist a procedure to periodically check the lamps for adequate alignmen Although Appendix R does not specifically state this as a requirement it is recommended that the licensee consider incorporating this item into a procedur Also reviewed was the manufacturer's data for the emergency lights. The Exide Electronics 1983-84 emergency lighting catalog shows that for Exide Model F-100 units, two 12 watt lamps will provide at least 85 percent illumination after 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The NRC inspector had the licensee perform an 8-hour battery service test on two lighting units (R-38 and R-42) located in the reactor building on elevation 931'-6". The inspector performed a visual check of the. illumination at the beginning and end of the tes There was no discernible reduction in illumination after 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of continuous discharg Post-Fire Safe ~ Shutdown Capability

~ Systems Required for Safe Shutdown

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The systems required for safe shutdown as essential are listed in the SER of April 16, 1984, and the licensee's submittals (see paragraph:2), but neither of these documents detailed the systems required by fire area or zon The.following paragraphs list the systems used for safe shutdown by l

i fire areas or zones according to the current analysis (dated April.1;8, 1986) produced by the licensee's consultant (EPM, Inc.).

(1) Reactivity Control Upon detection of a disabling fire, the control rods will be inserted using the scram switches in the control room, or automatically by the reactor protection system (RPS) upon loss of offsite powe This is true for a fire in any fire zone or are If neither action occurs, scram can be accomplished outside of the control room by opening the RPS MG set output breaker '

(2) Reactor Coolant Inventory Control and Decay Heat Removal For most fire areas or zones, (which do not require alternate

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shutdown), the reactor core isolation cooling (RCIC) system will f

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4~ be used.for. reactor coolant make-up. The source of water will be both the emergency condensate storage tanks (2), and the

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-suppression pool. The capability to switch suction sources will

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- ;for reactor make-up, and a combination of automatic depressurization system (ADS) and low pressure injection will be

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used. For fire area RB-A, the low pressure coolant injection

'(LPCI) system (train A) will be used for coolant make-up and the source of water will be the suppression pool. For fire areas, CB-A, CB-C, RB-E, and RB-I, the core spray system will be used

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for coolant make-up and the source will be the suppression poo High pressure coolant injection (HPCI) is not used for coolant inventory control except for those areas requiring alternate shutdow For all fire areas, decay heat removal is accomplished by suppression pool cooling using either train A or B of the residual heat removal (RHR) pumps, and a minimum of one vessel to the suppression pool as steam via the RCIC exhaust or the ADS system. This method of decay heat removal will be continued until the shutdown cooling mode of the RHR can be placed in service (~50 psig in the reactor vessel).

(3) Process Monitoring For fires requiring control room evacuation, the following instrumentation will be provided on the alternate shutdown panel:

o Reactor Pressure (HPCI Steam Inlet)

o Reactor Level o Torus Level (suppression pool)

o ECST Level (condensate storage)

o Torus Temperature (4) (suppression pool)

Diagnostic instrumentation for the HPCI system, including HPCI speed, water suction and discharge pressure, and HPCI flow will also be available at the alternate shutdown pane (4) Support System and Equipment For the safe shutdown systems described above, the following

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o Diesel Generator - at least one of two for RHR and SW pump powe ., . -_ _ ..

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o Service Water - at least one of four pumps for diesel engine cooling, REC (reactor building closed cooling system), and RHR heat exchange o REC - at least one train for RHR pump cooling and room coolers, or as an alternative the REC /SW intertie, allowing service water to be used instead, o Emergency switchgear one train for pump powe o D.C. batteries and battery charger - one trai (5) Cold Shutdown Cold shutdown is achieved by operating the RHR system in the shutdown cooling mode once the reactor pressure has been reduced to 50 psig. The present EPM study shows that the RHR suction valves (from the vessel) could fail to open on demand for certain fire In this case, cold shutdown will be achieved by filling the reactor vessel with water and discharging the water to the suppression pool (torus) via the ADS valves. Thus, a closed loop is created, transferring heat to the suppression pool. One train of core spray or one RHR pump is used for this purpose, and one RHR pump / heat exchanger is operated in the suppression pool cooling mode to transfer the heat to the ultimate heat sink which is the Missouri Rive Area Compliance with Appendix R,Section III. Although a general tour of all available fire areas pertaining to safe shutdown was conducted by the team, both together and individually, no conclusions can be reached regarding area compliance until the licensee's analysis is complete At the time of the inspection, the licensee was in the process of completing the associated circuit analysis. To determine the depth / effectiveness of the ongoing analysis, a limited random sample of cables was examined in the field and compared to the analysis data. Sample results were as follows:

Fire Area IS-A, Service Water Pump Area, contains the following redundant safe shutdown components:

Safe Shutdown Equipment Division Type Cable Cable Number Service Water Pump I Power H401 SW-P-A

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Service Water Pump II Power H521 SW-P-B Service Water Pump I Power H411 SW-P-C Service Water Pump II Power H531 SW-P-D The loss of cables H401, H521, H411 and H531/will cause a loss of service wate The pumps and cables are not in compliance with the separation requirements of Section III.G.2. The associated circuit analysis data documents the status of the cables and the existence of an exemption from the noncomplianc Fire Area RB-J, SWGR Room IF, and Fire Area RB-K, SWGR Room 1G, contain cabling for the following redundant safe shutdown components:

Safe Shutdown Equipment Division Type Cable Cable Number RHR Fan Coil Unit I & II Control MK 119 FC-R-IJ RHR Fan Coil Unit I & II Power MS 101 FC-R-1H These cables are located in both Fire Areas RB-J and RB-K. Fan coil unit FC-R-1H has start permissive contacts in RHR-P-B and RHR-P-D breaker close circuits. Fan coil unit FC-R-1J has start permissive contacts in RHR-P-A and RHR-P-C breaker close circuits. The loss of control cable MK 119 and power cable MS 101 will cause a loss of remote starting of all redundant RHR pumps. Cables MS 101 and MK 119 are not in compliance with Section III.G.2 separation requirement The associated circuit analysis data documents the status of cables MS 101 and MK 119. Resolution is pendin Fire Area CB-A, RHR Service Water Booster Pump and Service Air Compressor Areas, contains cabling for the following redundant safe shutdown components:

Safe Shutdown Equipment Division Type Cable Cable Number Service Water Pump I Power H401 SW-P-A

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j 15 Service Water Pump II Power H521 SW-P-B Service Water Pump I Power H411 SW-P-C j Service Water Pump II Power H531

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SW-P-D.

Diesel Generator' I Power DG1A, DG1B DG1

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Diesel Generator II Power DG37A, DG37B

! The loss of cables H401, H521, H411 and H531 will cause a loss of

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service water. The loss of cables DG1A, DG18, DG37A and DG378 will cause a loss of safe shutdown emergency power. These cables are not in compliance with Section III.G.2 separation requirements. The associated circuit analysis documents the status of the cables and the existence of an exemption from the noncompliance.

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Fire Area CB-8, Battery Room IB, contains cabling for the following  !

p redundant safe shutdown components:

Safe Shutdown

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Service Water Pump I Power H401 i SW-P-A i

Service Water Pump II Power H521 i

SW-P-B

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' Service Water Pump I Power H411 SW-P-C

! Service Water Pump II Power H531

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Diesel. Generator I Power DG1A, DG1B f DG1 7 (DieselGenerator II Power DG37A, DG37B

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.; Fire Area CB-B was found to be in compliance with the separation

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c. Alternate Shutdown -

R (1) Areas Requiring Alternate Shutdown-The areas requiring alternate shutdown have not changed due to the new EPM analysis. They include the following:

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o Control Room (Fire Area CB-D) *

o CableSpreadingRoom(FireAreaCB-D] *

o Cable Expansion Room (Fire Area CB-D, o Aux. Relay Room (Fire Area CB-D)

o Computer Room (Fire Area CB-D)-

o Reactor Bldg. 903' Elev. , N.E. corner (Fire Area RB-FN)*

In three of these areas, (noted by asterisks above), cabling for both trains of diesel generator power (output), and both trains of service water pump power are present. These cables, and a high percentage of all other cabling at Cooper are run in conduit. Since these are essential for safe shutdown and for alternate shutdown, they were brought to the inspection team's attention. Exemptions to the Appendix R requirements had previously been granted by the NRR fire protection reviewers on the basis of low combustible loading and adequate protection from a floor based exposure fire. The inspection team's fire protection specialist reviewed these areas and the findings are discussed in paragraph (2) Systems Used for Alternate Shutdown The EPM analysis and the design of the alternate shutdown system were reviewed. They were consistent and indicated that the HPCI system will be used for inventory control, with both the emergency condensate tanks (ECTs), and the suppression pool used as the source of coolant. The suppression pool cooling made of RHR will be used for decay heat removal. All of the support systems (one train), indicated in paragraph 5.a, will be required, and the shutdown cooling mode of RHR will be used for cold shutdown as described in paragraph The instrumentation listed in paragraph 5.a will be available on the alternate shutdown panels, as well as the necessary controls for HPCI, RHR, and three ADS valve (3) Modifications Required for Alternate Shutdown (New Panels)

The design of the alternate shutdown modifications is in the final stages of completion, with all parts ordered, but installation is just beginning. A schedule was presented by the licensee which showed completion by the end of the next refueling outage (end of December 1986).

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There will be three new control panels installed.in the southeast corner of the reactor building at elevation 903'. The three panels, one each for the HPCI, RHR.and ADS system, .will be housed in a new enclosure. This enclosure (room) will be accessible in two ways: One from the reactor building (elevation 903'), and from the outside using the reactor building roof and a caged ladde The panels will be equipped with isolation switches for all of the equipment controlled to ensure independence from the control roo . Procedure Safe Shutdown Procedures The licensee presently uses symptom-oriented procedures for shutdown in case of any transient or emergency, including fire. These procedures do not depend on any particular event or malfunction and can be used for shutdown in case of fire that does not require control room evacuation. The procedures reviewed included the following:

o 5. General Fire Procedure o 5.4. Battery Room Fire o Emergency Operating Procedures (E.0.P.)

o E.0.P.-C Operator Precautions o E.0.P.-1 Reactor Pressure Control Possible Core Uncovery Using ADS / Low Pressure Systems for Inventory Control Since for the five areas outlined in paragraph 5.a. the licensee intends to depressurize the vessel by using the ADS system and then inject coolant via the low pressure systems (core spray or LPCI), the possibility of core uncovery was reviewed. The licensee presented an analysis entitled " Report on Core Uncovery Due to Depressurization Using ADS in conjunction with the Core Spray System" dated July 15, 1985, by EPM. This report indicated that if depressurization (using 3 ADS valves) is begun within 6 minutes after rod injection (with Wdter level at normal operating level), the water level would not go below the top of active fuel:

'The procedure E.0.P.-1, hcwever, presently prescribes the use of this method of shutdown, unless all other methods are unavailable, and the Wdter level has fallen to the top of active fuel. Therefore, while the analysis presented was plant speci*ic, it could not be accepted as complete. The licensee was requested to include in the analysis a case using top of active fuel as the s N rting point. This will most probably result in partial core uncover; , and require an exemption request. The licensee was also asked to expand the analyses to

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include the use of LPCI and/or core spray. Pending the licensee's action this will be considered an unresolved item. (298/8615-01) Alternate Shutdown Procedures The alternate shutdown procedures have not been prepared yet, and will also be symptom oriented. They will, however, involve control room evacuation and the use of the new alternate shutdown panels, and therefore will be different from, and an addition to the present E.0.P.'s. These new procedures should be reviewed and walked down when available. The NRC review of the alternate shutdown procedures, hardware installations, and emergency lighting and communication equipment required by procedures will be considered an unresolved

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item. (298/8615-02) Repairs The licensee has proposed post-fire repairs for two systems, which he has designated as cold shutdown repairs. The first is a repair to cabling for the diesel fuel oil transfer pumps for fires in areas where the cables are located. The repair consists of replacing the damaged cable by re-routing a new cable on the outside of all areas affected. This repair would be required in 8-16 hours since the day tanks for the diesels each store enough fuel for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> at rated load. The licensee considers this a cold shutdown repair because he has the ability to reach cold shutdown in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. However, since the emergency procedures presently in use preclude shutdown in a manner which would reach cold shutdown in so short a time, this repair is considered a hot shutdown repair. The licensee was requested to review this repair and either change his proposal or ask for an exemption.

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The second is a repair to cabling for the battery chargers. Since

, the batteries are rated for only 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at full load, this is also considered a hot shutdown repair, and the licensee was requested to review this in the same manner as the diesel fuel pump cable repai '

Pending licensee action the above repairs will be considered an unresolved ite (298/8615-03)

7. Protection for Associated Circuits The Cooper Nuclear Station was inspected for compliance with the following associated circuit provisions of 10 CFR 50.48, Appendix R:

o Common Bus Concern o Spurious Signals Concern o Common Enclosure Concern

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l a. Common Bus Concern The common bus associated circuit concern is found in circuits, either non safety-related or safety-related, where there is a common power source with shutdown equipment and the power source is not electrically protected from the circuit of concer In order to inspect for this concern at Cooper Nuclear Station, the time-current curves developed during the licensee's bus coordination study were reviewed. The following randomly selected circuits were reviewed during the audit:

Licensee's Recommended Circuit Action to Achieve Coordination 125 V DC Panel DG2 None - Circuit Coordinated 125 V DC Panel AA2 None - Circuit Coordinated 250 V DC SWGR Bus 1B Replace 600 amp DB25 circuit breaker with 600 amp FRS fuse Service Water Pump SWP-1B None - Circuit Coordinated RHR Pump RHRP-10 None - Circuit Coordinated 4160 V Bus IF Change Breaker SS1F Phase Relay IAC (Relay) instantaneous setting from 105 to 120 Change Breaker 1FE IFC (Relay)

T.L. setting from 5 to 6 and

. IAC (Relay) instantaneous setting from 40 to 50 4160 V Bus 1G Change Breaker SSIG Phase Relay

. IAC 53 (Relay) Tap setting

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from 5 to 6 and instantaneous setting from 105 to 120 Change Breaker 1GE Phase Relay IFC (Relay) T.L. setting from 5 to 6, IAC (Relay) T.L. setting from 4 to 5 and IAC (Relay)

instantaneous setting from 40 to

480 V 1F Change MCC-LX Feeder 1000 amp DB-50 circuit breaker LSI (Amptector 1-A) Short Delay Time setting from .18 to .5

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Change MCC-K Feeder 1000 amp DB-50 circuit breaker LSI (Amptector 1-A) Short Time Delay setting from .18 to. 5 120 V Panel DP15-1A None - Circuit Coordinated The licensee's bus coordination study was comprehensive and recommended modifications (recommended modifications had not been accomplished at the time of the inspection) to achieve coordination for all circuits analyzed. The bus coordination study l recommendations are summarized below:

AC Bus recommended modifications / changes:

o Four fuses to be replaced with larger fuses, o Seven circuit breakers to be replaced with fuses, o Four DB-50 circuit breakers Amptector I-A trip settin'gs to'be changed, o Two DB-50 circuit breakers Amptector I-A tripping devices to be replaced, and o Six overcurrent relay settings to be change DC Bus recommended modifications / changes:

l o Twenty-two 08-25 circuit breakers to be replaced with fuses, o Four 08-50 circuit breakers to be replaced with fuses,

! o Two fuses to be replaced with higher amperage fuses, and l

l o One fuse to be replaced with a lower amperage fus Licensee action on-the above recommended modifications / changes is f

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pendin The licensee has a protective relay setting and testing program, Maintenance Procedure 7.3.1, which provides for relay testing at 1, 2,-3 and 5 year' intervals.

The licensee has a' procedure, Maintenance Procedure 7.3.2, for l

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setting'and testing circuit breakers; however, the procedure does not I

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specify the interval for performance.

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21 Spurious Signals The spurious signal concern is made up of two items:

o The false motor, control, and instrument readings such as occurred at the 1975 Brown's Ferry fire. These could be caused by fire initiated grounds, short or open circuits, o Spurious operation of safety-related or non safety-related components that would adversely affect shutdown capability (e.g., RHR/RCS isolation valves).

(1) Current Transformer Secondaries Licensee analysis for burned out current transformer secondaries including fires due to current transformer open circuits is incomplete. Licensee representatives stated that the current transformer secondary analysis will be completed during the l ongoing associated circuit analysi (2) High/ Low Pressure Interface Dcring the ongoing associated circuit analysis, 41 components have been identified as high/ low pressure interface component The high/ low pressure interface components and resolution status t are tabulated below:

High/ Low Pressure

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Interface Component Resolution Status

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HPCI-MOV-M014 Pending HPCI-M0V-M015 Pending HPCI-MOV-M016 Pending MS-A0V-738AV Pending MS-A0V-739AV Pending MS-A0V-A080A Pending MS-A0V-A080B Pending MS-A0V-A080C Pending

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MS-A0V-A0800 Pending l

MS-A0V-A086A Pending i

MS-A0V-A086B Pending l MS-A0V-A086C Pending MS-A0V-A086D Pending MS-MOV-M074 Pending l MS-M0V-M077 Pending l MS-MOV-M078 Pending l MS-MOV-M079 Pending MS-50V-SPV71A Pending MS-S0V-SPV71B Pending MS-50V-SPV71C Pending

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'MS-50V-SPV71D Pending

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',MS-50V-SPV71F Pending

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. 'MS-50V-SPV71G Pending

. . -MS-50V-SPV71H Pending

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MS-V-263X-45 Pending

'- RCIC-MOV-M0121 Pending

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RCIC-MOV-M015 Pending RCIC-MOV-M016 Pending RHR-A0V-PCV70A Pending RHR AV0-PCV708 Pending RHR-MOV-920MV Pending RHR-MOV-921MV Pending RHR-MOV-M017 Pending RHR-MOV-M018 Pending RHR-MOV-M025A Pending RHR-MOV-M025B Pending RHR-MOV-M0274A Pending RHR-MOV-M0274B Pending RWCU-MOV-M015 Pending RWCU-MOV-M018 Pending (3) Isolation of Fire Instigated Spurious Signals Licensee analysis for fire instigated spurious signals is incomplete. Licensee representatives stated that circuits that require isolation for fire instigated spurious signals have been identified and that resolution is pendin Proposed methods of resolution include:

o Prefire rackout of breakers o Rerouting of cables o Wrapping or boxing in cables o Installing fire breaks on cable trays o Installing isolation switches The following proposed alternate safe shutdown modifications j (DC86-21) were reviewed:

Component Proposed Isolation PT-83 HPCI-Pump Discharge Pressure Transfer switch with fuses

! Indication PT-89 HPCI-Turbine Steam Inlet Transfer switch with fuses Pressure l

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PT-100 HPCI-Pump Suction Pressure Transfer switch with fuses Indication SC-2792 HPCI-Turbine Speed Indication Transfer switch FT-82 HPCI Flow Control and Indication Transfer switch with fuses FT-1098 RHR Flow Indication B Loop Transfer switch with fuses LT-598 Reactor Vessel Level Transfer switch with fuses LT-918 Reactor Vessel Level Transfer switch with fuses LT-10 Suppression Vessel Level Transfer switch with fuses LT-681B Emerg. Cond. Storage Tank 1B Transfer switch with fuses Level Torus Temperature Indication Dedicated RTD Detectors Proposed isolation was satisfactory for the circuits reviewe c. Common Enclosure The common enclosure associated circuit concern is found when redundant circuits are routed together in a raceway or enclosure and they are not electrically protected, or fire can destroy both circuits due to inadequate fire protection mean Licensee representatives stated that:

o All circuits are electrically protected by breakers or fuses, o Cables for redundant safe shutdown divisions are not routed in common enclosur o Non-safety related cables routed in common enclosure with safety related cables are never routed between division Random cable selection in the field did not identify any cable routing in common enclosure that constituted a common enclosure Concer d. Multiple High Impedance Faults The multiple high impedance fault concern is found in the case where multiple high impedance faults exist as loads on a safe shutdown power supply and cause the loss of the safe shutdown supply prior to clearing the high impedance fault .

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Licensee analysis for multiple high impedance faults is incomplet Licensee representatives stated that the multiple high impedance faults analysis will be completed during the ongoing associated circuit analysis, Associated Circuit Analysis The licensee is conducting an associated circuit analysis. The analysis is being conducted thoroughly and has identified problems requiring resolution to achieve compliance with Appendix R requirement The completion of the associated circuit analysis, current transformer analysis, multiple high impedance faults analysis, high/ low pressure interface analysis, and outstanding modifications from the breaker-fuse relay coordination study is considered an unresolved ite (298/8615-04)

8. Fire Protection, Detection, Suppression The NRC inspector reviewed the exemptions to Appendix R that were granted by the NRC on September 21, 1983, in the following areas:

o Service Water Intake Structure o Cable Spreading Room o Cable Expansion Room o Reactor Building, Northeast Corner Room o Control Building Basement o Auxiliary Relay Room o Control Room o Fire Area Boundaries - Four Areas

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Reactor Building 932' Elevation (Critical Switchgear Rooms IF and 1G)

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Reactor Building 931' Elevation

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Reactor Building 903' Elevation (excluding northeast corner)

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Reactor Building 859' and 881' Elevations (quadrants and tours area)

The areas above were inspected to ensure that the level of fire protection, including detection and suppression, was adequate and as described in the exemptions. The level of protection was determined to be adequate with the following exceptions:

o The auxiliary relay room had only one smoke detector installed. This left one beam pocket without detection. An additional smoke detector was deemed necessar s

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o- The installation of a flame impingement shield beneath the Division 2 conduit bank was found to not extend far enough to protect all the conduits. The NRC inspector felt that some additional level.of protection under the conduit bank was neede Pending the licensee's action, the above two findings are considered'as one unresolved ite (298/8615-05) ,

~ Unresolved Item An unresolved item is a matter about which more 'information is required in order to detennine whether it is an acceptable item, a violation,' or a deviation. Five unresolved items are discussed in paragraphs 6.b, 6.c, 6.d, 7.e, and 8 of this repor . Exit Interview An exit interview was conducted on April 25, 1986, with those Nebraska Public Power District personnel denoted in paragraph 1 of this report. At this meeting, the scope of the inspection and the findings were

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sumarized.