ML20203F782
| 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: |
| References | |
| 50-298-86-15, TAC-61117, NUDOCS 8607310178 | |
| Download: ML20203F782 (25) | |
See also: IR 05000298/1986015
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APPENDIX B
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
NRC Inspection Report:
50-298/86-15
License:
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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'R.' P. Mul fikin,/Pr@ct Inspector, Project
Dste'
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Section B, Reactor Projects Branch
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K. E. Nurphy,/ Project #Tnspector, Project
Date '
Section B, Reactor Projects Branch
Participating
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in the
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inspection:
D..Notley, Office of Nuclear Reactor Regulation
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J. .Kudrick, Office of Nuclear Reactor Regulation
A. Coppola, Brook
en Nat onal Laboratory
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K. Parki so , Br
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tional Laboratory
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Approved:
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./ .Waudo , tiiief, Project Section A,
Date
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[Ractor
ojects Branch
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D. R. Hunter, Chief, Project Section B,
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Reactor Projects Branch
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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
compliance to the safe shutdown requirements of 10 CFR 50, Appendix R.
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
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criteria, paragraph 3.) Five unresolved items are identified in paragraphs 6.b,
6.c, 6.d
7.e, and 8.
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DETAILS
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Persons Contacted
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- 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
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- V. L. Wolstenholm, Quality Assurance Manager, CNS
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- J. V. Sayer, Technical Staff Manager
- H. T. Hitch, Acting Administrative Services Manager
- C. R. Goings, Regulatory Compliance Specialist
- W. Crawford, Maintenance Supervisor
M. Ward, Shift Supervisor
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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
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- 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
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A. Morisi, Electrical Engineer, EPH
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- Denotes those attending the exit interview cor. ducted on April 25, 1986.
The NRC inspectors also interviewed other CNS personnel during the
inspection.
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2.
List of Documents Reviewed
a.
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
DC Switchgear Rooms
05/23/84
MDC No. 84-5, Appendix R - Fire Protection for the Cable
06/07/84
Spreading Room
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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
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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
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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
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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
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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
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 CO ) - Annunciator 3
2
CNS Procedure 2.3.2.40, Revision 10, Fire Protection -
10/30/85
CNS Procedure 2.3.2.40, Revision 5, Fire Protection -
03/27/86
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)
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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
b.
Drawings
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Title
Date
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Flow Diagram - Residual Heat Removal System No. 2040,
02/11/74
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Revision 13
Flow Diagram - Reactor Core Isolation Coolant System
04/15/74
No. 2043, Revision 14
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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,
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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
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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. & Ind.
03/06/86
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 O
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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 standards.
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 sources.
The NRC inspector conducted a walkdown of the fire suppression water
system and verified that it was operable as required by technical
specifications.
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 zone.
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 satisfactory.
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
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fix this door was initiated.
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 doors.
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
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commitments under the Appendix R exemption requests.
The NRC inspector
informed the licensee that the applicable technical specifications,
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paragraph 3.19.8 does not recognize any other condition and only a
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continuous fire watch satisfies the technical specification LCO.
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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 "A.
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 violation.
(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-essential.
This is a 3-hour rated fire door for DC switchgear room 18.
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
frequencies.
Fire brigade training and drill records were reviewed as well as selected
personnel records.
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 requirements.
4.
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
-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 areas.
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 alignment.
Although Appendix R does not specifically state this as a requirement it
is recommended that the licensee consider incorporating this item into a
procedure.
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 test.
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 discharge.
5.
Post-Fire Safe ~ Shutdown Capability
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a.
Systems Required for Safe Shutdown
The systems required for safe shutdown as essential are listed in the
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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 zone.
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The.following paragraphs list the systems used for safe shutdown by
fire areas or zones according to the current analysis (dated
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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 power.
This is true for a fire in any fire zone or
area.
If neither action occurs, scram can be accomplished
outside of the control room by opening the RPS MG set output
breakers.
(2) Reactor Coolant Inventory Control and Decay Heat Removal
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For most fire areas or zones, (which do not require alternate
shutdown), the reactor core isolation cooling (RCIC) system will
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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
be maintained throughout hot shutdown.
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For five fire areas, there is no high pressure system available
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depressurization system (ADS) and low pressure injection will be
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used.
For fire area RB-A, the low pressure coolant injection
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'(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 pool.
High pressure coolant injection (HPCI) is not used for coolant
inventory control except for those areas requiring alternate
shutdown.
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 panel.
(4) Support System and Equipment
For the safe shutdown systems described above, the following
, support systems are required and will be available:
o
Diesel Generator - at least one of two for RHR and SW pump
power.
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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 exchanger.
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 power.
o
D.C. batteries and battery charger - one train.
(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 fires.
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 River.
b.
Area Compliance with Appendix R,Section III.G.2
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 completed.
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:
Equipment
Division
Type Cable
Cable Number
Service Water Pump
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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
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 noncompliance.
Fire Area RB-J, SWGR Room IF, and Fire Area RB-K, SWGR Room 1G,
contain cabling for the following redundant safe shutdown components:
Equipment
Division
Type Cable
Cable Number
RHR Fan Coil Unit
I & II
Control
MK 119
RHR Fan Coil Unit
I & II
Power
MS 101
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 requirements.
The associated circuit analysis data documents the status of
cables MS 101 and MK 119. Resolution is pending.
Fire Area CB-A, RHR Service Water Booster Pump and Service Air
Compressor Areas, contains cabling for the following redundant safe
shutdown components:
Equipment
Division
Type Cable
Cable Number
Service Water Pump
I
Power
H401
SW-P-A
__
_
, _
___ _ __
.
-
,
.
- _ - .
. -- -
.
~.
._.
. - .. -
-
,
,
.
,
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
SW-P-D.
,
Diesel Generator'
I
Power
DG1A, DG1B
DG1
Diesel Generator
II
Power
DG37A, DG37B
~
!
The loss of cables H401, H521, H411 and H531 will cause a loss of
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.
'
Fire Area CB-8, Battery Room IB, contains cabling for the following
!
p
redundant safe shutdown components:
!
Equipment
Division
Type Cable
Cable Number
Service Water Pump
I
Power
H401
'
i
SW-P-A
i
Service Water Pump
II
Power
H521
i
SW-P-B
'
' Service Water Pump
I
Power
H411
SW-P-C
!
Service Water Pump
II
Power
H531
SW-P-D
~.
.'
7
.
_
Diesel. Generator
I
Power
DG1A, DG1B
'
f
DG1
7
(DieselGenerator
II
Power
DG37A, DG37B
,DG2,
"
"
-
,
e
,
~
.; Fire Area CB-B was found to be in compliance with the separation
.
requirements of Section III.G.2.
"
-
- '
~
. .
i
l
-_---,...-..-.___.m..
_ _ _ _ _ , _ . _ _ _ . _ . . _ . _ _ _ , _ _ _ _ _ .
. . _ ,
. , _ . _ , , , . _ .
_ _ _ , . . - , . .
.
.
.
16
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:
o
Control Room (Fire Area CB-D)
-
CableSpreadingRoom(FireAreaCB-D]
o
o
Cable Expansion Room (Fire Area CB-D,
o
Aux. Relay Room (Fire Area CB-D)
Computer Room (Fire Area CB-D)-
o
Reactor Bldg. 903' Elev. , N.E. corner (Fire Area RB-FN)*
o
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 8.
(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 5.a.
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 valves.
(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).
.
. - .
-
_ _ _
. . - _ _
-
.
-
__
.
.
17
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 ladder.
The panels will be equipped with isolation switches for all of
the equipment controlled to ensure independence from the control
room.
6.
Procedure
a.
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 5.4.1
General Fire Procedure
o 5.4.2.1
Battery Room Fire
o 5.8
Emergency Operating Procedures (E.0.P.)
o
E.0.P.-C
Operator Precautions
o
E.0.P.-1
Reactor Pressure Control
b.
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
-
.
.
18
include the use of LPCI and/or core spray.
Pending the licensee's
action this will be considered an unresolved item.
(298/8615-01)
c.
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
item.
(298/8615-02)
'
d.
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.
'
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 repair.
Pending licensee action the above repairs will be considered an
'
unresolved item.
(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
1
-
,
is
t
-
}
.
.
19
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 concern.
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
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
.
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
50
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
_ .
. . .
.
20
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,
Four DB-50 circuit breakers Amptector I-A trip settin'gs to'be
o
changed,
o
Two DB-50 circuit breakers Amptector I-A tripping devices to be
replaced, and
o
Six overcurrent relay settings to be changed.
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 fuse.
f
Licensee action on-the above recommended modifications / changes is
pending.
'
- 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
7
l
'
setting'and testing circuit breakers; however, the procedure does not
I
specify the interval for performance.
'
t
..
s
y
p m i-th-
--
g
1---e.s
>&p---
-y
9
.y->.ygd+r
-g
yym-p-gw..g
p.3-.mipe9-,-
one y p - g
ig-yy.,
y-gm.e99--m.gs a t
gey-een
%.e P=w tr ap
a.r---y-
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+fy__
-
.
.
21
b.
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 analysis.
(2) High/ Low Pressure Interface
Dcring the ongoing associated circuit analysis, 41 components
have been identified as high/ low pressure interface components.
The high/ low pressure interface components and resolution status
are tabulated below:
t
High/ Low Pressure
Interface Component
Resolution Status
-
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
MS-A0V-A0800
Pending
,
l
MS-A0V-A086A
Pending
MS-A0V-A086B
Pending
i
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
.
-
- . _ _ _
-
.
--
. - - - -
--
.-
.
-.
.- - ___ _ __
.
'
.-
P
.
22
,
.
'MS-50V-SPV71D
Pending
--
MS-50V-SPV71E
Pending
',MS-50V-SPV71F
Pending
.
,
"
Pending
'MS-50V-SPV71G
.
'
. .
.
-MS-50V-SPV71H
Pending
MS-V-263X-45
Pending
RCIC-MOV-M0121
Pending
'-
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 pending.
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
l
t
. .
-,
.
.-
a
23
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 reviewed.
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 means.
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 enclosure.
o
Non-safety related cables routed in common enclosure with safety
related cables are never routed between divisions.
Random cable selection in the field did not identify any cable
routing in common enclosure that constituted a common enclosure
Concern.
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 faults.
.
.
24
Licensee analysis for multiple high impedance faults is incomplete.
Licensee representatives stated that the multiple high impedance
faults analysis will be completed during the ongoing associated
circuit analysis,
e.
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
requirements.
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 item.
(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
-
Reactor Building 932' Elevation (Critical Switchgear Rooms IF
and 1G)
-
Reactor Building 931' Elevation
-
Reactor Building 903' Elevation (excluding northeast corner)
-
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 necessary.
s
.:
25
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 needed.
Pending the licensee's action, the above two findings are considered'as
one unresolved item.
(298/8615-05)
,
9.
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 report.
10.
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
sumarized.
,