ML20044G343
| ML20044G343 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 05/12/1993 |
| From: | Constable G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20044G342 | List: |
| References | |
| 50-458-93-09, 50-458-93-9, NUDOCS 9306020373 | |
| Download: ML20044G343 (39) | |
See also: IR 05000458/1993009
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APPENDIX
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U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
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Inspection Report:
50-458/93-09
Operating License: NPF-47
Licensee: Gulf States Utilities
P.O. Box 220
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St. Francisville, Louisiana 70775
Facility Name: River Bend Station (RBS)
Inspection At: RBS, St. Francisville, Louisiana
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Inspection Conducted: March 29 to April 22, 1993
Team Leader (Acting): Amarjit Singh, Reactor Inspector, Plant Support Section
Division of Reactor Safety
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Team Members: Howard F. Bundy, Reactor Inspector, Plant Support Section
Division of Reactor Safety
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Michael E. Murphy, Reactor Inspector, Plant Support Section
Division of Reactor Safety
K. Sullivan, Electrical Systems Specialist, Brookhaven
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National Laboratory, Consultant
A. Fresco, Hechanical Systems Specialist, Brookhaven
National Laboratory, Consultant
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Approved:
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't;~Li CbHstable, Chief, Plant Support Sect!on
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Division of Reactor Safety
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9306020373 930524
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ADDCK 05000458
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TABLE OF CONTENTS
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EXECUTIVE SUMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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DETAILS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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l INTRODUCTION
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2 FIRE PROTECTION OF SAFE SHUTDOWN CAPABILITY (64100)
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2.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . .
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2.2 Systems Required for Safe Shutdown and Shutdown Methodology
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2.3 Associated Circuits
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2.4 Conclusions
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3 ALTERNATE SHUTDOWN CAPABILITY (64100)
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3.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . .
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3.2 Procedures . . . . . . . . . . . . . . . . . . . . . . . . . .
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3.3 Operator Training
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3.4 Inservice Testing of Remote Shutdown Capability
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3.5 Conclusions
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4 FIRE PROTECTION / PREVENTION PROGRAM (64704) . . . . . . . . . . . . . .
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4.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . .
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4.2 Program Review and Implementation
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4.3 Surveillances
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4.4 P l a n t W a l kd own . . . . . . . . . . . . . . . . . . . . . . . .
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4.5 Fire Brigade Training / Drills . . . . . . . . . . . . . . . . .
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4.6 Fire Protection Quality Assurance
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4.7 Conclusions
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5 FOLLOWUP ON CORRECTIVE ACTIONS FOR VIOLATIONS (92702)
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6 FOLLOWUP (92701) . . . . . . . . . . . . . . . . . . . . . . . . . . .
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6.1
(Closed) Unresolved Item 456/9204-01: River Bend Station
Specific Thermo-Lag Issues
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6.2 (Closed) Unresolved Item 458/9204-02: Appendix R Issues . . .
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6.3
(Closed) Inspection Followup Item 458/8937-02:
Motor-Operated Valves Not Deenergized - Conflicts with Fire
Hazards Analysis
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7 ONSITE REVIEW 0F LICENSEE EVENT REPORTS (92700)
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7.1
(Closed) Licensee Event Report 458/88-009: Unsealed Fire
Barrier Penetrations
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7.2 (Closed) Licensee Event Report 458/89-005: Fire Seal
Penetration IC2W19 Inadequate Due to Poor
Application / Inspection Technique
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7.3 (Closed) Licensee Event Report 458/90-017: Inadequate Fire
Barrier in Shake Space
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7.4 (Closed) Licensee Event Report 458/91-005: Design
Deficiencies in Fire Doors
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7.5 (0 pen) Licensee Event Report 458/92-003: Deviations From
Approved Designs in Structural Steel Fireproofing . . . . . .
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7.6 (Closed)
Licensee Event Report 458/91-008:
Fire Hazards
Analysis Deficiencies Including Lack of Fire Wrap / Inadequate
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7.7 (Open) Licensee Event Report 458/89-010: Missing or
Inadequate Penetration Seals Per Technical Specification 3.7.7.a . . . . . . . . . . . . . . . . . . . .
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7.8 (Closed) Licensee Event Report 458/92-007: Vulnerability to
Hot Shorts Discovered as a Result of Information Notice 92-18
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7.9 Conclusions
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ATTACHMENT 1 - Persons Contacted and Exit Meeting
ATTACHMENT 2 - Inspection Findings Index
ATTACHMENT 3 - Documents Reviewed
ATTACHMENT 4 - Table 1, Coordination of Electrical Protective Devices
Table 2, Circuit Breaker and Relay Test Procedures Reviewed
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EXECUTIVE SUMMARY
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In 1988, Gulf States Utilities (GSU) began a study to compare the River Bend
Fire Hazards Analysis (FHA) to plant procedures and valve lineup requirements.
The licensee discovered that 19 motor-operated valves required to be
deenergized by the FHA during operations, were actually energized because
procedural controls had not been established to fulfill FHA requirements. The
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NRC reviewed this issue during inspections conducted in October 1989 and
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January 1990, and subsequently issued a Notice Of Violation (EA 90-039) in
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April 1990. GSU stated in their response to the violation that a contributing
cause of the failure to implement fire protection requirements was a lack of
knowledge of fire protection issues within the-GSU engineering organization
and a lack of organizational maturity at the time of transition of
responsibility from the architect / engineer to GSU in 1985.
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Part of the extensive corrective actions included retaining a contractor to
independently review the FHA and its implementation at River Bend Station.
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This review was completed in January 1991 and identified 106 discrepancies
that required further review. The NRC conducted a followup inspection in
January 1992 to assess the adequacy and timeliness of corrective actions and
to assess a number of emerging Thermo-Lag issues. At that time, the
evaluation of the issues identified by the contractor was not complete, and as
a result, the FHA update had not been completed. The inspection findings were
identified as unresolved items and the licensee was requested to respond to
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the issues raised and to provide a schedule for completion of planned actions.
The inspection report noted that, "The many examples of fire protection
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weaknesses and inadequacies documented in this report demonstrate an apparent
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lack of management attention to the fire protection program at River Bend
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Station."
During the period March 29 through April 2,1993, the NRC conducted a planned
announced team inspection of the licensee's corrective actions- to the
previously identified fire protection findings. The inspection included a
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broad evaluation of the approveu River Ben 6 Station Fire Protection Program
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including the fire protection features and procedures used to achieve a post-
fire safe shutdown. Generic Thermo-Lag issues were not reviewed during this
inspection.
The significant findings identified by the team are summarized below.
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Strenaths:
The team concluded that the licensee's shutdown methodology properly
identified the components, instrumentation, and systems necessary to
achieve and maintain safe shut 3own conditions from either within or
outside the control room coincident with a loss of normal ac power
(Section 2.2.2).
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The coordination (selective tripping) of a sample of power supplies was
found to be acceptable when evaluated against Appendix R criteria. The
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scope of this inspection was limited to assessing the adequacy of
selective tripping between protective devices of a required power supply
in the event of fire initiated faults (Section 2.3.1.1).
River Bend Station had developed and implemented procedures for testing
and maintenance of circuit breakers and relays associated with power
supplies that are required to achieve post-fire safe shutdown. This
element of the circuit breaker and relay testing program at River Bend
Station was found to be acceptable (Section 2.3.1.2).
River Bend Station had developed compensatory procedures which provided
operator guidance in the event a power supply was lost due to the
occurrence of fire induced high-impedance faults on unprotected cabling.
These procedures provided sufficient guidance to permit operators to
identify affected power supplies and take corrective actions (i.e., non-
essential load shedding) necessary to restore operability
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(Section 2.3.1.3).
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The licensees method of control for valves identified as having
potential to comp 'ise high/ low pressure interface boundaries was found
to be acceptable (? r* Men 2.3.2.3).
Appropriate procedt a. cratrols were in place to reduce fire hazards and
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implement the required arveillance tests of fira systems and equipment
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(Section 4.2).
Fire brigade training and qualifications or personnel were considered
strengths, ara actual performance during a drill was good (Section 4.5).
Weaknesses:
In GSUs response to the Notice of Violation identified in NRC Inspection
Report 50-458/90-02 (Ref.: Letter dated May 7, 1990, From:
J. C. Deddens, GSU, To: U.S. NRC), the licensee stated:
"The
independent contractor is to provide detailed documentation of the
design basis and assumptions of the FHA." During this inspection, the
licensee could not provide the team the engineering design basis and
analysis methodology necessary to verify the adequacy of the results
presented in the FHA.
For example, while the FHA was found to identify
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passive valves and dampers, which could spuriously change position as a
result of fire damage to connected cabling, the FRA did not identify the
specific cables evaluated and type of faults considered. Additionally,
the FHA did not specifically discuss the potential effect of fire
induced faults on cables or equipment which could initiate false signals
or spurious operation of equipment other than passive valves and
d=~aars, such as, pumps, motors, and motor cea+rol centers, or the
potential effect of fire initiated spurious operation of equipment
associated with non-credited safe shutdown methods in a given fire area.
Documents necessary to verify the adequacy of the assumptions and
methodology that form the basis of the FHA results were not available
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for review. The licensee could not provide analysis to support
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conclusions in the FHA that required systems and equipment would be
available to safely shutdown the plant in the event of a fire. This
failure to take adequate corrective action is an apparent violation of
Sections III.G and III.L oi appendix R and Criterion XVI of 10 CFR Part 50, Appendix B (Escalated Enforcement Item 458/9309-01)
(Section 2.3.2).
The team concluded that the licensee had not performed an analysis of
the common enclosure associated circuits. This analysis is necessary to
identify required actions needed to safely shutdown the plant in the
event of a fire. This is an additional example of the apparent
violation noted in the above paragraph. (Escalated Enforcement
Item 458/9309-01) (Section 2.3.3).
The team's assessment of Procedure A0P-0031, " Shutdown From Outside Main
Control Room," identified potential deficiencies with the procedure,
primarily related to insufficient information available to support the
actions taken by the operators. The licensee indicated that these items
would be entered into their internal tracking system for further study.
These concerns will be followed up in a future inspection (Inspection
Followup Item 458/9309-02). The issues include:
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Insufficient procedural guidance to mitigate the possible
consequences of spurious initiation of reactor pressure vessel
makeup injection systems in the event of a control room fire
(Section 3.2.1).
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In the event of a fire in the control room, as an immediate
action, the operators would initiate high pressure core spray.
The procedure did not provide guidance on terminating high
pressure core spray. To prevent reactor overfill, they would rely
on the automatic shut-off high pressure core spray logic, but the
automatic logic circuitry was not evaluated for potential spurious
signals to determine if it would be available (Sections 2.3.2
and 3.2.1).
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The procedure states that there was 15 minutes available to verify
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diesel generator ventilation fans were running, but a calculation
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intended to justify the 15 minute time period indicated that from
an initial temperature of 100'F, it took only 29 seconds to reach
120*F and concluded that, " Based on the above, operator
verification should be immediate, within 10 minutes." The
licensee had not provided justification for the 15 minutes
specified as allowable to verify that the diesel generator
ventilation fans were running and apparently had not acted upon
the results of, and recommendations &wn f-
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calculation (Section 3.2.1).
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Inadequatt justification to demonstrate that repairs to the
automatic depressurization system air supply are not required to
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achieve and maintain hot shutdown (Section 3.2.1).
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The team identified surveillance test results that were closed as
" acceptable with comments" when problems were identified during the
conduct of a surveillance. This was considered a weakness because it
was not evident that appropriate followup action was taken or that
appropriate work control procedures were followed (Section 4.3).
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The team concluded that the licensee's corrective actions for fire
protection licensee event reports had not been timely. Most of the
major fire protection issues were identified 3 to 5 years ago and the
issues involved matters that have existed since the plant was licensed
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in 1985 (Section 7).
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Conclusions:
The licensee's overall effort in the area of fire protection provides defense
in depth to assure the health and safety of the public. However, significant
weaknesses were identified in the areas of engineering analysis of fire safety
hazards and timeliness of corrective actions when deficient fire protection
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issues were identified.
Based on the results of this inspection, the NRC staff identified an apparent
violation related to the failure to take adequate corrective actions in
response to a violation issued in 1990. As noted above, GSU has not completed
all engineering analysis necessary to provide the design basis and assumptions
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to support conclusions in the FRA. The team concluded that the GSU
engineering organization still lacks the organizational depth to effectively
resolve complex fire protection issu,es.
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DETAILS
1 IhTRODUCTION
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In 1988, Gulf States Utilities (GSU) began a study to compare the River Bend
Fire Hazards Analysis (FHA) to plant procedures and valve lineup requirements
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as a part of corrective actions following discovery of improperly installed
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fire walls at the remote shutdown panel. The licensee discovered that 19
motor-operate-d valves required to be deenergized by the FHA during operations
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were actually energized because procedural controls had not been established
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to fulfill FHA requirements (Condition Report 89-1117). The NRC initially
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reviewed this issue during an inspection conducted in October 1989 (NRC
Inspection Report 50-458/89-37) and identified the issue as an unresolved
item.
Followup inspection was conducted by fire protection specialist during
January 1990 (NRC Inspection Report 50-458/90-02) and a Notice Of Violation
(EA 90-039, Severity Level 3, no Civil Penalty) was issued in April 1990.
GSU
stated in their response to the violation that a contributing cause of the
failure to implement fire protection requirements was a lack of knowledge of
fire protection issues within the GSU engineering organization and a lack or
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organizational maturity at the time of transition of responsibility from the
architect / engineer to GSU in 1985. This resulted in a failure to translate
fire protection program requirements into procedural requirements in the
pl ant.
Part of the extensive corrective actions planned included retaining a
contractor to independently review the FHA and its implementation at River
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Bend Station. This review was completed in January 1991 and it identified
106 discrepancies that required further review. The NRC conducted a followup
inspection in January 1992 (NRC Inspection Report 50-458/92-04) to assess the
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adequacy and timeliness of corrective actions and to assess a number of
Thermo-Lag and Appendix R issues that had been identified during a site visit
by NRR staff members. At that time, the evaluation of the issues identified
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by the contractor was not complete and as a result the FHA update had not been
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completed. The inspection findings were identified as unresolved items
(Unresolved Item 458/9204-01 and -02) and the licensee was requested to
respond to the issues raised and to provide a schedule for completion of
planned actions. The inspection report noted that, "The many examples of fire
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protection weaknesses and inadequacies documented in this report demonstrate
an apparent lack of management attention to the fire protection program at
River Bend Station."
During the period March 29 through April 2, 1993, the NRC staff and two
contract specialists from Brookhaven National Laboratory performed a planned
announced team inspection of the licensee's corrective actions to the
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previously identified fire protection findings. The inspection included a
broad evaluation of the approved River Bend Station Fire Protection Program
including the fire protection features and procedures used to achieve a post-
fire safe shutdown.
Generic Thermo-Lag issues were not reviewed during this
inspection.
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The inspection effort included the selective evaluation and assessment of the
commitments and exceptions taken to 10 CFR Part 50, Appendix R, Secticn III G,
" Fire Protection of Safe Shutdown Capability";Section III L, " Alternative and
Dedicated Shutdown Capability"; Appendix A to Branch Technical Position
APCSB 9.5-1, " Guidelines for Fire Protection for Nuclear Power Plants Docketed
Prior to July 1,1976"; and, NRC Generic Letters 81-12, 83-33, 86-10, and
88-12. These commitments and exceptions are documented in the Updated Safety
Analysis Report (USAR) and have been reviewed and approved in NUREG-0989,-
" Safety Evaluation Report related to River Bend Station," through
Supplement 4.
The inspection team used as guidance an approved inspection
plan which closely paralleled the NRC Inspection Manual Inspection
Procedures 64704, " Fire Protection / Prevention Program," and 64100, "Postfire
Safe Shutdown, Emergency Lighting and Oil Collection Capability at Operating
and Near Term Operating Reactor Facilities." The inspection was principally
performance based in that substantial field evaluations of as-built
configurations were performed as well as evaluations of procedures and -
hardware. Significant effort was also expended reexamining the licensee's
methodologies, engineering analysis, and assumptions supporting its fire safe
shutdown analysis.
2 FIRE PROTECTION OF SAFE SHUTDOWN CAPABILITY (64100)
2.1 Overview
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The purpose of this portion of the inspection was to review the protection of
electrical circuits associated with systems and equipment that would be needed -
to achieve safe shutdown in the. event of a design basis fire. The regulations
require that a licensee have the capability to shutdown and cooldown the plant
if a destructive fire occurs in the control room in conjunction with a loss of
all offsite electrical power.
The team selectively assessed the licensee's commitments to 10 CFR Part 50,
Appendix R,Section III.G., " Fire Protection of Safe Shutdown Capability,"
Section III.L., " Alternative and Dedicated Shutdown Capability," and
Appendix A to Branch Technical Position APCSB 9.5-1, " Guidelines for Fire
Protection for Nuclear Power Plants Docketed Pri to July 1,1976," oy:
Performing a review of the systems required to maintain safe shutdown
and achieve cold shutdown as described in Sections 7.4 and 9.5 of the
USAR and evaluated in Section 9.5 of the Safety Evaluation Report.
Reviewing the licensee's methodology for protecting electrical circuitry
associated with systems and equipment relied upon to achieve safe
shutdown. The licensee's associated circuit methodology was reviewed
for common power supply, spurious operation, and common enclosure
vulnerabilities as defined in NRC Generic Letter 86-10, " Implementation
of Fire Protection Requirements."
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2.2 Systems Reouired for Safe Shutdown and Shutdown Methodoloav
2.2.1
Objectives
The team found River Bend Station to have the following fire safe shutdown
performance goals:
Reactivity control that is capable of achieving and maintaining shutdown
conditions.
Reactor coolant system pressure control to support coolant inventory
control and decay heat removal.
Reactor coolant system level control capable of maintaining level above
the top of the core.
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Reactor decay heat removal capable of achieving and maintaining reactor
coolant at shutdown temperatures.
Appendix R,Section III.L.1, states that during the post-fire shutdown, the
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reactor coolant system process variables shall be maintained within those
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predicted for a loss of normal ac power, and the fission product boundary
integrity shall not be affected, i.e., there shall be no fuel clad damage,
rupture of any primary coolant boundary, or rupture of the containment
boundary.
To achieve these goals, the licensee had identified the systems required for
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post-fire safe shutdown, methods used and the fire areas for which they were
applicable. These were described in " Fire Analysss Criteria and Evaluation
Method Including Results and Conclusions for 10 CFR 50 Appendix R Fire Hazards
Analysis," Revision 7, dated April 9,1987, which was prepared for the
licensee by the Stone & Webster Engineering Corporation. This document will
be referred to as the FHA in this report.
2.2.2
Findings
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The team reviewed the licensee's designs for fire safe shutdown reactivity
control, reactor coolant system pressure and level control, reactor decay heat
removal, process monitoring and support systems as described in Section 7.4 of
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the USAR and determined the design could support the stated goals and
objectives. The team ascertained that the licensee's shutdown methodology
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properly identified the components, instrumentation, and systems necessary to
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achieve and maintain safe shutdown conditions from either within or outside
the control room coincident with a loss of normal ac power.
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2.3
Associated Circuits
In order to adequately demonstrate the intent of Section III.G of Appendix R,
the licensee's analysis must consider the potential effect of fire on all
cables and circuits necessary to assure the successful achievement of safe
shutdown performance goals (e.g., reactor coolant makeup and decay heat
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removal). Additionally,Section III.G requires that this analysis include an
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evaluation of the potential effect of fire initiated cable faults (hot shorts,~
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open circuits, and shorts to ground) on non-essential associated circuits. As-
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defined by NRC Generic Letter 81-12, such associated circuits of concern may
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be categorized.into one of three distinct types:-
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Circuits associated by common power supply (i.e.,- non-essential circuits
which share a common switchgear, motor control center, or distribution
panel with circuits of equipment relied on to achieve post-fire safe =
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shutdown)
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Circuits associated by common enclosure (i.e. non-essential circuits
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which share a common cable tray, conduit, junction box, etc., with
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required circuits).
Circuits whose spurious. operation may adversely impact the achievement
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of a safe shutdown performance goal.
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Acceptable protection alternatives for each type of associated circuit
described above have been principally defined by Generic Letters 81-12 with
additional clarification provided by Generic letter 86-10..
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During this inspection, the potential effect of fire on each of the associated ~
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circuit configurations described above was evaluated on a sample basis. -'The
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overall approach of this assessment was based on a review of the licensee's
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analysis as documented in its FHA (Criterion 12210-240.201) and a selected
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sample of power, control, and instrument circuits which were then reviewed for
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potential fire initiated problems.
2.3.1
Review of Circuits Associated by Common Power Supply
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A common power supply associated circuit concern is found when unprotected
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circuits are connected to a common power' supply'(switchgear, motor control-
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center, panel, etc.) with equipment required to achieve post-fire safe
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shutdown. . In the absence of adequate fire protective barriers or electrical-
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coordination (selective tripping) between electrical protective devices -(i.e.. .
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circuit breakers, relays, fuses), fire initiated faults on branch / load cables
of an affected supply may propagate to a trip.of the upstream feeder'
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protective device of the power supply prior to isolation by the branch / load
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protective device (s) located nearest the fault. Such a scenario would result
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in' a loss of the entire ' supply.
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2.3.1.1 Coordination of Electrical Protective Devices
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Unless protected against fire damage, selective tripping of electrical-
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protective devices is necessar" Nncure that fire in$tiated faults on
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affected cables will be rapidly isolated by the protective device located
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nearest the fault, prior to the fault current propagating to a trip of any
protective device located upstream of the affected power supply.
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Selected electrical protection provided for power supplies of equipment relied
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on to achieve safe shutdown in the event of fire were reviewed. The specific
sample of circuits selected for review and the corresponding results of this
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evaluation are provided in Attachment 4, Table 1.
The coordination (selective
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tripping) of the sample of power supplies selected for review was found to be
acceptable when evaluated against Appendix R criteria. The scope of this
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inspection was limited to assessing the adequacy of selective tripping between
protective devices of a required power supply in the event of fire initiated
faults only. Within the context of Appendix R, the licensee can, and did,
take full credit for such fault current limiting factors as the length of
cable between the affected supply and the point of a postulated fire induced
fault.
2.3.1.2
Circuit Breaker and Relay Testing and Maintenance
Circuit breakers and relays may have adjustable settings and trip points. The
specific values selected for the setting of these devices is largely based on
the results of calculations performed during the plants coordination study.
An established program consisting of surveillance testing and periodic
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maintenance is, therefore, necessary to provide assurance that the selected
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settings will not drift or vary considerably over the life of the plant.
River Bend Station had developed and implemented procedures for the test and
maintenance of circuit breakers and relays associated with power supplies
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required to achieve post-fire safe shutdown. The specific procedures selected
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for review by the inspection team are provided in Table 2 of Attachment 4.
Based on the team's review of these procedures and discussions Cth plant
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maintenance personnel, the circuit breaker and relay testing program at River
Bend Station was found to be acceptable.
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2.3.1.3
High Impedance Faults
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As stated in Section 5.3.8 of Generic Letter 86-10, the NRC staff has
determined that to meet the seoaration criteria of Section III.G, simultaneous
high impedance faults (fault currents of a value that is just below the trip
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point of the protective device on each individual circuit) for all associated
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circuits located in a given fire area, should be considered in the evaluation
of safe shutdown capability. The significance of this issue is that an
inadvertent trip of a power supply relied on to achieve post-fire safe
shutdown cuid result from fire induced circuit damage to unprotected cables
of the affected supply. River Bend Station had developed compensatory
procedures which provided operator guidance in the event a power supply was
lost due to the occurrence of fire induced high impedance faults on
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unprotected cabling. During the inspection, the licensee was found to have
incorporated an acceptable level of procedural guidance into
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Procedures A0P-0031, " Shutdown From Outside The Control Room," and A0P-0052,
" Fire Outside the Main Control Room (In Areas Containing Safety Related
Equipment)." A review of these procedures found them to provide sufficient
guidance to permit operators to identify affected power supplies and take
corrective actions (i.e., non-essential load shedding) necessary to restore
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operability.
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2.3.1.4
Findings
The licensee had developed and implemented procedures for the test and
maintenance of circuit breakers and relays associated with power supplies
required to achieve post-fire safe shutdown. The coordination / selective
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tripping capability of power supplies relied upon to achieve and maintain safe
shutdown was found to be acceptable. The licensee also had developed
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compensatory procedures which provide operator guidance in the event a power
supply is lost due to the occurrence of fire induced high impedance faults on
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unprotected cabling.
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2.3.2 Review of the Spurious Signals Associated Circuit Concern
Specific circuits of concern include those which have a physical separation
that is less than that required by Section III.G and have a connection to
equipment whose spurious operation or mal-operation could adversely affect the
shutdown capability. This concern is principally comprised of two items:
The mal-operation of required equipment due to fire induced damage to
associated cabling.
Examples include false motor, control, and
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instrument readings which may be initiated as a result of fire induced
grounds, shorts, or open circuits.
The spurious operation of safety-related or non-safety-related
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components that could prevent the accomplishment of a safe shutdown
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function.
At the time of the inspection, licenseI representatives stated that potential
spurious operation circuits were evaluated by an independent contractor
during the revalidation of FHA Criterion 240.201. Additionally, River Bend
Station representatives stated that these circuits were evaluated in the same
manner as required circuits and, therefore, were provided with the same level
of protection as that required for redundant trains of equipment in accordance
with Section III.G.2 of Appendix R.
During the inspection, however, the
licensee could not provide any detailed documentation to support either this
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statement or the conclusions presented in its FHA Criterion 240.201. While
the FHA was found to identify and describe this issue and provide the results
of an analysis, detailed, plant-specific supporting documentation and
calculations, necessary to verify the adequacy of the assumptions and
methodology that form the basis of the documented results, were not available
for review. Additionally, it should be noted that in its response to a Notice
of Violation identified in NRC Inspection Report 50-458/90-02 (Ref.:
Letter
dated May 7, 1990, From:
J. C. Deddens, GSU, To:
U.S. NRC), the licensee
stated:
"The independent contractor is to provide detailed documentation of
the design basis and assumptions of the FHA."
In response to the team's concern for the apparent lack of such detailed
documentation, licensee representatives stated that the supporting documents
for this analysis were still in the possession of its contractor and had not
yet been incorporated into the FRA. At the inspection team's request, the
licensee attempted to obtain this information from its contractor but was not
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able to obtain an acceptable response. Subsequent to this inspection, the
licensee forwarded a contractor draft report to the team. This report was
initially represented by the licensee as a description of the evaluation of
the safe shutdown methodology. Hv ever, the licensee had previously stated
that 'All associated circuits of concern had not been analyzed by (the
contractor) in this document." This comment referred to associated circuits
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in common enclosures (Refer to Section 2.3.3).
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In addition to the lack of supporting documentation and calculations, the
inspection team was also concerned with the apparently. limited scope of the
licensee's analysis of this concern. Specifically, its analysis
(Criterion 240.201) was found to only address the potential effect of spurious
operations of passive valves and dampers.
Such components are defined as
those valves and dampers that are in a fixed position during normal operation
and must not change position. The analysis did not, however, appear to
address the manner in which the spurious operation of components that may have
an indirect affect on the plants ability to achieve safe shutdown, such as
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automatic logic circuits and equipment associated with non-credited systems,
were evaluated and dispositioned.
Subsequent review of the licensee's alternative shutdown methodology
identified an example which, in the team's view, demonstrated a lack of
completeness with regard to the licensee's analysis of this concern.
Specifically, during the inspection team's review and walkthrough of Alternate
Shutdown (control room fire) Procedure A0P-0031, it was determined that in the
event of a fire in the control room, as an immediate action (A0P-31,
Step 4.4), the operators would initiate high pressure core spray, a non-
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credited system, in addition to the reactor core halation cooling system, the
credited method of providing reactor coolant makeup. However, the procedure
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did not provide any guidance on terminating high pressure core spray. When
questioned about how high pressure core spray would be terminated to prevent
reactor overfill, licensed operators stated that they would rely on the
automatic shut-off high pressure core spray logic. . But high pressure core
spray was not a credited shutdown system for a control room fire, and its
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automatic ,sjic circuitry apparently was not evaluated for potential spurious
signals. Nevertheless, it was assumed to be available both in the procedure
and by the operators. The potential for spurious high pressure core spray
equipment actuation or mal-operation of its logic circuitry was not found to
be fully evaluated in the River Bend Station FHA or the Stone & Webster Safe
Shutdown Calculation, Criterion 240.201, and is an example of an apparent
weakness in the licensee's analysis of this concern. This issue is discussed
further in Section 3.2.1.
Based on the above, the inspection team was concerned that the licensee's lack
of a complete and well documented analysis of the associated circuits concern
may have resulted in ineffective procedural guidanca whi ^ e uld cause
operators to rely on equipment and' systems which may not be available or which
may spuriously operate due to fire damage. This lack of analysis is
considered to be an extmple of a failure to take adequate corrective actions
and appears to be an apparent violation of Sections III.G and III.L of
Appendix R and Criterion XVI of 10 CFR Part 50, Appendix B (Escalated
Enforcement Item 458/9309-01).
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Isolation of Fire Initiated Spurious Signals
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River Bend Station had developed various methods to prevent and isolate
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spurious equipment operations that may occur as a result of fire. Specific
examples noted during the inspection included:
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Administrative controls,
Isolation / transfer switches which incorporate redundant fusing schemes,
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Fire wrap, and
Manual operator actions governed by written procedures.
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2.3.2.2
Potential for Spurious Motor-Operated Valve Operations
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Section III.G of Appendix R requires that protection be provided for fire
initiated faults on circuits that could adversely impact the achievement and
maintenance of stable safe shutdown conditions. Adattionally, Appendix R safe
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shutdown criteria require a designated set of safe shutdown equipment to
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remain operable from the remote shutdown panel after a control room fire.
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As described in Information Notice 92-18, " Potential for Loss of Remote
Shutdown Capability During a Control Room Fire," dated February 28, 1992, a
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postulated fire in the control room or cable spreading room could create a
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single hot short in the control circuitry of various motor-operated valves
resulting in their spurious operation. Since the postulated fault would cause
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the valve position limit and torque switches to be bypassed, mechanical damage
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of the valve due to overtorque may occur, which could render the motor-
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operated valve inoperable.
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The licensee issued Licensee Event Report 92-07 as a result of findings
identified during their review of Information Notice 92-18.
Ir.spection of
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that area is documented in Section 7.8 of this report.
2.3.2.3 High/ Low Pressure Interfaces
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High/ low pressure interfaces were examined to determine if the licensee had
provided sufficient protection to prevent fire induced spurious signals from
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initiating an uncontrolled loss of reactor coolant.
The River Bend Station USAR was found to identify the following valves as
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high/ low pressure interfaces of concern:
Series residual heat removal /recirc system interface and containment
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isolation valves 1E12*MOV-F008 and IE12*MOV-F009; and
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Series main steam / reactor building valves IMSS*MOVF001 and IMSS*MOVF002.
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To prevent the fire initiated spurious opening of the high/ low pressure
interface boundaries, which these two sets of series valves comprise, the
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licensee had implemented administrative controls and operating procedures to
maintain one valve closed with its power supply isolated during plant
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operations. Additionally, Residual Heat Removal Valve M0VF008 was provided
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.with a keylocked switch on the remote shutdown panel which blocks the control
of the valve from both the main control room and the remote shutdown panel.
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The licensee's method of control for valves identified as comprising high/ low
pressure interface boundaries was found to be acceptable.
2.3.2.4
Findings
River Bend Station had taken acceptable corrective actions with regard to
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spurious motor-operated valve operability concerns described in information
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notice 92-18. The licensee also had implemented acceptable administrative
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controls for valves identified as comprising high/ low pressure interface.
The licensee's questionable procedural guidance could cause operators to rely
on equipment and systems which may not be available or which may spuriously
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operate due to fire damage as a result of an apparent failure to take complete
and adequate corrective actions in response to Violation 458/9002-02
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(EA 90-039). The team concluded that the licensee has not met Sections III G
and III L of 10 CFR Part 50, Appendix R and Criterion XVI of 10 CFR Part 50,
Appendix B.
This-is an apparent violation (Escalated Enforcement
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Item 458/9309-01).
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2.3.3 Review of the Common Enclosure Associated Circuit Concern
Fire induced damage to non-essential circuits that are associated by common
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enclosure with circuits required to achievo and maintain safe shutdown may
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create circuit faults in electrically unprotected cables. Such faults could
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be of sufficient magnitude to create secondary fires.
If such secondary fires
were to occur in an enclosure which contained cables required for safe
shutdown, the successful achievement of safe shutdown could be adversely
affected.
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During the inspection, the licensee was not able to provide any information
related to the method in which this concern was evaluated. Subsequent to the
inspection, the licensee's contractor forwarded to the inspection team an
undated, unreviewed. " Preliminary for Clients Comments Document" describing
the overall methodology of its revalidation effort related to,the River Bend
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Station safe shutdown methodology. As stated in Section 4.5.2 of Appendix A
to this document:
"No actual assessment of protective device adequacy has
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been performed for nonsafe-shutdown circuits that may be associated with safe
shutdown cables." The inspection team's concern for the licensee's lack of an
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analysis for this concern was underscored by a review of a small sample of
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enclosures known to contain cables of both required and non-essential circuits
which identified the following examples of apparently inadequate electrical
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protection (i.e., fuse sizes which exceed the current handling capability of
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the cable they are protecting):
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CABLE
SSD REQD
N ON-
CABLE
CABLE
ELECTRICAL
ENCLOSURE
CKTS IN
ESSENTIAL
FUNCTION
Size
PROTECTION
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3D
ENCLOSURE 7
CIRCUIT ID
& TYPE
[
1TX0010
YES:
ICSHAOXBOO
INST.
16AWG
35 AMP FUSE
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2 CONDUCTOR
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1CSHBOX800
INST.
16AWG
35 AMP FUSE
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2 CONDUCTOR
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1CSHCOX800
INST.
16AWG
35 AMP FUSE
2 CONDUCTOR
1TC044B
YES
1ENBBBC950
CONTROL
16AWG
35 AMP FUSE
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125VDC
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8 CONDUCTOR
1ENSBBCS50
CONTROL
16AWG
35 AMP FUSE
125VDC
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8 CONDUCTOR
1RHSBBC950
CONTROL
16AWG
35 AMP FUSE
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125 VDC
4 CONDUCTOR
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As shown in the preceding table, several Circuits were found to have 16 AWG
cable protected by a 35 amp fuse. This size fuse is approximately twice as
large as the maximum fuse size specified by the National Electrical Code for
16AWG wire (18 amps - assuming wire having a temperature rating of 90'C). At
the time of the inspection, licensee representatives concurred with the
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inspection team's determination that the common enclosure associated circuit
concern does not appear to have been adequately analyzed in the River Bend
Station FHA. During a followup telephone conference call on April 27, 1993,
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the licensee informed the NRC that some of the above information regarding
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fuse and wire size provided to the inspectors was incorrect.
2.4 Conclusions
The team concluded that the licensee's lack of analysis, oesign basie, sr.d
assumptions for the spurious operation of equipment caused by potential fire
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induced damage and the cocanon enclosure associated circuit concern is an
apparent violation as discussed in Sections 2.3.2 and 2.3.3 above. The team
concluded, based on the limited scope and lack of completeness of the
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information provided to the team during the inspection, that the GSU
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engineering organization lacks the organizational depth to effectively resolve
complex fire protection issues. The licensee should discuss any needed
correcticns to inforsation provided to the team that may be important to the
conclusions stated above during the scheduled enforcement conference
(Escalated Enforcement Item 458/9309-01).
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3 ALTERNATE SHilTDOWN CAPABILITY (64100)
3.1 Overview
Alternate or remote shutdown capability _provides a means to safely shutdown
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and cooldown the~ plant in the event of a fire in the control room coincident
with a loss of off-site power. The team addressed alternate shutdown
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capability primarily by assessing the:
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Abnormal Operating Procedure A0P-0031, Revision 8A, dated February 9,
1993, with change notice dated March 1, 1993;
The alternative shutdown methodology upon which Procedure A0P-0031 is
based;
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The licensee's ability to implement Procedure A0P-0031;
Training lesson plans and training records for licensed and non-licensed
operators for Procedure A0P-0031; and
Licensee programs for ensuring and maintaining operability of
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alternative shutdown transfer and control functions.
3.2 Procedures
Procedure A0P-0031, " Shutdown from Outside Main Control Room," Revision 8A,
effective date February 9, 1993, provides the operator actions for performing
alternate shutdown in the event of a fire in and forced evacuation of the
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main control room.
3.2.1
Procedure Review and Walkdown
Procedure A0P-0031 assumed that the initiating event was a fire in the control
room which resulted in a manual reactor trip followed by a turbine trip with a
loss of offsite power. The following concerns, related to Procedure A0P-0031,
were identified by the team during the procedure review and walkdown.
With the main steam isolation valves closed, the operator was directed
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to control reactor pressure vessel pressure. The procedure stated that,
"If ILSV*C3A PVLCS AIR COMPRESSOR A is not available due to a main
control room fire, refer to Enclosure - 7."
Enclosure 7 was a jumpering
procedure for the subject compressor. The object was to have an
adequate air supply to operate three safety relief valves to control
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reactor pressure vessel pressure between 1064 psig and 800 psig. The
operator was cautioned that if the safety relief valve operating air
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supply was determined to be, or would become inoperabl , the reactor
pressure vessel should be depressurized with continuous safety relief
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valve opening so as not to exceed the remaining air supply in the event
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it was needed,
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Since the procedure called for the jumpering of the air compressor to
occur while the reactor was in hot shutdown, i.e, pressure above
1064 psig, the licensee was asked to explain the apparent contradiction
with the River Bend Station Safety Evaluation Report, Supplement 3,
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i 9.5.1.4 concerning safe shutdown capability.
Specifically, the
Safety Evaluation Report stated: "For a fire in the main control room,
air compressor ILSV*C3A may have to be started by use of jumpers at
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standby motor control center IEHS*MCC2L if additional air is required
for cycling the automatic depressurization system / safety relief valves.
Since these valves have a qualified air accumulator to provide for
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cyclic operation, it is anticipated that the air compressor will not
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need to be jump started until well into the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, if at all." The
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licensee was not able, during this inspection, to provide information to
determine the air supply requirements for the postulated 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to
achieve cold shutdown as required by 10 CFR Part 50, Appendix R,
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6 III.L. Therefore, it could not be established that jumpering of the
air compressor would not be needed early in the event. The licensee
planned to further evaluate this issue. This issue is one of three
issues related to Procedure A0P-0031 identified for followup inspection
(Inspection Followup Item 458/9309-02).
Attachment 1, Item 11, of the current Revision 8 of Procedure A0P-0031
stated that there was 15 minutes available to verify diesel generator
ventilation fans were running. There was a calculation intended to
justify the 15 minute time period, which also appeared in Revision 1.
From an initial temperature of 100'F, the results showed that to reach
120'F took only 29 seconds. The review results were:
" Based on the
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above, operator verification should be immediate, within 10 minutes.
10 minutes is justified based on timing for other items requiring
immediate verification.
Immediate verification is consistent with
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A0P-0004." The licensee had not provided justification for the
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15 minutes specified as allowable to verify tisat the diesel generator
ventilation fans were running and apparently had not acted upon the
results of, and recommendations drawn from, its own calculation.
In
response to the team's concern, the licensee stated that they would
evaluate these calculations. This issue is Part 2 of Inspection
Followup Item 458/9309-02.
In Procedure A0P-0031, Step 4.0, "Immediate Operator Actions," the
operator was instructed to initiate high pressure core spray, reactor
core isolation cooling and low-pressure core spray, if possible, prior
to evacuating the main control room. Although Procedure A0P-0031
directed the operator to rely upon several sources of reactor pressure
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vessel water level makeup in order of preference, i.e., feedwater and
condensate, reactor core isolation cooling, high pressure core spray and
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low-pressure core injectiaa there were no instructions in the procedure
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regarding the termination of reactor pressure vessel makeup flow upon
achieving high reactor pressure vessel water level. During the
procedural walkdown, the team questioned licensee personnel concerning
the fact that the procedure did not call for the operator to
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specifically terminate flow from the high pressure core spray system.
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The licensee response was that the operator would rely upon automatic
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termination of high pressure core spray when Reactor Vessel Level 8 was
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reached. The team then questioned whether the control circuitry for.
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such automatic termination had been protected in the event of a fire in
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the main control room. The licensee's response was that Level 8 trip
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circuitry was not protected.
The effects of excessive reactor vessel water inventory was addressed-in
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Stone & Webster t:ngineering Corporation letter dated April 15, 1991.
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The concerns identified in the referenced-letter involving the ability
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of the main steam isolation valves to close and the structural integrity
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of the piping were addressed in the licensee's Calculation
No. G13.18.12.2*11. The team determined that'the referenced Stone &
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Webster letter indicated several' serious deficiencies in the existing
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Procedure A0P-0031. The letter stated that ". . . SWEC [ Stone & Webster
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Engineering Corporation) has reviewed the potential for a spurious
operation of reactor _ pressure vessel injection flow paths and for safety.
relief valve actuation, caused by a fire induced hot short, and has
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concluded' that spurious operations of this nature must be postulated . .
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. SWEC has reviewed the potential for spurious operation of those
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systems, connected to the reactor pressure vesse;,-that could cause
changes in the RPV [ reactor pressure vessel] water level inventory, and
has determined the following:
. . . The existing fire hazards analysis
does not address a spurious actuation in which an increase in reactor
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water inventory occurs, such as an actuttion of HPCS [high pressure core
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spray], LPCI [ low-pressure core injection], LPCS [ low-pressure core
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spray], RCIC [ reactor core isolation cooling), or feedwater control
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valve failure."
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During this inspection the team determined that the licensee had not
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adequately addressed this issue raised by Stone & Webster Engineering
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Corporation.
In response to the team's concerns, the licensee stated
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that they would evaluate this issue. This issue is Part 3 of Inspection
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Followup Item 458/9309-02.
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During the walkdown, the inspectors observed that under reduced lighting
conditions, while controlling from the Division I Remote Shutdown Panel, it
was difficult to read the recorder for suppression pool level and temperature
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and drywell pressure and temperature. This issue was conveyed to the licensee
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for their consideration.
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The ter.m identified several issues noted above that are viewed as potential
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weaknesses in the capability to achieve and maintain safe shutdown in the
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event of a fire in the control room. The licensee indicated that these items
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would be entered into their internal tracking system for further study. The
concerns identified in this section have been designated as Inspection
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Followup Item 458/9309-02.
3.3 Operator Trainino
The licensee maintained lesson plans for " Hot License Operator Systems
Training," designated HLO, which were for initial licensing purposes, and
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" License Requalification Program," designated REQ, which were for
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requalification purposes for licensed operators. The following were relevant
to Procedures A0P-0031 and A0P-0052
HLO-537-0, " Shutdown from Outside Main Control Rrom (A0P-0031)", dated
October 6, 1992;
HLO-544-0, "A0P-0052 Fire Outside the Main Control Room," dated
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January 4, 1993;
REQ-225-1, " Review of A0P-0031 Remote Shutdown /Walkdown," dated
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September 5, 1991; and
REQ-219-0, " Procedure Review," dated April 4, 1990.
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The lesson plans were comprehensive and detailed. They appeared to cover all
major steps in the procedures.
The training record system had been revised in 1990. The licensee was asked
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to provide training records for one individual who was licensed at plant
startup in 1986 and who is currently licensed, and also for a non-licensed
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operator who was on the operating staff in 1986 and was currently unlicensed.
For the licensed operator, the licensee was only able to provide training
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records for REQ-225 back to 1990. No records prior to 1990 were provided to
the team during the course of the inspection. Only a partial record of non-
licensed operator training was provided, which occurred in 1988 for one
individual.
The excellence of the lesson plans was considered a strength in the training
program. Training records for the period prior to 1990 were generally not
available.
3.4 Inservice Testina of Remote Shutdown Capability
The licensee maintained three procedures for inservice testing of the remote
shutdown panels:
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STP-200-0601, " Division I Remote Shutdown System Control Circuit
Operability Test," Revision 7, dated September 27, 1991, with
Revision 7A effective through November 24, 1992;
STP-200-0602, " Division 11 Remote Shutdown System Control Circuit
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Operability Test," Revision 7, dated February 5, 1993; and
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STD '60-0603, " Division III Remote Shutdown S~ tem Control Circuit
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Operability Test," Revision 5, dated September 26, 1990, with
Revision 5A effective through April 29, 1991.
These procedures were to be performed at least once every 18 months.
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These procedures were comprehensive and directed the plant staff to isolate
control of all functions of the remote shutdown panels from the' main control
.
room during Operational Modes 1 and 2.
,
Each time the above procedures were implemented,, signed-off copies were stored
in the microfilm system. The licensee was able to provide signed-off copies
for all three procedures dating back to 1985.
The licensee's recordkeeping and implementation of the procedures is
considered satisfactory.
3.5 Conclusions
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Fire protection lesson plans were considered a strength, but training records
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for the period prior to 1990 were generally not available.
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The implementation of the remote shutdown operability test procedures and
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related recordkeeping was considered satisfactory.
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The team identified several issues noted above that are viewed as potential
weaknesses in the capability to achieve and maintain safe shutdown in the
,
event of a fire in the control room. The licensee indicated that these items
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would be entered into their internal tracking system for further study. The
concerns identified in this section have been designated as Inspection
Followup Item 458/9309-02.
4 FIRE PROTECTION / PREVENTION PROGRAM (64704)
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4.1 Overview
The purpose of this part of the inspection was to determine if the licensee
had established and wa; implementing a program for fire protection and
,
prevention in conformance with regulatory requirements, Technical
Specifications, and industry guides and standards.
'
4.2 Procram Review and Imolementation
The team reviewed the licensee's fire protection procedures listed in'
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Attachment 3 to this report. This review indicated that the licensee had
technically acceptable procedures to implement the fire protection program.
Procedural guidance was provided to control combustible material and reduce
fire hazards. Administrative procedures also provided for maintenance and
surveillances on fire suppression, detection, and support equipment.
,
Personnel training, qualificctions, and responsibilities were satisfactorily
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provided. Maintenance evolutions that significantly increase fire risk were
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properly controlled.
4.3 Surveillances
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In this area, the team reviewed a sample of records for surveillances
conducted since the last inspection to verify that:
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The fire detection and suppression systems met the Technical-
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Specification operability testing requirements, and
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Operability for these systems had been satisfactorily demonstrated at
the required frequencies.
The team reviewed the surveillance procedure data packages listed in
Attachment 3.
The inspectors found that Technical Specification required
surveillances were being conducted at the required frequencies.
Prompt
actions tsad been taken to repair defective components and it appeared that
appropriate compensatory actions were taken when required. . However, several-
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surveillance test procedures were closed as " Acceptable With Comments." .In
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some instances it was not clear that appropriate actions were taken in
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response to the comments. For example, Procedure STP-251-3701, " Fire
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Protection' Sprinkler System 3 Year Air How Test," which was completed on
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October 31, 1990, was accepted subject to comments on 12 data sheets. No
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followup documentation was referenced. Some comments referenced apparent
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procedure errors, e.g., the number of nozzles found n s different than the
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number listed in the procedure. On Data Sheet 6, it was noted that 21
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sprinkler nozzles were found severely clogged and the nozzles were removed and
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cleaned by the test technician, apparently without issuance of a maintenance
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work order.
It was also noted that a sample of the " crud" was given to the
fire protection coordinator. However, from the documentation it was not clear
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what actions may have been taken to prevent recurrence.
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Procedure STP-251-3505, " Fire Protection Sprinkler System Functional Test,"
completed on August 6, 1992, there was a comment on a data sheet indicating
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that St.eps 7.4 and 7.5 could not be pgrformed as written.
It was not clear
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what action was taken on this comment. However, it did not appear to affect
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the acceptance criteria.
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The licensee was conducting surveillances at the required freouencies to
demonstrate the operability of Technical Specification required plant fire
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protection equipment. Although the above documentation problems did not-
impact the operability of Technical Specification required components, they
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did raise questions on whether appropriate work control procedures were
followed or appropriats followup action was taken. This issue was conveyed to
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the licensee for their consideration.
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4.4 Plant Walkdown
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The team toured the site area to observe the main fire water supply system.
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The tour included the fire water storage tanks, electric fire pump, both
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diesel fire pumps, and the system jockey pump. The suction and supply valves,
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tank isolation valves, and selected valvcs on the main loop were observed to.
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be properly aligned. Hose houses were in very good condition and along with
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.the fire hydrants, fully accessible.
Accessible areas in the plant were also toured to observe general area
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conditions, work activities in progress, and visual conditions of fire .
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protection systems and equipment. Combustible materials and~ flammable and
combustWie liquid and gas usage were restricted or properly controlled in
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. areas containing safety-related equipment and components.
Items checked
included positions of selected valves, fire barrier condition, hose stations,
fire lockers, and fire extinguishers for type, location, and condition. There
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was no welding, cutting, or use of open flame ignition sources found in the
areas toured. General housekeeping conditions were found to be good.
Fire
brigade equipment, including emergency breathing apparatus, was found to be
properly stored and maintained.
During the walkdown, the team noted that Fire Door CB-098-08 had a sill gap of
approximately 1 inch. This was a metal, hollow core, double leaf, 3-hour
rated fire door with a concrete sill. The National Fire Protection
Association Fire Code 80 (NFPA 80), stipulates a maximum sill gap of 3/8 inch.
This was pointed out to the licensee representative and the team was advised
that it would be evaluated. Subsequently, the licensee informed the team that
there was a history associated with the acceptance criteria for fire door
clearances and provided a copy of a page from Procedure STP-000-3602, " Fire
Barrier Visual Inspection," that detailed current acceptance criteria for fire
doors. The acceptance criteria dimensinns are not in agreement with NFPA 80.
The licensee accepts a maximum clearano between door _and door frame of
1/4 inch and a maximum clearance between the door bottom and the floor or
sill, if provided, of 1 inch. NFPA 80 stipulates the clearance between the
door and the frame shall not exceed 1/8 inch, and the clearance between the
bottom of the door and a raised noncombustible sill shall not exceed 3/8 inch,
where there is no sill the maximum clearance shall not exceed 3/4 inch. The
licensee has justified the variance to the NFPA Code through the acceptance of
a test report on a special test run by Warnock Hersey International, Inc., for
Bechtel Corporation in 1986 in conjunction with their work at Palo Verde.
This special test was run specifically to validate excessive gaps in this type
of fire door configuration. The office of Nuclear Reactor Regulation has
found these test reports acceptable for similar applications at other
utilities. The licensee's acceptance criteria of 1/4 inch and 1 inch,
respectively, is considered satisfactory because it is bounded by the special
test.
4.5 Fire Bricade Trainina/ Drills
The team reviewed fire brigade training and drill records. Therecordswere
in order and confiraed that training and drills were being conducted at the
specified intervals.
Initial and qualification maintenance fire brigade training was provided by
the Nuclear Training Representative for Fire Protection. An interviev of the
training representative determined that the brigade members also received
initial and refresher formal off-site training at the Louisiana State
University Fire School. Brigade members received annual refresher training at
the same facility.
The training representative provided fire watch
qualification training, and maintained the status of personnel qualification.
The team interviewed two roving fire watch personnel. This interview
determined that the individuals were familiar with the administrative
procedure for fire watches, and they were knowledgeable of fire watch duties
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and the basics of fire suppression. The individuals expressed satisfaction
with the level of training received and felt that it was adequate.
The licensee conducted an unannounced fire drill during this inspection and
the team observed the fire brigade response. The scenario was that a fire had
occurred in the 4160V high pressure core spray switchgear in the control
building Standby Switchgear Room IC. The five member fire brigade composed of
the fire brigade leader, two nuclear equipment operators, and two' security
personnel arrived at the assembly point within 5 minutes of the initial
announcement and alarm. Dress out was prompt and orderly. The leader
contacted the control room for the initial status, consulted the prefire
strategy for the affected area, and directed the brigade on what equipment to
bring. The brigade left the locker area and proceeded to the fire scene. The
leader contacted the control room to provide an update and directed the
brigade on approach and entry. Brigade conduct was appropriate and strategy
was satisfactory. Mutual aid was requested by the leader early in the
scenario.
Hutual aid is provided by an agreement between the licensee and three local
volunteer fire departments. These volunteer fire departments are trained by
the fire protection nuclear training representative. The training consists of
basic radiation protection and dosimetry information, plant-specific hazards,
plant layout, and access points. Annual participation in at least one drill
is satisfied by involving the departments in at least one emergency
preparedness drill each year.
The team interviewed the fire brigade leader. He had been a member of the
fire brigade for 4 years, was a licensed reactor operator and completed his
fire brigade leader training within the last month. This was his first drill
as a brigade leader. He was very satisfied with both the quantity and quality
of training and had no reservations acting as brigade leader.
4.6 Fire Protection Ouality Assurance
Quality assurance audits for the past 2 years were reviewed by the team.
These audits were identified as Audit 91-12-I-PFPP (FU), dated January 6,
1992, " Fire Protection Program Followup" and Audit No. 91-03-I-PFFP, dated
July 7, 1992, "GSU QA Audits of Operation River Bend Station Fire Protection
Program." The audits addressed fire brigade drills, fire watches,
organization and procedures, and overall adequacy and effectiveness of the
fire protection program at River Bend Station. Discrepancies identified were
formally presented to the responsible organizations.
Responses were tracked
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to closecut, and the actions taken were reviewed for adequacy by the
appropriate organizations.
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GSU uses fire protection specialists from other nuclear utilities every third
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year to assist in the conduct of the fire protection audits.
In general, the
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team concluded that the quality assurance audits were effective in identifying
problems in this area; however, the audits did not note the specific concerns
of incomplete analysis to support conclusions in the FHA or the timeliness of
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corrective actions on fire protection issues discussed in this report.
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4.7 Conclusions
The team concluded that, except as noted elsewhere in this report, the
licensee had established and was implementing an effective fire protection
program. Appropriate procedural controls were in effect to reduce fire
hazards and implement the required fire systems / equipment surveillance tests.
Fire brigade training and qualifications of personnel were considered
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strengths and actual performance during a drill of those perswnel Was
considered good.
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The team identified surveillance test results that were closed as " acceptable
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with comments" when problems were identified during the conduct of a
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surveillance. This was considered a weakness because it was not evident that
appropriate followup action was taken or that appropriate work control
procedures were followed.
Quality assurance audits of this area have been generally effective.
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5 FOLLOWP ON CORRECTIVE ACTIONS FOR VIOLATIONS (92702)
(Closed) Violation 458/9002-02:
Failure to Fully Imolement the Fire
Protection Prooram Aporoved by the NRC
The issue of failu e to fully implement the fire protection program approved
by the NRC has beer incorporated into the apparent violation identified in
Sections 2.3.2 and 2.3.3 of this report.
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6 FOLLOWP (92701)
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6.1
(Closedl Unresolved Item 458/9204-01:
River Bend Station Specific
Thermo-Lao Issues
This item was considered an unresolved item based on the licensee's initial
determination of the adequacy of Thermo-Lag application, installation, and
qualification testing.
The specific issues were defined in five items in the
inspection report and identifiea as paragraphs 4.2.1 through 4.2.5.
Generic
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Thermo-Lag issues will remain open as indicated in the following paragraphs.
Issue 4.2.1, Thermo-Lag Removal: This item involved the removal of
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Thermo-Lag in a number of plant locations around junction boxes, conduit
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seals, and wall penetrations. During the initial inspection, the
licensee stated that the Thermo-Lag had been removed in order to inspect
internal conduit seals and wall penetrations.
By letter dated May 6,
1992, the licensee stated that all the removed sections of Thermo-Lag
had been reinstalled and sections of Thermo-Lag removed in the future
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would be reinstalled upon completion of the work acti"ities
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necessitating the initial removal. During this inspection, the team
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verified that all the removed Thermo-Lag had been installed in
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accordance with the established procedures.
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Issue 4.2.2, Structural Integrity of Thermo-Lag Installations: This
item pertained to the possible inadvertent leakage from the fire
protection system that caused damage to the Thermo-Lag material. During
the initial inspection, visible damage to a portion of Thermo-Lag fire
barrier material in "F" tunnel was noted. The licensee was apparently
unaware that the condition existed prior to the inspection and was not
able to identify the source of water damage.
By letter dated May 6,
1992, the licensee stated that the damage to the "F" tunnel enclosure
was caused by a leak above the enclosure.
It was also determined by
the licensee that the leak had caused degradation to the trowel grade
miterial applied to the seams and joints of the enclosure and no damage
to the base Thermo-Lag material had occurred. These enclosures are
provided with drainage to ensure that the weight of water assumed in the
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design will not be exceeded. During this inspection, the team verified
that the degraded trowel grade material in the seams and joints had been ,
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repaired.
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Issues 4.2.3, Qualification Testing of Installed Configurations; 4.2.4,
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Electrical Cable Ampacity Derating; and, 4.2.5, Fire Test Acceptance
Criteria, are closed based on the fact that they are involved in the
generic concerns of Thermo-Lag. They will be tracked as a single
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Inspection Followup Item 458/9309-03.
6.2 (Closed) Unresolved Item 458/9204-02: Appendix R Issues
This was considered an unresolved item pending the licensee's response and
actions to the six issues identified in the inspection report and identified
as paragraphs 5.1.1 through 5.1.6.
This inspection determined that
issues 5.1.1, 5.1.2, 5.1.3, 5.1.5, and 5.1.6 could be closed. The remaining
issue, 5.1.4 is the subject of concern identified above as Escalated
Enforcement Item 458/9309-01. That issue remains opt.n and will be tracked
under that tracking number.
Issue 5.1.1, Electrical Separation for Spent Fuel Pool:
Spent fuel pool
equipment required for cooling was identified in Licensee Event
Report 91-008, Supplement 1, as not Laing adequately separated
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electrically from the control room.
Immediate corrective actions
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implemented by the licensee included the revision of Procedure A0P-0031
(Shutdown From Outside The Main Control Room) to provide operator
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guidance necessary to perforn manual actions to restore spent fuel pool
cooling in the event of a control room fire.
Long-term corrective
actions were described in Modification Request 92-0038.
The
modification included the installation of isolation / transfer switches
outside the main control room to obtain electrical independence between
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the control room and affected spent fuel pool cooling equipment.
Issue 5.1.2, Lack of Automatic Control of Dampers in Fuel Building:
Fuel Building Ventilation Dampers 1HVF*AOD037A, 102 and 122 were
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identified in the FHA as equipment required for spent fuel pool cooling.
Potential fire damage to electrical cables located in Fire Area FB-1
could cause the spurious operation of these dampers resulting in a loss
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of ventilation to the spent fuel pool cooling pump and, thus, the loss
of spent fuel pool cooling. The immediate corrective action taken by
River Bend Station was to treat the affected cabling as having missing
fire barriers and to post a continuous fire watch in accordance with the
plant Technical Specifications. The fire watch remained until the pre-
fire strategies were revised to identify manual actions required to
place the dampers in the required position. Attachment 6 of
Procedure A0P-0052, " Fire Outside Main Control Room," provided
appropriate procedural guidance to verify that spent fuel pool cooling
is maintained in the event of fire in the fuel building (Fire Area B-1).
Issue 5.1.3, 20-Foot Separation in Reactor Building:
The licensee
committed to have an independent contractor perform a detailed review
and verification of the plant FHA. During this review, it was found
that cables required for operation of Containment Unit Coolers IHVR*UCIA
and IHVR*UCIB did not meet the 20-foot horizontal separation criteria
stated in Section III.G.2 of 10 CFR Part 50, Appendix R.
The
containment unit coolers provide area cooling within the containment
building outside the drywell. The coolers were included in the original
FHA to assure that temperature in containment would remain below the
equipment qualification maximum temperature of 165'F. The immediate,
interim, corrective action taken by the licensee was to initiate an
hourly firewatch. As part of its permanent corrective action, the
licensee performed an analysis which demonstrated that the containment
unit coolers are not required for safe shutdown. This analysis was
documented in Condition Report 92-0031.
Issue 5.1.4, Lack of Fire Hazard Analysis: This issue involved a lack
of a FHA for a portion of the "D"
tunnel in the electrical cable room.
During the initial inspection, a preliminary analysis for this area was
completed.
By letter dated May 6, 1992, the licensee stated that as a
corrective action, an analysis was conducted on the cables which were
routed through the room. This analysis revealed that the high pressure
core spray system would not be affected by a fire in the room. The
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licensee also initiated a modification request to revise the FHA to
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incorporate the evaluation of this room. During this inspection, the
team verified that the modification request had been completed.
However, the licensee had not completed incorporation of all
documentation into a complete FHA. As of this inspection, the licensee
has not completed the associated circuits analysis in the final FHA
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evaluation.
This issue is the subject of concern identified above as
Escalated En.: cement Item 458/9309-01 and will be tracked by that
number.
Issue 5.1.5, lack of a Breaker / Fuse Coordination Analysis for
125VDC/120VAC Circuits: The licensee had prepared a comprehensive
breaker / fuse coordination evaluation for 125VDC and 120VAC circuits.
This analysis was found to be documented in Calculation
No. G13.18.3.6*5, " Coordination Study of Appendix R and Class IE Low
Voltage Protective Devices," prepared by Halliburton NUS Environmental
Corporation, an independent contractor to GSU. The team reviewed this
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document and conducted related discussions with plant engineering staff
members. The level of coordination was found acceptable in a sample of
circuits selected for review.
Issue 5.1.6, Lack of a High Impedance fault Analysis:
In response to
findings discussed in NRC Inspection Report 50-458/92-04, the licensee
had hired an independent contractor to evaluate this issue and provide
recommended corrective actions as necessary. As a result of this
review, the licensee had developed compensatory procedures which
provided operator guidance in the event a power supply was lost due to
the occurrence of fire induced high impedance faults on unprotected
cabling. During the inspection, the licensee was found to have
incorporated an acceptable level of procedural guidance into
Procedures A0P-0031, " Shutdown From Outside The Control Room," and
A0P-0052, " Fire Outside the Main Control Room (In Areas Containing
Safety Related Equipment)." A review of these procedures found them to
provide sufficient guidance to permit operators to identify affected
power supplies and take corrective actions (i.e., non-essential load
shedding) necessary to restore operability.
6.3 (Closed) Inspection Followuo Item 458/8937-02: Motor-0perated Valves
Not Deeneroized - Conflicts with Fire Hazards Analysis
This followup item was originally opened as an unresolved item to track the
issue that was later identified as Violation 458/9002-02 (EA 90-039) discussed
above.
7 DNSITE REVIEW OF LICENSEE EVENT REPORTS (92700)
7.1
(Closed) Licensee Event Report 458/88-009: Unsealed Fire Barrier
The licensee determined that there were 56 conduits with damaged seals or
uninstalled seals. The team reviewed the licensee event report and determined
that it was complete, accurate and submitted in a timely manner. The fire
barrier penetrations were located in the control building, diesel generator
building, auxiliary building
"D" tunnel, reactor building shield wall at
elevations 114 and 141 feet, and the fuel building. The licensee's imme .ite
corrective action was to establish fire watches in the affected areas in
accordance with Technical Specification requirements, and conduct a
100 percent inspection of all fire barrier penetrations. The licensee had
issued a modification request to repair and/or replace the damaged seals and
install the missing seals. The licensee's corrective actions were considered
appropriate and in sufficient ietail to prevent recurrence of this event. All
corrective actions related to this licensee event report are expected to be
completa hv December 1993.
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7.2 (Closed) Licensee Event Report 458/89-005: Fire Seal Penetration IC2W19
Inadeouate Due to Poor Aeolication/ Inspection Technicue
During this inspecticn, the team reviewed this licensee event report for
accuracy, completeness, and timeliness.
This issue involved the voids found
in the penetration seal caused by improper installation techniques and
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inadequate inspection. The seal was repaired under Maintenance Work Order 104235. The licensee also conducted a random inspection of similar
configurations to assure that no other inadequate seals existed. The licensee
stated that completion of this action had been delayed for 2 years due to
organizational changes at River Bend Station. The team evaluated the
licensee's corrective actions and determined that their approach was sound,
but not timely. The licensee has completed about two-thirds of the planned
inspections and expect all work activities to be completed by December 1993.
7.3
(Closed) Licensee Event Report 458/90-017: Inadecuate Fire Barrier in
Shake Space
During the performance of Surveillance Test Procedure STP-000-3602, the
.lcensee identified two voids in a fire barrier in the shake space between the
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auxiliary building and the containment shield wall. These voids were above
the main steam tunnel walls and extended through the thickness of the
elevation 141 feet slab. Apparently, the seal material was not installed
during initial construction through an oversight. The licensee immediately
added the area to the roving fire watch schedule. The penetration seal was
completed under Maintenance Work Order 135289. On the inside face, a seismic
gap seal was installed and the surveillance procedure was changed to include
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instructions to inspect the seals. This maintenance work order was properly
completed, although overall resolution of the issue was not completed in a
timely manner.
7.4
(Closed) Licensee Event Report 458/91-005: Desian Deficiencies in Fire
Doors
A licensee conducted quality assurance audit identified a design deficiency
with Fire Door CB-70-25. This is a double leaf, normally open, fused link
fire door. The deficiency in design was that no coordinating device was
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provided that would assure the proper closing sequence for the active and
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inactive leaves of the door.
Immediate compensatory measures were taken by
establishing an hourly fire watch. The licensee then reviewed other double
fire doors for a similar problem. This review identified one additional door,
CB-98-32, with the same configuration. This door was also added to the hourly
fire watch log. Long-term corrective action was to issue Temporary Change
Notice 91-0213 to Surveillance Procedure STP-000-3001, requiring that the
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inactive leaf be closed with latches engaged, and that the active leaf way be
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frae of obstructions. This was also extended to 4 clud: "rrifying that the
latch bolts on the inactive leaf of all normally closed double doors were
engaged. This review found that the corrective actions taken by the licensee
were acceptable and d'd not identify any concerns.
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7.5
(0 pen) Licensee Event Report 458/92-003: Deviations From Aporoved
Desions in Structural Steel Fireproofina
During this inspection, the team reviewed this licensee event report for
accuracy and completeness. The licensee's investigation determined that the
structural steel supporting required fire barrier walls and floors could not
be considered as being protected to a fire resistance rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> in
accordance with Underwriters Laboratories tested designs. Although this
condition was found by the licensee on February 22, 1992, it has existed since
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the plant startup. The primary root cause was identified as an inadequate
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level of engineering evaluation applied in the development of the fire barrier
designs.
The licensee declared the structural steel fireproofing doors / walls
Limiting condition for operation action statements specified by
Technical Specification were implemented. The licensee was in the process of
revising the design specification of fire proofing. The scheduled completion
date was December 1993. This licensee event report shall remain open until an
acceptable design change has been completed.
7.6 (Closed)
Licensee Event Reoort 458/91-008:
Fire Hazards Analysis
Deficiencies Includina lack of Fire Wrao/Inadecuate Fire Barrier
The team reviewed this licensee event reoort and determined that the licensee
had not completed the FRA.
During this inspection, the licensee stated that
the FHA will be completed at a later date by an independent contractor. This
action will be tracked as a part of the apparent violation as an inadequate
corrective action taken by the licensee stated in the Section 2.3.2.
7.7 (0 pen) Licensee Event Report 455/89-010: Missina or inadeouate
Penetration Seals Per Technical Soecification 3.7.7.a
The team reviewed this licensee event report and the corrective actions taken
by the licensee. At the time of this inspection, the licensee's 100 percent
reinspection was about two-thirds complete. The licensee had previously
indicated that all corrective action was scheduled to be complete on this
licensee event report by December 1993, but the licensee does not currently
expect for these actions to be cleplete until after the refueling outage in
the spring of 1994. This licensee event report shall remain open until
corrective actions are complete. The corrective actions for this licensee
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event report have not seen timely.
7.8 (Closed) Licensee Event Report 458/92-007: Vulnerability to Hot Shorts
Discovered as a Result of Information Notice 92-18
During its review of information notice 92-18, the licensee found that control
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circuits for motor-operated valves required for alternate shutdown of the
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plant could spuriously operate during a control room fire and on April 27,
1992, the licensee forwarded Licensee Event Repart 92-007 to notify the NRC of
its identification of this design discrepancy. As noted in a subsequent
revision of thic licensee event report (LER 92-007, Revision 1, dated
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September 4,1992) the .icensee had implemented modifications to rework the
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control circuit wiring of 46 motor-operated valves at the motor control
centers and remote shutdown panel so that the limit switches and torque
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switches cannot be bypassed by hot shorts in the control room. During this
inspection, the modifications associated with Motor-0perated Valve ISVV*HOVIA
as documented in Modification Request 92-0042 were reviewed in detail with
representatives of the licensee's staff. The licensees corrective actions-
with regard to spurious motor-operated valve operability concerns described in
information notice 92-18 were found to be acceptable.
7.9 Conclusions
The team concluded that the licensee's corrective actions associated with
licensee event reports identified above has not been timely. Most of the
major fire protection issues were identified 3 to 5 years ago and the issues
ir,volve matters that have existed since the plant was licensed in 1985.
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ATTACMMENT I
1 PERSONS CONTACTED
1.1 Licensee Personnel
- D. Andrews, Director, Quality Assurance
- T. Anthony, Supervisor, PP&T
- R. Backen, Supervisor, Quality Assurance Systems
- C. Ballard, Supervisor Contract Services
- R. Biggs, Supervisor, Operations Quality Control
- J. Blakley, Acting Assistant Plant Manager, Systems Engineering
- L. Borel, Senior Mechanical Engineer
S. Bougeus, Contract Services Staff Assistant
- R. Buell, Supervisor, Non Safety Systems
J. Burton, Supervisor, Prababilistic Risk Analysis
- J. Cook, Senior Technical Specialist
- W. Curran, Cajun Electric Site Representative
- B. Ellis, Maintenance Fire Protection Coordinator
L. England, Director, Nuclear Licensing
- C. Fantacci, Supervisor, Radiological Engineering
- C, Fisher, Quality Assurance Engineer
- D. Freehill, Assistant Plant Manager, Outage Management
K. Garner, Licensing Engineer
- A. Garrett, Senior Electrical Engineer
- J. Hamilton, Manager, Engineering
- W. Hardy, Supervisor, Radiation Protection
- D. Hartz, Outage Director
- T. Hoffman, Supervisor, Civil / Structural Design Engineering
- R. Kerar, Fire Protection Engineer
- G. Kimmell, General Maintenance Supervisor
- T. Knight, Licensing
- T. Lacy, Outage Director
- D. Lorfing, Supervisor, Nuclear Licensing
- J. Maher, Licensing Engineer
- l. Malik, Supervisor, Operations Quality Assurance
R. Malls, Superintendent, ANC0/ Maintenance
- J. Head, Supervisor, Electrical and Special Projects
- J. Miller, Director, Engineering Analysis
- W. Odell, Director, Radiological Programs
- S. Raderbaugh, Assistant Plant Manager, Maintenance
- J. Richmond, Senior Systems Engineer
- J. Salmon, Motor-0perated Valve Program Coordinator
- J. Spivey, Jr., Senior Quality Assurance Engineer
- M. Stein, Director, Design Engineering
D. Steinsiek, Senior Engineer
- A. Wells, Radiological Health Supervisor
- D. Williamson, Senior Nuclear Engineering Technologist
- L. Woods, Shift Supervisor
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1.2 NRC Personnel
- W. Smith, Senior Resident Inspector
2 EXIT MEETING
An exit meeting was conducted at the conclusion of the onsite inspection on
April 2, 1993. The findings were discussed with licensee representatives
identified in the attached report. Additional information was provided to and
reviewed by the NRC team after the onsite portion of the inspection.
Telephone conversations were held between Mr. Hamilton and other licensee
representatives of your staff and Mr. Constable and members of our staff to
clarify certain issues on April 16, 22 and 27,1993.
Proprie+ ary information
provided to the inspection team will be returned to the licensee and was not
reproduced in this report.
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ATTACHMENT 2
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INSPECTION FINDINGS INDEX
EEI 458/9309-01
Open
Section 2.3.2 and
Section 2.3.3
IFI 458/93-09-02
Open
Section 3.2.1
VIO 458/9002-02
Closed Section 5
URI 458/9204-01
Closed Section 6.1
IFI 458/9309-03
Open
Section 6.1
URI 458/9204-02
Closed Section 6.2
IFI 458/8937-02
Closed Section 6.3
Closed Section 7.1
Closed Section 7.2
Closed Section 7.3
Closed Section 7.4
Open
Section 7.5
Closed Section 7.6
Open
Section 7.7
Closed Section 7.8
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ATTACHMENT 3
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DOCUMENTS REVIEWED
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River Bend Station Updated Safety Analysis Report (USAR) Fire Protection
Program Evaluation Report
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License Change Notice (LCN) 9.5-67, "NRC Information Notice 86-35," dated
September 14, 1990
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LCN 9.5-74, " Plant Modification Request (PMR) 92-0005," dated March 21, 1992
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LCN 9A.2-20, "PMR 92-0005," not dated
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LCN 9A.2-23, " Change USAR Drawing 9A.2-12 for Services Building Fire
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Protection Arrangement," dated March 1, 1993
LCN 9A.3-3, "USAR Changes Associated with Revision of RBNP-038," dated
January 11, 1990
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LCN 9A.3-5, "USAR Change in Commitment to Section C.5.a(3) of BTP CMEB 9.5-1,"
dated January 13, 1992
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Procedure FPP-0010, " Fire Fighting Procedure," Revision 6
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Procedure FPP-0020, " Guidelines for Preparation of Pre-Fire Strategies and
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Pre-Fire Plans," Revision 78
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Procedure FPP-0030, " Storage of Combustibles," Revision 7
Procedure FPP-0040, " Control of Transient Combustibles," Revision 6A
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Procedure FPP-0060, " Hot Work Permit," Revision 6A
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Procedure FPP-0070, " Duties of Fire Watch," Revision 78
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Procedure FPP-0100, " Fire Protection System Impairment," Revision 6A
Procedure FPP-0101, " Monthly Fire Suppression System Inspection," Revision 4A
Procedure FPP-0103, "C02 System Functional Test," Revision 1
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Procedure FPP-0106, " Annual Halon System Functional Check," Revision 2
Procedure FPP-0090, " Fire-Fighting Equipment, Inventory, Inspections, and
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Maintenance," Revision 7
Surveillance Test Procedure (STP)-251-3100, " Weekly Diesel Fire Pump Battery
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Test," performance data for February 8, 19P- February 1, 1992; January 18,
1993; January 25, 1993; January 29, 1992; _
August 20, 1992
-1-
)
i
-
~e
. r
.
.
STP-251-3701, " Fire Protection Sprinkler System 3 Year Air Flow Test,"
performance data for October 31, 1990
STP-251-3203, " Motor Driven Fire Pump Monthly Operability Test," performance
data for December 7, 1992; November 9, 1992; January 5, 1993; and
February 1,1993
STP-253-3200, " Monthly PGCC Halon System Bottle Pressure," performance data
for August 24, 1992; July 24, 1992; and June 25, 1992
STP-251-3606, " Diesel Fire Pump 18 Month Inspection," performance data for
October 7, 1992
STP-251-3205, " Diesel Fire Pump Operational Test," performance data for
February 9, 1993; February 1, 1993; January 26, 1993; and January 11, 1993
STP-251-3502, " Fire Protection Water Valve Cycle Test," performance data for
June 22, 1992
STP-251-3505, " Fire Protection Sprinkler System Functional Test," performance
data for August 6, 1992
STP-201-3601, " Fire Protection Header Nozzle Inspection," performance data for
June l?, 1992
STP-251-3700, " Fire Protection Water System 3 Year Flow Test," performance
data for October 14, 1992
,,
STP-251-3602, " Fire Pump Annual Function Test," performance data for
August 28, 1992
STP-253-3203, "PGCC Halon System Actuation and Flow Test," performance data
for December 30, 1992
STP-253-3400, "PGCC Halon Storage Tank Weight / Pressure Check," performance
data
Procedure PMP-1019, " Preventive Maintenance and Periodic Testing of Faergency
Lighting," performance data for March 18, 1993; March 23, 1993; March 1, 1993;
March 15, 1993; and September 3, 1992
STP-251-0204, " Fire Protection Water System Monthly Valve Posit',on Check,"
performance data for March 2, 1993; February 1, 1993; and January 4, 1993
STP-251-3101, " Fire Protection Water System Minimum Water Volume Check,"
perfr---- e data for March 1, '.993; February 1, laa3; February 8, 1993; and
February 15, 1993
STP-251-3300, " Quarterly Diesel Fire Pump Battery Test," performance data for
May 26, 1991 and March 20, 1992
-2-
'
.
__
_
w.
- '
.,
1
.
STP-000-3604, " Fire Barrier 18 Month Visual Inspection," performance data for
March 27, 1992 and August 26, 1992
STP-250-3501, "Six Month Fire Detector Instrumentation Functional Test,"
performance data for August 13, 1992 and March 10, 1993
STP-251-3600, "18 Month Diesel Fire Pump Battery Surveillance," performance
data for July 28, 1991
STP-200-0601, " Division I Remote Shutdown System Control Circuit Operability
Test," Revision 7, dated September 27, 1991, with Revision 7A effective
through November 24, 1992
STP-200-0602, " Division II Remote Shutdown System Control Circuit Operability
Test," Revision 7, dated February 5, 1993
STP-200-0603, " Division III Remote Shutdown System Control Circuit Operability
Test," Revision 5, dated September 26, 1990, with Revision 5A effective
through April 29, 1991
HLO-537-0, " Shutdown from Outside Main Control Room (A0P-0031)", dated
October 10, 1992
HLO-544-0, "A0P-0052 Fire Outside the Main Control Room," dated January 4,
1993
REQ-225-1, " Review of A0P-0031 Remote Shutdown /Walkdown," dated September 5,
,
1991
REQ-219-0, " Procedure Review," dated April 4, 1990
.
-3-
.
e)#
Yk
ATTACHEENT 4
?
TABLE I
COORDINATION OF ELECTRICAL PROTECTIVE DEVICES
l
SSD METHOD
SELECTED COMP-
POWER SOURCE
REVIEW COMMENTS
1
STBY SW PUMP C
IE22*S004
ACCEPTABLE: Calc E200, Attach 3,
ISWP*P2C
Page 27
1
IEHS*MCC8A
ACCEPTABLE: Calc E200, Attach 3
ISOLATE EMERG
FED FROM
480V load ctr:
ISWP*MOV506B
IEJS*SWGIA
1
PRESSURE TX
IVBS*PNLOlA
ACCEPTABLE: Calc G13.18.3.6*5
ICMS*PT2A
CKT 11
2
IENS*SWGlB
ACCEPTABLE: Calc E200, Attach 3
IE12*PC002B
2
CONT BLDG CHILLED
1EHS*MCC14B
ACCEPTABLE: Calc E200, Attach 3
WATER COND. COOLING
FED FROM:
WTR PMP D
lEJS*SWGIB
ISWP*P3D
2
IEHS*McC22
ACCEPTABLE: Calc E200, Attach 3
YLV ISOLATION
lEJS*SWG2A
ISWP*MOV74A
2
ADS SYST LVL TX
IENB*PNLO2A CKT 7
ACCEPTABLE:
Calc E200, Attach 3,
~1821*LTN095B
AND lENB*PNLO2B
Page 88 AND 158
CKT 12
VIA OPTICAL
ISOLATOR
TABLE 2
CIRCUIT BREAKER AND RELAY TEST PROCEDURES REVIEWED
NUMBER'
TITLE
REV
DATE
MCP-1031
Testing and Calibration of GE Relays IFC66KD
4
8/24/92
PMP-1020,
PREVENTIVE MAINTENANCE OF THERMAL OVERLOAD RELAYS, UNITIZED
4
6/30/92
!
AND MOLDED CASE CIRCUIT BREAKERS
i
PMP-1014,
PREVENTIVE MAINTENANCE OF MOTOR CONTROL CENTER
2A
1/25/90
,
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