ML040090435
| ML040090435 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 12/22/2003 |
| From: | Ogle C NRC/RGN-II/DRS/EB |
| To: | Stall J Florida Power & Light Co |
| References | |
| FOIA/PA-2003-0358 IR-03-002 | |
| Download: ML040090435 (52) | |
See also: IR 05000335/2003002
Text
May XX, 2003
Florida Power and Light Company
ATTN: Mr. J. A.[Stall, Senior Vice President
Nuclear and Chief Nuclear Officer
P. 0. Box 14000
Juno Beach, FL 33408-0420
SUBJECT: ST. LUCIE NUCLEAR PLANT - NRC TRIENNIAL FIRE PROTECTION
INSPECTION REPORT 50-335/03-02 AND 50-389/03-02
Dear Mr. Stall: -)
On March 28, 2003, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection
at your St. Lucie Nuclear Plant Units 1 and 2. The enclosed inspection report documents the
inspection findings, which were discussed on March 28, 2003, with Mr. D. Jernigan and other
members of your staff.
The inspection examined activities conducted under your license as they relate to safety and
compliance with the Commission's rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
This report documents a finding concerning silicon oil filled transformers in the B Switchgear
Room which had not been considered or evaluated in the licensee's fire hazards analysis.
Additionally, a finding was identified concerning the crediting of manual operator actions outside
the main control room in lieu of physical protection of cables and equipment relied on to achieve
safe shutdown during a fire, without prior NRC approval, for areas designated as 10 CFR 50
Appendix R,Section III.G.2. These findings involved violations of NRC requirements. These
findings collectively have potential safety significance greater than very low significance.
However, a safety significance determination has not been completed. These findings did not
present an immediate safety concern. In addition, the report documents one NRC-identified
finding of very low safety significance (Green), which was determined to involve a violation of
NRC requirements. However, because of the very low safety significance and because it was
entered into your corrective action program, the NRC is treating this as a non-cited violation
(NCV) consistent with Section V.A of -the NRC Enforcement Policv. Additionalltwo.iicensee
identified yiolations Whic Were'determ&ned to.b verJow
e 'f safety' sUgificandarelisd
report. If you contest any NCV in this report, you should provide a response within 30 days of
the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory
Commission, ATTN.: Document Control Desk, Washington DC 20555-0001; with copies to the
Regional Administrator Region II; the Director, Office of Enforcement, United States Nuclear
- Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at St.
Lucie Nuclear Plant. - - -
FP&L 2
In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter and its
enclosure, and your response (if any) will be available electronically for public inspection in the
NRC Public Document Room or from the Publicly Available Records (PARS) component of
NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.pov/readina-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
Charles R. Ogle, Chief
Engineering Branch 1
Division of Reactor Safety
Docket Nos. 50-335, 50-389
Enclosure: Inspection Report 50-335, 389/03-02
w/Attachment: Supplemental Information
cc w/encl: (See page 3)
FP&L 3
cc:
Senior Resident Inspector
St. Lucie Plant Mr. Don Mothena
U.S. Nuclear Regulatory Commission - Manager, Nuclear Plant Support Services
P.O. Box 6090 Florida Power & Light Company
Jensen Beach, Florida 34957 P.O. Box 14000
Juno Beach, FL 33408-0420
Craig Fugate, Director
Division of Emergency Preparedness Mr. Rajiv S. Kundalkar
Department of Community Affairs Vice President - Nuclear Engineering
2740 Centerview Drive Florida Power & Light Company
Tallahassee, Florida 32399-2100 P.O. Box 14000
Juno Beach, FL 33408-0420
M. S. Ross, Attorney
Florida Power & Light Company Mr. J. Kammel
P.O. Box 14000 Radiological Emergency
Juno Beach, FL 33408-0420 Planning Administrator
Department of Public Safety
Mr. Douglas Anderson 6000 SE. Tower Drive
County Administrator Stuart, Florida 34997
St. Lucie County
2300 Virginia Avenue Attorney General
Fort Pierce, Florida 34982 Department of Legal Affairs
The Capitol
Mr. William A. Passetti, Chief Tallahassee, Florida 32304
Department of Health
Bureau of Radiation Control Mr. Steve Hale
2020 Capital Circle, SE, Bin #C21 St. Lucie Nuclear Plant
Tallahassee, Florida 32399-1741 Florida Power and Light Company
- 6351 South Ocean Drive
Mr. Donald E. Jernigan, Site Vice President Jensen Beach, Florida 34957-2000
St. Lucie Nuclear Plant
6501 South Ocean Drive -Mr. Alan P. Nelson
Jensen Beach, Florida 34957 Nuclear Energy Institute
.1776 I Street, N.W.,'Suite 400
Mr. R. E. Rose Washington, DC 20006-3708
Plant General Manager APN@NEI.ORG
St. Lucie Nuclear Plant
6501 South Ocean Drive David Lewis - , -
Jensen Beach, Florida 34957 Shaw Pittman, LLP
2300 N Street, N.W.
Mr. G. Madden Washington, D.C. 20037
Licensing Manager
St. Lucie Nuclear Plant Mr. Stan Smilan
6501 South Ocean Drive 5866 Bay Hill Cir.
Jensen Beach, Florida 34957 Lake Worth, FL 33463
-L
- F .11
li
U. S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos: 50-335, 50-389
Report No: 50-335/03-02, 50-389/03-02
Licensee: Florida Power and Light Company (FPL)
Facility: St. Lucie Nuclear Plant, Units 1 & 2
Location: 6351 South Ocean Drive
Jensen Beach, FL 34957
Dates: March 10-28, 2003
Inspectors: R. Deem, Consultant, Brookhaven National Laboratory
P. Fillion, Reactor Inspector
F. Jape, Senior Project Inspector
M. Thomas, Senior Reactor Inspector (Lead Inspector)
S. Walker, Reactor Inspector
G. Wiseman, Senior Reactor Inspector
Approved by: Charles R. Ogle, Chief
Engineering Branch 1
Division of Reactor Safety
SUMMARY OF FINDINGS
IR 05000335/2003-002, 05000389/2003-002; Florida Power and Light Company; 03/10 -
28/2003; St. Lucie Nuclear Plant, Units 1 and 2; Triennial Fire Protection.
The report covered a'two-week period of inspection by regional inspectors and a consultant.
Three Green non-cited violations (NCVs) and one unresolved item'with potential safety
significance greater than Green were identified. The significance of most findings is indicated
by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609,
"Significance Determination Process" (SDP). Findings for which the SDP does not apply may
be Green or be assigned a severity level after NRC management review. The NRC's program
for overseeing the safe'operation of commercial nuclear power reactors is described in NUREG 1649, "Reactor Oversight Process," Revision 3, dated July 2000.
A. NRC-Identified and Self-Revealing Findings
Cornerstone: Initiating Events
TBD. The team identified a violation of 10 CFR 50.48 and the St. Lucie Nuclear
Plant (PSL) Unit 2 Operating License Condition (OLC) 2.C.(20), Fire Protection.
The fire hazards analysis'(FHA) failed to consider and evaluate the combustibility
of 380 gallons of transformer silicone dielectric insulating fluid in each of six
transformers (installed in three Unit 2 fire areas) as contributors to fire loading
and effects on safe shutdown (SSD) capability, as required by Fire Protection
Program (FPP) commitments.
This finding is greater than minor because it affected the objective of the initiating
events cornerstone to limit the likelihood of those events that could upset plant
stability and challenge critical safety functions relied upon for SSD during a fire.
The six previously unidentified silicone oil-filled transformers represented an
increase in the ignition frequency of the'associated fire areas/zones. This
finding is unresolved pending completion of a significance determination. Also,
when assessed with other findings identified in this report, the significance could
be greater than very low significance. (Section 1R05.02)
Cornerstone: Mitigating Systems
TBD. A violation of 10 CFR 50, Appendix R,'Section Ill.G.2, was Identified for
failure to ensure that one train-of equipment necessary to achieve and maintain
safe shutdown would be free of fire damage. Train A 480 volt (V)vital load
center 2A5 and associated electrical'cables were located in the Train B
switchgear room (fire area C)without adequate spatial separation or fire barriers.
This load center.powered redundant equipment (via motor control center 2A6
which powered boric acid makeup pumps 2A and 2B) required for SSD in the
event of a fire. In lieu of providingWadequate physical protection for load center
2A5 and the'associated electrical cables, manual operator actions outside the
main control'rooim (MCR) were relied on and credited, without prior NRC
approval, for achieving and maintaining SSD.
2
This finding was greater than minor because fire damage to the unprotected
cables could prevent operation of the equipment from the MCR and challenge
the operators' ability to maintain adequate reactor coolant system (RCS)
inventory and reactor coolant pump (RCP) seal flow for SSD during a fire in the
B switchgear room.
Green. A non-cited violation of 10 CFR 50, Appendix R, Section III.G.2 was
identified concerning a lack of spacial separation or barriers to protect cables
against fire damage in containment could result in spurious opening of the
pressurizer power operated relief valve (PORV).
This finding is greater than minor because it affected the mitigating system
cornerstone objective of equipment reliability, in that, spurious opening of the
PORV during post-fire safe shutdown would adversely affect systems intended to
maintain hot shutdown. The finding is of very low safety significance because
the initiating event likelihood was relatively low, manual fire suppression
capability remained unaffected and all mitigating systems except for the PORV
and block valve were unaffected. (Section 40A5)
B. Licensee-Identified Violations
One violation for which the significance has not been determined and two violations of
very low safety significance, which were identified by the licensee and entered in the
corrective action program, were reviewed by the inspection team. (Section 40A7)
- TBD. Many local manual operator actions were used in lieu of the required
physical protection of cables for equipment relied on for SSD during a fire,
without obtaining prior NRC approval for these deviations from the approved fire
protection program. This condition applied to numerous fire areas, including the
areas selected for this inspection. This reliance on large numbers of local
manual actions, in place of the required physical protection of cables, could
potentially result in an increased risk of loss of equipment that was relied upon
for SSD from a fire. (Se1ction1R05.05)
A violation of PSL Unit 2 (OLC) 2.C.(20) and the Fire Protection Program was
identified. However, this finding is unresolved pending completion of a
significance determination. The finding is greater than minor because it could
potentially result in an increased risk of loss of equipment that was relied upon
for SSD from a fire. (lecntion ta05t.aXed)
Other violations of very low safety significance, which were identified by the licensee,
have been reviewed by the team. Corrective actions taken or planned by the licensee
have been entered into the licensee's corrective action program. These violations and
corrective action tracking numbers are listed in Section 4A07.
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REPORT DETAILS
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems and Barrier Integrity
1R05 FIRE PROTECTION
01. Systems Required to Achieve and Maintain Post-Fire Safe Shutdown
a. Inspection Scone
The team evaluated the licensee's fire protection program against applicable
requirements, including Operating License Condition (OLC) 2.C.20, Fire Protection; Title
10 of the Code of Federal Regulations Part 50 (10 CFR 50), Appendix R; 10 CFR 50.48;
Appendix A to Branch Technical Position (BTP) Auxiliary Systems Branch (ASB) 9.5-1,
Guidelines for Fire Protection for Nuclear Power Plants; related NRC Safety Evaluation
Reports (SERs); the St. Lucie Updated Final Safety Analysis Report (UFSAR); and plant
Technical Specifications (TS). The team evaluated all areas of this inspection, as
documented below, against these requirements. The team reviewed the licensee's
Individual Plant Examination for External Events (IPEEE) and performed in-plant walk
downs to choose three risk-significant fire areas for detailed inspection and review. The
three fire areas selected were:
- Unit 2 Fire Area B - Cable Spreading Room (Fire Zone 52). A fire in this area
would involve alternate shutdown from outside the main control room (MCR).
- Unit 2 Fire Area C - Train B Switchgear Room (Fire Zone 34) and Electrical
Equipment Supply Fan Room (Fire Zone 48). Fire Area C and the essential
equipment and cables within were evaluated by the licensee with respect to the
protection and separation criteria of 10 CFR 50, Appendix R, Section III.G.2, to
assure that the ability to safely shut down the plant was not adversely effected
by a single fire event. Safe shut down of Unit 2 from the MCR using Train A
equipment was credited for a fire in this area.
- Unit 2 Fire Area I - Fire Zone 51 West (Cable Loft), Fire Zone 21 (Personnel
Rooms), Fire Zone 32 (PASS and Radiation Monitoring Room), Fire Zone
331 (Instrument Repair Shop), and Fire Zone 23 (Train B Electrical
Penetration Room). Fire Area I and the essential equipment and cables within
were evaluated by the licensee with respect to the protection and separation
criteria of 10 CFR 50, Appendix R Section III.G.2 to assure that the ability to
safely shut down the plant was not effected by a single fire event. Safe
shutdown from the MCR using Train A equipment was credited for a fire in this
area.
The team reviewed the licensee's fire protection program documented in the St. Lucie
UFSAR (Appendix 9.5A, Fire Protection Program Report); safe shutdown analysis
2
(SSA); fire hazards analysis (FHA); SSD essential equipment list; and system flow
diagrams to identify the components and systems necessary to achieve and maintain
safe shutdown conditions. The objective of this evaluation was to assure the safe
shutdown equipment and post-fire safe shutdown analytical approach were consistent
and satisfied the Appendix R reactor performance criteria for safe shutdown. For each
of the selected fire areas, the team focused on the fire protection features, and on the
systems and equipment necessary for the licensee to achieve and maintain safe
shutdown conditions in the event of a fire in those fire areas. Systems and/or
components selected for review included the pressurizer PORVs; boric acid makeup
pumps 2A and 2B and gravity feed valves V-2508, V-2509; auxiliary feedwater'(AFW);
charging pumps and volume control tank discharge valve V-2501; shutdown cooling;
heating, ventilation, and air conditioning (HVAC); atmospheric dump valves (ADVs); and
component cooling water. This review also included verifying that manual valves
operated during post fire safe shutdown were included in the licensee's maintenance
program.
b. Findings
No findings of significance were identified.
.02 Fire Protection of Safe Shutdown Capabilitv
a. Inspection Scope
For the selected fire areas, the team evaluated the frequency of fires or the potential for
fires, the combustible fire load characteristics and potential fire severity, the 'separation
of systems necessary to achieve SSD, and the separation of electrical components and
circuits' located within the same fire area to ensure that at least one train of redundant
safe shutdown systems was free of fire damrage. The team also inspected the fire
protection features to confirm they were installed in accordance with the codes of record
to satisfy the applicable separation and design requirements of 10 CFR 50, Appendix R,
Section III.G, and Appendix A of BTP ASB 9.5-1. The team reviewed the following
documents which establish the controls and practices to prevent'fires and to control
combustible fire loads and ignition sources to verify that the objectives established by
the NRC-approved fire protection piogram (FPP) were satisfied:
- Updated Final Safety Analysis Report (UFSAR), Appendix 9.5A, Fire Protection
Program Report
- Plant St. Lucie (PSL) Individual Plant Examination of External Events (IPEEE)
- Administrative Procedure 1800022, Fire Protection Plan
- Administrative Procedure 0010434, Plant Fire Protection Guidelines
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Electrical Maintenance Procedure 52.01, Periodic Maintenance of 4160 Volt
Switchgear
The team toured the selected plant fire areas to'observe whether the licensee had
properly evaluated in-situ compartment fire loads and limited transient fire hazards in a
manner consistent with the fire prevention and combustible hazards control procedures.
In addition, the team reviewed fire protection inspection reports, and corrective action
program condition reports (CRs) resulting from fire, smoke, sparks, arcing, and
equipment overheating incidents for the years 2001-2002 to assess the effectiveness of
the fire prevention program and to identify any maintenance or material condition
problems related to fire incidents.
The team reviewed. the fire brigade response procedures, training procedures, and drill
program procedures. The team reviewed fire brigade initial training and continuing
training course materials to verify appropriate training was being conducted for the
station firefighting personnel. In addition, the team evaluated fire brigade drill training
records for the operating shifts from August 2001- February 2003. The reviews were
performed to determine whether fire brigade drills had been conducted in high fire risk
plant areas and whether fire brigade personnel qualifications, drill response, and
performance met the requirements of the licensee's approved fire protection program.
The team walked down the fire brigade staging and dress-out areas in the turbine
buildings and fire brigade house to assess the condition of fire fighting and smoke
control equipment. The team examined the fire brigade's personal protective
equipment, self-contained breathing apparatuses (SCBAs), portable communications
equipment, and various other fire brigade equipment to determine accessibility, material
condition and operational readiness of equipment. Also, the availability of supplemental
fire brigade SCBA breathing air tanks, and the capability for refill, was evaluated.
Additionally, the team observed whether emergency exit lighting was provided for
personnel evacuation pathways to the outside exits as identified in the National Fire
Protection Association (NFPA) 101', Life Safety Code andOccupational Safety and
Health Administration (OSHA) Part 1910, Occupational Safety and Health Standards.
This review also included an examination of backup emergency lighting availability on
pathways to and within the dress-out and staging areas to support fire brigade
operations during a fire-induced power failure. The fire brigade self-contained breathing
apparatuses were examined and assessed for adequacy.
Team members walked down the selected fire areas to compare the associated fire
fighting pre-fire strategies and drawings with as-built plant conditions. This was done to
verify that fire fighting pre-fire strategies and drawings were consistent with the fire
protection features and potential fire conditions described in the UFSAR Fire Protection
Program Report. Also, the team performed a review of drawings and engineering
calculations for fire suppression caused flooding associated with the floor and
equipment drain systems for the Train B Switchgear Room, Electrical Equipment Supply
Fan Room, and Train B Electrical Penetration Room. The review focused on
'4
ensuring that those actions required for SSD would not be inhibited by fire suppression
activities or leakage from fire suppression'systems.
The team reviewed design control procedures to verify that plant changes were
adequately reviewed for the potential impact on the'fire protection program, SSD
equipment, and procedures as required by PSL Unit 2 Operating License Condition
2.C(20). Additionally, the team performed an independent technical review of the
licensee's plant change documentation completed in support of 2002 temporary
modification, TSA 2-02-006-3, that placed two exhaust fans on a fire damper opening
between the cable spreading room and the Train B switchgear room. This TSA was
evaluated in order to verify that modifications to the plant were performed consistent
with plant design control procedures.
b. - Findings
Inadequate Fire Hazards Analysis
Introduction: The team identified a Green non-cited violation (NCV) associated with
failure to meet the fire protection program plan requirements. The team found that six
silicone oil filled transformers installed in three Unit 2 fire zones [Fire Zone 37, Train A
Switchgear Room; Fire Zone 34, Train B Switchgear Room; and Fire Zone 47, Turbine
Building Switchgear Room] were not'evaluated in the Fire Hazards Analysis (FHA) as
contributors'to fire loading and effects on SSD capability as required by fire protection
program commitments.
Description: 'At PSL, the indoor medium voltage power transformers installed in Unit 1
were of the dry type. However, six of the indoor medium voltage power transformers in
Unit 2 were cooled and insulated by a silicone-type fluid. The licensee provided the
team with information from the transformer vendor which indicated that the transformer
insulating fluid was Dow Coming (DC) 561, a dimethyl silicone insulating fluid. The
team performed an independent technical review of the licensee's engineering
calculations and maintenance documen6tation, transformer vendor technical information
manual, insulating fluid'manufacturer-information, Underwriters Laboratory (UL) and
Factory Mutual (FM) listing agencies' documentation, and Institute of Electrical and
Electronics Engineers (IEEE) Standards.
The DC 561 technical manual described the DC 561 fluid as a silicone liquid that will
bum, but was less flammable than' paraffin-type insulating oils. The technical manual
also stated that the DC 561 fluid had a flash'point of 324 oC, a total heat release rate
(HRR) of 140 kw/m 2 (per ASTM E 1354-90), and a fire point of 357 "C. In their Fire
Hazard Analysis the licensee evaluated the adequacy of their fire'area/zone and
electrical raceway fire barrier system (ERFBS) enclosure barrier features based on the
combustible hazard content and overall fire loading (analyzed fire duration) present
within the associated area/zone. Based on the above, the team concluded that the
transformer insulating fluid was a in-situ combustible.liquid not accounted for nor
evaluated in'the PSL FHA. Additionally, the team noted that the licensee had conducted
5
an UFSAR Combustible Loading Update, evaluation in 1997.. This evaluation was
documented in PSL-ENG-SEMS-97-070, but failed to identify that the transformers in
fire zone 37 contained combustible silicone insulating fluid. Also a PSL Triennial Fire
Protection Audit (documented in QA audit Report QSL-FP-01-07) conducted in 2001,
reviewed the FHA but did not identify any fire loading discrepancies.
The team determined that the previously unidentified six silicone oil-filled transformers
represented an increase in the ignition frequency of the associated fire areas/zones.
Also, the additional in-situ combustible fire load and fire severity represented by the
combustible transformer insulating fluid increased the likelihood of a sustained fire event
from a catastrophic failure of an effected transformer that may upset plant stability and
challenge critical safety functions during SSD operations.
The l-T-E Unit Substation Transformers Instruction Manual recommended that the
dielectric insulating fluid be sampled annually and the dielectric strength of the fluid be.
tested to ensure that it is at 26 KV or better. The licensee determined that except for
four tests conducted during the period 1990-1992, there were no records of the
transformers' fluid being sampled and tested. This issue was entered into the corrective
action program as CR 2003-0978 and will followed up by the NRC resident inspectors at
PSL.
Analysis: The team determined that this finding was associated with the "protection
against external factors" attribute and affected the objective of the initiating events
cornerstone to limit the likelihood of those events that could upset plant stability and
challenge critical safety functions relied upon for SSD from a fire, and is therefore
greater than minor. The six previously unidentified silicone oil-filled transformers in Unit
2 represented an increase in the ignition frequency of the associated fire areas/zones.
The finding was considered to have very low safety significance (Green) because it did
not involve the impairment or degradation of NRC approved fire protection features and
the overall SSD capabilities for the areas were evaluated by the licensee's SSA as
adequate to ensure SSD capability. However, when assessed in combination with other
findings identified in this report, the significance could be greater than very low
significance.
Enforcement: 10 CFR 50.48 states, in part, "Each operating nuclear power plant must
-have a fire protection program that satisfies Criterion 3 of Appendix A to this part." PSL
Unit 2 Operating License NPF'16, Condition 2.C.(4) specifies, in part, that the licensee
implement and maintain in effect all provisions of the approved FPP as described in the
UFSAR for the facility and as approved by the NRC letter dated July 17, 1984, and
subsequent supplements. The approved FPP is maintained and documented in the
PSL UFSAR, Appendix 9.5A, Fire Protection Program Report.
The Fire Protection Program Report stated, in part, that the PSL fire protection program
implements.the philosophy of defense-in-depth protection against fire hazards and
effects of fire on safe shutdown equipment. The PSL fire protection program is guided
by plant fire hazard analyses and by credible fire postulations. It further stated that the
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FHA performed for PSL Unit 2 considered potential fire hazards and their possible effect
on safe shutdown capability.
PSL administrative fire protection procedure, 1800022, Section 8.3 states that the FHA
is an individual study of each plant's design,-potential fire hazards in the plant, potential
of those threats occurring, and the effect of postulated fires on safe shutdown capability.
Further, Section 8.7.1.A of this procedure stated that in-situ combustible features were
evaluated in the FHA as contributors to fire loading in the respective fire zones.
Contrary to the above, the FHA for fire zones 34, 37, and 47 was not adequate and did
not meet FPP commitments. Specifically, 380 gallons of in-situ combustible transformer
silicone dielectric insulating fluid in each of six transformers located in Unit 2 was not
considered nor evaluated in the FHA as contributors to fire loading and possible effects
on SSD capability. This condition was contrary to the requirements of the PSL FPP as
outlined in UFSAR, Section 9.5A, and therefore did not meet the requirements as set
forth in 10 CFR 50.48 and PSL OLC 2.C.(20).
Because the failure to evaluate in-situ combustible transformer silicone dielectric
insulating fluid as a contributor to fire loading in the FHA is of very low safety
significance and has been entered into the corrective action program as CR 2003-0637,
this violation is being treated as an NCV in accordance with Section VI.A.1 of the NRC's
Enforcement Policy. This item is identified as NCV 50-389103-02-OX, Failure to'
Evaluate In-situ Combustible Transformer Dielectric Insulating Fluid as a
Contributor to Fire Loading in the FHA.
.03 Post-Fire Safe Shutdown Circuit Analysis
a. Inspection ScoDe
The team reviewed how systems would be used to achieve inventory control, reactor
coolant pump seal protection, core'heat removal and reactor coolant system (RCS)
pressure control during and following a postulated fire in the fire areas selected for
review. Portions of the licensee's Appendix R Safe Shutdown Analysis Report which
outlined equipment and components in th'e chosen fire areas, power sources, and their
respective cable functions and system flow diagrams were reviewed.' Control circuit
schematics were analyzed to identify and evaluate cables important to safe shutdown.
The team traced the routing of cables through fire areas selected for review by using
cable schedule, and conduit and tray drawings. The team walked down these fire areas
to compare the actual plant configuration to the layout indicated on'the'drawings. The
team evaluated the above information to determine if the requirements for protection of
control and power cables were met. The licensee's circuit breaker and fuse coordination
study was reviewed for adequate electrical scheme protection of equipment necessary
for safe shutdown.' The following equipment arid components were reviewed during the
inspection:'
- V1474 and V1475, Pressurizer PORVs
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- V1476 and V1477, Pressurizer Isolation Block Valves
- MV-09-03 and MV-09-04, Feedwater Bypass Valves
- 2HVE-1 3B, Control Room Booster Fan
- V2501, VCT Discharge Outlet Valve
- MV-07 -04, Containment Spray Isolation Valve
- LP-208, Lighting Panel 208
- LP-209, Lighting Panel 209
- HCV-3625, Safety Injection Block Valve
- V3444, Shutdown Cooling Block Valve
- P1-1107/1108, Pressurizer Pressure for Hot Shutdown Panel
- LI-1 104/1105, Pressurizer Level for Hot Shutdown Panel
- LI-9113/ 9123, Steam Generator Level for Hot Shutdown Panel
- SIAS Logic
MCC 2A5/2A6 and relative feeds, 480 Volt Motor Control Center
MCC 2B5/2B6 and relative feeds, 480 Volt Motor Control Center
- Load Center 2A5 480 Volt Switchgear
b. Findings
No findings of significance were identified.
04. Alternative Post-Fire Safe Shutdown Capability
a. Insgection Scope
The cable spreading room, which was one of two alternate shutdown (ASD) fire areas
listed in the St. Lucie SSA for Unit 2, was selected for detailed inspection of post-fire
SSD capability. Emphasis was placed on verification that hot and cold shutdown from
outside the control room could be implemented; and that transfer of control from the
main control room to the hot shutdown control panel (HSCP) and other equipment.
isolation locations could be accomplished within the performance goals stated in 10
CFR 50, Appendix R,Section III.L.3.
Electrical diagrams of power, control, and instrumentation cables required for ASD were
analyzed for fire induced faults that could defeat operation from the MCR or the HSCP.
The team reviewed the electrical isolation and protective fusing in the transfer circuits of
components (e.g., motor operated valves) required for post-fire SSD at the HSCP to
verify that the SSD components were physically and electrically separated from the fire
area. The team also examined the electrical circuits for a sampling of components
operable at the HSCP to ensure that a fire in the B Switchgear Room would not
adversely affect safe shutdown capability from the MCR. The team's review was
performed to verify that adequate isolation capability of equipment used for safe
shutdown implementation was in place, accessible, and that the hot shutdown control
panel was capable of controlling all the required equipment necessary to bring the unit
to a safe shutdown condition. This also included a review to verify that the shutdown
process met the performance goals of 10 CFR 50,Appendix R, Section lll.L.3 and
8
guidance in generic letter (GL) 86-10, by comparing it to the thermal hydraulic time line
analysis provided by the licensee.
b. Findings
No findings of significance were identified.
05. Operational Implementation of Post-Fire Safe Shutdown Canability
a. InsDection Scope
'The team reviewed off normal operating procedure 2-ONP-100.02, Control Room
Inaccessibility, Rev. 13B, the licensee's procedure for alternate safe shutdown, and
procedure 2-ONP-1 00.01, Response to Fire, Rev. 9, the licensee's operating procedure
for post-fire safe shutdown from the' MCR. The review focused on ensuring that all
required functions'for post-fire safe shutdown and the corresponding equipment
necessary to perform those functions were included in the procedures. The review also
examined the consistency between the operations shutdown procedures and other
procedure driven activities associated with post-fire safe shutdown (i.e., fire fighting
activities).
b. Findings -
The team noted that the licensee had identified that manual operator actions outside the
MCR were credited and used in lieu of physical protection of cables and equipment
relied on for SSD during a fire without obtaining prior NRC approval. Use of manual
operator actions outside the MCR for 10 CFR 50, Appendix R.Section III.G.2 areas
(Fire Area C and Fire Area I for this inspection) without prior NRC approval was not in
accordance with the licensee's approved Fire Protection Program. The licensee
identified this issue in CR 03-0153 prior to this inspection. This finding is More Than
Minor. This finding will be Unresolved pending completion of the SDP to determine the
risk associated with using manual operator'actions in lieu physical protection.' 10 CFR
50, Appendix R,Section II.G specified the need to identify equipment to achieve and
maintain safe shutdown functions; and the protection requirements for that equipment.
It also stated that one train of safe shutdown equipment should remain free of fire
damage for non-alternate shutdown' (l1l.G.2) designated fire areas. Two of the three fire
areas inspected were so designated. In these areas,. manual operator actions outside
the MCR were being used and credited in the SSA to achieve safe shutdown.
Determination of the licensinig basis and required NRC exemption to use manual
operations in lieu of protection for one shutdown train was addressed by another
inspection team member 'The in'spection team was also concerned whether all potential
spurious operations were properly accounted for in the shutdown procedures.
Subsequent review of the licensee's procedures for these-areas did demonstrate that
manual actions required to mitigate spurious signals on both units were properly
dispositioned. . -
I
9
06. Communications
a. Inspection Scooe
The team reviewed plant communications to verify that adequate communications were
available to support unit shutdown and fire brigade duties. This included verifying that
site paging (PA), portable radios, and sound-powered phone systems were available
consistent with the licensing basis. The team reviewed the licensee's communications
features to assess whether they were properly evaluated in the licensee's SSA
(protected from exposure fire damage) and properly integrated into the post-fire SSD
procedures. The team also walked down sections of the post-fire SSD procedures to
verify that adequate communications equipment would be available to support the SSD
process. The team also reviewed the periodic testing of the site fire alarm and PA
systems; maintenance checklists for the sound-powered phone circuits and amplifiers;
and inventory surveillance of post-fire SSD operator equipment to assess whether the
maintenance/surveillance test program for the communications systems was sufficient
to verify proper operation of the systems.
b. Findings
No findings of significance were identified.
a. Inspection Scope
The team reviewed licensee emergency lighting against the requirements of 10 CFR 50,
Appendix R, Section lll.J, to verify that eight hour emergency lighting coverage was
provided in areas where manual operator actions were required during post-fire safe
shutdown operations, including the ingress and egress routes. The team's review also
included verifying that emergency lighting requirements were evaluated in the licensee's
SSA and properly integrated into the Appendix R safe shutdown procedures as
described in UFSAR Appendix 9.5A, Section 3.7. During plant walk downs of selected
areas where operators performed local manual actions defined in the post-fire SSD
procedures, the team inspected area emergency lighting units (ELUs) for.operability and
checked the aiming of lamp heads to determine if adequate illumination was available to
correctly and safely perform the actions required by the procedures. The team also
inspected emergency lighting features along access and egress pathways used during
SSD activities for adequacy and personnel safety. The team checked the ELUs' battery
power supplies to verify that they were rated with at least an 8-hour capacity. In
addition, the team reviewed the manufacturer's information and the licensee's periodic
maintenance tests to verify that the ELUs were being maintained and tested in
accordance with the manufacturer's recommendations.
.. b. Findings
10
No findings of significance were identified.
08. Cold Shutdown Repairs
a. Insgection Scone
The team reviewed the licensee's SSA and existing plant procedures to determine if any
repairs were necessary to achieve cold shutdown, and if needed, the equipment and
procedures required to implement those'repairs was available onsite.
b. Findings
No findings of significance were identified.'
.09 Fire Barriers and Fire Area/Zone/Room Penetration Seals
a. 'Inspection Scope'
The team walked down the selected fire zones/areas to evaluate the adequacy of the
fire resistance of barrier'enclosure walls, ceilings, floors,'and cable protection. The
team randomly selected several fire barrier features for detailed evaluation and
inspection to verify proper installation and qualification. This evaluation included fire
barrier penetration fire stop seals, fire doors, fire dampers, fire barrier partitions, and
Thermo-Lag electrical raceway fire barrier system (ERFBS) enclosures to ensure that at
least one train of SSD equipment would be maintained free of fire damage from a single
fire.
The team observed the material condition and configuration of the selected fire barrier
features and also reviewed construction details and supporting fire endurance tests for
the installed fire barrier features.r' This review was performed to compared the observed
fire barrier penetration seal and ERFBS configurations to the design drawings and
tested configurations. The team also compared the penetration seal and ERFBS ratings
with the ratings of the barriers in-which they were installed.
The team reviewed licensing documentation, engineering evaluations of-Generic Letter
86-10 fire barrier features, and NFPA code deviations to verify that the fire barrier
installations met design requirements and license commitments. In addition, the team'
reviewed surveillance and maintenance'procedures for selected fire barrier features to
verify the fire barriers were being adequately maintained.
b. Findings
No findings of significance were identified. - -
.10 Fire Protection Systems. Features, and Equipment -
11
a. Inspection Scone
The team reviewed flow diagrams, electrical schematic diagrams, periodic test
procedures, engineering technical evaluations for NFPA code deviations, operational
valve lineup procedures, and cable routing data for the power and control circuits of the
electric motor-driven fire pumps and the fire protection water supply system yard mains.
The review was performed to assess whether the common fire protection water delivery
and supply components could be damaged or inhibited by fire-induced failures of
electrical power supplies or control circuits and subsequent possible loss of fire water
supply to the plant. Additionally, team members walked down the fire protection water
supply system piping and actuation valves for the selected fire areas to assess the
adequacy of the system material condition, consistency of the as-built configuration with
engineering drawings, and operability of the system in accordance with applicable
administrative procedures and NFPA standards.
The team walked down accessible portions of the fire detection and alarm systems in
the selected fire areas to evaluate the engineering design and operation of the installed
configurations. The team also reviewed engineering drawings for fire detector spacing
and locations in the four selected fire areas for consistency with the licensee's fire
protection plan, engineering evaluations for NFPA code deviations, and the
requirements in NFPA 72A and 72D.
The team also walked down the selected fire zones/areas with automatic sprinkler
suppression systems installed to verify the proper type, placement and spacing of the
heads/nozzles and the lack of obstructions. The team examined vendor information,
engineering evaluations for NFPA code deviations, and design calculations to verify that
the required suppression system density for each protected area was available.
The team reviewed the manual suppression standpipe and fire hose system to verify the
adequacy of their design, installation, and operation for the selected fire areas. The
team examined design flow calculations and evaluations to verify that the required fire
hose water flow and sprinkler system density for each protected area were available.
The team checked a sample of manual fire-hose lengths to determine whether they
would reach the SSD equipment. Additionally, the team observed placement of the fire
- hoses and extinguishers to assess consistency with the fire fightingpre- Ian drawings.
b. Findings
No findings of significance were identified.
4. Other Activities
40A2 Problem Identification and Resolution
a. Inspection Scone
12 . [ .
The team reviewed a sample of licensee audits,'self-assessmeihts, and plant condition
reports (CRs) to verify that items related to fire protection and safe shutdown were
appropriately entered into the licensee's corrective action program in accordance with
the licensee's quality assurance program arid procedural requirements. The items
selected'were also reviewed for classification and appropriateness of the corrective '
actions taken or initiated to resolve the Items. - ,
The team reviewed the licensee's applicability evaluations'and corrective actions for';
selected industry experience issues related to fire protection. The operating experience-
reports were reviewed to verify that the licensee's review and actions were appropriate.
The reports are listed in the List of Documents Reviewed Section.
b. Findings
No findings of significance were identified
40A3 Everit Followup ' '
.1 (Closed) LER 50-335. 389/00-01, Outside Design Bases Appendix R Hi-Lo Pressure
Interface and Separation Issues. ' -
.~~~~~~~~~~~~~~~~~~~
~ ":,^'t.v ..:
On March 9, 2000, the licensee identified seven'cases where the' plan't was not in'
compliance with 10 CFR 50, Appendix R, Sections IlI.'G.2.d and III.G.2. f. The first-'
case, involving the pressurizer PORVs, applied to Units 1 and 2, and is-discussed in ' - ..
Section 4AO5 of this report. The 6ther'six cases apply to Unit 2 only, and are discussed
as follows.
Shutdown cooling valves
'Shutdown cooling valves V3652 and V3481 could spuriously open due to fire induced
cable-to-cable short circuits. The location 'of vulnerability was a pull box (JB-2031) in the
annulus region of containment. The valves are motor operated type valves which'are '-
de-energized by procedure'during normal plant operation. The problem however Is that
the power cables for both these-valves were'routed through a pull box together'with
other three-phase power cables. Therefore,'the potential existed for fire induced cable'
to cable short circuiting which 'could inadvertently energize the motors to open these' '
valves. Both valves-would have to open to have a problem. Opening of these valves
directly connects the RCS to piping that is not rated for RCS normal operating pressure.
Should the valves open when the RCS is at operating-'pressure, a pressure relief valve
would open 'and RCS coolant would flow from the RCS to the containment sump. This '
situation is essentially a large break LOCA. Valve V3545 is a normally open motor - ' *
operated valve in series with V3652 and V3481.' Theoretically, V3545 could be closed
by the operator to stop the outflow, but the cables for V3545 'could have been damaged ' '
by the same fire. The licensee resolved the'problemrn by installing new power cables -
using armored cable. This precluded the possibility-of cable to cable short circuits. -
I.
13
Inspectors confirmed implementation of the modification through review of plant
modification PCMIO1028.
The reported condition was a violation of Appendix R requirements of more than minor
significance because it could adversely affect the equipment reliability objective of the
cornerstones of mitigating systems and barrier integrity as described above. Using
techniques described in NRC Procedure 0609, Appendix F, the inspectors determined
that the finding was of very low safety significance (Green). Specifically the SDP
worksheet for large break LOCA was evaluated. The conclusion was supported
primarily by the negligible probability of the initiating event occurring and the fact that
cables for mitigating systems for LOCA are located outside containment. The
enforcement considerations for this violation are given in Section 40A7.
Pressurizer pressure instrumentation affected by tray-conduit interaction
Lack of 20-foot separation or a radiant heat shield between a cable tray and two
conduits in containment meant that a fire which could start in the cable tray due to cable
self ignition could result in damage to a number of pressurizer pressure instrumentation
loops. PT-1 105, PT-1 106 and PT-1 107 are in cable tray L2224; and PT-1 103, PT-1 104
and PT-1108 are in conduits 25018Y and 23091A. PT-1107 and PT-1108 were the
instruments specified in the post-fire shutdown procedure. These instruments also
provide input to alarms, automatically initiate automatic actions, provide permissives,
computer inputs, input to calculations and indications of pressure at various locations.
The inspector reviewed the consequences and ramifications of instruments failing either
high or low. Also reviewed, was which pressurizer pressure instrumentations remain
unaffected by the fire. This information was analyzed by the inspector, and it was
concluded that the affected instrumentation would not lead to any transient nor to
change in core damage frequency. The finding is therefore of very low safety
significance. As corrective action, conduits 25018Y and 23091A were protected by a
radiant heat shield for twenty feet either side of the tray L2224 by plant modification
PCM99104, Supplement 1. The licensee reports the fact that both channels of
pressurizer pressure instruments specified in the post-fire shutdown procedure could
have been affected by one fire represents a violation of 10 CFR 50, Appendix R, Section
ll, G, 2. Refer to Section 40A7 of this report for enforcement aspects.
Pressurizer level instrumentation affected by tray-conduit interaction
Lack of 20-foot separation or a radiant heat shield between a cable tray and two
conduits in containment meant that a fire which could start in the cable tray due to cable
self ignition could result in damage to all pressurizer level instrumentation loops. LT-
i11OX and LT-1105 are in tray L2213; and LT-I1 OY and LT-1104 are in conduits
23320D and 23090A. LT-1 I1 OX & Y were specified in the post-fire shutdown
procedure. - It was determined that the failure mode for a short-circuit between the.
twisted pair or open circuit caused by fire exposure of the signal wires was level fails
low. Level failing low initiates several automatic actions some of which tend to cause
level to rise and some of which cause level to fall. The de-energization of pressurizer
I'
14
heaters dominates the situation and results in falling level. This leads to a reactor trip
with safety injection on low pressurizer pressure. When the safety injection pumps start,
the level will rise. Since the operator cannot .see level, he may not turn off the safety.
injection pumps. So it follows that the pressurizer will go solid. The post-fire safe
shutdown procedure directs the operator to place the PORVs in override due to r.
concerns about spurious opening. Therefore, rising level and concomitant pressure rise "c,'. ...
would be relieved by the safety relief valves. To obtain the risk significance of the fire
induced failure of pressurizer level instrumentation, the SDP worksheet for stuck open
relief valve was evaluated. The results indicated the finding was of very low safety
significance (Green) for the same reasons mentioned in Section 4A05.1 which deals
with spurious opening of PORVs- The licensee 'reports the fact that both channels of' r .. "".
pressurizer level instruments specified'in the post-fire shutdown procedure could have
been affected by one fire represents a violation of 10 CFR 50, Appendix R, Section 1II,
G, 2. Refer to Section 40A7 of this report for enforcement aspects.
Pressurizer level instrumentation affected by conduit to conduit interaction
I'., '*'
Lack of 20-foot separation or a radiant heat shield between two conduits in containment
containing cables for redundant channels of pressurizer level instrumentation meant that
the separation requirements of Appendix R were not met. The location of the Interaction .
is in the annulus area at an elevation' where there are no ignition sources other than the
cables themselves. It is not considered credible that low voltage,; low energy,
.,A.
instrumentation circuits could self-induce' cable ignition, and even if such occurred within
a conduit, the fire' would not affect another conduit. The reported problem was a
violation of Appendix R requirements with regard to s'eiarationrof cables. 'The
inspectors determined that, given the particular configuration at issue, it could not
credibly adversely affect'any cornerstone. 'The licensee corrected the separation
problem by installing a radiant heat shield on'one of the conduits per plant modification ' .,'.',
PCM99104, Supplement 1 This licensee identified issue'constitutes a violation of minor
significance that is not subject to enforcement action in accordance with Section IV of
the NRC's Enforcement Policy.
Circuits related to automatic pressurizer pressure control affected by conduit to conduit
interaction '
Lack of separation or a radiant heat shield between ceitain conduits in containment
related to automatic pressurizer pressure control meant that the separation
requirements of Appendix R were not met. The circuits involved were for the PORV
and the auxiliary spray isolation valves. The concern was that, if one fire could affect
both these circuits, two diverse subsystems designed to reduce pressure when
necessary may not function. There are other ways to reduce pressure, but the above .
f.
mentioned ones were the systems'designated in the'post-fire shutdown procedure for
this function. The location of the interaction is in the annulus area at an elevation where
'there are no ignition sources other than the cables themselves. It is not'considered
credible that a fire starting within one conduit would expand to affect other nearby
conduits. The reported problem was a violationhof Appendix R requirements with regard
15
to separation of cables. The inspectors determined that, given the particular
configuration at issue, it could not credibly adversely affect any cornerstone. The
licensee corrected the separation problem by installing a radiant heat shield on a
sufficient number of the conduits per plant modification PCM99104, Supplement 2. This
licensee identified issue constitutes a violation of minor significance that is not subject to
enforcement action in accordance with Section IV of the NRC's Enforcement Policy.
Radiant heat shields not installed per Apoendix R accegted deviation
Inside containment in the area between the containment wall and the bioshield four
groups of cable trays are installed. There are five trays in each group. These trays run
horizontally along the circumference of the containment to carry cables from the
penetration area to their various ultimate destinations in the containment. Train B
cables are in trays near the containment wall, and Train A cables are in trays near the
bioshield. There is at least seven foot horizontal separation between these two sets of
trays in the area of interest. Both the Train A set and the Train B set consists of a group
running above the 45-foot elevation grating and a group running above the 23-foot
elevation grating. Examples of cable trays involved are instrumentation trays L2223
(Train A) and L2224 (Train B); or control trays C2223 (Train A) and C2224 (Train B).
According to the safety evaluation report each of the four groups should have had a
radiant heat shield installed directly below the group. This is actually an accepted
deviation, or exemption, from the requirement to have a heat shield between the
redundant cables. The licensee reported in the LER that the radiant heat shields below
the groups at the 45-foot elevation were not installed. The missing radiant heat shields
have now been installed per PCM01028.
The inspector evaluated the risk significance of the lack of radiant heat shield below the
45-foot elevation groups of trays. The conclusion of this evaluation was that the
problem was of very low safety significance (Green). Some of the dominant factors
considered were:
- Fire brigade capability for a fire in containment was not impaired.
- In-situ ignition sources were negligible, and transient ignition sources and
combustibles are not present during normal plant operation.
- Only the top tray in each group contains power cables (480 volt) carrying
sufficient energy capable of self ignition of IEEE 383 flame tested cable. Most of
the power cables in containment are not energized during normal plant.
operation. These trays are solid metallic bottom and cover type trays. This
construction inherently limits the spread of internal tray fire, and effectively
provides a shield limiting the radiant heat energy.
- The "target" cable trays have a minimum spatial separation of 15 feet vertical
and 7 feet horizontal from the potentially burning cable tray. The target trays
have solid metallic bottoms. Radiant energy flowing between source and target
16
is blocked to a great extent by intervening HVAC ducts, large pipes, tanks and
building steel. Hot gas layer is not a factor in the part of containment under
' consideration.'
- The target cables would be instrumentation cables, and various scenarios
involving'damage to these same-instrumentation cables discussed Inrelation to
other findings within this report Section were shown to be of very low safety
significance.
A very similar configuration in the'Unit 1 containment was analyzed by the
'licensee and reviewed by the NRC in great detail, and found to be an acceptable
configuration from the fire protection viewpoint. The Unit I study had a safety
factor of at least two, which provides margin to account for geometry and other
unknown differences between the two units.
Failure to adhere to the configuration of cable trays and radiant heat shields described
in an exception to 10 CFR 50, Appendix R,Section III.G.2 represents a licensee'
identified violation. Refer to Section 4AO7 of this report for enforcement aspects.
.2 (Closed) LER 50-335/00-04, Pressurizer Level Instrumentation Conduit Separation
Outside Appendix R Design Bases
Lack of 20-foot separation or a radiant heat shield between a cable tray and a conduit in
Unit 1 containment meant that a fire which could start in the cable tray due to cable self
-ignition could result in damage to all pressurizer level instrumentation. The discussion
of risk'significance and requirements for this issue would be identical to the discussion
of essentially the same issue on Unit 2 in Section .1 above under the heading:
Pressurizer level instrumentation affected by tray-conduit interaction. Refer to Section
4AO7 of this report for enforcement aspects.
40A5 Other Activities
.1 (Closed) URI 335.389/99-08-03. PORV Cabling May Not be- Protected from Hot-Shorts
Inside Containment
Introduction: A Green NCV was identified for failure to comply with 10 CFR 50,
Appendix R,Section III, G, 2.d and f, related to spurious opening of the pressurizer
PORV. ' -.
Decriptiori: During conduct of an inspection in the area of fire protection (NRC
Inspection Report 50-335, 389/99-08, dated January 31, 2000) the inspectors identified
the possibility that the PORV cables inside containment were not protected from fire
induced cable to cable short circuits-'.The'Issue was identified through review of the
licensee's analysis. However, the analysis referred to a study which showed that the
cable to cable short circuit-leading to spurious opening of the PORV was not credible.
Since the study could not be located at the time of the inspection, an unresolved item
17
was initiated to track this issue. Subsequently LER 50-335, 389/00-01 reported that the
pressurizer PORVs could open due to fire induced short circuits that could occur in a
cable tray in containment. In addition, cables for the associated block valve were routed
in the same cable tray. This meant the block valve may not be available to counter the
spurious opening of the PORV. Cables for one PORV and its block valve were in a tray
near the containment wall and cables for the other set were in a tray near the bioshield.
The condition applied to both units.
The licensee resolved the problem by installing new PORV cables using armored cable.
This precluded the possibility of cable to cable short circuits. The potential for spurious
opening due to spurious pressure signal had already been offset by having the operator
place the control switch in override in response to a fire in containment. Inspectors
confirmed the modification was implemented through review of plant modification
package PCM00059 (Unit 1) and PCM99104, Rev 4 (Unit 2).
LER 00-01 mentioned above also reported licensee identified findings in the area of
Appendix R. In addition, Unit 1 LER 00-04 reported similar problems. Refer to Section
40A3 for discussion of these findings.
Analysis: The finding was a performance deficiency because it represented a violation of
Appendix R requirements. It was considered greater than minor because it could
adversely affect the cornerstones of mitigating systems and barrier integrity. It affects
mitigating systems in the sense that systems designated for post-fire shutdown would
be adversely affected by an open PORV during the early stages of post-fire shutdown.
It affects the cornerstone of barrier integrity in the sense that a spuriously open PORV
represents a breach of the RCS pressure boundary which is one of the barriers. Using
techniques described in NRC Procedure 0609, Appendix F, the inspectors determined
that the finding was of very low safety significance (Green). Specifically, the SDP
worksheet for stuck open relief valve was evaluated. A key factor leading to this
conclusion was that the initiating event likelihood was relatively low. It was less likely
than the likelihood for stuck open PORV due to non-fire induced causes. Manual
suppression of fires in the containment was in the normal state because the plant had
fire detectors, a fire plan and there were no automatic valves in the water source that
could be affected by the fire. Even though no credit could be given for the block valve,
other mitigating systems were unaffected. This was primarily due to the fact that the
associated cables were all outside containment.
Enforcement: Because this violation of 10 CFR 50, Appendix R, Section 1II,G.2.d. and f,
is of very low safety significance, has been entered into the CAP (CROO-0386) and the
problem has been corrected through a plant modification it is being treated as an NCV,
consistent with Section VL.A of the NRC Enforcement Policy. The number and title of
this NCV are: NCV 50-335, 389/03-02-01, Failure to Meet 10 CFR 50, Appendix R,
Section 1II,G, 2, for Protection of the PORV Cables in Containment.
40A6 Meetings
18
On March 28, 2003, the team presented the inspection results to Mr. D. Jemigan and
other members of your staff, who acknowledged the findings. The team confirmed that
proprietary information is included in this report.
40A7 Licensee-identified Violations
The following findings of very low safety significance (Green) were identified by the
licensee and are violations of NRC requirements which meet the criteria of Section VI of
the NRC Enforcement Policy, NUREG-1600, for being dispositioned as NCVs.
- 10 CFR 50, Appendix R, Fire Protection Program,-Section l1l, Specific
Requirements, Subpart G, Fire protection of safe shutdown capability, requires
that for cables, that could prevent operation or cause maloperation due to hot
shorts, open circuits or shorts to ground, of redundant trains of systems
necessary to achieve and maintain hot shutdown conditions and located inside
noninerted containments, one of the following fire protection means shall be
provided:
1. Separation of cables of redundant trains by a horizontal distance of more
than 20-feet with no intervening combustibles or fire hazards; or
2. - Separation of cables of redundant trains by a non-combustible radiant
energy shield.
Contrary to this, since the requirement became effective, the required fire
protection was not provided for the following redundant cables:
1. Shutdown cooling valves V3652 and V3481 on Unit 2.
2. Pressurizer pressure instrumentation PT-1 107 and PT-1 108 on Unit 2
3. Pressurizer level instrumentation LT-111OX and LT-11iOY on Units 1 & 2
4; Cables
.~~otle
contained. in
in
cable
a
trays L2223 (Train A) and L2224 (Train B4
2 -_Tan B
These findings have been entered into the CAP (CR 99-1963, Rev. 2, and CR
00-0386), corrected by plant modifications, and are of very low safety*
significance for reasons given in Sections 4AO3.1 and .2.
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel : : -- - -
D. Albritton, Assistant Nuclear Plant Supervisor
19
P. Barnes, Fire Protection Engineering Supervisor
R. De La Esprella, Site Quality Manager
B. Dunn, Site Engineering Manager
K. Frehafer, Licensing Engineer
J. Hoffman, Design Engineering Manager
D. Jernigan, Site Vice President
G. Madden, Licensing Manager
R. Maier, Protection Services Manager
R. McDaniel, Fire Protection Supervisor
T. Patterson, Operations Manager
R. Rose, Plant General Manager
V. Rubano, Engineering Special Projects Manager
S. Short, Electrical Engineering Supervisor
NRC Personnel
C. Ogle, Branch Chief
R. Rodriguez, Nuclear Safety Intem (Trainee)
T. Ross, Senior Resident Inspector
S. Sanchez, Resident Inspector
List of Documnents and Drawings' Reviewedduring Inspection
2998-B-048.,Safe Shutdown-Analvsis Fire Area reDort:'
2998-B1049 .Essentia ulDment Listw ev 662/14/02
dated
Procedure 2-ONP-100.02; Control Room Inaccessibilitv". Rev.13B. dated 10/29/02.
Procedu'ire 2-ONP100.01 ,"Reso 6
nsd to Fire..Rev.9. ' 128/01-t
Elec6tirical
PS-'I FJM i-961-"""P..S-L'-I`RA'BPSL-i3
Equipment Et-~
-a0cR date-d-;1I2at
Rooms Ac Computer, model Data
Inputssand Outputs ,-Rev.1-, dated:10/5/92.; --- _ _
St Luciej;Unit2 Flow, Diagrams:
2998-G-078.SH1 a oumeControl System! Re l6.
2998-G-879.,SH 1&2.-'HVAC Flow'and Control Diararns, dated 10/20/89.
2998-G-079;' SH 1.'2'& 7.Main Steam Svstem. ,Rev.
2998-G-080. SH 2A'& 2B. .F~eedwat'rand Condens te Svstem.,,Rev 25.
2998-G-082, SH 1 & 2.-Circuiatinat and Intake Coolina WaterSystem,' Rev37
2998-G-083. SH '1& 2.'l-Coibonent Cooin' Water Svstem-. Rev'.28.
2998-'G-078. 'SH I07.-I108'.I09110.' Reictor Coolant Svstem.' Rev;1.
2998-G-0i8.' SH I 30A'I30B.1311 32.7SafetvIniection Svsteem;. Rev. 12.
2998-G-088, SH 1, 'Containmhent Spray and Refuelinrg Water System, Rev. 35.
1
.. ... !I..I ................. !
- ,
. -... - i-1:-"--
i -;:
. " I i-':'I
. -w I- -
-.- . 6::411
.. -., . 1 !.
- . . .. - - --.. . r - - .. .
l- '. -.- -
2
LIST OF DOCUMENTS REVIEWED
3
ATTACHMENT 2
LIST OF ACRONYMS USED
AMP Aging Management Program
AMR Aging Management Review
ASME American Society of Mechanical Engineers
CASS Cast Austenitic Stainless Steel
CCW Component Cooling Water
CR Condition Report
CST Condensate Storage Tank
EDG Emergency Diesel Generator
EQ Environmental Qualification Program
FAC Flow Accelerated Corrosion
FPL Florida Power and Light Company
GALL Generic Aging Lessons Learned report
ICW Intake Cooling Water System
ILRT Integrate Leak Rate Test
ISI Inservice Inspection
LRA License Renewal Application
LRAMR License Renewal Aging Management Review report
LRBD License Renewal Basis Document
NRR NRC Office of Nuclear Reactor Regulation
OE Operating Experience
PM Preventive Maintenance
PMAI Plant Management Action Ite m
RAB Reactor Auxiliary Building
RAI Request for Additional Information
RV Reactor Vessel
RVH Reactor Vessel Head
RVI Reactor Vessel Internals
SSC Systems, Structures, and Components
SSMP Systems and Structures Monitoring Program
TCW Turbine Cooling Water
UFSAR Updated Final Safety Analysis Report
4
FORST;AUCIEINSPECTION REORO3O~f__,6
INPU
.Z - -~P 'ul tFill-bl
dateF
WORDS FOR-COVERkLETTER
The reDort documents one.,NRC-identified findn of~verv.low safetv sioni ican re
waadete~rmined t4, invlve ' iltino NRC. reurmns'5 oee. easio te.vr'lx...
i~t sia the entered into vourcorrective action'irooram and
the-NRC isttrbnaits as a non-cited-violation (NCV) co'sisttwlvhSection IVb
corre
t~ie, NRC +Enf&6orceent -. Polacy6Ei>,y~t~di1.6 identified victior
4807ofq~,repoj~
~ i~: Tin
i-5 acc psrppou'esuri
snpco
&~~~PR dnldad ~Self-Reveasafe shutdown wudavrl fet'yt neddt
dam<'a'ae;inecofnI
nrannt'oultreultin, atspurious,-openirnnioftt fthsiep
ornerstone 'a allgatgthe msan ex sepm
capabberrhaini n ededhb
6LVic~prdit6te
PORV!.durinafdtV sidnificance.s ensee'-ldrenbfiedwhich were,
aiolap lv. aftecens6inte Ons
have
Violations ofvrvlw' been'tiabn1aenthikd
aev was, relabvelv low. byi fien'
6't ><!,6,49.§.J in<1; 4,sy-*tr-t?,-iS~w.,,,^,1o.?~i'-~
-;v--.s-..,..= fica'te-dwhich 'werdnife ¢vth-1ese~a s<!$ ,^-,;<7; .
=IST.._OFITEMS.OPEN
i PaR~P~~S ED',. CLSDAD - DISCUSSED
. ~_-
...,---.. _... i-
5.
ed
3891030--2L0
FE7-33f5;,-T NCV .a~i~6 Meet-IO;CFR:50.'A~Defd* :S7ci Wl
o ectio~n'bf-the ?ORV-.ablei
E
slosed
o35 : 89I9-O8- I O3 .
F035F,33819-08-03- RI 'ORV.6linma .Not-be'F.'rotected ,from .t-Shots
nsd]otlmn (Scin:4A
50 5, 389i00-01 iitside iDesianiBases 'A~nedixlR Hi-Lo:Pi'.essur
utid6-b--,
aes nSparation - ssues J F .
(Section 4A3)
_O-:_335OO4, Jrssu~zrizeeveI strrentation CondULinSearatio32
Jusd4l~mnxRDsg Bssletonio3
L.IST
.. I~OF..DOUMEM,_SRE.'./EWEb
s.....e
F . R WED
Ss4AO3 Event lI6wW d 'Scti&ri 4AO4:Othei Atiitie
2998-G-0B4 g'Sv'nl'. .nitl2'.Elow.Diagram'Domezs~~~~~~~~~~~~~ .c' i .pp
~'eosz E R'eCnaimn~e
eraL
!
Psgian Ba'sis Docent
DonetFunctions folpreorizer Wid6 R e'P ressuire to 6&Se" ion7C?2;
ecitrument
om~ponent Functions~'for-rsneresat njei~lt onPO-bnturnnOX&Yt4o`ei6ri',7.823
[' ~~~~~.^ ~~~~~.'.6t" onttar Jrie'nt'Lo5o.:cS
fisceIlaneous
~~~ _126~~&@eto. @ oto~@ e~ gS~&TdLvl
4 ~ E,4~ .tf .g
corrective: action Droaram
ER
. de' ;of Fede Relations.
EEE nstitute of IE6tgc n EIe i gineers
KERv, icensee, event.repdort
6
r
NRC U.S2:Nucleag&Ratoe&" Coemission
PvCI~~~~~~
lat n hnc aimficabon
WOR\ owrperated reli&,'IQl
RCS reactor _ooantsvst
p5 ~signl ce'determination process
7
FIRE PROTECTION BASELINE INSPECTIONl
St]ic P0WER STATION
INP.UTrF.ORINSPECTION REPORT'NOC: 5-335;'389/2OO3'02
INPETOR~i--.
- r. P.roiect Matnae
t een
U~&6I~e~ib~ :~'TiN NIAi-,F.IRE-",PROTECTION
_ ; _ ~~~~~_
,BASELINE!INSPCIN
N PECTIONRPR]NU
WiPic
I S te
je dnicen-ar sproded
ype.8of
In'spection.idTsRIE ve'PRs2OT BASELIN waSPconE'CTedItON-e
patlria 6onitin~roblems rela~ted to ,fird'indidernt~.~Addition-IV t5.,hb~aMeviieWed
.protectionactorean tostfire roea.neptin krogram C bS§ipi t,:arid'
Oea ><,~tF:~,_,-¢,F-@!j,
e
ij_ ~bCt$ ~tvt tlj- s < jptlfllssb - t
iTindws
~ ,. tem V e rd6r.,_cess
se .yt):kW>1 a+,H , <
sb6disc6usseuthfit&?D6t6cti'o:'ersone:W6th remriecv exitiliohtinaWas~
Drotoi d~for 6etsnfibl'edts!Etior idethtifie .dithd
voriNithi eth. olwetei'cornsistents-w'ithigath' 6g temt'4
Fi'e!rhteam
iovn~ Wef,'SSd~scr ofand exatniiri'd 'ith6mdata
sheets foreewtoi~diu~6c currentheDleyenisrgeiicy Jighlnlsysem she;-co tained, water
8
powered unitsitThe i thebatte, rated'wth'at
least an
(,
8 h6 urcaicitvas
~~~~~~~- i required'.
j * ~v~;>4 ectio ll' ,' A
,;j.-..-\l - t* 1>_-!<1F>n%(* .E o
Kf.-
endix .Te team res iei
ceri6dic tes mianditeance-rocedures-and re t if.cdeute
su'r-v~eilan~-"e-dtn,~'cwvas' in~blcet6as'sret ftd'ELsj1
T firteat t iethee
wed
recordsmwere also reviewed to-ensfvthat the fire b'adD-fersonnelo'ualIfrcaii
alcense ddsapproved EPP
ad 66_
jesion 6ntrlt...~~~~~~~~~~~~~~~~~
=~bt d _a~As . t6'd'enfv th'at 'Dlant-chanaes.wereoadu
reviewedf t1ia bt on the FFPmSSD equipment and proeuresas'
eqdired byfe g e-ns bon
Audit Rep0i
_ .__ Mehnia M.~tnne Por~
,.~
PM'S
~~Fi~ ial F39:F1P&
2M00 187 Q c oz
___i"F-'"ec'ni'c, 'lten'nce^aj lte'rvnttvMiteacto
EOSP-15 10--Sel'otaie Em rrezinni Flow MsaiteRev:L
9
bodton pot-Q'Raingi p~rierncq
CR OOz1 514' failkr6f '5ooKVMa7rrasfdrier;'-SEN.215
R 01218577R~ir N218
Q01-.2459.r.4-kV Breaker~g~allure^SER.E'0n
PRLO2-1 6~19'> Potenitia! Problems wihWa olcor'.R nomto Notice .2002 24
po~ndition .Report
. ............
bR 02'20981,P. nSLChj'Cfiverl;i s - *
,R 0 , urttgbamR :Review of Se~veral Pio cedure Ch~ar~gs
I
10
FREPOTCION BASELINE NSETIN
ST., LUCIE
jSECTOR-1
Week 2b ojnj s' ~ f ~ o
T~~ofhi1pection,,,,TRIENNI.L ~F-IRE PROTETIN'BASELI NE INSPECTION:ir
REAC~~Fi~
2998 G411'.eacto Auxiliai Bu-dna -El',1 9'50, Codi.Lvu h .Rev
299WG'-.f'.ReacoruXiMfarv.Buildihca El' 90 %C6hd itC utt.~~ KrA5b.,R
2998-GL~ifzReactoiAuxil I rv.-Buildih'a"EI" 19'50 Condctlayu; W~RV
29981G-4 11--.Reabctoir Auk~ B'idr l.10 "bndii~Vuh ;,1O.6VR~
2998'-G4~11i Rect6rAxiliaVvBbildinadEI'19' Codit aotfs. ~Rv
2~98~G~4W~a66iixiiaiy Bi i~di~ El 9'50,~ C a,4,tUshdit~T ,iReV
x i6--.~ EI"19'50,Codi Layot: h~ 8-,'.Rv
99 41*1-React - - 1A
29958-G:"4I1.'1.; R&6ctorAdxiliarv BuiJd~ El 1 50 Q.C-n-du-iftLavjt. sh.. Rey,
2998:G~4 1.1; ~Reaicto'r Aukiliary BLIild~n Electrical d~it:ravpt f.-8 W
29667.64"lO~Cable Vait';Trbvq.,- KeyPa sh6~Rv' -_____
2998- G-394'- Reactor Aui77~ bidnQE 4'OCnd
A dx; Ianq:E-A30 Cridit-;Trays.&Grounhihn-.sh,.l';XtRevi'2?7
2998:G392:, RedtorA~uxiii Iidino El- 196, Conduit Tas&rdhi si.~617
2998 G'-071:.GeneraPF~rr;ii~a~neehtei RebactoF&AH6N~ Au~fiP~iSet3~v2
29981-13272A- Comfbined Main"rafid:AL'jxiliarvOrhe Line, Diaqerarin.: RbV'.
29 8B37 IIbls~aii.ave V-1477T'-sh~. ,118:,Rev?.--1.4
2998,~ 3,.L3 27 res~siu'nz'eI'r Re'ief i66i'~~V6~'V-6 ,h.'1
ffb-51.-.Rv~ Rev:.14
29984.B-37.` LIPS Pbm -2A_Suction'aV6a.V~V344&;-. 131 Rew
2998~-327:~si:I6~c~it~oVaIVeHCV-3625"'.:~h.260R.v.,16
2~998b-B732,' e'ssujrze'r R'e~lie~f .;V"alve V- 475, s. 1630, Revl 0
- 11
Unpe
dit ,R
~RSafe;Shutdown W:sis iFi ea Report
di~~~~Es~~~~nifueiC6di iio
rth'erb~ii~cents
~~b~D~iflr6at'ior~i EItrdic Cab P~ic1#LO'298.292.~;'dated ;,101~28/17j.____
/M-CE.91 7^,.xbo'ro .SDcific tionf 200 ControlSvsteniManuaI .79N-36291 'idated______
SEiieeinqSfdtV
m y-Qd
_ V~bbaK Revr.k;Or~der
II.eobtTll
P'C/M~zi 74-295MRerbuteo'ftCable'21 702 :,-'Rev.:1' ,TIdat dŽ10/29/95
- N.O3100661301 . TS/8/ 040ASDG02Rdev22 2 ibiatioi8dte01
i 0 '044BaS/G'-2B 0/1 OTdS.Y tdatedi1 1/1
N'O3'8734101221t ','S' 21.Carqinc 'PrDFozaibr'ir.dt' t2102
Nfi\012210t'iT.S41cil . ...InqPtiumn'Dis'darePi-22l2
.. Calibr fion, dj 01
at
W~9'.3i3200736501S S'Ts; Le'V','(P.1 i 108/l fdat'10 03
03 ~~~~~~~~~~~~~~~~~~~
Nt)-e3'0093JTSn
ies-'ve il: fO11 8?is ).Clbaio' adJ20
N.-oX665290p-g~a.~k.^.>. jii.Z-X,e<CO;-.3 ;(Ll0il~'l~
.te-uB- ;< -; w__~3:3
-r-BS_
j~echnical~pei f:c. ons, S- Unit 2~SR o-
=~~~~~~~~~~~~~~~~~ll_,
4.3.35.-1 J :5 i.2
Docuenttsom~o
12
?I'M- . ,,
!4FSAR"86'606'n- -'Ele'btfibbl,-P6W&
- -_:a
13
FIRE 'PROTECTION BASELINE INSPECTIONi
ST. LUCIE POWER STATION
INPUT FOR IINSPECTION REPORT NO.: 50-335, 389/2003-02
INSPECTOR:. 'Gerry Wisernan
Sr. Reactor Ins ecor-Fire Protection Systems
Engineering Branch, DRS
INSPECTION DATES: Wee'k,1 of onsite inspection - March 10 - 14, 203
d. We~j~ek 2 of onsite inspection - March24 28,2003-
Tpe of. Inspection: TRIENNIAL FIRE PROTECTION BASELINE INSPECTION'. Fire
Protection Features and Post-Fire SSafe Shutdown tapability
A. INSPECTION REPORT INPUT
A. Insctor ]etifiedFindins
';e Green., The Fire Hazards Analysis' (FHA) forthree Plant St.,Lucie (PSL) Unit 2 fire
areas/zones was inadeauate, Th~e PSELFHSAfaied to consider and evaluat:e the
combustibilitv of 380 qallons of transformer silicone dielectric insulatinq fluid, in each of
six'.transformers installed in three rUnit 2 fire zones Iascontributorslto fire loadino and
effcts onSSD, capabil~ityas required by Fire Protection Program (FPP) commitments.
A n~on-cited violation of 10 CFR. 50.48 and jPS L; Unit 2cOeratinq License Condition
6OiLC;2'`C.(20), was identified. The fin:dina is greater than minor because itwas.
associated with the `protection ,aqainst external factors' attribute and affected the
Obi"e'ctive of the initiatinqeevents cornerstone to limit the likelihood of those events that
could upset plant stability and challende critical safety functions relied upon for SSD
from afire.. The previously unidentified six silicone oil-filled transformers represented an
in an increase in the iqnition freauency of the associated fire areas/zones. The finding
was considered to have verY low safetv siqnificance (Green) because it did not involve
thef impairment or dedradation of NRC fire nrotection features and the overall
,aoroved
SSD capabilities for the areas were evaluated by the licesee's SSIA as adequate to
ensure SSD capability. (Secion 1R05.02)
f. TBD. Manv local manual ooerator actions were used in Dlace of the reouired
Dhvsical Drotection of cables for ecuiDmentrelied on for SSD durina a fire.
without obtainina NRC aDDroval for these deviations from the aworoved fire
Drotection Droaram. This condition aoolied to all areas that were insoected. This
reliance on large numbers of local manual actions, in place of the required
14
ohysicalg~~~iRW orotectiortof .x.,egi cabiesi
ZZ
lt;-&- could ootentlaliv result-ilz-itiea,3-g>
,~s. :>67aTe.#-
.as2v.1 in an increased rlsk'~of loss ~eos-;4
- A.idlatI6n of. i'S Unit.2;I6i C)2,C.(20)'and the' Fire Protection Proaranfwas
of ul~mbnt izdr#,f!ssoeumn~f
Idntfld.Hoevr.li~fndnaisunresolved Dndrna'comolel i ~__a-
sianlifca'nce.det rn~in'tlon'.JThe~ fijidino Is areater..than minor because it couild
no>ewta ~tesultij,,,,,~s~,rsrk oflos ,~ eEpmn tha, t wat r~elied" 6ij
io rSDfro .~fre (Section .1RO5XXX
rnerstoness:. I-,nitiatig. Eve nts;' Mitigating;Systems and B.rier,,lntet
n_
R0 IRE PROTECTION
S. ~e
r~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ Z4
fieddvd, .edbbho'fe
.rddulf d'd entI
crmDneft~S-5fld~rC~itS?,IOCtedWit M ifl; 6
fAe.the nbuA-b1
k otnriir~eiy oinvercaatnstsn
Brane'elchtireAPCasB 9.1 $Thetearneautethrevwe Theuenv flowinirs-orcte-Doentlah f
p atcope anspecbo
aond Iriti onintsion ai circu'is ~ ef~ locatedcwitthen..the,
HN 6fiRWns i1 same'
~6ffiWdaae;tlld~~a firesratfesrtta
t es o
tirs' om fire;d ra'teristio c and 6MDtentoI'flethbefseaoe
- s:rfl,s~t*the--- -il~eyrovi',c,, .c4,3,.fr ,,'d-';Vat -rdnc.
manerconsestozeiit'wi,o~asvth~fect6iplnt~lreaA' ~ entonicand seuralon-.an
(1O~C P) bdes th-enII* tiremnt;-f. IIT
cnbre
Rn';Seetii'
irand.nditi x~Atofea Brevnewe Tchical Pose~ih~r~i(Tec utibnreor andPowrerctive' acior
ganoredth C1ndmitiotrafe'o P votemse, Os1Or4e Pftfire _rbTeteon :,al __
P~ Prbq~arfi Re ion( ret.,,n
15
eauiDment overheatin 'incidents for the Vears,200i-2002 to assess the effectiveness of
the fire Prevention Drocram and to identify any maintenance or material condition
problems relatedjto fire incidents.
The team reviewed the fire.briaade response procedures. traininino procedures, and drill
proaram6' rocedures. The tejamreviewed Fire Briaade Initial Traininqan d Fire Briaade
Continuina Trainino course materials to verify mDoropriate trainina was beinq conducted
for the station firefiqhtinq pDersonnel., In addition, the team evaluated fire briaade drill
trainina report records for the oieratinq shifts from Auoust 2001- February 2003. The
Sreviews were..~erformed to determine whether fire briqade drills had :been conducted in
hiah fire ~risk plant areas and whether fire bri ade oersonnel qualifications, drill
response, and p~erforjm~ance met the tsof tlicensee's approved fire
protection~ program'.'.'
The'team walked down' the fire briqade staqina and dress-out areas in the turbine
buildinqs and fire briqade house to assess the, condition of fire fiahtind and ssmoke
chntrol equiment. 'The team examined the fire bnrqade's personal protective
eduipmrent. self-contained bre(athina apparatusk (SCBA). portable communications
eauipment, and various'other fire briqade eauipment to determine accessibilitvy. material
conditionrand operational readiness of equipment. Also, the availabilityof supplemental
fire bri'qade SCBA breathind air tanks, and the capability for refill was- evaluated,
AdditionallY.' the team observed whether, emermency exit iihtin was provided for
personnel evacuation pathwavs to the outside exits as identified in the National Fire
Protection Association fNFP.A) 101. Life Safety Code and Occupational Safety and
Health Administration (OSHA) Part 191 0. OccuDational Safetv and Health Standards.
This review aiso included an examrination of backup emeraencv liahtina availability on
pathways to and within the dress-out and staainoq areas to support fire briqade
operations durina a fire-induced power failure., The fire brigade self-contained breathing
apparatuses were examined and assessed for adequacy.:
er
... wa I.ke ., ., s .,o co . h , . i ...
,...,,
......
,f?..
re;........
te
Team members walked down the selected fire areas to compare the associated-fire
fiqhtina pre-.fire strateies and; dravwinqs with as-built Plant conditions. This was done to
verfy that fire fiqhtinq pre-fire strateqies and drawinas were consistent with the fire
protiction features and potential fire conditions described in the UFSAR Fire Protection
Proqram 'Report.' Also,o the team rerformed a review of drawinas and enaineering
calculations for fire suppression caused floodina associated with the floor and
'equipment drain systems for the Train "B"Switchqear Room, Electrical Euipjment
Suz~lv Fan :Roomi',and iTrain "B"Electrical Penetration Room. The review focused on
ensurng thati those, actions required for SSD would not be inhibited by fire suppression
activities or leakage from fire suppression systems.
The team reviewed desian control procedures to verify that plant chanaes were
adeauately reviewed forthe' potential impact on the fire protection rroaram. SSD
eauipment, and orocedures as reauired bv PSL Unit 2 Operatina License Condition
2.C(20). Additionaliv, the team performed an independent technical review of the
licensee's plant change documentation completed in support of 2002 temporary
16
modification. TSA 2-02-006-3. that placed two exhaust fans on a fire damp~er opening
between the cable spreadina room andtkhe Train B switchkcear room. This chanae
implemented by the licensee was evaluated in order to verifv that mbodification to the
plant were performed consistent with plant design control procedures.
b jFindings
Inadequate Fire iHazards Analysis
Introduction: The'team identified a Green non-cited violation MNMV) associated with.
failure to meet the fire pr~otection broaram plan reauirements contained in the 10 CFR
50.48 and PSL Unit 2 Operatina License Condition (OL 12.0G.420. The team ound
that six silicone oil filled transformers installed in three 2fire zones Wire Zone 37,.
2Unit
TrainA Switchaear Room. Fire ZOne:34, Traiin B Switchqear Room B.and Fire Zone 47,
Turbine Buildina Switchaear Rooml]were ~not evaluated in the Fire 0Hazards Anaiysis
fFHA) as contributors to fire loadina and effects on safeIshutdown (SSD) capability as
required by Fire Protection .pProgram commitments.
- ,,
.ire ,, ; . .. Pr gra
Descrip tion: At PSL the indoorbmedium voltacie power transformers installed in Unit 1
are of ithe dry tye.; However, six of the indoor medium ivoltacie oower transformers in
Unit,2 are cooled: and in'sulated' by a silicone-tvye fluid. The licensee provided to the
team information from the transformer manufacturer that the transformer ,insulating flu'id
was Dow Comnin O(DC) 561. a dimethvl silicone insulatina fluid.: The team oerformed an
independent technical review of the licensee's enciineerina calculations-and
maintenance documentation. transformer vendor technical information manual,
insulatinci fluid manufacturer information. Underwriters Laboratorv (UL) and FactorY
M~f'utual (FMlistinc., aciencies' documentation, andInstitute of Electrical and Electronics
Engineers l(IEEED Standards. Documents reviewed are listedl in the Attachment.
The DC 561 technical manual 'described the DC 561 fluidJ as. a silicone-liauid that will
burnbut was Iless flammable than paraffin-tvye insulating oils. The tenical manual
also stated that the DC561 fluid had a flash point of 324 o'C:a total heat release rate
(HRR) o~f; 140 WIm2 (ber ASTM E 1354-90), and a fire point iof 357 oC. In their Fire
Hazard Analvsis thealicensee evaluated the adeauacv o7f their fire area/zone and
electrical racewav fire barrier svstem (ERFBS) enclosure barrier features based on the
combustible hazard content and overall fire loadind (ainalvzed fire duration) oresent
within the associated area/zone. Based on the above, the team concluded that the
transformer insulatin fluid was a in-situ combustible liauid not 'accounted for nor
evaluated in the PSL FHA. Additionallv, the team noted that the licensee' had conducted
an UFSAR Combustible Loadinq Update evaluation in 1997. This evaluation was
documented in PSL-ENG-SEMS-97-070,. but failed to identifv that the transformers in
fire zone 37 contained combustible silicone insulatina fluid. Also a PSL Triennal Fire
Protection Audit (documented in QA audit Report QSL-FP-01-07) conducted in 2001,
reviewed the FHA but did not identify any fire loading discrepancies.
17
The team determined that the breviouslv unidentified six silicone 'oil-filled transformers
represented an in an increase> in the 'inition freauency of the associated fire
areaslzones. Also,: the'additional in-s'itu 0combustible fire6 load and fire severity
represented by the;combustible traansformer insulating fldid increased the likelihood of a
sustained fire' event from a catastrophic failure4of an effected transformer that may upset
I
plant stability and challenge ritical safety functions during SSD operations.
The i-T-E Unit Substatio Transformerst Instruction Manua-lrecommended tha the
dielectric insuiatina fluid be sampled annually and the dielectric strencth of the fluid 'be
tested to ensure that it is at 26 KV or better. The licensee determined that except`for
four tests conducted durina the beriod 1990-1992 there woereno records of the
transformersj fluid beina sampled and tested. This issue was entered into the corrective
actionprogram asCR 2003-0978 and willfollow p b th
staff.
Analysis: The team determined that this findinq was associated with the botection
aoains't.external factor.s" aftribute zandaffected the'obiective of'the initiatino events
cornerstone to limit the likelihood of those events that could uoset plant stability and
challende critical safetv functions relied. u6on for SSD from a fire, and is therefore
oreater than minor. The previouslv unidentified six silicone oil-filled transformers in Unit
2,represented an in an increase in the ignition freouenc of the associated fire
areas/zones. The findinci was considered to have verv low safety siqnificance 4(Green)
because it did not involve .the impairment, or deqradation of NRC aPbroved fire
protection features and the overall SSID capabilities for the areas were evaluated bv the
licensee s SSA as adeauate to ensure SSDI capabilitv. However. when assessed in
combination with other findings identified in this' report, the ,significance could be greater
th..n vr low significance.
Enforcement: 10 CFR 50.48 states, in part. "Each operatins 'nuclear Power: lant must
have a fire protection prrooram -that satisfies Criterion 3 of Appendix A to this part." PSL
Unit 2 Operatinq License NPF-16. Condition 2.C.(4) specifies. in part, that the licensee
implement and maintain in effect all provisions of the approved FPP as described in the
UFSAR for the facilitv'and as awcroved by the NRC letter dated JuIY 17,1984, and
subseauent suppblements. ' The approved FPP is maintained and documented in the
PSL UFSAR, Appendix 9.5A, Filre Protection Program Report.
TheU FSAR. Fire Protection Procram Report, states, in rart. that the: PSL Fire
Protection Prooram described in the report imple ments the philosoDhv of defense-in-
depath Drotection acainst fire hazards and effects of fire on safe shutdown eauipment.
The PSL fire protection proaram is auided bv Dlant fire hazard analvses and by credible
fire postulations. Itfurther stated that the Fire Hazard Analvses performed for St. Lucie
iUnit 2 considered potential fire hazards and their possible effect on safe shutdown
capability.
PSL administrative fire protection procedure, 1800022. Section 8.3 states that the FHA
for Unit 2 are individual studies of each plant's designs, potential fire hazards in the
18
lanb'Dt,nil of l~hbsveithreats: o'ccurrn th ebzff66t 6f. 66tltdfielo~a
hutdo cabilitv Further. this sttthat insit
bou'tUsteb e features.u aal'cottpldr:toflarezavd
loading inthe'res Ve~firezoneso
e FPP commin S cfiaaiI.38b aii& f in-sitLI ornbUblelrmeL
silicoI'dllct
n f eachof locatedin;UnW2Vsiitanf6
was no
cnsideran elute~dir I the ;HA~hs,'.contnibutors §tofire.Ioadi Qaa O~beefc
on
SD ca6 i Itv-;.Th is 'confd itio -w"as';c'on'trarv-to th e re a Uire m entftePCFPa
outlined in UFSAR, Sectihnj'9.5A. and therefore did noot mebtts
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- ATTACHMENT
ENGINEERING BRANCH 1 FIRE PROTECTION INSPECTION DEBRIEF
Inspection of: St. Lucie Nuclear Plant Report Number: 50-335,389/03-02
Inspection Dates: March 10-14 and 24-28, 2003 (onsite inspection)
Type of Inspection: TRIENNIAL FIRE PROTECTION BASELINE INSPECTION: Fire
Protection Features and Post-Fire Safe Shutdown Capability.
Inspectors: M. Thomas, Lead/Operations Inspector; G. Wiseman, Fire Protection Inspector; S.
Walker, Electrical Inspector; P. Fillion, Electrical Inspector (Open Items Followup); F. Jape,
Operations Inspector (Training);'R. Deem, Contractor (Mechanical Systems/Operations);
Accompanying'Personnel: R. Rodriguez, Nuclear Reactor Safety Intern, will be in training and
support the open items followup/Electrical areas.
Inspection Scope: This inspection was conducted in accordance with revised Inspection
Procedure 71111.05, Fire Protection, dated 03/23/01, and the NRC Reactor Oversight
Process. The inspection team focused their review on the separation of the systems and
equipment necessary to achieve and maintain safe shutdown and fire protection features of
these plant areas. The team used IPEEE data, with assistance from the RII Senior Risk
Analyst, to identify risk significant plant areas and components among those with the
highest CDFs and CCDPs. The fire areas/fire zones chosen for review during this
inspection are:
3. Unit 2 Fire Area B - Cable Spreading Room (Fire Zone 52). A fire in this area could
result in evacuation of the Unit 2 main control room (MCR) and the plant could be brought
to cold shutdown from a remote location even with the loss of all unprotected equipment
and cables in Fire Zone 52. Use of Train "A" equipment is credited for a fire in this area.
2. Unit 2 Fire Area C - Dual elevation fire 'areaencompassing Fire Zone 34 (Train "B"
Switchgear Room) and Fire Zone 48 (Electrical Equipment Supply Fan Room). Fire
Area C and the essential equipment and cables within, have been evaluated with respect to
the protection and separation criteria of Appendix R, Section IlI.G.2 to assure that the
ability to safely shut down the plant is not adversely effected by a single fire event. Safe
shut down of Unit 2 from the MCR using Train KAn equipment is credited for a fire in this
area.
3. Unit 2 Fire Area I consists of Fire Zone 51 West (Cable Loft), Fire Zone 21
(Personnel Rooms), Fire Zone 32 (PASS and Radiation Monitoring Room), Fire Zone
331 (Instrument Repair Shop), and Fire Zone 23 (Train "B" Electrical Penetration,'
Room). Fire Area I and the essential equipment and cables within, have been evaluated
with respect to the protection and separation criteria of Appendix R Section III.G.2 to
assure that the ability to safely-shut down the plant is not effected by a single fire event.
ATTACHMENT
Safe shut down of Unit 2 from the MCR using Train "A" equipment is credited for a fire in
this area.
INSPECTION RESULTS: Two Findings were identified.
Finding No. 1
Silicone oil filled transformers in Unit 2 fire areas were not evaluated in the Fire Hazards
Analysis (FHA) as, required by the Fire Protection Program commitments. The affected fire
areas were Fire Area A (Fire Zone 37, A SWGR Rm); Fire Area C (Fire Zone 34, B SWGR
Rm); and Fire Area QQ (Fire Zone 47, Turbine Bldg SWGR Rm). This finding is More
Than Minor. The 380 gallons of transformer silicone dielectric cooling fluid In each
transformer was not evaluated in the FHA as contributors to fire loading and effects on SSD
In FZ 34, 37 or 47.
Note: This finding affects:
1. Existing fire protection licensing bases (deviations to Appendix R granted by the NRC)
2. Current engineering evaluations allowed under GL 86-10 for fire protection barriers or
systems not submitted to the NRC (CR 02-0396, Derated Thermo-Lag fire barrier wall
partition separating the CSR and B Switchgear Room)
3. IPEEE Risk Analysis for Fire Events (the transformers were likely not accounted for in ISDS
and could affect total CDF for the fire areas.
4. The maintenance and surveillance programs for transformer related fluid sampling and
condition evaluations. (Note: Will be followed up by Resident inspectors).
The licensee initiated CRs 03-0637 and 03-0978 to address this finding
Missed Ognortunities For Identification:
- In 1997 the licensee conducted an UFSAR Combustible Loading Update evaluation
documented in PSL-ENG-SEMS-97-070 but failed to identify that the transformers in fire
zone A37 contained combustible silicone fluid.
the FHA but did not identify any fire loading discrepancies.
Finding No. 2
Use of Manual Operator actions outside the MCR for Ill.G.2 areas (Fire Area C and Fire Area I)
without prior NRC approval. Many manual operator actions were used id lieu of physical
protection of cables and equipment relied on for SSD during a fire. This was a deviation
ATTACHMENT
from the approved Fire Protection Program. The licensee identified this issue in CR 03-
0153 prior to this inspection. This finding is More Than Minor. This finding will be
Unresolved pending completion of the SDP to determine the risk associated with using the
manual operator actions in lieu physical protection. (NOTE: The NRC and the Nuclear
industry are working to resolve this issue on a generic basis).'
In addition to the two findings, eight condition reports (CRs) were written as a result of
this inspection. The CRs were evaluated against and determined to meet the NRC
criteria for minor issues and will not be discussed in the report details.
CR 03-0847 Hot shutdown repairs using tools to achieve safe shutdown in the event of a
fire
CR 03-0888 Update UFSAR to delineate that Deviation C6 previously approved bythe
NRC for fire areas A & C is no longer required
CR 03-0942 Discrepancies between the safe shutdown analysis (SSA), essential
equipment list (EEL), and the breaker/fuse coordination study
CR 03-0964 Rubatex insulation installed on instrument lines in the U2 intake (fire area R-
R)is not considered in the FHA
CR 03-0965 Combustible fire load for Ul and U2 intake fire areas same in the field but
different values listed each unit's FHA
CR 03-0966 Temp Mod (installation of fans between cable spreading room and B SWGR
room) did not sufficiently evaluate potential impact on fire protection
CR 03-0986 Discrepancies between SSA and EEL. Determined that EEL was in error
CR 03-1010 Cold shutdown repairs identified in licensee procedures, but UFSAR states
that no credit is taken for post-fire repair of cold shutdown equipment
Open Items Reviewed: Three open items assigned to EB1 were -reviewed for closure.
URI 50-335,389/99-08-03, PORV Cabling May Not Be Protectedfrom Hot Shorts Inside . .
Containment (Closed - Green NCV) -
LER 50-335,389/00-001, Outside Design Bases Appendix R Hi-Lo Pressure Interface and
Separation Issues
LER 50-335/00-004, Pressurizer Level Instrumentation Conduit Separation Outside Appendix R
Design Bases
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ATTACHMENT
I
LESSONS LEARNED: -
Successes:
- Followed up on three open items
- Nuclear Safety Intem (Reinaldo Rodriguez) involvement and support on open items
- Experience/knowledge of Fire Protection Inspector
- Resident inspector followup of licensee's sampling of transformer oil
Challenges:
- Better coordination by team leader with licensee for open item followup
- Completing SDP for the open items
- Effect of fire on instrumentation needs to be reviewed in more depth and detail
ATTACHMENT