ML040090432
| ML040090432 | |
| 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: ML040090432 (25) | |
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.O. 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.Jemigan 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.
r
This report documents a finding concerning silicone oil-filled transformers in the B Switchgear
Room which had not been considered or evaluated in the licensee's fire hazards analysis. X
Additionally, a finding was identified concerning the use 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 and,
combined, 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 VL.A of
the NRC Enforcement Policy. 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.
I P
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
httn://www.nrc.oov/reading-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)
I
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
U. S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos: 50-335, 50-389
Report No: 50-335/03-02 and 50-389/03-02
Licensee: Florida Power and Light Company (FPL)
Facility: St. Lucie Nuclear Plant
Location: 6351 South Ocean Drive
Jensen Beach, FL 34957
Dates: March 10- 14, 2003 (Week 1)
March 24 - 28, 2003 (Week 2)
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
. ;
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SUMMARY OF FINDINGS
IR 05000335/2003-002,0500038912003-002; Florida Power and Light Company; 03/10 -
28/2003; St. Lucie Nuclear Plant, Units ISand 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-ldentified and Self-Revealing Findings
Cornerstone: Initiating Events
- TBD. The team identified a violation of 10 CFR 50.48 and the St. Lucie Nuclear
Plant Unit 2 Operating License Condition 2.C.(20), Fire Protection. The fire
' hazards analysis 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
commitments.
This finding is'unresolved pending completion of a significance determination.
The 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. Also, when
assessed with other findings identified in this report, the significance could be
greater than very low significance. (Section 1R05.02.b(1))
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 480V 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 safe shutdown
(SSD) in the event of a fire. In lieu of providing adequate physical protection for
load center 2A5 and associated electrical cables, manual operator actions
outside the main control room (MCR) were relied on and credited, without prior
NRC approval, for achieving and maintaining SSD.
2
This finding is unresolved pending completion of a significance determination.
The finding was greater than minor because fire damage to the unprotected
cables could prevent operation of SSD equipment from the MCR and challenge
the operators' ability to maintain adequate reactor coolant system inventory and
reactor coolant pump seal flow during a fire in the B switchgear room. (Section
1R05.02.b(2))
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 systems
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 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, were reviewed by the
inspection 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 40A7 of this report
... a...;jrr .
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 Scope
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;
AppendixkA 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 Plant St. Lucie (PSL) 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, including the
essential equipment and cables within, was 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 affected by a single fire event. Train A equipment would be used to
achieve safe shutdown from the Unit 2 MCR during 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 , including the essential equipment and cables
within, was 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 affected by a single fire
event. Train A equipment would be used to achieve safe shutdown from the Unit
2 MCR during a fire in this area.
The team reviewed the licensee's fire protection program (FPP) documented in the PSL
UFSAR (Appendix 9.5Aj Fire Protection Program Report); safe shutdown analysis
2
(SSA); fire hazards analysis (FHA); safe shutdown (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
SSD equipment and post-fire SSD analytical approach were consistent with and
satisfied the Appendix R reactor performance criteria for SSD. 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 SSD in the event of a fire
in those fire areas. 'Systems and/or components selected for review included the
pressurizer power operated relief valves (PORVs); boric acid makeup pumps 2A and
2B; boric acid gravity feed valves V2508 and V2509; auxiliary feedwater (AFW);
charging pumps and volume control tank outlet valve V2501; shutdown cooling; heating,
ventilation, and air conditioning (HVAC); atmospheric dump valves (ADVs); and
component cooling water. The team also reviewed the licensee's maintenance program
to determine if a sample of manual valves used to achieve SSD were included.
b. Findings
No findings of significance were identified.
.02 Fire Protection of Safe Shutdown Capability
a. Inspection Scooe
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 damage. 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 established 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 FPP were satisfied:
- UFSAR, Appendix 9.5A, Fire Protection Program Report
- PSL Individual Plant Examination of External Events (IPEEE)
- Administrative Procedure 1800022, Fire Protection Plan
- Administrative Procedure 0010434, Plant Fire Protection Guidelines
- Electrical Maintenance Procedure 52.01, Periodic Maintenance of 4160 Volt (V)
Switchgear-'-;'
3
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, 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, training, and drill program procedures.
The team reviewed fire brigade initial and continuing training course materials to verify
that appropriate training was being conducted. 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 FPP.
The team walked down the fire brigade staging and dress-out areas in the turbine
building 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 caDabilitv for refill, was evaluated. In' addition;-the ;
team' observed whethe'r emerency, exit lihtina was provided forersonnel evacuation
pathwavs tlthe outside exits as ientief inthe' National Fire"Protection Association
(NFPA :101. Life Safetv.Code and Occupational Safety and Heat -Administration
(OSHA)'Part'191O,-Occup a Safety ahd H alth Standards: This review also
tional
included an examination of backup emergency lighting units along pathways to, and.
within, the dress-out and staging areas in support of fire brigade operations during a fire- .
induced power failure.
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 I
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, the electrical equipment
supply fan room, and the Train B electrical penetration room. The review focused on
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
4
2.C(20). Additionally, the team performed an independent technical review of the
licensee's plant change documentation completed in support of 2002 temporary system
alteration (TSA) 2-02-006-3, which 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. Findinas
Fire Area C - Train B Switchgear Room
(1) Inadequate Fire Hazards Analysis
Introduction: A violation was identified concerning failure to meet the FPP 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 FHA as
contributors to fire loading, and their effects on SSD capability, as required by the FPP.
Description: During a pre-inspection plant walk down on February 26, 2003, the team
found six Unit 2 indoor medium-voltage power transformers that 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
Corning (DC) 561, a dimethyl silicone insulating fluid. The team performed an
independent technical review of the licensee's engineering calculations and
maintenance documentation, 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 would
burn, 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 o, a total heat release rate,
'~~~~ 'a:-rii';'~+.i-'-;,.:jA*,- '-. .
- -' i .. '-.-z-I ' .,
in their FHfthe
(HRR) of 140 kw/m_ (ser ASTM E 1354-90). and a fire point of 357 oC.G!;g~;slL ;~ d _
li6nse 'valuated thadeauaci of their fire areazone-and eectrdI'4rcwaVfire
barier stem (ERFBS encoIsure barrier features ased on the combustible hazard
conten andoverall fire oading, (analyzed fire durati) present' with n the c
area/zone. Based on the above, the team concluded that the transformer insulating fluid
was an in-situ combustible liquid that had not been accounted for nor evaluated in the
PSL FHA. Additionally, the team noted that the licensee had conducted an UFSAR
Combustible Loading Update evaluation in 1997. This evaluation, documented in PSL-
ENG-SEMS-97-070, 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.
5
The team determined that the previously unidentified six silicone oil-filled transformers
represented an increase in the ignition frequency 6f 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 -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 nto 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
subseq pplements. 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
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.
I
6
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).
Failure to evaluate in-situ cobustible'transformer silicoie' dielectric insulatinfljid as-a
contributor tofire ladind in the FHA. when assessed inriobination with other finding
identified inthis reDort. could b'e areatr than very low sicnificance ' This fir'dina has
been entered into the licensee s cective tion roaramr as condition reort (CR) 03-
0637
063 . Howeve~ 'ei' .. 'i't -_- t "'Th--"" " h-- dentified in this
-dI
rfi !dings
reDort. the' si6gificance could be areater tan verv low siificance -,Thlg firdinji Is,
unresolved !t6KURl)_ 50-389103-02-OX. Failud to, Evaluate 1n-sI[u C-6mbu-ibr
Transformer Dlelectrc Insulating Fluid a aContribtuto' Fl oig nh
FHA.
(2) Use of Manual Operator Actions Outside the MCR for 10 CFR 50. Appendix R.
Section III.G.2 Areas
.03 Post-Fire Safe Shutdown Circuit Analysis
a. Inspection Scope
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-the chosen fire areas power sources, and their
respective cable functions and system flow diagrams were reviewed. Controtcircul
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 and components were reviewed during the
inspection:
V1474 and V1475, Pressurizer PORVs
V1476 and V1477, Pressurizer Isolation Block Valves
7
- MV-09-03 and MV-09-04, Feedwater Bypass Valves
- 2HVE-13B, 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:
- P-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 Capabilitv
a. Inspection 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 lll.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 Capability
a. Inspection Scooe
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-100.01, 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 inCR 03-0153 prior to this inspection. This finding is More Than
Minor. This finding will'beUnresolvedpending ompleiion-ofthe SDP to determine the
risk associated with using manual operator actions in lieu physicafprotection 10 CFR -
50, Appendix R, Section IIL.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 (III.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 licensing 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 inspection 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.
- 9
06. Communications -
a. Inspection Scone
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.
07. Emergencv Lighting
a. Inspection Scope
The team reviewed licensee emergency lighting against the requirements of 10 CFR 50,
Appendix R, Section 1ll.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. Inspection 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. 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 Ltter
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 Eauipment
11
a. Inspection Scope
The team reviewed flow diagrams, electrical schematic diagrams, periodic test
procedures, engineering technical evaluations for NFPA code deviatidns,>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
-adequacyof 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. Additionallyjthe team observed placement of the fire
hoses and extinguishers to assess consistency with the fire fighting pre-plan 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-assessments, 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 and 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 Event Followu2
.1 (Closed) LER 50-335. 389/00-01, Outside Design Bases Appendix R Hi-Lo Pressure
Interface and Separation Issues.
On March 9, 2000, the licensee identified seven cases where the plant 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 other 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 problem by installing new power cables
using armored cable. This precluded the possibility of cable to cable short circuits.
13
Inspectors confirmed implementation of the modification through review of plant
modification PCM01028.
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 arid 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-1105, PT-1106 and PT-1107 are in Cable tray L2224; and PT-I103, PT-1104
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 23091Awere 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
IlIl, 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 toall pressurizer level instrumentation loops. LT-
111OX and LT-1105 are in tray L2213; and LT-111OY and LT-1104 are'in conduits
23320D and 23090A. LT-11I0X &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
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
concerns about spurious opening. Therefore, rising level and concomitant pressure rise
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
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 I,
G, 2. Refer to Section 40A7 of this report for enforcement aspects.
Pressurizer level instrumentation affected by conduit to conduit interaction
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,.
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 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 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 certain 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
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 violation of 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 Appendix R accepted 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
I
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 in relation 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 1 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 Il.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.
Description: 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 coOld 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 pres'sure 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 -offines 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 coidbe 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 II, 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. Jernigan 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 liI, 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-I 108 on Unit 2
3. Pressurizer level instrumentation LT-I11OX and LT-110Yon Unit.-I-& 2-_-
4.--- Cabl6s contained in cable trays L2223 (Train A) and L2224 (Train 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 Intern (Trainee)
T. Ross, Senior Resident Inspector
S. Sanchez, Resident Inspector
. I-