ML040090436

Draft IR 05000335-03-002 and IR 05000389-03-002 on 03-10-28/03, St. Lucie Nuclear Plant, Units 1 and 2, Revision 1. Violations Noted

ML040090436
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: ML040090436 (32)

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 two findings that, 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, however a fire watch

is in place as a compensatory measure.

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. Also, two licensee-identified violations which were determined to

be of very low safety significance are listed in this 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

httD://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

License Nos.: DPR-67, NPF-16

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

U.S. Nuclear Regulatory Commission

P.O. Box 6090

Jensen Beach, Florida 34957

Craig Fugate, Director

Division of Emergency Preparedness

Department of Community Affairs

2740 Centerview Drive

Tallahassee, Florida 32399-2100

M. S. Ross, Attorney

Florida Power & Light Company

P.O. Box 14000

Juno Beach, FL 33408-0420

Mr. Douglas Anderson

County Administrator

St. Lucie County

2300 Virginia Avenue

Fort Pierce, Florida 34982

Mr. William A. Passetti, Chief

Department of Health

Bureau of Radiation Control

2020 Capital Circle, SE, Bin #C21

Tallahassee, Florida 32399-1741

Mr. Donald E. Jernigan, Site Vice President

St. Lucie Nuclear Plant

6501 South Ocean Drive

Jensen Beach, Florida 34957

Mr. R. E. Rose

Plant General Manager

St. Lucie Nuclear Plant

6501 South Ocean Drive

Jensen Beach, Florida 34957

Mr. G. Madden

Licensing Manager

St. Lucie Nuclear Plant

6501 South Ocean Drive

Jensen Beach, Florida 34957

Mr. Don Mothena

Manager, Nuclear Plant Support Services

Florida Power & Light Company

P.O. Box 14000

Juno Beach, FL 33408-0420

Mr. Rajiv S. Kundalkar

Vice President - Nuclear Engineering

Florida Power & Light Company

P.O. Box 14000

Juno Beach, FL 33408-0420

Mr. J. Kammel

Radiological Emergency

Planning Administrator

Department of Public Safety

6000 SE. Tower Drive

Stuart, Florida 34997

Attorney General

Department of Legal Affairs

The Capitol

Tallahassee, Florida 32304

Mr. Steve Hale

St. Lucie Nuclear Plant

Florida Power and Light Company

6351 South Ocean Drive

Jensen Beach, Florida 34957-2000

Mr. Alan P. Nelson

Nuclear Energy Institute

1776 I Street, N.W., Suite 400

Washington, DC 20006-3708

APN@NEI.ORG

David Lewis

Shaw Pittman, LLP

2300 N Street, N.W.

Washington, D.C. 20037

Mr. Stan Smilan

5866 Bay Hill Cir.

Lake Worth, FL 33463

U. S. NUCLEAR REGULATORY COMMISSION

REGION II

Docket Nos.:

License Nos.:

Report No.:

Licensee:

Facility:

Location:

Dates:

Inspectors:

50-335, 50-389

DPR-67, NPF-16

50-335/03-02 and 50-389/03-02

Florida Power and Light Company (FPL)

St. Lucie Nuclear Plant

6351 South Ocean Drive

Jensen Beach, FL 34957

March 10- 14, 2003 (Week 1)

March 24 - 28, 2003 (Week 2)

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; 3/10-28/2003;

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.

One Green non-cited violation (NCV) and two unresolved items 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: Mitigating Systems

TBD. The St. Lucie 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. Three of the six

transformers were located in the Train B Switchgear Room (Fire Area C). As a

result, the transformers' contribution to fire loading and their effects on SSD

capability had not been assessed as required by the Fire Protection Program.

This finding is unresolved pending completion of a significance determination.

The finding is greater than minor because it affected the mitigating systems

cornerstone objective to ensure the availability, reliability and capability of

systems that respond to initiating events to prevent undesirable consequences.

When assessed with the inadequate equipment physical protection finding (also

discussed in this report), the finding may have potential safety significance

greater than very low significance. (Section 1 R05.02.b.1)

TBD. The physical protection of equipment relied upon for a safe shutdown

(SSD) of Unit 2 during a fire in the Train B Switchgear Room (Fire Area C) was

not adequate. The Train A 480V vital load center 2A5, and its associated

electrical cables, was located in the Train B Switchgear Room without adequate

spatial separation or fire barriers as required by the Fire Protection Program.

Local, manual operator actions (which had not been reviewed and approved by

NRC) would be used to achieve and maintain SSD of Unit 2 in lieu of providing

adequate physical protection for load center 2A5 and its associated electrical

cables.

This finding is unresolved pending completion of a significance determination.

The finding is greater than minor because fire damage to the unprotected cables

could prevent operation of SSD equipment from the main control room and

because it affects the mitigating systems cornerstone objective. When assessed

with the silicone oil-filled transformer finding (also discussed in this report), the

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finding may have potential safety significance greater than very low significance.

(Section 1 R05.02.b.2)

Green. The inspectors identified a non-cited violation for the licensee's failure to

comply with 10 CFR 50, Appendix R, Criterion III.G.2. This finding is related

to a lack of spacial separation or barriers to protect cables in containment which

could result in spurious opening of the pressurizer power operated relief valve

(PORV) during a fire.

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 the ability to

achieve and maintain the reactor in a hot shutdown condition. 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

Six violations of very low safety significance (previously 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.

REPORT DETAILS

1.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems and Barrier Integrity

1 R05 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;

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 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)

Unit 2 Fire Area C, Train B Switchgear Room (Fire Zone 34) and Electrical

Equipment Supply Fan Room (Fire Zone 48)

Unit 2 Fire Area I, Cable Loft (Fire Zone 51 West), Personnel Rooms (Fire Zone 21),

PASS and Radiation Monitoring Room (Fire Zone 32), Instrument Repair Shop (Fire

Zone 331), and Train B Electrical Penetration Room (Fire Zone 23)

The team reviewed the licensee's fire protection program (FPP) documented in the PSL

UFSAR (Appendix 9.5A, Fire Protection Program Report); safe shutdown analysis

(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:

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 (VCT) outlet valve V2501; shutdown cooling;

heating, ventilation, and air conditioning (HVAC); atmospheric dump valves (ADVs); and

component cooling Water (CCW). The team also reviewed the licensee's maintenance

program to determine if a sample of manual valves used to achieve SSD were included.

2

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

SSD 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

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, corrective action

program (CAP) condition reports (CRs) resulting from fire, smoke, sparks, arcing, and

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

3

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. In addition, the

team examined personnel evacuation pathways to verify that emergency exit lighting

was provided to the outside in accordance with the National Fire Protection Association

(NFPA) 101, Life Safety Code, and the Occupational Safety and Health Administration

(OSHA) Part 1910, Occupational Safety and Health Standards. This review 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

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

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 in 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

1. Inadeauate Fire Hazards Analysis

Introduction: A finding was identified in that six silicone oil-filled transformers were not

identified or evaluated in the FHA as contributors to fire loading and fire ignition

frequency, as well as their effects on the SSD capability of Unit 2. These transformers

were located in three separate fires areas including the Train B Switchgear Room. This

is an unresolved item (URI) pending completion of the significance determination

process (SDP).

4

Description: During a pre-inspection plant walk down on February 26, 2003, the team

found six oil-filled transformers installed in three Unit 2 fire areas/fire zones. [One

transformer in Fire Area A/Fire Zone 37 (Train A Switchgear Room); three transformers

in Fire Area C/Fire Zone 34 (Train B Switchgear Room); and two transformers in Fire

Area QQ/Fire Zone 47 (Turbine Building Switchgear Room).] The team found these

transformers had not been evaluated in the FHA as contributors to fire loading, and for

their effects on SSD capability, as required by the FPP. Each indoor medium-voltage

power transformer is cooled and insulated by about 380 gallons of Dow Corning 561, a

dimethyl silicone-type insulating fluid. This finding was entered into the licensee's CAP

as CR 03-0637. The team also noted that the licensee had several opportunities over

the past six years but failed to recognize this condition. [A 1997 UFSAR Combustible

Loading Update evaluation (PSL-ENG-SEMS-97-070) and a 2001 PSL triennial fire

protection audit (QA Audit Report QSL-FP-01-07).]

-

The team also identified that the transformer insulating fluid had not been annually

sampled to confirm its dielectric strength as recommended by the l-T-E Unit Substation

Transformers Instruction Manual. The licensee determined that, except for four tests

conducted during the period 1990-1992, there were no records of the transformer fluid

being sampled and tested. This issue regarding failure to sample the transformer fluid

in accordance with the vendor's manual was entered into the CAP as CR 03-0978.

Analysis: The team determined that this finding was associated with the "protection

against external factors" and "equipment performance" attributes. It affected the

objective of the mitigating systems cornerstone to ensure the availability, reliability, and

capability of systems that respond to initiating events, and is therefore greater than

minor. In combination with other findings identified in this report, the team determined

the finding had potential safety significance greater than very low, safety significance

because the higher fuel loading in the associated fire areas/zones could increase the

duration and severity of postulated fires in those areas beyond that previously analyzed.

However, this finding is unresolved pending completion of a significance determination.

Enforcement: 10 CFR 50.48 states, in part, that each operating nuclear power plant

must have a fire protection program that satisfies Criterion 3 of 10 CFR 50, Appendix A.

PSL Unit 2 Operating License NPF-16, Condition 2.C.(20) states, in part, that the

licensee shall implement and maintain in effect all provisions of the approved FPP as

described in the UFSAR, and supplemented by licensee submittals dated July 14,1982,

February 25, 1983, July 22, 1983, December 27, 1983, November 28, 1984, December

31, 1984, and February 21, 1985 for the facility; and as approved in the NRC Safety

Evaluation Report Supplement 3 dated April 1983, and supplemented by NRC letter

dated December 5, 1986. The approved FPP is maintained and documented in the PSL

UFSAR, Appendix 9.5A, Fire Protection Program Report.

The Fire Protection Program Report states, in part, that the PSL fire protection program

implemented the philosophy of defense-in-depth protection against fire hazards and

effects of fire on SSD equipment. The PSL fire protection program is guided by the

5

plant FHA and by credible fire postulations. Further, it states that the FHA performed for

PSL Unit 2 considered potential fire hazards and their possible effect on SSD capability.

Contrary to the above, the licensee failed to meet 10 CFR 50.48 and their FPP

commitments, in that, they did not adequately evaluate the combustible fire loading in

the FHA for Fire Area A/Fire Zone 37, Fire Area C/Fire Zone 34, and Fire Area QQ/Fire

Zone 47. 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 its possible effects on SSD

capability. Pending determination of the finding's safety significance, this finding is

identified as URI 50-389/03-02-01, Failure to Provide Adequate Protection for

Redundant Safe Shutdown Equipment and Cables in the Event of a Fire in the Unit 2

Train B Switchgear Room (Fire Area C).

2. Inadequate Protection of EcuiDment and Cables Required for Safe Shutdown

Introduction: A finding was identified in that physical protection of the Train A 480V vital

load center 2A5, and its associated electrical cables, located in the Train B Switchgear

Room (Fire Area C) did not meet the requirements of 10 CFR 50, Appendix R, Criterion

Ill.G.2. Instead, the licensee substituted the use of local, manual operator actions,

which had not received NRC approval, to achieve and maintain SSD. This is a URI

pending completion of the SDP.

Description: On January 22, 2003, the licensee identified that PSL relied on manual

operator actions outside the MCR for SSD in non-alternative shutdown fire areas (i.e.,

areas designated as complying with 10 CFR 50, Appendix R, Criterion III.G.2) and that

the manual actions did not have prior NRC approval. The licensee documented this

issue in CR 03-0153. The team reviewed the local, manual operator actions for the

Criterion III.G.2 areas selected for this inspection (Fire Area C and Fire Area I). The

finding related to physical protection deficiencies in Fire Area C is discussed in this

section of the inspection report. The finding related to physical protection deficiencies

relative to Fire Area I is discussed in Section 40A7 of this inspection report.

The team found that 480V vital load center 2A5 (a Train A component) and its

associated electrical cables were located in the Train B Switchgear Room without

adequate spatial separation or fire barriers. Load center 2A5 provides power to boric

acid makeup (BAM) pumps 2A and 2B via motor control center (MCC) 2A6. MCC 2B5,

also located in the Train B Switchgear Room, provides power to the boric acid gravity-

feed motor operated valves V2508 and V2509. The licensee's SSA stated that the BAM

pumps and the boric acid gravity feed valves were redundant to each for achieving and

maintaining SSD. However, rather than providing adequate physical protection for

redundant trains of systems necessary to achieve and maintain SSD (as specified for

Appendix R, Criterion III.G.2 areas), the licensee substituted the use of manual operator

actions outside the MCR. The use of local manual operator actions, in fire areas

designated as complying with the provisions of Appendix R, Criterion III.G.2, requires

6

prior NRC review and approval. These local manual actions had not received NRC

approval.

The licensee developed an SSD time line of local, manual operator actions for the Train

B Switchgear Room (Fire Area C) to determine when the SSD actions were needed and

if personnel, and procedural guidance, were adequate to perform these actions. The

licensee determined that the assistance of one operator from the Unit 1 operating shift

staff was needed to perform the actions required for SSD of Unit 2 during a fire in this

area. The team reviewed the time line, personnel needed (compared to TS shift staffing

requirements), and the SSD procedures. Based on this review, the team determined

that potential maloperation was properly accounted for in the SSD procedures and that

adequate personnel were available to perform the specified local, manual operator

actions. The team assessed the feasibility of these local, manual operator actions using

the guidance provided in Enclosure 2 of NRC Revised Oversight Program (ROP)

Procedure 71111.05, Fire Protection, dated March 6, 2003. The team determined that

the manual actions were reasonable and met the criteria in Enclosure 2 of Procedure

71111.05.

Analysis: The team determined that this finding was associated with the "equipment

performance" attribute of the mitigating systems cornerstone. It affected this

cornerstone's objective to ensure the availability, reliability, and capability of systems

that respond to initiating events, and is therefore greater than minor. In combination

with other findings in this report, the team determined that this finding had potential

safety significance greater than very low, safety significance 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. However,

this finding is unresolved pending completion of a significance determination.

Enforcement: The licensee's Fire Protection Program commits to 10 CFR 50, Appendix

R,Section III.G. Criterion III.G.2 states in part, that,

"...where cables or equipment, including associated non-safety circuits 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 are located within the same fire area outside of

primary containment, one of the following means of ensuring that one of the

redundant trains is free of fire damage shall be provided:

(a) Separation of cables and equipment of redundant trains by a fire barrier

having a 3-hour rating.

(b) Separation of cables and equipment of redundant trains by a horizontal

distance of more than 20 feet with no intervening combustibles or fire

hazards. In addition, fire detectors and an automatic, fire suppression

system shall be installed in the fire area.

7

(c) Enclosure of cable and equipment of one redundant train in a fire barrier

having a 1-hour rating. In addition, fire detectors and an automatic, fire

suppression system shall be installed in the fire area."

Contrary to the above, on March 28, 2003, the team found that the licensee failed to

protect redundant equipment (powered by Train A load center 2A5 and Train B MCC

2B5) located within the Train B Switchgear Room (Fire Area C) with an adequate fire

barrier or to provide 20 feet of separation. Pending determination of the finding's safety

significance, this finding is identified as URI 50-389/03-02-02, Failure to Provide

Physical Protection for Redundant Safe Shutdown Equipment and Cables in the Event

of a Fire in the Unit 2 Train B Switchgear Room (Fire Area C).

.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. 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 schedules, and conduit and tray drawings. The team walked down the chosen 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

MV-09-03 and MV-09-04, Feedwater Bypass Valves

2HVE-1 3B, Control Room Booster Fan

V2501, Volume Control Tank 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

Safety Injection Actuation System Logic

8

2A5/2A6 and related feeds, 480V Motor Control Center

2B512B6 and related feeds, 480V Motor Control Center

Load Center 2A5 480V Switchgear

b.

Findings

No findings of significance were identified.

04.

Alternative Post-Fire Safe Shutdown Capability

a.

Inspection Scope

The Cable Spreading Room (Fire Area B), one of two alternative shutdown (ASD) fire

areas listed in the licensee's SSA, 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 MCR 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. This review also included a comparison of actions in procedures with the

licensee's thermal hydraulic time line analysis.

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 SSD capability from the MCR. The team's review was performed to

verify that adequate isolation capability of equipment used for SSD implementation was

in place, accessible, and that the HSCP was capable of controlling all the required

equipment necessary to bring the unit to a SSD.

b.

Findings

No findings of significance were identified.

05.

Operational Implementation of Post-Fire Safe Shutdown Capability

a.

Inspection Scope

The team reviewed off-normal operating procedures 2-ONP-100.02, Control Room

Inaccessibility, Rev. 13B [the licensee's procedure for ASD] and 2-ONP-100.01,

Response to Fire, Rev. 9 [the licensee's procedure for post-fire SSD from the MCR].

The review focused on ensuring that all required functions for post-fire SSD and the

corresponding equipment necessary to perform those functions were included in the

9

procedures. The review also examined the consistency of the operator's shutdown

procedures with other procedure-driven post-fire SSD activities (i.e., fire fighting

activities).

b.

Findings

No findings of significance were identified. The licensee identified that manual operator

actions outside the MCR were used in lieu of physical protection of equipment and

cables relied on for SSD during a fire, without obtaining prior NRC approval. Findings

related to this issue are discussed in Section 1 R05.02.b.2 of this inspection report for

Fire Area C, and in Section 40A7 of this inspection report for Fire Area I.

06.

Communications

a.

Inspection Scope

The team reviewed plant communication capabilities to verify that they were adequate

to support unit shutdown and fire brigade duties. This included verifying that site paging

(PA), portable radios, and sound-powered phone systems were consistent with the

licensing basis and would be available during fire response activities. 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. In addition, the team reviewed the periodic

testing of the site fire alarm and PA systems, the maintenance checklists for the sound-

powered phone circuits and amplifiers, and the 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.

Emergency Lighting

a.

InsDection Scope

The team compared the installation of the licensee's emergency lighting systems to the

requirements of 10 CFR 50, Appendix R, Criterion II.J, to verify that 8-hour emergency

lighting coverage was provided in areas where manual operator actions were required

during post-fire SSD 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 post-fire SSD procedures as

described in the UFSAR, Appendix 9.5A, Section 3.7. During plant walk downs of

selected areas where local manual operator actions would be performed, the team

10

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 directed by the procedures. The team also inspected emergency

lighting features along access and egress pathways that would be used during SSD

activities for adequacy and personnel safety. The team checked a sample of ELU

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

No findings of significance were identified.

08.

Cold Shutdown ReDairs

a.

Insoection ScoDe

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 ScoDe

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 selected several fire barrier features for detailed evaluation and inspection to verify

proper installation and qualification. These features included fire barrier penetration fire

stop seals, fire doors, fire dampers, fire barrier partitions, and Thermo-Lag ERFBS

enclosures.

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 verify that the observed

fire barrier penetration seal and ERFBS configurations conformed with 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

11

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

a.

Inspection Scope

The team reviewed flow diagrams, electrical schematic diagrams, periodic test

procedures, engineering technical evaluations of NFPA code deviations, 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

team assessed the common fire protection water delivery and supply components to

determine if they 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 the operation of the

installed configurations. The team also reviewed engineering drawings for fire detector

spacing and locations in the three 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 to verify the proper type, placement and spacing of the

heads/nozzles as well as the lack of obstructions for effective functioning. 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

adequate design, installation, and operation in 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 confirm consistency with the fire fighting pre-plan drawings.

12

b.

Findings

No findings of significance were identified.

4.

Other Activities

4OA2 Problem Identification and Resolution

a.

Inspection Scoge

The team reviewed a sample of licensee audits, self-assessments, and CRs to verify

that items related to fire protection and to SSD were appropriately entered into the

licensee's CAP in accordance with the PSL 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. In

addition, the team reviewed the licensee's applicability evaluations and corrective

actions for selected industry experience issues related to fire protection. The operating

experience (OE) reports were reviewed to verify that the licensee's review and actions

were appropriate.

b.

Findings

No findings of significance were identified

40A3 Event Followup

.01

(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, Criterions IlI.G.2.d and Ill.G.2.f. The first

case, involving the pressurizer PORVs, applied to Units 1 and 2, and is discussed in

Section 4A05 of this report. The other six cases apply to Unit 2 only, and are discussed

below.

a.

Shutdown Cooling Valves

Shutdown cooling (SDC) system valves, V3652 and V3481, isolate the SDC piping from

the RCS while the plant is operating. The SDC piping is not rated for RCS normal

operating pressure. Hence, these valves are procedurally de-energized in the closed

position during normal plant operation. Only one valve needs to remain closed to

effectively isolate the SDC piping from RCS pressure. The licensee found that the

power cables for these valves were routed through a pull box (JB-2031), located in the

annulus region of containment, which also contained other three-phase power cables.

During a fire, one or both of these motor-operated valves could spuriously open due to

fire-induced cable-to-cable short circuits. Should both valves open when the RCS is at

13

normal operating pressure, a pressure relief valve would open and RCS coolant would

flow from the RCS to the containment sump causing a loss of coolant accident (LOCA).

SDC valve V3545 is a normally open motor-operated valve in series with V3652 and

V3481 which could be closed by the operator to re-isolate the SDC piping. However,

the power cables for V3545 also could be damaged by the fire. The licensee corrected

the problem by installing new power cables using armored cable. The inspectors

confirmed implementation of the modification through review of plant modification

PCM01028. This finding is more than minor because it could adversely affect the

equipment reliability objective of the mitigating systems and barrier integrity

cornerstones. Using Appendix F of the SDP, the inspector determined that the finding

was of very low safety significance (Green) because the likelihood of an LOCA event

occurring was very low and cables for systems used to mitigate an LOCA were located

outside containment. This licensee-identified finding involved a violation of 10 CFR 50,

Appendix R, Criterion III.G.2 requirements. The enforcement considerations for this

violation are given in Section 40A7. This issue is closed.

.b

Pressurizer Pressure Instrumentation Affected by Tray-Conduit Interaction

The licensee identified that cable tray L2224, located in containment, lacked 20-foot

separation or a radiant heat shield to prevent interaction with conduits 25018Y and

23091A during a fire as required by 10 CFR 50, Appendix R, Criterion III.G.2.

Pressurizer pressure instruments PT-1105, PT-1106 and PT-1107 were routed in cable

tray L2224; and pressure instruments PT-1 103, PT-1 104 and PT-1 108 were routed in

conduits 25018Y and 23091A. Instruments PT-1107 and PT-1108 would be used to

achieve and maintain SSD during a fire. The licensee corrected this finding by

protecting conduits 25018Y and 23091A with a radiant heat shield 20 feet on either side

of cable tray L2224 (plant modification PCM99104, Supplement 1). The inspector

evaluated the consequences and ramifications of these instruments failing high or low,

as well as the availability of pressurizer pressure instruments which would remain

unaffected by the fire. This finding is more than minor because it could adversely affect

the equipment reliability objective of the mitigating systems cornerstone. Using

Appendix F of the SDP, the inspector determined that the finding was of very low safety

significance (Green) because the affected instrumentation would not lead to any

transient or a change in core damage frequency. This licensee-identified finding

involved a violation of 10 CFR 50, Appendix R, Criterion III.G.2 requirements. The

enforcement considerations for this violation are given in Section 40A7. This issue is

closed.

.c

Pressurizer Level Instrumentation Affected by Tray-Conduit Interaction

The licensee identified that cable tray L2213, located in containment, lacked 20-foot

separation or a radiant heat shield to prevent interaction with conduits 23320D and

23090A during a fire as required by 10 CFR 50, Appendix R, Criterion III.G.2.

Pressurizer level instruments LT-11 1OX and LT-1105 were routed in cable tray L2213;

and level instruments LT-1 11 Y and LT-1 104 were routed in conduits 23320D and

23090A. Instruments LT-111OX & Y would be used to achieve and maintain SSD during

14

a fire. The inspector determined that level failing low was the most limiting effect of a

fire-induced fault with these cables. Low indicated pressurizer level would initiate

several automatic actions, some cause level to rise while others cause level to fall. The

loss of pressurizer heaters dominates the situation and causing actual pressurizer level

and pressure to decrease. Low pressurizer pressure would initiate a reactor trip and a

safety injection (SI) actuation. Safety injection flow would increase actual pressurizer

level. Because actual pressurizer level cannot be determined, the operator may not

secure the safety injection pumps resulting in the pressurizer completely filling. The

post-fire SSD procedure directs the operator to place the PORVs in override due to

concerns about spurious opening. Therefore, once the pressurizer completely fills, the

associated pressure increase would be relieved by the safety relief valves. This finding

is more than minor because it could adversely affect the equipment reliability objective

of the mitigating systems and barrier integrity cornerstones. Using Appendix F of the

SDP, the inspector determined that the finding was of very low safety significance

(Green) because the likelihood of a stuck open safety valve event occurring was very

low, manual suppression systems for fires in containment were in a normal state, and

cables for systems used to mitigate this event were located outside containment. This

licensee-identified finding involved a violation of 10 CFR 50, Appendix R, Criterion

III.G.2 requirements. The enforcement considerations for this violation are given in

Section 40A7. This issue is closed.

.d

Pressurizer Level Instrumentation Affected by Conduit to Conduit Interaction

The licensee identified that two conduits in containment, containing cables for redundant

channels of pressurizer level instrumentation, did not have 20-foot separation or radiant

heat shield protection as required by 10 CFR 50, Appendix R, Criterion III.G.2. The

conduits were located in the containment annulus at an elevation where there were no

ignition sources other than the cables themselves. The licensee corrected the

separation problem by installing a radiant heat shield on one of the conduits (plant

modification PCM99104, Supplement 1). The inspector determined that self-induced

cable ignition of low voltage, low energy, instrument circuits, was not a credible event.

The inspector noted even if a fire occurred within a conduit, the fire would not affect

another conduit. The inspector determined that, given the particular configuration at

issue, it could not credibly adversely affect any reactor safety cornerstone. No new

findings were identified in the inspector's review. This finding constitutes a violation of

minor significance that is not subject to enforcement action in accordance with Section

IV of the NRC's Enforcement Policy. The licensee documented the problem in CR 99-

1963, Rev. 2 and CR 00-0386. This issue is closed.

.e

Circuits Related to Automatic Pressurizer Pressure Control Affected by Conduit to

Conduit Interaction

The licensee identified that certain conduits in containment, containing cables for the

pressurizer PORV and the auxiliary spray isolation valves, did not have 20-foot

separation or radiant heat shield protection as required by 10 CFR 50, Appendix R,

Criterion III.G.2. The licensee's SSA considered these two systems to be separate and

15

independent, and would be used in the post-fire SSD procedures as diverse methods to

reduce RCS pressure when necessary. The conduits were located in the containment

annulus at an elevation where there were no ignition sources other than the cables

themselves. The licensee corrected the separation problem by installing a radiant heat

shield on one of the conduits (plant modification PCM99104, Supplement 2). The

inspector determined that a fire in one conduit could not credibly expand to affect other

nearby conduits. The inspector determined that, given the particular configuration at

issue, it could not credibly adversely affect any reactor safety cornerstone. No new

findings were identified in the inspector's review. This finding constitutes a violation of

minor significance that is not subject to enforcement action in accordance with Section

IV of the NRC's Enforcement Policy. The licensee documented the problem in CR 99-

1963, Rev. 2 and CR 00-0386. This issue is closed.

.f

Radiant Heat Shields Not Installed ner Accepted Appendix R Deviation

The licensee identified that radiant heat shields had not been installed directly below

four groups of cable trays, running above the 45-foot elevation grating inside

containment [in the space between the containment wall and the bioshield], as required

by a NRC-approved deviation in the Unit 2 Operating License. The licensee corrected

the problem by installing the missing radiant heat shields [plant modification

PCM01028]. Train B cables are in trays near the containment wall, and Train A cables

are in trays near the bioshield. At certain locations, these cable trays are only separated

by seven feet. This finding is more than minor because a fire could adversely affect the

equipment reliability objective of the mitigating systems cornerstone by damaging

redundant trains of SSD equipment. The finding was considered to have very low safety

significance (Green) using Appendix F of the SDP because:

Fire brigade capability for a fire in containment was not impaired.

In-situ ignition sources are negligible and transient ignition sources and combustibles

are not present during normal plant operation.

Only the top tray in each group contains power cables carrying sufficient energy

(480V) for IEEE 383 cables to self-ignite. These trays are solid metallic bottom and

cover-type trays. This construction inherently limits the spread of an internal tray

fire, and provides a shield limiting the radiant heat energy release. Most of these

power cables are not energized during normal plant operation.

A very similar configuration in the Unit 1 containment was analyzed by the licensee,

reviewed by the NRC in great detail, and found to be an acceptable configuration.

The Unit 1 study had a safety factor of at least two, which provides a margin to

account for geometry and other unknown differences between the two units.

This licensee-identified finding involved a violation of PSL Unit 2 Operating License

NPF-16, Condition 2.C.(20) and 10 CFR 50, Appendix R, Criterion III.G.2 requirements.

The enforcement aspects of the violation are discussed in Section 4AO7. This issue is

closed.

- --- ----

16

.02

(Closed) LER 50-335/00-04, Pressurizer Level Instrumentation Conduit Separation

Outside Appendix R Design Bases

The licensee identified that a Unit 1 cable tray in containment containing pressurizer

level instrumentation, lacked 20-foot separation or a radiant heat shield to prevent

interaction with conduit containing redundant pressurizer level instrumentation, as

required by 10 CFR 50, Appendix R, Criterion III.G.2. A fire in the cable tray could result

in damage to all pressurizer level instrumentation causing the pressurizer to completely

fill and causing the safety valves to lift. [This is essentially the same issue as discussed

for Unit 2 in Section 4OA3.01.c above.] This finding is more than minor because it could

adversely affect the equipment reliability objective of the mitigating systems and barrier

integrity cornerstones. Using Appendix F of the SDP, the inspector determined that the

finding was of very low safety significance (Green) because the likelihood of a stuck

open safety valve event occurring was very low, manual suppression systems for fires in

containment were in a normal state, and cables for systems used to mitigate this event

were located outside containment. This licensee-identified finding involved a violation of

10 CFR 50, Appendix R, Criterion IIl.G.2 requirements. The enforcement

considerations for this violation are given in Section 4OA7. This LER is closed.

4OA5 Other Activities

.01

(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 provide 20-foot separation or

radiant heat shield protection for the pressurizer PORV cables inside containment as

required by 10 CFR 50, Appendix R, Criterion III.G.2.

Description: During an NRC fire protection inspection (Inspection Report 50-335,

389/99-08, dated January 31, 2000), the inspectors identified that the PORV cables

inside containment were not protected from fire-induced cable-to-cable short circuits.

The licensee's SSA referred to a study which documented that spurious opening of the

PORV due a cable-to-cable short circuit was not credible. Because the study could not

be located at the time of the inspection, the inspectors initiated this URI to track the

issue. Subsequently, the licensee determined that either pressurizer PORV could

spuriously open as a result of fire-induced short circuits in a cable tray in containment.

In addition, cables for the associated PORV block valve were routed in the same cable

tray and could be damaged by the same fire. Cables for one PORV and its block valve

were routed in a tray near the containment wall. Cables for the other PORV and its

block valve were routed in a tray near the bioshield. The condition applied to both units.

The licensee reported this finding in LER 50-335, 389/00-01. The licensee corrected

this problem by installing new PORV cables using armored cable [plant modification

PCM00059 (Unit 1) and PCM99104, Rev. 4 (Unit 2)].

Analysis: The finding was a performance deficiency because it represented a violation of

10 CFR 50, Appendix R, Criterion IllI.G.2 requirements. It was considered greater than

17

minor because it affects the 'equipment performance" attribute of the mitigating systems

and barrier integrity cornerstones. Using Appendix F of the SDP, the inspector

determined that the finding was of very low safety significance (Green) because the

initiating event likelihood was relatively low, manual suppression of fires in the

containment was in the normal state, and other mitigating systems were unaffected

because their associated cables were outside of containment.

Enforcement: The licensee's Fire Protection Program commits to 10 CFR, Appendix

R, Criterion III.G. For noninerted containments, Criterion IlI.G.2.d. and IlI.G.2.f, state, in

part "one of the following fire protection means shall be provided:

Separation of cables and equipment and associated non-safety circuits of redundant

trains by a horizontal distance of more than 20-feet with no intervening combustibles

or fire hazards;

Separation of cables and equipment and associated non-safety circuits of redundant

trains by a noncombustible radiant energy shield."

Contrary to the above, the cabling for redundant trains of pressurizer PORVs, and their

associated block valves did not meet this requirement. Because the failure to protect

these cables is of very low safety significance, has been entered into the CAP (CR 00-

0386) and the problem has been corrected through a plant modification, this violation is

being treated as an NCV consistent with Section VL.A of the NRC Enforcement Policy.

This finding is identified as NCV 50-335,389/03-02-02, PORV Cables in Containment

Fail to Meet 10 CFR 50, Appendix R, Criterion III.G.2 Requirements.

4OA6 Meetings

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 not 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.

(1) On January 22, 2003, the licensee documented in CR 03-0153 that PSL relied on

manual operator actions outside the MCR for SSD in non-alternative shutdown fire

areas (i.e., areas designated as 10 CFR 50, Appendix R, Criterion III.G.2) and the

manual actions did not have prior NRC approval.

During this inspection, the team found that the licensee used manual operator

actions outside the MCR for a number of fire areas which the licensee had

designated as 10 CFR 50, Appendix R, Criterion III.G.2 areas. The team reviewed

the manual operator actions for the III.G.2 areas selected for this inspection (Fire

18

Area C and Fire Area I). Enforcement related to using manual operator actions for

Fire Area I are discussed in this report section. Findings related to using manual

operator actions for Fire Area C are discussed in Section 1 R05.02.b(2) of this

inspection report.

10 CFR 50, Appendix R, Criterion III.G.2, requires in part, that, where cables or

equipment, including associated non-safety circuits 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 are located within the same fire area outside of primary containment, one

of the following means of ensuring that one of the redundant trains is free of fire

damage shall be provided:

Separation of cables and equipment of redundant trains by a fire barrier having a

3-hour rating.

Separation of cables and equipment of redundant trains by a horizontal distance

of more than 20 feet with no intervening combustibles or fire hazards.

Enclosure of cable and equipment of one redundant train in a fire barrier having

a 1-hour rating.

Manual operator actions to respond to maloperations are not listed as an acceptable

method for satisfying this requirement.

Contrary to the above, the licensee did not provide adequate protection to ensure

that redundant trains of systems and equipment necessary to achieve and maintain

SSD were maintained free of fire damage in the event of a fire in Fire Area I. In lieu

of providing adequate physical protection, the licensee used manual operator

actions outside the MCR without obtaining prior NRC approval.

This finding was entered into the licensee's CAP as CR 03-0153. The licensee

performed a SSD manual operator action time line for Fire Area I to determine the

time requirements when the SSD functions were needed, and, if personnel and

procedural guidance were adequate to perform the actions. The licensee

determined that the assistance of one operator from the Unit 1 operating shift staff

was needed perform the actions required for Unit 2 SSD in the event of a fire in Fire

Area I. The team reviewed the time line, personnel needed versus Unit 1 TS shift

staffing requirements, and the SSD procedures. Based on this review, the team

determined that manual actions to mitigate potential maloperations were properly

addressed in the SSD procedures and the personnel needed to perform the manual

actions (including the use of an operator from Unit 1) for Fire Area I was reasonable.

The team assessed the manual operator actions used for this fire area against the

guidance provided in Enclosure 2 of NRC ROP Procedure 71111.05, Fire Protection,

dated March 6, 2003. The team determined that the manual actions were

reasonable and met the criteria in Enclosure 2 of Procedure 71111.05.

19

(2) 10 CFR 50, Appendix R, Criterion III.G.2, 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:

Separation of cables of redundant trains by a horizontal distance of more than

20-feet with no intervening combustibles or fire hazards; or

Separation of cables of redundant trains by a non-combustible radiant energy

shield.

Contrary to the above, since the requirement became effective, the required fire

protection was not provided for the following redundant cables:

-

Shutdown cooling valves V3652 and V3481 on Unit 2.

-

Pressurizer pressure instrumentation PT-1 107 and PT-I 108 on Unit 2

-

Pressurizer level instrumentation LT-1 11 OX and LT-1 11 OY on Units 1 & 2

Cables contained in cable trays L2223 (Train A) and L2224 (Train B)

These findings have been entered into the licensee's 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 4AO3.2.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

D. Albritton, Senior Reactor Operator

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

R. Lamb, Senior Reactor Operator

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

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

ODened

50-389/03-02-01

50-389/03-02-02

50-335,389/03-02-03

URI

Failure to Provide Adequate Protection for Redundant

Safe Shutdown Equipment and Cables in the Event of a

Fire the Unit 2 Train B Switchgear Room (Fire Area C)

(Section 1 R05.02.b.1)

URI

Failure to Provide Physical Protection for Redundant Safe

Shutdown Equipment and Cables in the Event of a Fire in

the Unit 2 Train B Switchgear Room (Fire Area C)

(Section 1 R05.02.b.2)

NCV

PORV Cables in Containment Fail to Meet 10 CFR 50,

Appendix R, Criterion III.G.2 Requirements (Section

40A5)

Closed

50-335,389/99-08-03

50-335,389/00-001

50-335/00-004

50-335,389/03-02-03

URI

PORV Cabling May Not be Protected from Hot-Shorts

Inside Containment (Section 40A5.01)

LER

Outside Design Bases Appendix R Hi-Lo Pressure

Interface and Separation Issues (Section 40A3.01)

LER

Pressurizer Level Instrumentation Conduit Separation

Outside Appendix R Design Bases (Section 40A3.02)

NCV

PORV Cables in Containment Fail to Meet 10 CFR 50,

Appendix R, Criterion III.G.2 Requirements (Section

40A5)

Discussed

None

LIST OF DOCUMENTS REVIEWED

Section 1R05: Fire Protection

Procedures:

2-ADM-03.01, Unit 2 Power Distribution Breaker List, Rev. 6C

2-ONP-100.01, Response to Fire, Rev.9

2-ONP-100.02, Control Room Inaccessibility, Rev.13B

2-M-0018D, Mechanical Maintenance Safety-Related PM Program (Dampers), Rev. 11

Administrative Procedure 0005729, Fire Protection Training, Qualification, and Requalification,

Rev. 17

Administrative Procedure 0010239, Fire Protection System Impairment, Rev. 13B

Administrative Procedure 0010434, Plant Fire Protection Guidelines, Rev. 37C

Administrative Procedure 1800022, Fire Protection Plan, Rev. 35

EMP 50.10, Self-Contained Emergency Lighting Unit Maintenance and Inspection, Rev. 9

EMP 52.01, Periodic Maintenance of 4160 Volt Switchgear, Rev. 14

General Maintenance Procedure 2-M-0018F, Safety-Related PM Program (Fire PMs), Rev. 25B

Protection Services Guidelines, PSG-15.01, Monitoring Fire Protection System Failures, Rev. 0

QI-3-PSL-1, Design Control, Rev. 11

0-OSP-1 5.11, Fire Protection System Quarterly Alignment Verification, Rev. 6

0-OSP-1 5.17, Fire Protection System Triennial Flow Test, Rev. 1

2-OSP-100.16, Remote Shutdown Components 18 Month Functional Test, Rev. 2

2-IMP-69.02, ESFAS Monthly Channel Functional Test, Rev. 4A

Drawings:

2

2998-G-078, Sheets 107, 108, 109, 110, Unit 2 Reactor Coolant System, Rev. 1

2998-G-078, Sheets 121A, 121B & 122, Unit 2 Chemical and Volume Control System, Rev. 16

2998-G-078, Sheets 130A, 130B, 131, 132, Unit 2 Safety Injection System, Rev. 12

2998-G-079, Sheets 1, 2 & 7, Unit 2 Main Steam System, Rev. 20

2998-G-080, Sheets 2A & 2B, Unit 2 Feedwater and Condensate System, Rev. 25

2998-G-082, Sheets 1 & 2, Unit 2 Circulating and Intake Cooling Water System, Rev. 37

2998-G-083, Sheets 1 & 2, Unit 2 Component Cooling Water System, Rev. 28

2998-G-084, Unit 2 Flow Diagram Domestic & Make-up Water Systems, Rev. 33

2998-G-088, Sheet 1, Unit 2 Containment Spray and Refueling Water System, Rev. 35

2988-G-275 series, 480 V. Switchgear One Line Wiring Diagrams, Rev. 4

2988-G-424, Reactor Auxiliary Building Fire Detectors and Emergency Lights, Rev 9.

2988-G-890, Reactor Auxiliary Building Plumbing and Drainage Plan, Rev. 8

2988-G-891, Reactor Auxiliary Building Plumbing and Drainage Plan El. 43', Rev. 10

2998-B-733, Unit 2 Fire Protection Penetration Schedule, Rev. 6

2998-G-785, Reactor Auxiliary Building Room and Door Schedule, Rev. 8

2998-G-882, HVAC Equipment Schedule and Details, Rev. 1

2998-16082, Air Balance Inc. SL-2121 List of Materials, 319 ALV & 319 ALH Fire Dampers,

Rev. 0

8770-B-327, Control Wiring Diagrams for Fire Water Pumps, Rev. 14

2998-G-424, Sheet 7, Unit 2 Reactor Containment Fire Detectors and Emergency Lights, Rev 1

2998-G-879, Sheets 1 & 2, Unit 2 HVAC Flow and Control Diagrams, dated 10/20/89

2998-G-411, Reactor Auxiliary Building El' 19'50 Conduit Layout, Sheet 14, Rev. 8

2998-G-41 1, Reactor Auxiliary Building El' 19'50 Conduit Layout, Sheet 15, Rev. 6

2998-G-41 1, Reactor Auxiliary Building El' 19'50 Conduit Layout, Sheet 19, Rev. 5

2998-G-41 1, Reactor Auxiliary Building El' 19'50 Conduit Layout, Sheet 10, Rev. 6

2998-G-41 1, Reactor Auxiliary Building El' 19'50 Conduit Layout, Sheet 4, Rev. 5

2998-G-41 1, Reactor Auxiliary Building El' 19'50 Conduit Layout, Sheet 3, Rev. 6

2998-G-41 1, Reactor Auxiliary Building El' 19'50 Conduit Layout, Sheet 13, Rev. 5

2998-G-41 1, Reactor Auxiliary Building El' 19'50 Conduit Layout, Sheet 7, Rev. 9

2998-G-41 1, Reactor Auxiliary Building El' 19'50 Conduit Layout, Sheet 8, Rev. 8

2998-G-41 1, Reactor Auxiliary Building El' 19'50 Conduit Layout, Sheet 9, Rev. 8

2998-G-41 1, Reactor Auxiliary Building Electrical Pen Area Conduit Layout, Sheet 20, Rev. 9

2998-G-410, Cable Vault Trays - Key Plan, Sheet 6, Rev. 6

2998-G-394, Reactor Auxiliary Building El' 43'0 Conduit, Trays & Grounding, Sheet 1, Rev. 27

2998-G-392, Reactor Auxiliary Building El' 19'6 Conduit, Trays & Grounding, Sheet 1, Rev. 17

2998-G-374, Reactor Auxiliary Building Pen Area Conduit, Trays & Grounding, Sheet 1, Rev.

11

2998-G-076, Reactor Auxiliary Building Misc. Plans & Sections, Rev. 19

2998-G-071, General Arrangement Reactor Auxiliary Building Plan Sheet 3, Rev. 24

2998-G-272A, Combined Main and Auxiliary One Line Diagrams, Rev. 7

2998-B-327, Pressurizer Relief Isolation Valve V-1477, Sheet 118, Rev. 14

2998-B-327, Pressurizer Relief Isolation Valve V-1476, Sheet 120, Rev. 14

2998-B-327, LPSI Pump 2A Suction Valve V-3444, Sheet 1531, Rev. 6

2998-B-327, LPSI Flow Control Valve HCV-3625, Sheet 260, Rev. 16

2998-B-327, Pressurizer Relief Valve V-1475, Sheet 1630, Rev. 10

2998-B-327, Pressurizer Relief Valve V-1474, Sheet 1624, Rev. 10

3

2998-B-327, Pressurizer Level Channel L-1 110, Sheet 139, Rev. 13

2998-B-400, Lighting Panel Details, Sheet 209, Rev. 8

2998-B-325, Bill of Material, Sheet 026-01, Rev. 5

2998-B-327, Steam Generator 2A/2B Pressure & Level, Sheet 369, Rev. 12

2998-B-327, Pressurizer Pressure & Level , Sheet 370, Rev. 12

2998-B-327, Measurement Channels F2212, P2212, P2215, T2229, T2221, Sheet 150, Rev. 15

C-13172-412-522, Process Instruments Remote Nest Interconnection Diagram, Sheet 1, Rev. 3

C-13172-412-523, Process Instruments Remote Nest Interconnection Diagram, Sheet 1, Rev. 2

Calculations and Evaluations

2998-B-048, St. Lucie Unit 2, Appendix R Safe Shutdown Analysis Fire Area Report

2998-B-049, St. Lucie Unit 2 Essential Equipment List, Rev. 6

2998-2-FJE-98-002, Review of Circuit Breaker and Fuse Coordination for St. Lucie Unit 2

Appendix R Essential Equipment List Circuits, Rev. 0

PSL-2-FJE-90-0020, St. Lucie Unit 2 2A & 2B EDG Electrical Loads, Rev. 7

PSL-IFJM-91-001, PSL-1 RAB Electrical Equipment Rooms HVAC Computer Model Data

Inputs and Outputs, Rev. 1

PSL-FPER-00-004, Disposition of Unit 2 Fire Detection System Nonconformance, Rev. 1

PSL-BFSM-98-004, Hose Station Supply Piping (Standpipe) Hydraulic Analysis, Rev. 0

PSL-ENG-97-070, UFSAR Combustible Loading Update for Unit 2, Rev. 0

PSL-FPER-99-008, Two-sided Cable Tray Fire Stop Redesign, Rev. 1

PSL-FPER-99-01 1, Disposition of Unit 2 NFPA Code Nonconformance, Rev. 1

PSL-FPER-00-0126, Evaluation of Fire Barrier Rating for Barriers Containing Two-sided Fire

Stops, Rev.0

Calculation to determine the capacity of diked areas surrounding Unit 2 transformers 2A5, 2B5

and 2B2, dated March 12, 2003

Evaluation to determine compliance with DC 561 Technical Manual 'Use Restrictions" for Unit 2

transformers 2A5, 2B5 and 2B2, dated March 10, 2003

Desiqn Basis Documents:

Component Functions for Pressurizer Wide Range Pressure Instrument Loop, Section 7.22

Component Functions for Pressurizer Instrument Loop P-11OOX&Y, Section 7.23

Component Functions for Pressurizer Pressure/Safety Injection Instrument Loop, Section 7.28

DBD-ESF-2, Engineering Safety Features Actuation System, Rev. 1

DBD-CVCS-2, Chemical and Volume Control System, Rev. 1

Applicable Codes and Standards:

IEEE Standard 100, Standard Dictionary of Electrical and Electronics Terms, Fourth Edition

NFPA 13, Standard for the Installation of Sprinkler Systems, 1973 Edition

NFPA 14, Standard for the Installation of Standpipe and Hose Systems, 1973 Edition

NFPA 20, Standard for the Installation of Centrifugal Fire Pumps, 1972 Edition

NFPA 72A, Standard on Local Protective Signaling Systems, 1972 Edition

4

NFPA 72D, Standard for the Installation, Maintenance, and Use of Proprietary Protection

Signaling Systems, 1973 Edition

NFPA 80, Standard on Fire Doors and Windows, 1973 Edition

NFPA 90A, Standard on Air Conditioning and Ventilating Systems, 1981 Edition

NUREG-1552, Supplement 1, Fire Barrier Penetration Seals in Nuclear Power Plants, dated

January 1999

Underwriters Laboratories, Fire Resistance Directory, January 1998

OSHA Standard 29 CFR 1910, Occupational Safety and Health Standards,

Audit Reports:

QSL-FP-00-07, Annual Fire Protection Functional Area Audit

QSL-FP-01-07, Triennial Fire Protection Functional Audit

QSL-FP-02-05, Fire Protection Functional Audit

Condition Reports:

CR 98-0260, Evaluate Deviations from NFPA 72 Code

CR 98-0405, Evaluate Deviations from NFPA 13-1975 Code

CR 98-0563, Assess Currently Installed Fire Hose Nozzles in Both Units

CR 00-1514, Failure of 500KV Main Transformer, SEN 215

CR 01-0577, Circuit Breaker Failure and Fire, SEN 218

CR 01-2296, Assess Deviations from NFPA 72 Code addressed in QA Audit QSL-FP-01-07

CR 01-2459, 4-kV Breaker Failure, SER 5-01

CR 02-0396, Assess Qualifications of Thermo-Lag Walls at PSL

CR 02-1619, Potential Problems with Heat Collectors, NRC Information Notice 2002-24

CR 02-2081, Design Change Checklist

CR 02-2098, PSL CARS

CR 02-3145, Failure to Obtain FRG Review of Several Procedure Changes

Condition Reports Generated During this Inspection

CR 03-0153, Use of manual actions in Appendix R, lll.G.2 areas without prior NRC approval

CR 03-0637, Silicone oil-filled transformers installed in Unit 2 interior rooms

CR 03-0847, Hot shutdown repairs using tools to achieve safe shutdown in the event of a fire

CR 03-0888, Update UFSAR to show previously approved Deviation C6 no longer required

CR 03-0942, Discrepancies between the SSA, EEL, and the breaker/fuse coordination study

CR 03-0964, Rubatex insulation installed in U2 intake (fire area R-R) not considered in the FHA

CR 03-0965, Combustible fire load for Ul and U2 intake fire areas different for each unit's FHA

CR 03-0966, Temp Mod did not sufficiently evaluate potential impact on fire protection

CR 03-0978, Transformers' oil not being sampled and tested in accordance with vendor manual

CR 03-0986, Discrepancies between SSA and EEL, determined that EEL was in error

CR 03-1010, Discrepancy between UFSAR and procedure regarding cold shutdown repairs

Work Orders/Job Tasks

5

WO 3201713801, T.S. 044A SIG 2A Level Loop Calibration, dated 1/7/03

WO 3100661301, T.S. 044A SIG 2A Level Loop Calibration, dated 8/8/01

WO 3101259101, T.S. 044B SIG 2B Level Loop Calibration, dated 11/03/01

WO 3181734101, T.S. F-2212 Charging Pump Flow Calibration, dated 4/24/02

WO 3101222101, T.S. Charging Pump Discharge P-2212 Calibration, dated 9/7/01

WO 3201736501, T.S. Pressurizer Level (P1107/1108/1116) Calibration, dated 11/10/03

WO 3100693301, T.S. Pressurizer Level (P1107/1108/1116) Calibration, dated 7/12/01

WO 3261652901, T.S. Pressurizer & Quench Tank Level (L1 103/4/5/1116) Calibration, dated

1/10/03

WO 3100682601, T.S. Pressurizer & Quench Tank Level (L1 103/4/5/11) Calibration, dated

7/11/01

Technical Manuals/vendor Information

Dow Corning 561 Silicone Transformer Liquid, Material Safety Data Sheet 01496247, 1/27/97

Dow Corning 561 Silicone Transformer Fluid Technical Manual,10-453-97, 1997

Data Sheet Issue C Duraspeed, Automatic Sprinklers, Grinnell Sprinkler Corporation

Data Sheet Model F950, Upright and Pendent Sprinklers, Grinnell Sprinkler Corporation

Data Sheet Model L-205-EB, Industrial Electrical Non-Shock Fog Nozzles, Elkhart Brass

Manufacturing Co. Inc.

IB-PD-1001, Gould Inc. l-T-E Unit Substation Transformers Instruction Manual

S2000, Protecto-wire Fire Systems Fire System 2000 Fire Alarm Control Panel, Rev. 1998

Sheet 5-4/14-8, Factory Mutual Research Approval Guide-Transformer Fluids

Miscellaneous

0711206, Reactor Operator Lesson Pressurizer Pressure and Level Control, Rev.12

1/M-CE 917 Foxboro Specification 200 Control System Manual # 79N-36291, dated 8/20/98

Consumer Product Safety Commission (CPSC) Recall Alert, Invensys Building Systems Recall

of Siebe Actuators in Building Fire/Smoke Dampers, dated October 2, 2002

Ebasco Specification - Electric Cables, Project 10 # FLO 298.292, dated 10/28/77

IPEEE Submittal for St. Lucie Units 1 and 2, Rev. 0, dated December 15, 1994

Fire Brigade Drill Training Reports for operating shifts, August 2001- February 2003

Letter from Ebasco to Florida Power and Light, on the subject of UL Qualification Test for

Pullman Industries Internal Expansion Damper Assembly, dated April 16, 1986

NRC Supplemental Safety Evaluation Report SSER 3, for St. Unit 2

PC/M 174-295M, Reroute of Cable 21702C, Rev. 1

Pre-fire Strategy No. 4, A Switchgear Room, Fire Area A, Rev. 23

Pre-fire Strategy No. 6, Cable Spread Room, Fire Area B, Rev. 23

Pre-fire Strategy No. 7, B Switchgear Room, Fire Area C, Rev. 23

Pre-fire Strategy No. 8, Electrical Equipment Supply Fan Room, Fire Area C, Rev. 23

Pre-fire Strategy No. 25, Personnel Monitoring Area & Health Physics Area, Fire Area I, Rev. 23

Pre-fire Strategy No. 26, Electrical Penetration Room B, Fire Area I, Rev. 23

Pre-fire Strategy No. 57, Turbine Building, Fire Area QQ, Rev. 23

Technical Specifications, St. Lucie Unit 2, LCO 3.3.3.5

Technical Specifications, St. Lucie Unit 2, SR 4.3.3.5.1 I 4.3.3.5.2

6

UFSAR Section 8, Electrical Power

UFSAR Appendix 9.5A, Fire Protection Program Report

Underwriters Laboratories, Report File R4708, Fire Test of 3HR Curtain Type Fire Damper

Utilizing an Alternate Method of Installation, Air Balance, Inc., dated December 5, 1984