ML050380428
| ML050380428 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 02/03/2005 |
| From: | Ogle C NRC/RGN-II/DRP/RPB1 |
| To: | Scarola J Carolina Power & Light Co |
| References | |
| EA-00-022, EA-01-310, FOIA/PA-2004-0277 IR-03-007 | |
| Download: ML050380428 (17) | |
See also: IR 05000400/2003007
Text
EA-01 -310
Carolina Power & Light Company
ATTN: Mr. James Scarola
Vice President - Harris Plant
Shearon Harris Nuclear Power Plant
P. 0. Box 165, Mail Code: Zone 1
New Hill, North Carolina 27562-0165
SUBJECT:
SHEARON HARRIS NUCLEAR POWER PLANT - NRC FIRE PROTECTION
INSPECTION REPORT NO. 05000400/2003007
Dear Mr. Scarola:
On _
_, 2003, the U.S. Nuclear Regulatory Commission (NRC) completed an in-office
review of the significance of the triennial fire protection inspection findings of inspection report
05000400/2002011 related to your Shearon Harris Nuclear Power Plant. The enclosed
inspection report documents the results of our significance determination, which was discussed
on
,2003, by telephone with Mr.
and other members of your staff.
This report documents two NRC-identified findings of very low significance (Green). Both of
these findings were determined to involve violations of NRC requirements. However, because
of the very low safety significance and because they are entered into your corrective action
program, the NRC is treating these two findings as non-cited violations (NCVs) consistent with
Section VI.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 the Shearon Harris Nuclear Power Plant.
In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter, 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 Website at
httD://www.nrc.aov/readina-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
Charles R. Ogle, Chief
Engineering Branch 1
Division of Reactor Safety
Docket No.: 50-400
2
License No.: NPF-63
Enclosure: Inspection Report 05000400/200307
cc w/encl:
James W. Holt, Manager
Performance Evaluation and
Regulatory Affairs
CPB 9
Carolina Power & Light Company
Electronic Mail Distribution
Robert J. Duncan II
Director of Site Operations
Carolina Power & Light Company
Shearon Harris Nuclear Power Plant
Electronic Mail Distribution
Benjamin C. Waldrep
Plant General Manager--Harris Plant
Carolina Power & Light Company
Shearon Harris Nuclear Power Plant
Electronic Mail Distribution
Terry C. Morton, Manager
Support Services
Carolina Power & Light Company
Shearon Harris Nuclear Power Plant
Electronic Mail Distribution
(cc w/encl cont'd - See page 3)
(cc w/encl cont'd)
John R. Caves, Supervisor
Licensing/Regulatory Programs
Carolina Power & Light Company
Shearon Harris Nuclear Power Plant
Electronic Mail Distribution
William D. Johnson
Vice President & Corporate Secretary
Carolina Power & Light Company
Electronic Mail Distribution
John H. O'Neill, Jr.
Shaw, Pittman, Potts & Trowbridge
2300 N. Street, NW
Washington, DC 20037-1128
Beverly Hall, Acting Director
,I
3
Division of Radiation Protection
N. C. Department of Environmental
Commerce & Natural Resources
Electronic Mail Distribution
Peggy Force
Assistant Attorney General
State of North Carolina
Electronic Mail Distribution
Public Service Commission
State of South Carolina
P.O. Box 11649
Columbia, SC 29211
Chairman of the North Carolina
Utilities Commission
P. O. Box 29510
Raleigh, NC 27626-0510
Robert P. Gruber
Executive Director
Public Staff NCUC
4326 Mail Service Center
Raleigh, NC 27699-4326
(cc w/encl cont'd - See page 4)
(cc w/encl cont'd)
Linda Coleman, Chairman
Board of County Commissioners
of Wake County
P. O. Box 550
Raleigh, NC 27602
Gary Phillips, Chairman
Board of County Commissioners
of Chatham County
Electronic Mail Distribution
Distribution w/encl:
L. Slack, EICS
C. Patel, NRR
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U.S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket No.:
License No.:
Report No.:
Licensee:
Facility:
Location:
50-400
Carolina Power & Light (CP&L)
Shearon Harris Nuclear Power Plant
5413 Shearon Harris Road
New Hill, NC 27562
Dates:
Inspectors:
Approved by:
W. Rogers, Senior Reactor Analyst, Region II
R. Schin, Senior Reactor Inspector, Region II
Charles R. Ogle, Chief
Engineering Branch 1
Division of Reactor Safety
Enclosure
SUMMARY OF FINDINGS
IR 05000400/2003-007; .J.J2003 - .J.J2003; Shearon Harris Nuclear Power Plant; Fire
Protection.
The in-office review was conducted by a regional inspector, a regional senior reactor analyst,
and NRC Headquarters Senior Reactor Analysts. Two Green findings, each a Non-Cited
Violation (NCV), were identified. The significance of issues is indicated by their color (Green,
White, Yellow, Red) using IMC 0609 "Significance Determination Process" (SDP). Findings for
which the SDP does not apply may be uGreen" 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-1 649, "Reactor Oversight Process," Revision 3,
dated July 2000.
A.
InsDector Identified Findings
Cornerstones: Mitigating Systems and Initiating Events
Green. An NCV of Operating License Condition 2.F, the Fire Protection Program, and
Technical Specification (TS) 6.8.1, Procedures and Programs, was identified for
inadequate implementation of the fire protection program. Physical and procedural
protection for equipment that was relied on for safe shutdown (SSD) during a fire in fire
safe shutdown analysis (SSA) areas 1 -A-BAL-B-B1, 1 -A-BAL-B-B2, 1 -A-BAL-B-B3, 1-A-
BAL-B-B4, 1-A-EPA, and 1 -A-BAL-C of the reactor auxiliary building was inadequate.
Consequently, a fire in one of these SSA areas could result in a reactor coolant pump
seal loss of coolant accident (LOCA), a main steam line break (MSLB) event, a loss of
high pressure safety injection, and/or a loss of component cooling water to the reactor
coolant pump seals. The licensee has initiated corrective actions including assigning an
additional operator to be available to perform post-fire safe shutdown actions and
performing a complete review of the safe shutdown analysis and related operating
procedures.
This finding was greater than minor because it could initiate a LOCA or MSLB event and
could result in a loss of equipment that was relied upon for SSD from a fire. The finding
was of very low significance (Green) because of the low fire ignition frequencies and
lack of combustible materials in critical locations and because of the effectiveness of the
fire protection features and the remaining SSD equipment to mitigate a fire in each of
the affected fire zones/areas. (Section 1 R05.03.b.1)
Green. An NCV of Operating License Condition 2.F, the Fire Protection Program, and
Technical Specification (TS) 6.8.1, Procedures and Programs, was identified for
inadequate corrective action for previous Violation 50-400/02-08-01. Physical and
procedural protection for equipment that was relied on for safe shutdown (SSD) during a
fire in the new auxiliary control panel fire area 1-A-ACP was inadequate. Consequently,
a fire in area 1-A-ACP could result in a loss of auxiliary feedwater and a main steam line
break (MSLB) event. The licensee has initiated corrective actions including assigning
an additional operator to be available to perform post-fire safe shutdown actions and
performing a complete review of the safe shutdown analysis and related operating
procedures.
This finding was greater than minor because it could initiate a MSLB event and could
result in a loss of equipment that was relied upon for SSD from a fire. The finding was
of very low significance (Green) because of the low fire ignition frequencies and lack of
combustible materials and because of the effectiveness of the fire protection features
and the remaining SSD equipment to mitigate a fire in area 1 -A-ACP. (Section
1 R05.03.b.1)
B.
Licensee-Identified Violations
None
REPORT DETAILS
1.
REACTOR SAFETY
Cornerstones: Initiating Events and Mitigating Systems
1R05 FIRE PROTECTION
.01
Significance Determination for Triennial Fire Protection Inspection Findings
a.
Inspection Scope
In inspection report 50-400/02-11, nine findings had been identified as unresolved items
(URls) pending completion of the NRC significance determination process (SDP). This
inspection report documents the results of the in-office completion of the NRC SDP with
respect to those nine URIs. The significance determination was accomplished as
described in NRC Inspection Manual Chapter (IMC) 0609, Signification Determination
Process; IMC 0609A, Significance Determination of Reactor Inspection Findings for At-
Power Situations; and IMC 0609F, Determining Potential Risk Significance of Fire
Protection and Post-Fire Safe Shutdown Inspection Findings. This involved evaluating
the significance of a potential fire in each of the seven affected fire safe shutdown
analysis (SSA) areas, considering all examples of the findings that could be involved in
each fire.
In addition, the performance deficiencies which could result in the loss of a safety
function were evaluated by the Office of Nuclear Reactor Regulation (NRR) using the
Phase IlIl portion of the SDP. Inclusive in this evaluation were extensive walkdowns of
the applicable fire SSA areas by two contractors to observe ignition sources and
possible fire propagation from these ignition sources that could affect the unprotected
cables of concern. Also, electrical circuit drawings and the latest information on cable
hot short failure mechanisms and probabilities were used to develop cable failure
probabilities that could cause a loss of function for the unprotected cables of concern.
b.
Findings
(1)
Inadeguate Implementation of the Fire Protection Program
Introduction: An overall finding was identified in that the implementation of the fire
protection program was inadequate. Eight of the nine URis described in IR 50-400/02-
11 were considered to include examples of this overall finding. Based on evaluating the
eight examples of this overall finding for their effects during a fire that could occur in
each of the six affected fire SSA areas, this overall finding was determined to have a
very low significance (Green).
Description: The licensee's implementation of the fire protection program for ensuring
the ability to safely shut down the plant during a fire was inadequate, in that:
The fire safe shutdown analysis (SSA) failed to identify some cables that were
relied upon for safe shutdown (SSD) during a fire. Consequently, those cables
2
were not provided with the required protection from fire damage. A fire could
cause hot shorts in the cables which would result in maloperation of equipment
that was relied upon for SSD during that fire.
The SSA identified many cables that were relied upon for SSD during a fire, but
the licensee generally failed to provide the required physical protection from fire
damage. Instead, the SSA designated that operator actions would be taken to
prevent or mitigate the effects of the fire damage. However, the licensee did not
obtain NRC approval for these deviations from the approved fire protection
program.
Some of the operator actions that were designated by the SSA were not
incorporated into operating procedures for SSD. Also, the operator actions in
procedures differed in many respects from the operator actions that were
analyzed in the SSA. For example, the operating procedures directed operators
to use some different flowpaths than those analyzed in the SSA.
Some operator actions in the SSD procedures would not work. They were too
challenging, involved entering the area of the fire, were not adequately analyzed,
or were too numerous for the available SSD non-licensed operator to perform.
The eight examples of this overall finding that are described in IR 50-400/02-11 include:
Physical and procedural protection for equipment that was relied on for safe
shutdown (SSD) during a fire in safe shutdown analysis (SSA) areas 1-A-BAL-
B1,
1 -A-BAL-B2, and 1-A-EPA of the reactor auxiliary building was inadequate.
Motor-operated valve 1 CS-1 65, volume control tank outlet to charging/safety
injection pumps was not protected physically or procedurally from maloperation
due to a fire. Also, per the SSA, component cooling (CC) to the RCPs was not
protected from the fire; and the control cable for 1 CC-207, CC to RCP seals, was
in the same cable tray with the control cable for 1 CS-165. Consequently, a fire
in one of the three SSA areas could result in a reactor coolant pump seal loss of
coolant accident (LOCA) with no high pressure safety injection available.
Physical and procedural protection for equipment that was relied on for SSD
during a fire in SSA area 1 -A-BAL-B-B5 of the reactor auxiliary building was
inadequate. Motor-operated valves 1 CS-1 69, charging/safety injection pump
(CSIP) suction cross-connect; 1 CS-214, CSIP mini-flow isolation; 1 CS-21 8,
CSIP discharge cross-connect; and 1 CS-219, CSIP discharge cross-connect;
were not protected physically or procedurally from maloperation due to a fire.
Consequently, a fire in SSA area 1 -A-BAL-B-B5 could result in a loss of all
charging and high pressure safety injection.
Physical and procedural protection for equipment that was relied on for SSD
during a fire in SSA area 1-A-BAL-B-B4 of the reactor auxiliary building was
inadequate. Motor operated valves 1 CS-1 66, volume control tank outlet to
CSI Ps; and 1 CS-1 68, CSIP suction cross-connect; were not protected physically
or procedurally from maloperation due to a fire. Consequently, a fire in SSA
3
area 1 -A-BAL-B-B4 could result in a loss of all charging and high pressure safety
injection.
Physical and procedural protection for equipment that was relied on for SSD
during a fire in SSA area 1-A-BAL-C of the reactor auxiliary building was
inadequate. Motor operated valves 1 CC-208, component cooling water (CC)
supply to reactor coolant pump (RCP) seals; and 1 CC-251, CC return from RCP
seals; were not protected physically or procedurally from maloperation due to a
fire. The SSA did not analyze or credit RCP seal injection but subsequent
analysis after the inspection determined that RCP seal injection function would
not be affected by a fire in this area. Consequently, a fire in SSA area 1-A-BAL-
C could potentially result in only a loss of component cooling to the RCP seals.
Many local manual operator actions were used in place of the required physical
protection of cables for equipment relied on for SSD during a fire, without
obtaining NRC approval for these deviations from the approved fire protection
program. This condition applied to the six inspected SSA areas that are listed
below. This reliance on large numbers of local manual actions, in place of the
required physical protection of cables, could potentially result in an increased risk
of loss of equipment that was relied upon for SSD from a fire.
The procedure for SSD from a fire was inadequate. For a fire in certain safe
shutdown analysis areas of the reactor auxiliary building, there were too many
SSD procedure contingency actions to respond to potential spurious actuations
for the one designated SSD non-licensed operator to perform. Consequently,
equipment that was relied on for SSD may not be available.
The procedure for SSD from a fire was inadequate. For a fire in safe shutdown
analysis areas near the boric acid tank (BAT) in the reactor auxiliary building, the
SSD procedure directed operators to take CSIP suction from the BAT even if
BAT level indication were lost. However, the charging volume needed for reactor
coolant system cooldown would have emptied the BAT and damaged the CSIP.
Consequently, a loss of charging/high pressure safety injection could have
resulted.
Required battery-backed emergency lights were not provided in locations where
operators were required to perform actions for SSD from a fire. This condition
affected SSD during fires in all of the areas inspected in the reactor auxiliary
building. The lack of required lighting could result in an increased risk of
operators failing to perform the SSD actions in a timely and accurate manner.
The inspectors evaluated the effects of the eight examples of this overall finding during
a fire that could occur in each of the following six affected fire SSA areas of the reactor
auxiliary building (RAB):
Fire SSA area 1-A-BAL-B-B1, on the 261 foot level of the RAB, including the 'A'
chiller and motor-driven auxiliary feedwater (AFW) pump flow control valves
(FCVs).
4.
Fire SSA area 1 -A-BAL-B-B2, on the 261 foot level of the RAB, including the 'B'
chiller and turbine-driven AFW pump FCVs.
Fire area 1-A-EPA, on the 261 foot level of the RAB, including the electrical
penetration room 'A'.
Fire SSA area 1-A-BAL-B-B4, on the 261 foot level of the RAB, including 480V
motor control center (MCC) 1 B35-SB.
Fire SSA area 1-A-BAL-B-B5, on the 261 foot level of the RAB, including 480V
Fire area 1 -A-BAL-C, on the 286 foot level of the RAB, including 480V MCC
1 B31 -SB.
The inspectors and analysts assessed that the in situ and allowed transient
combustibles were sufficient to support a credible fire scenario in each fire SSA area.
During this analysis, the inspectors considered the following combustibles:
The 'B' chiller, which was located below certain cable trays of concern, contained
an ignition source of a compressor with 12 gallons of lubricating oil (containing
approximately 1.7 million BTU). The 'B' chiller also included flammable thermal
insulation on its piping. This 12 gallons of oil and flammable thermal insulation
were apparently not included in the licensee's IPEEE fire analysis.
The transient combustible control program allowed transient combustibles, up to
one million BTU above the analyzed combustible loading, to be temporarily
stored in any fire zones without a fire watch or other compensatory measures.
The transient combustible control program also allowed rubber mats up to 150
square feet (containing approximately 1.8 million BTU) to be installed without
continuous attendance.
The transient combustible control program included staging areas for 55 gallon
drums of oil (containing approximately 7.7 million BTU each) in or near three of
the fire SSA areas of concern. The program also allowed up to five gallons of
transient combustible liquids (containing approximately 0.7 million BTU) and two
gallons of transient flammable liquids (containing approximately 0.27 million
BTU) with no transient combustible permit required.
Inspector analysis indicated that, for the smaller enclosed fire SSA areas of concern
(e.g., 1-A-EPA, 1-A-BAL-C), a fire could start in any part of the room and cause a hot
gas layer that could ignite the IEEE 383 cables near the ceiling throughout the room.
For the larger areas (e.g., 1 -A-BAL-B-B1, 1 -A-BAL-B-82), a fire was unlikely to cause a
hot gas layer that could ignite all of the IEEE 383 cables in the area; however, a fire
starting below certain cables could generate a fire plume sufficiently large and hot to
ignite the cables that were in the plume.
5
The inspectors and analysts assessed that many fire protection features and event
mitigation functions would not be affected by a fire in the SSA areas of concern.
Examples included:
Based on no identified degradation of the automatic sprinklers and fire brigade
for each of the affected fire areas/zones, full credit was given for the
effectiveness of those fire protection features.
Based on the types of RCP seals installed, an RCP seal LOCA was judged to
result in a small break LOCA and not a larger LOCA.
For the fire areas/zones where a fire could cause an RCP seal LOCA, at least
one train of the following mitigation functions would be unaffected by the fire:
high pressure injection (HPI), including a pump and flowpath from the RWST to
the RCS; primary feed and bleed (FB), including a pressurizer power-operated
relief valve; low pressure injection (LPI), including an RHR pump and a flowpath
from the RWST to the RCS; low pressure recirculation (LPR), including a suction
path from the containment sump; high pressure recirculation (HPR), including a
flowpath from the RHR pump to the HPI pump; and auxiliary feedwater (AFW).
In addition, both trains of emergency power would be unaffected.
Also, for the fire areas/zones where a fire could cause an RCP seal LOCA, the
abnormal operating procedures would provide adequate direction, after the loss
of a CSIP, for operators to recover charging by aligning suction from the RWST
and starting the standby CSIP.
For a fire in SSA fire areas 1-A-BAL-B-B4 and 1-A-BAL-B-B5, auxiliary feedwater
would not be affected. Also, for a fire in fire area 1-A-BAL-C, CSIPs and RCP
seal injection would be unaffected.
Analysis: This finding had more than minor safety significance because it involved a
lack of required fire barriers for equipment that was relied upon for safe hot shutdown
following a fire. The finding also had more than minor safety significance because it
affected the objectives of the Mitigating Systems and Initiating Events Cornerstones of
Reactor Safety. The finding affected the availability and reliability of systems that
mitigate initiating events to prevent undesirable consequences. It also affected the
likelihood of occurrence of initiating events that challenge critical safety functions. The
finding did not have more than very low safety significance (Green) because of the low
fire ignition frequencies and lack of combustible materials in critical locations and
because of the effectiveness of the fire protection features and the unaffected SSD
equipment to mitigate a fire in each of the affected fire zones/areas.
Enforcement: As described in IR 50-400/02-11, Operating License Condition (OLC) 2.F
required that the licensee implement and maintain in effect all provisions of the
approved Fire Protection Program (FPP) as described in the Final Safety Analysis
Report (FSAR). The Updated FSAR, Section 9.5.1, FPP, stated that outside
containment, where cables or equipment (including associated non-essential circuits
that could prevent operation or cause maloperation due to hot shorts, open circuits, or
6
shorts to ground) of redundant safe shutdown divisions of systems necessary to achieve
and maintain cold shutdown conditions are located within the same fire area outside of
primary containment, one of the redundant divisions must be ensured to be free of fire
damage. Section 9.5.1 further stated that if both divisions are located in the same fire
area, then one division is to be physically protected from fire damage by one of three
methods: 1) a three-hour fire barrier, 2) a one-hour fire barrier plus automatic detection
and suppression, or 3) a 20-foot separation with no intervening combustibles and with
automatic detection and suppression. The licensee had received no NRC approvals for
deviating from these requirements.
Also, OLC 2. F. and UFSAR Section 9.5.1 stated that BTP 9.5-1 was used in the design
of the fire protection program for safety-related systems and equipment and for other
plant areas containing fire hazards that could adversely affect safety-related systems.
BTP 9.5-1, Section C.5.g, "Lighting and Communication," paragraph (1), required that
fixed self-contained lighting consisting of fluorescent or sealed-beam units with
individual eight-hour-minimum battery power supplies should be provided in areas that
must be manned for safe shutdown and for access and egress routes to and from all fire
areas.
In addition, TS 6.8.1, Procedures and Programs, required procedures as recommended
by Regulatory Guide (RG) 1.33 and procedures for fire protection program
implementation. RG 1.33 recommended procedures for combating emergencies,
including fires. The licensee's interpretation of their fire protection program was that
they could and would rely on proceduralized operator actions in place of physically
protecting SSD equipment from fire damage (see Section 1 R05.04.b.1).
Contrary to the above requirements, the licensee failed to adequately implement and
maintain in effect all of the provisions of the approved FPP. The licensee failed to
ensure that one of the redundant safe shutdown divisions of systems necessary to
achieve and maintain cold shutdown conditions was protected from fire damage; failed
to have adequate procedures for combating fire emergencies; and failed to provide the
required emergency lighting in areas that must be manned for safe shutdown; as
described above in the eight examples of this overall finding. Because the identified
examples of this failure to adequately implement and maintain in effect all of the
provisions of the approved FPP are of very low safety significance and have been
entered into your corrective action program (ARs 76260, 80212, 80089, 69721, 80215,
75065, and 79047), this violation is being treated as an NCV, consistent with Section
VL.A of the NRC Enforcement Policy: NCV 50-400/03-07-01; Failure to Adequately
Implement the Approved Fire Protection Program for Safe Shutdown, Eight Examples.
(2)
Inadequate Corrective Action for a Previous White Fire Protection Finding
Introduction: An overall finding was identified in that the corrective action for previous
Violation (VIO) 50-400/02-08-01 was inadequate. Four of the nine URIs described in IR
50-400/02-11 were considered to be examples of this overall finding. Based on
evaluating the three examples of this overall finding for their effects during a fire that
could occur in the six affected fire SSA area, this overall finding was determined to have
a very low significance (Green).
7
Description: The licensee's corrective action for a previous White fire protection finding
(VIO 50-400/02-08-01), associated with a Thermo-Lag fire barrier assembly between the
'B' train switchgear room / auxiliary control panel and the 'A' train cable spreading room,
was inadequate. Examples described in IR 50-400/02-11 include:
Many local manual operator actions were used in place of the required physical
protection of cables for equipment relied on for SSD during a fire, without
obtaining NRC approval for these deviations from the approved fire protection
program. This condition applied to all areas that were inspected, including the
new auxiliary control panel fire area that had been recently created as corrective
action for previous Violation 50-400/02-08-01. This reliance on large numbers of
local manual actions, in place of the required physical protection of cables, could
potentially result in an increased risk of loss of equipment that was relied upon
for SSD from a fire. (Section 1 R05.04.b.1)
Procedure steps for safe shutdown (SSD) from a fire and related corrective
action for previous Violation 50-400/02-08-01 were inadequate. For a fire in the
new auxiliary control panel fire area, certain cables were not physically protected
from the fire and certain SSD procedure steps, that were used in place of
physical protection of cables, involved excessive challenges to operators.
Consequently, a fire in the ACP fire area could result in a loss of all auxiliary
feedwater. (Section 1 R05.04.b.2)
A procedure for SSD from a fire and related corrective action for previous
Violation 50-400/02-08-01 were inadequate. For a fire in certain safe shutdown
analysis areas of the reactor auxiliary building, including the new auxiliary control
panel fire area, there were too many SSD procedure contingency actions to
respond to potential spurious actuations for the one designated SSD non-
licensed operator to perform. Consequently, equipment that was relied on for
SSD may not be available. (Section 1 R05.04.b.3)
Required battery-backed emergency lights were not provided in locations where
operators were required to perform actions for SSD from a fire. This condition
affected SSD during fires in all of the areas inspected in the reactor auxiliary
building, including the new auxiliary control panel fire area that was created as
corrective action for previous Violation 50-400/02-08-01. The lack of required
lighting could result in an increased risk of operators failing to perform the SSD
actions in a timely and accurate manner. (Section 1 R05.06.b)
Analysis: This finding had more than minor safety significance because it involved a
lack of required fire barriers for equipment that was relied upon for safe hot shutdown
following a fire. The finding also had more than minor safety significance because it
affected the objectives of the Mitigating Systems Cornerstone of Reactor Safety. The
finding affected the availability and reliability of systems that mitigate initiating events to
prevent undesirable consequences. The finding did not have more than very low safety
significance (Green) because of the low fire ignition frequencies and lack of
combustible materials in the ACP fire area and because of the effectiveness of the fire
8
protection features and the unaffected SSD equipment to mitigate a fire in the ACP fire
area.
Enforcement: OLC 2.F and the UFSAR, Section 9.5.1, FPP, included quality assurance
requirements for fire protection. The FPP stated that a QA program was being used to
identify and rectify any possible deficiencies in design, construction, and operation of the
fire protection systems. Also, as described in Section 1 R05.01.b.1 above, OLC 2.F
required that one of the redundant divisions would be free of fire damage. Further, if
both divisions were located in the same area, then one of the divisions was to be
physically protected from fire damage by one of three specified methods. Further,
OLC.2.F required that battery-backed emergency lights be provided in locations where
operators were required to perform actions for SSD from a fire. In addition, TS 6.8.1,
Procedures and Programs, required procedures for implementing the fire protection
program and for combating fires.
Contrary to the above requirements, the licensee's corrective actions for previous VIO
50-400/02-08-01 were inadequate because they failed to rectify deficiencies in design,
construction, and operation related to SSD from a fire in the area of the ACP room. The
licensee failed to protect various equipment either physically or procedurally from the
effects of a fire where that equipment was relied on for SSD. The licensee entered the
finding into the corrective action program as AR 80215. Because the identified
examples of this failure to adequately implement and maintain in effect all of the
provisions of the approved FPP are of very low safety significance and have been
entered into the corrective action program, this violation is being treated as an NCV,
consistent with Section Vl.A of the NRC Enforcement Policy: NCV 50-400/03-07-02;
Inadequate Corrective Action for a Previous White Fire Protection Finding.
40A6 Meetings
Exit Meeting Summary
The team presented the inspection results to Mr. _
_
and members of his staff
at the conclusion of the inspection on.
, 2003. The licensee acknowledged the
findings presented. Proprietary information is not included in this inspection report.
SUPPLEMENTAL INFORMATION
Partial List of Persons Contacted
Licensee
D. Baksa, Supervisor, Equipment Perfromance
J. Caves, Licensing Supervisor
R. Duncan, Director of Site Operations
M. Fletcher, Manager, Fire Protection Program
Attachment 1
9
P. Fulford, Superintendent, Design Engineering
C. Georgeson, Supervisor, El&C Design
W. Gregory, Operations Fire Protection Specialist
W. Gurganious, Manager, NAS
T. Hobbs, Manager, Operations
A. Khanpour, Manager, Engineering
F. Lane, Jr., Senior Nuclear Work Management Specialist
J. Laque, Manager, Maintenance
T. Morton, Site Services Manager
J. Scarola, Site Vice President
B. Waldrep, Plant General Manager
NRC
J. Brady, Senior Resident Inspector, Shearon Harris
H. Christensen, Deputy Director, Division of Reactor Safety (DRS), Region II (R1l)
C. Ogle, Chief, Engineering Branch 1, DRS, Rll
Items Opened. Closed, and Discussed
Opened
50-400/03-07-01
50-400/03-07-02
Failure to Adequately Implement the Approved Fire
Protection Program for Safe Shutdown, Eight Examples
(Section 1 R05.01.b.1)
Closed
50-400/02-11-01
Failure to Protect Charging System MOV 1 CS-1 65, VCT
Outlet to CSIPs, From Maloperation Due To a Fire
(Section 1 R05.01.b.1)
50-400/02-11-02
50-400/02-11-03
URI
Failure to Protect Charging System MOVs 1 CS-1 69, 1 CS-
214, 1CS-218, and 1CS-219 From Maloperation Due To a
Fire (Section 1 R05.01.b.1)
Failure to Protect Charging System MOVs 1 CS-1 66, 1 CS-
168, and 1CS-217 From Maloperation Due To a Fire
(Section 1 R05.01.b.1)
Attachment 1
10
50-400/02-11-04
50-400/02-11-05
50-400/02-11-06
50-400/02-11-07
50-400/02-11-08
50-400/02-11-09
URI
URI
URI
Failure to Protect Component Cooling MOVs 1 CC-251 and
1CC-208, CC for RCP Seals, From Maloperation Due To a
Fire (Section 1 R05.01.b.1)
Reliance on Manual Actions in Place of Required Physical
Separation or Protection From a Fire (Section
1 R05.01.b.2)
Fire SSD Operator Actions With Excessive Challenges
(Section 1 R05.01.b.2)
Too Many Fire SSD Actions for Operators to Perform
(Section 1 R05.01.b.2)
Using the Boric Acid Tank Without Level Indication
(Section 1 R05.01.b.1)
Failure to Provide Required Emergency Lighting for SSD
Operator Actions (Section 1 R05.01.b.2)
Discussed
50-400/02-08-01
V10
Failure to Implement and Maintain NRC Approved Fire
Protection Program Safe Shutdown System Separation
Requirements (Section 1 R05.01.b.2)
Attachment 1