ML040090428
| ML040090428 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 12/22/2003 |
| From: | Ogle C NRC/RGN-II |
| To: | Jamil D Duke Energy Corp |
| References | |
| FOIA/PA-2003-0358 IR-03-007 | |
| Download: ML040090428 (32) | |
See also: IR 05000369/2003007
Text
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
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REGION II
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SAM NUNN ATLANTA FEDERAL CENTER
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61 FORSYTH STREET SW SUITE 23T85
ATLANTA, GEORGIA 30303-8931
Duke Energy Corporation
ATTN: Mr. D. Jamil
Vice President
McGuire Nuclear Station
12700 Hagers Ferry Road
Huntersville, NC 28078-8985
SUBJECT:
MCGUIRE NUCLEAR STATION - NRC TRIENNIAL FIRE PROTECTION
INSPECTION REPORT 50-369/03-07 AND 50-370/03-07
Dear Mr. Jamil:
On May 23, 2003, the U.S. Nuclear Regulatory Commission (NRC) completed an Inspection at
your McGuire Nuclear Station, Units 1 and 2. The enclosed report documents the inspection
findings which were discussed on May 22, 2003, with you 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 three findings that 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 was initiated'on
June 10, 2003, as a compensatory measure one of the findings.
If you contest any violation in this report, you should provide a response with the basis for your
-denial,-within 30 days of the date of this inspection report, to the United States Nuclear
Regulatory CommissionAfN--Docu rent Control Desk, Washington, D.C. 20555-0001,'with
copies to the Regional Administrator, Regionil; the Director, Office of Enforcement, United
States Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident
Inspector at the McGuire facility.
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
ALA
2
NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at
hftt://www.nrc.aov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
Charles R. Ogle, Chief,
Engineering Branch I
Division of Reactor Safety
Docket Nos.: 50-369, 50-370
Enclosure: Inspection Report 50-369, 370/03-07
w/Attachment: Supplemental Information
cc w/encl:
C. J. Thomas
Regulatory Compliance Manager (MNS)
Duke Energy Corporation
Electronic Mail Distribution
M. T. Cash, Manager
Regulatory Issues & Affairs
Duke Energy Corporation
526 S. Church Street
Charlotte, NC 28201-0006
Lisa Vaughn
Legal Department (E iUX)
Duke Energy Corporation
-
422 South Church Street
Charlotte, NC 28242
Anne Cottingham
Winston and Strawn
Electronic Mail Distribution
Beverly Hall, Acting Director
Division of Radiation Protection
N. C. Department of Environmental
Health & Natural Resources
Electronic Mail Distribution
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(cc w/encl cont'd - See page 3)
(cc wlencl cont'd)
County Manager of Mecklenburg County
720 East Fourth Street
Charlotte, NC 28202
Peggy Force
Assistant Attorney General
N. C. Department of Justice
Electronic Mail Distribution
Distribution w/encl:
B. Martin, NRR
L. Slack, R11 EICS
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SIGNATURE
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NAME
MThomas
PFillion
RMaxey
RSchin
BMelly
CPayne
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DATE
7/
12003
7/
12003
7/
/2003
7/
/2003
7/
/2003
7/
/2003
7/
/2003
E-MAIL COPY?
YES
NO
YES
NO
YES
NO
YES .
NO
YES
NO
YES
.NO
YES
NO
PUBLIC DOCUMENT
YES
NOI
OFFICIAL RECORD COPY
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DOCUMENT NAME: S:XDRSXEng Branch 1-Fire ProtecUon\\ReportsXMc~uIreXMcG 0307
./
U.S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos.:
License Nos.:
Report No.:
Licensee:
Facility:
Location:
Dates:
Inspectors:
50-369, 50-370
50-369/03-07 and 50-370/03-07
Duke Energy Corporation
McGuire Nuclear Station
12700 Hagers Ferry Road
Huntersville, NC 28078
May 5 - 9, 2003 (Week 1)
May 19 - 23, 2003 (Week 2)
P. Fillion, Reactor Inspector
R. Maxey, Reactor Inspector
B. Melly, Fire Protection Engineer (Consultant)
R. Schin, Senior Reactor Inspector (April 14-17, 2003)
M. Thomas, Senior Reactor Inspector (Lead Inspector)
Approved by:
Charles R. Ogle, Chief
Engineering Branch 1
Division of Reactor Safety
Enclosure
SUMMARY OF FINDINGS
IR05000369/03-07, IR05000370/03-07; Duke Energy Corporation; 05/05-09/2003 and 05/19-
23/2003; McGuire Nuclear Station, Units 1 and 2; Triennial Fire Protection
The report covered a two-week period of inspection by regional inspectors and a consultant.
Three 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.
Inspector Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
TBD. The team identified a violation because Train A and Train B cables associated
with the reactor protection system were located in the same fire area and were not
protected from fire damage, as required by.McGuire's fire protection program.
This finding is unresolved pending determination of the systems affected and completion
of a significance determination. This finding is greater than minor because it was
associated with the equipment performance attribute and affected the objective of the
mitigating systems cornerstone to ensure the availability, reliability and capability of
systems that respond to initiating events in that instrumentation important for post-fire
safe shutdown could be lost. This finding did not present an immediate safety concern,
however, a fire watch was initiated on June 10, 2003, as a compensatory measure.
When assessed in combination with the finding related to inadequate protection of
auxiliary feedwater system cables and equipment required for safe shutdown in Fire
Area 16/18 (also discussed in this inspection report), this finding may have potential
safety significance greater than very low significance. (Section 1 R05.03.b.1)
TBD. The team identified a violation in that the turbine driven auxiliary feedwater
(TDAFW) pump suction supply valve 2CA0007A was not evaluated in the licensee's
safe shutdown analysi formpotential impact on safe shutdown in the event of a fire
where the TDAFW pump is require-dfar safe shutdown. The valVe-could spuriously
operate, due to fire damage and adversely affect the -TDAFW pump.
The finding is unresolved pending completion of a significance determination. The
finding is greater than minor because it was associated with the equipment performance
attribute and affected the objective of the mitigating systems cornerstone to ensure the
availability, reliability and capability of systems that respond to initiating events in that
spurious closure of the valve could damage the TDAFW pump and seriously degrade
the decay heat removal function. This finding may have potential safety significance
greater than very low significance. (Section 1 R05.04.b.2)
2
B.
Licensee Identified Violations
TBD. The physical protection of cables and equipment relied upon for safe shutdown
(SSD) of Unit 2 during a fire in the Train A Electrical Penetration Room (Fire Area 16/18)
was not adequate. Train B electrical cables, associated with the 2B motor driven
auxiliary feedwater pump discharge valve 2CA0042B to steam generator 2D, were
located in the Train A Electrical Penetration Room (Fire Area 16/18) without adequate
spatial separation or fire barriers as required by the McGuire 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 the electrical cables associated with valve 2CA0042B.
This finding is unresolved pending completion of a significance determination. The
finding is greater than minor because it was associated with the equipment performance
attribute and affected the objective of the mitigating systems cornerstone to ensure the
availability, reliability and capability of systems that respond to initiating events in that
fire damage to the unprotected cables could prevent operation of SSD equipment from
the main control room. When assessed in combination with the inadequate reactor
protection system cable separation finding (also discussed in this inspection report), this
finding may have potential safety significance greater than very low significance.
(Section 1R05.03.b.2)
Report Details
1.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems and Barrier Integrity
1R05 FIRE PROTECTION
The purpose of this inspection was to review the McGuire Nuclear Station (MNS) fire
protection program (FPP) for selected risk-significant fire areas. Emphasis was placed
on verification that the post-fire safe shutdown (SSD) capability and the fire protection
features provided for ensuring that at least one redundant train of safe shutdown
systems is maintained free of fire damage. The inspection was performed in
accordance with the Nuclear Regulatory Commission (NRC) Reactor Oversight Program
using a risk-informied approach for selecting the fire areas and attributes to be
inspected. The team used the licensee's Individual Plant Examination for External
Events (IPEEE) and performed in-plant walk downs to choose four risk-significant fire
areas for detailed inspection and review. The four fire areas selected were:
Fire Area 4, Auxiliary Building (AB) Common Area; AB +716 feet elevation
Fire Area 13, Battery Rooms; AB +733 feet elevation common area
Fire Area 16/18, Unit 2 Train A Electrical Penetration Room/2ETA 4160 volt
Switchgear Room; AB +750 feet elevation
Fire Area 24, Main Control Room (MCR); AB +767 feet elevation
For each of the selected fire areas, the team focused the inspection on the fire
protection features, and on the systems and equipment necessary for the licensee to
achieve and maintain safe shutdown conditions in the event of a fire in those fire areas.
The team evaluated the licensee's FPP against applicable requirements, Including
Operating License Conditions 2.C.4 and 2.C.7Fire Pr6tection Program, for Units 1 and
2, respectively; Title 10 of the Code'of Federal Regulations Part 50 (10 CFR 50),
Appendix R, Sections 1I1. G.-J, L, and O0; IOCFR 50.48; Appendix A to Branch Technical
Position Auxiliary and Power Conversion Systems Branch 9.5-1, Guideline for Fire
Protection for Nuclear Power'Plants; related NRC Safety Evaluation Reports (SERs);
MNS Updated Final Safety Analysis Report (UFSAR), Section 9.5.1; UFSAR Section
16.9, Selected Licensee Comrriitments (SLC); and plant Technical Specifications (TS).
The team evaluated all areas of this inspection, as documented below,'against these
requirements.
-
.01
Systems Required to Achieve and Maintain Post-Fire Safe Shutdown
i'
a.
Inspection Scope
2
The team reviewed the licensee's FPP described in UFSAR Section 9.5.1; the MNS Fire
Protection Review; safe shutdown analysis (SSA); fire hazards analysis (FHA); SSD
essential equipment list; and system flow diagrams to identify the components and
systems necessary to achieve and maintain SSD conditions. 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. The following Unit 2 systems and components were selected for
review:
Standby Shutdown System (SSS)
Standby makeup pump (SMP) 2NVPU0046
SMP suction supply valve 2NV842AC
Auxiliary feedwater (AFW) suction supply valves 2CA007A and 2CA009B
Reactor Coolant Pump (RCP) seal water return isolation valve 2NV94AC
Pressurizer power operated relief valve (PORV) 2NC34A
PORV isolation valves 2NC33A
Pressurizer heaters No. 28, 55, 56
Reactor vessel head vent valves 2NC272AC and 2NC273AC
Heating, ventilation, and air conditioning (HVAC)
Specific licensee documents, calculations, and drawings reviewed during this inspection
are listed in the attachment.
b.
Findings
No findings of significance were identified.
.02
Fire Protection of Safe Shutdown Capability
a.
Inspection Scope
The team reviewed the fire detection system protecting Fire Areas 4, 13, 16, 18 and 24
to assess the adequacy of the design and installation. This was accomplished by
reviewing design drawings, ceiling beam location drawings, and National Fire Protection
Association (NFPA) 72E (code of record 1974 edition) for detector location
requirements. The team reviewed the McGuire Fire Protection Code Deviation
Calculation to determine if there were any outstanding code detector deviations for the
selected areas. The team walked down the fire detection and alarm systems in Fire
Areas 13, 16, and 18 to evaluate the installed detector locations relative to the NFPA
72E location requirements. Additionally, the team reviewed the surveillance test
procedures for the detection and alarm systems to determine compliance with UFSAR
Sections 9.5.1 and 16.9.
The team reviewed the adequacy of the design and installation of the fire suppression
system protecting the nuclear service water (RN) pump area in Fire Area 4. This was
accomplished by reviewing the engineering design drawings, suppression system
13
hydraulic calculations, as-built system configuration and NFPA 13 (code of record 1978
edition) for sprinkler system location requirements. The team also reviewed the
McGuire Fire Protection Code Deviation Calculation for the RN pump sprinkler system to
determine the adequacy of the system to control a fire in this area utilizing the 2-1/2 inch
by-pass lines as the'sole means of supplying the sprinkler system.
The team reviewed the fire hose stations in Fire Areas 4,13,16, 18 and 24 to assess
the adequacy of the design and installation. This was accomplished by reviewing the
fire plan drawings, engineering mechanical equipment drawings, pre-fire strategies and
NFPA 14 (code of record 1976 edition)'for hose station location requirements and
effective reach capability. Team members also performed a field walkdown of the
selected fire areas to ensure that hose stations were not blocked and to compare hose
station location drawings with as-built plant locations.
b.
Findinas
The team identified an unresolved Item (URI) involving the adequacy of the suppression
system for Fire Area 4. Dedicated shutdown (DSD) using the SSS was designated by
the licensee for a fire in this area. 10 CFR 50, Appendix R, Section III.G.3 (altemative
or dedicated shutdown) requires that fire detection and a fixed fire suppression system
shall be installed in the area, room, or zone under consideration. The fire suppression
system for Fire Area 4 was not installed in accordance with 10 CFR 50, Appendix R,
Section III.G.3. The system in Fire Area 4 was a partial automatic sprinkler system
effectively protecting the RN pumps and 20 feet north of these pumps. The area
protected by this sprinkler system was located between column lines 54-58 and EE-GG.
The majority of Fire Area 4 was not provided with automatic sprinkler protection as
required by 10 CFR 50, Appendix R, Se6tion III.G.3.
This issue was previously identified by the NRC (URI 50-369/84-28-01, 370/84-25-01) in
1984 during an Appendix R inspection. The licensee considered this Issue to be a
potential backfit per 10 CFR 50.109 (letter dated September 4, 1984, from H.B. Tucker,
Duke Power Company, to H.R. Denton, NRC Office of Nuclear Reactor Regulation).
The URI was closed in NRC inspection report (IR) 50-369,370/87-34. The team noted
that, subsequent to closure of the URI, licensee Fire Protection Functional Audit SA-99-
04(MC)(RA)(FPFA) dated April 9, 1999, identified that MNS did not meet separation and
detection/suppression criteria for alterriative or dedicated shutdown capability-required
by 10 CFR 50, Appendix R, Section III.G.3. During the current inspection, the team
questioned whether the previous reviews of the sprinkler system for this fire area
included an evaluation of the risk impact associated with not providing adequate
sprinkler coverage for the RN cabling in this fire area. The team informed the licensee
that this issue would be reviewed to determine if the lack of sprinkler coverage in this
fire area has an impact on' risk. The team noted that a similar condition exists in other
fire areas where dedicated shutdown capability using the SSS was designated by the
licensee. Pending determination of whether a backfit evaluation is warranted, this issue
is identified as URI 50-369, 370/03-07-01, Fire Suppression System for Dedicated
Shutdown Areas not in Accordance with 10 CFR 50, Appendix R, Section III.G.3.
4
.03
Post-Fire Safe Shutdown Circuit Analysis
a.
Inspection Scope
The team reviewed the adequacy of separation and fire barriers provided for the power
and control cabling of equipment relied on for SSD during a fire in the selected fire
areas. On a sample basis, the team reviewed the SSA and the electrical schematics for
power and control circuits of SSD components, and looked for the potential effects of
open circuits, shorts to ground, and hot shorts. This review focused on the cabling of
selected components of the charging/makeup system, reactor coolant system (RCS)
and AFW system.. The team traced the routing of cables by using the cable schedule
and conduit and cable tray drawings. The team walked down the selected fire areas to
compare the actual plant configuration to the cable layout on the drawings. Circuit and
cable routings were reviewed for the following equipment:
ORN4AC, Turbine Driven AFW Suction Supply Valve
2CA0007A, Turbine Driven AFW Suction Isolation Valve
2CAO09B, Motor Driven AFW Suction Isolation Valve
2CFLT6080, 6090, 6100, 6110, Steam Generator Level Transmitters
2NCLT5151, Pressurizer Level Transmitter
2NC34A, Pressurizer PORV
v
2NC33A, PORV Isolation Valve
2NC272AC, 273AC, Reactor Vessel Head Vent Valves
2NVPU0046, Standby Makeup Pump
2NV94AC, RCP Seal Water Return Isolation Valve
2NV842AC, SMP Suction Isolation Valve
2NV1012C, SMP Discharge to Containment Sump Isolation Valve
Pressurizer heaters No. 28, 55, 56
The team also reviewed licensee studies of overcurrent protection for alternating current
and direct current systems to identify whether fire-induced faults could result in
defeating the SSD functions.
b.
Findings
Findings associated with valves 2CA0007A, 2NC34A, and 2NC33A are discussed in
Section .04 of this IR.
1.
Introduction: A finding with potentially greater than very low safety significance was
identified in that redundant instrumentation (and possibly other equipment) important to
SSD could be damaged by a fire in Fire Area 16/18. This finding involved a violation of
NRC requirements. This finding is a URI pending completion of the SDP.
5
Descrigtion: Fire Area 16/18 is the Unit 2 Train A electrical penetration room/2ETA 4160
volt (V) switchgear room. Train B equipment controlled from the MCR room was
designated as the SSD train for a fire in this area according to the SSA and plant
procedures. During a walkdown of Fire Area 16/18, the team identified that room 805A
lacked fire detection and fire suppression. -Room 805A is the HVAC equipment room
which supplies ventilation to the Unit 2 Train A 4160V switchgear room 2ETA. The team
also observed that Train B cables were routed through room 805A. Many of the
identified cables were in cable-trays near the ceiling and were going from/to the cable
spread room, which was on the same elevation; and to/from 'the control room,;which
was above room 805A. The licensee had not been aware these Train B cables passed
through room 805A, and initiated Problem Investigation Process (PIP) M-03-02106 and
M-03-02588. [The team identified that a similar condition also existed in room 803A
(Fire Area '17), which is the HVAC equipment room providing ventilation for the Unit 1
Train A 4160V switchgear room 1 ETA]. On June 10, 2003, the licensee reported that
these cables did not meet the separation'criteria of Appendix R and represented an
unanalyzed condition (Event No. 39915), and initiated a fire watch as a compensatory
measure.
Preliminary investigation by the licensee revealed that cables for primary and backup
power supplies for all four reactor protection system (RPS) channels were routed in
close proximity and could be damaged during a severe fire.' As many as 74 Train B
RPS cables may be involved. Onie6consequence of this finding is that fire-induced cable
damage may cause many RPS protective functions to spuriously go to the trip condition.
Consequently, a safety injection signal could be generated due to spurious high
containment pressure. The safety injection signal could in tum trigger a reactor trip and
Phase A isolation. [At the same time, many main control panel instruments necessary
to achieve and maintain hot shutdown would be lost, including pressurizer level and all
four steam generator (SG) level instruments.] The licensee also stated that similar
effects could occur for a fire in the Unit I Train A switchgear room I ETA (Fire Area 17).
Analysis: The team determined that this finding was associated with the equipment
performance attribute and 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. The finding did not present an immediate
safety concern, however, the licensee initiated a fire watch'on June 10, 2003, as a
- compensatory measure. The licensee is-analyzing the manner in which plant systems
would be affected by fire drmrge to the Train B cables and is reviewing plant abnormal
procedures (APs) in light of the degraded instrumentation and any automatic actions
that would be initiated. Once the equipment degradations and relevant procedures are
understood, the significance determination process (SDP) will be used to determine the
level of significance. When assessed incombination with the finding related to
inadequate protection of AFW cables and equipment required for SSD In Fire Area
- -
16/18 (Section .03.b.2); this finding may have potential safety significance greater than
- verylo 'significance.:
-
6
Enforcement: The licensee's FPP commits to 10 CFR 50, Appendix R, Section III.G.
Section IlI.G.1.a. states, in part, that one train of systems necessary to achieve and
maintain hot shutdown shall be free of fire damage.
Contrary to the above, redundant trains of instrumentation necessary to achieve and
maintain hot shutdown could be damaged during a fire in room 805A (Fire Area 16/18).
Pending determination of the safety significance, the finding is identified as URI 50-369,
370/03-07-02, Failure to Protect Redundant Trains of Reactor Protection System Cables
From the Effects of Fire.
2.
Inadequate Protection of AFW Cables and Equipment Required for Safe Shutdown
Introduction: A finding was identified in that physical protection of the associated
electrical cables for valve 2CA0042B (2B motor driven AFW pump discharge supply to
SG 2D) did not meet the requirements of 10 CFR 50, Appendix R, Section III.G.2.
Instead, the licensee used a local manual operator action, which had not received prior
NRC approval, to achieve and maintain SSD. This is a URI pending completion of the
SDP.
Description: On April 2, 2003, the licensee identified that MNS relied on local, manual
operator actions outside the MCR for SSD in non-dedicated shutdown fire areas (i.e.,
areas designated as complying with 10 CFR 50, Appendix R, Section III.G.2). These
local, manual operator actions did not have prior NRC approval. The licensee
documented this issue in PIP M-03-02311. The team reviewed the local, manual
operator action for the Appendix R,Section III.G.2 fire area selected for this inspection
(Fire Area 16/18).
The team found that the associated electrical cables for Train B valve 2CA0042B were
located in the Unit 2 Train A electrical penetration room (Fire Area 16/18) without
adequate spatial separation or fire barriers. Rather than providing adequate physical
protection for redundant trains of equipment/systems necessary to achieve and maintain
SSD (as specified for Appendix R, Section lll.G.2 areas), the licensee substituted the
use of a manual operator action outside the MCR. The licensee's SSA stated that de-
energizing this valve, after verifying that it was open, was a time critical action because
spurious closure of this valve would limit the secondary heat sink to only one SG (rather
than the two required to achieve and maintain SSD). The use of local manual operator
actions, in fire areas designated as complying with the provisions of Appendix R,
Section III.G.2, requires prior NRC review and approval. This local, manual operator
action had not received NRC approval.
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. When assessed in
combination with the inadequate RPS cable separation finding (Section .03.b.1), this
finding may have potential safety significance greater than very low significance.
7:
Enforcement: The licensee's FPP commits to 10 CFR 50,- Appendix R,Section III.G.
Section lll.G.2 states in part, that,
U...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: (1)
separation of cables and equipment of redundant trains by a fire
barrier having a 3-hour rating; (2) 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; (3) enclosure of cables 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 May 23, 2003, the licensee failed to protect cables of
redundant equipment located within the Unit 2 Train A electrical penetration room/4160V
switchgear room 2ETA (Fire Area 16/18) with an adequate barrier or to provide 20 feet
,of separation. Pending determination of the finding's safety significance, this finding is
identified as URI 50-370/03-07-05, Failure to Provide Adequate Protection for Cables of
Redundant Safe Shutdown Equipment in Fire Area 16/18.
.04
Alternative Post-Fire Safe Shutdown Capability
a.
Inspection Scope
-The team reviewed the licensee's procedures for fire response, APs for DSD, and the
licensee's Appendix R fire area failure analysis and compliance strategy for a fire in Fire
Areas 4,13, and 24. The team also walked down selected portions of the procedures in
the'plant. The reviews focused on ensuring that the required -functions for post-fire safe
shutdown and the corresponding equipment necessary to perform those functions were
included in the procedures. The review also included assessing whether hot and cold
shutdown from outside the MCR could be implemented, and that transfer of control from
the MCR to the standby shutdown facility (SSF) could be accomplished within the
performance goals stated in 10 CFR 50, Appendix R, Section III.L. The components
listed in Section .03.a. of this IR were also reviewed in relation to DSD capability. The
team reviewed the most recently completed surveillances for selected instruments
required during SSS operation to verify that these surveillances were being completed in
accordance with MNS SLC 16.9.7,.Standby Shutdown System. The team walked down
- DSD procedures to determine if they could be performed within the required times given
the minimum required staffing level of operators, with or without offsit power available.
8
The team also reviewed the electrical isolation of selected motor operated valves from
the control room to verify that operation of the SSS from the SSF, and other remote
plant locations, would not be prevented by a fire-induced circuit fault.
b.
Findings
I .
Requirements Relative to the Number of Sgurious Operations that Must be Postulated
Introduction: The team identified an issue involving the number of concurrent spurious
operations associated with a particular component or set of components that must be
postulated during SSD analysis of a fire area. This issue is a URI pending review by
NRC staff.
Description: The licensee's SSA included the concept that only one spurious operation
due to fire damage need be postulated. This concept became evident during review of
the pressurizer PORVs. There are three sets of PORVs and PORV isolation valves on
the pressurizer of each unit. Should operators in the control, room become aware of a
fire in any plant area (from a fire alarm or the plant communications system), they would
respond by implementing procedure AP101NA55001045, Plant Fire. Depending on the fire
location, procedure AP/0/A15500/045 directed the operator to close the PORV isolation
valves within ten minutes. The basis for this time critical action is that spurious opening
of the PORV, or damage to the isolation valve circuit would not occur in the first ten
minutes of a fire being detected. With the isolation valve closed, it would then take two
spurious operations to breach the RCS pressure boundary (i.e., the isolation valve
opening and its associated PORV also opening). This concept of postulating only one
spurious operation meant that closing the isolation valve was sufficient to ensure RCS
pressure boundary integrity. The licensee considered that there was no need to take
any other action such as de-energizing the isolation valve after it was closed.
Application of this concept is not consistent with NRC's cable protection requirements of
Appendix R,Section III.G.
The team reviewed the control circuits and cable routing information for pressurizer
PORV 2NC34A, and its associated isolation valve 2NC33A. They observed that cables
for both the PORV and isolation valve were routed through Fire Areas 13, 16/18 and 24.
The team determined that, for these three fire areas, spurious opening of the PORV
could only occur for a MCR fire (Fire Area 24). If more than one spurious operation
were to occur, the dedicated shutdown capability (SSS) would not be independent from
the MCR in that a fire in the control room could result in conditions outside those
specified in Appendix R,Section III.L.
Analysis: The team determined that this finding was associated with the equipment
performance attribute of the mitigating systems cornerstone. Because it affected this
cornerstone's objective to ensure the availability, reliability, and capability of systems
that respond to initiating events, this finding is greater than minor. If more than one
spurious operation were to occur, the dedicated shutdown capability (SSS) would not be
'9
independent from the MCR in that a fire in the MCR could result in conditions outside of
those specified in Appendix R,Section III.L.
Enforcement: In the case of the PORV and PORV isolation valve circuits, operation of
the SSS may not be independent'of the fire area as required by Appendix R, Section
III.G.3. Review of this matter by the NRC will determine whether a violation has
occurred. Pending the issuance of new NRC inspection guidance regarding associated
circuits, the issue is identified as URI'50-369, 370/03-07-03, Requirements Relative to
the Number of Spurious Operations That Must be Postulated.
2.
Auxiliary Feedwater Valve 2CA0007A Not Included in Safe Shutdown Analysis
Introduction: A finding with potentially greater than very low safety significance' was
identified in that AFW suction supply valve 2CA0007A, which could spuriously operate
during a MCR fire, was not included in the SSA. Spurious closure of this valve could
damage the turbine driven auxiliary feedwater (TDAFW) pump, thus seriously degrading
the secondary decay heat removal function of the SSS. This is a URI pending
completion of the SDP.
Description: Valve 2CA0007A is a motor operated valve in the suction flow path from
the 300,000 gallon AFW storage tank to the TDAFW pump. 'The valve Is open during
normal plant operation. 2CA0007A is important to safe shutdown for fire areas where
the'SSS will be used. The importance is derived from the fact that the SSS relies on the
TDAFW pump for secondary decay heat removal. Spurious closure of the valve would
immediately reduce suction 'pressure and quickly shut off all flow through the pump
causing severe damage. For a severe fire in the MCR requiring evacuation and transfer
of plant shutdown to the SSS, the ability to remove decay heat would be seriously
degraded if the TDAFW pump were damaged. The team found that the'SSA did not
include valve 2CA0007A. The valve was not listed in Appendix E; Unit 1 and Unit 2
Safe Shutdown Equipment; nor Appendix F, Fire Area Failure Analysis and Compliance
Strategy, of the SSA (Specification MCS-1465.00-00-0022, Design Basis Specification
for Appendix R).
The licensee initiated PIPs M-03-02084, M-03-02118, and M-03-02311 for this issue
and took'prompt action to prevent spurious operation of this valve. Procedure
-AP/O/A/5500/045 was revised to specify~t the operator ensure, within'the first ten
minutes of an active fire, that valve 2CA0007A was open and then remove power from
2CA0007A.
The team noted that system design provided for automatic transfer to alternate suction
sources initiated by pressure switches in the TDAFW pump suction line. There were
three separate alternate suction flow paths. Path 1 was through valves 2CA1 61 C,
.2CA162C and ORN4AC; Path 2 was through valves 2CA086A and 2RN069A; and Path
.3 was tirough valves 2CA116B and 2RN162B. However, key information related to
these automatic'transfers was not available to the team during the inspection.
10
Information was subsequently provided to the team, however, this information has not
yet been fully reviewed.
Analysis: The team determined that this finding was associated with the equipment
performance attribute and 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. For a severe fire in the MCR, the MCR
would be evacuated and the SSF would be used to achieve and maintain hot shutdown.
Because the SSF relies on the TDAFW pump for the decay heat removal, the decay
heat removal function would be seriously degraded if the TDAFW pump were damaged
due to closure of valve 2CA0007A.
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.
MNS Unit 2 Operating License NPF-17, Condition 2.C.(7) states, in part, that the
licensee shall implement and maintain in effect all provisions of the approved FPP as
described in the UFSAR for the facility, and as approved in the SER dated March 1978
and SER Supplements 2, 5, and 6 dated March 1979, April 1981, and February 1983,
respectively, and the safety evaluation dated May 15i 1989.
The UFSAR states that the overall concept and details of the FPP are presented in the
MNS Fire Protection Review (MCS-1465.00-00-0008). The FPP, which includes the
SSA (MCS-1465.00-00-0022) for MNS, states in part, that the FPP implemented the
philosophy of defense-in-depth protection against fire hazards and effects of fire on SSD
equipment. It further states that the SSA performed for MNS considered potential fire
hazards and their possible effects on SSD capability. The licensee's SSA designated
the MCR (Fire Area 24) and Fire Area 4 as dedicated shutdown areas. Appendix R,
Section III.G.3 requires that the alternative/dedicated shutdown capability, and its
associated circuits, be independent of cables, systems or components in the area under
consideration.
Contrary to these requirements, valve 2CA0007A was not included in the SSA resulting
in the dedicated shutdown system (SSS) not being independent from Fire Area 24, in
that, a fire in these areas could result in spurious closure of this valve and damage to
the .TDAFW pump. Pending determination of the safety significance, this finding is
identified as URI 50-370/03-07-06, Spurious Closure of Valve 2CA0007A Could Lead to
Damage of the TDAFW Pump.
.05
Operational Implementation of Post-Fire Safe Shutdown Capability
a.
Inspection Scope
The team reviewed the operational implementation of the SSD capability for a fire in Fire
Areas 4,13, 16/18, or 24 to verify that: (1) the training program for licensed personnel
included dedicated safe shutdown capability; (2) personnel required to achieve and
maintain the plant in hot standby following a fire using the'SSS could be provided from
11;
normal onsite staff, exclusive of the fire brigade; (3) the licensee had incorporated the
operability of dedicated shutdown transfer and control functions into plant TS and/or
SLCs; and (4) the licensee periodically performed operability testing of the dedicated
shutdown instrumentation, and transfer and control functions. The team reviewed
procedures APII/A/5500/24 and AP12/A15500/024, Loss of Plant Control Due to Fire or
Sabotage, and APIO/AJ5500/045, Plant Fire. 'The reviews focused on ensuring that all
required functions for post-fire safe shutdown, and the corresponding equipment
necessary to perform those functions, were included in the procedures.
b.
Findings
The licensee identified that local, manual operator actions outside the MCR were used
in lieu of physical protection of equipment and cables relied upon for SSD during a fire
without obtaining prior NRC approval. Findings related to this issue for Fire Area 16/18
are discussed in Section 03.b.2 of this IR.
-
The team identified a URI regarding the adequacy of the licensee's method for
controlling RCS pressure during operation from the SSF in the event of a fire. During
review of procedures AP/1/AJ5500/024 and AP/2/A/5500/024, the team questioned the
adequacy of the 70 kilowatts (kW) pressurizer heater capacity (per unit) powered from
the SSF to maintain and control RCS pressure in hot standby during a fire in plant areas
which require use of the SSS. A procedural note in both AP/1/A5500/024 and
AP12/AN55001024 provided guidance to the operators which stated that it was acceptable
to allow the pressurizer to go water solid in order to maintain subcooling, and with the
pressurizer water solid, the reactor vessel head vents would be used to control
pressure. Allowing the pressurizer to go water solid for controlling RCS pressure during
hot standby conditions while operating from the SSF was not consistent with Appendix
R, Section lll.L, for dedicated shutdown capability, nor the.design basis description for
the SSF as stated in the licensee's letter to the NRC dated March 31, 1980. Also, solid
plant operation from the SSF for controlling RCS pressure was neither reviewed nor
discussed in any NRC SER/SER Supplements relative to acceptability of the SSF
design for dedicated shutdown capability. The team requested information from the
licensee (e.g., analyses, calculations, etc.) which demonstrated the following:
Adequacy of the 70 kW pressurizer heater capacity powered from the SSF for
maintaining and controlling RCS pressure in hot standby.
Validity of the assumptions for pressurizer heat loss stated in the October 21,
1980, letter (based on insulation degradation and/or degraded capacity of the
heaters powered from SSF) for current pressurizer, heat loss and for determining
when the heaters will be needed.
. .
SSMP capacity to'achieve and control solid plant operation from the SSF within
the' required time to maintain subcooling.
12
Operator training Gob performance measures, simulator, etc.) on solid plant
operation from the SSF.
The licensee indicated that there were no specific calculations documented which
provided the basis for the number of heaters to be powered from the SSF. The licensee
further stated that there was no calculation which demonstrated the performance
capability of the SMP during solid plant operation from the SSF. The licensee also
indicated that training provided to operators on solid plant operation from the SSF
consisted primarily of classroom discussions and tabletop discussions of procedures
AP/1/A/5500/024 and AP/2/A15500/024. The team concluded that sufficient information
was not provided to resolve the questions raised above nor to determine the licensee's
ability to safely operate the SSF with the pressurizer in a water solid condition during
fire events in areas where the SSF is used to achieve SSD. Pending further NRC
review of additional licensee information, this issue is identified as URI 50-369,370/03-
07-04, Reactor Coolant System Pressure Control During SSF Operation.
.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
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.
b.
Findings
No findings of significance were identified.
.07
a.
Inspection Scope
The team compared the installation of the licensee's emergency lighting systems to the
requirements of 10 CFR 50, Appendix R, Section III.J, to verify that 8-hour emergency
lighting coverage was provided in areas where manual local operator actions were
required during post-fire SSD operations, including the access and egress routes. The
team's review also included verifying that emergency lighting requirements were
- i
evaluated in the licensee's SSA and properly integrated into the post-fire SSD
--
rocedures. During team walk downs of the selected areas where local, manual
operator actions would be performed, area emergency lighting units were inspected for
13
operability and the aiming of lamp heads was checked to determine if adequate
illumination would be available to correctly and safely perform the actions directed by
the procedures.
b.
Findings
-
No findings of significance were identified.
.08
Cold Shutdown Repairs
a.
Inspection Scope
'
The team reviewed the licensee's SSA'an'd 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 were 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 reviewed the selected fire'areas to evaluate the adequacy of the fire
resistance of fire area barrier enclosure walls, ceilings, floors, fire'barrier mechanical
and electrical penetration'seals, fire'doors, and fire dampers. This was accomplished by
observing the material condition and configuration of the installed fire barrier features,
as well as construction details and supporting fire endurance tests for the installed fire
barrier features, to verify the as-built configurations were qualified by appropriate fire
endurance tests. The team also reviewed the fire hazards analysis to verify the fire
loading used by the licensee to determine the fire resistive rating of the fire barrier
enclosures. The team also reviewed the design specification for mechanical and
electrical penetrations, fire flood and pressure seals, penetratiori seal database and
Generic Letter (GL) 86-10 evaluations and the calculation for the technical basis of fire
barrier penetration seals to verify that the fire barrier installations met licensing basis
commitments.
The team reviewed fire barriers 'shown on the fire plan drawings for the selected fire
'
areas. The team noted that MNS has"eliminated selected fire barriers from the
1Ag
approved fire protection program and'designated these fire barriers as "Sealed Firewall -
Non Committed". These barriers are no longer included in any surveillance and testing
program. Therefore, doors, dampers, fire proofing, etc. that exist in these declassified
barriers are no longer included in any station surveillance procedures and effectively
cannot be relied upon' for the'fire protedtion'program. Two walls'associated with Fire
Area 16/18 have been declassified. The wall between the Unit 2 switchgear room 2ETA
14
(Fire Area 18) and the Unit 2 electrical penetration room (Fire Area 16) was declassified
in Revision 9 (2000). The wall between the Unit 2 switchgear room 2ETA (Fire Area 18)
and the Unit 2 HVAC equipment room 805A (Fire Area 18) was declassified in Rev. 3
(1982). For the purposes of the inspection of Fire Area 18, the electrical penetration
room (Fire Area 16) was included in the inspection plan because the fire wall separating
these areas has been declassified and is no longer a "Fire Sealed - NRC Committed"
fire barrier. The similar wall at Unit 1 Room 803A was also declassified from a "Sealed
Firewall - NRC Committed" to a "Sealed Firewall - Non Committed."
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 electrical
raceway fire barrier system (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 GL 86-10 fire
barrier features, and NFPA code deviations to verify that the fire barrier installations met
design requirements and license commitments. In addition, the team reviewed
surveillance and maintenance procedures for selected fire barrier features to verify the
fire barriers were being adequately maintained.
b.
Findings
No findings of significance were identified.
.10
Fire Protection Systems. Features, and Eguipment
a.
Inspection Scoge
The team reviewed UFSAR Section 9.5.1, the fire protection design basis specification,
fire protection code deviations, and administrative procedures used to prevent fires and
control combustible hazards and ignition sources. This review was performed to verify
that the objectives established by the NRC-approved FPP were satisfied. The team also
toured the selected plant fire areas to observe the licensee's implementation of these
procedures.
The team reviewed the adequacy of the design and installation of the automatic wet
pipe sprinkler system protecting the RN pumps in Fire Area 4. Team members
.15
performed a walk down of the system to ensure proper placement and'spacing of the
sprinkler heads and the extent of the'sprinkler head obstructions. Selected engineering
evaluations for NFPA code deviations were reviewed and compared with the physical
configuration of the system. The team reviewed the sprinkler system hydraulic
calculations for this system to ensure that the system could be supplied sufficient
pressure and volume utilizing the two by-pass lines without opening the deluge valves.
The team also inspected one of the by-pass lines located in an outside pit to determine
the piping and fitting equivalent length to confirm the accuracy of the design input to the
RN pump calculation. The team reviewed the fire protection code deviations calculation
for automatic suppression systems relative to the selected fire areas.
The team reviewed the adequacy of the design and installation of the automatic
detection and alarm system for the selected fire areas. This was accomplished by
reviewing the ceiling reinforcing plans and beam schedule drawings to determine the
location of ceiling bays. After the ceiling bay locations were identified, the team
conducted a plant tour to confirm that each bay was protected by a fire detector in
accordance with the Code of Record requirements - NFPA 72E, 1974. .Field tours were
conducted in fire areas 13, 16/18 to confirm detector locations. Minor modification
-package MM-1 2907 was reviewed where 10 new detectors were added to Fire Area 13
to conform the detection system to NFPA 72E location requirements.
The team reviewed the fire protection code deviations calculation for automatic
detection systems relative to the selected areas to determine if there were any code
deviations cited for the selected fire areas. The team reviewed the fire protection pre-
plans and fire strategies to ensure that hose locations could sufficiently reach the
selected fire areas for manual fire fighting efforts. Hose stations in the selected area
were inspected to ensure that hose lengths depicted on the engineering documents
were also the hose lengths located in the field. This was done to ensure that manual
fire fighting efforts could be accomplished in the selected fire areas.
b.
Findins-
No findings of significance were identified.
4.
OtherActivities -
'
-r
40A2 Problem Identification and Resolution
a.
Inspection Scope
The team reviewed a sample of licensee audits, self-assessments, and PIPs to verify
that items related to fire protection and to SSD were appropriately entered into the
licensee's corrective action program in accordance with the MNS quality assurance
program and procedural requirements The items selected were reviewed for
classificatiorni appropriateness, and timeliness of the corrective actions taken, or
Initiated, to resolve the issues. Included in this'review were PlPs G-99-001 10, M-99-,-
lI
16
01884, M-99-01886, M-03-01675, and minor modification MM-1 2907 related to the
McGuire Fire Protection Functional Audit SA-99-04(MC)(RA)(FPFA).
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 reports
were reviewed to verify that the licensee's review and actions were appropriate.
b.
Findings
No findings of significance were identified.
40A5 Other Activities
.01
(Closed) URI 50-369.370/00-09-04: Adequacy of the Fire Rating of Mineral Insulated
Cables in Lieu of Thermo-Lag Electrical Raceway Fire Barrier Systems
The NRC had opened this URI for further NRC review of the adequacy of the fire
resistance rating of certain mineral insulated cables that the licensee had installed. The
licensee had replaced an inadequate 3-hour Thermo-Lag fire barrier with mineral
insulated cables for charging pump 1A in the Unit 1 Train B switchgear room. However,
the adequacy of the testing of the mineral insulated cables, to assure their 3-hour fire
resistance ability, had not been reviewed by the NRC.
The inspectors reviewed the NRC SER of January 13, 2003, on the licensee's use of
mineral insulated cables and also reviewed the licensee's 10 CFR 50.59 safety
evaluation for the modification. The NRC SER evaluated the licensee's installation and
fire testing of the mineral insulated cables and concluded that the licensee had
adequately demonstrated that the protection provided by the mineral insulated cables in
the specific application was equivalent to the protection provided by a 3-hour rated fire
barrier. The NRC SER further concluded that this change to the approved fire
protection program did not adversely affect the ability to achieve and maintain safe
shutdown in the event of a fire and, therefore, did not require prior approval of the NRC.
The inspectors concluded that the licensee's 50.59 safety evaluation for the change had
adequately considered that the change did not adversely affect the ability to achieve and
maintain safe shutdown in the event of a fire. Consequently, the licensee's installation
of mineral insulated cables was not a violation of NRC requirements.-ThIsgURI is
closed.
40A6 Meetings
On May 23, 2003, the team presented the inspection results to you and other members
of your staff, who acknowledged the findings. The team confirmed that proprietary
information is not included in this report.
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
D. Bailey, Mechanical and Civil Engineering (MCE) - Civil
J. Boyle, Training Manager
S. Bradshaw, Superintendent of Operations
H. Brandes, Consulting Engineer, General Office Fire Protection Program
J. Bryant, Regulatory Compliance Engineer.
B. Dolan, Safety Assurance Manager
J. Hackney, Operations
T. Harrell, McGuire Station Manager
D. Henneke, Engineer, General Office Probabilistic and Risk Assessment Group
D. Herrick, Civil Engineering Supervisor
D. Jamil, Site Vice President, McGuire Nuclear Station
R. Johansen, Standby Shutdown Facility System Engineer
J. Lukowski, Reactor Electrical Systems (RES) - Power
E. Merritt, RES - Instrumentation and Controls
J. Oldham, Fire Protection Engineer, MCE - Civil
B. Peele, Station Engineering Manager
G. Peterson, Site Vice President, Catawba Nuclear Station
C. Thomas, Regulatory Compliance Manager
NRC Personnel
J. Brady, Senior Resident Inspector, Shearon Harris
E. DiPaolo, Resident Inspector
R. Fanner, Nuclear Safety Intern (Trainee)
C. Ogle, Chief, Engineering Branch 1, Division of Reactor Safety, Region II
R. Rodriguez, Nuclear Safety Intem (Trainee)
S. Shaeffer, Senior Resident Inspector
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
50-369,370103-07-01
Fire Suppression System for Dedicated Shutdown Areas
Not in Accordance with 10 CFR 50, Appendix R, Section
III.G.3 (Section 1R05.02.b)
50-369,370/03-07-02
Failure to Protect Redundant Trains of Reactor Protection
System Cables From the Effects of Fire (Section
1R05.03.b.1)
Attachment
I
2
50-369,370/03-07-03
50-369,370103-07-04
50-370/03-07-05
50-370/03-07-06
Requirements Relative to the Number of Spurious
Operations that must be Postulated (Section 1 R05.04.b.1)
Methods for Reactor Coolant System Pressure Control
During SSF Operation (Section 1 R05.05.b)
Failure to Provide Adequate Protection for Cables of
Redundant Safe Shutdown Equipment In Fire Area 16/18
(Section 1RR05;03.b.2)
Spurious Closure of Valve 2CA0007A Could Lead to
Damage of the TDAFW Pump (Section 1 R05.04.b.2)
Closed
50-369,370/00-09-04
Adequacy of the Fire Rating of Mineral Insulated Cables in
Lieu of Thermo-Lag Electrical Raceway Fire Barrier
Systems (Section 40A5.01)
Discussed
None
Attachment
-
APPENDIX
LIST OF DOCUMENTS REVIEWED
Section 1 R05: Fire Protection
Procedures"'
AP10/A15500/045, Plant Fire, Rev. 0 and Rev. 2
AP111A15500/024, Loss of Plant Control Due to Fire or Sabotage, Rev. 21
AP12/A155001024, Lois of Plant Control Due to Fire or Sabotage, Rev. 20
NSD 112, Fire Brigade Organization, Training, and Responsibilities, Rev. 5
NSD 313, Control of Combustible'and Flammable Material, Rev. 4
NSD 314, Hot Woik Authorization ,'Rev. 2
NSD 316, Fire Protection Impairment and Surveillance, Re'v. 6'
MP/01/A76501122, Inspection of Fire Hose and Hydrant Houses, Rev. 5
OP/0/A/6100/020, Operational Guidelines Following a Fire In Aux Bldg or Vital Area, Rev. 16
PTI0/A/4250/004, Fire Barrier Inspection, Rev. 19
PTI01A14250/01 1,'Fire Door Inspections, Rev. 14
PT/0/AN4250/020, Roll-Up Fire Door Semi-Annual Inspection/Test, Rev. 2
PT/0/A/4400/001A, Fire Protection System Periodic Test, Rev. 24
PT/O/A14400/001 C, Fire Protection System Monthly Test, Rev. 54
PT/0NA/4400/001 K, Fire Protection Annual Valve Test, Rev. 35
PT/O/A/4400/001M, Fire Protection System' Flow Test, Rev. 14
PT/01A/4400/008,'Fire Hose Hydrostatic Test SLC-Committed Hose Stations, Rev. 11
PT/0/A/44001010A, Main Fire Pump A, Rev. 15
PT/0/A14400/O1OB,'Main Fire Pump B, Rev. 10'
PT/0/A/4400/01 DC, Main Fire Pump C, Rev. 11
PT/0/A14400/017, Fire Pump A and B Operability Test, Rev. 13
PT/0/A/4400/018, Fire Pump C Operability Test,;Rev.11 I
PT/11/A4400/OOIL, Fire Protection Containment HeaderTest, Rev. 9
PTI1/A/4400/001 N, Halon 1301 System Periodic Test, Rev. 29
PT/2/A/4400/001 L, Fire Protection Containment Header Test, Rev. 7
PT/0/AN4600/016A, Fire Detection System'Operational Tests, Rev. 18
PT/0/B/4600/015, Fire Detection System Monthly Test, Rev. 14
PT/0/A14700/049, SLC Fire Hose Inspection,- Rev; 1
PT/11/A4700/042, SLC Fire-Hose Station Valve Operability Test, Rev. 3
PT/2/A/4700/043, SLC Fire Hose StationValve'Operability Test, Rev. 3
PT/11A/4150/001 B, Reactor Coolant Leakage Calculation, Rev. 47
Drawings
MC-1042-4, General ArrangementAuxilary Building, -Elevation 750+0, Rev.6
MC-1201-2-A, General Arrangement, Auxiliary Building, Elevation 716+0, Rev.-67
MC-1201-3-A, General Arrangement, Auxiliary Building, Elevation 716+0, Rev. 67
Attachment
2
MC-1201-4, General Arrangement, Auxiliary Building, Elevation 733+0, Rev. 27
MC-1223-38, Auxiliary Building, Unit 1 & Unit 2, Beam Schedule at Elevation 733+0, Concrete
and Reinforcing, Sheet 1, Rev. 4
MC-1223-39, Auxiliary Building, Unit 1 & Unit 2, Beam Schedule at Elevation 733+0, Concrete t
and Reinforcing Sheet 2, Rev. 6
MC-1223-6, Auxiliary Building, Unit 1, Plan at Elevation 733+0, Reinforcing Sheet 1, Rev. 8
MC-1223-7, Auxiliary Building, Unit 2, Plan at Elevation 733+0, Reinforcing Sheet 2, Rev. 5
MC-1223-8, Auxiliary Building, Unit 1, Plan at Elevation 733+0, Reinforcing Sheet 3, Rev. 6
MC-1223-9, Auxiliary Building, Unit 2, Plan at Elevation 733+0, Reinforcing Sheet 4, Rev. 6
MC-1223-27, Auxiliary Building, Units 1 & 2, Sections at Elevation 733+0, Concrete Sheet 3-1,
Rev. 27
MC-1224-9, Auxiliary Building Unit 1, Plan at Elevation 750+0, Reinforcing Sheet 3, Rev. 9
MC-1224-10, Auxiliary Building Unit 1, Plan at Elevation 750+0, Reinforcing Sheet 4, Rev. 10
MC-1224-39, Auxiliary Building, Beam Schedule at Elevation 750+0, Concrete & Reinforcing
Sheet 1, Rev. 6
MC-1 225-1 0, Auxiliary Building Unit 2, Plan at Elevation 767+0, Reinforcing Sheet 4, Rev. 5
MC-1225-11, Auxiliary Building, Plan at Elevation 767+0, Reinforcing Sheet 5, Rev. 4
MC-1225-39, Auxiliary Building, Beam Schedule at Elevation 767+0, Concrete & Reinforcing,
Rev. 6
MC-1225-40, Auxiliary Building, Beam Schedule at Elevation 767+0, Concrete & Reinforcing,
Sheet 2, Rev. 5
MC-1226-8, Auxiliary Building, Plan at Elevation 784+0, Reinforcing Sheet 3, Rev. 1
MC-1226-9, Auxiliary Building, Plan at Elevation 784+0, Reinforcing Sheet 4, Rev. 2
MC-1226-19, Auxiliary Building, Beam Schedule at Elevation 784+0, Concrete and Reinforcing,
Rev. 1
MC-1315-01.02-105, General Arrangement, Fire, Flood & HVAC Boundaries, Elevation 716+0,
Rev. 0
MC-1384-06.02, Fire Protection Layout, Plan at Elevation 716+0, Rev. 7
MC-1 384-06.03, Fire Protection Layout, Plan at Elevation 733+0, Rev. 7
MC-1384-06.04, Fire Protection Layout, Plan at Elevation 750+0, Rev. 7
MC-1 384-06.05, Fire Protection Layout, Plan at Elevation 767+0, Rev. 7
MC-1384-07.12-00, Fire
MC-1384-07.01-00, Fire
MC-1384-07.13-00, Fire
MC-1384-07.13-01, Fire
MC-1384-07.14-00, Fire
MC-1384-07.14-01, Fire
MC-1384-07.14-02, Fire
MC-1384-07.14-03, Fire
MC-1384-07.15-00, Fire
MC-1384-07.15-01, Fire
MC-1384-07.15-01, Fire
MC-1384-07.15-01, Fire
MC-1384-07.15-02, Fire
Plan, Auxiliary Building, Elevation 695+0, Rev. 3
Plan, Unit 1 Turbine Building, Elevation 739+0, Rev. 11
Plan, Auxiliary Building, Elevation 716+0, Rev. 12
Plan, Auxiliary Building, Elevation-716+0, Rev. 9.
Plan, Auxiliary Building, Elevation 733+0, Rev.J 2
Plan, Auxiffary. Building, Elevation 733+0, Rev. 9
Plan, Auxiliary Building, Elevation 733+0 & 736+6, Rev. 9
Plan, Auxiliary Building, Elevation 733+0 & 736+6, Rev. 9
Plan, Auxiliary Building, Elevation 750+0, Rev. 10
Plan, Auxiliary Building, Elevation 750+0, Rev. 2
Plan, Auxiliary Building, Elevation 750+0, Rev. 3
Plan, Auxiliary Building, Elevation 750+0, Rev. 9
Plan, Auxiliary Building, Elevation 750+0, Rev. 10
MC-1 384-07-.16-00, Fire Plan, Auxiliary Building, Elevation 760+6, Rev. 7
Attachment
3
MC-1384-07.17-00, Fire Plan, Auxiliary Building,' Elevation 767+0, Rev. 10
MC-1384-07.17-01, Fire Plan, Auxiliary Building,'Elevation 767+0, Rev. 9
'
MC-1 384-07.18-01, Fire Plan, Auxiliary Building, Elevation 778+10, Rev. 8
MC-1 518-06.43-00, Piping Layout, Interior Fire Protection, Nuclear Service Water Pumps,
Sprinkler Addition, Rev.'1
MC-1518-06.43-01, Piping Layout, Interior Fire Protection, Component Cooling Pumps,
Sprinkler Addition, Rev.
'
MC-1518-25.85-01, Piping Layout, Service Water Piping, Outside Pumphouse, Rev. 29
MC-1 710-01.00, Plan, Control Room Computer Room, Elevation 767+0, Rev. 49
MC-1710-04.08, Battery Room Junction Points Elevation 747, Rev. 15
MC-1710-04.09, Battery Room Junction Points Elevation 746, Rev. 23
MC-1710-04.10, Battery Room Junction Points Elevation 745, Rev. 20'
MC-1710-04.11, Battery Room Junction Points Elevation 744, Rev. 24
MC-1710-04.12, Battery Room Junction Points Elevation 743, Rev. 22
MC-1710-04.13, Battery Room Junction Points Elevation 742, Rev.24
MC-1710-04.14, Battery Room Junction Points Elevation 741, Rev. 23
MC-1710-04.15, Battery Room Junction Points Elevation 740, Rev. 23
MC-1762-01.00-02, Location Diagram, Fire Detectors Located on Elevation 716+0, Rev. 7
MC-1762-01.00-03, Location Diagram, Fire Detectors Located on'Elevations 733+0'& 739+0,
Rev. 10
MC-1762-01.00-04, Location Diagram, Fire Detectors Located on Elevation 750+0, Rev. 10
MC-1762-01.00-06, Location Diagram, Fire Detectors Located on Elevations 760+6 & 767+0,
Rev. 13
MC-2901-01.01, Auxiliary Building Plan Below Elevation 733'+0, Rev. 44
MC-2907-01.01, Penetration and Switchgear Rooms Plan Below Elevation 776'+0, Rev. 25
MCEE-138-00.02, Turbine Driven AFW Suction Supply Valve, Rev. 5
MCEE-138-00.04, Turbine-driven AFW Suction'Supply Valve, Rev. 11
MCEE-138-00-01, Turbine Driven AFW Suction Supply Valve, Rev. 5
MCEE-211-00.52, Pressurizer Heaters, Rev. 2
MCEE-211-00.52-01, Pressurizer Heaters, Rev 9
MCEE-211-00.52-02, Pressurizer Heaters,-Rev. 87
MCEE-211-00.52-03, Pressurizer Heaters, Rev. 9
MCEE-211-00.52-04, Pressurizer Heaters, Rev. 4
MCEE-211-00.52-05, Pressurizer Heaters, Rev. 3-
MCEE-244-02.01, Steam Generator Level and Pressurizer Level, Rev. 4
MCEE-247-10.00, Motor Driven AFW Isolation Valve, Rev. 0
MCEE-247-20.00, Turbine Driven AFW Isolation Valve, Rev.-0 -
MCEE-247-20.01, Turbine Driven AFW Isolation Valve, Rev. 0
MCEE-247-32.00, Turbine-driven AFW Isolation Valve, Rev. 1
MCEE-247-33.00, Turbine Driven AFW Isolation Valve, Rev. OA
MCEE-250-00.03, Pressurizer Power-operated Relief Valve
MCEE-250-00.03-01, Pressurizer Power-operated Relief Valve
MCEE-250-00.06, Pressurizer Power-operated Relief Valve Isolation Valve
MCEE-250-00.24, Unit 2.Chemical and Volume Control Isolation Valve, Rev. 01
MCEE-250-00.28, Reactor Vessel Head Vent Valves, Rev. 6
Attachment
4
MCEE-250-00.29,
MCEE-250-00.33,
MCEE-257.00.54,
MCEE-257-00.24,
MCEE-257-00.50,
MCEE-257-00.52,
MCEE-257-00.55,
MCFD-1 574-01.00
MCFD-1 574-01.01
MCFD-1 599-01.00
MCFD-1 599-01.01
MCFD-1599-02.00
MCFD-1599-02.01
MCFD-1599-02.02
MCFD-1599-02.03
MCFD-1599-03.00
MCFD-1599-03.01
MCFD-2574-02.00
MCFD-2574-02.01
MCFD-2592-01.01
MCFD-2592-02.00
Reactor Vessel Head Vent Valves, Rev. 5
Reactor Vessel Head Vent Valves, Rev. 5
Chemical and Volume Control Containment Isolation'
Chemical and Volume Control Containment Isolation'
Unit 2 Chemical and Volume Control Isolation Valve,
Chemical and Volume Control Isolation Valve, Rev. 1
Standby Makeup Pump, Rev. 1
), Nuclear Service Water, Rev. 6
Nuclear Service Water, Rev. 10
), P&ID, Flow Diagram of Fire Protection, Rev. 13
P&ID, Flow Diagram of Fire Protection, Rev. 14
), P&ID, Flow Diagram of Fire Protection, Rev. 15
P&ID, Flow Diagram of Fire Protection, Rev. 15
, P&ID, Flow Diagram of Fire Protection, Rev. 5
, P&ID, Flow Diagram of Fire Protection, Rev. 6
I, P&ID, Flow Diagram of Fire Protection, Rev. 7
P&ID, Flow Diagram of Fire Protection, Rev. 3
, Nuclear Service Water, Rev. 12
Nuclear Service Water, Rev. 2
Auxiliary Feedwater System, Rev. 13
, Auxiliary Feedwater System, Rev. 2
Valve, Rev. 3
Valve, Rev. 5
Rev. 6
MCM.1206.07-0074.001, McNeary Insurance Consulting Services, FP-12
MCM.1206.07-0087.001, McNeary Insurance Consulting Services, FP-18
Compgleted Maintenance And Surveillance Test Procedures/Records
Work Order 98410020, PT 2NCLP5151, SSF Pressurizer Level, dated 3/13/02
Work Order 98410021, PT 2NCLP5121 NC Loop D Hot Leg W/R Pressure, dated 3/13/02
Work Order 98410083, PM 2CFLP61 10, S/G D W/R Level, dated 2/28/02
Work Order 98410084, PM 2CFLP6100, S/G C W/R Level, dated 3/5/02
Work Order 98410085, PM 2CFLP6090, S/G B W/R Level, dated 3/1/02
Work Order 98410086, PM 2CFLP6080, S/G A W/R Level, dated 2/28/02
Cable Installation Data for the Following Components
2CA0007A
2CA009B
2CFLT6080, 6090, 6100, 6110
2NC272AC, 273AC
2NC33A, 35B
2NCLT5151
2NV1012C
2NV842AC
2NV94AC
2NVPU0046
Attachment
5
ORN4AC
Calculations and Evaluations
MCC-1223.04-00-0010, Determine the Reactor Coolant Pump Sealwater Flow Requirements
for the'SSF Auxiliary Makeup Pump,Type II.
MCC-1223.42-00-0030, Documentation of the Adequacy of the Assured Suction Sources to the
CA Pumps, Rev. 8
MCC-1223.49-00-0030, Sprinkler System for Nuclear Service Water Pumps @'Elevation 716-0,
Rev. 0
MCC-1435.00-00-0006, Calculation for the Technical Basis of Fire Barrier Penetration Seals,
Rev.1
MCC-1435.03-00-0002, Fire Exposure to Unprotected Steel Hangers for HVAC Ducts, Rev. 2
MCC-1435.03-00-0004, Supports for Cable Tray Penetrating Fire Barriers, Rev. 0
MCC-1435.03-00-0012, MNS Penetration Seal Database and GL 86-10 Evaluations, Rev. 0
MCC-1435.03-00-0013, Fire Protectionode Deviations, Rev. 0'
MCS-1 435 00-00-0001, Fire Protection Accepta hc6 Specification, Rev. 17
MCS-1435.00.00-0003, Design Specification for Mechanical and Electrical Penetrations; Fire
Flood and Pressure Seals
National Fire Codes - Volume 1, Codes & Standards: NFPA 13 - Standard for the Installation of
Sprinkler Systems, 1978 Edition
Design Basis Document
MCS-1223.SS-00-0001, Design Basis Specification for the Standby Shutdown System,- Rev. 12
MCS-1465.00-00-0008, Design Basis Specification for Fire Protection, Rev. 4.
MCS-1465.00-00-0022, Design Basis Specification for Appendix R, Rev. 2
Problem Investigation Process Reports Reviewed
G-99-001 10, McGuire Fire Protection Functional Audit (SITA) SA-99-04(MC)(RA)(FPFA).
M-97-0331 1, All three CA pumps may have been dead headed during the U1 Rx trip recovery.
M-99-01884, GL 86-10 guidance for circdit failure modes, hot short duration, and design basis
-transientsfor-dedicated shutdown not evaluated for applicability to MNS methodology.
M-99-01886, NFPA cdeaviations not'doctmented in UFSAR'or FHA as per GL 86-10.:
M-99-03926, Effect of warmer seal -injection water on RCP seals du'ring SSF event not
adequately taken into consideration bn-SMP-capacity. Ev6aluate-applicability to McGuire.
M-00-01900, Unit 1 CA pumps normal suction sources inadvertently isolated following a reactor
trip and automatically aligned to RN.
M-00-04466, Evaluate UFSAR Section'9.5-1 Clarifications for Fire Suppression Systems.'
M-00-04469, Evaluate Fire Pump Loss Due to Fire in Fire Area 19 and Main Control Room.
M-00-04483, The fire protection RY by-pass lines around 1RY 113 and 1 RY 114 do not Pe'rmit
the Maximum Flow for the Largest Sprinkler Demand.'
M-00-04487, Fire Brigade Drills'Had Not Been Performed Within 10 Years in Areas Considered
-Safety Significant.
Attachment
6
M-00-04491, NRC Appendix R inspection in certain fire areas determined the potential for NC
PORV and block valve actuation. We need to evaluate this cabling as to "if' this will occur.
M-00-04516, Adequacy of Pzr heater capacity at SSF due to increase safety valve leakage.
M-02-01708, It has been discovered that pressurizer ambient heat losses are greater than
calculated in OSC-3144 impacting SSF ASW system operability (TS 3.10.1 and TS 3.4.9).
M-02-03214, SSS and NC DBDs identified errors related to pressurizer heater requirements.
M-02-05031, RO closed 1 CA-0002, resulted in temp low suction flow to running 1 B CA pump.
M-02-05096, Information on system problem [PIP M-02-05031] not documented for resolution.
M-03-01675, Fire Detection System Not Installed to NFPA Codes.
M-03-01748, Smoldering fire on roof of Unit 1 Diesel Generator building.
Prblem Investigation Process Reports Generated During This Inspection
M-03-02084, Fire scenarios that could cause suction loss to U2 TDCA pump for SSF areas.
M-03-02086, Discrepancy between Appendix R DBD and Procedure AP12/A15500124.
M-03-02091, Unit 1 and Unit 2 HVAC areas do not have fire detectors.
M-03-02092, Discrepancy between drawings and fire pre-plans for fire hose lengths.
M-03-02093, Drawing discrepancy for as-built configuration of HVAC Equipment Room 805A.
M-03-02106, B train cables in A SWGR room Fire Area which are not previously identified.
M-03-02115, Appendix R logic diagrams not updated to show function of valve 2CA002.
M-03-02118, Appendix R logics for AFW do not show valve 2CA0007A.
M-03-02249, Detector zones 203 and 204 not in SLC 16.9.6, Table 16.9.6-1.
M-03-02275, Calculation (MCC 1223.48-00-0030) in support of sprinkler system design over
the nuclear service water pumps needs revising.
M-03-02294, SLC Table 16.9.7-1 appears to be missing some information.
M-03-0231 1; Evaluate May 2003 NRC Fire Protection Inspection items.
M-03-02327, Calc MCC-1435.03-00-0002 contains deleted pages not marked as being deleted.
M-03-02588, Apparent Appendix R violation in the 1 ETA and 2ETA switchgear HVAC rooms.
Miscellaneous
MNS Units 1 and 2 Safety Evaluation Report (SER), March 1978.
SER Supplement 2 (SSER 2), Appendix D, Fire Protection Review, Units 1 & 2, March 1979
SSER 5, Appendix B, McGuire SER, Fire Protection Review, Unit I & 2 (Revised), April 1981
SSER6, Appendix C, McGuire SER - Standby Shutdown System, -ebruary 1983U'
MNS Updated Final Safety Analysis Report (UFSAR)_Sectionr9.5.1, Fire Protection System
UFSAR Section 16.9.7, Selected Licensee Commitments (SLC), Standby Shutdown System
Letter from W.O. Parker, Duke Power Co., to H.R. Denton, NRC, McGuire Nuclear Station Fire
Protection, dated January 9, 1981
Letter from D.S. Hood, NRC, to H. B. Tucker, Duke Power Co., Fire Protection Deviations,
McGuire Nuclear Station, Units 1 and 2, dated May 15, 1989
Fire Area Ventilation Rates, Fire Areas 4, 13, 18 & 24
Fire Area Oil Quantities, Fire Area 4, 13, 18 & 24
Fire Area 4 Correlation List between Rooms Number vs. Detection Zones
Fire Qualification Test on Silicone Foam Floor Pen Seals, Slab No. 5, Project No. 03-5656-001
Attachment
Applicable Codes and Standards
NFPA 13, Standard for the Installation of Sprinkler Systems, 1978 Edition
NFPA 14, Standard for the Installation of Standpipe and Hose Systems, 1976 Edition
NFPA 72E, Standard on Automatic Fire Detectors, 1974 Edition
Modifications
Minor Modification MM-12907A thru F
-
!i
-X.
5 -
Attachment
-
LIST OF ACRONYMS
Auxiliary Building
Abnormal Procedure
DSD
Dedicated Shutdown
Electrical Raceway Fire Barrier System
Fire Hazards Analysis
Fire Protection Review
GL
Generic Letter
Heating Ventilation and Air Conditioning
Individual Plant Examination for External Events
IR
Inspection Report
kW
Kilowatt
Main Control Room
McGuire Nuclear Station
NC
National Fire Protection Association
NRC
Nuclear Regulatory Commission
NRC Office of Nuclear Reactor Regulation
NSD
Nuclear System Directive
NV
Chemical and Volume Control
Problem Investigation Process
Power Operated Relief Valve
Reactor Coolant Pump
RN
Nuclear Service Water
Significance Determination Process
Safety Evaluation Report
Selected Licensee Commitment
Standby Makeup Pump
Safe Shutdown Analysis
SSD
SSF
Standby Shutdown Facility
Standby Shutdown System
Turbine-Driven Auxiliary Feedwater
TS
Technical Specifications
Updated Final Safety Analysis Report
Unresolved Item
V
Volt
Attachment