ML033020068
| ML033020068 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 10/10/2003 |
| From: | Ogle C NRC/RGN-II/DRS/EB |
| To: | Gordon Peterson Duke Energy Corp |
| References | |
| FOIA/PA-2003-0358 IR-03-007 | |
| Download: ML033020068 (30) | |
See also: IR 05000369/2003007
Text
- - -ws ~~~~~UNITED STATES
- NUCLEAR REGULATORY COMMISSION
go g { ~~~~~~~REGION II
S z ma
t ~~~SAM NUNN ATLANA FEDERAL CENTER
Be sLW,{
a ~~~61FORSYTH STREET SW SUITE 23T85
o t~ ATLANTA, GEORGIA 30303-8931
Duke Energy Corporation
ATTN: Mr. G. Peterson
Vice President
MoGuire 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. Peterson:
On May 23, 2003, the Nuclear Regulatory Commission (NRC) completed an inspection at your
McGuire Nuclear Station, Units 1 and 2. An interim exit was held with Mr. D. Jamil and other
-members of your staff on May 22, 2003, to discuss the results of that effort. Following
completion of additional review in the Region II office, a final exit was held with you and other
members of your staff on July 2,2003. The enclosed report documents our findings from this
inspection.
The inspection examined activities conducted under your licenses as they relate to safety and
compliance with the Commission's rules and regulations and with the conditions of your
licenses. The inspectors reviewed selected procedures and records, ob~served activities, and
interviewed personnel. 25sc7IAZte (7,e 4.-F /;/
This report documents three findings that have pot tal safety significance greater than very
low significance, however, a safety significance d termination has not been completed. These
findings did not present an immediate safety con emn at the time of the interim exit. However,
your subsequent analyses of one of the findings resulted in identification of additional cables
associated with reactor protection system in .tnruentatikn (and possibly other equipment)
required for safe shutdown located in the same fire area that could be susceptible to fire
damage. Upon discovery of this condition on June 10, 2003, a fire watch was established as a
compensatory measure.
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 Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001, with
copies to the Regional Administrator, Region II; 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 bRules 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 b
DEC 2
NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.gov/readina-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
Charles R. Ogle, Chief,
Engineering Branch 1
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 (EC11X)
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
(cc w/encl cont'd - See page 3)
DEC 3
(cc w/encl 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
U.S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos.: 50-369, 50-370
Report No.: 50-369/03-07 and 50-370/03-07
Licensee: Duke Energy Corporation
Facility: McGuire Nuclear Station
Location: 12700 Hagers Ferry Road
Huntersville, NC 28078
Dates: May 5 - 9, 2003 (Week 1)
May 19 - 23, 2003 (Week 2)
Inspectors: 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. NRC-Identified and Self-Revealing Findings ,(' ,i /6/b)
Cornerstone: Mitigating Systems
- TBD The team identified a violation in that Train A and Train B ables associated with
redundant reactor protection system instrumentation (and poss ly other equipment)
important to safe shutdown 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. 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
1R05.03.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 analysis for potential impact on safe shutdown in the event of a fire
where the TDAFW pump is required for 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. This
finding may have potential safety significance greater than very low significance
because the standby shutdown system relies on the TDAFW pump for decay heat
removal, and the decay heat removal function would be seriously degraded if the
TDAFW pump were damaged due to closure of valve 2CA0007A. (Section 1R05.04.b.2)
2
B. Licen e-e tified 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-informed approach for selecting the fire areas and attributes to be
inspected. The team used the licensee's Individual Plant Examination for Extemal
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 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
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.7, Fire Protection Program, for Units 1 and
2, respectively; Title 10 of the Code of Federal Regulations, Part 50 (10 CFR 50),
Appendix R, Sections III. G, J, L, and 0; 10 CFR 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 Commitments (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
a. Inspection Scone
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
2
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 valve 2NC33A
- Pressurizer heaters Nos. 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 and 16/18 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
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.
3
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. Findings
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 Ill.G.3 (alternative
or dedicated shutdown) requires that fire detection and a fixed fire suppression system
shall be installed in the area, room, or zone under consideration. However, the fire
suppression system in Fire Area 4 was a partial automatic sprinkler system designed to
protect the RN pumps and the area 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, Section III.G.3.
This issue was previously identified by the NRC in 1984 during an Appendix R
inspection (URI 50-369/84-28-01, 370/84-25-01). 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 alternative or dedicated shutdown 2
capability as required by 10 CFR 50, Appendix R, Sect.ri G,.; Dubie In e c-cttrcete-- /
inspection, the team questioned whether the previousreviews of the srinkler system for
this fire area included an evaluation of the risk impact associated with not providing
adequate sprinkler coverage for the RN cab jing 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.
.03 Post-Fire Safe Shutdown Circuit Analysis
a. Inspection ScoDe
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
4
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. Circuit and cable routings were reviewed for the
following equipment:
- ORN4AC, Turbine Driven AFW Suction Supply Valve
- 2CA0007A, Turbine Driven AFW Suction Isolation Valve
- 2CA009B, Motor Driven AFW Suction Isolation Valve
- 2CFLT6080, 6090, 6100, 6110, Steam Generator Level Transmitters
- 2NCLT5151, Pressurizer Level Transmitter
- 2NC34A, Pressurizer PORV
- 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
- 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. Findinas
Findings associated with valves 2CA0007A, 2NC34A, and 2NC33A are discussed in
Section .04 of this IR.
1. Inadequate Separation of Cables Associated With Safe Shutdown Instrumentation
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 File Area 16/18. This finding involved a violation of
NRC requirements. This finding is a URI pending a determination of the systems
affected by the licensee and completion of the significance determination process.
(SDP).
Description: Fire Area 16/18 is the Unit 2 Train A electrical penetration room/2ETA
4160 volt (V)switchgear room and the associated HVAC equipment room 805A. 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 (i.e., this fire area complies with 10
CFR 50, Appendix R,Section III.G.2). 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 was not aware
that these Train B cables passed through room 805A, and initiated Problem
5
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 supplying ventilation for the Unit 1 Train A 41 60V switchgear room
1ETA]. 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). The licensee subsequently 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 in room 805A and could be damaged during a severe fire. As many as
74 Train B RPS cables may be involved. One consequence 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. The
safety injection signal could in turn trigger a reactor trip and Phase A isolation. [At the
same time, many main control panel instruments necessary to achieve and maintain hot
s1tddQ.wehld 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 1 Train A switchgear room 1ETA (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 licensee is analyzing the manner in
which plant systems would be affected by fire damage 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 SDP will be used to determine the level of
significance. When assessed in combination 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 very low
significance.
Enforcement: The licensee's FPP commits to 10 CFR 50, Appendix R.Section III.G.
Section III.G.2 requires in part, that cables or equipment for one of the redundant trains
of a system necessary to achieve and maintain hot shutdown (located in the same fire
area) shall be ensured to be free of fire damage by one of the following: (1) separated
by a 3-hour rated fire barrier; (2) separated by 20-feet or more horizontal distance with
no intervening combustibles or fire hazards, and having suppression and detection; or
(3) enclosure of the cables in a 1-hour rated fire barrier and having suppression and
detection.
Contrary to the above, electrical cables associated with redundant trains of RPS
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 systems
affected and the safety significance, the finding is identified as URI 50-369, 370/03-07-
02, Inadequate Separation and Protection of Cables Associated With Redundant Trains
of Instrumentation Located in the Same Fire Area.
6
2. Inadequate Protection of AFW Cables and Equipment Required for Safe Shutdown
Introduction: A finding was identified in that physical protection of the electrical cables
associated with 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: The licensee identified (April 2003) 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-03-0231 1. 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 physical protection
for redundant trains of equipment/systems necessary to achieve and maintain SSD (as
specified for Appendix R,Section III.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.
Enforcement: The licensee's FPP commits to 10 CFR 50, Appendix R, Section III.G.
Section III.G.2 requires in part, that cables or equipment for one of the redundant trains
of a system necessary to achieve and maintain hot shutdown (located in the same fire
area) shall be ensured to be free of fire damage by one of the following: (1) separated
by a 3-hour rated fire barrier; (2) separated by 20-feet or more horizontal distance with
no intervening -combustibles or fire hazards, and having suppression and detection; or
(3) enclosure of the cables in a 1-hour rated fire barrier and having suppression and
detection.
Contrary to the above, on May 23, 2003, the licensee failed to protect electrical cables
associated with redundant equipment located within the Unit 2 Train A electrical
penetration room (Fire Area 16/18) with an adequate barrier or to provide 20 feet of
separation. Instead, the licensee used a local manual operator action, which had not
7
received prior NRC approval, to achieve and maintain SSD. Pending determination of
the finding's safety significance, this finding is identified as URI 50-370/03-07-03, Use of
a Local Manual Operator Action in Lieu of Providing Physical Protection for Cables of
Redundant Safe Shutdown Equipment in Fire Area 16/18.
.04 Alternative Post-Fire Safe Shutdown Capability
a. Inspection Scone
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 offsite power available.
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
1. Requirements Relative to the Number of Spurious Operations that Must be Postulated
-introduction: The team identified an issue involving t number of concurrent spurious
operations associated with a particular component r set of components that must be
postulated during SSD analysis of a fire are,. Thi issue is a URI pending review-bythe
tflCeeft the issuance zf mew NRC ioepeetn idance ric
r A.
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 AP/O/A/5500/045, Plant 'Fire. Depending on the
fire location, Procedure AP/0/A/5500/045 directed the operator to close the PORV
isolation valves within ten minutes. The basis for this time critical action is the licensee's
assumption 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
8
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 (i.e., the SSS) would not be
independent from the MCR in that, during a fire in the MCR, pressurizer level may not
remain within the indicating range which 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 reater than minor. If more than one//
spurious operation were to occur, the dedi tedhutdown capability (i.e., the SS) , /
would not be independent from the MCR that fire in the MCR could result in
pressurizer level not remaining within tI range.
Enforcement: In the case of the PORV an ORV isolation valve circuits, operation of
the SSS may not be independent of the ie 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 isuanee of now NRC iespectien guidance Fgarding as
eifeuits, the issue is identified as URI 50-369/03-07-03, 370/03-07-04, 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.
Descriotion: 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. Valve 2CA0007A is important to safe shutdown for fire areas
where the SSS will be used because 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 normal AFW flow through the pump. Closure of this
valve could cause severe damage to the pump if automatic transfer to the alternate
suction sources does not initiate within sufficient time. For a severe fire in the MCR
requiring evacuation and transfer of plant shutdown to the SSS, the ability to remove
9
decay heat would be seriously degraded ifthe TDAFW pump were damaged. The team
found that the SSA did not include valve 2CAOOO7A. 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 (MCS-1465.00-O0-0022, Design
Basis Specification for Appendix R).
The licensee initiated PIPs-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/0I/N5500/045 was revised to specify that the operator ensure, within the first ten
minutes of an active fire, that valve 2CAOOO7A was open and then remove power from
2CAOOO7A.
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 610,
2CAI 620 and ORN4AC; Path 2 was through valves 2CA086A and 2RN069A; and Path
3 was through valves 2CA1 16B and 2RN1 62B. However, key information related to
these automatic transfers was not available to the team during the inspection.
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 MOR, the MOR
would be evacuated and the SSF would be used to achieve and maintain hot shutdown.
The finding was also determined to have potential safety significance greater than very
low significance because the SSF relies on the TDAFW pump for decay heat removal,
and the decay heat removal function would be seriously degraded if the TDAFW pump
were damaged due to closure of valve 2CAOOO7A.
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 15, 1989.
The McGuire FPP, which includes the SSA (MCS-1465.00-00-0022), 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.
10
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 50-370/03-07-05, 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
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 AP/1/A/5500/24 and AP/2/AN5500/024, Loss of Plant Control Due to Fire or
Sabotage, and AP/0/A/5500/045, 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. A specific finding related to this issue for Fire
Area 16/18 is 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/A/5500/024 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/AN5500/024 and
AP/2/A/5500/024 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 IlI.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:
11
- 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.
- SMP capacity to achieve and control solid plant operation from the SSF within
the required time to maintain subcooling.
- 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/1IA/5500/024 and AP/2/A/5500/024. 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 50-369/03-07-04,
370/03-07-06, Methods for Reactor Coolant System Pressure Control During SSF
Operation.
.06 Communications
a. Inspection Scone
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. Findinas
No findings of significance were identified.
12
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
evaluated in the licensee's SSA and properly integrated into the post-fire SSD
procedures. During team walk downs of the selected areas where local, manual
operator actions would be performed, area emergency lighting units were inspected for
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 Scone
The team reviewed the licensee's SSA and existing plant procedures to determine if any
repairs were necessary to achieve cold shutdown, and if needed, the equipment and
procedures required to implement those repairs were available onsite.
b. Findings
No findings of significance were identified.
.09 Fire Barriers and Fire Area/Zone/Room Penetration Seals
a. Inspection Scope
The team 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, penetration 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.
13
The tea v r ariers shown on the fire plan drawings for the selected fire
area The amroted at MNS has eliminated selected fire barriers from the
approved f e pr .ection program and designated these fire barriers as "Sealed Firewall -
Non Corn itted .hese arriers are no longer included in any surveillance and testing
program. he ore, d rs, dampers, fire proofing, etc. that exist in these declassified
barriers arno longer ncluded in any. station surveillance procedures and effectively
cannot be re n for the fire protection program. Two walls associated with Fire
Area 16/18 have been declassified. The wall between the Unit 2 switchgear room 2ETA
(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, and fire barrier partitions.
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 configurations conformed with the design drawings and
tested configurations. The team also compared the penetration seal 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 Equipment
a. Inspection ScoDe
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
14
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
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. Findings
No findings of significance were identified.
4. OTHER ACTIVITIES
40A2 Problem Identification and Resolution
a. Inspection Scope
The team reviewed a sample of licensee audits, self-assessments, and PlPs 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
15
classification, appropriateness, and timeliness of the corrective actions taken, or
initiated, to resolve the issues. Included in this review were PIPs G-99-00110, M-99-
01 884, 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 IA 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. This URI is
closed.
40A6 Meetings
The team presented the interim inspection results to Mr. D. Jamil and other members of
the licensee's staff on May 22, 2003. A final exit meeting was held via telephone with
Mr. G. Peterson, and other members of the licensee's staff on July 2, 2003, to present
the final results of the inspection. The licensee acknowledged the findings presented.
Proprietary information is not included in the inspection 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
M. Dicks, Engineer, Reactor and Electrical Systems (RES)
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, MCE
D. Jamil, Site Vice President, McGuire Nuclear Station
R. Johansen, Standby Shutdown Facility System Engineer
J. Lukowski, 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, McGuire Nuclear Station
C. Thomas, Regulatory Compliance Manager
K. Thomas, Manager, RES
NRC Personnel
J. Brady, Senior Resident Inspector,
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 Intern (Trainee)
S. Shaeffer, Senior Resident Inspector
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
ODened
50-369,370/03-07-01 URI 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 URI Inadequate Separation and Protection of Cables
Associated With Redundant Trains of
Instrumentation Located in the Same Fire Area
(Section 1R05.03.b.1)
Attachment
2
50-370/03-07-03 URI Use of a Local Manual Operator Action in Lieu of
Providing Physical Protection for Cables of
Redundant Safe Shutdown Equipment in Fire Area
16/18 (Section 1R05.03.b.2)
50-369/03-07-03, 370/03-07-04 URI Requirements Relative to the Number of Spurious
Operations That Must be Postulated (Section
1R05.04.b.1)
50-370/03-07-05 URI Spurious Closure of Valve 2CA0007A Could Lead
to Damage of the TDAFW Pump (Section
1R05.04.b.2)
50-369/03-07-04, 370/03-07-06 URI Methods for Reactor Coolant System Pressure
Control During SSF Operation (Section 1R05.05.b)
Closed
50-369,370/00-09-04 URI 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 1R05: Fire Protection
Procedures
AP/0/A/5500/045, Plant Fire, Rev. 0 and Rev. 2
AP/1/A15500/024, Loss of Plant Control Due to Fire or Sabotage, Rev. 21
AP/2/A/5500/024, Loss 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 Work Authorization, Rev. 2
NSD 316, Fire Protection Impairment and Surveillance, Rev. 6
MP/0/AN7650/122, 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
PT/0/A/4250/004, Fire Barrier Inspection, Rev. 19
PT/0/A/4250/01 1, Fire Door Inspections, Rev. 14
PT/0/A/4250/020, Roll-Up Fire Door Semi-Annual Inspection/Test, Rev. 2
PT/0/A/4400/OO1A, Fire Protection System Periodic Test, Rev. 24
PT/0/A/4400/001 C, Fire Protection System Monthly Test, Rev. 54
PT/0/A/4400/001 K, Fire Protection Annual Valve Test, Rev. 35
PT/0/A/4400/001 M, Fire Protection System Flow Test, Rev. 14
PT/0/A/4400/008, Fire Hose Hydrostatic Test SLC-Committed Hose Stations, Rev. 11
PT/0/AN4400/01 OA, Main Fire Pump A, Rev. 15
PT/0/A/4400/O1OB, Main Fire Pump B, Rev. 10
PT/OA/4400/01 0C, Main Fire Pump C, Rev. 11
PT/0/A/4400/017, Fire Pump A and B Operability Test, Rev. 13
PT/0/A/4400/018, Fire Pump C Operability Test, Rev. 11
PT/1 /A/4400/001 L, Fire Protection Containment Header Test, Rev. 9
PT/1 /A/4400/001 N, Halon 1301 System Periodic Test, Rev. 29
PT/2/A/4400/001 L, Fire Protection Containment Header Test, Rev. 7
PTM0,'A/460:011 6A, Fire Detection System Operational Tests, Rev. 18
PT/0/B/4600/015, Fire Detection System Monthly Test, Rev. 14
PT/0/A/4700/049, SLC Fire Hose Inspection, Rev. 1
PT/1/A/4700/042, SLC Fire Hose Station Valve Operability Test, Rev. 3
PT/2/A/4700/043, SLC Fire Hose Station Valve Operability Test, Rev. 3
PT/1/A/4150/001 B, Reactor Coolant Leakage Calculation, Rev. 47
Drawings
MC-1042-4, General Arrangement, Auxiliary 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
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
Attachment
2
MC-1 223-39, Auxiliary Building, Unit 1 & Unit 2, Beam Schedule at Elevation 733+0, Concrete
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-1 223-8, 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-1 223-27, 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-1 224-1 0, 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-1384-06.03, Fire Protection Layout, Plan at Elevation 733+0, Rev. 7
MC-1 384-06.04, Fire Protection Layout, Plan at Elevation 750+0, Rev. 7
MC-1384-06.05, Fire Protection Layout, Plan at Elevation 767+0, Rev. 7
MC-1384-07.12-00, Fire Plan, Auxiliary Building, Elevation 695+0, Rev. 3
MC-1 384-07.01-00, Fire Plan, Unit 1 Turbine Building, Elevation 739+0, Rev. 11
MC-1384-07.13-00, Fire Plan, Auxiliary Building, Elevation 716+0, Rev. 12
MC-1 384-07.13-01, Fire Plan, Auxiliary Building, Elevation 716+0, Rev. 9
MC-1384-07.14-00, Fire Plan, Auxiliary Building, Elevation 733+0, Rcv. 12
MC-1384-07.14-01, Fire Plan, Auxiliary Building, Elevation 733+0, Rev. 9
MC-1384-07.14-02, Fire Plan, Auxiliary Building, Elevation 733+0 & 736+6, Rev. 9
MO-1384-07.14-03, Fire Plan, Auxiliary Building, Elevation 733+0 & 736+6, Rev. 9
MC-1 384-07.15-00, Fire Plan, Auxiliary Building, Elevation 750+0, Rev. 10
MC-1 384-07.15-01, Fire Plan, Auxiliary Building, Elevation 750+0, Rev. 2
MC-1 384-07.15-01, Fire Plan, Auxiliary Building, Elevation 750+0, Rev. 3
MC-1384-07.15-01, Fire Plan, Auxiliary Building, Elevation 750+0, Rev. 9
MC-1 384-07.15-02, Fire Plan, Auxiliary Building, Elevation 750+0, Rev. 10
MC-1384-07.16-00, Fire Plan, Auxiliary Building, Elevation 760+6, Rev. 7
MC-1 384-07.17-00, 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+1 0, Rev. 8
MC-1518-06.43-00, Piping Layout, Interior Fire Protection, Nuclear Service Water Pumps,
Sprinkler Addition, Rev. 1
Attachment
...
3
MC-1518-06.43-01, Piping Layout, Interior Fire Protection, Component Cooling Pumps,
Sprinkler Addition, Rev. 1
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-1 710-04.1 0, 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-1 762-01.00-03, Location Diagram, Fire Detectors Located on Elevations 733+0 & 739+0,
Rev. 10
MC-1 762-01.00-04, 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-1 38-00.04, 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. 8
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
MCEE-250-00.29, Reactor Vessel Head Vent Valves, Rev. 5
MCEE-250-00.33, Reactor Vessel Head Vent Valves, Rev. 5
MCEE-257.00.54, Chemical and Volume Control Containment Isolation Valve, Rev. 3
MCEE-257-00.24, Chemical and Volume Control Containment Isolation Valve, Rev. 5
MCEE-257-00.50, Unit 2 Chemical and Volume Control Isolation Valve, Rev. 6
MCEE-257-00.52, Chemical and Volume Control Isolation Valve, Rev. 1
MCEE-257-00.55, Standby Makeup Pump, Rev. 1
Attachment
4
MCFD-1 574-01.00, Nuclear Service Water, Rev. 6
MCFD-1574-01.01, Nuclear Service Water, Rev. 10
MCFD-1599-01.00, P&ID, Flow Diagram of Fire Protection, Rev. 13
MCFD-1599-01.01, P&ID, Flow Diagram of Fire Protection, Rev. 14
MCFD-1 599-02.00, P&ID, Flow Diagram of Fire Protection, Rev. 15
MCFD-1599-02.01, P&ID, Flow Diagram of Fire Protection, Rev. 15
MCFD-1599-02.02, P&ID, Flow Diagram of Fire Protection, Rev. 5
MCFD-1599-02.03, P&ID, Flow Diagram of Fire Protection, Rev. 6
MCFD-1599-03.00, P&ID, Flow Diagram of Fire Protection, Rev. 7
~ MCFD-1599-03.01, P&ID, Flow Diagram of Fire Protection, Rev. 3
/J
-MCFD-2574-02.00, Nuclear Service Water, Rev. 12
MCFD-2574-02.01, Nuclear Service Water, Rev. 2
MCFD-2592-01.01, Auxiliary Feedwater System, Rev. 13
MCFD-2592-02.00, Auxiliary Feedwater System, Rev. 2
MCM.1206.07-0074.001, McNeary Insurance Consulting Services, FP-12
MCM.1206.07-0087.001, McNeary Insurance Consulting Services, FP-18
&) Completed 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, SIG 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
ORN4AC
Calculations and Evaluations
MCC-1 223.04-00-001 0, Determine the Reactor Coolant Pump Sealwater Flow Requirements
for the SSF Auxiliary Makeup Pump, Type II
MCC-1 223.42-00-0030, Documentation of the Adequacy of the Assured Suction Sources to the
CA Pumps, Rev. 8
Attachment
5
MCC-1223.49-00-0030, Sprinkler System for Nuclear Service Water Pumps @ Elevation 716-0,
Rev. 0
MCC-1 435.00-00-0006, 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 Protection Code Deviations, Rev. 0
MCS-1435.00-00-0001, Fire Protection Acceptance 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-1 465.00-00-0022, Design Basis Specification for Appendix R, Rev. 2
Problem Investigation Process Reports Reviewed
G-99-00110, 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 Ul Rx trip recovery.
M-99-01884, GL 86-10 guidance for circuit failure modes, hot short duration, and design basis
transients for dedicated shutdown not evaluated for applicability to MNS methodology.
M-99-01886, NFPA code deviations not documented in UFSAR or FHA as per GL 86-10.
M-99-03926, Effect of warmer seal injection water on RCP seals during SSF event not
adequately taken into consideration on SMP capacity. Evaluate applicability to McGuire.
M-00-01 900, 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 1RY 114 do not Permit
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.
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 1CA-0002, resulted in temp low suction flow to running 1B 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.
Attachment
6
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 AP/2/AN5500/24.
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 1ETA 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 1 & 2 (Revised), April 1981
SSER6, Appendix C, McGuire SER - Standby Shutdown System, February 1983
MNS Updated Final Safety Analysis Report (UFSAR) Section 9.5.1, 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
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-1 2907A thru F
Attachment
LIST OF ACRONYMS
AB Auxiliary Building
AP Abnormal Procedure
DSD Dedicated Shutdown
FHA Fire Hazards Analysis
FPP Fire Protection Review
GL Generic Letter
HVAC Heating Ventilation and Air Conditioning
IPEEE Individual Plant Examination for External Events
IR Inspection Report
kW Kilowatt
MCR Main Control Room
MNS McGuire Nuclear Station
NFPA National Fire Protection Association
- NRC Nuclear Regulatory Commission
NRR NRC Office of Nuclear Reactor Regulation
NSD Nuclear System Directive
NV Chemical and Volume Control
PIP Problem Investigation Process
PORV Power Operated Relief Valve
RCP Reactor Coolant Pump
RN Nuclear Service Water
SER Safety Evaluation Report
SG Stearm Generator
SLC Selected Licensee Commitment
SMP Standby Makeup Pump
SSA Safe Shutdown Analysis
SSD Safe Shutdown
SSF Standby Shutdown Facility
SSS Standby Shutdown System
TDAFW Turbine-Driven Auxiliary Feedwater
TS Technical Specifications
UFSAR Updated Final Safety Analysis Report
URI Unresolved Item
V Volt
Attachment