ML062140118
| ML062140118 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 07/31/2006 |
| From: | Julio Lara Engineering Branch 3 |
| To: | Crane C Exelon Generation Co, Exelon Nuclear |
| References | |
| IR-06-002 | |
| Download: ML062140118 (42) | |
See also: IR 05000254/2006002
Text
July 31, 2006
Mr. Christopher M. Crane
President and Chief Nuclear Officer
Exelon Nuclear
Exelon Generation Company, LLC
Quad Cities Nuclear Power Station
4300 Winfield Road
Warrenville, IL 60555
SUBJECT: QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2
NRC TRIENNIAL FIRE PROTECTION BASELINE INSPECTION
INSPECTION REPORT 05000254/2006002(DRS); 05000265/2006002(DRS)
Dear Mr. Crane:
On May 12, 2006, the U.S. Nuclear Regulatory Commission (NRC) completed a triennial fire
protection baseline inspection at your Quad Cities Nuclear Power Station, Units 1 and 2. The
enclosed inspection report documents the inspection results, which were discussed at the end
of the on-site activities on May 12, 2006, and during a re-exit meeting held by telephone at the
conclusion of the inspection on June 29, 2006, with Mr. T. Tulon and other members of your
staff.
The inspection examined activities conducted under your license, as they relate to safety and to
compliance with the Commissions rules and regulations, and with the conditions of your
license. The inspectors reviewed selected procedures and records, observed activities, and
interviewed personnel.
Based on the results of this inspection, six NRC-identified findings of very low safety
significance, all of which involved violations of NRC requirements were identified. However,
because these violations were of very low safety significance, and because the findings were
entered into the licensee's corrective action program, the NRC is treating these findings as
Non-Cited Violations in accordance with Section VI.A.1 of the NRCs Enforcement Policy.
Additionally, two licensee identified violations are listed in Section 4OA7 of this report.
If you contest the subject or severity of a Non-Cited Violation, you should provide a response
within 30 days of the date of this inspection report, with the basis for your denial, to the U. S.
Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC
20555-0001, with a copy to the Regional Administrator, U. S. Nuclear Regulatory Commission -
Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of
Enforcement, U. S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the
Resident Inspector Office at the Quad Cities Nuclear Power Station facility.
C. Crane -2-
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter,
its enclosure, and your response (if any), will be available electronically for public inspection in
the NRC Public Document Room or from the Publicly Available Records (PARS) component of
NRC's document system (ADAMS), accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Julio F. Lara, Chief
Engineering Branch 3
Division of Reactor Safety
Docket Nos. 50-254; 50-265
Enclosure: Inspection Report 05000254/2006002(DRS); 05000265/2006002(DRS)
w/Attachment: Supplemental Information
cc w/encl: Site Vice President - Quad Cities Nuclear Power Station
Plant Manager - Quad Cities Nuclear Power Station
Regulatory Assurance Manager - Quad Cities Nuclear Power Station
Chief Operating Officer
Senior Vice President - Nuclear Services
Senior Vice President - Mid-West Regional
Operating Group
Vice President - Mid-West Operations Support
Vice President - Licensing and Regulatory Affairs
Director Licensing - Mid-West Regional
Operating Group
Manager Licensing - Dresden and Quad Cities
Senior Counsel, Nuclear, Mid-West Regional
Operating Group
Document Control Desk - Licensing
Vice President - Law and Regulatory Affairs
Mid American Energy Company
Assistant Attorney General
Illinois Emergency Management Agency
State Liaison Officer, State of Illinois
State Liaison Officer, State of Iowa
Chairman, Illinois Commerce Commission
D. Tubbs, Manager of Nuclear
MidAmerican Energy Company
C. Crane -2-
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its
enclosure, and your response (if any) will be available electronically for public inspection in the
NRC Public Document Room or from the Publicly Available Records (PARS) component of
NRCs document system (ADAMS). ADAMS is accessible from the NRC Website at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
Julio F. Lara, Chief
Engineering Branch 3
Division of Reactor Safety
Docket Nos. 50-254; 50-265
Enclosure: Inspection Report 05000254/2006002(DRS); 05000265/2006002(DRS)
w/Attachment: Supplemental Information
cc w/encl: Site Vice President - Quad Cities Nuclear Power Station
Plant Manager - Quad Cities Nuclear Power Station
Regulatory Assurance Manager - Quad Cities Nuclear Power Station
Chief Operating Officer
Senior Vice President - Nuclear Services
Senior Vice President - Mid-West Regional
Operating Group
Vice President - Mid-West Operations Support
Vice President - Licensing and Regulatory Affairs
Director Licensing - Mid-West Regional
Operating Group
Manager Licensing - Dresden and Quad Cities
Senior Counsel, Nuclear, Mid-West Regional
Operating Group
Document Control Desk - Licensing
Vice President - Law and Regulatory Affairs
Mid American Energy Company
Assistant Attorney General
Illinois Emergency Management Agency
State Liaison Officer, State of Illinois
State Liaison Officer, State of Iowa
Chairman, Illinois Commerce Commission
D. Tubbs, Manager of Nuclear
MidAmerican Energy Company
DOCUMENT NAME:E:\Filenet\ML062140118.wpd
G Publicly Available G Non-Publicly Available G Sensitive G Non-Sensitive
To receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl "N" = No copy
OFFICE RIII RIII RIII RIII
NAME GHausman: ls MRing JLara
DATE 07/28/06 07/27/06 07/31/06
OFFICIAL RECORD COPY
C. Crane -3-
ADAMS Distribution:
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ROPreports@nrc.gov
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket Nos: 50-254; 50-265
Report No: 05000254/2006002(DRS); 05000265/2006002(DRS)
Licensee: Exelon Generation Company, LLC
Facility: Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2
Location: Cordova, IL
Dates: April 24, 2006 through May 12, 2006
Inspectors: G. Hausman, Senior Reactor Inspector, Lead
A. Dahbur, Reactor Inspector
D. Schrum, Reactor Inspector
Approved by: Julio F. Lara, Chief
Engineering Branch 3
Division of Reactor Safety
Enclosure
TABLE OF CONTENTS
SUMMARY OF FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
REPORT DETAILS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
1. REACTOR SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
1R05 Fire Protection (71111.05T) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
.1 Systems Required to Achieve and Maintain Post-Fire SSD . . . . . . . 4
a. Inspection Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
b. Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
b.1 Safe Shutdown Makeup Pump . . . . . . . . . . . . . . . . . . . . 5
b.2 RHR Service Water Cross-tie . . . . . . . . . . . . . . . . . . . . . 8
.2 Fire Protection of SSD Capability . . . . . . . . . . . . . . . . . . . . . . . . . . 11
.3 Post-Fire SSD Circuit Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
.4 Alternative SSD Capability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
.5 Operational Implementation of SSD Capability . . . . . . . . . . . . . . . 13
a. Inspection Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
b. Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
b.1 Fuse Repair . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
.6 Communications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
.7 Emergency Lighting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
.8 Cold Shutdown Repairs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
.9 Fire Barriers and Fire Zone/Room Penetration Seals . . . . . . . . . . . 16
.10 Fire Protection Systems, Features, and Equipment . . . . . . . . . . . . 18
a. Inspection Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18
b. Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18
b.1 Fire Pre-Plans . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18
b.2 Standpipes with Hose Connections . . . . . . . . . . . . . . . 20
b.3 Class A Fire Extinguishers . . . . . . . . . . . . . . . . . . . . . . 22
.11 Compensatory Measures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26
4. OTHER ACTIVITIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26
4OA5 Other Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26
4OA6 Meetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27
4OA7 Licensee-Identified Violations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27
.1 Appendix R SSD Battery Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . 27
.2 Post-Fire Operator Manual Actions . . . . . . . . . . . . . . . . . . . . . . . . 28
SUPPLEMENTAL INFORMATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
KEY POINTS OF CONTACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . A-2
LIST OF DOCUMENTS REVIEWED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-3
LIST OF ACRONYMS USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-7
i Enclosure
SUMMARY OF FINDINGS
IR 05000254/2006002(DRS), 05000265/2006002(DRS); 04/24/06 - 05/12/06; Quad Cities
Nuclear Power Station (QCNPS), Units 1 and 2; Triennial Fire Protection Baseline Inspection.
This report covers an announced triennial fire protection baseline inspection. The inspection
was conducted by Region III inspectors. Based on the results of this inspection, six Green
findings associated with six non-cited violations 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 NRCs 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-Revealed Findings
Cornerstone: Mitigating Systems
- Green. The inspectors identified a Non-Cited Violation (NCV) of 10 CFR Part 50,
Appendix R,Section III.G.2, having very low safety significance (Green) involving the
licensees failure to ensure, in the event of a severe fire, that one redundant train of
systems necessary to achieve and maintain hot shutdown conditions was free of fire
damage. Specifically, the licensee failed to ensure, in the event of a fire in any of the
III.G.2 fire areas, that one redundant train of reactor coolant inventory makeup water
remained free of fire damage. Instead the licensee credited the dedicated safe
shutdown makeup pump (SSMP) for reactor coolant inventory makeup water in the
III.G.2 fire areas. This finding was entered into the licensees corrective action program
as Issue Report (IR) 00502702, NRC Inspection Finding Concerning App R Redundant
Trains, dated June 22, 2006. The licensee plan to review the options for resolving this
issue, and pursue the appropriate resolution.
The finding was more than minor because this failure could have affected the mitigating
systems cornerstone objective and safe shutdown (SSD). Specifically, the licensee
failed to ensure one redundant train of reactor coolant inventory makeup water was
available, and instead relied on an alternate shutdown system without an analyses and
procedures that demonstrated full compliance with all of the requirements of 10 CFR
Part 50, Appendix R,Section III.G.3, and Section III.L, or requesting prior NRC
approval. The finding was not suitable for SDP evaluation, but has been reviewed by
NRC management and was determined to be a finding of very low safety significance.
(Section 1R05.1b.1)
- Green. The inspectors identified a NCV of 10 CFR Part 50, Appendix R, Section III.G.2,
having very low safety significance (Green) involving the licensee failure to ensure, in
the event of a severe fire, that one redundant train of systems necessary to achieve and
maintain hot shutdown conditions was free of fire damage. Specifically, the licensee
failed to ensure, in the event of a fire in Fire Areas TB-III, 13-1 or 24-1, that one
redundant train of residual heat removal service water (RHRSW) remained free of fire
damage. Instead the opposite units RHRSW train was cross-tied (i.e., an alternative
SSD activity) and credited for torus cooling during hot shutdown for a III.G.2 fire area.
1 Enclosure
In addition, the licensee failed to have an analyses and procedures that demonstrated
full compliance with all of the requirements of 10 CFR Part 50, Appendix R,
Section III.G.3, and Section III.L. This finding was entered into the licensees corrective
action program as IR 00502702, NRC Inspection Finding Concerning Appendix R
Redundant Trains, dated June 22, 2006. The licensee plan to review the options for
resolving this issue, and pursue the appropriate resolution.
The finding was more than minor because the failure to ensure one redundant train of
RHRSW was available for torus cooling for hot shutdown could have affected the
mitigating systems cornerstone objective and SSD. The finding was not suitable for
SDP evaluation, but has been reviewed by NRC management and was determined to be
a finding of very low safety significance. (Section 1R05.1b.2)
- Green. The inspectors identified a NCV of 10 CFR Part 50, Appendix B, Criterion V,
Instruction, Procedures, and Drawings, having very low safety significance (Green)
involving inadequate procedure steps. Specifically, The licensee failed to provide
adequate procedure steps in-accordance with Appendix R requirements for hot
shutdown and allowed the replacement (i.e., a repair) of breaker fuses prior to attaining
hot shutdown. Specifically, QCNPSs Procedure QOP 6500-10 Local Control of 4160
and 480 Volt Motor Operated Circuit Breaker, Revision 8, included a hot shutdown
repair to replace any circuit breakers control fuses that were believed to be blown due
to a fire-induced failure. This fuse replacement constituted a hot shutdown repair which
was not allowed by 10 CFR Part 50, Appendix R. Once identified, the licensee revised
procedure QOP 6500-10 and added steps to manually close breakers using a local
pushbutton. This finding was entered into the licensees corrective action program as
IR 00485702, Required SSDA Actions Not Contained in QCARP 0030-01, dated
May 2, 2006. The licensee revised procedure QOP 6500-10.
The finding was more than minor because the failure to include adequate procedure
steps could have affected the mitigating systems cornerstone objective and SSD.
Performing the repair activities could have delayed and/or complicated shutdown of the
plant. The finding was of very low safety significance because the licensee could have
manually charged the breakers spring and closed the breaker using the pushbutton
located at the breaker. (Section 1R05.5b.1)
having very low safety significance (Green) involving the lack of complete and accurate
information in the QCNPSs fire pre-plans for various plant fire areas. Specifically, the
licensee failed to include important information in the fire pre-plans, such as hydrogen
and electrical hazards, to assist the fire brigade to fight a fire within those plant fire
areas. This finding was entered into the licensees corrective action program as
IR 00489175, Quality of Fire Preplans (2006 FP Triennial), dated May 11, 2006.
The finding was more than minor because the failure to provide adequate warnings and
guidance related to hydrogen and electrical hazards in the fire pre-plans could have
adversely impacted the fire brigades ability to fight a fire, thereby, increasing the
likelihood of a fire which would challenge SSD and could have affected the mitigating
systems cornerstone objective. The inspectors determined that this issue also affected
the cross-cutting area of Problem Identification and Resolution because the licensee
2 Enclosure
failed to identify the presence of hydrogen and oxygen hazards in Fire Areas RB-7 and
RB-19 during their review as part of the fire pre-plan improvement effort conducted as a
result of previously identified corrective action (IR 00221528). The finding was of very
low safety significance because of the extensive training provided to the fire brigade
members to deal with unexpected contingencies. (Section 1R05.10b.1)
having very low safety significance (Green) involving adequacy of water pressure and
flow rate at standpipes with hose connections. Specifically, the licensee failed to
provide calculations to ensure that an adequate water pressure and flow rate were
available to meet the QCNPSs FPP requirements. Once identified, the licensee
entered the finding into their corrective action program as IR 00489160, Justification of
Fire Hose Pressure and Flow Meeting NFPA, dated May 11, 2006, and planned to
perform calculations to verify water flow at all affected standpipes with hose
connections.
The finding was more than minor because the failure to provide an adequate water
pressure and flow rate at standpipes with hose connections could hamper the fire
brigades ability to fight a fire, thereby, increasing the likelihood of a fire which would
challenge SSD and could have affected the mitigating systems cornerstone objective.
The finding was of very low safety significance because other defense-in-depth fire
protection elements remained unaffected in all fire areas. (Section 1R05.10b.2)
having very low safety significance (Green) involving adequacy of number of Class A
fire extinguishers. Specifically, the licensee failed to have an adequate number of
Class A fire extinguishers available where significant fire hazards existed to meet the
NFPA 10 Code requirements to suppress and/or extinguish Class A fire hazards. This
finding was entered into the licensees corrective action program as IR 00489426,
Class A Fire Extinguisher Placement Improvements, dated May 12, 2006. The
licensee planned to evaluate putting more Class A fire extinguishers into the plant.
The finding was more than minor because failure to have an adequate number of
Class A fire extinguishers available could potentially escalate a small fire into a larger
fire since only standpipes with hose connections were available and their use required a
trained fire brigade to extinguish the fire. As a result, non-fire brigade personnel would
be prevented from moving quickly to suppress and/or extinguish a small fire and the
potential for an escalated fire could have affected the mitigating systems cornerstone
objective. The finding was of very low safety significance because most fire areas and
zones have fire detectors that would alarm in the control room and the fire brigade
would respond to a fire in these areas. In addition, other defense-in-depth fire
protection elements remained unaffected and a fire in these areas would not result in a
loss of dedicated SSD systems. (Section 1R05.10b.3)
B. Licensee-Identified Violations
Two violations of very low safety significance, which were identified by the licensee,
have been reviewed by the inspectors. Corrective actions taken or planned by the
licensee have been entered into the licensees corrective action program. These
violations and the licensees corrective action tracking numbers are listed in
Section 4OA7 of this report.
3 Enclosure
REPORT DETAILS
1. REACTOR SAFETY
Cornerstones: Initiating Events and Mitigating Systems
1R05 Fire Protection (71111.05T)
The purpose of this inspection was to review the QCNPSs Fire Protection Program
(FPP) for selected risk-significant fire areas. Emphasis was placed on determining that
the post-fire safe shutdown (SSD) capability and the fire protection (FP) features were
maintained free of fire damage to ensure that at least one post-fire SSD success path
was available. The inspection was performed in accordance with the Nuclear
Regulatory Commissions (NRCs) regulatory oversight process using a risk-informed
approach for selecting the areas and attributes to be inspected. The inspectors with
assistance from a senior reactor analyst used the QCNPSs Individual Plant
Examination for External Events (IPEEE) to choose several risk-significant areas for
detailed inspection and review. The inspectors review of the following fire areas and/or
zones represented completion of four samples for the triennial FP inspection:
Fire Area Fire Zone Description
TB-III 8.2.6.A Unit 1 Turbine Building Ground Floor (South)
TB-III 8.2.7.A Unit 1 Turbine Building Mezzanine Floor (South)
13-1 8.2.8.B Unit 1 4-kV Bus 13-1 Switchgear Area (North)
24-1 8.2.8.C Unit 2 4-kV Bus 24-1 Switchgear Area (South)
For each of the selected fire areas and/or fire zones, the inspectors focused on the fire
protection features, the systems and equipment necessary to achieve and maintain safe
shutdown conditions, determination of licensee commitments, changes to the FPP, and
evaluated the licensees FPP against applicable NRC requirements.
.1 Systems Required to Achieve and Maintain Post-Fire SSD
Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix R,Section III.G.1,
required the licensee to provide FP features that were capable of limiting fire damage to
structures, systems, and components (SSCs) important to SSD. The SSCs that were
necessary to achieve and maintain post-fire SSD were required to be protected by FP
features that were capable of limiting fire damage to the SSCs so that:
- one train of systems necessary to achieve and maintain hot shutdown conditions
from either the CR or emergency control station(s) was free of fire damage; and
- systems necessary to achieve and maintain cold shutdown from either the CR or
emergency control station(s) could be repaired within 72-hours.
4 Enclosure
Specific design features for ensuring this capability were specified by 10 CFR Part 50,
Appendix R,Section III.G.2.
a. Inspection Scope
The inspectors reviewed the plant systems required to achieve and maintain post-fire
SSD to determine if the licensee had properly identified the components and systems
necessary to achieve and maintain SSD conditions for each fire zone selected for
review. Specifically, the review was performed to determine the adequacy of the
systems selected for reactivity control, reactor coolant inventory makeup, reactor heat
removal, process monitoring, and support system functions. This review included the
FP Safe Shutdown Analysis (SSA).
The inspectors also reviewed the operators ability to perform the necessary manual
actions for achieving SSD by reviewing procedures, the accessibility of SSD equipment,
and the available time for performing the actions.
The inspectors reviewed the QCNPSs Updated Final Safety Analysis Report (UFSAR)
and the licensees engineering and/or licensing justifications (e.g., NRC guidance
documents, license amendments, technical specifications, safety evaluation
reports (SERs), exemptions, and deviations) to determine the licensing basis.
b. Findings
b.1 Safe Shutdown Makeup Pump
Introduction: The inspectors identified a Non-Cited Violation (NCV) of 10 CFR Part 50,
Appendix R,Section III.G.2, having very low safety significance (Green) involving the
licensees failure to ensure, in the event of a severe fire, that one redundant train of
systems necessary to achieve and maintain hot shutdown conditions was free of fire
damage. Specifically, the licensee failed to ensure, in the event of a fire in any of the
III.G.2 fire areas, that one redundant train of reactor coolant inventory makeup water
remained free of fire damage. Instead, the licensee credited the dedicated safe
shutdown makeup pump (SSMP) for reactor coolant inventory makeup water in these
fire areas without having an analysis that demonstrated full compliance with all of the
requirement of 10 CFR Part 50, Appendix R, Section III.G.3, III.L or requesting prior
NRC approval.
Description: During the inspectors review of the licensees Appendix R analysis for the
selected fire zones, the inspectors identified that contrary to the NRC approved licensing
bases, as discussed in the SER dated December 1981, the licensee was relying upon
the SSMP as a redundant train for the reactor coolant inventory makeup in the event of
a severe fire in III.G.2 fire areas. Specifically, the SER dated December 1981, for the
QCNPS, Section 2.2, Areas Where Alternate Safe Shutdown Capability is Required,
stated . . . for those areas where a fire results in loss of the RCIC system, the licensee
has proposed to provide a new safe shutdown makeup pump as an alternate.
Section 2.3, (Section III.G.2 of Appendix R), stated . . . the licensee also indicated that
all other areas of the plant not required to have an alternate safe shutdown system will
comply with the requirements of Section III.G.2 of Appendix R, unless an exemption
5 Enclosure
request has been approved by the staff. Section 2.4, Alternate Safe Shutdown
System, stated . . . the alternate safe shutdown system required for those areas not
meeting Section III.G.2 or included in the exemption requests consists of a new safe
shutdown makeup pump and its associated support systems and instrumentation.
Section 3.1.2, Reactor Coolant Inventory, stated The RCIC and shutdown makeup
pump initially take suction from the CCST. Additional supply for the RCIC pump is
provided by the suppression pool and for the makeup pump by the service water
system. However, the NRC in April of 1988 revised Section 3.1.2 of this SER and
stated that Backup water supply source for the safe shutdown makeup pump will be
provided by the fire water system instead of the service water system as originally
indicated in the earlier SER Section 3.1.2. The inspectors could not find any other
documents from the NRC to the licensee indicating that the NRC had accepted the use
of the SSMP as a redundant system for Appendix R.
Prior to November 2000, the fire areas which were selected during this triennial FP
inspection, TB-III, 13-1 and 24-1, were classified as alternate shutdown areas because
of the potential fire damage to redundant RCIC system and the use of the alternate
SSMP system. As a result of the FP Optimization Project at QCNPS, in November of
2000, the licensee completed safety evaluation SE-00-098. Part of the objectives of the
optimization project were to reduce the number of the areas that require leaving the
main control room and reduce reliance on 10 CFR Part 50, Appendix R exemptions. In
order to accomplish these objectives, the licensee reclassified several fire areas from
Alternate to Redundant by reclassifying the SSMP from Dedicated to Redundant
per safety evaluation SE-00-098. The reclassification was based primarily on the
functional equivalence between RCIC and SSMP. The licensee concluded that this
change had no adverse impact on SSD and, therefore, did not require prior NRC
approval. The licensee also modified the control for the SSMP system, where injection
could be manually initiated from either the control room or remotely. This was changed
from the initial installation of the pump. After the change, all of the circuitry needed to
operate the SSMP was included in the SSA.
In Generic Letter (GL) 86-10 Implementation of Fire Protection Requirements, the
NRCs response to Question 3.8.3 Redundant Trains/Alternates Shutdown, stated, in
part, if the system is being used to provide its design function, it generally is considered
redundant. If the system is being used in lieu of the preferred system because the
redundant components of the preferred system do not meet the separation criteria of
Section III.G.2, the system is considered an alternate shutdown capability.
The inspectors review of SE-00-098 concluded that the licensees basis for the
reclassification was not appropriate, in that, based on the response to Question 3.8.3 in
GL 86-10, and that the backup water supply source for the SSMP was from the fire
water system, which was not a preferred source for reactor coolant inventory makeup
water, the SSMP met the criteria for an alternate and not a redundant system to RCIC.
The licensee response was that based on General Electric (GE) Nuclear Energy (NE)
document GE-NE-T43-00002-00-03-R01, BWROG [Boiling Water Reactor Owners
Group] Position on the Use of Safety Relief Valves and Low Pressure Systems as
Redundant Safe Shutdown Paths, a system was considered to be redundant when it
is used to provide its design function. The GE-NE document also indicated that based
on the information provided in NRC GL 81-12, the BWROG sees no requirement or
6 Enclosure
basis for limiting the systems that may be used as redundant SSD systems. The
inspectors reviewed the GE documents and could not find any justification in these
documents that considered the use of the SSMP, with the FP system as a backup
source, a redundant system for reactor coolant inventory makeup.
In addition, Subsection 3 of Section III.L Alternative and Dedicated Shutdown
Capability to 10 CFR Part 50 Appendix R required, in part, that the shutdown capability
shall be independent of the specific fire area(s) and shall accommodate postfire
conditions where offsite power is available and where offsite power is not available for
72-hours. Procedures shall be in effect to implement this capability. During the
inspection, the licensee did not have analyses or procedures that satisfied these
requirements. Instead, the licensee depended on offsite power supplied from the
opposite unit equipment. Therefore, the inspectors also concluded that the change
made by SE-00-098 was not appropriate and had adversely affected safe shutdown,
because the licensee did not have the required analyses and procedures. The licensee
entered this finding into the QCNPSs corrective action program as IR 00502702, NRC
Inspection Finding Concerning App R Redundant Trains, dated June 22, 2006. The
licensee plan to review the options for resolving this issue and pursue the appropriate
resolution.
Analysis: The inspectors determined that failure to ensure that one redundant train of
systems necessary to achieve and maintain hot shutdown conditions was free of fire
damage resulted in a performance deficiency warranting a significance evaluation. The
inspectors concluded that the finding was greater than minor in accordance with
IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, issued
on September 30, 2005. The finding involved the attribute of protection against external
factors (i.e., Fire) and could have affected the mitigating systems cornerstone objective
of ensuring the availability, reliability, and capability of systems that respond to initiating
events to prevent undesirable consequences (i.e., core damage). Specifically, the
licensees failure to ensure, in the event of a fire in any of the III.G.2 fire areas, that one
redundant train of reactor coolant inventory makeup water remained free of fire damage
and instead credited the dedicated SSMP without demonstrating compliance with
Appendix R, Sections III.G.3 and III.L, did not provide the adequate level of safety
required per Appendix R to ensure SSD capability.
Since, the SSD path using the SSMP system did not meet the requirement of
Appendix R,Section III.G.2, but the SSD path was approved by the NRC as alternate,
the inspectors determined that the finding was not suitable for SDP evaluation. The
finding was reviewed by NRC management and was determined to be of very low safety
significance (Green).
Enforcement: 10 CFR Part 50.48, Fire Protection, and 10 CFR Part 50, Appendix R,
Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1,
1979, established specific FP features required to satisfy 10 CFR Part 50, Appendix A,
General Design Criterion 3, Fire Protection. Appendix R applies to licensed nuclear
power electric generating stations that were operating prior to January 1, 1979, which
included QCNPS.Section III.G.2 of Appendix R to 10 CFR Part 50 required, in part,
that where cables or the equipment of a redundant train of systems necessary to
achieve and maintain hot shutdown conditions are located within the same fire area
7 Enclosure
outside of primary containment, one of a specified means of ensuring that one of the
redundant trains is free of fire damage shall be provided.Section III.G.3 of Appendix R
to 10 CFR Part 50 required, in part, that alternative of dedicated shutdown capability
should be provided where the protection of systems whose function is required for hot
shutdown does not satisfy the requirement of paragraph III.G.2. Subsection 3 of
Section III.L. Alternative and Dedicated Shutdown Capability to 10 CFR Part 50,
Appendix R required, in part, that the shutdown capability shall be independent of the
specific fire area(s) and shall accommodate postfire conditions where offsite power is
available and where offsite power is not available for 72-hours. Procedures shall be in
effect to implement this capability.
Contrary to the above, in the event of a fire in any of these fire areas (TB-III, 13-1 or
24-1), the licensee failed to ensure that one of the redundant trains of reactor coolant
inventory makeup water remained free of fire damage. Instead, the licensee credited
the use of the dedicated SSMP for reactor coolant inventory makeup without having
analyses and procedures that demonstrated their full compliance with III.G.3 and III.L, or
requesting prior NRC approval. Once identified, the licensee entered the finding into
their corrective action program as IR 00502702, NRC Inspection Finding Concerning
App R Redundant Trains, dated June 22, 2006. Because this violation was of very low
safety significance and it was entered into the licensees corrective action program, this
violation is being treated as a NCV, consistent with Section VI.A.1 of the NRC
Enforcement Policy (NCV 05000254/2006002-01(DRS);05000265/2006002-01(DRS)).
b.2 RHR Service Water Cross-tie
Introduction: The inspectors identified a NCV of 10 CFR Part 50, Appendix R,
Section III.G.2, having very low safety significance (Green) involving the licensees
failure to ensure, in the event of a severe fire, that one redundant train of systems
necessary to achieve and maintain hot shutdown conditions was free of fire damage.
Specifically, the licensee failed to ensure, in the event of a fire in any of these Fire
Areas TB-III, 13-1 or 24-1 (III.G.2 Fire Area), that one redundant train of the residual
heat removal service water (RHRSW) system remained free of fire damage. Instead the
licensee credited the cross-tie of the RHRSW train from the opposite unit for torus
cooling during hot shutdown.
Description: The RHRSW system at QCNPS for each unit, consisted of two
independent and redundant subsystems (A Loop and B Loop). Each subsystem is
made of a header, two pumps, a suction source, valves, piping heat exchanger and
associated instrumentation. Each loop provided cooling water flow to the respective
loop of the RHR heat exchangers in the torus cooling and shutdown modes. The A
subsystems for Unit 1 and Unit 2 could be connected by a normally locked, isolated
cross-tie line. Similarly, the B subsystems could be connected by another normally
isolated cross-tie line. These lines could be used to supply RHRSW from Unit 1 to Unit
2 and vice-versa when required.
The inspectors reviewed QCNPSs SSA which indicated that, in the event of a severe
fire in Fire Area 13-1, a redundant shutdown method and/or systems would be utilized
for SSD of Unit 1. The analysis also indicated that, in the event of a fire in Fire
Area 13-1, Unit 1 4KV Switchgear 13-1 (Division I) and/or both divisions of 480 V
8 Enclosure
Switchgear (18 and 19) would be exposed. Division II of the Unit 1 RHR system would
be available for torus cooling and alternate shutdown cooling. However, the Unit 1
Division II RHRSW system was unavailable for a fire in Fire Area 13-1. This was due to
the 480V power feeds for the pump cubicle coolers being routed in this area. Therefore,
the analysis stated that Unit 1 RHRSW pumps may not be available for decay heat
removal. However, Unit 1 Division II RHRSW can be supplied by Unit 2 RHRSW Pump
via a Division II mechanical cross-tie. The analysis also verified that cubical coolers for
Unit 2, Division II RHRSW pumps were free of fire damage. The inspectors also
identified a similar issue in the event of a fire in Unit 2 Fire Area 24-1, in that the
shutdown method for this area used the Unit 1 RHRSW to shutdown Unit 2.
In addition, the inspectors noticed that the analysis also showed that, in the event of a
severe fire in Fire Area TB-III, Division I of the RHR system would be available for torus
cooling and alternate shutdown cooling. However, the AC power for RHR system would
have been provided from offsite power via the opposite unit (Unit 2 Bus 23) and cross-
tied to Unit 1 Bus 13-1. Similarly, the analysis also indicated that the Unit 1 Division I
RHRSW system may not be available for a fire in TB-III. This was due to the 4KV
power feeds for the pumps being routed in TB-III. Therefore, the analysis indicated that
Unit 1 RHRSW pumps may not be available for decay heat removal. However, the
analysis credited Unit 2 Division I RHRSW via a Division I mechanical crosstie valve.
Prior to November of 2000, Fire Areas TB-III, 13-1 and 24-1 were classified as an
alternate shutdown areas. As a result of the FP Optimization Project at QCNPS, the
licensee reclassified these fire areas from Alternate to Redundant when they
reclassifying the SSMP from Dedicated to Redundant per Safety Evaluation
SE-00-098. However, the licensee failed to evaluate whether RHRSW cross-tie from
the opposite unit met the requirement of 10 CFR Part 50, Appendix R, Section III.G.2.
10 CFR Part 50.48 Fire Protection required, in part, that each operating nuclear power
plant must have a fire protection plan that satisfies Criterion 3 of appendix A of 10 CFR Part 50. 10 CFR Part 50, Appendix R establishes fire protection features required to
satisfy Criterion 3 of appendix A with respect to certain generic issues for nuclear power
plants licensed to operate before January 1, 1979. 10 CFR Part 50, Appendix R, also
required, in part, that a fire protection program be established at each nuclear power
plant.Section III.G.2 of Appendix R stated, in part, that where cables or equipment 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 three means of ensuring that one of the redundant trains is free of
fire damage shall be provided. The inspectors determined that the licensees shutdown
method in the event of a fire in any of these Fire Areas TB-III, 13-1 or 24-1, did not meet
the requirement of 10 CFR Part 50, Appendix R, Section III.G.2. Specifically, the
licensees shutdown methods in Fire Areas TB-III, 13-1 and 24-1 which used the
RHRSW system from the opposite unit instead of ensuring one of the redundant trains
of RHRSW from the respective unit with fire Unit 1 RHRSW was available. In addition,
Subsection 3 of Section III.L Alternative and Dedicated Shutdown Capability to 10 CFR Part 50, Appendix R required, in part, that the shutdown capability shall be independent
of the specific fire area(s) and shall accommodate postfire conditions where offsite
power is available and where offsite power is not available for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Procedures
9 Enclosure
shall be in effect to implement this capability. During the inspection, the licensee did
not have analyses or procedures that satisfied these requirements for Appendix R III.L.
The licensee justification was that Fire Areas TB-III,13-1 and 24-1 were alternate
shutdown areas based on the reliance on SSMP and not due to the RHRSW cross-tie.
The licensee also indicated that this method for SSD, which credited the RHRSW
cross-tie, was previously communicated to the NRC staff by letter dated December 18,
1984. The inspectors reviewed the licensees submittal and noticed that the RHRSW
cross-tie was only credited in alternate shutdown fire areas.. Therefore, the inspectors
concluded that the use of RHRSW cross-tie was only permitted for alternate shutdown
areas (i.e.,Section III.G.3), and would not meet the requirements of 10 CFR Part 50,
Appendix R,Section III.G.2. See Section 1R05.1b.1 of this inspection report for further
discussion of SSMP reclassification.
Analysis: The inspectors determined that failure to ensure that one redundant train of
systems necessary to achieve and maintain hot shutdown conditions was free of fire
damage resulted in a performance deficiency warranting a significance evaluation. The
inspectors concluded that the finding was greater than minor in accordance with
IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, issued
on September 30, 2005. The finding involved the attribute of protection against external
factors (i.e., Fire) and could have affected the mitigating systems cornerstone objective
of ensuring the availability, reliability, and capability of systems that respond to initiating
events to prevent undesirable consequences (i.e., core damage). Specifically, the
licensees failure to ensure, in the event of a fire in any of these Fire Areas TB-III, 13-1
or 24-1, that one redundant train of RHRSW remained free of fire damage and instead
credited the RHRSW from the opposite unit, did not provide the adequate level of safety
required per Appendix R to ensure SSD capability.
Since, the SSD path using the RHRSW system from the opposite unit did not meet the
requirement of Appendix R,Section III.G.2, but it was credited in alternate shutdown
areas at Quad Cities plant, the inspectors determined that the finding was not suitable
for SDP evaluation. The finding was reviewed by NRC management and was
determined to be of very low safety significance (Green).
Enforcement: 10 CFR Part 50.48, Fire Protection, and 10 CFR Part 50, Appendix R,
Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1,
1979, establish specific FP features required to satisfy 10 CFR Part 50, Appendix A,
General Design Criterion 3, Fire Protection. Appendix R applies to licensed nuclear
power electric generating stations that were operating prior to January 1, 1979, which
includes QCNPS.Section III.G.2 of Appendix R to 10 CFR Part 50 required, in part, that
where cables or equipment 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 a specified means of ensuring that one of the redundant
trains is free of fire damage be provided.
Contrary to the above, in the event of a fire in any of these Fire Areas TB-III, 13-1 or 24-
1 (i.e., a III.G.2 area), the licensee failed to meet the requirement of Section III.G.2 of
10 CFR Part 50, Appendix R. Specifically, in the event of a fire in any of these areas,
the licensee credited the opposite Unit RHRSW system instead of ensuring one of the
10 Enclosure
redundant trains of RHRSW from the respective unit with fire was available. Once
identified, the licensee entered the finding into their corrective action program as
IR 00502702, NRC Inspection Finding Concerning App R Redundant Trains, dated
June 22, 2006. Because this violation was of very low safety significance and it was
entered into the licensees corrective action program, this violation is being treated as a
NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy
(NCV 05000254/2006002-02(DRS);05000265/2006002-02(DRS)).
.2 Fire Protection of SSD Capability
Title 10 CFR Part 50, Appendix R, Section III.G.2, required separation of cables and
equipment and associated circuits of redundant trains by a fire barrier having a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />
rating. If the requirements cannot be met, then alternative or dedicated shutdown
capability and its associated circuits, independent of cables, systems or components in
the area, room, or zone under consideration should be provided in accordance with
10 CFR Part 50, Appendix R, Section III.G.3.
a. Inspection Scope
For each of the selected fire areas, the inspectors reviewed the licensees Safe
Shutdown Capability Analysis (SSCA) and Safe Shutdown Systems Analysis (SSSA) to
ensure that at least one post-fire SSD success path was available in the event of a fire.
This included a review of manual actions required to achieve and maintain hot shutdown
conditions and make the necessary repairs to reach cold shutdown within 72-hours.
The inspectors also reviewed procedures to verify that adequate direction was provided
to operators to perform these manual actions. Factors, such as, timing, access to the
equipment, and the availability of procedures were considered in the review.
The inspectors also evaluated the adequacy of fire suppression and detection systems,
fire area barriers, penetration seals, and fire doors to ensure that at least one train of
SSD equipment was free of fire damage. To accomplish this, the inspectors observed
the material condition and configuration of the installed fire detection and suppression
systems, fire barriers, and construction details and supporting fire tests for the installed
fire barriers. In addition, the inspectors reviewed license documentation, such as,
deviations, detector placement drawings, fire hose station drawings, carbon dioxide
pre-operational test reports, smoke removal plans, fire hazard analysis reports, SSD
analyses, and National Fire Protection Association (NFPA) codes to verify that the fire
barrier installations met license commitments.
b. Findings
No findings of significance were identified.
.3 Post-Fire SSD Circuit Analysis
Title 10 CFR Part 50, Appendix R, Section III.G.1, required that SSCs important to SSD
be provided with FP features capable of limiting fire damage to ensure that one train of
systems necessary to achieve and maintain hot shutdown conditions is free of fire
damage. Options for providing this level of FP were delineated in 10 CFR Part 50,
11 Enclosure
Appendix R,Section III.G.2. Where the protection of systems whose function was
required for hot shutdown did not satisfy 10 CFR Part 50, Appendix R, Section III.G.2,
an alternative or dedicated shutdown capability and its associated circuits, was required
to be provided that was independent of the cables, systems, and components in the
area. For such areas, 10 CFR Part 50, Appendix R, Section III.L.3, specifically required
the alternative or dedicated shutdown capability to be physically and electrically
independent of the specific fire areas and capable of accommodating post-fire
conditions where offsite power was available and where offsite power was not available
for 72-hours.
a. Inspection Scope
On a sample basis, the inspectors evaluated the adequacy of separation provided for
the power and control cabling of redundant trains of shutdown equipment. This
investigation focused on the cabling of selected components in systems important for
SSD. The inspectors review also included a sampling of components whose
inadvertent operation due to fire may adversely affect post-fire SSD capability. The
purpose of this review was to determine if a single exposure fire, in one of the fire areas
selected for this inspection, could prevent the proper operation of both SSD trains.
The inspectors evaluated selected portions of licensees fuse/breaker coordination
analysis for ground faults on the 4160 Volt alternating current (Vac) and 480Vac
systems and the vital low-voltage ac and direct current (dc) power sources to determine
whether fire-induced faults on distribution system cables or buses could degrade
post-fire SSD capability. Specifically, the inspectors determined if selective coordination
existed between branch circuit protective devices and the upstream distribution panel
fuse/breaker feeders to ensure that in the event of a fire-induced short circuit, the fault
would be isolated before the upstream feeder fuse/breaker tripped.
b. Findings
No findings of significance were identified.
.4 Alternative SSD Capability
Title 10 CFR Part 50, Appendix R, Section III.G.1, required that SSCs important to SSD
be provided with FP features capable of limiting fire damage to ensure that one train of
systems necessary to achieve and maintain hot shutdown conditions is free of fire
damage. Options for providing this level of FP were delineated in 10 CFR Part 50,
Appendix R,Section III.G.2. Where the protection of systems whose function was
required for hot shutdown did not satisfy 10 CFR Part 50, Appendix R, Section III.G.2,
an alternative or dedicated shutdown capability independent of the area under
consideration was required to be provided. Additionally, alternative or dedicated
shutdown capability must be able to achieve and maintain hot standby conditions and
achieve cold shutdown conditions within 72-hours and maintain cold shutdown
conditions thereafter. During the post-fire SSD, the reactor coolant process variables
must remain within those predicted for a loss of normal alternating current power, and
the fission product boundary integrity must not be affected (i.e., no fuel clad damage,
rupture of any primary coolant boundary, or rupture of the containment boundary).
12 Enclosure
a. Inspection Scope
The inspectors reviewed the licensees systems required to achieve alternative SSD to
determine if the licensee had properly identified the components and systems necessary
to achieve and maintain SSD conditions. The inspectors also focused on the adequacy
of the systems to perform reactor pressure control, reactivity control, reactor coolant
inventory makeup, decay heat removal, process monitoring, and support system
functions.
b. Findings
No findings of significance were identified.
.5 Operational Implementation of SSD Capability
Title 10 CFR Part 50, Appendix R, Section III.L.2.d, required that the process monitoring
function should be capable of providing direct readings of the process variables
necessary to perform and control the functions necessary to achieve reactivity control,
reactor coolant inventory makeup, and decay heat removal.
a. Inspection Scope
The inspectors reviewed a sample of the actions defined in procedures QCOA 0010-12,
Fire/Explosion, Revision 27, QCARP 0050-01, SB-1-1 Injection with SSMP and Bring
the Unit to Cold Shutdown, Revision 10, and other procedures which were referenced
by procedure QCOA 0010-12. Procedure QCARP 0050-01 was the procedure for
performing a plant alternative shutdown from outside the Unit 1 CR. The inspectors
reviewed, on a sample bases, the ability of operators to perform procedure actions
within applicable plant shutdown time requirements. The inspectors also focused on the
feasability of the actions described in the procedure.
The inspectors reviews of the adequacy of communications and emergency lighting
associated with these procedures are documented in Sections 1R05.6 and 1R05.7 of
this report.
b. Findings
b.1 Fuse Repair
Introduction: The inspectors identified a NCV of 10 CFR Part 50, Appendix B,
Criterion V, Instruction, Procedures, and Drawings, having very low safety significance
(Green) involving inadequate procedure steps. Specifically, the inspectors identified
that Procedure QOP 6500-10, Local Control of 4160 and 480 Volt Motor Operated
Circuit Breaker, Revision 8, included repair steps to replace fuses for hot shutdown in
the event of a fire. This fuse replacement constituted a hot shutdown repair which did
not meet 10 CFR Part 50, Appendix R requirements.
Description: The breakers for Unit 1's Division 1 RHR pumps and Bus 13-1/23-1
cross-tie are required to be opened and closed to safely shutdown the unit in the event
of a fire in Fire Area TB-III. Each breaker had a cable routed in Fire Area TB-III, which
13 Enclosure
had the potential to cause the breakers control power fuses to be blown. This would
have prevented the electrical operation of the breakers but would not have prevented
local manual operation as documented in QCNPSs Appendix R analysis. During the
inspectors review of Procedure QCARP 0030-01 TB-III Injection With SSMP and
Bringing the Unit to Cold Shutdown, Revision 8, the inspectors noted that in
Steps D.11.e (e.g, closes Bus 23-1 and 13-1 cross-tie breaker) and D.11.f (e.g, closes
one of the RHR pumps 1A or 1B) the operator(s) were required to close the circuit
breaker per procedure QOP 6500-10. Procedure QOP 6500-10 included a step to
replace any fuses believed to be blown prior to operating the breaker from a local
control box. The inspectors determined that since one train of RHR system was
required for torus cooling, the fuse replacement specified in this step constituted a hot
shutdown repair.
Title 10 CFR Part 50, Appendix R, Section III. G.1.a, stated One train of systems
necessary to achieve and maintain hot shutdown conditions from either the control room
or emergency control station(s) is free of fire damage; . . . and Section III.G.1.b, stated
Systems necessary to achieve and maintain cold shutdown from either the control or
emergency station(s) can be repaired within 72-hours. Based on the above, the
inspectors determined that the breakers fuse replacement did not meet Appendix R
requirements, because it constituted a hot shutdown repair, and repair is only allowed
for systems necessary to achieve and maintain cold shutdown. As a result, the licensee
entered this finding into the stations corrective action program as IR 00485702,
Required SSDA Actions Not Contained in QCARP 0030-01, dated May 2, 2006. The
licensee revised Procedure QOP 6500-10 and added steps to manually charge
breakers spring and close the breaker using a pushbutton if the fuses are not available
due to fire damage.
Analysis: The inspectors determined that the failure to include adequate steps in
Procedure QOP 6500-10 that did not meet Appendix R requirements was a
performance deficiency warranting a significance evaluation. The inspectors concluded
that the finding was greater than minor in accordance with IMC 0612, Power Reactor
Inspection Reports, Appendix B, Issue Screening, issued on September 30, 2005.
The finding involved the attribute of Procedure Quality and could have affected the
mitigating systems cornerstone objective of ensuring the availability, reliability, and
capability of systems that respond to initiating events (fire) to prevent undesirable
consequences (i.e., core damage). Specifically, the failure to include adequate steps in
the procedure to manually operate switchgear breakers and instead included a repair to
achieve and maintain hot shutdown in the event of a fire could have delayed and
complicated shutdown of the plant.
The inspectors completed a significance determination of this finding using IMC 0609,
Significance Determination Process (SDP), dated November 22, 2005, Appendix F,
Fire Protection Significance Determination Process, dated February 28, 2005. The
inspectors assigned a degradation rating of low safety significance (Green) because the
licensees R analysis did not require the repair, and the operator(s) could have manually
closed the breaker instead of replacing the fuses. Therefore, this finding screened out
as having very low safety significance (Green).
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion V, stated, in part, that
activities affecting quality shall be prescribed by documented instructions, procedures,
14 Enclosure
or drawings of a type appropriate to the circumstances and shall be accomplished in
accordance with these instructions, procedures, or drawings. Title 10 CFR Part 50,
Appendix R,Section III. G.1.a, stated One train of systems necessary to achieve and
maintain hot shutdown conditions from either the control room or emergency control
station(s) is free of fire damage; . . . and Section III.G.1.b, stated Systems necessary
to achieve and maintain cold shutdown from either the control or emergency station(s)
can be repaired within 72-hours.
Contrary to the above, Procedure QOP 6500-10, Local Control of 4160 and 480 Volt
Motor Operated Circuit Breaker, Revision 8, included steps that were not appropriate to
the circumstances. Specifically, Procedure QOP 6500-10, included steps to replace the
switchgear breakers fuses prior to operation of the breaker from a local control box.
This fuse replacement constituted a hot shutdown repair which did not meet the
requirement of 10 CFR Part 50, Appendix R and could have delayed SSD of the plant in
the event of a fire. Once identified, the licensee entered the finding into their corrective
action program as IR 00485702, Required SSDA Actions Not Contained in
QCARP 0030-01, dated May 2, 2006. Because this violation was of very low safety
significance and it was entered into the licensees corrective action program, this
violation is being treated as a NCV, consistent with Section VI.A.1 of the NRC
Enforcement Policy (NCV 05000254/2006002-03(DRS);05000265/2006002-03(DRS)).
.6 Communications
For a fire in an alternative shutdown fire area, CR evacuation may be required and a
shutdown is performed from outside the CR. Radio communications are relied upon to
coordinate the shutdown of both units and for fire fighting and security operations.
Title 10 CFR Part 50, Appendix R, Section III.H., required that equipment provided for
the fire brigade include emergency communications equipment.
a. Inspection Scope
The inspectors reviewed, on a sample bases, the adequacy of the communication
system to support plant personnel in the performance of alternative SSD functions and
fire brigade duties.
b. Findings
No findings of significance were identified.
Title 10 CFR Part 50, Appendix R, Section III.J., required that emergency lighting units
with at least an 8-hour battery power supply be provided in all areas needed for
operation of SSD equipment and in access and egress routes thereto.
15 Enclosure
a. Inspection Scope
The inspectors performed a plant walkdown of areas in which a sample of the actions
would be performed as described in procedure QCARP 0050-01, SB-1-1 Injection with
SSMP and Bring the Unit to Cold Shutdown, Revision 10, and other procedures which
were referenced by procedure QCARP 0050-01. As part of the walkdowns, the
inspectors focused on the existence of sufficient emergency lighting for access and
egress to areas and for performing necessary equipment operations.
b. Findings
No findings of significance were identified.
.8 Cold Shutdown Repairs
Title 10 CFR Part 50, Appendix R, Section III.L.5, required that equipment and systems
comprising the means to achieve and maintain cold shutdown conditions should not be
damaged by fire; or the fire damage to such equipment and systems should be limited
so that the systems can be made operable and cold shutdown achieved within 72-hours.
Materials for such repairs shall be readily available onsite and procedures shall be in
effect to implement such repairs.
a. Inspection Scope
The inspectors reviewed the licensees procedures to determine if any repairs were
required to achieve cold shutdown. The inspectors determined that the licensee did
require repair of some equipment to reach cold shutdown based on the SSD methods
used. The inspectors reviewed the procedures for adequacy. The inspectors also
reviewed completed surveillances of the tools and equipment needed to reach cold
shutdown.
b. Findings
No findings of significance were identified.
.9 Fire Barriers and Fire Zone/Room Penetration Seals
Title 10 CFR Part 50, Appendix R, Section III.M, required that penetration seal designs
be qualified by tests that are comparable to tests used to rate fire barriers.
a. Inspection Scope
The inspectors reviewed the test reports for three-hour rated barriers installed in the
plant and performed visual inspections of selected barriers to ensure that the barrier
installations were consistent with the tested configuration. In addition, the inspectors
reviewed the fire loading for selected areas to ensure that existing barriers would not be
challenged by a potential fire.
16 Enclosure
17 Enclosure
b. Findings
No findings of significance were identified.
.10 Fire Protection Systems, Features, and Equipment
a. Inspection Scope
The inspectors reviewed the material condition, operations lineup, operational
effectiveness, and design of fire detection systems, fire suppression systems, manual
fire fighting equipment, fire brigade capability, and passive FP features. The inspectors
reviewed deviations, detector placement drawings, fire hose station drawings, carbon
dioxide system pre-operational test reports, and fire hazard analysis reports to ensure
that selected fire detection systems, sprinkler systems, portable fire extinguishers, and
hose stations were installed in accordance with their design, and that their design was
adequate given the current equipment layout and plant configuration.
b. Findings
b.1 Fire Pre-Plans
Introduction: The inspectors identified a NCV of QCNPSs Operating Licenses DPR-29
and DPR-30, Section h.3.F, having very low safety significance (Green) involving the
lack of pertinent information in the QCNPSs fire pre-plans for various plant fire areas.
Specifically, the licensee failed to include important information in the fire pre-plans,
such as, hydrogen and electrical hazards, to assist the fire brigade to fight a fire within
those plant fire areas.
Description: The inspectors reviewed the licensees corrective action documents
IR 00221528, Fire Brigade Turnout Gear /SCBA Location, dated May 12, 2004, and
IR 00478821, Fire Pre-plans [Fire Area] RB-19 and [Fire Area] TB-71 Have Incorrect
Information, dated April 14, 2006. Both issue reports were initiated as a result of the
NRC resident inspector identified problems associated with the QCNPSs fire pre-plans.
One of the licensees corrective actions associated with IR 00221528 was the
implementation of a plan to improve the fire pre-plans over a four year period. Each
cycle would be six-months with the first cycle beginning July 2004 thru December 2004.
The licensee maintained a log of the fire pre-plans reviewed during each cycle. The
licensees review of IR 00478821 noted that Fire Areas RB-7 and RB-19 needed
hydrogen and oxygen hazards added to the fire pre-plans.
The inspectors observed that Fire Area RB-7 was previously reviewed on July 8, 2004,
and Fire Area RB-19 was previously reviewed on December 20, 2005, as part of the fire
pre-plan improvement effort. The inspectors review identified that the licensees
corrective actions to improve the QCNPSs fire pre-plans were not adequate. Major fire
hazards, such as, hydrogen and oxygen hazards were missed during the licensees
review. In addition, the inspectors identified additional concerns regarding fire pre-plans
as follows: 1) pre-plans did not identify all potential hazards (e.g., the location of
electrical panels that could be a risk to the fire brigade); 2) pre-plans lacked adequate
18 Enclosure
information on damper and ventilation controls (e.g., which damper and/or ventilation
controls needed to be isolated or turned off and which must be left on to cool SSD
equipment); 3) some pre-plan drawings contained errors (e.g., the quantity of fire
fighting equipment available); 4) pre-plans failed to identify the electrical panels that
were not sealed (e.g., to prevent water, if used during fire fighting, from entering
unsealed electrical panels to eliminate potential damage to both trains of equipment;
5) pre-plans did not identify which hose reels were primary or secondary hose reels to
ensure that the correct hose reel was used in a fire area (e.g., correct nozzle, length of
hose, etc.); 6) pre-plans listed transient combustibles as a hazard, when transient
combustibles may no longer be in the fire area and should have been removed from the
fire pre-plan; 7) pre-plans listed cable insulation as a hazard when cable insulation is
normally located in most fire areas; and 8) pre-plans allowed a person with a radio to be
positioned at a sprinkler control valve, where in many cases the control valve located in
the fire area .
Fire pre-plan requirements were identified in the licensees, Fire Protection Report
Volume 1, Revision 17, dated October 2005, in Section 2.0, Fire Protection Program,
Paragraph 2.5.4, Fire Fighting Strategies, which stated Pre-fire plans are provided for
all safety-related areas of the plant. The fire pre-plans provided necessary information,
including a diagram showing where fire equipment is located to aid the fire brigade in
executing manual fire fighting operations. In addition, in procedure CC-AA-211, Fire
Protection Program, Paragraph 4.9, Fire Pre-Plans, the licensee stated, The plans
are designed to provide as much useful information as possible in a short amount of
time, . . .. and . . . they provide useful information for quickly determining the
emergency response strategies based on hazards and equipment in the area.
Analysis: The inspectors determined that failure to maintain complete and accurate fire
pre-plans was a performance deficiency warranting a significance evaluation. The
inspectors concluded that the finding was greater than minor in accordance with
IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, issued
on September 30, 2005. The finding involved the attribute of protection against external
factors (i.e., fire), where failure to provide adequate warnings and guidance related to
hydrogen and electrical hazards in the fire pre-plans could have adversely impacted the
fire brigades ability to fight a fire. This would increase the likelihood of a fire which
would challenge SSD and could have affected the mitigating systems cornerstone
objective of ensuring the availability, reliability, and capability of systems that respond to
initiating events to prevent undesirable consequences (i.e., core damage). The
inspectors determined that this issue also affected the cross-cutting area of Problem
Identification and Resolution because the licensee failed to ensure that issues
potentially impacting nuclear safety were identified and fully evaluated. The licensee
failed to identify and revise the presence of hydrogen and oxygen hazards in Fire Areas
RB-7 and RB-19 during their review as part of the fire pre-plan improvement effort
conducted as a result of previously identified corrective action (IR 00221528).
The inspectors completed a significance determination of this finding using IMC 0609,
Significance Determination Process, Appendix F, Fire
Protection Significance Determination Process, dated February 28, 2005. The finding
affected the Fire Prevention and Administrative Controls Category in the area of
19 Enclosure
compliance documentation. The inspectors assigned a degradation rating of low safety
significance because extensive training was provided to fire brigade members to deal
with unexpected contingencies. In addition, other defense-in-depth FP elements
remained unaffected and fire in this area would not result in a loss of dedicated SSD
systems. The inspectors review of the Initial Qualitative Screening concluded that this
finding was considered to be of very low safety significance (Green).
Enforcement: The QCNPSs Operating Licenses DPR-29 and DPR-30, Section h.3.F,
stated that the licensee shall implement and maintain in effect all provisions of the
approved QCNPSs FPP as described in the UFSAR for the facility and as approved in
the SER dated July 27, 1979, and subsequent SER supplements. Section h.3.F, also
stated that the licensee may make changes to the approved QCNPSs FPP without prior
approval of the Commission only if those changes would not adversely affect the ability
to achieve and maintain SSD in the event of a fire. The licensees procedure
CC-AA-211, Fire Protection Program, Paragraph 4.9, Fire Pre-Plans stated The
plans are designed to provide as much useful information as possible in a short amount
of time, . . . . and . . . they provide useful information for quickly determining the
emergency response strategies based on hazards and equipment in the area.
Contrary to the above, on May 11, 2006, the licensees fire pre-plans failed to provide
useful information for quickly determining the emergency response strategies based on
hazards and equipment in the area for effective fire fighting. Specifically, the licensee
failed to maintain acceptable fire pre-plans by not providing accurate and complete and
guidance related to hydrogen and electrical hazards in the fire pre-plans which could
have adversely impacted fire brigades ability to fight a fire. The inspectors concluded
this was a violation of the QCNPSs License Condition, Section h.3.F. Once identified,
the licensee entered the finding into their corrective action program as IR 00489175,
Quality of Fire Preplans (2006 FP Triennial), dated May 11, 2006. Because this
violation was of very low safety significance and it was entered into the licensees
corrective action program, this violation is being treated as a NCV, consistent with
Section VI.A.1 of the NRC Enforcement Policy (NCV 05000254/2006002-04(DRS);05000265/2006002-04(DRS)).
b.2 Standpipes with Hose Connections
Introduction: The inspectors identified a NCV of Operating Licenses DPR-29 and
DPR-30, Section h.3.F, having very low safety significance (Green) involving adequacy
of water pressure and flow rate at standpipes with hose connections. Specifically, the
licensee failed to provide calculations to ensure that an adequate water pressure and
flow rate were available to meet the QCNPSs FPP requirements.
Description: The inspectors were concerned that standpipes with hose connections had
the potential for an inadequate water pressure and flow rate to effectively fight a fire.
The licensee stated that calculations had been made to ensure that an adequate water
pressure and flow rate were available, however, the licensee could not locate the
supporting calculations.
20 Enclosure
In NFPA 14, Standpipe and Hose Systems, Revision 1974, the requirements for
standpipe systems in Class II service required that each standpipe shall be sized for a
minimum flow of 100 gallons per minute and that standpipes in excess of 50 feet in
height shall be at least 21/2 inches in size.
The inspectors review of the licensees NFPA Code Deviation 14-06, indicated that a
few standpipes serving multiple hose connections were less than four inches in diameter
and several connections to single hose stations were less than 21/2 -inch diameter. The
licensees justification stated that calculations had been completed to verify the
adequacy of the water supply from these small pipes. In addition, the inspectors noted
that in the QCNPSs Fire Protection Report comparison against NRC Branch Technical
Position 9.5.1, Appendix A to 9.5-1 Guidelines for Fire Protection, Paragraph E.3.(d),
the licensee stated that several standpipes serving single hose connections were less
than 21/2 inch in diameter and that calculations had been completed to verify the
adequacy of the water supply from the smaller pipes.
During this inspection, the licensee completed Calculation QDC-4100-M1534, Fire
Protection Hose Reel Supply Piping Pressure Drop; dated May 11, 2006, for one of the
more limiting two inch FP supply pipelines to demonstrate that the NFPA Code
requirements were met. However, the inspectors concluded that this calculation failed
to account for the more limiting 11/2 inch piping to the standpipes with hose connections
used in the QCNPSs turbine building. The inspectors review revealed that the
calculation did not account for the water supply pressure drop that would occur at either
the fire area sprinklers or the standpipes with hose connections when a concurrent
actuation was initiated. The inspectors review of the two inch piping calculation also
revealed that there was little or no margin in water flow rate in the 11/2 inch piping. As a
result, the licensee initiated IR 00489160, Justification of Fire Hose Pressure and Flow
Meeting NFPA, dated May 11, 2006, to include in their corrective actions a plan to
perform more extensive calculations to verify water pressure and flow rate at all effected
standpipes with hose connections.
Analysis: The inspectors determined that failure to provide supporting calculations to
demonstrate adequate water pressure and flow rate at standpipes with hose
connections was a performance deficiency warranting a significance evaluation. The
inspectors concluded that the finding was greater than minor in accordance with
IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, issued
on September 30, 2005. The finding involved the attribute of protection against external
factors (i.e., fire), where the lack of water pressure and flow rate at standpipes with hose
connections could hamper the fire brigades ability to fight a fire, thereby, increasing the
likelihood of a fire which would challenge SSD and could have affected the mitigating
systems cornerstone objective of ensuring the availability, reliability, and capability of
systems that respond to initiating events to prevent undesirable consequences
(i.e., core damage).
The inspectors completed a significance determination of this finding using IMC 0609,
Significance Determination Process, dated November 22, 2005, Appendix F, Fire
Protection Significance Determination Process, dated February 28, 2005. The finding
affected the Fire Prevention and Administrative Controls Category in the area of
compliance documentation. The inspectors assigned a degradation rating of low safety
significance because other defense-in-depth FP elements remained unaffected in all fire
21 Enclosure
areas. The inspectors review of the Initial Qualitative Screening concluded that this
finding was considered to be of very low safety significance (Green).
Enforcement: The QCNPSs Operating Licenses DPR-29 and DPR-30, Section h.3.F,
stated that the licensee shall implement and maintain in effect all provisions of the
approved QCNPSs FPP as described in the UFSAR for the facility and as approved in
the SER dated July 27, 1979, and subsequent SER supplements. Section h.3.F, also
stated that the licensee may make changes to the approved QCNPSs FPP without prior
approval of the Commission only if those changes would not adversely affect the ability
to achieve and maintain SSD in the event of a fire. In addition, the licensee stated in
Appendix A to 9.5-1 Guidelines for Fire Protection, Paragraph E.3.(d) that several
standpipes serving single hose connections were less than 21/2 inch in diameter and that
calculations had been completed to verify the adequacy of the water supply from these
smaller pipelines.
Contrary to the above, on May 11, 2006, the licensee failed to provide calculations to
ensure adequate water pressure and flow rate were available at standpipes with hose
connections for effective fire fighting. The inspectors concluded this was a violation of
the QCNPSs License Condition, Section h.3.F. Once identified, the licensee entered
the finding into their corrective action program as IR 00489160, Justification of Fire
Hose Pressure and Flow Meeting NFPA, dated May 11, 2006, and planned to perform
calculations to verify water flow at all effected standpipes with hose connections.
Because this violation was of very low safety significance and it was entered into the
licensees corrective action program, this violation is being treated as a NCV, consistent
with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000254/2006002-05(DRS);05000265/2006002-05(DRS)).
b.3 Class A Fire Extinguishers
Introduction: The inspectors identified a NCV of QCNPSs Operating Licenses DPR-29
and DPR-30, Section h.3.F, having very low safety significance (Green) involving
Class A fire extinguishers. Specifically, the QCNPSs FPP did not have an adequate
number of Class A fire extinguishers available where significant fire hazards existed to
meet the NFPA 10 Code requirements to suppress and/or extinguish Class A fire
hazards.
Description: The inspectors reviewed Design Analysis WDC-4100-0691, Combustible
Loading Calculation for the Power Block, SBO Building, and Cribhouse, dated
February 18, 2005. This document showed large quantities of Class A combustibles
located in many areas of the plant. For example, Fire Zones 1.1.1.5, 1.1.1.6, and
8.2.7.A contained greater than 1000 pounds of Class A combustible materia..
During the inspectors plant walkdown, the inspectors observed that there were an
inadequate number of Class A fire extinguishers located throughout QCNPS where
22 Enclosure
significant fire hazards existed to meet the NFPA 10 Code requirements. For example,
fire zones 1.1.1.5, 1.1.1.6, and 8.2.7.A did not have Class A fire extinguishers present.
The NFPA 10 Code requirements were endorsed by the QCNPSs FPP. Specifically,
there were substantial amounts of Class A combustibles located in many areas of the
plant that would require a Class A fire extinguisher to suppress and/or extinguish this
type of fire hazard.
In Amendment Number 52 to the Facility Operating License (SER dated July 27, 1979),
Paragraph 4.3.3, Portable Fire Extinguishers, the SER stated in part, that portable dry
chemical and carbon dioxide extinguishers have been distributed throughout the plant.
The fire extinguishers meet the NFPA 10 Code requirements. The fire extinguisher
selections were made in accordance with the type of fire hazards to be protected
against at the QCNPS. In addition, the licensees responses to the Guidelines of
Appendix A to APCSB 9.5-1," Section E.6, Portable Extinguishers, stated in part, that
QCNPS complied with the NFPA 10 Code requirements for fire extinguishers. The NRC
originally approved the licensees statement of compliance with the NFPA Code during
the NRCs review of the licensee response to APCSB 9.5.1. The licensee subsequently
revised their statement of compliance with the APCSB 9.5-1 section for Class A fire
extinguishers and instead took credit for standpipes with hoses in lieu of the Class A fire
extinguishers. However, when the inspectors requested the licensees evaluation that
made this change to the QCNPSs FPP, the licensee could not locate the associated
documentation.
The inspectors noted during a review of NFPA 10, Portable Fire Extinguishers,
Revision 1975, that up to one-half of the complement of fire extinguishers may be
replaced by uniformly spaced small hose stations for use by the building occupants. In
addition, Section 1-1, Standard for Portable Fire Extinguishers, stated in part, that
portable fire extinguishers were intended as a first line of defense to cope with fires of
limited size. The fire extinguishers were needed even though the property is equipped
with automatic sprinklers, standpipes and hose, and other fixed fire protection
equipment. The NFPA Code further stated that Class A fires were fires of ordinary
combustible materials, such as wood, cloth, paper, rubber, and many plastics. In
Section 2-2.1, the NFPA Code required that fire extinguishers shall be selected for the
specific class or classes of fire hazards. Fire extinguishers for protecting Class A fire
hazards shall be selected from among water types, foam, loaded stream, and
multipurpose dry chemical.
The inspectors review of the QCNPSs NFPA Code Deviation 10-01 revealed that fire
extinguishers for Class A fire hazards were not distributed in accordance with the NFPA
Code. The licensees justification, for the NFPA Code deviation, stated that at QCNPS
the quantity of Class A combustibles was limited primarily to cable insulation and/or
transient combustibles, that proper distribution of standpipes with hose connections
were provided, and since trained on-site fire brigade personnel were present that the
limited distribution of Class A fire extinguishers was considered acceptable. As a result
of the inspectors review of Design Analysis WDC-4100-0691, the inspectors concluded
that the licensees NFPA Code Deviation 10-01, as stated above, contained an
inadequate justification for the number of Class A fire extinguishers in the plant.
23 Enclosure
Analysis: The inspectors determined that failure to have an adequate number of
Class A fire extinguishers available where significant fire hazards existed to meet the
NFPA 10 Code requirements to suppress and/or extinguish Class A fire hazards was a
performance deficiency warranting a significance evaluation. The inspectors concluded
that the finding was greater than minor in accordance with IMC 0612, Power Reactor
Inspection Reports, Appendix B, Issue Screening, issued on September 30, 2005.
The finding involved the attribute of protection against external factors (i.e., Fire), where
failure to have an adequate number of Class A fire extinguishers available could
potentially escalate a small fire into a larger fire since only standpipes with hose
connections were available and their use required a trained fire brigade to extinguish the
fire. As a result, non-fire brigade personnel would be prevented from moving quickly to
suppress and/or extinguish a small fire and the potential for an escalated fire could have
affected the mitigating systems cornerstone objective of ensuring the availability,
reliability, and capability of systems that respond to initiating events to prevent
undesirable consequences (i.e., core damage).
The inspectors completed a significance determination of this finding using IMC 0609,
Significance Determination Process, dated November 22, 2005, Appendix F, Fire
Protection Significance Determination Process, dated February 28, 2005. The finding
affected the Fire Prevention and Administrative Controls Category in the area of
compliance documentation. The inspectors assigned a degradation rating of low safety
significance because most fire areas and fire zones have fire detectors that would alarm
in the control room. In addition, other defense-in-depth FP elements remained
unaffected and fire in this area would not result in a loss of dedicated SSD systems.
The inspectors review of the Initial Qualitative Screening concluded that this finding
was considered to be of very low safety significance (Green).
Enforcement: The QCNPSs Operating Licenses DPR-29 and DPR-30, Section h.3.F,
stated that the licensee shall implement and maintain in effect all provisions of the
approved QCNPSs FPP as described in the UFSAR for the facility and as approved in
the SER dated July 27, 1979, and subsequent SER supplements. Section h.3.F, also
stated that the licensee may make changes to the approved QCNPSs FPP without prior
approval of the Commission only if those changes would not adversely affect the ability
to achieve and maintain SSD in the event of a fire. Amendment Number 52 to the
Facility Operating License (SER dated July 27, 1979), required the QCNPS to maintain
Class A fire extinguishers in accordance with NFPA 10 Code requirements and that
Class A fire extinguishers would be available for the type of fire hazard present.
Contrary to the above, from May 24, 1985, to May 12, 2006, the licensee failed to
maintain Class A fire extinguishers in accordance with NFPA 10 Code requirements and
that Class A fire extinguishers were not available for the type of fire hazard present.
Specifically, no Class A fire extinguishers were located within Fire Zone 1.1.1.5, 1.1.1.6
and 8.2.7.A even though the fire zones contained more than 1000 ponds of Class A
combustible materials. Additionally, the licensees code evaluation 10-01 was not
consistent with the results in Design Analysis WDC-4100-0091. The inspectors
concluded this was a violation of the QCNPSs License Condition, Section h.3.F. Once
identified, the licensee entered the finding into their corrective action program as
IR 00489426, Class A Fire Extinguisher Placement Improvements, dated May 12,
2006. Because this violation was of very low safety significance and it was entered into
24 Enclosure
the licensees corrective action program, this violation is being treated as a NCV,
consistent with Section VI.A.1 of the NRC Enforcement Policy
(NCV 05000254/2006002-06(DRS);05000265/2006002-06(DRS)).
25 Enclosure
.11 Compensatory Measures
a. Inspection Scope
The inspectors conducted a minimal review to verify that adequate compensatory
measures were put in place by the licensee for out-of-service, degraded or inoperable
FP and post-fire SSD equipment, systems, or features. The inspectors also conducted
a minimal review on the adequacy of short term compensatory measures to compensate
for a degraded function or feature until appropriate corrective actions were taken.
b. Findings
No findings of significance were identified.
4. OTHER ACTIVITIES
4OA2 Identification and Resolution of Problems
a. Inspection Scope
The inspectors reviewed the corrective action program procedures and samples of
corrective action documents to verify that the licensee was identifying issues related to
the FPP at an appropriate threshold and entering them in the corrective action program.
The inspectors reviewed these issues to verify an appropriate threshold for identifying
issues and to evaluate the effectiveness of corrective actions related to the FPP. In
addition, corrective action documents written on issues identified during the inspection
were reviewed to verify adequate problem identification and incorporation of the problem
into the corrective action system. The specific corrective action documents that were
sampled and reviewed by the team are listed in the attachment to this report.
b. Findings
No findings of significance were identified.
4OA5 Other Activities
(Closed) Unresolved Item 05000254/2003012-01(DRS);05000265/2003012-01(DRS):
Cable Ampacity Calculation Methodology
A Unresolved item (URI) was opened during the 2003 triennial FP inspection regarding
Calculation QDC-0000-E-0853, SLICE Cable Ampacity Multiplying Factors for Quad
Cities, Revision 0, dated December 2, 1999. Specifically, the inspectors were
concerned that the calculations methodology used to account for cable derating and/or
cable tray loading was correctly applied to potential overpowered cables. The URI was
opened pending further NRC review of the licensees calculation.
During this inspection, the inspectors reviewed the licensees activities associated with
cable ampacity derating and cable tray ampacity loading. The results of the inspectors
review indicated that the QCNPSs Calculation QDC-0000-E-0853 was not utilized at the
26 Enclosure
QCNPS. As a result, no further inspector concerns were identified with this calculation
at the present time. Since the calculation was not utilized at QCNPS, the inspectors
considered this URI closed.
4OA6 Meetings
.1 Exit Meeting
On May 12, 2006, at the end of the on-site inspection activities, the inspectors
presented the inspection results to Mr. T. Tulon and other members of licensee
management. The inspectors asked the licensee whether any materials examined
during the inspection should be considered proprietary. No proprietary information was
identified.
On June 29, 2006, at the conclusion of the inspection, a re-exit meeting conference call
was with Mr. T. Tulon and other members of licensee management to present the
inspection findings.
.2 Interim Exit Meetings
No interim exits were conducted.
4OA7 Licensee-Identified Violations
The following violations of very low safety significance (Green) were identified by the
licensee and are violations of NRC requirements which meet the criteria of Section VI of
the NRC Enforcement Policy, NUREG-1600, for being dispositioned as NCVs.
Cornerstone: Mitigating System
.1 Appendix R SSD Battery Loads
Criterion III Design Control, of 10 CFR Part 50, Appendix B, requires, in part, that
measures shall be established to assure that applicable regulatory requirements and the
design basis are correctly translated into specifications, drawings, procedures, and
instructions. Contrary to the above, the licensee did not have a design basis calculation
to verify that the 125Vdc battery/system can support the SSD loads during an
Appendix R fire for the period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Specifically, the licensee did not consider the
additional loads lineup to battery 2 as a result of a fire in Fire Area 13-1. During normal
plant operating conditions, the 125Vdc battery 1 supplied Bus 1A (Unit 1 Division I) and
Bus 2B (Unit 2 Division II), battery 2 supplied Bus 2A (Unit 2 Division I) and Bus 1B
(Unit 1 Division II). In the event of a fire in Fire Area 13-1, both battery chargers for
battery 1 were considered unavailable. Procedure QCARP 0040-01 13-1 Injection with
SSMP and Bringing the Unit to Cold Shutdown included steps connected the Unit 2
Division II, 125Vdc Bus 2B, to battery 2. These steps in the procedure would have
lined-up all three 125Vdc divisions to battery 2. A calculation was not performed to
ensure that the 125Vdc battery/charger could have provided power to three divisions
simultaneously. The inspectors determined that the finding was more than minor
because it was associate with the Mitigating System Cornerstone attribute of Design
27 Enclosure
Control and affected the cornerstone objective of ensuring the capability of systems
needed to respond to initiating events to prevent undesirable consequences.
Specifically, the failure of the fuse or the battery/charger to supply the necessary loads
for 72-hours could potentially have complicated shutdown during a fire event. The
licensee discovered this finding during their Focus Area Self Assessment (FASA) and
tracked it through Action Tracking AT 426687-15. The licensee also entered this finding
into the stations corrective action program as IR 00492546. The inspectors determined
that the finding was of very low safety significance because the licensee completed
technical evaluation EC 360895 which ensured that the battery, main fuse and charger
number 2 were adequate to supply the loads lineup included in procedure
QCARP 0040-01.
.2 Post-Fire Operator Manual Actions
On March 6, 2006, the NRC published a Federal Register Notice (FRN, Vol. 71, No. 43,
Page 11169, 10 CFR Part 50, RIN 3150 AH54) that announced the withdrawal of a
proposed rule to 10 CFR Part 50, Appendix R, Paragraph III.G.2. The proposed rule
would have revised Paragraph III.G.2 of Appendix R to allow licensees to implement
acceptable operator manual actions combined with fire detectors and automatic fire
suppression capability as an acceptable method for ensuring the capability of a licensee
to bring a reactor to, and maintain it in, a hot shutdown condition. The NRC withdrew
the proposed rule stating that 10 CFR Part 50, Appendix R, Paragraph III.G.2, cannot
be reasonably interpreted to permit reliance upon operator manual actions in lieu of the
specific methods provided in the subparagraphs of Paragraph III.G.2, to ensure that one
of the redundant SSD trains in the same fire area is free of fire damage. Therefore, any
pre-1979 licensee (i.e., QCNPS) that is using operator manual actions instead of the
specific methods provided in the subparagraphs of Paragraph III.G.2, without an
NRC-approved exemption, is not in compliance with the regulations.
Contrary to the requirements of 10 CFR Part 50, Appendix R, Paragraph III.G.2, the
QCNPSs FPP permitted reliance upon operator manual actions in lieu of the specific
methods provided in the subparagraphs of Paragraph III.G.2, to ensure that one of the
redundant SSD trains in the same fire area is free of fire damage. The licensee
recognized that many of the QCNPSs FPP operator manual actions were without
NRC-approved exemptions. As a result, the licensee generated Issue Report
00464665, NRC Terminates Fire Protection Manual Action Rulemaking, date
March 10, 2006, to acknowledge withdrawal of the proposed rule and QCNPSs
nonconformance to the regulatory requirements. Since the licensee did not dispute that
a violation of regulatory requirements had occurred, enforcement discretion has been
exercised in accordance with EGM 98-002, Revision 2, dated February 2, 2000. The
licensee has initiated corrective actions to resolve this finding within a reasonable time
frame in accordance with the published EGM and FRN. Therefore, this finding is of very
low safety significance.
ATTACHMENT: SUPPLEMENTAL INFORMATION
28 Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
W. Beck, Regulatory Assurance Manager
D. Boyles, Operations
D. Bucknell, FP Engineer
J. Burkhead, Quad Cities Nuclear Oversight
R. Buttke, Design Engineering
T. Fuhs, Regulatory Assurance
J. Garrity, Outage Control
L. Geerts, Fire Marshal
R. Gideon, Plant Manager
D. Gullott, Corporate Licensing
T. Hanley, Dresden Director of Site Engineering
K. Moser, Quad Cities Director of Site Engineering
C. Pragman, Corporate Fire Protection
J. Rathman, Design Engineering
S. Reynolds, Fire Protection System Engineer
T. Scott, Operations
P. Simpson, Corporate Licensing
M. Taylor, Corporate Fire Protection
T. Tulon, Site Vice President
M. Wagner, Regulatory Assurance
D. Wolf, Design Engineering
NRC
L. Kozak, RIII Senior Reactor Analyst
M. Kurth, Resident Inspector
J. Lara, RIII Engineering Branch 3 Chief
K. Stoedter, Senior Resident Inspector
A-1 Attachment
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
05000254/2006002-01(DRS); NCV SSMP Credited as a Redundant System for an
05000265/2006002-01(DRS) Appendix R III.G.2 Fire Area (Section 1R05.1b.1)05000254/2006002-02(DRS); NCV Failure to Ensure One Redundant Train of RHRSW
05000265/2006002-02(DRS) Free of Fire Damage (Section 1R05.1b.2)05000254/2006002-03(DRS); NCV Procedure Included Unapproved Fuse Repair for
05000265/2006002-03(DRS) Appendix R (Section 1R05.5b.1)05000254/2006002-04(DRS); NCV Failure to Maintain Acceptable Pre-Fire Plans05000265/2006002-04(DRS) (1R05.10b.1)05000254/2006002-05(DRS); NCV Failure to Have a Calculation for Hose Stations That Did
05000265/2006002-05(DRS) Not Meet Code Requirements to Ensure Adequate
Water Pressure and Flow Rate (1R05.10b.2)05000254/2006002-06(DRS); NCV Failure to Meet NFPA Code Requirements for Class A
05000265/2006002-06(DRS) Fire Extinguishers (1R05.10b.3)
Closed
05000254/2006002-01(DRS); NCV SSMP Credited as a Redundant System for an
05000265/2006002-01(DRS) Appendix R III.G.2 Fire Area (Section 1R05.1b.1)05000254/2006002-02(DRS); NCV Failure to Ensure One Redundant Train of RHRSW
05000265/2006002-02(DRS) Free of Fire Damage (Section 1R05.1b.2)05000254/2006002-03(DRS); NCV Procedure Included Unapproved Fuse Repair for
05000265/2006002-03(DRS) Appendix R (Section 1R05.5b.1)05000254/2006002-04(DRS); NCV Failure to Maintain Acceptable Pre-Fire Plans05000265/2006002-04(DRS) (1R05.10b.1)05000254/2006002-05(DRS); NCV Failure to Have a Calculation for Hose Stations That Did
05000265/2006002-05(DRS) Not Meet Code Requirements to Ensure Adequate
Water Pressure and Flow Rate (1R05.10b.2)05000254/2006002-06(DRS); NCV Failure to Meet the NFPA Code Requirements for
05000265/2006002-06(DRS) Class A Fire Extinguishers (1R05.10b.3)05000254/2003012-01(DRS); URI Cable Ampacity Calculation Methodology
05000265/2003012-01(DRS) (Section 4OA5)
Discussed
None.
A-2 Attachment
LIST OF DOCUMENTS REVIEWED
The following is a list of documents reviewed during the inspection. Inclusion on this list does
not imply that the NRC inspectors reviewed the documents in their entirety, but rather, that
selected sections of portions of the documents were evaluated as part of the overall inspection
effort. Inclusion of a document on this list does not imply NRC acceptance of the document or
any part of it, unless this is stated in the body of the inspection report.
CALCULATIONS
Number Description or Title Date or Revision
9198-13-19-1 Calc for SLICE Cable Ampacity Multiplying Factors for the August 24, 1994
Dresden and Quad Cities Stations, Rev 0
DQAMPAC Revised Cable Tray Power Cable Ampacities (S-141A) April 24, 2006
Ver 1.1
QDC-0000- SLICE Cable Ampacity Multiplying Factors for QCNPS, Rev 0 December 2, 1999
E-0853
QDC-3300- Determine Usable Volume in CCSTs for SSMP or RCIC 1
M-0542 following an App R Fire Event
WDC-4100- Combustible Load Calc for Pwr Block, SBO Bldg and Crib February 18, 2005
0691 Hse
CORRECTIVE ACTION PROGRAM DOCUMENTS ISSUED DURING INSPECTION
Number Description or Title Date or Revision
00483345 FP Report Volume 1 Part 2.1-2 Needs Correction April 26, 2006
00483355 Inconsistent Information in the SSDR April 26, 2006
00483752 Discrepancies Have Been Identified in the FHA Table 2.1-2 April 27, 2006
00485387 NFPA Code Deviations Improvement May 1, 2006
00485413 Measure Currents Flowing in Cables May 1, 2006
00485702 Required SSDA Actions Not Contained in QCARP 0030-01 May 2, 2006
00487902 SLICE Program Cable Ampacity Issues May 8, 2006
00488593 Update FP NFPA Deviations May 10, 2006
00489145 Missing Reference Letter for FP NFPA Code Deviations May 11, 2006
00489160 Justification of Fire Hose Pressure and Flow Meeting NFPA May 11, 2006
00489175 Quality of Fire Preplans (2006 FP Triennial) May 11, 2006
00489285 Relay Setting Changed Without Revising Calculation May 11, 2006
00489426 Class A Fire Extinguisher Placement Improvements May 12, 2006
00492546 No Calc for 125Vdc QCARP Lineup May 22, 2006
00502702 NRC Inspection Finding Concerning App R Redundant Trains June 22, 2006
CORRECTIVE ACTION PROGRAM DOCUMENTS ISSUED PRIOR TO INSPECTION
Number Description or Title Date or Revision
00040209 Q2000-04344 Certain FP Commitments in the 1979 December 12, 2000
00113343 Incomplete App R Revisions to Tray Routing Numbers June 26, 2002
00180384 Cable Ampacity Input Discrepancies in SLICE Database October 10, 2003
00182702 Deficiency Identified in Calculation QDC-0000-E-0853 October 24, 2003
A-3 Attachment
CORRECTIVE ACTION PROGRAM DOCUMENTS ISSUED PRIOR TO INSPECTION
Number Description or Title Date or Revision
00218683 Existing Cables Not Populated in the SLICE Database May 4, 2004
00221528 Fire Brigade Turnout Gear/SCBA Location May 12, 2004
00230758 Wrong Unit Designation on Fire Pre-Plan TB-71 June 23, 2004
00293853 Discrepancy On Fire Pre-plan TB 112 Information January 25, 2005
00464665 NRC Terminates FP Manual Action Rulemaking March 10, 2006
00478821 Fire Pre-plans RB-19 and TB-71 Have Incorrect Information April 14, 2006
DRAWINGS
Number Description or Title Date or Revision
4E-7573C S/D ATWS Recirc Pump Trip Sys Div I and II - 0
Part 3
4E-6613A S/D MOVs 1/2-2901-6 and 7 SSD System F
4E-1344 S/D 4160V Buses 13-1 and 14-1 Main Feed J
Breakers
M-3 General Arrangement Main Floor Plan K
M-4 General Arrangement Mezzanine Floor Plan H
M-6 General Arrangement Basement Floor Plan C
M-70 Diagram of SSMP System W
QDC-0000-E-1038, App R AC Distribution Diagram 0
Attach F
QDC-0000-E-1038, App R 250Vdc Distribution Diagram, Rev 0 November 14, 2000
Attach F, Pg 2 of 3
QDC-0000-E-1038, App R 125Vdc Distribution Diagram, Rev 0 November 13, 2000
Attach F, Pg 3 of Final
ENGINEERING ACTION PLANS
Number Description or Title Date or Revision
00-00-24, Rev 0 SLICE Program and Backlog Reduction February 19, 2001
00-00-24, Rev 1 SLICE Program and Backlog Reduction March 28, 2001
00-00-24, Rev 2 SLICE Program and Backlog Reduction October 15, 2004
ENGINEERING CHANGES (ECs)
Number Description or Title Date or Revision
354100 001 Abandonment of CO2 Hose Reels December 9, 2005
0000360804 Address Triennial Fire Inspection Tray Ampacity Issues May 9, 2006
Related to Tray Nodes 261M1, 320B, 323M, 325B, and 329B
0000360805 Evaluate SLICE Cable Ampacity Issues in Response to May 8, 2006
Triennial FP Inspection Request
EVALUATIONS
Number Description or Title Date or Revision
Portable Fire Extinguishers (NFPA 10 and 10A) May 24, 1985
A-4 Attachment
EVALUATIONS
Number Description or Title Date or Revision
Std for Installation of Standpipe and Hose Systems (NFPA 14) June 19, 1985
Fire Doors in Nuclear SR Areas April 9, 1987
FP Survey of HVAC Systems in Computer/Nuclear SR Areas 8
IMPAIRMENT/REMOVAL PERMITS
Number Description or Title Date or Revision
List of Plant FP Impairments April 26, 2006
PROCEDURES
Number Description or Title Date or Revision
CC-AA-211 Fire Protection Program 2
CC-AA-302 Control of the Cable Management Database 2
QCOA 0010-12 Fire/Explosion 27
QCARP 0030-01 TB-III Injection with SSMP and Bringing the Unit to Cold 8
QCARP 0040-01 13-1 Injection with SSMP and Bringing the Unit to Cold 7
QCARP 0040-02 24-1 Injection with SSMP and Bringing the Unit to Cold 9
QCARP 0050-01 SB-1-1 Injection with SSMP and Bring the Unit to Cold 10
QCMMS 4100-61 Fire Door Inspection 11
QCOP 2900-02 SSMP System Start Up 18
QOP 6500-10 Local Control of 4160 and 480V MOV Circuit Breakers 8 and 9
REFERENCES
Number Description or Title Date or Revision
Pre-fire Strategies for Reactor Bldg and Turbine Bldg April 26, 2006
FPR Vol 1 and 2 QCNPSs Fire Protection Report October 2005
GE Spec 22A2501 GE Specification for Domestic Turnkey Projects 0
GE-NE-T43-00002 Original SSD Paths for the BWR 1
-00-01-R01
GE-NE-T43-00002 BWROG Position on the Use of Safety Relief Valves 1
-00-03-R01 and Low Pressure Systems as Redundant SSD Paths
NFPA 14 Standpipe and Hose Systems 1974
NFPA 10 Portable Fire Extinguishers 1975
Q-ECDS-960134 QCNPS Cable Ampacity Report July 30, 1996
SandL LTR ComEd SLICE Discrepancy Resolution Project December 23, 1999
D-3477E
SandL LTR ComEd QCNPS Ampacity Evaluation Info July 2, 1996
Q-2148E
NUREG/CR-6681 Ampacity Derating and Cable Functionality for August 2000
(SAND2000-1825) Raceway Fire Barriers
SandL Program DQAMPAC Users Manual - Revised SLICE Ampacity August 31, 1995
No. 03.7.528-1.1 Software
A-5 Attachment
A-6 Attachment
LIST OF ACRONYMS USED
AC or ac Alternating Current
ADAMS Agency-Wide Document Access and Management System
App Appendix
ATTN Attention
BWROG Boiling Water Reactor Owners Group
CFR Code of Federal Regulations
CR Control Room
DC or dc Direct Current
DPR Demonstration Power Reactor
DRP Division of Reactor Projects
DRS Division of Reactor Safety
FP Fire Protection
GL Generic Letter
FRN Federal Register Notice
IMC Inspection Manual Chapter
IP Inspection Procedure
IPEEE Individual Plant Examination of External Events
IR Inspection Report
k kilo
LLC Limited Liability Company
MOV Motor Operated Valve
NFPA National Fire Protection Association
NRC Nuclear Regulatory Commission
NRR Office of Nuclear Reactor Regulation
NUREG NRC Technical Report Designation
PARS Publicly Available Records
QCNPS Quad Cities Nuclear Power Station
RIII Region III
RHRSW Residual Heat Removal Service Water
S/D Schematic Diagram
SandL Sargent and Lundy
SD Shutdown
SDP Significance Determination Process
SER Safety Evaluation Report
SLICE Sargent and Lundy Interactive Cable Engineering
SR Safety Related
SSA Safe Shutdown Analysis
SSCA Safe Shutdown Capability Assessment
SSCs Structures, Systems and Components
SSD Safe Shutdown
SSMP Safe Shutdown Makeup Pump
SSSA Safe Shutdown System Analysis
UFSAR Updated Final Safety Analysis Report
URI Unresolved Item
A-7 Attachment
A-8 Attachment