ML060330616
| ML060330616 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 02/01/2006 |
| From: | Laura Smith Division of Reactor Safety IV |
| To: | Muench R Wolf Creek |
| References | |
| IR-05-008 | |
| Download: ML060330616 (32) | |
See also: IR 05000482/2005008
Text
February 1, 2006
Rick A. Muench, President and
Chief Executive Officer
Wolf Creek Nuclear Operating Corporation
P.O. Box 411
Burlington, KS 66839 Wolf Creek Nuclear Operating Corporation
SUBJECT: WOLF CREEK GENERATING STATION - INSPECTION REPORT
Dear Mr. Muench:
On December 29, 2005, the Nuclear Regulatory Commission (NRC) completed an inspection at
the Wolf Creek Generating Station. The enclosed report documents the inspection findings,
which were discussed in a debrief meeting at the end of the onsite inspection on
December 2, 2005, with you and other members of your staff and again in an exit meeting
conducted via conference call on December 29, 2005.
During this triennial fire protection inspection, the inspection team examined activities
conducted under your license related to safety and compliance with the Commissions rules and
regulations and the conditions of your license. The inspection consisted of selected
examination of procedures and records, observations of activities and installed plant systems,
and interviews with personnel.
During the inspection, two apparent violations related to compliance with the requirements of
the approved Fire Protection Program were identified. These findings involved analysis and
procedure inadequacies related to fire damage induced spurious actuations of components.
These circuit vulnerabilities, could, under certain postulated fire scenarios, adversely affect the
ability to achieve and maintain safe shutdown of the facility. It is the NRCs understanding that
you do not consider these vulnerabilities to be violations of NRC requirements. In order to allow
the industry to develop an acceptable approach to resolving this issue, that the NRC can
endorse, the NRC will defer any enforcement action relative to these matters while the staff
evaluates NEIs proposed resolution methodology for circuit vulnerabilities and you have time to
implement the resolution methodology, once approved, provided you take adequate
compensatory measures for the identified vulnerabilities.
Based on the results of this inspection, the NRC has also identified two findings that were
evaluated under the risk significance determination process as having very low safety
significance (Green). The NRC has determined that these findings involve violations of NRC
requirements. These violations are being treated as noncited violations, consistent with
Section VI.A of the Enforcement Policy. These noncited violations are described in the subject
inspection report. If you contest the violations or their significance, you should provide a
Wolf Creek Nuclear Operating Corporation -2-
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 copies to the Regional Administrator, U.S. Nuclear Regulatory Commission,
Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011; the Director, Office of
Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the
NRC Resident Inspector at the Wolf Creek facility.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its
enclosure, and your response will be made 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 Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
//RA//
Linda Joy Smith, Chief
Engineering Branch 2
Division of Reactor Safety
Docket: 50-482
License: NPF-42
Enclosure:
NRC Inspection Report 05000482/2005008
w/attachment: Supplemental Information
cc w/enclosure:
Vice President Operations/Plant Manager
Wolf Creek Nuclear Operating Corp.
P.O. Box 411
Burlington, KS 66839
Jay Silberg, Esq.
Shaw Pittman, LLP
2300 N Street, NW
Washington, DC 20037
Supervisor Licensing
Wolf Creek Nuclear Operating Corp.
P.O. Box 411
Burlington, KS 66839
Wolf Creek Nuclear Operating Corporation -3-
Chief Engineer
Utilities Division
Kansas Corporation Commission
1500 SW Arrowhead Road
Topeka, KS 66604-4027
Office of the Governor
State of Kansas
Topeka, KS 66612
Attorney General
120 S.W. 10th Avenue, 2nd Floor
Topeka, KS 66612-1597
County Clerk
Coffey County Courthouse
110 South 6th Street
Burlington, KS 66839-1798
Vick L. Cooper, Chief, Air Operating
Permit and Compliance Section
Kansas Department of Health and
Environment
Bureau of Air and Radiation
1000 SW Jackson, Suite 310
Topeka, KS 66612-1366
Wolf Creek Nuclear Operating Corporation -4-
Electronic distribution by RIV:
Regional Administrator (BSM1)
DRP Director (ATH)
DRS Director (DDC)
DRS Deputy Director (RJC1)
Senior Resident Inspector (SDC)
Resident Inspector (TBR2)
SRI, Callaway (MSP)
Branch Chief, DRP/B (WBJ)
Senior Project Engineer, DRP/B (RAK1)
Team Leader, DRP/TSS (RLN1)
RITS Coordinator (KEG)
J. Dixon-Herrity, OEDO RIV Coordinator (JLD)
ROPreports
WC Site Secretary (SLA2)
SUNSI Review Completed: __Yes_ ADAMS: / Yes G No Initials: __LJS___
/ Publicly Available G Non-Publicly Available G Sensitive / Non-Sensitive
R:REACTORS\WC\2005\WC 2005-008RP-JMM.wpd
RIV:DRS/EB2 RIV:DRS/EB2 RIV:DRS/EB2 RIV:DRS/EB2
JMMateychick DLLivermore RMullikin BTindell
/RA/ /RA/ /RA/ /RA/
1/12 /06 1/12/06 1/12 /06 1/18/06
RIV:DRS/EB2 C:DRP/B C:DRS/PEB
DHOverland WBJones LJSmith
/RA/ /RA/ /RA/
1/12/06 1/18/06 2/1/06
OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket: 50-482
License: NPF-42
Report: 05000482/2005008
Licensee: Wolf Creek Nuclear Operating Corporation
Wolf Creek Generating Station
Location: 1550 Oxen Lane NE
Burlington, Kansas
Dates: October 24 through December 29, 2005
Team Leader J. M. Mateychick, Senior Reactor Inspector, Engineering Branch 2
Inspectors: D. L. Livermore, Reactor Inspector, Engineering Branch 2
D. H. Overland, Reactor Inspector, Engineering Branch 2
B. Tindell, Reactor Inspector, Engineering Branch 2
Accompanying R. Mullikin, Consultant
Personnel:
Approved By: Linda Joy Smith, Chief
Engineering Branch 2
Division of Reactor Safety
Enclosure
SUMMARY OF FINDINGS
IR 500482/2005008; 10/24/05 - 12/29/05; Wolf Creek Nuclear Operating Corporation; Wolf
Creek Generating Station; Fire Protection (Triennial)
The NRC conducted an inspection with a team of four regional inspectors and one contractor.
The inspection identified two apparent violations, two Green noncited violations (NCV) and two
unresolved items (URI). The significance of most findings is indicated by their color (Green,
White, Yellow, Red) using MC 0609 Significance Determination Process (SDP). Findings for
which the significance determination process does not apply may be Green or may be assigned
a severity level after NRC management review. The NRC describes its program for overseeing
the safe operation of commercial nuclear power reactors in NUREG-1649, Reactor Oversight
Process, Revision 3, dated July 2000.
A. NRC-Identified and Self Revealing Findings
Cornerstone: Mitigating Systems
C Green. The team identified a noncited violation (NCV) for failure to comply with
Technical Specification 5.4, Procedures, in that a procedure required for post-fire safe
shutdown was found to be inadequate. Procedure OFN RP-014, Hot Standby to Cold
Shutdown from Outside the Control Room, was inadequate because it did not provide a
method to provide sufficiently borated water to the reactor coolant system so that cold
shutdown could be achieved and maintained within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after a control room fire.
Procedure OFN RP-014 requires monitoring of the boron concentration in the reactor
and, if necessary, starting the acid transfer pumps to draw borated water from the boric
acid tanks. However, this procedure did not include sufficient instructions for refilling
and borating the Refueling Water Storage Tank for a potential loss of offsite power or
fire induced damage to circuits related to the pumps.
This finding is greater than minor because it impacted the mitigating systems
cornerstone objective to ensure the availability, reliability, and capability of systems that
respond to external events (such as fire) to prevent undesirable consequences (i.e.,
core damage). The inspectors evaluated the finding using MC 0609, Appendix F, and
determined that it screens as very low safety significance (Green) because it is related
to the ability to achieve and maintain cold shutdown. (Section 1R05.1.b.(1))
C TBD. The team identified an Apparent Violation of Wolf Creek License
Condition 2.C.(5)(a) concerning an inadequate alternative shutdown analysis. The
licensees alternative shutdown analysis was inadequate in that it used an acceptance
criteria which was inconsistent with and less conservative than that required by the
approved Fire Protection Program. The licensee developed Calculation
Number AN-02-021, Revision 0, OFN RP-017, Control Room Evacuation,
Consequence Evaluation, to demonstrate alternative shutdown capability for Wolf
Creek in response to NRC-identified Noncited Violation 2002008-01, Inadequate
alternative shutdown procedure. The calculation predicted that during an alternative
shutdown, the reactor coolant system subcooling margin would not be maintained,
significant voiding would occur in the core, and a steam void would form in the reactor
Enclosure
-2-
vessel head. The licensee found the results of the calculation to be acceptable since it
demonstrated that the void formation would be limited, natural circulation in the reactor
coolant system would be maintained, sufficient decay heat removal would be
maintained, and no fuel damage would occur. This is not consistent with the license
condition to meet the technical requirements of 10 CFR Part 50, Appendix R.
Section III.L of 10 CFR Part 50, Appendix R, Alternative and dedicated shutdown
capability, that states in part, During the postfire shutdown, the reactor process
variables shall be maintained within those predicted for a loss of normal a.c. power.
This finding is greater than minor because it impacted the mitigating systems
cornerstone objective to ensure the availability, reliability, and capability of systems that
respond to external events (such as fire) to prevent undesirable consequences (i.e.,
core damage). It is the NRCs understanding that the licensee does not consider these
circuit vulnerabilities to be violations of NRC requirements. The licensee considers the
spurious operation of multiple components to be outside of the plant licensing basis for
the Fire Protection Program. Specifically, in this case, both pressurizer power-operated
relief valves are assumed to spuriously open because of fire induced circuit damage.
The NRC staff and the industry are currently working on developing a resolution
methodology to address these types of potential fire induced circuit failures. The team
concluded that this violation meets the criteria of the NRC Enforcement Manual
Section 8.1.7.1 for deferring enforcement actions for postulated fire induced circuit
failures. (Section 1R05.1.b.(2))
C Green. The team identified a noncited violation of License Condition 2.C.(5), Fire
Protection (Section 9.5.1, SER; Section 9.5.1.8, SSER #5), for failure to ensure that
redundant trains of safe shutdown systems in the same fire area were free of fire
damage. The licensee credited manual actions to mitigate the effects of fire damage in
lieu of providing the physical protection required by 10 CFR Part 50, Appendix R,
Section III.G.2.
SNUPPS FSAR Appendix 9.5E provided the design comparison between the plants fire
protection program and 10 CFR Part 50, Appendix R. The comparison to Section III.G,
Fire Protection of Safe Shutdown Capability, states, Redundant trains of systems
required to achieve and maintain hot standby are separated by 3-hour-rated fire
barriers, or the equivalent provided by III.G.2, or else a diverse means of providing the
safe shutdown capability exists that is unaffected by the fire. Wolf Creek has
interpreted diverse means as by any reasonable means including local valve and
breaker operations as long as they are within the scope of normal operator duties. The
team disagrees with this interpretation. The NRC staff does not recognize the use of
manual actions as meeting the technical requirements of Appendix R,Section III.G.2.
The components being operated are identified as required for operation of safe
shutdown systems or are subject to potential spurious operation impacting the
shutdown. The local manual actions are being performed because of fire damage to
electrical cables related to those components and are meant to compensate for damage
or maloperation of safe shutdown equipment caused by fire.
Enclosure
-3-
This finding is greater than minor because it impacted the mitigating systems
cornerstone objective to ensure the availability, reliability, and capability of systems that
respond to external events (such as fire) to prevent undesirable consequences (i.e.,
core damage). The team found that the manual operator actions implemented to
mitigate the effects of fire damage were reasonable (as defined in Enclosure 2 of NRC
Inspection Procedure 71111.05T, Fire Protection (Triennial)), and could be performed
within the analyzed time limits. Therefore, in accordance with Enclosure 2 of
NRC Inspection Procedure 71111.05T, the finding was determined to be of very low
safety significance (Green), and the significance determination process was not entered.
(Section 1R05.2)
C TBD. The team identified an Apparent Violation of Technical Specification 5.4,
Procedures, due to an inadequate alternative shutdown procedure that is required for
implementation of the Fire Protection Program. The team found that some time critical
actions required to safely shutdown the plant following a control room fire could not be
accomplished within the required time periods. Specifically, the team found that the
recommendations by Westinghouse Owners Group for assuring reactor coolant
pump seal reliability and avoiding component cooling water thermal barrier water
hammer concerns would not be met if the operators had to respond to multiple spurious
operations. The procedure was developed and verified based on a time line assuming
operators only have to respond to one spurious operation from the fire induced damage
during the scenario. The team disagrees with this limitation of potential spurious
operations.
This finding is greater than minor because it impacted the mitigating systems
cornerstone objective to ensure the availability, reliability, and capability of systems that
respond to external events (such as fire) to prevent undesirable consequences (i.e.,
core damage). It is the NRCs understanding that the licensee does not consider these
circuit vulnerabilities to be violations of NRC requirements. The licensee considers the
spurious operation of multiple components to be outside of the plant licensing basis for
the Fire Protection Program. The NRC staff and the industry are currently working on
developing a resolution methodology to address these types of potential fire induced
circuit failures. The team concluded that this violation meets the criteria of the NRC
Enforcement Manual Section 8.1.7.1 for deferring enforcement actions for postulated
fire induced circuit failures. (Section 1R05.6.b.(2))
B. Licensee-Identified Violations
None
Enclosure
REPORT DETAILS
1 REACTOR SAFETY
1R05 Fire Protection
The purpose of this inspection was to review the Wolf Creek Generating Stations fire
protection program for selected risk-significant fire areas. Emphasis was placed on
verification of the post-fire safe shutdown capability. The inspection was performed in
accordance with the NRC regulatory oversight process using a risk-informed approach
for selecting the fire areas and attributes to be inspected. The team used the
Individual Plant Examination for External Events for the Wolf Creek Generating Station
to choose risk-significant areas for detailed inspection and review. Inspection
Procedure 71111.05T, Fire Protection (Triennial), requires selecting three to five fire
areas for review. The four areas reviewed during this inspection were:
Fire Area A-8: Auxiliary Building - 2000 Elevation, General Area
Fire Area A-18: Auxiliary Building - 2026' Elevation, Electrical Penetration Room
(North)
Fire Area A-27: Auxiliary Building - 2026' Elevation, Reactor Trip Switchgear
Room
Fire Area C-9: Control Building Elevation - 2000', ESF Switchgear Room (North)
For each of these fire areas, the inspection focused on fire protection features, systems
and equipment necessary to achieve and maintain safe shutdown conditions, and
licensing basis commitments.
Documents reviewed by the team are listed in the attachment.
.1 Shutdown From Outside Main Control Room
a. Inspection Scope
The team reviewed the functional requirements identified by the licensee as necessary
for achieving and maintaining hot shutdown conditions to ensure that at least one
post-fire safe shutdown success path was available in the event of fire in each of the
selected areas and alternative shutdown for the case of control room evacuation. The
team reviewed piping and instrumentation diagrams of systems credited in
accomplishing safe shutdown functions to independently verify whether the shutdown
methodology had properly identified the required components. The team focused on the
following functions that must be available to achieve and maintain safe shutdown
conditions:
Reactivity control capable of achieving and maintaining cold shutdown reactivity
conditions;
Enclosure
-2-
Reactor coolant makeup capable of maintaining the reactor coolant inventory;
Reactor heat removal capable of achieving and maintaining decay heat removal;
Supporting systems capable of providing other services necessary to permit extended
operation of equipment necessary to achieve and maintain hot shutdown conditions; and
Verification that a safe shutdown can be achieved and maintained with and without
off-site power.
A review was also conducted to ensure that all required components in the selected
systems were included in the safe shutdown analysis. The team identified the systems
required for each of the primary safety functions necessary to achieve and maintain
shutdown conditions. These systems were then evaluated to identify the systems that
interfaced with the selected fire areas and were the most risk significant systems
required for reaching hot shutdown conditions.
b. Findings
(1) Failure to Provide Adequate Post-Fire Shutdown Procedures
Introduction. The team identified a Green noncited violation (NCV) for failure to comply
with Technical Specification 5.4, Procedures. Procedure OFN RP-014, Hot Standby to
Cold Shutdown from Outside the Control Room, was inadequate because it did not
provide a method to provide sufficiently borated water to the reactor coolant system so
that cold shutdown could be achieved and maintained within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after a control
room fire.
Description. Wolf Creek utilizes Procedure OFN RP-014, Hot Standby to Cold
Shutdown from Outside the Control Room, to satisfy the fire protection program
requirement to achieve and maintain cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after a control room
fire. Following the fire, borated water must be injected into the reactor coolant system to
make up for reactor coolant pump seal leakage, control reactor coolant system inventory
during the cooldown and maintain cold shutdown reactivity conditions.
Procedure OFN RP-017, Control Room Evacuation, provides instructions for
performing an alternative shutdown from outside of the control room to establish stable
hot shutdown conditions. Procedure OFN RP-017 includes steps to mitigate potential
spurious actuations that could divert required inventory of borated water from the
Reactor Water Storage Tank. For example, operation of the containment spray system
would divert water to the containment until the spuriously operating pump was secured.
The team identified that in this case the Reactor Water Storage Tank would not contain
enough borated water to maintain reactivity less than 0.99 for the required 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
assuming that the containment spray system spuriously operates along with the
assumed loss of offsite power during a control room fire. Procedure OFN RP-014
requires monitoring of the boron concentration in the reactor and, if necessary, starting
Enclosure
-3-
the boric acid transfer pumps to draw borated water from the boric acid tanks. However,
this procedure did not include any instructions under the Response Not Obtained
column should the operation not be accomplished because of a loss of offsite power or
fire induced damage to circuits related to the pumps.
Analysis. The inspectors referred to the guidance of MC 0612 and determined that the
finding is greater than minor in that it affected the ability to makeup borated water to the
reactor coolant system following a control room fire and a spurious operation of the
containment spray system. This finding is associated with the Mitigating Systems
cornerstone and the respective attribute of procedure quality. This finding impacted the
mitigating systems cornerstone objective to ensure the availability, reliability, and
capability of systems that respond to external events (such as fire) to prevent
undesirable consequences. The inspectors evaluated the finding using MC 0609,
Appendix F, and determined that it screens as very low safety significance (Green)
because it is related to the ability to achieve and maintain cold shutdown. The licensee
documented the teams concern in PIR 2005-3033. The licensee has revised
Procedure OFN RP-014 to include steps to use the diesel driven fire pump to refill the
Reactor Water Storage Tank as needed and detailed instructions how to isolate boric
transfer pump circuits from the control room and restore operability. The licensee has
also pre-staged the required electrical jumpers and fuses.
Enforcement. Wolf Creek Technical Specifications, Section 5.4.1 states, in part,
Written Procedures shall be established, implemented, and maintained covering the
following activities:.... d. Fire Protection Program implementation. License
Condition 2.C.(5)(a) states, The Operating Corporation shall maintain in effect all
provisions of the approved fire protection program as described in the SNUPPS Final
Safety Analysis Report for the facility through Revision 17, the Wolf Creek site
addendum through Revision 15, and as approved in the SER through Supplement 5,
subject to provisions b & c below. Safety Evaluation Report, Section 9.5.1.7,
Appendix R Statement, states, The staff will condition the operating license to require
the applicant to meet the technical requirements of Appendix R to 10 CFR Part 50, or
provide equivalent protection.Section III.L.3 of Appendix R states, The shutdown
capability for specific fire areas may be unique for each such area, or it may be one
unique combination of systems for all such areas. In either case, the alternative
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 shall be in effect to implement this
capability.
Contrary to the above, Procedure OFN RP-014 did not contain adequate instructions to
assure an adequate supply of borated water. Because this finding is of very low safety
significance and the licensee has already completed corrective actions, this violation is
being treated as a noncited violation, consistent with Section VI.A of the NRC
Enforcement Policy: NCV 05000482/2005008-01, Failure to Provide Adequate
Post-Fire Shutdown Procedures.
Enclosure
-4-
(2) Failure to Maintain Reactor Coolant System Subcooling During the Alternative Shutdown
Introduction. The team identified an Apparent Violation of License Condition 2.C.(5)(a)
concerning an inadequate alternative shutdown analysis. The alternative shutdown
analysis was inadequate in that it used acceptance criteria which was inconsistent with
and less conservative than that required by the approved Fire Protection Program.
Description. The licensee developed Calculation Number AN-02-021, Revision 0,
OFN RP-017, Control Room Evacuation, Consequence Evaluation, to demonstrate
alternative shutdown capability for Wolf Creek in response to NRC-identified Noncited
Violation 2002008-01, Inadequate alternative shutdown procedure. The original basis
for the time critical actions in Procedure OFN RP-017 was the phased procedural
approach outlined in Licensee Letter SLNRC 84-0109, dated August 23, 1984. This
alternative shutdown methodology was found acceptable by the NRC as documented in
Supplemental Safety Evaluation Report 5. No detailed thermal-hydraulic analysis of the
plant response during the alternative shutdown had been performed at that time. In
developing Calculation Number AN-02-021, the licensee used no fuel damage as an
acceptance criteria. The calculation predicted that during an alternative shutdown, the
reactor coolant system subcooling margin would not be maintained, significant voiding
would occur in the core, and a steam void would form in the reactor vessel head. The
licensee found the results of the calculation to be acceptable since it demonstrated that
the void formation would be limited, natural circulation in the reactor coolant system
would be maintained, sufficient decay heat removal would be maintained, and no fuel
damage would occur.
The teams review of the approved Fire Protection Program noted that the plant must
meet the technical requirements of 10 CFR Part 50, Appendix R, Fire Protection
Program for Nuclear Power Facilities Operating Prior to January 1, 1979.Section III.L
of 10 CFR Part 50 Appendix R, Alternative and dedicated shutdown capability, states
in part, During the postfire shutdown, the reactor process variables shall be maintained
within those predicted for a loss of normal a.c. power. The predicted plant response
documented in Wolf Creek UFSAR, Chapter 15, Section 15.2.6, Loss of
non-emergency AC power to the station auxiliaries (blackout), maintains reactor coolant
system subcooling margin and no void formation in the reactor vessel head occurs.
Therefore, the team considered the acceptance criteria used in Calculation Number
AN-02-021 to not be in compliance with the approved Fire Protection Program.
Analysis. The inspectors referred to the guidance of MC 0612 and determined that the
finding is greater than minor in that it affected the ability to achieve and maintain hot
shutdown following a control room fire. This finding is associated with the Mitigating
Systems cornerstone and the respective attribute of protection against external factors
(e.g., fire). This finding impacted the mitigating systems cornerstone objective to ensure
the availability, reliability, and capability of systems that respond to external events (such
as fire) to prevent undesirable consequences.
During the inspection, the licensee contended that the evaluation was overly
conservative in that it assumed multiple fire induced spurious operations, while their
Enclosure
-5-
licensing basis only required one worst case spurious operation for the design of
alternative shutdown capability. Calculation Number AN-02-021 assumed the spurious
operation of both pressurizer power-operated relief valves. However, the licensee
initiated compensatory measures consisting of stationing additional fire watch personnel
in the control room to increase surveillance for potential fire hazards and fires in the
incipient stage. The team did not enter the Significance Determination Process at this
time because the enforcement is being deferred as discussed below and the licensee
has established adequate compensatory measures. Therefore, the significance will be
determined after the NRC endorses a path to resolution for fire induced circuit failures.
Enforcement. License Condition 2.C.(5)(a) states, The Operating Corporation shall
maintain in effect all provisions of the approved fire protection program as described in
the SNUPPS Final Safety Analysis Report for the facility through Revision 17, the Wolf
Creek site addendum through Revision 15, and as approved in the SER through
Supplement 5, subject to provisions b & c below. The Safety Evaluation Report,
Section 9.5.1.7, Appendix R Statement, states, The staff will condition the operating
license to require the applicant to meet the technical requirements fo Appendix R to
10 CFR 50, or provide equivalent protection. Wolf Creek SER, Supplement 3 states,
Based on our review, the staff concludes that the alternative shutdown capability for the
control room meets the requirements of Appendix R,Section III.L, and is therefore
acceptable. Title 10 CFR Part 50, Appendix R, Section III.L. 1 specifies, in part, that
during alternative post-fire shutdown, the reactor coolant system process variables shall
be maintained within those predicted for a loss of normal a.c. power.
Contrary to the above, the alternative shutdown methodology in Procedure OFN RP-017
as evaluated in Calculation Number AN-02-021 fails to maintain reactor coolant process
variables (e.g., pressure, temperature, and subcooling margin) within those predicted for
a normal loss of AC power. It is the NRCs understanding that the licensee does not
consider these vulnerabilities to be violations of NRC requirements. The licensee
considers the spurious operation of multiple components to be outside of the plant
licensing basis for the Fire Protection Program. Specifically, in this case, both
pressurizer power-operated relief valves are assumed to spuriously open because of fire
induced circuit damage. The NRC staff and the industry are currently working on
developing a resolution methodology to address these types of potential fire circuit
failures. The teams review concluded that this violation meets the criteria of the NRC
Enforcement Manual Section 8.1.7.1 for deferring enforcement actions for postulated
fire induced circuit failures. This violation is being treated as an apparent violation:
AV 05000482/2005008-02, Failure to Maintain Reactor Coolant System Subcooling
During the Alternative Shutdown.
Enclosure
-6-
.2 Protection of Safe Shutdown Capabilities
a. Inspection Scope
The team reviewed the piping and instrumentation diagrams, safe shutdown equipment
list, safe shutdown design basis documents, and the post-fire safe shutdown analysis to
verify whether the shutdown methodology had properly identified the components and
systems necessary to achieve and maintain safe shutdown conditions for equipment in
the fire areas selected for review. The team also reviewed and observed walkdowns of
the procedures for achieving and maintaining safe shutdown in the event of a fire to
verify that the safe shutdown analysis provisions were properly implemented. The team
focused on the following functions that must be ensured to achieve and maintain
post-fire safe shutdown conditions: (1) reactivity control capable of achieving and
maintaining cold shutdown reactivity conditions, (2) reactor coolant makeup capable of
maintaining the reactor coolant level within the level indication in the pressurizer,
(3) reactor heat removal capable of achieving and maintaining decay heat removal,
(4) supporting systems capable of providing all other services necessary to permit
extended operation of equipment necessary to achieving and maintaining hot shutdown
conditions, and (5) process monitoring capable of providing direct readings to perform
and control the above functions.
The team reviewed the separation of safe shutdown cables, equipment, and
components within the same fire areas, and reviewed the methodology for meeting the
requirements of 10 CFR 50.48, Appendix A to Branch Technical Position 9.5-1 and
10 CFR Part 50, Appendix R, Section III.G. Specifically, this was to determine whether
at least one post-fire safe shutdown success path was free of fire damage in the event
of a fire in the selected areas. The evaluation focused on the cabling of selected
components for the chemical and volume control system, high pressure safety injection
system, and the auxiliary feedwater system. A sample of components was selected
whose inadvertent operation could significantly affect the shutdown capability credited in
the safe shutdown analysis. The specific components selected are listed in the
attachment. In addition, the team reviewed license documentation, such as NRC safety
evaluation reports, the Wolf Creek Updated Final Safety Analysis Report, submittals
made to the NRC by the licensee in support of the NRC's review of their fire protection
program, and deviations from NRC regulations to verify that the licensee met license
commitments.
b. Findings
Introduction. The team identified a noncited violation of License Condition 2.C.(5), Fire
Protection (Section 9.5.1, SER; Section 9.5.1.8, SSER #5), for failure to ensure that
redundant trains of safe shutdown systems in the same fire area were free of fire
damage. The licensee credited manual actions to mitigate the effects of fire damage in
lieu of providing the physical protection required by 10 CFR Part 50, Appendix R,
Section III.G.2. The team determined that the violation was of very low safety
significance (Green).
Enclosure
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Description. License Condition 2.C.(5)(a) states, The Operating Corporation shall
maintain in effect all provisions of the approved fire protection program as described in
the SNUPPS Final Safety Analysis Report for the facility through Revision 17, the Wolf
Creek site addendum through Revision 15, and as approved in the SER through
Supplement 5, subject to provisions b & c below. SER Section 9.5.1.7, Appendix R
Statement, states, "The staff will condition the operating license to require the applicant
to meet the technical requirements fo Appendix R to 10 CFR 50, or provide equivalent
protection.Section III.G.2 of 10 CFR Part 50, Appendix R, describes three acceptable
methods for protecting at least one safe shutdown train when redundant trains are
located in the same fire area. The Section III.G.2 requirements are based on the
combination of physical barriers, spacial separation, fire detection and automatic
suppression systems.
SNUPPS FSAR Appendix 9.5E provided the design comparison between the plants fire
protection program and 10 CFR Part 50, Appendix R. The comparison to Section III.G,
Fire Protection of Safe Shutdown Capability, states, Redundant trains of systems
required to achieve and maintain hot standby are separated by 3-hour-rated fire
barriers, or the equivalent provided by III.G.2, or else a diverse means of providing the
safe shutdown capability exists that is unaffected by the fire. Wolf Creek has
interpreted diverse means to mean by any reasonable means including local valve and
breaker operations as long as they are within the scope of normal operator duties. The
team disagrees with this interpretation. The NRC staff does not recognize the use of
manual actions as meeting the technical requirements of Appendix R. The components
being operated are identified as required for operation of safe shutdown systems or are
subject to potential spurious operation impacting the shutdown. The local manual
actions are being performed because of fire damage to electrical cables related to those
components and are meant to compensate for damage or maloperation of safe
shutdown equipment caused by fire. Manual actions are not a method of satisfying
Appendix R,Section III.G.2 requirements. Plant specific manual actions may be
acceptable based on detailed specific exemptions or deviations for each case identified.
Analysis. This finding is of greater than minor safety significance because it impacted
the mitigating systems cornerstone objective to ensure the availability, reliability, and
capability of systems that respond to external events (such as fire) to prevent
undesirable consequences. The team reviewed Procedure OFN KC-016, Fire
Response, and stepped through the manual actions directed in the procedure with
licensee operations personnel. The team found that the manual operator actions were
reasonable (as defined in Enclosure 2 of Inspection Procedure 71111.05T), and could
be performed within the analyzed time limits. Since the manual operator actions were
considered reasonable, the significance determination process was not entered. The
team determined that this finding is of very low safety significance (Green) in
accordance with the guidance in Enclosure 2 to Inspection Procedure 71111.05T.
Enforcement. The Fire Hazard Analysis states that it will comply with the technical
requirements of Appendix R or utilize a diverse means to do so. Appendix R,
Section III.G.2 to 10 CFR Part 50 requires that cables whose fire damage could prevent
the operation or cause maloperation of safe shutdown functions be physically protected
Enclosure
-8-
from fire damage. Contrary to this requirement, the licensee implemented a
methodology that utilized manual operator actions as a diverse means to mitigate the
effects of fire damage in lieu of providing physical protection from fire damage. This is a
violation of License Condition 2.C.(5)(a) for failing to meet the technical requirements of
10 CFR Part 50, Appendix R, as required by SER Section 9.5.1.7. Because this finding
is of very low safety significance, this violation is being treated as a noncited violation,
consistent with Section VI.A of the NRC Enforcement Policy:
NCV 05000482/2005008-03, Failure to Ensure Redundant Safe Shutdown Systems
Located In the Same Fire Area Are Free of Fire Damage.
.3 Passive Fire Protection
a. Inspection Scope
For the selected fire areas, the team evaluated the adequacy of fire area barriers,
penetration seals, fire doors, electrical raceway fire barriers and fire rated electrical
cables. The team observed the material condition and configuration of the installed
barriers, seals, doors, and cables. The team compared the as-installed configurations
to the approved construction details and supporting fire tests. In addition, the team
reviewed license documentation, such as NRC safety evaluation reports, and deviations
from NRC regulations and the National Fire Protection Association code to verify that
fire protection features met license commitments.
b. Findings
No findings of significance were identified.
.4 Active Fire Protection
a. Inspection Scope
For the selected fire areas, the team evaluated the adequacy of fire suppression and
detection systems. The team observed the material condition and configuration of the
installed fire detection and suppression systems. The team reviewed design documents
and supporting calculations. In addition, the team reviewed license basis
documentation, such as NRC safety evaluation reports, and deviations from NRC
regulations and the National Fire Protection Association codes to verify that fire
suppression and detection systems met license commitments.
The team also observed an announced site fire brigade drill and the subsequent drill
critique using the guidance in Inspection Procedure 71111.05AQ. Team members
observed the fire brigade simulate fire fighting activities in plant Fire Area T-4 (Lube Oil
Storage Room). The inspectors verified that the licensee staff identified deficiencies,
openly discussed them in a self-critical manner at the drill debrief, and took appropriate
corrective actions. Specific attributes evaluated were: (1) proper wearing of turnout
gear and self-contained breathing apparatus; (2) proper use and layout of fire hoses; (3)
employment of appropriate fire fighting techniques; (4) sufficient fire fighting equipment
Enclosure
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brought to the scene; (5) effectiveness of fire brigade leader communications,
command, and control; (6) search for victims and propagation of the fire into other plant
areas; (7) smoke removal operations; (8) utilization of pre-planned strategies; (9)
adherence to the pre-planned drill scenario; and (10) drill objectives.
b. Findings
No findings of significance were identified.
.5 Protection From Damage From Fire Suppression Activities
a. Inspection Scope
For the sample areas, the team verified that redundant trains of systems required for hot
shutdown were not subject to damage from fire suppression activities or from the
rupture or inadvertent operation of fire suppression systems including the effects of
flooding.
b. Findings
No findings of significance were identified.
.6 Alternative Shutdown Capability
a. Inspection Scope
The team reviewed the alternative shutdown methodology to determine if the licensee
properly identified the components, systems, and instrumentation necessary to achieve
and maintain safe shutdown conditions from the auxiliary shutdown panel and
alternative shutdown locations. The team focused on the adequacy of the systems
selected for reactivity control, reactor coolant makeup, reactor heat removal, process
monitoring and support system functions. The team verified that hot and cold shutdown
from outside the control room could be achieved and maintained with offsite power
available or not available. The team verified that the transfer of control from the control
room to the alternative locations was not affected by fire induced circuit faults by
reviewing the provision of separate fuses for alternative shutdown control circuits.
The team also reviewed the operational implementation of the alternative shutdown
methodology. Team members observed a walk-through of the control room evacuation
procedures with that days watchstanders consisting of both licensed reactor and senior
reactor operators. The team observed operators simulate performing the steps of
Procedure OFN RP-017 that provided instructions for performing an alternative
shutdown from the auxiliary shutdown panel and for manipulating equipment in the
plant. The team verified that the minimum number of available operators, exclusive of
those required for the fire brigade, could reasonably be expected to perform the
procedural actions within the applicable plant shutdown time requirements and that
equipment labeling was consistent with the procedure. Also, the team verified that
Enclosure
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procedures, tools, dosimetry, keys, lighting, and communications equipment were
available and adequate to support successfully performing the procedure as intended.
The team also reviewed records for operator training conducted on this procedure.
b. Findings
(1) Lack of Evaluations of Changes to The Approved Fire Protection Program
Introduction. The team identified an unresolved item related to unanalyzed changes to
approved Wolf Creek Generating Station fire protection program. Specifically, the team
identified that the licensee had revised Procedure OFN RP-017 without documentation
demonstrating that the changes would not adversely affect the ability to achieve and
maintain safe shutdown in the event of a fire. This will be treated as an unresolved item
pending further evaluation by the license. NRC inspection of the results of the licenses
evaluations and determination of safety significance.
Description. In Letter SLNRC 84-0109, the licensee made time commitments for
specific items required to achieve and maintain hot shutdown conditions from outside
the control room that would be completed in six phases. Phase A items would be
completed in 5 minutes. Phase B items would be completed in 10 minutes. Phase C
items would be completed in 20 minutes. Phase D items would be completed in
30 minutes. Phase E items would be completed in 60 minutes. Phase F items would be
completed in 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. These phased time commitments were approved by the NRC staff
in SER Supplement 5.
Future revisions to OFN RP-017 consolidated the approved number of phases from six
to four. Phases B and C were consolidated into a new Phase B with an item completion
time of 20 minutes. Phases D and E were consolidated into a new Phase C with an item
completion time of 60 minutes. Review of the procedure revisions identified changes
that resulted in actions having allowable completion times longer that the approved time
commitments per SLNRC 84-0109. The changes of concern allowed:
a. An item with a 5 minute commitment per Letter SLNRC 84-0109 to become a
10 minute action. The step to verify EDG running (Step C10) was initially a
Phase A item, which per Letter SLNRC 84-0109, allowed 5 minutes for
completion. Step C10 is now a Phase B item, which per the current revision of
the procedure, allows 20 minutes for completion. The actual step was performed
in 7 minutes and 25 seconds when the response not obtained column was
invoked.
b. Six items that were initially Phase B items, which per Letter SLNRC 84-0109,
allowed 10 minutes for completion, are now allowed longer completion times.
Steps B10, C18, C21, and C22 are all currently Phase B items, which per the
current revision of the procedure, allows 20 minutes for completion. Timed
walkthroughs of the procedure confirmed that completion of these steps would
require more than 10 minutes. Step B10 to isolate RHR Pump A was completed
at time 10:45. Step C18 to ensure room cooling for EDG room was completed at
Enclosure
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time 11:18. Step C21 to ensure room cooling for ESW room was completed at
time 12:24. Step C22 to isolate B RHR pump was completed at time 12:40.
Steps C30 and D10 are currently Phase C items, which per the current revision
of the procedure, allows 60 minutes for completion. Step C30 to ensure A
containment spray pump stopped was completed at time 18:46. Step D10 to
ensure room cooling for the electrical penetration room was completed at
time 22:15.
Analysis. This finding is unresolved pending the completion of further inspection and
completion of a significance determination. The license must complete a records search
for any documentation evaluating the changes to Procedure OFN RP-017 described
above. The license must perform evaluations for changes where no previous
evaluations can be identified. The NRC will review the results of the licenses efforts.
This finding is of greater than minor safety significance because it impacted the
mitigating systems cornerstone objective to ensure the availability, reliability, and
capability of systems that respond to external events (such as fire) to prevent
undesirable consequences. Specifically, the license did not evaluate all changes to the
approved fire protection program to assure that the changes would not adversely affect
the ability to achieve and maintain safe shutdown in the event of a fire.
Enforcement. License Condition 2.C(5)(b) states, The licensee may make changes to
the approved fire protection program without prior approval of the Commission only if
those changes would not adversely affect the ability to achieve and maintain safe
shutdown in the event of a fire. However, the team could not identify evaluations
showing that changes to OFN RP-017 would not adversely affect the ability to achieve
and maintain safe shutdown in the event of a fire. Pending completion of further
inspection of the impact of these changes and a significance determination, this finding
is identified as URI 05000482/2005008-04, Lack of Evaluations of Changes to The
Approved Fire Protection Program.
(2) Inadequate Alternative Shutdown Procedure
Introduction. The team identified an Apparent Violation of Technical Specification 5.4,
Procedures, because of an inadequate alternative shutdown procedure which is
required for implementation of the Fire Protection Program. The team found that some
time critical actions required to safely shutdown the plant following a control room fire
could not be accomplished within the planned time periods.
Description. Wolf Creek utilized Procedure OFN RP-017 to satisfy the fire protection
program requirement to be able to achieve and maintain hot standby in the case of a
control room fire. During the procedure, the operators must respond to a loss of reactor
coolant pump seal injection, and a loss of component cooling water thermal barrier
cooling.
The Westinghouse Owners Group released the Assessment of RCP Operation During
Loss of Seal Cooling for members in February 2000. The assessment states that if
reactor coolant pump seal injection is lost and then restored, it should be restored in a
Enclosure
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short period of time. If seal injection is restored after the seals have heated, there is a
possibility that the seals will leak reactor coolant excessively. Also, the letter states a
concern that when flow is stopped to the component cooling water thermal barrier in the
reactor coolant pump, that voiding may occur in the component cooling water system,
and if flow is re-established, then it could cause a water hammer leading to system
damage.
The licensee timed a practice run of the control room evacuation and concluded that
they met the recommendations by Westinghouse for assuring reactor coolant pump seal
reliability and avoiding component cooling water thermal barrier water hammer
concerns. However, the team found that the methodology assumed only one spurious
operation from the fire during the scenario. This method minimized the number of
spurious operations the operators had to respond to and correspondingly minimized the
procedure completion time.
The team performed an independent timed walkthrough of the control room evacuation
procedure during the inspection. The team asked the operators to mitigate almost all of
the spurious operations that might be caused by the fire, including manually opening
motor operated valves and starting the emergency diesel generator. This lengthened
the operators response times significantly, such that the Westinghouse
recommendations were no longer being met for the steps in the procedure addressing
the reactor coolant pump seals and the thermal barrier.
Analysis. The inspectors referred to MC 0612 and determined that the finding is greater
than minor in that it affected the ability to achieve and maintain hot shutdown following a
control room fire. This finding is associated with the Mitigating Systems cornerstone
and the respective attribute of protection against external factors (e.g., fire). This finding
impacted the mitigating systems cornerstone objective to ensure the availability,
reliability, and capability of systems that respond to external events (such as fire) to
prevent undesirable consequences.
The licensee recognized that the assumption of multiple spurious actuations would
affect the validity of their previous timing results. However, the licensees position is that
their licensing basis only requires one spurious operation to be assumed during a
control room fire. However, the licensee did initiate compensatory measures consisting
of stationing additional fire watch personnel in the control room to increase surveillance
for potential fire hazards and fires in the incipient stage. The team did not enter the
Significance Determination Process at this time because the enforcement is being
deferred as discussed below and the licensee has established adequate compensatory
measures. Therefore, the significance will be determined after the NRC endorses a
path to resolution for fire induced circuit failures.
Enforcement. Technical Specification 5.4.1 states, in part, Written Procedures shall be
established, implemented, and maintained covering the following activities:.... d. Fire
Protection Program implementation. License Condition 2.C.(5)(a) states The
Operating Corporation shall maintain in effect all provisions of the approved fire
protection program as described in the SNUPPS Final Safety Analysis Report for the
Enclosure
-13-
facility through Revision 17, the Wolf Creek site addendum through Revision 15, and as
approved in the SER through Supplement 5, subject to provisions b & c below. Safety
Evaluation Report, Section 9.5.1.7, Appendix R Statement, states "The staff will
condition the operating license to require the applicant to meet the technical
requirements fo Appendix R to 10 CFR Part 50, or provide equivalent protection.
Appendix R,Section III.L.7, states The safe shutdown equipment and systems for each
fire area shall be known to be isolated from associated non-safety circuits in the fire
area so that hot shorts, open circuits, or shorts to ground in the associated circuits will
not prevent operation of the safe shutdown equipment. The separation and barriers
between trays and conduits containing associated circuits of one safe shutdown division
and trays and conduits containing associated circuits or safe shutdown cables from the
redundant division, or the isolation of these associated circuits from the safe shutdown
equipment, shall be such that a postulated fire involving associated circuits will not
prevent safe shutdown.
Contrary to the above, the licensee could not perform some time critical actions required
for safe shutdown following a control room fire within the required time periods using
Procedure OFN RP-017. The licensee considers the spurious operation of multiple
components to be outside of the plant licensing basis for the Fire Protection Program.
The licensees position is that the original procedure timing method with one spurious
operation is valid and the teams assumption of multiple spurious operations is overly
conservative and an increase in regulatory requirements. The NRC staff and the
industry are currently working on developing a resolution methodology to address these
types of potential fire induced circuit failures. The teams review concluded that this
violation met the criteria of the NRC Enforcement Manual Section 8.1.7.1 for deferring
enforcement actions for postulated fire induced circuit failures. This violation is being
treated as an apparent violation: AV 05000482/2005008-05, Inadequate Alternative
Shutdown Procedure.
.7 Circuit Analyses
a. Inspection Scope
The team reviewed the post-fire safe shutdown analysis to verify that the licensee had
identified circuits that may impact safe shutdown. On a sample basis, the team verified
those cables for equipment required to achieve and maintain hot shutdown conditions in
the event of fire in selected fire zones had been properly identified. The evaluation
focused on the cabling of selected components for the chemical and volume control
system, high pressure safety injection system, and the auxiliary feedwater system.
Included in this evaluation were a sample of components whose inadvertent operation
could significantly affect the shutdown capability credited in the safe shutdown analysis.
In addition, the team verified that these cables had either been adequately protected
from the potentially adverse effects of fire damage, mitigated with approved manual
operator actions, or analyzed to show that fire induced faults (e.g., hot shorts, open
circuits, and shorts to ground) would not prevent safe shutdown. In order to accomplish
this, the team reviewed electrical schematics and cable routing data for power and
control cables associated with each of the selected components.
Enclosure
-14-
In addition, the team verified, on a sample basis, that circuit breaker coordination and
fuse protection have been analyzed, and are acceptable as means of protecting the
power source of the designated redundant or alternative safe shutdown component.
For the selected fire areas, the team also reviewed the location and installation of
diagnostic instrumentation that was necessary for achieving and maintaining safe
shutdown conditions to ensure that in the event of a fire, this instrumentation would
remain functional.
b. Findings
No findings of significance were identified.
.8 Communications
a. Inspection Scope
The team reviewed the adequacy of the communication system to support plant
personnel in the performance of alternative safe shutdown functions and fire brigade
duties. The team verified that phones were available for use and maintained in working
order. The team reviewed that the electrical power supplies and cable routing for the
phone system would allow them to remain functional following a fire in the control room
fire area.
b. Findings
No findings of significance were identified.
a. Inspection Scope
The team reviewed the emergency lighting system required to support plant personnel
in the performance of alternative safe shutdown functions to verify it was adequate to
support the performance of manual actions required to achieve and maintain hot
shutdown conditions, and for illuminating access and egress routes to the areas where
manual actions are required. The locations and positioning of emergency lights were
observed during a walkthrough of the control room evacuation procedure.
b. Findings
No findings of significance were identified.
Enclosure
-15-
.10 Cold Shutdown Repairs
a. Inspection Scope
The team reviewed Procedure OFN RP-014 to determine whether repairs were required
to achieve cold shutdown. The team also verified that the repair material was available
on the site.
b. Findings
No findings of significance were identified.
.11 Compensatory Measures
a. Inspection Scope
The team reviewed the program with respect to compensatory measures in place for
out-of-service, degraded, or inoperable fire protection and post-fire safe shutdown
equipment, systems or features.
The team reviewed AP 10-103, Fire Protection Impairment Control, Revision 19 to
determine whether the procedures adequately controlled compensatory measures for
fire protection systems, equipment and features (e.g., detection and suppression
systems and equipment, and passive fire barriers). The team also walked down
compensatory measures in effect at the time of the inspection.
b. Findings
No findings of significance were identified.
4OA2 Problem Identification and Resolution
a. Inspection Scope
The team reviewed a sample of Problem Identification Reports to verify that the licensee
was identifying fire protection-related issues at an appropriate threshold and entering
those issues into the corrective action program. A listing of Problem Identification
Reports reviewed is provided in the attachment to this report.
b. Findings
Introduction. The team identified an unresolved item related to the evaluation of
conditions adverse to fire protection, which is a provision of the Wolf Creek Generating
Station fire protection program. This will be treated as an unresolved item pending
further inspection of the extent of condition and determination of safety significance.
Enclosure
-16-
Description. The NRC issued Information Notice 92-18, Potential for Loss of Remote
Shutdown Capability During a Control Room Fire, on February 28, 1992, to all holders
of operating licenses. This notice was issued to alert licensees to conditions found at
several reactors that could result in the loss of capability to maintain the reactor in a safe
shutdown condition because of a control room fire that caused operators to evacuate
the control room. A fire in the control room could cause hot short circuits between
control wiring and power sources, for certain motor-operated valves needed for safe
shutdown. If a fire in the control room forces operators to leave the control room, these
motor-operated valves can be operated from the remote/alternative shutdown panel.
However, hot short circuits combined with the absence of thermal overload, torque
switch and limit switch protection, could cause valve damage before the operator shifted
control of the valves to the remote/alternative shutdown panel.
The licensee evaluated Information Notice 92-18 via Industry Technical Information
Program (ITIP)1906 on April 15, 1992, and determined that the notice was not
applicable to Wolf Creek. The disposition and closure of ITIP 1906 relied upon
evaluations performed during initial licensing as discussed in documents from 1984 and
1985. The documents referenced in the ITIP are Letter SLNRC 84-0108,
dated August 24 1985; Letter SLNRC 84-0109, dated August 10, 1984; and Safety
Evaluation Report, NUREG 0881, Supplement 5. Based upon the NRCs acceptance of
the response plan to spurious actuations resulting from control room fires, as discussed
in the referenced documents, the licensee deemed the information contained in
Information Notice 92-18 as having previously been evaluated.
The licensee subsequently reevaluated their position in regard to Information Notice 92-18 in 1999 based upon questions raised by the NRC during an inspection at
the Callaway Plant. The licensee initiated Performance Improvement Request 99-1245
on April 4, 1999, to validate their position as described in ITIP 1906. The performance
improvement request stated that engineering had compiled a list of motor-operated
valves which are susceptible to inadvertent failure because of a control room fire, and
could potentially jeopardize plant safe shutdown. It also stated that further evaluation
and investigation was being done to narrow down the list of valves requiring
modifications. Performance Improvement Request 99-1245 was closed based on an
NRC/industry initiative in place at the time to address dealing with multiple hot shorts in
associated circuits resulting in spurious actuations. The NRC temporarily suspended
the associated circuit portion of the triennial fire protection inspection in November 2000,
but restarted the inspections in January 2005.
At the time of the inspection, the licensee had not determined which motor-operated
valves could be susceptible to mechanistic damage because of having the torque and
limit switches, and the thermal overloads bypassed because of fire induced short
circuits. The inspectors reviewed a sample of valves and determined that they could
have their protection bypassed. Four motor operated valves was selected from control
room evacuation Procedure OFN RP-017 for review of Information Notice 92-18
applicability. The four valves, BN-LCV112E, EM-HV8803B, EM-HV8801A, and
BN-HV8812A, were all found to be susceptible to having their torque and limit switch
protection bypassed as a result of a control room fire. All four valves were also required
Enclosure
-17-
by Procedure OFN RP-017 to be positioned after a control room fire. However, the
inspectors could not determine whether damage could occur to the valves rendering
them inoperable.
Analysis. This finding is unresolved pending the completion of further inspection of the
extent of condition and completion of a significance determination. The licensee must
evaluate the motor operated valves relied upon during a post-fire shutdown outside of
the control room. The licensee must review control circuits to identify any valves which
could spuriously operate because of fire damage with the normal protective devices
bypassed. The licensee must determine if any such valves would be susceptible to
damage which would prevent the planned electrical or manual operation of the valve
during the shutdown from outside of the control room. This finding is of greater than
minor safety significance because it impacted the mitigating systems cornerstone
objective to ensure the availability, reliability, and capability of systems that respond to
external events (such as fire) to prevent undesirable consequences. Specifically, the
licensee did not perform a timely or technically adequate evaluation to determine if the
Wolf Creek configurations were subject to the potential loss of capability to maintain the
reactor in a safe shutdown condition following a control room fire described in NRC
Enforcement. License Condition 2.c(5) of the Wolf Creek Generating Station Operating
License states that the Operating Corporation shall maintain in effect all provisions of
the approved fire protection program as described in the SNUPPS Final Safety Analysis
Report. The Wolf Creek Generating Station Updated Safety Analysis Report,
Appendix 9.5A, Section 8, states that failures, malfunctions, deficiencies, deviations,
defective components, uncontrolled combustible material and nonconformances which
affect fire protection are promptly identified, reported, evaluated and corrected.
However, the team found that the licensee failed to evaluate the potential for fire
induced damage to motor operated valves relied upon for safe shutdown following a
control room evacuation as described in NRC Information Notice 92-18. The licensee
entered this finding in their corrective action program as Performance Improvement
Request 2005-3314. Pending completion of further inspection for extent of condition
and a significance determination, this finding is identified as URI 05000482/2005008-06,
Failure to Adequately Evaluate Fire Protection Program Deficiencies
4OA6 Management Meetings
Debrief Meeting Summary
The team leader presented the inspection results to Mr. Rick A. Muench, President and
Chief Executive Officer, and other members of licensee management at the conclusion
of the onsite inspection on December 2, 2005.
During this meeting, the team leader confirmed to the licensee management that
materials considered to be proprietary had been examined during the inspection and
had been returned to the licensee.
Enclosure
-18-
Exit Meeting Summary
The team leader presented the inspection results to members of licensee management
at the conclusion of the inspection in a conference call on December 29, 2005.
Enclosure
KEY POINTS OF CONTACT
Licensee
T. M. Anselmi, Manager Design Engineering
W. Aregood, Fire Protection
R. Badenhamer, Operations
T. Card, Supervisor Support Engineering
D. Dixon, Design Engineering - Electrical
R. D. Flannigan, Manager Nuclear Engineering
K. Fredrickson, Regulatory Affairs
S. Hedges, VP Operations & Plant Manager
S. A. Henry, Superintend of Operations
P. Herrmann, Fire Protection
D. M. Hooper, Regulatory Affairs
W. Ketchum, Probabilistic Risk Analysis
T. Krause, Manager Quality
J. B. Makar, Manager Systems Engineering
K. J. Moles, Manager Regulatory Affairs
R. A. Muench, President & CEO
W. Muilenburg, Regulatory Affairs
G. L. Pendergrass, Manager Support
D. Phelps, Owner Company Representative
L. Ratzlaff, Fire Protection
E. A. Ray, Manager Operations
W. Selbe, Design Engineering
M.W.Sunseri, VP Oversite
J. Suter, Fire Protection
W. Wagner, Safety Analysis
NRC
S. Cochrum, Senior Resident Inspector
A-1 Attachment
ITEMS OPENED AND CLOSED
Opened
05000482/2005008-02 AV Failure to Maintain Reactor Coolant System
Subcooling During the Alternative Shutdown
(Section 1R05.1.b(2))05000482/2005008-04 URI Lack of Evaluations of Changes to The Approved Fire
Protection Program (Section 1R05.6.b(1))05000482/2005008-05 AV Inadequate Alternative Shutdown Procedure
(Section 1R05.6.b(2))05000482/2005008-06 URI Failure to Adequately Evaluate Fire Protection
Program Deficiencies (Section 4OA2)
Opened and Closed
05000482/2005008-01 NCV Failure to Provide Adequate Post-Fire Shutdown
Procedures (Section 1R05.1.b(1))05000482/2005008-03 NCV Failure to Ensure Redundant Safe Shutdown Systems
Located In the Same Fire Area Are Free of Fire
Damage (Section 1R05.2)
Closed
None
Discussed
None
A-2 Attachment
LIST OF DOCUMENTS REVIEWED
The following documents were selected and reviewed by the team to accomplish the objectives
and scope of the inspection.
COMPONENTS SELECTED FOR REVIEW
Component Description
ALHV0030 Auxiliary Feedwater Pump Suction Isolation Valves
ALHV0031
ALHV0032
ALHV0033
ALHV0034
ALHV0035
ALHV0036
DPAL01A Auxiliary Feedwater Pump A
DPAL01B Auxiliary Feedwater Pump B
BGLCV112B Volume Control Tank Outlet Valves
BGLCV112C
BGHV8110 Centrifugal Charging Pump A Mini-Flow Isolation Valve
BGHV8111 Centrifugal Charging Pump B Mini-Flow Isolation Valve
BNHV8812A Refueling Water Storage Tank To Residual Heat Removal Suction
BNHV8812B Isolation Valves
DPBG05A Centrifugal Charging Pump A
DPBG05B Centrifugal Charging Pump B
DPEF01A Essential Service Water Pump A
DPEF01B Essential Service Water Pump B
EFHV0023 Service Water To Essential Service Water Loop Isolation Valves
EFHV0024
EFHV0025
EFHV0026
EGHV0058 Component Cooling Water To Reactor Coolant Pump Isolation Valves
EGHV0071
EGHV0126
EGHV0127
EJHV8701A Residual Heat Removal Suction Isolation Valves
EJHV8701B
A-3 Attachment
EJH8811A Containment Sump Isolation Valves
EJHV8811B
CALCULATIONS
Number Title Revision
AN-02-021 OFN RP-017 Control Room Evacuation Consequence 0
Evaluation
E-H-8 System NB Protective Relays 5
FL-03 Flooding of Individual Aux Bldg Rooms 0
FL-08 Control Building Flooding 0
LE-M-004 Flooding In Class 1E Switchgear Rooms 3301 & 3302 00
and Battery Room # 2 (3411) & Battery Room # 3
(3413)
XX-E-013 Post-Fire Safe Shutdown (PFSSD) Analysis 0
DRAWINGS
Number Title Revision
E-1F9910 Post-Fire Safe Shutdown Fire Area Analysis 0
E-1R1441(Q) Raceway Plan - Auxiliary Building Area-4 6
EL. 2026'-0"
E-1R1443A Exposed Conduit - Auxiliary Building Area-4 7
EL. 2026'-6"
E-1R1443B Exposed Conduit - Auxiliary Building Area-4 11
EL. 2026'-0"
E-1R1443C Exposed Conduit - Auxiliary Building Area-4 9
EL. 2026'-0"
E-1R1444A Exposed Conduit - Auxiliary Building Partial Plan 4
Area-4 EL. 2026'-0"
E-1R1444B Exposed Conduit - Auxiliary Building Partial Plan 7
Area-4 EL. 2026'-0"
E-1R1444C Exposed Conduit - Auxiliary Building Partial Plan 12
Area-4 EL. 2026'-0"
E-11NG01 Low Voltage System Class IE 480 V. Single Line 9
Meter & Relay Diagram
A-4 Attachment
Number Title Revision
E-11NG02 Low Voltage System Class IE 480 V. Single Line 8
Meter & Relay Diagram
E-11NG20 Motor Control Center Summary 234
E-11NK01 Class IE 125V DC System Meter & Relay Diagram 9
E-11NK02 Class IE 125V DC System Meter & Relay Diagram 7
E-13AB01 Schematic Diagram - Main Steam Supply Valve To 2
Turbine Driven Aux Feedwater Pump
E-13AB18 Schematic Diagram - Main Steam High Pressure 0
Trap Bypass Valves
E-13AL03A Schematic Diagram - Auxiliary Feedwater Pumps, 4
Discharge Control - Motor Operated Valves
E-13AL04B Schematic Diagram - Supply From ESS Service 8
Water System
E-13AL05A Schematic Diagram - Auxiliary Feedwater Pumps, 2
Discharge Control - Air Operated Valves
E-13BB04 Schematic Diagram - Seal Water Injection Isolation 3
Valves
E-13BB12A Schematic Diagram - RHR Loop 1 Inlet Isolation 6
Valve
E-13BB12B Schematic Diagram - RHR Loop 2 Inlet Isolation 4
Valve
E-13BB30 Schematic Diagram - RCS Head Vent Valves 2
E-13BB39 Schematic Diagram - Pressurizer Relief Isolation 8
Valves
E-13BB40 Schematic Diagram - Pressurizer Power Relief 3
Valves
E-13BG01 Schematic Diagram - Centrifugal Charging Pump A 3
E-13BG01A Schematic Diagram - Centrifugal Charging Pump B 1
E-13BG10 Schematic Diagram - Letdown Line Isolation Valves 3
E-13BG12 Schematic Diagram - Volume Control Tank Outlet 3
Isolation Valve
E-13BG12A Schematic Diagram - Volume Control Tank Outlet 4
Isolation Valve
A-5 Attachment
Number Title Revision
E-13BG48 Schematic Diagram - Excess Letdown Line Isolation 1
Valves
E-13BN01 Schematic Diagram - Refueling Water Storage Tank 3
To Charging Pump MOV
E-13BN03 Schematic Diagram - Refueling Water Storage Tank 7
E-13EG09 Schematic Diagram - Component Cooling Water 4
Containment Isolation Valve
E-13EG18 Schematic Diagram - Component Cooling Water 7
Containment Isolation Valves
E-13EJ05A Schematic Diagram - RHR Loop 1 Inlet isolation 4
Valve
E-13EJ06A Schematic Diagram - Sump To No. 1 Residual Heat 6
Removal Pump
E-13EJ06B Schematic Diagram - Sump To No. 2Residual Heat 7
Removal Pump
KD-7496 One Line Diagram 27
M-12AB01 P&ID - Main Steam System 10
M-12AB02 P&ID - Main Steam System 9
M-12AB03 P&ID - Main Steam System 18
M-12AL01 P&ID - Auxiliary Feedwater System 10
M-12BB01 P&ID - Reactor Coolant System 24
M-12BB02 P&ID - Reactor Coolant System 14
M-12BB03 P&ID - Reactor Coolant System 9
M-12BB04 P&ID - Reactor Coolant System 10
M-12BG01 P&ID - Chemical and Volume Control System 12
M-12BG03 P&ID - Chemical & Volume Control System 36
M-12BN01 P&ID - Borated Refueling Water Storage System 12
M-12EF01 P&ID - Essential Service Water System 19
M-12EF02 P&ID - Essential Service Water System 22
M-12EG01 P&ID - Component Cooling Water System 14
A-6 Attachment
Number Title Revision
M-12EG02 P&ID - Component Cooling Water System 17
M-12EG03 P&ID - Component Cooling Water System 8
M-12EJ01 P&ID - Residual Heat Removal System 31
M-K2EF01 P&ID - Essential Service Water System 48
PERFORMANCE IMPROVEMENT REQUESTS (PIRs)
99-1245 20010046 20053025* 20053176* 20053314* 20053331*
20003699 20010210 20053033* 20053209* 20053317* 20053333*
20010045 20052757 20053054* 20053305* 20053319*
- PIR written as a result of inspection activities
PROCEDURES
Number Title Revision
AP 10-100 Fire Protection Program 9
AP 10-103 Fire Protection Impairment Control 19
AP 10-105 Fire Protection Training and Drills 9
AP 21-003 Operations 7A
OFN KC-016 Fire Response 13
OFN KJ-032 Local Emergency Diesel Startup 6
OFN RP-013 Control Room Not Habitable 10A
OFN RP-014 Hot standby to Cold Shutdown From Outside the 8
Control Room
OFN RP-017 Control Room Evacuation 21
STN GP-009 Emergency Radio and Equipment Check and Inventory 41
STN FP-206 Spray and Sprinkler System Functional Testing 9
STN FP-207 Visual Inspection of Pipe Headers and Nozzle/Sprinkler 2
Areas
STN FP-400B Halon Sys/North Pene Rm (KC-244) 5
STN FP-452 Fire Barrier Penetration Seals Inspection 4
A-7 Attachment
STN FP-817F Trip Act. Device Oper. Test for Bechtel Zones 306, 307 6
and 314-317
MISCELLANEOUS DOCUMENTS
Number Title Revision
AP 10-106 Fire Preplans 4
APF 10-105-02 Fire Drill Scenario and Critique Report 1
E-1F9905 Fire Hazards Analysis 0
E-1F9910 Post-Fire Safe Shutdown Area Analysis 0
ITIP No. 01906 Industry Technical Information Program Report - 4/15/92
NRC Information Notice 92-18: Potential For Loss Of
Remote Shutdown Capability During A Control Room
Fire
LER 42146 Potential Failure to Meet Required Response Times 11/16/05
For Shutdown Outside Control Room
License No. NPF-42 Facility Operating License, Wolf Creek Generating Amendment
Station, Unit No. 1 No. 151
M-663-00017 Penetration Seal Typical Details W20
M-663-00017A Fire Protection Evaluations For Unique or Unbounded W01
Fire Barrier Configurations
Self Assessment NFPA Code Compliance 0
SEL 01-027
SLNRC 84-0109 SNUPPS Letter to H. R. Denton From N. A. Petrick - 8/23/1984
Subject: Fire Protection Review
Specification No. Technical Specification For Contract For Furnishing, 7
16577-M-658 Installing, and Testing Halogenated Agent
Extinguishing System for The Standardized Nuclear
Unit Power Plant System (SNUPPS) Wolf Creek Only
NUREG 0881, Safety Evaluation Report Related to the Operation of April 1982
Volume 1 Wolf Creek Generating Station Unit No. 1
NUREG 0881, Safety Evaluation Report Related to the Operation of August 1983
Supplement No. 3 Wolf Creek Generating Station Unit No. 1
NUREG 0881, Safety Evaluation Report Related to the Operation of March 1985
Supplement No. 5 Wolf Creek Generating Station Unit No. 1
PIR 1998-0600 NFPA Code Deficiency Tracking Sheet 09/21-2005
A-8 Attachment
USAR - 7.4 Updated Safety Analysis Report - Section 7.4 - 16
Systems Required For Safe Shutdown
USAR - 9.5.1 Updated Safety Analysis Report - Section 9.5.1 - Fire 16
Protection System
USAR - 15.2.6 Updated Safety Analysis Report - Section 15.2.6 - 16
Loss of Non-Emergency AC Power to the Station
Auxiliaries (Blackout)
WCNOC-76 Design Guide for Medium and Low Voltage AC and 2
Low Voltage DC Overcurrent Protection Coordination
for Wolf Creek Generating Station
Cable Routing Data for Various Components and Fire
Areas
WCGS Approved Fuse List 7
Wolf Creek Fire Protection Program Regulatory 1
Bases
Time - Current Curves for Various 480Vac and
125Vdc Components
MODIFICATIONS
Number Title Revision
DCP 011038 Install Fire Wrap on Raceway in Fire Areas A-1 & A-18 4
WORK ORDERS
04-258679-000 04-258728-000 04-263755-000 05-270020-000
A-9 Attachment