ML060330616

IR 05000482-05-008 on 10/24/2005 - 12/29/2005 for Wolf Creek Nuclear Operating Corporation; Wolf Creek Generating Station; Fire Protection (Triennial)

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

05000482/2005008

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)

DRS STA (DAP)

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

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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

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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

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(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

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.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

-7-

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

-9-

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

-10-

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

-11-

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

-12-

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.

.9 Emergency Lighting

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

Information Notice 92-18.

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

To RHR Pump MOV

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