IR 05000298/2007008

IR 05000298-07-008; Nebraska Public Power District; on 05/21/07 - 12/26/07; Cooper Nuclear Station: Triennial Fire Protection Inspection

ML080350425
Person / Time
Site:	Cooper
Issue date:	02/01/2008
From:	Smith L J NRC/RGN-IV/DRS/EB2
To:	Minahan S B Nebraska Public Power District (NPPD)
References
EA-07-204 IR-07-008
Download: ML080350425 (33)

Text

February 1, 2008

EA-07-204

Stewart B. Minahan, Vice President-Nuclear and CNO Nebraska Public Power District P.O. Box 98 Brownville, NE 68321

SUBJECT: COOPER NUCLEAR STATION - NRC TRIENNIAL FIRE PROTECTION INSPECTION REPORT 05000298/2007008 AND EXERCISE OF ENFORCEMENT DISCRETION

Dear Mr. Minahan:

On June 15, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Cooper Nuclear Station. The enclosed report documents the inspection findings, which were discussed in a debrief meeting at the end of the onsite inspection on June 15, 2007, with Mr. M. Colomb, General Manager of Plant Operations, and other members of your staff and again in an exit meeting conducted via conference call on December 26, 2007, with you and other members of your staff. During this triennial fire protection inspection, the inspection team examined activities conducted under your license related to safety and compliance with the Commission's 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. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/NFO'Keefe for

Linda J. Smith, Chief Engineering Branch 2 Division of Reactor Safety Nebraska Public Power District - 2 -

Docket: 50-298 License: DPR-46

Enclosure:

NRC Inspection Report 05000298/2007008

w/attachment:

Supplemental Information cc w/

Enclosure:

Gene Mace Nuclear Asset Manager Nebraska Public Power District P.O. Box 98 Brownville, NE 68321 John C. McClure, Vice President and General Counsel Nebraska Public Power District P.O. Box 499 Columbus, NE 68602-0499 David Van Der Kamp Licensing Manager Nebraska Public Power District P.O. Box 98 Brownville, NE 68321

Michael J. Linder, Director Nebraska Department of Environmental Quality P.O. Box 98922 Lincoln, NE 68509-8922

Chairman Nemaha County Board of Commissioners Nemaha County Courthouse 1824 N Street Auburn, NE 68305 Julia Schmitt, Manager Radiation Control Program Nebraska Health & Human Services Dept. of Regulation & Licensing Division of Public Health Assurance 301 Centennial Mall, South P.O. Box 95007 Lincoln, NE 68509-5007

Nebraska Public Power District - 3 -

H. Floyd Gilzow Deputy Director for Policy Missouri Department of Natural Resources P. O. Box 176 Jefferson City, MO 65102-0176 Director, Missouri State Emergency Management Agency P.O. Box 116 Jefferson City, MO 65102-0116 Chief, Radiation and Asbestos Control Section Kansas Department of Health and Environment Bureau of Air and Radiation 1000 SW Jackson, Suite 310 Topeka, KS 66612-1366 Melanie Rasmussen, State Liaison Officer/Radiation Control Program Director Bureau of Radiological Health Iowa Department of Public Health Lucas State Office Building, 5th Floor 321 East 12th Street Des Moines, IA 50319

John F. McCann, Director, Licensing Entergy Nuclear Northeast Entergy Nuclear Operations, Inc. 440 Hamilton Avenue White Plains, NY 10601-1813 Keith G. Henke, Planner Division of Community and Public Health Office of Emergency Coordination 930 Wildwood, P.O. Box 570 Jefferson City, MO 65102 Paul V. Fleming, Director of Nuclear Safety Assurance Nebraska Public Power District P.O. Box 98 Brownville, NE 68321

Nebraska Public Power District - 4 -

Ronald L. McCabe, Chief Technological Hazards Branch National Preparedness Division DHS/FEMA 9221 Ward Parkway, Suite 300 Kansas City, MO 64114-3372 Daniel K. McGhee, State Liaison Officer Bureau of Radiological Health Iowa Department of Public Health Lucas State Office Building, 5th Floor 321 East 12th Street Des Moines, IA 50319 Ronald D. Asche, President and Chief Executive Officer Nebraska Public Power District 1414 15th Street Columbus, NE 68601

Nebraska Public Power District - 5 -

Electronic distribution by RIV:

Regional Administrator (EEC) DRP Director (DDC) DRS Director (RJC1) DRS Deputy Director (ACC) Senior Resident Inspector (NHT) Branch Chief, DRP/C (MCH2) Senior Project Engineer, DRP/C (WCW) Team Leader, DRP/TSS (CJP) RITS Coordinator (MSH3) DRS STA (DAP) ROPreports D. Pelton, OEDO RIV Coordinator (DLP) CNS Site Secretary (SEF1) D. Starkey, OE (DRS) OE Mail M. Ashley, NRR (MAB) M. Vasquez (GMV) V. Dricks (VLD) W. Maier (WAM)

SUNSI Review Completed: LJS ADAMS:

~ Yes No Initials: LJS

~ Publicly Available Non-Publicly Available Sensitive

~ Non-Sensitive SRI/EB2 RI/EB2 DRS/EB2 DRS/EB2 C:DRP/C JMMateychick HAbuseini RMullikin BCorrell MCHay /RA/ /RA/ /RA/ /RA/ /RA/ 1/04/08 1/4/08 1/8/08 1/7/08 1/14/08 C:DRS/EB2 C:/DRS ACES C:DRS/EB2 NFO'Keefe RJCaniano MGVasquez NFO'Keefe /RA/ /RA/ /RA/

/RA/ 1/30/08 1/31/08 1/31/08 2/1/08 OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax

Enclosure - 1 - U.S. NUCLEAR REGULATORY COMMISSION REGION IV

Docket.: 50-298 License: DPR-46 Report: 05000298/2007008

Licensee: Nebraska Public Power District Facility: Cooper Nuclear Station Location: P.O. Box 98 Brownville, Nebraska Dates: May 21 through December 26, 2007 Team Leader: J. M. Mateychick, Senior Reactor Inspector, Engineering Branch 2

Inspectors: H. Abuseini, Reactor Inspector, Engineering Branch 2 Accompanying R. Mullikin, Consultant Personnel: B. Correll, Reactor Inspector, Engineering Branch 2

Approved By: Linda Joy Smith, Chief Engineering Branch 2 Division of Reactor Safety

Enclosure - 2 -

SUMMARY OF FINDINGS

IR 05000298/2007008; 05/21/07 - 12/26/07; Cooper Nuclear Station: Triennial Fire Protection Inspection

The report covered a 2-week period of inspection by region-based inspectors and a contractor. No findings of significance were identified. The significance of most findings is indicated by its color (Green, White, Yellow, Red) using Inspection Manual Chapter 0609, "Significance Determination Process." Findings for which the significance determination process does not apply may be green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3, dated July 2000.

Cooper Nuclear Station formally committed to converting their fire protection program to comply with the requirements of 10 CFR Part 50.48.(c) and National Fire Protection Association Standard 805. This involves using a risk-informed methodology. Because the conversion and licensing process are expected to identify and address a variety of difficult issues that are normally the subject of triennial fire protection inspections, and because any findings in this area would have to be addressed under the new, rather than the existing, program, the NRC has adapted its inspection and enforcement of certain issues for plants in this situation. As a result, the scope of this inspection was modified, and some issues raised in this inspection are documented but subject to enforcement discretion.

A. NRC-Identified and Self-Revealing Findings

No findings of significance were identified during this inspection.

B. Licensee-Identified Findings None.

REPORT DETAILS

1R05 Fire Protection

The purpose of this inspection was to review the Cooper Nuclear Station fire protection program for selected risk-significant fire areas. The inspection was performed in accordance with Inspection Procedure 71111.05TTP, "Fire Protection-NFPA 805 Transition Period (Triennial)," dated May 9, 2006, for a plant in transition to National Fire Protection Association (NFPA) Standard 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition. The NRC reduced the scope of this inspection for plants transitioning to NFPA Standard 805 by not specifically targeting safe shutdown circuit configurations for inspection. 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 Cooper Nuclear Station to choose risk-significant areas for detailed inspection and review. Inspection Procedure 71111.05TTP, "Fire Protection-NFPA 805 Transition Period (Triennial)," requires selecting a minimum of three fire areas for review. The four fire areas reviewed during this inspection were:

• Fire Area CB-C/V; Fire Zone 8B (Control Building Reactor Protection Room 1B)

• Fire Area CB-D/VII; Fire Zone 9A (Control Building Cable Spreading Room)

• Fire Area RB-CF / I; Fire Zones 1C, 2A-2, 2A-3, 2B (Reactor Building North/North-West 903' el, North-West Quad 889' and 859' el, and RHR HX A Room)

• Fire Area IS-A/XI; Fire Zone 20A (Intake Structure - Service Water Pump Area)

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

- 4 -accomplishing safe shutdown functions to independently verify whether licensee's 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,

• 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,

• Safe shutdown can be achieved and maintained with and without offsite power.

A review was also conducted to ensure that all required components in the selected systems were included in the licensee's 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

No findings of significance were identified.

.2 Protection of Safe Shutdown Capabilities

a. Inspection Scope

The team reviewed the safe shutdown equipment list, safe shutdown design basis documents, the post-fire safe shutdown analysis, and conducted plant walk downs to verify whether the shutdown components and systems necessary to achieve and maintain safe shutdown conditions for equipment in the fire areas selected for review were separated or protected so as to remain available in the event of a fire. The team also reviewed and observed walk downs of the post-fire procedures for achieving and maintaining safe shutdown to verify that the safe shutdown analysis provisions were properly implemented.

The team focused on the following functions required 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 top of active fuel,

(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

- 5 -monitoring capable of providing direct readings to perform and control the above functions.

In accordance with Inspection Procedure 71111.05TTP, "Fire Protection (Triennial)," dated March 9, 2006, as modified for a plant in transition to NFPA Standard 805, the NRC reduced the scope of this inspection by not specifically targeting safe shutdown circuit configurations for inspection. However, the team reviewed the separation of safe shutdown cables, equipment, and components within the same fire areas for a reduced sample, 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 team compared the results of this review with the results of the licensee's efforts. 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.

b. Findings

(1) Inadequate Circuit Protection

Introduction.

The team identified a violation of 10 CFR Part 50, Appendix R, Section III.G, "Fire Protection of Safe Shutdown Capability," for failure to ensure that redundant trains of safe shutdown systems in the same fire area were free of fire damage. The licensee had been granted an exemption from the requirements of 10 CFR Part 50, Appendix R, Section III.G, for some power circuits in conduits routed through the cable spreading room and relied upon for safe shutdown for a fire in that area. The team identified that the routing of one conduit containing required circuits was not consistent with the conditions of the exemption and, therefore, did not provide a level of fire protection equivalent to the technical requirements of Section III.G of Appendix R. However, discretion is being exercised to refrain from taking any enforcement action, in accordance with the NRC Enforcement Policy, "Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48)."

Description.

The licensee is required by 10 CFR 50.48, "Fire Protection," to satisfy the requirements of Section III.G of 10 CFR Part 50, Appendix R, "Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979." The NRC granted eight exemptions from the requirements of 10 CFR Part 50, Appendix R, "Fire Protection," in a Safety Evaluation Report dated September 21, 1983. The specific fire protection issues addressed in these exemption requests were found to be acceptable to the NRC staff based on the licensee's description that the configuration provided a level of fire protection equivalent to the technical requirements of Section III.G of Appendix R.

The cable spreading room did not meet the requirements of Section III.G because 20 feet of separation free of intervening combustibles or 1-hour barriers were not provided between redundant trains of safe shutdown equipment. An exemption was requested in Nebraska Public Power District letter dated June 28, 1982, with additional information submitted in Nebraska Public Power District letters dated March 18, 1983, and June 2, 1983, involving Division II 125/250V dc power feeds and

- 6 -Division II 4160V ac power feeds. These cables were routed through the southeast corner of cable spreading room in conduits. Alternate shutdown capability, which relies on Division II equipment, is used to assure post-fire safe shutdown for a fire in the cable spreading room.

In the letter dated March 18, 1983, the licensee stated, "the District committed to provide additional suppression beneath the path of the 125/250 Volt power feeds to their penetrations at the south wall of the cable spreading room. This suppression coverage will be achieved by extending down the existing ceiling-based suppression. It is envisioned that four heads will be added, one each in the vicinity of the penetration areas (i.e., 125/250 Volt dc Division II into and out of the cable spreading room, and two beneath the path of the dc power conduits)." In the letter dated June 2, 1983, the licensee provided a description and sketch of the 125/250 Volt dc Division II power conduit routing along with photos of the conduits routed through the area.

The licensee's submittals did not identify the specific conduits of concern or the functions of the required circuits. The team obtained the specific conduits of concern from other fire protection program documents and conducted a walk down of the area to determine if the conditions specified in the exemption were being maintained. Only two Division II 125/250V dc Power Conduits DC-25 and DC-333 were involved. The walk down confirmed that Conduit DC-333 and all of the Division II 4160V ac power conduits were consistent with the exemption. However, the routing of Conduit DC-25 was not consistent with the written description and sketch provided in the basis for the exemption. Conduit DC-25 was actually routed lower in the room and followed a different path to the south wall of the room. The team concluded that for Conduit DC-25, the modified pre-action suppression system would not be as effective in protecting the circuit from floor level threats as the approved configuration.

Analysis.

This finding was 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 the physical separation of circuits in the area and potential ignition sources. The equipment in the southeast corner of the cable spreading room was predominately electrical conduits and wall mounted electrical cabinets. Fires which could potentially impact Conduit DC-25 were limited to floor-based fires. Limitations and administrative controls on transient combustibles in the cable spreading room minimize the threat of a single fire of sufficient size to both damage circuits in Conduit DC-25 and spread to damage circuits in the main portion of the room. Intervening combustibles with the main portion of the room are limited to cable trays with asbestos-cement board bottoms.

Conduit DC-25 contains power cables for Valve RHR-MO-25B, the inboard isolation valve for the low pressure coolant injection mode. For a fire in the cable spreading room, stable hot shutdown is initially achieved using the high pressure coolant injection system. Transition to cold shutdown requires Valve RHR-MO-25B to be opened while aligning the residual heat removal system to the low pressure coolant injection mode. Based on the limited potential for fire damage and the time available to align the residual heat removal system into the low pressure coolant injection mode during the transition to cold shutdown, the team concluded that the risk of the as-installed configuration was of

- 7 -very low safety significance using the Fire Protection Significant Determination Process Phase 1 worksheet in NRC Inspection Manual Chapter 0609, Appendix F.

As a compensatory measure, the licensee has revised Emergency Procedure 5.4 Fire-S/D, "Fire Induced Shutdown From Outside Control Room," to require local manual operation of Valve RHR-MO-25B using the hand wheel for fires in the cable spreading room which eliminates reliance on the circuits in Conduit DC-25. The licensee has entered this issue into their corrective action program as Condition Report CR-CNS-2007-03972.

Enforcement.

License Condition 2.C.4 states, "The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Cooper Nuclear Station (CNS) Updated Safety Analysis Report (USAR) and as approved in the Safety Evaluations dated November 29, 1977; May 23, 1979; November 21, 1980; April 29, 1983; April 16, 1984; June 1, 1984; January 3, 1985; August 21, 1985; April 10, 1986; September 9, 1986; November 7, 1988; February 3, 1989; August 15, 1995; and July 31, 1998, subject to the following provision:

"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." The licensee is required by 10 CFR 50.48, "Fire Protection," to satisfy the requirements of Section III.G of 10 CFR Part 50, Appendix R, "Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979." The NRC granted eight exemptions from the requirements of 10 CFR Part 50, Appendix R, "Fire Protection," in a Safety Evaluation Report dated September 21, 1983. An exemption was requested for the cable spreading room because the area does not meet 10 CFR Part 50, Appendix R, Section III.G because 20 feet of separation free of intervening combustibles or 1-hour fire barriers are not provided between redundant trains of circuits. This exemption was granted for power circuits routed in conduits based on the licensee's submittals. Contrary to the above, the team concluded that, since original construction, Conduit DC-25 was routed in a manner not bounded by the conditions of the exemption and that the existing conditions provide a lower level of fire protection than that described in the exemption.

Because the licensee committed, prior to December 31, 2005, to adopting NFPA Standard 805 and changing their fire protection program license basis to comply with 10 CFR 50.48(c), this issue is covered by enforcement discretion in accordance with the NRC Enforcement Policy. Specifically, this issue would have been expected to be identified and addressed during the licensee's conversion to NFPA Standard 805, was entered into the licensee's corrective action program and will be corrected, and was of very low safety significance. The procedural requirements to manually operate the valve for a fire in the cable spreading room are to remain in effect as compensatory measures until the issue is resolved and compliance is restored under the revised fire protection

- 8 -program. Since this violation meets the criteria for enforcement discretion for plants in transition to a risk-informed, performance-based fire protection program as allowed per 10 CFR 50.48(c), the NRC is exercising enforcement discretion for this issue.

(2) Reliance On Unapproved Manual Actions

Introduction.

The team identified an apparent violation of License Condition 2.C.4, "Fire Protection," 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 that did not receive prior NRC approval to mitigate the effects of fire damage in lieu of providing physical protection consistent with the technical requirements of 10 CFR Part 50, Appendix R, Section III.G.2. The team considered the manual actions to be feasible and reliable; therefore, the finding was determined to be of very low safety significance. Specific procedural errors related to manual operation of select motor-operated valves are addressed separately in Section 1R05.2.b.3 below. Discretion is being exercised to refrain from taking any enforcement action, in accordance with the NRC Enforcement Policy, "Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48)."

Description.

License Condition 2.C.4 states, "The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Cooper Nuclear Station Updated Safety Analysis Report and as approved in the Safety Evaluations dated November 29, 1977; May 23, 1979; November 21, 1980; April 29, 1983; April 16, 1984; June 1, 1984; January 3, 1985; August 21, 1985; April 10, 1986; September 9, 1986; November 7, 1988; February 3, 1989; August 15, 1995; and July 31, 1998, subject to the following provision:

"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."

The Cooper Nuclear Station Updated Safety Analysis Report, Section X-18, "Appendix R Safe Shutdown," states the following:

• "The Safe Shutdown Systems shall be physically separated and protected in accordance with the requirements of 10 CFR Part 50, Appendix R, Section III.G.2, such that, a single postulated fire will not disable redundant systems and prevent the achievement of a safe shutdown." * "The Safe Shutdown Systems are operated from the Control Room or locally as required to accomplish the performance goals." * "Exemptions from the specific requirements of Appendix R, Sections III.G.2 and III.G.3 were granted by the NRC for plant areas where the detailed requirements of Appendix R are not met but where the existing configuration provides an equivalent level of fire protection safety.

- 9 -"A more detailed description of these exemptions and the affected areas can be found in the CNS Fire Hazards

Analysis.

" The NRC granted eight exemptions from the requirements of 10 CFR Part 50, Appendix R, "Fire Protection," in a Safety Evaluation Report dated September 21, 1983.

The specific fire protection issues addressed in these exemption requests were found to be acceptable to the NRC staff based on the licensee's description that the configurations provided a level of fire protection equivalent to the technical requirements of Section III.G of Appendix R. The exemptions did not include the use of local manual actions in lieu of meeting the requirements of Appendix R.

At the time of the inspection, the fire protection program relied on manual actions for fires outside of the control room for achieving and maintaining hot shutdown, as documented in the licensee's document titled, "10 CFR Part 50, Appendix R Post-Fire Safe and Alternative Shutdown Analysis Report," and Emergency Procedure 5.4POST-FIRE, "Post-Fire Operational Information." For example, for a fire in Fire Area CB-C, Core Spray System Train A is used to establish hot shutdown.

Circuits required to allow control of Motor-Operated Valve CS-MO-12A, the inboard injection throttle valve, from the control room are not physically protected from fire damage. Emergency Procedure 5.4POST-FIRE, Attachment 6, requires operation of the valve from a motor-control center. Operators must remove control power fuses to prevent spurious operations due to fire damage then open the valve by depressing the open contactor in the motor-control center cubical. A second example is a fire in Fire Area RB-CF, Core Spray System Train B and Suppression Pool Cooling (residual heat removal) Train B are used to establish hot shutdown. Emergency Procedure 5.4POST-FIRE, Attachment 12, requires operators to remove fuses to prevent the spurious operation of 12 automatic depressurization system valves and four low-low set valves. Control power is de-energized to allow the fuse removal then reenergized. The control power is required for the operation of the core spray and residual heat removal systems from the control room.

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. Specifically, relying on manual actions to react to fire induced maloperation is inherently less reliable than separating or protecting the same equipment from fire damage. The team reviewed Emergency Procedure 5.4POST-FIRE, "Post-Fire Operational Information" and walked down the manual actions directed in the procedure with licensee operations personnel. The team found that the manual operator actions were feasible and reliable based on the criteria in Enclosure 2 of Inspection Procedure 71111.05TTP, dated May 9, 2006. The manual operator actions could be performed within the analyzed time limits, with the exception of select valves addressed in Section 1R05.2.b.3 below. The team determined that the manual operator actions were acceptable as compensatory measures until this non-compliance is addressed during the transition to NFPA-805.

Enforcement.

The Cooper Nuclear Station License Condition 2.C.4 requires that the licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Safety Analysis Report. The Updated

- 10 -Safety Analysis Report, Section X-18, "Appendix R Safe Shutdown," states "The Safe Shutdown Systems shall be physically separated and protected in accordance with the requirements of 10 CFR Part 50, Appendix R, Section III.G.2, such that a single postulated fire will not disable redundant systems and prevent the achievement of a safe shutdown."

Section III.G.2 of Appendix R lists three options for satisfying the requirements for separation and protection of equipment needed to ensure safe shutdown remains free of fire damage. The requirements of this section do not allow using manual actions in lieu of protection and separation. Contrary to this, the team concluded that, since original construction, the licensee failed to protect and separate cables and equipment necessary to ensure safe shutdown in the event of a fire in multiple fire areas. Specifically, the licensee relied on manual actions to overcome the effects of spurious operations or circuit damage due to the effects of fire for the equipment required to achieve and maintain hot shutdowns in the event of a fire. The licensee entered this issue into their corrective action program under Condition Report CNS-2004-03034.

Because the licensee committed, prior to December 31, 2005, to adopting NFPA Standard 805 and changing their fire protection program license basis to comply with 10 CFR 50.48(c), this violation is eligible for enforcement discretion in accordance with the NRC Enforcement Policy. Specifically, this issue would have been expected to be identified and addressed during the licensee's conversion to NFPA Standard 805, was entered into the licensee's corrective action program and will be corrected, and was of very low safety significance. The manual actions are to remain in effect as compensatory measures until the issue is resolved and compliance restored. Since this violation meets the criteria for enforcement discretion for plants in transition to a risk-informed, performance-based fire protection program as allowed in accordance with 10 CFR 50.48(c), the NRC is exercising enforcement discretion to refrain from taking any enforcement action.

(3) Inadequate Post-Fire Safe Shutdown Procedures

Introduction.

The team identified an unresolved item associated with Technical Specification Section 5.4.1.d concerning failure to maintain adequate written procedures covering fire protection program implementation (EA-07-204). Specifically, emergency procedures used to achieve post-fire safe shutdown contained errors, which would have challenged operators performing actions to align 10 motor-operated valves to the positions required to achieve and maintain post-fire safe shutdown.

Description.

Technical Specification Section 5.4.1.d states, "Written Procedures shall be established, implemented and maintained covering the following activities:" and "d. Fire Protection Program Implementation." Post-fire safe shutdown at the Cooper Nuclear Station requires operations to be performed in accordance with one of two emergency procedures. For most fire areas, plant shutdown is performed from the control room using Emergency Procedure 5.4POST-FIRE, "Post-Fire Operational Information," in conjunction with other plant procedures. Emergency Procedure 5.4POST-FIRE provides information concerning systems and instrumentation available for shutdown, systems and instrumentation potentially damaged because of the fire and manual actions operators must perform outside of the control room because of fire damage. Two fire

- 11 -areas require alternate shutdown to be performed from outside of the control room in accordance with Emergency Procedure 5.4FIRE-S/D, "Fire Induced Shutdown From Outside Control Room." Emergency Procedure 5.4FIRE-S/D is a stand alone procedure.

The team performed a walkthrough of Emergency Procedure 5.4POST-FIRE for the sample fire areas. As discussed in Section 1R05.2.b.2 above, the fire protection program relies on manual actions for fires outside of the control room for achieving and maintaining hot shutdown as documented in the "10 CFR Part 50 Appendix R Post-Fire Safe and Alternative Shutdown Analysis Report" and Emergency Procedure 5.4POST-FIRE, "Post-Fire Operational Information." Emergency Procedure 5.4POST-FIRE requires operators to stroke numerous motor-operated valves to their required positions from each motor-operated valve's motor starter. The procedure steps direct operators to open the motor-operated valve motor starter cabinet, remove the control power fuses then press an open or closed contactor for a specified amount of time to stroke the valve to the required position. The operator at the motor starter has no indication confirming that the valve has stroked to desired position and the procedure does not direct the operator to verify the valve's position locally. Once the control power fuses are removed, valve position indication in the control room is not available.

During the walkthrough, the licensee did not open some motor-operated valve motor starter cabinets because of concerns of potential plant trip hazard. The team performed a table top review of the motor-operated valve circuit drawings. As a result, four 125 Vdc motor-operated valves were identified for which the operators would not have been able to perform the procedure steps as written (RHR-MO-67, RHR-921MV, RWCU-MO-18, and MS-MO-77). These four valves had motor starter cubicles that were different than most of the other cubicles in that they had motor starters designed without separate control power fuses. The team concluded that operators would have followed the operating instructions, but the valves would not stroke because removing the fuses would remove motive power. There were no indications available to indicate whether the actions were successful or not, and the procedure did not require local verification that the valves were actually in the correct position.

The licensee's extent of condition review identified procedural errors for six additional motor-operated valves. These six valves had motor starters with separate control power fuses and three or four contactors. The procedure directed operators to depress the open or closed contactor. The operators could have thought they performed the procedure steps as written, but the valves would not have actually stroked to the required positions because additional contactors needed to be operated to get the valves to stroke. Five of the valves require two contactors to be pressed simultaneously to stroke the valve (HPCI-MO-14, HPCI-MO-16, RHR-MO-25A, RHR-MO-25B and RR-MO-53A). The sixth valve (RHR-MO-17) requires three contactors to be pressed simultaneously to stroke the valve. One of the five valves with three contactors (RHR-MO-25B) is operated in the same manner during alternative shutdown in accordance with Emergency Procedure 5.4FIRE-S/D, which contains the same procedural error. The procedural errors impacted the response to fires in 14 fire areas involving one to five valves in numerous combinations (CB-A, CB-A-1, CB-B, CB-C, CB-D, RB-F, RB-FN, RB-DI (SE), RB-DI (SW), RB-J, RB-K, RB-M, RB-N and TB-A).

- 12 -The major impact of being unable to operate the valves as required could be that operators would think they aligned the valves, they would eventually be directed to perform and emergency blowdown, but then be unable to reflood the core because the injection valves were not open.

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 NRC Enforcement Policy provides enforcement discretion for certain violations of the requirements in 10 CFR 50.48, "Fire Protection,"

(or fire protection license conditions) that are identified as a result of the transition to a new risk-informed, performance-based fire protection approach included in 10 CFR 50.48(c) and lists four criteria to qualify for discretion. The NRC Enforcement Policy also states, "The NRC may take enforcement action when these conditions are not met or when a violation that is associated with a finding of high safety significance is identified." Additional information is needed to complete an evaluation of the safety significance in order to determine how this issue should be treated within the reactor oversight process, and whether enforcement discretion is appropriate for this violation.

Upon identification of this issue, the licensee took immediate compensatory actions to notify operations of the procedural problems, establish a roving fire watch, issue a night order to communicate to all operating crews, and change the procedures. Both emergency procedures have been revised to assure correct valve alignment. The licensee issued Licensee Event Report 2007-005-00 related to this issue. The licensee has entered this issue into their corrective action program as Condition Report CNS-2007-04155.

Enforcement.

Technical Specification, Section 5.4.1.d states, "Written Procedures shall be established, implemented and maintained covering the following activities: d. Fire Protection Program Implementation." The fire protection program relies on manual actions performed outside of the control room for achieving and maintaining hot shutdown as documented in the "10 CFR Part 50, Appendix R, Post-Fire Safe and Alternative Shutdown Analysis Report." The post-fire safe shutdown at the Cooper Nuclear Station requires these operations to be performed in accordance with one of two Emergency Procedures 5.4POST-FIRE, "Post-Fire Operational Information," or 5.4FIRE-S/D, "Fire Induced Shutdown From Outside Control Room." Contrary to the above, the team concluded that, from 1998 until June, 2007, inadequate procedural guidance was provided to allow operators to successfully perform required post-fire safe shutdown manual operations. Specifically, inadequate procedural guidance was provided in the procedures for the manual operation of 10 motor-operated valves from their motor starters as required by the fire protection program.

Because the licensee committed to adopting NFPA Standard 805 and changing their fire protection program license basis to comply with 10 CFR 50.48(c), this issue may be subject to enforcement discretion in accordance with the NRC Enforcement Policy. Pending completion of additional analyses to determine whether this finding is less than high safety significance, and thus whether it should be treated as a violation, this issue is being treated as an unresolved item: URI 05000298/2007008-01, Inadequate Post-Fire Safe Shutdown Procedures. (EA-070204)

- 13 -

.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 documentation, such as NRC safety evaluation reports, and deviations from NRC regulations and the National Fire Protection Association codes 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 NFPA 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. The fire brigade simulated fighting a motor control center fire in the auxiliary relay room on the Control Building 903'-6" elevation. Team members observed the fire brigade simulate fire fighting activities in the plant. 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 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.

- 14 -.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 licensee's 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 licensee's alternative shutdown methodology. Team members observed a walk-through of the control room evacuation procedures with both licensed reactor and senior reactor operators. The team observed operators simulate performing the steps of Procedure 5.4 FIRE-S/D, "Fire Induced Shutdown from Outside the Control Room," which provided instructions for performing an alternative shutdown from the dedicated 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 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

No findings of significance were identified.

- 15 -.7 Circuit Analyses This inspection activity was not performed because it has been suspended for plants in transition to NFPA Standard 805.

.8 Communications

a. Inspection Scope

The team verified through inspection of the contents of designated emergency storage lockers and review of emergency control station alternative shutdown procedures, that the portable communication equipment are available, operable, and adequate for alternative shutdown procedure performance. The inspection considered communication issues, such as ambient noise levels, clarity of reception, reliability, and coverage patterns.

b. Findings

No findings of significance were identified.

.9 Emergency Lighting

a. Inspection Scope

The team reviewed emergency lighting systems required to support plant personnel in the performance of alternative safe shutdown functions to verify it was adequate to 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.

b. Findings

No findings of significance were identified.

.10 Cold Shutdown Repairs

a. Inspection Scope

The team reviewed licensee's safe shutdown analysis and Procedures 5.4Post-Fire and 5.4 Fire-S/D to determine whether repairs were required to achieve cold shutdowns. The licensee designated four systems potentially requiring repair. The repairs included the 125/250Vdc battery chargers power cables replacement, the diesel fuel oil transfer pump power and control cables replacement, the battery room fan cable replacement, and restoration of the service air compressor to provide control air for long-term operation of the automatic depressurization system valves. The repairs were potentially required in order to reach cold shutdown based on the safe shutdown methodology implemented. The team verified that the replacement cables and tools were available and the procedure to install it was adequate. The team also evaluated whether cold shutdown could be achieved within the required time using the licensee's procedures and repair methods

- 16 -

b. Findings

No findings of significance were identified.

.11 Compensatory Measures

a. Inspection Scope

The team reviewed the licensee's 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 the Technical Requirements Manual sections applicable to active and passive fire protection equipment and Procedures 0.23, "CNS Fire Protection Plan" and 0.39.1, "Fire Watches and Fire Impairments." The team also reviewed the current fire impairment log and a sample of fire impairments 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 reviewed Procedure 0.49, "Schedule Risk Assessment" to determine whether the procedures adequately controlled compensatory measures for out-of-service, degraded, or inoperable equipment that could affect post-fire safe shutdown equipment, systems or features.

b. Findings

No findings of significance were identified.

4OA2 Problem Identification and Resolution

a. Inspection Scope

The team reviewed a sample of condition reports associated with the licensee's fire protection program to verify that the licensee had an appropriate threshold of identifying deficiencies. In addition the team reviewed the corrective actions proposed and implemented to verify that they were effective in correcting identified deficiencies. In this sample there were condition reports written to address the following NRC findings identified in the 2004 Triennial Fire Protection Inspection:

• Condition Report CNS-2004-05510, "Failure to ensure redundant trains of safe shutdown n the same area were free of fire damage"

• Condition Report CNS-2004-05511, "Failure to provide adequate instructions in Emergency Procedure 5.4 Fire-S/D"

- 17 -The team also reviewed findings from the 2004 Triennial Fire Protection Inspection regarding Procedure 5.4 FIRE-S/D, and documented in NRC Inspection Report 05000298/2004008-02, to verify that these were addressed and adequate corrective actions implemented.

b. Findings

No findings of significance were identified.

4OA6 Management Meetings

Debrief Meeting Summary On June 15, 2007, the team leader presented the inspection results to Mr. M. Colomb, General Manager of Plant Operations, and other members of licensee management. The team destroyed or returned all proprietary information reviewed during the inspection to the licensee.

Exit Meeting Summary

The team leader presented the inspection results to Mr. S. Minahan, Vice-President-Nuclear and CNO, and other members of licensee management at the conclusion of the inspection in a conference call on December 26, 2007.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

J. Bebb, Security Manager

R. Beilke, Chemistry Manager

V. Bhardwaj, Engineering Support Manager

M. Boyce, Director of Projects

D. Buman, Acting Director of Engineering

D. Chambliss, Director of Nuclear Safety Assurance

M. Colomb, General Manager of Plant Operations

P. Donahue, Information Technology Manager

L. DuBois, Quality Assurance Auditor

R. Dyer, Heat Exchanger Program Engineer

J. Dykstra, Electrical Engineering Programs Supervisor

R. Estrada, Corrective Action and Assessment Manager

J. Flaherty, Senior Staff Licensing Engineer

P. Fleming, Director of Nuclear Safety Assurance

S. Freborg, Acting Engineering Support Department Manager

J. Furr, Assistant Outage Manager

V. Furr, Risk Management Engineer

P. Gritton, Business Services Manager

D. Kimball, Planning, Scheduling and Outages Manager

C. Long, Fire Protection System Engineer

J. Long, Acting Assistant Operations Manager - Operations Shift

M. Matheson, Senior Staff Engineer - Design Engineering

S. Minahan, Vice-President-Nuclear and CNO

A. Mitchell, Design Engineering Manager

D. Oshlo, Radiation Protection Manager

T. Ratzlaff, Operations Specialist - Operations

R. Shaw, Operations Shift Manager

T. Shudak, Fire Protection Program Engineer

R. Stephan, Risk Assessment Engineer

D. VanDerKamp, Licensing Supervisor

J. Waid, Training Manager

D. Werner, Operations Training Superintendent

D. Willis, Operations Manager

K. Wright, Administrative Services

NRC

N. Taylor, Senior Resident Inspector

M. Chambers, Resident Inspector

A-19

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000298/2007008-01 URI Inadequate Post-Fire Safe Shutdown Procedures (Section 1R05.2.b.3)

Opened and Closed

None

Closed

None

Discussed

None

LIST OF DOCUMENTS REVIEWED

Cable Routing Data Component Component Component Component

HPCI-MOV-014 HPCI -MOV-016

RCIC-MOV-015

RCIC-MOV-0131

Calculations

Number Title Revision NEDC 91-03 Qualification of Fire Barrier Penetration Seal Details 5

NEDC 07-078 System 5 Hydraulic Calculation 0

0840-045-C06 Silicone Foam Duct Penetration Seals 1

0840-045-C07 Pipe/Duct Sealed With Grout or Foam and Sheet Metal Plates 1

Condition Reports

(CRs)

CR-CNS-2004-04915

CR-CNS-2006-02995

CR-CNS-2007-03902*

CR-CNS-2004-05510

CR-CNS-2006-03258

CR-CNS-2007-03969

CR-CNS-2004-03034

CR-CNS-2006-04815

CR-CNS-2007-03972*

CR-CNS-2004-03595

CR-CNS-2006-04829

CR-CNS-2007-04080*

CR-CNS-2004-05511

CR-CNS-2007-01136

CR-CNS-2007-04041*

CR-CNS-2004-06307

CR-CNS-2007-01139

CR-CNS-2007-04155*

CR-CNS-2005-01785

CR-CNS-2007-01263

CR-CNS-2007-04156*

CR-CNS-2005-02167

CR-CNS-2007-01324

CR-CNS-2007-04168*

CR-CNS-2005-02231

CR-CNS-2007-01448

CR-CNS-2007-04171*

CR-CNS-2005-02543

CR-CNS-2007-01449

CR-CNS-2007-04173*

CR-CNS-2005-04652

CR-CNS-2007-01451

CR-CNS-2007-04175*

CR-CNS-2005-05392

CR-CNS-2007-01484

CR-CNS-2007-04178

CR-CNS-2006-00112

CR-CNS-2007-01658

CR-CNS-2007-04179*

CR-CNS-2006-00138

CR-CNS-2007-03536

CR-CNS-2007-04181*

CR-CNS-2006-00222

CR-CNS-2007-03678*

CR-CNS-2007-04184*

CR-CNS-2006-01118

CR-CNS-2007-03714*

CR-CNS-2007-04187*

CR-CNS-2006-01710

CR-CNS-2007-03734* DEPTD

1-05100

CR-CNS-2006-02008

CR-CNS-2007-03825* SCAQ 95-0047

CR-CNS-2006-02362 CR-CNS-2007-03826*

• CR initiated due to inspection activities.

Drawings

Number Title Revision Automatic Sprinkler 34-1243 SH, Sht. 13 N.P.P.D. Contract No. E-69-20 N04

CNS-EE-178 Safe Shutdown Component Locations & Emerg Route Lighting 903'-6" Reactor Building

CNS-EE-181 Safe Shut Down Component Locations & Emerg Route Lighting 932'-6" Critical Switchgear Rooms

CNS-EE-186 Safe Shut Down Component Locations & Emerg Route Lighting 903'-6" Diesel Gen Building

CNS-EE-189 Safe Shutdown Component Locations & Emerg Route Lighting 882'-6", 877'-6" & 903'-6" Control Bldg

CNS-FP-62 Fire Area Boundary Drawing, Cable Spreading Room East Wall 918'-0" Control Building

N08

CNS-FP-64 Fire Area Boundary Drawing, Cable Spreading Room South Wall 918'-0" Control Building

N03

CNS-FP-65 Fire Area Boundary Drawing, Cable Spreading Room South Wall 918'-0" Control Building

N03

CNS-FP-70 Fire Area Boundary Drawing, Cable Spreading Room West Wall 918'-0" Control Building

N03

A-21CNS-FP-72 Fire Area Boundary Drawing, Cable Spreading Room North Wall 918'-0" Control Building

N01

CNS-FP-75 Fire Area Boundary Drawing, Cable Spreading Room Floor Plan Section 1, 918'-0" Control Building

N05

CNS-FP-76 Fire Area Boundary Drawing, Cable Spreading Room Floor Plan Section 2, 918'-0" Control Building

N04

CNS-FP-77 Fire Area Boundary Drawing, Cable Spreading Room Floor Plan Section 3, 918'-0" Control Building

N05

CNS-FP-79 Fire Area Boundary Drawing, Cable Spreading Room Floor Plan Section 5, 918'-0" Control Building

N05

CNS-FP-86 Fire Area Boundary Drawing, Reactor Protection System Room 1B East Wall, 903'-6" Control Building

N04

CNS-FP-87 Fire Area Boundary Drawing, Reactor Protection System Room 1B South Wall, 903'-6" Control Building

N03

CNS-FP-88 Fire Area Boundary Drawing, Reactor Protection System Room 1B West Wall, 903'-6" Control Building

N06

CNS-FP-89

Fire Area Boundary Drawing, Reactor Protection System Room 1B North Wall, 903'-6" Control Building

N04

CNS-FP-90

Fire Area Boundary Drawing, Reactor Protection System Room 1B Floor Plan, 903'-6" Control Building

N05

CNS-FP-91 Fire Area Boundary Drawing, Reactor Protection System Room 1B Ceiling Plan, 903'-6" Control Building

N02

CNS-FP-213

Fire Protection Pre-Plan, Reactor Building Northwest Quadrant, 881'-9" and 859'-9"

CNS-FP-215 Fire Protection Pre-Plan, Reactor Building First Floor, Elevation 903'-6" 3

CNS-FP-226 Fire Protection Pre-Plan, Control Building RPS Room 1B, Elevation 903'-6" 4

CNS-FP-229 Fire Protection Pre-Plan, Control Building Cable Spreading Room, Elevation 918'-0" 3

CNS-FP-256 Fire Protection Pre-Plan, Intake Structure, Elevation 903'-6" 3

CNS-FP-285 Sheet 1 CNS Fire Barrier Penetration Seal Details N03

CNS-FP-285 Sheet 2 CNS Fire Barrier Penetration Seal Details N03

CNS-FP-285 Sheet 3 CNS Fire Barrier Penetration Seal Details N05

CNS-FP-310 Appendix A Barrier Drawing, Intake Structure, Service Water Pump Room / Elev. 903', Fire Zone 20A

N02 GEN08-004668 Sheet No. 2 Field Wiring, Halon 1301 System, Water Service Pump Room N.P.P.D 2 X1001 Fire Protection - FP Control Bldg. Cable Spreading Room N01 791E271, Sheet 3 HPCI System Elementary Diagram N21

791E271, Sheet 6 HPCI System Elementary Diagram N19

791E271, Sheet 7 HPCI System Elementary Diagram N19

2791E271, Sheet 8 HPCI System Elementary Diagram N19 791E271, Sheet

HPCI System Elementary Diagram N20 2001

SH.1 Flow Diagram Symbols and Abbreviations N16 2001

SH.2 Flow Diagram Symbols and Abbreviations N07

2002

SH.1 Flow Diagram Main, Exhaust & Auxiliary Steam

Systems N39

2002

SH.2 Flow Diagram Main, Exhaust & Auxiliary Steam

Systems N36

2006

SH.4 Flow Diagram - Control Building Service Water System N45

2010

SH.1 Flow Diagram - Instrument Air Control & Turbine

Building N87

2010

SH.2 Flow Diagram - Instrument Air Control & Turbine

Building N89

2010

SH.1A Flow Diagram - Instrument Air Control & Turbine

Building N12

2016 Sh. 1B Flow Diagram - Fire Protection Cont. RDW & ARDW Bldg.'s 1

2016 Sht. 4 Flow Diagram - Halon and Cardox System 2

28 Flow Diagram - Reactor Building & Drywell Equipment Drain System N47 2036

SH.1 Flow Diagram - Reactor Building Service Water System N93 2040

SH.1 Flow Diagram - Residual Heat Removal System N76

2040 Sh 2 Flow Diagram - Residual Heat Removal System Loop "B" N15

2041 Flow Diagram - Reactor Bldg Main Steam System N78

2042 Sh1 Flow Diagram - Reactor Water Clean-Up System N32

2042

SH.2 Flow Diagram - Reactor Water Clean-Up System N12

2043 Flow Diagram - Reactor Core Isolation Coolant and Reactor Feed Systems

N50 2044 Flow Diagram - High Pressure Coolant Injection and Reactor Feed Systems

N69 3001 Main One Line Diagram N15 3002 SH. 1 Auxiliary One Line Diagram MCC Z, SWGR BUS 1A, 1B, 1E, & Critical SWGR BUS 1F 1G

N41 3003 SH. 2 Auxiliary One Line Diagram Motor Control Centers A, B, F &

G N39 3004 SH. 3 Auxiliary One Line Diagram MCC

C, D, H, J DG1 & DG2 N20 3005 SH. 4 Auxiliary One Line Diagram Motor Control Centers M, N, P, U, V & W N48 3006

SH.5 Auxiliary One Line Diagram Starter Racks LZ and TZ MCC'S K, L, LX, RA, RX, S, T, TX, X

N67

A-233007 SH 6 Auxiliary One Line Diagram Motor Control Centers E, Q, R, &

Y N79 3010 SH 1 Cooper Nuclear Station Vital One Line Diagram N67 3020 SH 4

4160V Switchgear Elementary Diagrams

N19

23 SH 7

4160V Switchgear Elementary Diagrams N1

27 SH 1

480V Switchgear Elementary Diagrams Sheet # 1 N25

28 SH 2 480V Switchgear Elementary Diagrams Sheet # 2

N17

3050 Sheet 11A Cable and Conduit Schedule N09 3050 Sheet 11S Cable and Conduit Schedule N03 3050 Sheet 26D Cable and Conduit Schedule N03 3058 DC One Line Diagram N47

29 Sh. 1 Control Building, Cable & Control Rooms Conduit, Cable Tray & Grounding Plans

N25

PROCEDURES

Number Title Revision Administrative Procedure 0.1

Procedure

Use and Adherence 31 Administrative Procedure 0.16 Control of Doors 36 Administrative Procedure 0.23 CNS Fire Protection Plan 47 Administrative Procedure 0.29.4 Other Regulatory Reviews 10 Administrative Procedure 0.36.8 Electrical Safety Rule Book 6 Administrative Procedure 0.39 Hot Work 35 Administrative Procedure 0.39.1 Fire Watches and Fire Impairments 0 Administrative Procedure 0.49 Schedule Risk Assessment 19 Administrative Procedure 0.7.1 Control of Combustibles 22 Administrative Procedure 2.0.1.2 Operations Procedure Policy 27 Administrative Procedure 2.0.3 Conduct of Operations 58 Emergency

Procedure

5.4 Fire General Fire Procedure 14

A-24Number Title Revision Emergency

Procedure

5.4

Post-Fire Post-Fire Operational Information 12 & 13 Emergency

Procedure

5.4 Fire-S/D Fire Induced Shutdown From Outside Control Room 14 & 15 Engineering Procedure 3.6.1 Fire Barrier Control 14

STP 94-075 Appendix 'R' Emergency Lighting Verification Test 6-15-98 Surveillance Procedure

6.ADS.202 ADS Manual Valve Circuit Continuity from

ASD-ADS Panel 8 Surveillance Procedure 6.ASDR.301 ASD Panel Instrument Checks 5 Surveillance Procedure 6.ASDR.301 Surveillance Procedure 6.FP.203Fire Damper Assembly Examination (Fire Protection System 18 Month Examinations)

Surveillance Procedure 6.FP.302 Automatic Deluge and Pre-Action Systems Testing 14 Surveillance Procedure

6.FP.303 Operations Deluge and Pre-Action Systems Testing 11 Surveillance Procedure 6.FP.304 Fire Detection System Circuitry Operability 6 Surveillance Procedure 6.FP.305 Halon 1301 Service Water Pump Room Fire Suppression Surveillance Checks Surveillance Procedure 6.FP.306 Fire Detection Systems Semi-Annual Examination 11 Surveillance Procedure 6.FP.601 Fire Protection System 31 Day Examination 14 Surveillance Procedure

6.FP.602 Engineers Fire Protection Examination 6 Surveillance Procedure 6.FP.604 Fire Door Annual Examination 17 Surveillance Procedure 6.FP.605 System Number 5 Flow Verification Test 5 Surveillance Fire Barrier/Fire Wall Visual Examination 10

A-25Number Title Revision Procedure 6.FP.606 Surveillance Procedure 6.HPCI.102 HPCI Test Mode Surveillance Operation From

ASD-HPCI

Panel 16 Surveillance Procedure

6.HPCI.106 FCU

FC-R-1G Operability Test From

ASD-HPCI Panel 6 Surveillance Procedure

6.HPCI.202 HPCI Motor Operated Valve Operability Test From

ASD-HPCI

Panel 4 Surveillance Procedure 6.HPCI.319 HPCI Gland Seal Condenser Hotwell Level Channel Calibration From

ASD-HPCI Panel Surveillance Procedure 6.HPCI.320 HPCI Turbine Oil Pressure Functional Test From

ASD-HPCI

Panel 7 Surveillance Procedure 6.HPCI.321 HPCI Turbine Trip and Operability Test From

ASD-HPCI

Panel 3 Surveillance Procedure

6.HPCI.323 HPCI Control System Readout Test From

ASD-HPCI Panel 8 Surveillance Procedure 6.PC.309 Suppression Chamber Temperature Indication Channel Calibration Test From

ASD-ADS Panel Surveillance Procedure 6.REC.102 REC Pump Operability Test From

ASD-ADS Panel 5C1 Surveillance Procedure 6.RHR.201 RHR Motor Operated Valve Operability Test From

ASD-ADS

Panel 6 Surveillance Procedure 7.3.12.2 Safe Shutdown BBESI Emergency Lighting Unit Examination and Maintenance

MISCELLANEOUS DOCUMENTS

Number Title Revision

Cooper Nuclear Station Appendix R to 10 CFR Part 50 Post-Fire Repair Required

Implementation Times Cooper Nuclear Station 10 CFR Part 50 Appendix R Post-Fire Safe and

Alternative

Shutdown Analysis Report Volumes 1 & 2 06/30/2004 Cooper Nuclear Station 10 CFR Part 50 Appendix R Post-Fire Safe and Alternative Shutdown Analysis Report, January 1998

A-26Number Title Revision Volume 1, Appendix H - Safe Shutdown System Functional Requirements

Cooper Nuclear Station Appendix R Post-Fire Safe Shutdown Topical Design Criteria Document - Appendix L: Appendix R Emergency Lighting January 1998 Engineering Evaluation 01-008 Assessment of RJA Hydraulic Calculations for CNS Sprinkler Systems Fire Hazards Analysis Appendix B Suppression Effects Analysis 06/20/02 Job Performance Measure SKL034-10-60 Energizing

MCC-E From 4160V Bus 1F 04.02 Job Performance Measure SKL034-10-88 Shutdown From Outside of the Control Room, Control Room Operator Action (5.2.1) (Control Building/Critical Switchgear Operator Actions)

Lesson Plan

COR0090103 Licensed and STE Personnel Requal Cycle 01-03 for 2007 08

Lesson Plan OTH015-92-02 Post Fire and Shutdown Outside the Control Room Procedures (5.4POST-FIRE, 5.4FIRE-S/D,5.1ASD)

Lesson Plan SKL009-04-01 Miscellaneous OJT/TPE Qualification Card (Initial) 28

Lesson Plan SKL010-01-01 OPS NRC RO Qualification Guide 18

Lesson Plan SKL051-51-59 Battery Explosion/Shutdown from Outside the Control Room 12 LO-CNSLO-2006-

0059 CNS Focused Self-Assessment Report - Fire Protection Program August 2006

MWR 95-0145

EE-LTG-R12 Repair or Replace

950109

MWR 95-0130

EE-LTG-R82 Repair or Replace

950109

MWR 95-0109

EE-LTG-R64 Repair or Replace

950109

NPPD Letter LQA8200158 Fire Protection Rule 10 CFR Part 50, Appendix R 6/28/82 NPPD Letter LQA8300109 Fire Protection Rule 10 CFR Part 50, Appendix R, Preliminary Supplemental Response (Revised) 3/18/83 NPPD Letter Fire Protection Rule 10 CFR Part 50, Appendix R, Preliminary Supplemental Response (Revision 2) 6/02/83 NRC Letter Cooper Nuclear Station - Fire Protection Technical Specifications 11/29/77 NRC Letter Amendment No. 56 to Facility Operating License No.

DPR-46 5/23/79 NRC Letter Amendment No. 59 to Facility Operating License No.

DPR-46 9/18/79 NRC Letter Amendment No. 66 to Facility Operating License No.

DPR-46 11/21/80

A-27Number Title Revision NRC Letter Amendment No. 82 to Facility Operating License No.

DPR-46 4/29/83 NRC Letter Exemption Requests - 10 CFR Part 50.48 Fire Protection and Appendix R to 10 CFR Part 50 9/21/83 NRC Letter Safety Evaluation for Appendix R to 10 CFR Part 50, Items

III.G.3 and

III.L, Alternate or Dedicated Shutdown Capability 4/16/84 NRC Letter Amendment No. 86 to Facility Operating License No.

DPR-46 6/01/84 NRC Letter Amendment No. 90 to Facility Operating License No.

DPR-46 01/03/85 NRC Letter Outstanding Fire Protection Modifications 8/21/85 NRC Letter Amendment No. 98 to Facility Operating License No.

DPR-46 4/10/86 NRC Letter Amendment No. 101 to Facility Operating License No.

DPR-46 9/09/86 NRC Letter Amendment No. 126 to Facility Operating License No.

DPR-46 11/07/88 NRC Letter Amendment No. 127 to Facility Operating License No.

DPR-46 02/03/89 NRC Letter Revocation of Exemption From 10 CFR Part 50, Appendix

R 8/15/95 NRC Letter Amendment No. 178 to Facility Operating License No.

DPR-46 7/31/98 QA Audit 07-01 Fire Protection Program 02/2007

TDQ 0201GEN Generic Operator Qualification 2007

TDQ 0501 Shift Technical Engineer - Initial 12/8/2006

Technical Specification 5.4 Procedures Amend. No. 178 Technical Specification 3.3.3.2 Alternate Shutdown System Amend. No. 178 Training Program Description 0528 Shift Technical Engineer - Requal 2/7/01 Training Program Description 0631 Shift Communicator 8/3/98 TRM Section

T 3.11.1 Fire Detection Instrumentation 12/18/03 TRM Section

T 3.11.2 Fire Suppression Water System 10/10/01 TRM Section

T 3.11.3 Sprinkler Systems 10/10/01 TRM Section

T 3.11.4 High Pressure Carbon Dioxide Extinguishing System 10/10/01

A-28Number Title Revision TRM Section

T 3.11.5 Halon 1301 Fire Suppression System 10/10/01 TRM Section

T 3.11.6 Fire Hose Stations 10/10/01 TRM Section

T 3.11.7 Fire Barrier and Fire Wall Penetration Fire Seals 10/10/01 USAR Section

VII-18 Alternative Shutdown Capability 07/24/01 USAR Section X-9 Fire Protection System 01/11/07

USAR Section X-18 Appendix R Safe Shutdown 01/29/03

USAR Section XIII-

Fire Protection Program 11/30/04

MODIFICATIONS

Number Title Revision

CED 1998-0191 Appendix "R" Emergency Battery Lights Upgrade 0

DC 94-075B Appendix R Safe Shutdown Lighting Addition Jan 1995