IR 05000298/2003003

IR 05000298-03-003, on 05/03/2003 - 05/23/2003, Cooper Nuclear Station; Safety System Design and Performance Capability

ML031960507
Person / Time
Site:	Cooper
Issue date:	07/14/2003
From:	Marschall C Division of Reactor Safety IV
To:	Warren C Nebraska Public Power District (NPPD)
References
IR-03-003
Download: ML031960507 (29)

Text

uly 14, 2003

SUBJECT:

COOPER NUCLEAR STATION - NRC INSPECTION REPORT 05000298/2003003

Dear Mr. Warren:

On May 23, 2003, the NRC completed an inspection at your Cooper Nuclear Station. The enclosed report documents the inspection findings, which were discussed on May 23 and July 2, 2003, with you and other members of your staff.

This inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.

Within these areas, the inspection consisted of selected examination of procedures and representative records, observations of activities, and interviews with personnel.

Based on the results of this inspection, the NRC has identified one finding of very low safety significance (Green). If you contest the violation or significance of the noncited violation, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with 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 Cooper Nuclear Station facility.

In accordance with 10 CFR 2.790 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 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/ by RLN for CSM Charles S. Marschall, Chief Engineering and Maintenance Branch Division of Reactor Safety

Nebraska Public Power District -2-Docket: 50-298 License: DPR-46

Enclosure:

NRC Inspection Report 5000298/2003003

REGION IV==

Docket: 50-298 License: DPR 46 Report No.: 05000298/2003003 Licensee: Nebraska Public Power District Facility: Cooper Nuclear Station Location: P.O. Box 98 Brownville, Nebraska Dates: May 5-23, 2003 Lead Inspector: W. McNeill, Senior Reactor Inspector, Engineering and Maintenance Branch Inspectors: P. Goldberg, Senior Reactor Inspector, Engineering and Maintenance Branch T. McConnell, Reactor Inspector, Engineering and Maintenance Branch R. Mullikin, Senior Reactor Inspector, Engineering and Maintenance Branch S. Schwind, Senior Resident Inspector, Projects Branch C J. Taylor, Reactor Inspector, Engineering and Maintenance Branch Accompanying C. Baron, Contractor, Beckman and Associates Personnel:

Approved By: Charles S. Marschall, Chief Engineering and Maintenance Branch Division of Reactor Safety

-2-SUMMARY OF FINDINGS IR 05000298/2003003; 5/3-23/2003; Cooper Nuclear Station; Safety System Design and Performance Capability The NRC conducted an inspection with six regional inspectors and one contractor. The inspection identified one green finding. The NRC indicates the significance of most findings by their color (green, white, yellow, red) using IMC 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 described the program for overseeing the safe operation of commercial nuclear power reactors in NUREG-1649, "Reactor Oversight Process," Revision 3, dated July 2000.

Cornerstone: Barrier Integrity

• Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, regarding the surveillance test procedures associated with Technical Specification Surveillance Requirement 3.6.4.3.4. The surveillance test procedures used to periodically verify that bypass flow through the idle train of the standby gas treatment system did not include adequate allowances for test measurement uncertainty in the acceptance criteria. The damper provided some flow in the idle train to prevent fire in the charcoal filter medium, but idling the train means a lower filtering efficiency in the idle train.

The finding is greater than minor because the standby gas treatment system bypass flow did not meet the design limits for control room dose rate concerns (See Example 3.i of Appendix E of Inspection Manual Chapter 0612). The licensees engineering staff recalculated the maximum allowable flow. The new analysis demonstrated that control room habitability remained assured. The inspectors considered this finding to be of very low safety significance because it did not represent an actual loss-of-safety function (Section 1R21.6).

Report Details 1. REACTOR SAFETY Introduction The NRC conducted an inspection to verify that the licensee adequately preserved the facility safety system design and performance capability and that the licensee preserved the initial design in subsequent modifications of the systems selected for review. The scope of the review also included any necessary nonsafety-related structures, systems, and components that provided functions to support safety functions. This inspection also reviewed the licensees programs and methods for monitoring the capability of the selected systems to perform the current design basis functions. This inspection verified aspects of the initiating events, mitigating systems, and barrier cornerstones.

The licensee based the probabilistic risk assessment model for the Cooper Nuclear Station on the capability of the as-built safety systems to perform their intended safety functions successfully. Inspectors determined the area and scope of the inspection by reviewing the licensees probabilistic risk analysis models to identify the most risk significant systems, structures, and components. Inspectors establish this according to their ranking and potential contribution to dominant accident sequences and/or initiators.

The inspectors also used a deterministic effort in the selection process by considering recent inspection history, recent problem area history, and all modifications developed and implemented.

Inspectors reviewed in detail the containment and dc systems. The primary review prompted parallel review and examination of support systems, such as, electrical power, instrumentation, and related structures and components.

The inspectors assessed the adequacy of calculations, analyses, engineering processes, and engineering and operating practices that the licensee used for the safety systems selected and the necessary support systems during normal, abnormal, and accident conditions. Acceptance criteria used by the NRC inspectors included NRC regulations, the technical specifications, applicable sections of the Updated Safety Analysis Report, applicable industry codes and standards, and industry initiatives implemented by the licensees programs.

1R02 Evaluations of Changes, Tests, or Experiments (71111.02)

a. Inspection Scope Inspectors reviewed six licensee-performed 10 CFR 50.59 evaluations to verify that the licensee had appropriately considered the conditions under which the licensee may make changes to the facility or procedures or conduct tests or experiments without prior NRC approval.

-2-The inspectors reviewed an additional ten licensee-performed 10 CFR 50.59 screenings, in which the licensee determined that evaluations were not required, to ensure that the licensees exclusion of a full evaluation was consistent with the requirements of 10 CFR 50.59.

The inspectors reviewed and evaluated the most recent licensee 10 CFR 50.59 program audit to determine whether the licensee conducted sufficient in-depth analyses of their program to allow for the identification and subsequent resolution of problems or deficiencies.

b. Findings No findings of significance were identified.

1R21 Safety System Design and Performance Capability (71111.21)

.1 System Requirements a. Inspection Scope The inspectors inspected the following attributes of the containment and dc systems:

(1) process medium (water, steam, and air), (2) energy sources, (3) control systems, and (4) equipment protection. The inspectors examined the procedural instructions to verify instructions as consistent with actions required to meet, prevent, and/or mitigate design basis accidents. Inspectors also considered requirements and commitments identified in the Updated Safety Analysis Report, technical specifications, design basis documents, and plant drawings.

b. Findings No findings of significance were identified.

.2 System Condition and Capability a. Inspection Scope Inspectors reviewed the periodic testing procedures for the containment and dc systems to verify that the licensee periodically verified the capability of the systems. The inspectors also reviewed the systems' operations by conducting system walkdowns; reviewing normal, abnormal, and emergency operating procedures; and reviewing the Updated Safety Analysis Report, technical specifications, design calculations, drawings, and procedures.

-3-b. Findings No findings of significance were identified.

.3 Identification and Resolution of Problems a. Inspection Scope The inspectors examined a sample of problems identified by the licensee in the corrective action program to evaluate the effectiveness of corrective actions related to design issues. The sample included open and closed condition reports for the past 3 years that identified issues affecting the selected systems. Inspectors reviewed older condition reports that the inspectors identified while performing other areas of the inspection.

b. Findings No findings of significance were identified.

.4 System Walkdowns a. Inspection Scope The inspectors performed walkdowns of the accessible portions of the containment and dc systems, and required support systems. Inspectors focused on the installation and configuration of switchgear, motor control centers, manual transfer switches, field cabling, raceways, piping, components, and instruments. During the walkdowns, the inspectors assessed:

• The placement of protective barriers and systems;

• The susceptibility to flooding, fire, or environmental conditions;

• The physical separation of trains and the provisions for seismic concerns;

• Accessibility and lighting for any required local operator action;

• The material condition and preservation of systems and equipment; and

• The conformance of the currently-installed system configurations to the design and licensing bases.

b. Findings No findings of significance were identified.

-4-

.5 Design Review a. Inspection Scope The inspectors reviewed the current as-built instrument and control, electrical, and mechanical design of the containment and dc systems. These reviews included an examination of design assumptions, calculations, required system thermal-hydraulic performance, electrical power system performance, protective relaying, control logic, and instrument setpoints and uncertainties. The inspectors also performed selected single-failure evaluations of individual components and circuits to determine the effects of such failures on the capability of the systems to perform their design safety functions.

The inspectors inspected calculations, drawings, specifications, vendor documents, the Updated Safety Analysis Report, technical specifications, emergency operating procedures, and temporary and permanent modifications.

b. Findings No findings of significance were identified.

.6 Safety System Inspection and Testing a. Inspection Scope The inspectors reviewed the program and procedures for testing and inspecting selected components in the containment and dc systems. The review included the results of surveillance tests required by the technical specifications and selective review of Class 1E control circuits for capability to test system functions.

b. Findings Introduction The inspectors identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the inspectors identified that the surveillance test procedures used to verify the standby gas treatment system cross-tie damper position did not include allowances for test measurement uncertainty in the acceptance criteria.

Description The inspectors observed that Technical Specification Surveillance Requirement 3.6.4.3.4 requires verification of the standby gas treatment systems cross-tie damper throttle position. Operators verified this requirement by the performance of a flow test. The inspectors also determined that the engineers based the dose calculation for the design basis fuel handling accident on a flow value only 3 percent greater than the upper limit of the surveillance test acceptance criteria. As a result, the standby gas treatment cross-tie flow could have exceeded the value used in the fuel handling accident radiological dose

-5-calculation, resulting in higher than expected control room doses during an accident.

Engineers acknowledged that the uncertainty was greater than 3 percent.

Engineers designed this system with a cross tie between the two standby gas treatment trains to allow air flow in the idle train to reduce the risk of a charcoal fire due to excessive decay heat from Iodine buildup. The design limited the maximum post-accident flow through the idle train because the idle train had lower efficiencies for iodine removal. Calculation NEDC 99-032, Control Room Habitability and Offsite Dose for a Fuel Handling Accident, Revision 3, addressed the upper design limit of the standby gas treatment cross-tie flow. This calculation used a standby gas treatment cross-tie flow of 288 cubic feet per minute.

Inspectors also reviewed the surveillance procedures associated with Technical Specification Surveillance Requirement 3.6.4.3.4. Surveillance Procedures 6.1SGT.401,

"Standby Gas Treatment A Fan Capacity Test, Standby Gas Treatment B Cooling Flow Test and Check Valve IST (Division 1)," Revision 9, and 6.2SGT.401, "Standby Gas Treatment B Fan Capacity Test, Standby Gas Treatment A Cooling Flow Test and Check Valve IST (Division 2)," Revision 8, included a maximum standby gas treatment cross-tie flow acceptance criteria of 280 cubic feet per minute.

In response to the inspectors concerns regarding the cross-tie flow acceptance criteria, the licensee initiated Notification 10248715 on May 22, 2003. This notification stated that the engineering staff failed to review the standby gas treatment cross-tie flow test acceptance criteria when the engineering staff revised and approved Calculation NEDC 99-032, on March 20, 2003.

A subsequent informal analysis determined that the design flow limit could be increased from 288 to approximately 316 cubic feet per minute without exceeding the control room dose limit. This would result in over 12 percent available flow margin versus an estimated 7 percent test instrument uncertainty. Based on this analysis, the licensee concluded that no operability issue existed.

Analysis The inspectors determined that this condition affected the barrier integrity cornerstone because of the potential of degraded secondary containment integrity. The inspectors considered this finding more than minor since the finding was similar to Example 3.i of Appendix E of Inspection Manual Chapter 0612. The finding was greater than minor because engineers had to re-perform the calculation to ensure the control room remained habitable after adding flow instrument uncertainty. The engineers had to recalculate the maximum flow because the margin between the maximum design flow and the acceptance criteria was less than the test measurement uncertainty.

The inspectors assumed a potential fuel handling accident release event for the risk assessment. The inspectors found this finding resulted from a performance deficiency of very low safety significance (Green: Question 1 of Appendix E to the Inspection Manual

-6-Chapter 0612, regarding degraded standby gas treatment). Inspectors determined no other cornerstones degraded as a result of this finding.

The inspectors assessed this finding as green because it does not represent an actual loss of the standby gas treatment system or secondary containment safety functions.

The licensee implemented appropriate corrective actions to ensure continued operability of these systems.

Enforcement Criterion III, Design Control, of 10 CFR Part 50, Appendix B, requires correct translation of design requirements into procedures. Contrary to this requirement, since March 20, 2003, the licensees engineering staff did not correctly translate the design basis into the surveillance test procedures associated with Technical Specification Surveillance Requirement 3.6.4.3.4. Specifically, the licensee failed to include adequate allowances for test measurement uncertainty in the acceptance criteria. As a result, the actual standby gas treatment cross-tie flow could have exceeded the value used in the fuel handling accident radiological dose calculation, resulting in higher than expected control room doses during an accident.

Because of the very low safety significance of the finding and because the licensee entered this issue into their corrective action program as Notification 10248715 on May 22, 2003, the inspectors treated this as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000298/2003003-01).

4. OTHER ACTIVITIES (OA)

4OA6 Management Meetings Exit Meeting Summary The team leader presented the inspection results to Mr. Clay C. Warren, Vice President of Nuclear Energy, and other members of licensee management at the conclusion of the onsite inspection on May 23, 2003 who acknowledged the findings. In addition, the team leader held a final a telephone exit meeting on July 2, 2003, with Mr. Gary Kline, General Manager Engineering, and other members of licensee management.

At the conclusion of these meetings, the team leader asked the licensee's management whether any materials retained by the inspectors was proprietary. The licensee identified that the inspectors had no proprietary information.

KEY POINTS OF CONTACT Licensee D. E. Buman, Assistant Design Engineering Manger P. V. Fleming, Licensing and Regulatory Affairs Manager T. E. Hottovy, Assistant Plant Engineering Manager J. A. Hutton, Plant Manager G. J. Kline, Senior Manager of Engineering D. F. Kunsemiller, Senior Manager Quality Assurance T. P. McClure, Mechanical Engineering Supervior M. R. McCormack, Design Engineering Supervisor-Electrical C. C. Warren, Chief Nuclear Officer A. L. Williams, Engineering Support Manager R. L. Wulf, Plant Engineering Manager ITEMS OPENED AND CLOSED Opened and Closed NCV (05000298/2003003-01) Failure to implement Criterion III of 10 CFR 50, Appendix B, regarding the control of acceptance criteria (Section 1R21.6).

DOCUMENTS REVIEWED The inspectors selected and reviewed the following documents to accomplish the objectives and scope of the inspection and to support any findings:

CALCULATIONS Number Title Revision 90-319 Post-Accident Monitoring Containment Level Calculation 1 97-090I Plant Specific Technical Guidelines/Severe Accident 1 Technical Guidelines Primary Containment Pressure Limits 97-090N Plant Specific Technical Guidelines/Severe Accident 2 Technical Guidelines Reactor Pressure Vessel Level Instruments 97-090O Plant Specific Technical Guidelines/Severe Accident 1 Technical Guidelines Net Positive Suction Head Limits03-010 Evaluation of Penetration X-12 for Potential 1 Overpressurization NEDC 87-131A 250 Volt DC Division 1 Load and Voltage Study 8 NEDC 87-131A 250 Volt DC Division 1 Load and Voltage Study 9 CED/EE EE02-038

-2-Number Title Revision NEDC 87-131B 250 Volt DC Division 2 Load and Voltage Study 9 CED/EE EE02-040 NEDC 87-131C 125Volt Division 1 Load and Voltage Study 9 CED/EE EE02-041 NEDC 87-131D 125Volt Division 2 Load and Voltage Study 9 CED/EE EE02-042 NEDC 88-209 250 Volt Battery, Rack Mounting, Charger Mountings, and 2 Test Reviews NEDC 89-1966 Drywell Cooler Heat Removal Capacity 0 NEDC-91-069 Moderate Energy Line Break and Flooding Calculations 5 NEDC 93-128 Flooding Interaction between Torus Area and Quads 3 NEDC 93-184 Residual Heat Removal Heat Exchangers Thermal 1 Performance and Tube Plugging Margin NEDC 94-034A Containment Analysis Input Parameters 0 NEDC 94-034B Containment System Response for Net Positive Suction 0 Head NEDC 94-261 Calculation for Safety Analysis Report Question 5.17 and 2C1 Surveillance Procedure 6.PC.503 NEDC 95-058 Evaluation of the Overpressurization Potential for Isolated 3 Penetrations in Accordance with Generic Letter 96-06 NEDC 96-058 Evaluation of the Overpressurization Potential for Isolated 3 Penetrations in Accordance with Generic Letter 96-06 NED C 97-044A Net Positive Suction Head Margins for the Residual Heat 3 Removal and Core Spray Pumps NEDC 98-017 PC-PS-12A, B, C, D and PC-PS-101A, B, C, D Setpoints H 0 Margins for the Residual Heat Removal and Core Spray Pumps NEDC 98-042 Estimate of Containment Volumes 0 NEDC 98-043 Containment Flooding Volumes 0 NEDC 99-032 Control Room Habitability and Offsite Dose for a Fuel 1 Handling Accident NEDC-00-080 Flood Door Gap Analysis 3 NEDC 00-095A Equipment Qualification Normal Temperature, Relative 0 Humidity, Pressure and Radiation NEDC 01-080 Drywell Normal Equipment Qualification Temperature 0

-3-DESIGN CRITERIA DOCUMENTS Number Description Dated DCD-05 DC Electrical Distribution System July 7, 2002 DCD-09 Primary Containment (PC) System November 20, 2002 DCD-31 Secondary Containment Topical July 1, 2002 DCD-36 High Energy Line Break/Moderate Energy Line January 23, 2003 Break DCD-38 Internal Flooding September 4, 2002 DRAWINGS Drawing Number Title Revision

/Sheet Number 2010/3 Service Air N40 2022/1 Primary Containment Cooling & Nitrogen Inerting System N78 2022/2 Primary Containment Cooling & Nitrogen Inerting System N01 2022/3 Primary Containment Cooling & Nitrogen Inerting System N02 2027/1 Loop A Reactor Recirculation & Suppression Chamber N61 Vent Systems & Connections 2027/2 Loop A Reactor Recirculation & Suppression Chamber N10 Vent Systems & Connections 2028 Reactor Building & Drywell Equipment Drain System N43 2029 Reactor Building Demineralized Water System N33 2031/1 Reactor Building Closed Cooling Water System N20 2031/2 Reactor Building Closed Cooling Water System N61 2031/3 Reactor Building Closed Cooling Water System N23 2039 Control Rod Drive Hydraulic System N49 2045/1 Core Spray System N54 2045/2 Standby Liquid Control System N18 2047 Drywell & Suppression Chamber Composite Systems N03 2048 Drywell & Suppression Chamber Composite Systems N01 3006/5 Auxiliary One Line Diagram Starter Racks LZ and TZ, N68 Motor Control Centers K, L, LX, RA, RX, S, T, TX, X 3050/1 Wire & Cable Description & Schedule Index N08 3050/49D Cable and Conduit Schedule N03 3050/49E Cable and Conduit Schedule N03 3050/56D Cable and Conduit Schedule N03 3050/56E Cable and Conduit Schedule N03 3058 DC One Line Diagram N43 450208882 Electrical Penetration Assy NA CNS-HV-39 Reactor Building Drywell Cooling Developed Flow 1 Diagram w/Measurement & Damper Locations

-4-ENGINEERING EVALUATIONS Number Description Revision 01-035 Equipment Qualification Temperature Profile in Containment Based 0 on Small Steam Line Break and Design Basis Accident-Loss of Coolant Accident 01-080 Effect of Loss of Coolant Accident & Small Steam, Line Break 0 Accident Conditions on Drywell Fan Coil Units in Containment and Containment Penetration Piping for Closure of Generic Letter 96-06 Issue 02-067 Design Basis Accident Radiological Dose Assessment 0 Methodologies (License Amendments 187 and 196)

LICENSEE EVENT REPORTS Number Title Dated 2000-006 Torus to Drywell Vacuum Breaker Misalignment Places Plant April 4, 2000 in Condition Prohibited by Technical Specifications 2000-007 Failed Valve Motor Places Plant in Condition Prohibited by April 3, 2000 Technical Specifications 2000-007, Failed Valve Motor Places Plant in Condition Prohibited by May 10, 2000 Sup. 1 Technical Specifications 2000-008 Non-conservative Drywell Temperature Profile Places Plant May 1, 2000 in Condition Outside of Design Basis 2000-008, Non-conservative Drywell Temperature Profile Places Plant August 28, Sup. 1 in Condition Outside of Design Basis 2000 2001-001 Confusing or Incomplete Standards and Administrative March 12, 2001 Controls Results in Failure to Test an Excess Flow Check Valve 2001-003 Failure to Adequately Revise Procedures Resulted in July 6, 2001 Inadequate Fire Watches Under Certain Battery/Battery Charger Configurations and an Unanalyzed Condition 2001-006 Scheduling Error and Oversight Results in Loss of Reactor December 31, Building-to-Suppression Chamber Vacuum Relief Function 2001 2001-007 Excessive Primary Containment Leakage Discovered During January 8, Local Leak Rate Testing of Reactor Feedwater Check 2002 Valves

-5-MODIFICATIONS Number Title Dated CED 1998-0179 Upgrade of Air Supply to Control Valves December 30, 1998 HV-SOV-(SPV-259) and HV-SOV-(SPV-261)

CED 1998-0183 Dragon Model 500F Generic Valve Replacement March 25, 1999 CED 1998-0292 Replacement of Non-Essential Solenoid Valve December 21, 1999 PC-SOV-(AD-R-1A) with ASCO Model 8344B5 CED 2000-0077 Installation of Washers on Torus to Drywell June 5, 2000 Vacuum Breakers CED 2000-0098, Installation of Washers on Torus to Drywell May 24, 2000 Vacuum Breakers CED 2000-0184 Replacement of Miscellaneous Drywell Gaskets October 18, 2000 where the Material Changed from Silicone Rubber to EPDM Rubber CED 2001-0028 Revision of Programs to Include Residual Heat December 10, 2001 Removal Containment Spray Mode Components DC 93-050 Appendix J Testing in the Accident Direction October 31, 2000 DC 94-212D Penetrations X-21 and X-22 Upgrades July 21, 1998 DC 94-212F Primary Containment Nitrogen/Air Supply July 27, 1998 Penetrations MP 97-039 Thermal Overpressure Protection X-18, X-19, and September 9, 1997 X-20 NOTIFICATIONS 10086268 10181989 10234216 10246429 10247879 10088673 10183776 10238357 10246484 10247897 10093347 10184978 10242650 10246992 10248138 10093348 10185834 10246035 10247173 10248714 10093350 10192051 10246036 10247253 10248715 10093648 10222545 10246041 10247224 10248734 10100621 10227490 10246130 10247228 10248847 10103253 10229475 10246155 10247456

-6-10110550 10232224 10246260 10247530 10124002 10233899 10246301 10247740 10129097 10233900 10246402 10247771 PROBABILISTIC RISK/SAFETY ASSESSMENTS Number Title Revision PRA-PFN001 Primary Containment Pedestal Cavity 1 PRA-PFN002 Primary Containment Safety/Relief Valves and Tailpipe Vacuum 1 Breakers PRA-PFN003 Primary Containment Vent Lines and Vacuum Relief Systems 0 PRA-PFN004 Primary Containment Drywell and Wetwell 1 PRA-PFN005 Secondary Containment Reactor Building and Steam Tunnel 1 PRA-PFN006 Secondary Containment Torus Area 1 PRA-PFN007 Containment Venting and Standby Gas Treatment System 1 PRA-SN002 Containment Isolation System 1 PRA-SN006 Electrical Power System 1 PRA-SN021 Standby Gas Treatment System 2 PSA-ES022 Containment Bypass Due to a High-Energy Line Break Inside 0 Containment PROBABILISTIC RISK ASSESSMENT SYSTEM NOTEBOOKS Number Title Revision PRA-PFN005 Secondary Containment Reactor Building and Steam Tunnel 1 PRA-PFN007 Containment Venting and Standby Gas Treatment System 1 PRA-SN002 Containment Isolation System 1 PRA-SN006 Electric Power System 1

-7-PROCEDURES Number Title Revision 0.10 Operating Experience Program 11 0.5 Conduct of the Problem Identification and Resolution Process, 37 0.5.BCO Basis for Continued Operation 2 0.5.NAIT Corrective Action Implementation and Nuclear Action Item 17 Tracking 0.5.OPS Operations Review of Notifications/Operability Determinations 15 0.5.PIR Problem Identification, Review, and Classification 10 0.5.TRND Trending of Problem Identification Report Results 1 0.8 10CFR50.59 Reviews 11 1.7 Warehouse Storage 16 1.8 Warehouse Goods Issue, Return, and Shipping 34 1.6 Warehouse Marking and Tagging 13 1.5 Warehouse Receiving 32 2.0.1 Plant Operations Policy 48 2.0.2 Operations Logs and Reports 63 2.1.1 Startup Procedure 105 2.2.24A 250 Volt DC Power Checklist 1 2.2.24.1 250 Volt DC Electrical System (Division 1) 3 2.2.24.2 250 Volt DC Electrical System (Division 2) 5 2.2.25A 125 Volt DC Power Checklist 4 2.2.25.1 125 Volt DC Electrical System (Division 1) 3 2.2.25.2 125 Volt DC Electrical System (Division 2) 3 2.2.26 24 Volt DC Electrical System 18 2.2.26A 24 Volt DC Power Checklist 1 2.2.60 Primary Containment Cooling and Nitrogen Inerting System 69C1 2.2.60A Primary Containment Cooling and Nitrogen Inerting System 18 Component Checklist

-8-Number Title Revision 2.2.60B Primary Containment Cooling and Nitrogen Inerting System 1 Instrument Valve Checklist 2.2.60.1 Containment H2/O2 Monitoring Systems 14 2.2.61 Primary Coolant Leakage Detection System 23 2.2.61A Primary Coolant Leakage Detection System Component Checklist 8 2.2.63 Plant Management Information System Uninteruptible Power 9 Supply System 2.2.63A Plant Management Information System Uninteruptible Power 2 Supply Component Checklist 2.2.65.1 Reactor Equipment Cooling Operations 35 2.4FPC Fuel Pool Cooling Trouble 0 3.4 Configuration Change Control 34 3.4.4 Temporary Configuration Change 1 4.6.3 Reactor Vessel Top Head Flange Leak Detection 16 5.3DC125 Loss of 125 Volt DC 4 5.8.2 Alternate Emergency Depressurization Systems (Table 2) 16 5.8.7 Primary Containment Flooding/Spray Systems 13 5.8.17 Primary Containment Venting 6 5.8.18 Primary Containment Venting for Primary Containment Pressure 10 Limit, Pressure Suppression Pressure, or Primary Containment Flooding 5.8.21 Primary Containment Venting and Hydrogen Control (Less than 8 Combustible Limits)

5.8.22 Primary Containment Venting and Hydrogen Control (Greater than 10 Combustible Limits)

5.9H2O2 Primary Containment Combustible Gas Control (Severe Accident 1 Guideline Number 3)

6.PC.207 Torus to Drywell Vacuum Breaker Operation, Revision 4 6.SC.602 Reactor Building Roof Access 3 7.0.7 Scaffolding Construction and Control 15

-9-Number Title Revision 7.2.26.2 Bolted of Screwed Bonnet Check Valve Disassembly, Inspection, 8 and Reassembly 13.17 Residual Heat Removal Heat Exchanger Performance Testing 15 15.OG.601 Off Gas Loop Seal Blowdown and Fill 3 PROCEDURE CHANGE REQUESTS Number Title Dated 2.2.24A 250 Volt DC Power Checklist June 19, 2000 2.2.24.1 250 Volt DC Electrical System (Division 1) February 21, 2000 2.2.24.1 250 Volt DC Electrical System (Division 1) March 28, 2000 2.2.24.1 250 Volt DC Electrical System (Division 1) August 6, 2001 2.2.24.1 250 Volt DC Electrical System (Division 1) December 2, 2001 2.2.24.1 250 Volt DC Electrical System (Division 1) April 2, 2002 2.2.24.2 250 Volt DC Electrical System (Division 2) February 21, 2000 2.2.24.2 250 Volt DC Electrical System (Division 2) March 28, 2000 2.2.24.2 250 Volt DC Electrical System (Division 2) June 12, 2001 2.2.24.2 250 Volt DC Electrical System (Division 2) August 6, 2001 2.2.24.2 250 Volt DC Electrical System (Division 2) December 2, 2001 2.2.24.2 250 Volt DC Electrical System (Division 2) April 2, 2002 2.2.24.2 250 Volt DC Electrical System (Division 2) March 12, 2003 2.2.25A 125 Volt DC Power Checklist June 26, 2000 2.2.25A 125 Volt DC Power Checklist February 1, 2001 2.2.25A 125 Volt DC Power Checklist August 27, 2001 2.2.25A 125 Volt DC Power Checklist February 11, 2002 2.2.25.1 125 Volt DC Electrical System (Division 1) February 21, 2000 2.2.25.1 125 Volt DC Electrical System (Division 1) December 2, 2001 2.2.25.1 125 Volt DC Electrical System (Division 1) February 11, 2002 2.2.25.1 125 Volt DC Electrical System (Division 1) April 2, 2002

-10-Number Title Dated 2.2.24A 250 Volt DC Power Checklist June 19, 2000 2.2.24.1 250 Volt DC Electrical System (Division 1) February 21, 2000 2.2.25.1 125 Volt DC Electrical System (Division 1) October 11, 2002 2.2.25.2 125 Volt DC Electrical System (Division 2) February 21, 2000 2.2.25.2 125 Volt DC Electrical System (Division 2) June 12, 2001 2.2.25.2 125 Volt DC Electrical System (Division 2) December 2, 2001 2.2.25.2 125 Volt DC Electrical System (Division 2) April 2, 2002 2.2.25.2 125 Volt DC Electrical System (Division 2) October 11, 2002 2.2.26 24 Volt DC Electrical System (Division 2) June 19, 2000 2.2.26 24 Volt DC Electrical System (Division 2) October 30, 2000 2.2.26 24 Volt DC Electrical System (Division 2) October 25, 2002 2.2.26 24 Volt DC Electrical System (Division 2) November 1, 2002 2.2.26 24 Volt DC Electrical System (Division 2) November 15, 2002 2.2.26A 24 Volt DC Power Checklist June 19, 2000 2.2.60 Primary Containment Cooling and Nitrogen Inerting May 2, 2000 System 2.2.60 Primary Containment Cooling and Nitrogen Inerting September 13, 2000 System 2.2.60 Primary Containment Cooling and Nitrogen Inerting October 19, 2000 System 2.2.60 Primary Containment Cooling and Nitrogen Inerting March 8, 2001 System 2.2.60 Primary Containment Cooling and Nitrogen Inerting May 10, 2001 System 2.2.60 Primary Containment Cooling and Nitrogen Inerting November 16, 2001 System 2.2.60 Primary Containment Cooling and Nitrogen Inerting May 8, 2002 System 2.2.60 Primary Containment Cooling and Nitrogen Inerting May 27, 2002 System

-11-Number Title Dated 2.2.24A 250 Volt DC Power Checklist June 19, 2000 2.2.24.1 250 Volt DC Electrical System (Division 1) February 21, 2000 2.2.60A Primary Containment Cooling and Nitrogen Inerting May 1, 2001 System Component Checklist 2.2.60A Primary Containment Cooling and Nitrogen Inerting February 12, 2003 System Component Checklist 2.2.60B Primary Containment Cooling and Nitrogen Inerting June 26, 2000 System Instrument Valve Checklist 2.2.60.1 Containment H2/O2 Monitoring Systems September 11, 2000 2.2.60.1 Containment H2/O2 Monitoring Systems February 7, 2001 2.2.60.1 Containment H2/O2 Monitoring Systems July 6, 2001 2.2.60.1 Containment H2/O2 Monitoring Systems April 6, 2002 2.2.60.1 Containment H2/O2 Monitoring Systems May 7, 2002 2.2.61 Primary Coolant Leakage Detection System April 12, 2000 2.2.61 Primary Coolant Leakage Detection System March 8, 2001 2.2.61A Primary Coolant Leakage Detection System Component June 26, 2000 Checklist 2.2.61A Primary Coolant Leakage Detection System Component May 1, 2001 Checklist 2.2.61A Primary Coolant Leakage Detection System Component February 27, 2003 Checklist 2.2.61A Primary Coolant Leakage Detection System Component March 9, 2003 Checklist 2.2.63 Plant Management Information System Uninteruptible August 6, 2001 Power Supply System 2.2.63 Plant Management Information System Uninteruptible October 22, 2002 Power Supply System 2.2.63 Plant Management Information System Uninteruptible November 27, 2002 Power Supply System 2.2.63A Plant Management Information System Uninteruptible July 19, 2001 Power Supply System Component Checklist 2.4PC Primary Containment Control March 8, 2001

-12-Number Title Dated 2.2.24A 250 Volt DC Power Checklist June 19, 2000 2.2.24.1 250 Volt DC Electrical System (Division 1) February 21, 2000 2.4PC Primary Containment Control April 30, 2001 2.4PC Primary Containment Control July 12, 2001 2.4PC Primary Containment Control December 16, 2002 2.4PC Primary Containment Control April 10, 2003 5.3DC125 Loss of 125 Volt DC October 18, 2001 5.3DC125 Loss of 125 Volt DC June 5, 2002 5.3DC125 Loss of 125 Volt DC November 14, 2002 5.3DC125 Loss of 125 Volt DC December 12, 2002 5.3DC125 Loss of 125 Volt DC April 2, 2003 5.3DC125 Loss of 125 Volt DC April 10, 2003 5.8.7 Primary Containment Flooding/Spray Systems April 19, 2000 5.8.7 Primary Containment Flooding/Spray Systems August 17, 2000 5.8.7 Primary Containment Flooding/Spray Systems October 5, 2000 5.8.7 Primary Containment Flooding/Spray Systems May 8, 2001 5.8.7 Primary Containment Flooding/Spray Systems April 2, 2003 5.8.17 Primary Containment Venting July 6, 2001 5.8.18 Primary Containment Venting for Primary Containment April 14, 2000 Pressure Limit, Pressure Suppression Pressure, or Primary Containment Flooding 5.8.18 Primary Containment Venting for Primary Containment April 17, 2001 Pressure Limit, Pressure Suppression Pressure, or Primary Containment Flooding 5.8.18 Primary Containment Venting for Primary Containment November 7, 2001 Pressure Limit, Pressure Suppression Pressure, or Primary Containment Flooding 5.8.21 Primary Containment Venting and Hydrogen Control October 30, 2000 (Less than Combustible Limits)

5.8.22 Primary Containment Venting and Hydrogen Control October 30, 2000 (Greater than Combustible Limits)

-13-Number Title Dated 2.2.24A 250 Volt DC Power Checklist June 19, 2000 2.2.24.1 250 Volt DC Electrical System (Division 1) February 21, 2000 5.8.22 Primary Containment Venting and Hydrogen Control November 7, 2001 (Greater than Combustible Limits)

5.9H2O2 Primary Containment Combustible Gas Control (Severe March 22, 2000 Accident Guideline Number 3)

OPERABILITY EVALUATIONS Number Title Date or Revision 93-000-047 Fire Door Evaluation November 17, 1993 4-00683 Containment Structure and Spray Valves Revision 00 4-08041 Containment Structure and Spray Valves Revision 00 4-08332 Containment Structure and Spray Valves Revision 00 4-09592 Residual Heat Removal-PS-119A-D May 26, 2000 4-10631 Primary Containment Oxygen Concentration July 31, 2000 4-12618 Standby gas treatment-AOV-270AV November 17, 2000 4-12745 125/250 Volt DC Batteries November 27, 2000 4-13618 125 Volt DC Batteries March 1, 2001 4-13806 Torus Shell September 17, 2001 SELF-ASSESSMENTS SA-01-005, ASME Section XI Inservice Testing (IST) Program SA-01-006, Motor Operated Valve (MOV) Program SA-02-012, Design Modifications SA-02-026, Appendix J Program

-14-SETPOINT CHANGES82-017, Pump Around Discharge Overpressure Protection Sutorbuilt Blower 1B 82-018, Pump Around Discharge Overpressure Protection Sutorbuilt Blower 1A 84-03, Torus Water Temp. Recorder 85-11, Drywell Hi/Lo Pressure Alarm 85-12, Suppression Chamber Pressure Recorder 88-38, Primary Containment Hydrogen and Oxygen Monitor - Oxygen Alarm 88-39, Primary Containment Hydrogen and Oxygen Monitor - Oxygen Alarm 88-59, Primary Containment Hydrogen and Oxygen Monitor - Hydrogen Alarm 89-05, Drywell Zone 2C Temperature Alarm 90-07, Drywell Temperature Zone 1 Recorder and Annunciator 90-08, Drywell Temperature Zone 1 Recorder and Annunciator 92-057, Torus Hard Pipe Vent Pressure Switch 92-065, Suppression Chamber Reactor Building Vacuum Breakers93-026, Suppression Chamber Hi/Lo Alarm - Narrow Range 94-23, Primary Containment High Pressure 98-01, Suppression Chamber Hi/Lo Alarm - Narrow Range 98-10, Drywell High Pressure - Reactor Scram, Groups 2 and 6 Isolation & Emergency Diesel Generator Start 98-11, Drywell High Pressure - High Pressure Coolant Injection, Core Spray, and Residual Heat Removal (Low Pressure Coolant Injection Mode) Initiation, N/A 98-12, Drywell High Pressure - Reactor Scram, Groups 2 and 6 Isolation & Emergency Diesel Generator Start, N/A 2000-016, Containment Spray Drywell Permissive Interlock Pressure Switches, N/A

-15-TEMPORARY CONFIGURATION CHANGES 4301609, "Locked Open Device for Mechanical Overspeed Trip Butterfly Valve on Diesel Generator No. 1," Revision 1 4287004, Temporary Repair of Equipment Drain Line Leak, Revision 0 TESTING REPORTS Procedure Title Dated 6.PC.205 Instrument Line Excess Flow Check Valve Test April 9, 2003 6.PC.302 Calibration Test Results July 24, 2002 6.1SGT.401 Standby Gas Treatment A Fan Capacity Test, Standby December 1, Gas Treatment B Cooling Flow Test and Check Valve 2001 IST (Division 1)

6.1SGT.401 Standby Gas Treatment A Fan Capacity Test, Standby March 7, Gas Treatment B Cooling Flow Test and Check Valve 2003 IST (Division 1)

6.2EE.305 Distribution System Breaker Alignment (Division 1) March 2, 2003 6.2EE.305 Distribution System Breaker Alignment (Division 1) April 6, 2003 6.2EE.305 Distribution System Breaker Alignment (Division 2) December 7, 2002 6.2EE.305 Distribution System Breaker Alignment (Division 2) January 5, 2003 6.2EE.305 Distribution System Breaker Alignment (Division 2) February 2, 2003 6.2EE.305 Distribution System Breaker Alignment (Division 2) March 2, 2003 6.2EE.305 Distribution System Breaker Alignment (Division 2) April 6, 2003 6.2EE.601 125Volt /250Volt Station and Diesel Fire Pump Battery 7 March 2, Day Check 2003 6.2EE.602 125Volt /250Volt Station and Diesel Fire Pump Battery March 21, 92 Day Check 2003 6.2EE.603 125Volt Battery Service Test April 21, 2003 6.2EE.604 125Volt Battery Charger Performance Test August 7, 2002

-16-Procedure Title Dated 6.2EE.605 250Volt Battery Service Test March 22, 2003 6.2EE.606 250Volt Battery Charger Performance Test April 21, 2003 4/21/03 6.2EE.607 125Volt Station Battery Performance Discharge Test December 14, 2001 6.2EE.608 250Volt Station Battery Performance Discharge Test December 8, 2001 6.2EE.609 125Volt /250Volt Station Battery. Intercell Connection March 12, Testing 2003 6.2EE.609 125Volt /250Volt Station Battery. Intercell Connection March 15, Testing 2003 6.2EE.610 Off-Site AC Power Alignment March 2, 2003 6.2EE.610 Off-Site AC Power Alignment April 6, 2003 6.2EE.611 125V/250V Battery Cell and Rack Examination July 25. 2002 6.2EE.611 125V/250V Battery Cell and Rack Examination October 23, 2002 6.2EE.611 125V/250V Battery Cell and Rack Examination January 15, 2003 6.2SGT.401 Standby Gas Treatment B Fan Capacity Test, Standby December 1, Gas Treatment A Cooling Flow Test and Check Valve 2001 IST (Division 2)

6.2SGT.401 Standby Gas Treatment B Fan Capacity Test, Standby March 6, Gas Treatment A Cooling Flow Test and Check Valve 2003 IST (Division 2)

TRAINING MANUALS Number Title Revision COR002-03-02 OPS Containment 16 COR002-03-02 OPS DC Electrical Distribution 21 COR002-03-03 Containment 5 COR002-07-02 OPS DC Electrical Distribution 21

-17-COR002-07-03 DC Distribution 5 WORK ORDERS 00-1125 4175512 4261312 00-1264 4230577 4263826 00-1265 4235345 4274285 00-1266 4235345 4278260 00-1267 4235346 4299173 00-2031 4235346 10 CFR 50.59 EVALUATIONS 1993-024, Diesel Generator Upgrades, Revision 0 1993-0050, Appendix J Testing in the Accident Direction, dated August 25, 1995 2001-0013, DC 91-121A, Installation of 69kV Capacitor Bank, Revision 0 2001-0017, Controllers Modification CED 2001-0017, Revision 2 2001-0044, EE01-035 Equipment Qualification Temperature Profile in Containment based on Small Steam Line Break and Design Basis Accident-Loss of Coolant Accident, Revision 0 2002-0008, Implementation of NLS2001064 (Amendment 192) via EE 01-023, Revision 0 2003-0006, TCC4301609 - Lock Open Device for Mechanical Overspeed Butterfly Valve on DG No.1, Revision 0 10 CFR 50.59 SCREENINGS Revision of Calculation NEDC 02-026 for Issue as a Status 1 Document.

CED 1998-0179, Upgrade of Air Supply to Control Valves HV-SOV-(SPV-259) and HV-SOV-(SPV-261), dated December 5, 1998 CED 2000-0098, Installation of Washers on Torus to Drywell Vacuum Breakers, dated April 11, 2000 CED 6008504, Replacement Evaluation for AS-V-112 and AS-V-209, dated December 3, 2002

-18-Engineering Evaluation 01-080, Effect of Loss of Coolant Accident & Small Steam Line Break Accident Conditions on Drywell Fan Cooling Units and Containment Penetration Piping for Closure of Generic Letter 96-06, Revision 0 Engineering Evaluation 02-014, Evaluation for the Use of either Service Water Pump Discharge or River Well Pump Discharge as the Normal Supply for the Gland Water System for the Service Water Pumps, Revision 0 Engineering Evaluation EE 02-047, Permanent Change Documenting Belzona Coating of Specific Internal Areas of Service Water Pump Intermediate and Lower Columns, Revision 0 Engineering Evaluation 02-070, Develop an Engineering Evaluation to Change the High Cooling Water Temperature Alarm Ronan Setpoint from 140 degrees to 250 degrees F, Revision 0 Engineering Evaluation 02-079, Effect of Removal and Replacement of Snubber, Revision 0 MP 97-039, Thermal Overpressure Protection X-18, X-19, and X-20, April 23, 1997 Procedure 13.15.1, Reactor Equipment Cooling Heat Exchanger Performance Analysis, Revision 19 Procedure 2.2.69, Residual Heat Removal System, Revision 58 Procedure 13.17, Residual Heat Removal Heat Exchanger Performance Testing, Revision 9 Temporary Configuration Change 4287004 MISCELLANEOUS DOCUMENTS Updated Safety Analysis Report, dated February 28, 2003 Technical Specifications, Amendment 198 Critical AC Bus Coordination Study, dated May 1994 List of Environmentally Qualified Components in Primary Containment and Electrical DC Master Equipment List (EQ) for Primary Containment Components, Revision 20.

Report of the Fire Endurance and Hose Stream Testing of Two Single, Fire Rated Door Assemblies with Excessive Clearances Installed in a Concrete Block Wall Cooper Nuclear Station Plant Engineering Department System Health Report, dated March 2003 NPPD Letter NLS20022122, 10CFR50.59(d)(2) Summary Report Cooper Nuclear Station NRC Docket No. 50-298, DPR-46, dated October 17, 2002

-19-NPPD Letter CNSS877236,from NPPD to USNRC Regarding NPPD Response to Generic Letter 87-05, dated May 12, 1987 NPPD Memo from Radloff to Fleming Regarding Evaluation of Commitments Associated with Generic Letter 87-05, dated March 20, 2003 White Paper, Current Condition of the 250 Volt DC Station Batteries Related to the Five Under Charged Cells, dated April 11, 2003 Facsimile, 2001 from Flowserve Co. to NPPD Regarding 20 inch Series 300 y-Globe Lift Check Valves, dated August 30 C&D Technical Manual For LCR And LCY Batteries Purchase Order 4500014228 5.1.3.1, TIP Action Plan - External Regulatory Communications, Revision 2a 5.3.3.1, TIP Action Plan - Design Basis Information/Licensing Basis Information (DBI/LBI)

Translation Project, Revision 2 License Amendment No. 187, Cooper Nuclear Station - Issuance of Amendment Regarding Revised Radiological Dose Assessment and Technical Specification Changes (TAC NO.

MB1419), dated October 23, 2001 License Amendment No. 189, Cooper Nuclear Station - Issuance of Amendment to revise the Technical Specifications Surveillance Test Requirement SR 3.6.1.3.8, for Excess Flow Check Valves (EFCVs) (TAC NO. MB1820), dated October 26, 2001 License Amendment No. 192, Cooper Nuclear Station - Issuance of Amendment Re:

Containment Overpressure to Ensure Sufficient Net Positive Suction Head (NPSH) for the Emergency Core Cooling System (ECCS) Pumps Following a Loss-of-Coolant Accident (LOCA) (TAC NO. MB 2896), dated July 19, 2002 Reactor Equipment Cooling Containment Isolation Requirements Position Paper, dated November 29, 1994 Third Interval Inservice Testing Program, Revision 4 NCR 94-057, Nonconformance Report 94-057, dated May 5, 1994 NSL2001064, NPPD Letter - Proposed Licensing Amendment, dated July 30, 2001 PIR 4-07997, During Work on Standby Gas Treatment-AOV-AO270 it was Determined that the Valve Opened 78.38 Degrees, Versus the Required 90 Degrees, dated April 2, 2000 PIR 4-08332, Initiation of Drywell Spray and then the Subsequent Return to Primary Containment Condition for the Drywell Spray Valves May Not Be a Function They Can Perform, dated April 11, 2000

-20-EE-DC System Health Indicators, dated April 23, 2003 Relief Requests, Relief Requests from Inservice Inspection Requirements, dated May 19, 1983 Relief Request RV-10, Relief Request for the Exercise Frequency for Excess Flow Check Valves in the Pump and Valve Testing Program (TAC NO. MB1820), dated October 26, 2001 Safety Evaluation Report 9.3.2, Reactor Building Closed Cooling Water System (RBCCWS)