IR 05000313/2003007

IR 05000313-03-007 and IR 05000368-03-007 on 07/14/2003 - 08/01/2003; Arkansas Nuclear One, Units 1 and 2; Plant Design Pilot, Enclosures 1 and 3

ML032240044
Person / Time
Site:	Arkansas Nuclear
Issue date:	08/11/2003
From:	Marschall C Division of Nuclear Materials Safety IV
To:	Anderson C Entergy Operations
References
IR-03-007
Download: ML032240044 (23)

Text

ust 11, 2003

SUBJECT:

ARKANSAS NUCLEAR ONE, UNITS 1 and 2 - NRC INSPECTION REPORT 05000313/2003007; 05000368/2003007

Dear Mr. Anderson:

On August 1, 2003, the NRC completed an inspection at your Arkansas Nuclear One, Units 1 and 2, facility. The enclosed report documents the inspection findings, which were discussed on August 1, 2003, with you and other members of your staff.

This inspection examined activities conducted under your licenses 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). The finding did not present an immediate safety concern. Because of the very low significance and because you entered it into your corrective action program, the NRC is treating it as a noncited violation, consistent with Section VI.A of the Enforcement Policy.

The noncited violation is described in the subject inspection report. 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 Arkansas Nuclear One, Units 1 and 2.

Entergy Operations, Inc. -2-In accordance with 10 CFR 2.790 of the NRCs "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/

Charles S. Marschall, Chief Engineering and Maintenance Branch Division of Reactor Safety Dockets: 50-313; 50-368 Licenses: DPR-51; NPF-6

Enclosure:

NRC Inspection Report 05000313/2003007; 05000368/2003007

REGION IV==

Dockets: 50-313, 50-368 Licenses: DPR-51, NPF-6 Report : 05000313/2003007, 05000368/2003007 Licensee: Entergy Operations, Inc.

Facility: Arkansas Nuclear One, Units 1 and 2 Location: Junction of Hwy. 64W and Hwy. 333 South Russellville, Arkansas Dates: July 14 through August 1, 2003 Team Leader: C. Paulk, Senior Reactor Inspector, Engineering and Maintenance Branch Inspectors: P. Goldberg, Reactor Inspector, Engineering and Maintenance Branch J. Mateychick, Reactor Inspector, Engineering and Maintenance Branch W. McNeill, Reactor Inspector, Engineering and Maintenance Branch G. Miller, Reactor Inspector, Engineering and Maintenance Branch Accompanying C. Baron, Contractor, Beckman and Associates Persons: S. Meyers, Engineering Associate Approved By: Charles S. Marschall, Chief Engineering and Maintenance Branch Division of Reactor Safety

-2-SUMMARY OF FINDINGS IR 05000313/2003007, 05000368/2003007; 07/14/2003 - 08/01/2003; Arkansas Nuclear One, Units 1 and 2; Plant Design Pilot, Enclosures 1 and 3 The NRC conducted an inspection with five regional inspectors and one contractor. The inspection identified one green noncited violation. The significance of most findings is indicated by their 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 NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3, dated July 2000.

Cornerstone: Barrier Integrity

• Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the inspectors identified four examples of failures to correctly translate the design basis into specifications, procedures, and instructions. The inspectors considered the barrier integrity cornerstone affected because of the potential of containment and engineered safety features integrity being degraded by these conditions.

The inspectors considered this finding greater than minor because it paralleled Example 3.i of Appendix E to Inspection Manual Chapter 0612. The licensee's engineering staff had to perform reanalyses and operability evaluations due to these conditions. The inspectors considered this finding of very low safety significance because it did not represent an actual loss-of-safety function (Section 1RDS1.5).

Report Details 1. REACTOR SAFETY Introduction The NRC has undertaken a pilot inspection program to determine if efficiencies or resource savings can be gained by consolidating selected baseline inspection procedures. This inspection report documents the performance of Attachment 71111.DS, "Plant Design - Pilot," Enclosures 1 and 3. These enclosures are normally performed using Attachments 71111.21, "Safety System Design and Performance Capability," and 71111.02, "Evaluation of Changes, Tests, or Experiments."

The NRC conducted an inspection to verify the adequacy of the original design and subsequent modifications to safety systems and to monitor the capability of the selected systems to perform their design basis functions. The inspection was also conducted to monitor the effectiveness of the licensees implementation of changes to facility structures, systems, and components, risk-significant normal and emergency operating procedures; test programs; and the updated final safety analysis reports in accordance with 10 CFR 50.59, "Changes, Tests, and Experiments."

The team reviewed in detail the containment structures. The primary review prompted parallel review and examination of support systems, such as, high and low pressure injection systems; building spray systems; penetrations; electrical power; air supplies; instrumentation; and related structures and components.

The team assessed the adequacy of calculations, analyses, engineering processes, and engineering and operating practices that the licensee used for the selected safety system 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.

1RDS Plant Design (71111.DS)

1RDS1Enclosure 1: Safety System Design and Performance

.1 System Requirements a. Inspection Scope The team inspected the following attributes of the reactor containment structures:

(1) process medium (water, steam, and air), (2) energy sources, (3) control systems, and (4) equipment protection. The team examined the procedural instructions to verify instructions as consistent with actions required to meet, prevent, and/or mitigate design basis accidents. The team also considered requirements and commitments identified in

-2-the Updated Safety Analysis Report, technical specifications, design basis documents, and plant drawings.

The reviews also include support systems required for the containment structures to perform their mitigating function. These systems included high and low pressure injection systems, building spray systems, hydrogen control systems, and penetrations.

b. Findings No findings of significance were identified.

.2 System Condition and Capability a. Inspection Scope The team reviewed the periodic testing procedures for the containment and support systems to verify that the capabilities of the systems were verified periodically. The team also reviewed the systems operations by conducting system walkdowns; reviewing normal, abnormal, and emergency operating procedures; and reviewing the Updated Final Safety Analysis Reports, technical specifications, design calculations, drawings, and procedures.

b. Findings No findings of significance were identified.

.3 Identification and Resolution of Problems a. Inspection Scope The team reviewed 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 samples included open and closed condition reports for the past 3 years and are listed in the attachment to this report. Inspection Procedure 71152,

"Identification and Resolution of Problems," was used as guidance to perform this part of the inspection. Older condition reports, identified while performing other areas of the inspection, were also reviewed.

b. Issues and Findings No findings of significance were identified.

.4 System Walkdowns a. Inspection Scope The team performed walkdowns of the accessible portions of the containment structures and support systems. During the walkdowns, the team assessed:

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• 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 materiel 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.

.5 Design Review a. Inspection Scope The team reviewed the current as-built instrument and control, electrical, and mechanical design of the containment structures and support systems. These reviews included an examination of design assumptions, calculations, required system thermal-hydraulic performance, electrical power system performance, control logic, and instrument setpoints and uncertainties. The team also performed selected single-failure evaluations of individual components and circuits to determine the effects of such failures on the capability of the system to perform its design safety functions. The team also reviewed the licensees calculations and methodology for ensuring the component cooling water system was protected against seismic, flooding, fire, and high energy line break events.

The team reviewed calculations, drawings, specifications, vendor documents, Final Safety Analysis Report, technical specifications, emergency operating procedures, and temporary and permanent modifications.

b. Findings Introduction The team 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 four examples of failures to correctly translate the design basis into specifications, procedures, and instructions.

Description Arkansas Nuclear One, Unit 1, Calculation 97-E-0212-01, "BWST Draindown Analysis,"

Revision 2, addressed the flowrate from the borated water storage tank during the post-

-4-accident transfer of the engineered safety features pumps from the tank to the containment sump. The inspectors noted that the calculation did not consider the potential single active failure of one of the borated water storage tank outlet valves to close, potentially allowing air to reach the suction of the engineered safety features pumps. In response, the Arkansas Nuclear One staff initiated Condition Reports CR-ANO-1-2003-00755 on July 16, 2003 and CR-ANO-1-2003-00769 on July 18, 2003. Condition Report CR-ANO-1-2003-00755 included an operability evaluation, and stated that air would not reach the suctions of the engineered safety features pumps.

Arkansas Nuclear One, Unit 2, Calculation 98-E-0044-01, "RWT Draindown Analysis,"

Revision 2, addressed the flow rate from the refueling water tank during the post-accident transfer of the engineered safety features pumps from the tank to the containment sump (similar to the Unit 1 calculation above). The calculation determined that the water level remaining in the tank at the completion of the transfer would be adequate to prevent air entrainment in the system. However, the team noted that the calculation did not consider the potential single active failure of one of the refueling water tank outlet valves to close. In response, the licensees staff initiated Condition Report CR-ANO-2-2003-00977 on July 15, 2003. This condition report included an operability evaluation, and stated that air could potentially enter the piping, but would not reach the suctions of the engineered safety features pumps.

Arkansas Nuclear One, Unit 1, Updated Safety Analysis Report, Section 14.2.2.6.6, stated that the decay heat vaults (containing engineered safety features pumps) were sealed rooms. The radiological analyses did not consider leakage from equipment in these rooms. However, the inspectors noted that the leakage acceptance criteria for the closed room drain valves (ABS-13/14) was 0.43 gpm. In response, the Arkansas Nuclear One staff initiated Condition Report CR-ANO-1-2003-00761 on July 17, 2003.

This condition report included an operability evaluation in which licensee engineers concluded that there was negligible effect on the off-site dose. Also, this condition report included an action to develop a leakage acceptance criterion that is consistent with the licensing/design basis of Arkansas Nuclear One, Unit 1.

Calculation 88-E-0100-33, "ANO U1 Spent Fuel Cooling System P/T Calculation," did not consider the maximum pressure in a section of the spent fuel cooling system due to potential leakage from the decay heat system through Check Valve SF-21. The inspectors questioned the boundary between these systems. In response, the Arkansas Nuclear One staff initiated Condition Report CR-ANO-1-2003-00814 on July 29, 2003.

This condition report included an operability evaluation in which licensee engineers concluded that there was no immediate operability concern. As a result of the operability evaluation, the licensee engineers included actions to update Calculation 88-E-0100-33 and corresponding pipe stress calculations.

Analysis The team considered the barrier integrity cornerstone affected because of the potential of containment and engineered safety features integrity being degraded by these conditions. The team considered this finding more than minor since the findings fit with

-5-Example 3.i of Appendix E of Manual Chapter 612. The licensees engineering staff had to perform reanalyses and operability evaluations due to these conditions.

The team found these issues resulted from a performance deficiency of very low safety significance. The team determined no other cornerstones were degraded as a result of this finding.

The team assessed this finding as green because it does not represent an actual loss of the containment or engineered safety features safety functions. The specific accident conditions that could have challenged the systems have not existed. The licensee has implemented appropriate corrective actions to ensure continued operability.

Enforcement Criterion III of 10 CFR Part 50, Appendix B, Design Control, states, in part, that measures shall be established to assure that the design basis is correctly translated into specifications, procedures, and instructions. Contrary to the Appendix B, Arkansas Nuclear One engineering did not correctly translate the design basis into these design documents. As a result, the subject analyses and test criteria were non-conservative.

After the identification of these issues by the inspectors, the licensee implemented appropriate corrective actions. The Arkansas Nuclear One staff initiated condition reports and entered this finding into its corrective action program.

Because of the very low safety significance of the finding and because the licensee has entered these issues into their corrective action program, the inspectors treated this as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000313/2003007-001; 05000368/2003007-001).

.6 Safety System Inspection and Testing a. Inspection Scope The team reviewed the program and procedures for testing and inspecting selected components for the containment structures and support systems. The review included the results of surveillance tests required by the technical specifications and selective review of in-service tests.

b. Findings No findings of significance were identified.

1RDS3Enclosure 3: Changes, Tests, or Experiments a. Inspection Scope The team reviewed the licensees procedures for performing evaluations and screenings for changes to the facility, procedures and tests in accordance with 10 CFR 50.59. The team reviewed samples of plant modifications, operating procedures, test procedures, and plant analysis methods.

-6-The team reviewed 7 evaluations to verify that the licensee personnel 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. The team reviewed 13 screenings, in which the licensee personnel determined that evaluations were not required, to ensure that the exclusion of a full evaluation was consistent with the requirements of 10 CFR 50.59.

b. Findings No findings of significance were identified.

4. OTHER ACTIVITIES (OA)

4OA6 Management Meetings Exit Meeting Summary On August 1, 2003, the team leader presented the inspection results to Mr. C. G.

Anderson, Vice President Operations, and other members of licensee management and staff who acknowledged the findings. The team leader confirmed that no proprietary information was provided or examined during the inspection.

ATTACHMENT PARTIAL LIST OF PERSONS CONTACTED Licensee:

C. Anderson, Vice President Operations G. Ashley, Manager Licensing D. Bice, Licensing Specialist IV E. Blackard, Design Engineering Supervisor M. Byram, Senior Lead Engineer H. Chadbourn, Supervisor Engineering M. Chism, System Engineering Manager M. Cooper, Licensing Specialist R. Cooper, Supervisor Control Room W. Cottingham, Senior Staff Engineer J. Cotton, Senior Engineer R. Cuilty, Senior Operations Specialist C. Eubanks, General Manager D. Fouts, Supervisor Engineering M. Fuller, Senior Engineer W. Greeson, Supervisor Engineering J. Hale, Senior Engineer D. Hawkins, Licensing Specialist II W. Hinton, Senior Engineer P. Kearney, Technical Assistant II J. Kowalewski, Design Engineering Director D. MacPhee, Senior Staff Engineer S. McKissack, Senior Lead Engineer R. McWilliams, Senior Engineer D. Phillips, Supervisor Engineering J. Richardson, Senior Engineer W. Rowlett, Senior Lead Engineer S. Smith, Senior Engineer M. Smith, Engineer R. To, Senior Engineer R. Wilson, Senior Staff Engineer C. Zimmerman, Plant Manager, Support LIST OF ITEMS OPENED AND CLOSED Opened and Closed 05000313/2003007-001 NCV Failure to correctly translate a design basis into calculations (Section 1RDS1.5). 05000368/2003007-001 NCV Failure to correctly translate a design basis into calculations (1RDS1.5).

-2-DOCUMENTS REVIEWED Calculations NUMBER DESCRIPTION REVISION 00-E-0035-01 Allowable Leakage in the DHR Vaults 0 01-EQ-1001-01 MFW Critical Crack HELB Analysis 0 2CCB-48-3-H009 Evaluation of Pipe Support 2CCB-48-3-H009 for 0 Additional Pipe Loads due to the Addition of Relief Valve Piping 2HCD-1-H2 Evaluation of Pipe Support 2HCD-1-H2 for Additional 3 Loads due to the Addition of Relief Valve Piping 2HCB-5-H4 Evaluation of Pipe Support 2HCB-5-H4 for Additional 0 Pipe Loads due to the Addition of Relief Valve Piping 83-D-1153-01 Error and Setpoint Analyses for BWST Instrumentations 4 Loops 86-D-1105-37 Letdown Piping Analysis for MOVAT Modification of 5 Valves CV-1213 and CV-1215 86-D-1106-39 ASME Class 1 Stress Report - Letdown System 3 86-EQ-0002-06 Loop Accuracy Analysis for Shutdown Cooling and Low 4 Pressure Safety Injection Flow 87-E-0085 Combustion Engineering Calculation for Reduction of 0 HPSI Differential Pressure Requirement 88-E-0098-20 Arkansas Nuclear One - 1 Design Basis Accident 1 Reanalysis 88-E-0144-01 ANO-2 EDG Loading for Buses 2A3 and 2A4 4 88-E-0200-07 Pressure Temperature Calculation for Unit 2 High 1 Pressure Safety Injection System 89-E-0010-26 LPI Pump NPSH 5

-3-Calculations NUMBER DESCRIPTION REVISION 89-E-0010-28 P-34A/B and P-35A/B Net Positive Suction Head from 0 BWST 89-E-0018-06 DHR Heat Exchanger Performance 2 89-E-0164-06 Spray Lambdas and LOCA Radiation Doses Offsite with 1 Reduced Spray Flows 89-E-0164-08 Maximum and Minimum Spray and Sump pH 0 90-D-1043-02 Hazards Calc for RG 1.97 Upgrade of LPI and RBS Flow 1 indication Loops 90-D-2015-06 Sizing Calculation for LPSI Runout Orifice 2FO-5090 0 90-D-2017-11 Backside Anchor Loading for 2DCD-5-H5 and 2DCD-5- 2 H6 90-E-0041-10 Minimum ANO-1 RB Pressure at Time of Swapover to 0 Recirculation 90-E-0046-01 Arkansas Nuclear One-1 Reactor Building Spray Pump 5 Net Positive Suction Head 90-E-0058-01 Allowed Operator Tolerance Error and Loop Error 1 Analysis 90-E-0100-03 Total Contained Volume of Reactor Water Tank 4 90-E-0116-01 ANO-2 EOP Setpoint Document 11 90-E-0116-04 Evaluation of HPSI Minimum Flow Strategy 0 91-E-0016-53 Qualification of R.C. Pump Seal Return to CVC Header 2 Piping System 91-E-0016-183 Qualification of Line CCA-13-1 & 3/4 on Isos SA-273, 1 SA-233, SA-216 91-E-0019-01 Loop Error Analysis for Sodium Hydroxide Tank T-10 5 Level

-4-Calculations NUMBER DESCRIPTION REVISION 91-E-0021-01 Parametric TORC Model 0 91-E-0035-08 Allowable Initial Containment Conditions Accounting for 3 Instrument Errors 91-E-0116-01 NPSH Calculation for HPSI and RB Spray 4 91-R-1010-02 ANO-1 EOP Setpoint Basis Document 7 91-R-1018-02 ANO-1 EOP Setpoint Basis Document 8 91-R-2013-01 Service Water Performance Testing Methodology 9 92-E-0005-01 Required HPSI System single Pump Flow 0 92-E-0077-01, Hydraulic Model of the Arkansas Nuclear One-1 Reactor 1 Building Spray System 92-E-0077-02 ANO-1 HPI System Pump Performance Requirements 0 92-E-0077-03 ANO-1 LPI System Pump Performance Requirements 0 92-E-0078-08 LPSI Pump NPSH Calculation 0 92-E-0078-09 ANO-2 LPSI Pump Runout Calculation 0 92-E-0079-01 Determination of SW Cooled Room Heat Loads Under 1 PC-2 Various Operating Conditions 92-E-0079-01, Determine Service Water Cooled Room Heat Loads 2 Under Various Operating Conditions 92-R-1017-23 Unit One Setpoint Document Package for the Reactor 3 Building Spray System 92-R-2016-01 ANO-2 Post Accident Operator Doses Performing EOP 2 Local Actions 93-E-0058-01 HPI NPSH from the BWST 0

-5-Calculations NUMBER DESCRIPTION REVISION 93-R-0010-01 Evaluation of Safety Related Power Operated Gate 0 Valves for Thermal Binding and Hydraulic Locking 94-E-0038-01 Code Qualification for R.B. Sump Drain 2HCB-5 to Aux, 2 Bldg. Rad. Waste System 94-E-0095-18 Room 2007/2009 Heat Load Evaluation 1 94-E-0095-19 Room Heat Load Evaluation 1 94-E-0095-20 Room 2013/2014 Heat Load Evaluation 0 97-E-0009-15 Containment Basemat Design Investigation Report 2 97-E-0009-17 Rebar Strength Data for Unit 2 Containment 1 97-E-0045-01 HPSI Pump Suction Pressure Required for Adequate 0 NPSH - AOP Setpoint 97-E-0211-01 BWST Level Analysis 0 97-E-0212-01 BWST Draindown Analysis 2 97-R-0001-01 ECCS Leakage SAR Clarification 1 97-R-1002-01 ECCS Leakage Quantities to the Auxiliary Building 0 97-R-2002-01 ECCS Leakage Quantities to the Auxiliary Building 4 974813D101-01 Setpoint Determination for Thermal Relief Valves 0 Installed as a Result of NRC Generic Letter 96-06 (Mechanical)

974813D101-02 Pressure Design of Piping & Valves as a Result of 0(2)

Overpressure Protection in Response to Generic Letter 96-06 974813D101-03 Evaluation of the Effects of Thermal Relief Vent Line on 0 Small Bore Containment Piping Associated with Penetrations P-9, P-10, and P-12

-6-Calculations NUMBER DESCRIPTION REVISION 974813D101-04 Qualification of Large Bore Piping Systems and Supports 0, 1 Affected by the Addition of Thermal Relief Valves 974813D101-05 Qualification of One (P14) Large Bore Piping System and 0 Supports Affected by the Addition of Thermal Relief Valve 974814D201-01 Evaluation of the Effects of Thermal Relief Vent Lines 0(1)

Associated with Penetrations 2P-51, 2P-59, and 2P-69 98-E-0041-02 Model for Evaluating Injected Volume from BWST (T3) 0 98-E-0044-01 RWT Draindown Analysis 2 99-R-0002-01 Evaluation of High/Low Pressure Interface Valves with 0 Respect to 10CFR50 Appendix R EBD-19-H14 Qualification of Pipe Support EBD-19-H14 2 EBD-19-H75 Qualification of Pipe Support EBD-19-H75 2 G-286-5 Volume of Water in BWST when Suction is Transferred 0 to RB Sump for Recirc.

Condition Reports CR-ANO-1-2000-00011 CR-ANO-2-1991-00557 CR-ANO-2-2003-00640 CR-ANO-1-2001-00084 CR-ANO-2-2000-00245 CR-ANO-2-2003-00977 CR-ANO-1-2001-00208 CR-ANO-2-2000-00270 CR-ANO-2-2003-00991 CR-ANO-1-2001-00350 CR-ANO-2-2000-00511 CR-ANO-2-2003-00992 CR-ANO-1-2001-00486 CR-ANO-2-2000-00585 CR-ANO-2-2003-01044 CR-ANO-1-2002-00101 CR-ANO-2-2000-00622 CR-ANO-2-2003-01053 CR-ANO-1-2002-01268 CR-ANO-2-2000-00624 CR-ANO-C-1991-00103 CR-ANO-1-2002-01342 CR-ANO-2-2000-01059 CR-ANO-C-1996-00135 CR-ANO-1-2003-00626 CR-ANO-2-2001-00277 CR-ANO-C-1996-00210 CR-ANO-1-2003-00755 CR-ANO-2-2001-01027 CR-ANO-C-2001-00183 CR-ANO-1-2003-00760 CR-ANO-2-2001-01114 CR-ANO-C-2002-00101 CR-ANO-1-2003-00761 CR-ANO-2-2001-01384 CR-ANO-C-2003-00558 CR-ANO-1-2003-00764 CR-ANO-2-2002-00245 CR-ANO-C-2003-00565 CR-ANO-1-2003-00765 CR-ANO-2-2002-00779 CR-ANO-C-2003-00568 CR-ANO-1-2003-00769 CR-ANO-2-2002-00933 CR-ANO-C-2003-00576 CR-ANO-1-2003-00811 CR-ANO-2-2002-00978 CR-ANO-C-2003-00612 CR-ANO-1-2003-00814 CR-ANO-2-2002-00993 CR-ECH-2001-00113 CR-ANO-1-2003-00827 CR-ANO-2-2003-00381 CR-ANO-1-2003-00830 CR-ANO-2-2003-00451

-7-Drawings NUMBER DESCRIPTION REVISION 7-DH-106 Large Pipe Isometric Borated Water Storage Tank 0 Level Detection to LT-1421 74-2680, Drw 1 General Plan, Dome Roof Tank 6 74-2680, Drw 10 Pad Details for Bottom Connections 3 74-2680, Drw 20 Vortex Breaker for 24 Outlet Pipe 2 74-3486 General Plan, Dome Roof Tank, Drawing 1 7 C-46 Field Erected Tanks 16 C-46 Field Erected Tanks, Borated Water Storage Tank N Details, Sheet 2 CA-307 Small Pipe Isometric Boric Acid Supply to Borated 4 Water Tank E-692 Equipment Arrangement Borated Water Storage Tank 2 Area , Sheet 3 E-692 Equipment Arrangement Plant System Outdoor Areas, 1 Sheet 4 E-2198 Schematic Diagram Low Pressure Safety Injection 18 Pump 2P60A, Sheet 1 E-2115 Schematic Diagram Containment Spray Pump 2P35A, 25 Sheet 1 E-2217 Schematic Diagram Spray Header Isolation Valve 20 2CV5612-1, Sheet 1 FSK-C-847 Documentation of Tank T-3 Borated Water Storage 0 Tank JN-D37178 Borated Water Storage Tank, Sheet 1 5 M-206 Steam Generator Secondary System, Sheet 1 123

-8-Drawings NUMBER DESCRIPTION REVISION M-213 Laundry Waste and Containment and Aux Building 23 Sump Drainage, Sheet 2 M-214 Clean Liquid Radioactive Waste, Sheet 3 17 M-219 Fire Water, Sheet 1 77 M-220 Plant Heating and Start-up Boiler, Sheet 3 14 M-222 Chilled Water System, Reactor and Auxiliary Buildings, 68 Sheet 1 M-230 Reactor Coolant System, Sheet 1 106 M-230 Reactor Coolant System, Sheet 2 35 M-231 Makeup and Purification System, Sheet 1 107 M-231 Makeup and Purification System, Sheet 2 43 M-232 Decay Heat Removal System, Sheet 1 96 M-233 Piping & Instrument Diagram, Chemical Addition 74 System, Sheet 1 M-234 Intermediate Cooling System, Sheet 1 88 M-234 Intermediate Cooling System, Sheet 2 41 M-235 Spent Fuel Cooling System, Sheet 1 61 M-236 Reactor Building Spray and Core Flooding Systems, 87 Sheet 1 M-237 Sampling System, Sheet 1 52 M-237 ANO-1 P&ID Post Accident Containment Atmosphere 15 Sampling System, Sheet 4 M-2213 Liquid Radioactive Waste System, Sheet 1 59

Drawings NUMBER -9-DESCRIPTION REVISION M-2214 Boron Management System, Sheet 1 84 M-2220 Plant Heating System, Sheet 1 64 M-2222 Chilled Water System, Containment, Turbine, and Aux 54 Buildings, Sheet 1 M-2230 Reactor Coolant System, Sheet 1 73 M-2230 Reactor Coolant System, Sheet 2 36 M-2231 Chemical And Volume Control System, Sheet 1 138 M-2232 Safety Injection System, Sheet 1 110 M-2235 Fuel Pool System, Sheet 1 66 M-2236 Containment Spray System, Sheet 1 89 M-2236 Containment Spray System, Sheet 2 18 M-2237 Sampling System, Sheet 1 63 M-2260 HVAC Control Room Expansion Facility, Sheet 5 2 M-2263 Units 1 & 2 Control & Computer Rooms HVAC, 72 Sheet 1, M-2422 Functional Description and Logic Diagram 15 Containment Spray System, Sheet 3 M-2505 Level Setting Diagram, Sheet 95 2 P-200 Instrumentation, Component Symbols, and Drawing 0 Index Sheet, Sheet 1 P-232 Boundary Diagram, Decay Heat Removal System, 0 Sheet 1

-10-Engineering Requests NUMBER DESCRIPTION REVISION 002311-B201 Review Inline Instruments for Impacts due to Pressure 0 and Temperature Changes in Calculation 88-E-0200-09, Containment Spray System 002311-E201 Unit 2 Containment Spray System Calculation 88-E- 0 0200-09, Revision 0, Discrepancy Resolution 002311-E206 Review Valve Body ANSI ratings for Unit 2 Calculation 0 88-E-022-09 002311-E207 Review Stress Calculations for Impact Due to Pressure 0 and Temperature Changes to Containment Spray System form Calculation 88-E-0200-09, Revision 002311-E209 Evaluate Vendor Piping to Determine Effects Due to 0 Changes in Pressure and Temperature in Reactor Building Spray Calculation 88-E-0200-09 002311-I202 Calculation 88-E-0200-09 Evaluation for Potential 0 Missiles 002311-R202 Review Fire Barriers for Impact Due to 0 Design/operating Temperature Increase in Calculation 88-E-0200-09 002415-E102 Arkansas Nuclear One-1 Spray Pump Lube Oil 0 Evaluation 002415-E103 Prediction of Pump 35A/B Bearing Temperatures at 0 Elevated Service Water Temperatures 002415-E104 Temporary Alteration Work Plan 1409,713 Screening 0 002415-E105 Operability Evaluation for Pump 35A and Pump 35B 0 010263-E101 Hydrogen Recombiner Reference Power Re- August 1, calculation 2000 963137-R201 Add Hinges to Front Panels on 2C182 & 184 0 991572-E201 Evaluation of Hydrogen Recombiners for an Increase 0 in Containment Pressure to 59 psig

-11-Engineering Requests NUMBER DESCRIPTION REVISION 991864-E229 Containment Structural Analysis for Uprate @ 59 psig 0 992054 E103 Provide the Instrument Loop Error Applicable to the 0 Differential Pressure Calculated via Subtracting the SPDS Suction Pressure from the SPDS Discharge Pressure for LPI Pumps P-34A and P-34B ER 002773 E101 Reverse Testing of Crosby Omni Series 900 Style 0 9551814B Pressure Relief Valves in Unit 1 Containment Penetrations ER 002971 E201 Engineering evaluation for Alternative 2P-89A Flow 0 Test ER-ANO-0528-005 HPSI Pump NPSH Margin Improvement 0 ER002612N101 ANO-1 GL 96-06 Phase II Modifications 0 ER2003-0332-012 Add Additional Footnote to LPI Flow Assumption Table 0 for LBLOCA in ANO-1 Groundrules Document CALC-A1-NE ER991864E229 Containment Analysis for Uprate to 59 psig 0 PEAR-95-0170 Extend the Scale for LG-1616, Sodium Hydroxide June 8, Tank Level Gauge 1995 Miscellaneous Documents NUMBER DESCRIPTION REVISION Containment (Building) Spray - Arkansas Unit 2 June 24, 2003 ESI - EMD Owners Group Recommended 3 Maintenance Program - Mechanical Letter - NRC to ANO - Issuance of Amendment October 3, Nos. 185 and 176 to Facility Operating License 1996 Reactor Building Spray - Arkansas Unit 1 June 24, 2003

-12-Miscellaneous Documents NUMBER DESCRIPTION REVISION 0CAN019702 Letter - ANO to NRC - 120-Day Response to January 28, Generic Letter 96-06 1997 0CAN019903 Additional Information Pertaining to Generic Letter January 25, 96-06 1999 0CAN019903 Letter - ANO to NRC - Additional Information January 25, Pertaining to Generic Letter 96-06 1999 0CAN049602 Letter - ANO to NRC - Tech Spec Change April 11, 1996 Request Concerning Implementation of 10CFR50, Appendix J, Option B 0CAN060301 Letter - ANO to NRC - ANO-1 & ANO-2 June 11, 2003 Commitment Change Summary Report and ANO-1 10CFR50.59 Summary Report 0CAN079710 Final Resolution of Generic Letter 96-06 July 31, 1997 0CAN089606 Letter - ANO to NRC - Modification of Proposed August 23, Tech Spec Change Request Concerning 1996 Implementation of 10CFR50, Appendix J, Option B

0CAN129703 Letter - ANO to NRC - Response to Generic Letter December 18, 96-06, Supplement 1 1997 0CNA020005 Completion of Licensing Action for Generic Letter February 7, 96-06 2000 0CNA069716 Letter - NRC to ANO - NRC Inspection Report 50- June 28, 1997 313/97-13; 50-368/97-13 and Notice of Violation and Notice of Deviation 2CNA067837 Testing and Inspection of Piping Systems June 29, 1978 Penetrating Containment CEP-IST-2 IST Plan, Valve Summary List, ANO-1 Appendix 2 and ANO-2 Appendix DCN 96-02193 Drawing Revision Notice As-Built M-237 for LCP 0 94-5034 Rev 0

-13-Miscellaneous Documents NUMBER DESCRIPTION REVISION DRN 03-01205 Calculation Change CALC-92-E-0078-09 July 30, 2003 EN-S Nuclear 10 CFR 50.59 Review Program 3 Management Manual LI-101 EN-S Nuclear ER Response Development 3 Management Manual DC-115 Engineering Report ANO-1 LOCA Analysis Summary Report 0 No. 02-R-1002-01 HES-02 Containment Leak Rate Testing Program 9 LBD Change SAR Section 6.2.3.2.2.2 Does not Reflect the July 14, 2003 2-6.2-0087 Current System Design MAI 25517 SIS Drn from SI Tank to RWT Valve Op November 5, 2000 MAI 31490 Personnel Lock September 16, 2000 MAI 34204 DH Suction Relief April 5, 2001 MAI 44158 DH Suction Relief March 27, 2001 STM 1-08 Reactor Building Spray and Containment Building 7 STM 2-08 Containment Spray System 8 TD W120 2200 Technical Manual for Electric Hydrogen 2 Recombiner Unit No. 2 TD W120.2230 Installation, Setup and Troubleshooting Hydrogen 0 Recombiner Power Supply Panel TD W120.3450 Instruction Manual for Electric Hydrogen 0 Recombiner Model B

-14-Miscellaneous Documents NUMBER DESCRIPTION REVISION Technical ANO-1 Technical Specification, Reactor Building 215 Specification 3.6.1 Technical ANO-2 Technical Specification, Containment 226 Specification 3/4.6 Systems Technical ANO-1 Technical Specification, Reactor Building 219 Specification 5.5.16 Leak Test Program ULD-0-TOP-14 Containment Isolation and Containment Leak Rate 1 Testing ULD-1-STR-02 ANO-1 Reactor Building 2 ULD-1-SYS-05 Arkansas Nuclear One-1 Reactor Building Spray 3 System ULD-1-SYS-18 ANO-1 Containment Hydrogen Control System 3 ULD-1-TOP-04 ANO-1 Containment Response to Design Basis 7 Accidents ULD-2-STR-02 ANO-2 Containment Building 1 ULD-2-SYS-05 ANO-2 Containment Spray System 3 ULD-2-SYS-06 ANO-2 Containment Heating and 2 Ventilation/Purge System ULD-2-SYS-18 ANO-2 Containment Hydrogen Control System 2 ULD-2-TOP-03 ANO-2 Containment Response to Design Basis 3 Accidents USAR Section 1.2.2 ANO-2 USAR, Concise Plant Description 17 USAR Section 1.4 ANO-1 USAR, General Design Criteria 18 USAR Section 14 ANO-1 USAR, Safety Analysis 18 USAR Section 15 ANO-2 USAR, Accident Analysis 17

-15-Miscellaneous Documents NUMBER DESCRIPTION REVISION USAR Section 5.2 ANO-1 USAR, Reactor Building 18 USAR Section 6 ANO-2 USAR, Engineered Safety Features 17 USAR Section 6 ANO-1 USAR, Engineered Safeguards 18 Workplan 1409.731 Control Room Envelope Unfiltered Air Inleakage 0 Measurement Test Modifications NUMBER DESCRIPTION REVISION DCP 97-4813-D101 Install Pressure Relieving Devices on May 6, Containment Penetrations to Comply with NRC 1999 Generic Letter 96-06 ER-ANO-2000-2255-001 Filter Addition to Cabinet C-178 and C-179 0 ER-ANO-2002-0357-0000 Replacement of ACW Boundary Isolation Valve 0 CV-3643 ER-ANO-2002-0363-000 MFW Pump Lube Oil Pump P-26A/B and P- 0 27A/B Motor Equivalents ER-ANO-2002-0929-000 Upgrade of overload protection of all ICW Pumps 0 (P-33A, B, C)

ER-ANO-2002-1223-001 Add Room Flooding Alarms to Decay Heat Vaults 0 ER-ANO-2003-0099-000 Install splice on 2PM4C power cables 0 LCP 94-5034 ANO-1 Hydrogen Analyzer Modification 1, 2

-16-Procedures NUMBER DESCRIPTION REVISION 1000.131 10CFR50.59 Review Program 003-04-0 1015.003A Unit 1 Operations Logs 050-04-0 1022.011 Reactor Core Monitoring Activities 005-00-0 1032.037 Inspection and Evaluation of Boric Acid Leaks 000-05-0 1102.010 Plant Shutdown and Cooldown 053-02-0 1104.005 Reactor Building Spray System Operation 042-04-0 1104.031 Containment Hydrogen Control 014-01-0 1202.010 Engineered Safety Feature Actuation System 005-01-0 1202.012 Repetitive Tasks 004-02-0 1203.024 Loss of Instrument Air 010-07-0 1203.028 Loss of Decay Heat Removal 016-02-0 1203.030 Loss of Service Water 013-00-0 1307.031 Unit 1 Hydrogen Recombiners (M55A & B Surveillance Testing) 004-03-0 1307.037 Unit 1 Plant Freeze Protection Testing 014-00-0 1309.013 Unit One Service Water Flow Test 009-06-0 1403.007 Unit 1 Heat Trace System Maintenance 004-03-0 1412.001 Preventive Maintenance of Limitorque SB/SMB Motor 012-03-0 Operators 2102.002 Plant Heatup 051-02-0 2104.005 Containment Spray 041-07-0 2104.033 Containment Atmosphere Control 042-00-0 2104.039 HPSI System Operation 041-06-0

-17-Procedures NUMBER DESCRIPTION REVISION 2106.032 Unit 2 Freeze Protection Guide 009-04-0 2202.003 Unisolated Loss of Coolant Accident 006-00-0 2203.012T Annunciator 2K20 Corrective Action 014-01-0 2304.029 Unit 2 Hydrogen Purge System Analyzer 2AITS-8371-1 020-00-0 2304.031 Hydrogen Recombiner Temperature Calibration 009-00-0 2305.006 Cold Shutdown Valve Testing 017-01-0 2305.009 Containment spray System Integrity Test and Leak Rate N/A Determination 2311.008 EDG Heat Exchanger Performance Test 004-00-0 2403.016 Unit Two Hydrogen Recombiner Inspection and Electrical 008-04-0 Testing 2409.707 2P-89A HPSI Pump Alternate Testing 0 5010.004 Design Document Changes 005-01-0 5120.402 Unit 1 Primary Containment Leak Rate Running Total 008-00-0 5120.403 Unit 2 Primary Containment Leak Rate Running Total 008-00-0 5120.422 Containment Recirc Fan Flow Rate Surveillance Test 001-01-0 LI-102 Corrective Action Process 2 SES16 Spring Can Setting Tolerances 0 Safety Evaluations NUMBER DESCRIPTION REVISION 1995 - 225 ANO-1 Hydrogen Analyzer Modification 11/30/95 2001 - 46 Reanalysis of the ANO-1 Main Feedwater HELB 10/25/01

-18-Safety Evaluations NUMBER DESCRIPTION REVISION 2001 - 48 Control Room Envelope Unfiltered Air Inleakage 10/18/01 Measurement Test Plan)

2002 - 22 Unit 1 ITS 3.5.4 Bases BWST Temperature Limit 6/27/02 Surveillance 2002 - 24 ANO-1 LOCA Analysis Summary Report 7/25/02 2002 - 31 EDG Test Requirements Frequency 9/18/02 FFN-00-080 Install Pressure Relieving Devices on Containment 0 Penetrations to Comply with NRC GL 96-06 - Unit 2 FFN-01-018 ANO-1 GL 96-06 Phase II Modifications 0 FFN-99-074 Install Pressure Relieving Devices on Containment 0 Penetrations to Comply with NRC GL 96-06 LCP 94-5034 ANO-1 Hydrogen Analyzer Modification 1 Safety Evaluation Sceenings NUMBER DESCRIPTION REVISION 1032.037 Inspection and Evaluation of Boric Acid Leaks 000-05-0 1102.010 Plant Shutdown and Cooldown 053-02-0 1202.010 ESAS 005-01-0 2305.006 Cold Shutdown Valve Testing 017-01-0 2311.008 EDG Heat Exchanger Performance Test 004-00-0 ER-ANO-2000-2255-001 Filter Addition to Cabinet C-178 and C-179 0 ER-ANO-2002-0357-0000 Replacement of ACW Boundary Isolation Valve 0 CV-3643

-19-Safety Evaluation Sceenings NUMBER DESCRIPTION REVISION ER-ANO-2002-0363-000 MFW Pump Lube Oil Pump P-26A/B and P- 0 27A/B Motor Equivalents ER-ANO-2002-0929-000 Upgrade of overload protection of all ICW Pumps 0 (P-33A, B, C)

ER-ANO-2002-1223-001 Add Room Flooding Alarms to Decay Heat Vaults 0 ER-ANO-2003-0099-000 Install splice on 2PM4C power cables 0 ER991864E238 Civil Uprate of Containment Structure from 1 54 psig to 59 psig OP 2409.707 2P-89A HPSI Pump Alternate Testing 0 OP-1022.011 Reactor Core Monitoring Activities 005-00-0