IR 05000458/2004008
| ML041680240 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 06/16/2004 |
| From: | Clark J Division of Reactor Safety IV |
| To: | Hinnenkamp P Entergy Operations |
| References | |
| IR-04-008 | |
| Download: ML041680240 (24) | |
Text
June 16, 2004
SUBJECT:
RIVER BEND STATION - NRC SAFETY SYSTEM DESIGN AND PERFORMANCE CAPABILITY INSPECTION REPORT 05000458/2004-008
Dear Mr. Hinnenkamp:
On May 21, 2004, the U.S. Nuclear Regulatory Commission (NRC) completed the onsite portion of an inspection at your River Bend Station. The enclosed report documents the inspection findings, which were discussed 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, no findings of significance were identified.
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/
Jeff Clark, Chief Engineering Branch Division of Reactor Safety Docket: 50-458 License: NPF-47
Entergy Operations, Inc.
-2-cc w/enclosure:
Senior Vice President and Chief Operating Officer Entergy Operations, Inc.
P.O. Box 31995 Jackson, MS 39286-1995 Vice President Operations Support Entergy Operations, Inc.
P.O. Box 31995 Jackson, MS 39286-1995 General Manager Plant Operations Entergy Operations, Inc.
River Bend Station 5485 US Highway 61N St. Francisville, LA 70775 Director - Nuclear Safety Entergy Operations, Inc.
River Bend Station 5485 US Highway 61N St. Francisville, LA 70775 Wise, Carter, Child & Caraway P.O. Box 651 Jackson, MS 39205 Mark J. Wetterhahn, Esq.
Winston & Strawn 1401 L Street, N.W.
Washington, DC 20005-3502 Manager - Licensing Entergy Operations, Inc.
River Bend Station 5485 US Highway 61N St. Francisville, LA 70775 The Honorable Charles C. Foti, Jr.
Attorney General Department of Justice State of Louisiana P.O. Box 94005 Baton Rouge, LA 70804-9005
Entergy Operations, Inc.
-3-H. Anne Plettinger 3456 Villa Rose Drive Baton Rouge, LA 70806 Burt Babers, President West Feliciana Parish Police Jury P.O. Box 1921 St. Francisville, LA 70775 Michael E. Henry, State Liaison Officer Department of Environmental Quality Permits Division P.O. Box 4313 Baton Rouge, LA 70821-4313 Brian Almon Public Utility Commission William B. Travis Building P.O. Box 13326 1701 North Congress Avenue Austin, TX 78711-3326
Entergy Operations, Inc.
-4-Electronic distribution by RIV:
Regional Administrator (BSM1)
DRP Director (ATH)
DRS Director (DDC)
Senior Resident Inspector (PJA)
Branch Chief, DRP/B (DNG)
Senior Project Engineer, DRP/B (RAK1)
Staff Chief, DRP/TSS (PHH)
RITS Coordinator (KEG)
Rebecca Tadesse, OEDO RIV Coordinator (RXT)
RBS Site Secretary (LGD)
ADAMS: Yes
No Initials: _JAC_____
Publicly Available
Non-Publicly Available
Sensitive
Non-Sensitive SRI:EB RI:EB RI:EB RI:EB RI:EB C:EB C:PBB C:EB LEEllershaw/lmb JPAdams JMMateychick WMMcNeill BWHenderson CSMarschall DNGraves JClark
/RA/
/RA/
/RA/
/RA/
/RA/
/RA/
/RA/
/RA/
06/08/04 06/09/04 06/09/04 06/09/04 06/10/04 06/10/04 06/10/04 06/16/04 OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket Nos:
50-458 License Nos:
NPF-47 Report No:
05000458/2004-008 Licensee:
Entergy Operations, Inc.
Facility:
River Bend Station Location:
5485 U.S. Highway 61 St. Francisville, Louisiana Dates:
May 3-21, 2004 Team Leader:
L. E. Ellershaw, Senior Reactor Inspector, Engineering Branch Inspectors:
J. P. Adams, Reactor Inspector, Engineering Branch J. M. Mateychick, Reactor Inspector, Engineering Branch W. M. McNeill, Reactor Inspector, Engineering Branch B. W. Henderson, Reactor Inspector, Engineering Branch Approved by:
Jeff Clark, Chief Engineering Branch Division of Reactor Safety
-2-SUMMARY OF FINDINGS IR 05000458/2004008; 05/03/2004 through 05/21/2004, River Bend Station; Evaluation of Changes, Tests, or Experiments, and Safety System Design and Performance Capability The NRC conducted an inspection with five regional inspectors. No findings of significance were identified. 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.
NRC-Identified Findings and Self-Revealing Findings No findings of significance were identified.
Report Details 1.
REACTOR SAFETY Introduction The NRC conducted an inspection to verify that licensee personnel adequately preserved the facility safety system design and performance capability and that licensee personnel 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 personnel based the probabilistic risk assessment model for the River Bend Station on the capability of the as-built safety systems to perform their intended safety functions successfully. The team 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. The team established this according to their ranking and potential contribution to dominant accident sequences and/or initiators. The team also used a deterministic approach in the selection process by considering recent inspection history, recent problem area history, and all modifications developed and implemented.
The minimum sample size for this procedure is one risk-significant system for mitigating an accident or maintaining barrier integrity. The team completed the required sample size by reviewing the containment structures. The primary review prompted parallel review and examination of support systems, such as, containment atmosphere control, standby gas treatment, residual heat removal, and related structures and components.
The team assessed the adequacy of calculations, analyses, engineering processes, and engineering and operating practices that licensee personnel used for the selected safety system and the necessary support systems during normal, abnormal, and accident conditions. Acceptance criteria used by the team included NRC regulations, the technical specifications, applicable sections of the Updated Final 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 The minimum sample size for this procedure is 5 evaluations and 10 screenings. The team reviewed 7 licensee-performed 10 CFR 50.59 evaluations to verify that 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. These evaluations had been performed since the last NRC inspection of 10 CFR 50.59 activities.
-2-The team reviewed 10 licensee-performed 10 CFR 50.59 screenings, in which licensee personnel determined that evaluations were not required to ensure that exclusion of a full evaluation was consistent with the requirements of 10 CFR 50.59. Additionally, the team reviewed 10 licensee-performed applicability determinations, in which licensee personnel determined that neither screenings nor evaluations were required, to ensure consistency with the requirements of 10 CFR 50.59 in the licensees exclusion of screenings and evaluations.
The team reviewed and evaluated the most recent licensee 10 CFR 50.59 program self assessment and a sample of 10 corrective action documents written since the last NRC 10 CFR 50.59 inspection to determine whether licensee personnel 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 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 are consistent with actions required to meet, prevent, and/or mitigate design basis accidents. The team also considered requirements and commitments identified in the Updated Final Safety Analysis Report, technical specifications, design basis documents, and plant drawings. In conjunction with the primary review of the reactor containment structures, a parallel review and examination of support systems, such as, containment atmospheric control, standby gas treatment, residual heat removal (shutdown cooling mode), penetrations, and related structures and components was also conducted.
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
-3-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 associated with containment structures and support systems that were identified by licensee personnel 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 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 that were identified while performing other areas of the inspection were also reviewed.
b.
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. The team 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 team 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 operator action,
The material conditions and preservation of systems and equipment, and
The conformance of the currently-installed system configurations to the design and licensing bases.
-4-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, environmental qualifications, required system thermal-hydraulic performance, electrical power system performance, control logic, and instrument setpoints and uncertainties. The team assessed the adequacy of calculations, analyses, test procedures, and operating procedures that licensee personnel used during normal and accident conditions.
The team also reviewed the adequacy of the combustible gas control systems original design to control hydrogen concentrations in the drywell and containment during post-accident conditions, including maintaining the capability of the selected support systems to perform their design basis functions. The support systems reviewed in detail were the drywell and containment hydrogen analyzer system, hydrogen igniter system, hydrogen recombiner system and drywell purge system.
b.
Findings No findings of significance were identified.
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 inservice tests.
b.
Findings No findings of significance were identified.
4OA3 Event Followup (Closed) Licensee Event Report 05000458/2003003, Primary Containment Airlock Breach Due to Door Interlock Malfunction On March 10, 2003, the upper airlock door interlocks at the 171-foot elevation of the primary containment did not properly function, such that both the inner and outer doors were unsealed for a period of approximately 36 minutes. At the time the reactor was operating at 87 percent power in end-of-cycle coastdown. This event was reported in
-5-accordance with 10 CFR 50.73(a)(2)(v)(c) as a condition that could have prevented the primary containment from performing its safety function. The licensee repaired the 171-foot upper airlock door interlocks to restore proper function. A root-cause analysis was performed following the event and resulted in the licensee changing maintenance practices and replacing the cables which operate the interlocks with a new and improved design. Identical actions were taken on the 113-foot elevation lower airlock door interlocks to prevent similar problems. The team reviewed the licensee event report and the root-cause analysis, and no findings of significance were identified. The licensee documented this condition in Condition Report CR-RBS-2003-0882. This licensee event report is closed.
4OA6 Management Meetings Exit Meeting Summary The inspection findings were acknowledged during an exit meeting presented by the team leader on May 21, 2004, to Mr. T. E. Trepanier, and other members of licensee management staff. The team leader confirmed that proprietary information, while reviewed, had not been retained by the team.
ATTACHMENT PARTIAL LIST OF PERSONS CONTACTED Licensee:
M. Ballard, Supervisor, Quality Audits T. Burnett, Superintendent, Chemistry C. Forpahl, Manager, Corrective Action and Assessments R. Hebert, Manager, Materials, Procurement, and Contracts R. King, Director, Nuclear Safety Assurance J. Malara, Director, Engineering W. Mashburn, Manager, Programs and Components T. Trepanier, General Manager, Plant Operations J. Clark, Assistant Manager, Operations A. Roshto, Superintendent, Electrical Maintenance B. Fountain, Licensing Specialist NRC M. Miller, Resident Inspector LIST OF ITEMS CLOSED Closed 50-458/2003-003 LER Airlock door interlocks of the primary containment did not properly function such that both the inner and outer doors were unsealed and could have prevented the primary containment from performing its safety function (Section 4OA3)
DOCUMENTS REVIEWED Calculations G13.18.2.3*147, G.L. 89-10 Design Basis Review for E12-MOVF003A/B, Revision 1 G13.18.2.3*148, G.L. 89-10 Design Basis Review for E12-MOVF004A/B, Revision 4 G13.18.2.3*155, G.L. 89-10 Design Basis Review for E12-MOVF024A/B, Revision 5 G13.18.2.3*160, G.L. 89-10 Design Basis Review for E12-MOVF042A/B, Revision 3
G13.18.15.2*052, Cat. 1 Maximum Thrust Force for Valve 1E12*MOVF064A, 1E12*MOVF064B, & 1E12*MOVF064C, Revision 0 G13.18.15.2*053, Maximum Thrust Force for Valves: 1E12*MOVF042A, 042B, & 042C and 1E21*MOVBF005, Revision 1
-2-
G13.18.15.2*107, Maximum Thrust Force for Valves: 1E12*MOVF048A & F048B, Revision 0 ES-170-6, Addendum C (ES-170-6C), Accident Environmental Conditions In The Containment, Auxiliary Building, and Steam Tunnel, Revision 0 ES-170-6, Addendum D (ES-170-6D), Accident Environmental Conditions In The Containment, Auxiliary Building, and Steam Tunnel, Revision 0 ES-170-6, Addendum E (ES-170-6E), Accident Environmental Conditions In The Containment, Auxiliary Building, and Steam Tunnel, Revision 0 ES-225, Post DBA Hydrogen Concentration in Drywell and Containment for FSAR Section 6.2.5, Revision 0 PB-326, Effect of LOCA on Containment Unit Coolers, Revision 0 PN-268, RHR System Pumps TDH and NPSHA Except LPCI (Mode A-2) Operation, Revision 5 7221-438-312-002C, Weak Link Analysis for 14" ANSI Class 300 PermaSeat Valve, Revision D BV-45.20-1, Fan External Total Pressure 1GTS*FN1A 7 FN1B Containment/Drywell Purge Exhaust Normal Operation Mode III, Revision 1 BV-45.21, Fan External Total Pressure 1GTS*FN1A and FN2B Decay Heat Exhaust Fans Normal Operation Mode II, Revision 2 BV-45.22-1, Fan External Total Pressure Fans 1GTS*FN1A & FN1B Annulus and Auxiliary Building Exhaust Accident Mode, Revision 0 ES-073, Wind Velocity to Offset Building Vacuum, Revision 2 ES-194, Auxiliary Building Pressure Following Loss of Coolant Accident for Updated Safety Analysis Report Sections 6.2.3, Revision 4 ES-0205, Technical Specification Secondary Containment Integrity Drawdown Times During Normal Operation, Revision 1 G13.18.2.1*079, Evaluation of Standby Gas Treatment System Drawdown Data, Revision 0 G13.18.2.7*023, Shield Building Annulus Pressure Following Loss of Coolant Accident for Updated Safety Analysis Report Section 6.2.3, Revision 2 G13.18.14.0*01, Standby Gas Treatment System and Fuel Building Charcoal Filter Decay Heat, Revision 0
-3-G13.18.9.5*051, Loss of Coolant Accident Doses for Updated Safety Analysis Report Chapter 15, Revision 2 G.13.18.9.5*061, Alternate Source Term Loss of Coolant Accident Off-Site and Control Room Dose Analysis, Revision 0 PB-206, Sizing of Emergency Charcoal Filters 1GTS*FLT 1A and 1B, Revision 1 PB-361, Charcoal Weight and Iodine Loading Filtration Units, 1HVF*FLT2A-2B and 1GTS*FLT1A-1B, Revision 0 PR-C-087, Halogen Loading on the Standby Gas Treatment System and Fuel Building Charcoal Filters due to a Loss of Coolant Accident, Revision 0 PR-C-344, Peak Radiological Decay Heat Rate for Standby Gas Treatment System Filters during Loss of Coolant, Revision 1A Condition Reports (CRs)
CR-RBS-2003-03431 CR-RBS-2002-01177 CR-RBS-1993-00403 CR-RBS-2002-00751 CR-RBS-1998-00437 CR-RBS-1997-00703 CR-RBS-2002-00738 CR-RBS-1998-00430 CR-RBS-1997-02154 CR-RBS-2004-01325 CR-RBS-2003-00009 CR-RBS-2004-00945 CR-RBS-2004-01493 CR-RBS-2003-01407 CR-RBS-2004-00172 CR-RBS-2002-01404 CR-RBS-2004-01374 CR-RBS-2003-03341 CR-RBS-2003-02661 CR-RBS-2004-01346 CR-RBS-2003-00961 CR-RBS-2004-00172 CR-RBS-2004-01346 CR-RBS-2003-00605 CR-RBS-2002-02057 CR-RBS-2004-01331 CR-RBS-2003-00009 CR-RBS-2003-00605 CR-RBS-2004-01328 CR-RBS-2002-02057 CR-RBS-2002-00751 CR-RBS-2004-01320 CR-RBS-2002-01404 CR-RBS-2002-00738 CR-RBS-2004-01305 CR-RBS-2004-00194 CR-RBS-1997-00526 CR-RBS-2002-01177 CR-RBS-2004-01086 CR-RBS-2002-00443 CR-RBS-2004-00197 CR-RBS-2002-00353 CR-RBS-2002-00131 CR-RBS-2002-01343 CR-RBS-2003-00364 CR-RBS-2003-02028 CR-RBS-2004-01045 CR-RBS-2003-00882 CR-RBS-2002-00443 CR-RBS-2004-00122 CR-RBS-2004-00004 CR-RBS-2004-01478 CR-RBS-2004-01479 CR-RBS-2004-01480 Design Basis Documents CSD-27-13, Control System Description for Engineered Safety Features Hydrogen Recombiner System Diagram 27-13, April 6, 1984 CSD-27-21, Control System Description for Engineered Safety Features Containment Hydrogen Purge Diagram 27-21, July 26, 1985 CSD-27-24, Control System Description for Engineered Safety Features Hydrogen Mixing Diagram 27-24, September 17, 1985 SDC-204, Residual Heat Removal System Design Criteria System Number 204, Revision 3
-4-SDC-403, 404, & 409, Reactor Plant Ventilation System Design Criteria System Numbers 403, 404, & 409, Revision 4 SDRD-E2, System Design Requirements Document - Hydrogen Control, Revision 0 SDRD-P12, System Design Requirements Document - Containment Hydrogen Control, Revision 0 Drawings ESK-06HVR09, Elementary Diagram 480V SWGR Containment Unit Cooler*UC1A, Revision 22 ESK-06HVR10, Elementary Diagram 480V SWGR Containment Unit Cooler*UC1B, Revision 20 PID-27-21A, Engineering P & I Diagram System 254 Hydrogen Mixing Purge & Recombiner, Revision 5 PID-33-02A, Engineering P & I Diagram System 552 Containment Atmosphere and Leakage Monitoring, Revision 16 PID-33-02B, Engineering P & I Diagram System 552 Containment Atmosphere and Leakage Monitoring, Revision 1B PID-33-02C, Engineering P & I Diagram System 552 Containment Atmosphere and Leakage Monitoring, Revision 7 LSK 22-1.6B, Logic Diagram Reactor Plant Ventilation Annulus Mixing, Revision 12 LSK 22-1.6C, Logic Diagram Reactor Plant Ventilation Annulus Mixing, Revision 12 LSK 27-15A, Logic Diagram Standby Gas Treatment System, Revision 121 LSK 27-15D, Logic Diagram Standby Gas Treatment System, Revision 12 LSK 27-15F, Logic Diagram Standby Gas Treatment System, Revision 11 PID-22-01B, Engineering Process and Instrument Diagram System 403 Heating, Ventilation and Air Conditioning - Containment Building, Revision 16 PID-22-01C, Engineering Process and Instrument Diagram System 403 Heating, Ventilation and Air Conditioning - Containment Building, Revision 13 PID-27-15A, Engineering Process and Instrument Diagram System 257 Standby Gas Treatment, Revision 15 TLD HVR-041, Annulus Pressure Control System Suction Flow, Sheet 1, Revision 0 TLD HVR-041, Annulus Pressure Control System Suction Flow, Sheet 2, Revision 0
-5-828E535AA, Elementary Diagram Low Pressure Core Spray System, Sheet 6, Revision 22 828E535AA, Elementary Diagram Low Pressure Core Spray System, Sheet 10, Revision 234 Root-Cause Analysis Reports 171 ft. Containment Airlock Inadvertent Breach Event Miscellaneous Documents NRC Letter J. F. Harold to R. K. Edington, Subject: River Bend, Unit 1 - Issuance of Amendment Re: Increase in Maximum Allowable Thermal Power to 3039 Megawatts Thermal (TAC MA6185),
dated October 6, 2000 and attached Safety Evaluation.
Modification Request 91-0101, To Override LOCA Signal for Manual Initiation of the System, September 23, 1991 Modification Request 95-0016, RHR Test Return Lines Vibration Reduction, August 24, 1995 Documentation of Telecom with NRC Regarding Fuel Reloads for GGNS&RBS From Ron Byrd, April 9, 2001. (Conversation between Adrienne Smith, Ron Byrd (both Entergy) and Pat Sekerak (NRC-NRR) dated April 3, 2001)
Technical Specification 3.3.6.3, Containment Unit Cooler System Instrumentation, Amendment 81 Technical Specification 3.6.3.1, Primary Containment Hydrogen Recombiners, Amendment 81 Technical Specification 3.6.3.2, Primary Containment and Drywell Hydrogen Igniters, Amendment 81 Technical Specification 3.6.3.3, Primary Containment/Drywell Hydrogen Mixing System, Amendment 89 Technical Specification 5.5.13, Primary Containment Leakage Rate Testing Program, Amendment 132 Technical Specification Bases B 3.6.3.1, Primary Containment Hydrogen Recombiners, Revision 4-3 Technical Specification Bases B 3.6.3.2, Primary Containment and Drywell Hydrogen Igniters, Revision 3-3 Technical Specification Bases B 3.6.3.3, Primary Containment/Drywell Hydrogen Mixing System, Revision 4-3 Entergy Licensing Position, Evaluation and Resolution of Degraded and Nonconforming Conditions, Revision 1
-6-Loop Calibration Report No. 1.ILGTS.012, Standby Gas Treatment Filter Train A Electric Heater Temperature Loop, Revision 5 LO-RLO-204-00113, Assessment Report: Evaluation of River Bend Station 10 CFR 50.59 Changes, Tests, or Experiments Program, dated April 16, 2004 NFTA-NTR423, Field Test Reports, dated October 15, 1985 SDRD-P50, System Design Requirements Document, Revision 0 Safety Evaluation Report, Increase in Maximum Allowable Thermal Power to 3039 Megawatts Thermal, dated October 6, 2000 Specification 216.160, Shop Fabrication of Ventilation and Air-Conditioning Systems Safety-Related Areas, Revision 2 Specification 225.220, Standby Gas Treatment Units, dated January 7, 1976 Stone & Webster Letter C-RBS-04444, dated June 25, 1986 NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, dated July 26, 1995 Procedures ADM-0050, Primary Containment Leakage Rate Testing Program, Revision 14 ADM-0085, Repetitive Task Program, Revision 5 SEP-APJ-001, Primary Containment Leakage Rate Testing Program (App. J), Revision 0 CEP-IST-1, Inservice Testing Bases Document, RBS Appendix, Revision 3 CEP-IST-2, Inservice Testing Plan (including RBS Appendix, Valve Summary Listing and Pump Summary Listing), Revision 3 CEP-IST-3, EN-S IST Cross-Reference Document (including listing of procedures by pump and a listing of procedures by valve), Revision 1 CEP-IST-4, Entergy Nuclear South Standard on Inservice Testing, Revision 0 LI-102, Corrective Action Process, Revision 4 OE-100, Operating Experience Program, Revision 1 EOP-0005, Emergency Operating and Severe Accident Procedures Enclosures, Revision 15 EDG-AA-014, Technical Evaluations for Operability Determinations, Revision 0
-7-ENG-3-028, Processing of System Design Criteria (SDC) Documents, Revision 6 EOP-0002, Emergency Operating Procedure - Primary Containment Control, Revision 13 EOP-0003, Emergency Operating Procedure - Secondary Containment and Radioactive Release Control, Revision 13 ENS-DC-126, Engineering Calculation Process, Revision 3 ENS-DC-141, Design Inputs, Revision 2 G12.1.22, Preparation, Review, Approval and Control of Revisions to the River Bend Station Environmental Design Criteria Database (RBS-EDC), Revision 06 PEP-0240, Performance Monitoring Program for the Residual Heat Removal Heat Exchangers E12-EB001B and E12-EB001D (Div II), Revision 03 SAP-0001, Severe Accident Procedure - RPV and Primary Containment Control, Revision 04 SAP-0002, Severe Accident Procedure - Containment and Radioactive Release Control, Revision
SOP-0031, Residual Heat Removal (Sys #204), Revision 41 SOP-0040, Hydrogen Mixing, Purge, Recombiners, and Ignitors, Revision 10A SOP-0084, Containment Atmospheric Monitoring System (Sys #552), Revision 11 STP-051-4279, Containment Unit Cooler System Instrumentation, Unit Cooler A - Containment to Annulus Differential Pressure High Channel Calibration and Logic System Functional Test (HVR-ESZ60A, HVR-ESY60A, HVR-ESX60A, HVR-PDT60A), Revision 9 STP-254-0601, Containment/Drywell H2 Mixing System Flow Test, Revision 8 STP-254-1401, Division 1 Hydrogen Igniter Train Current and Voltage Check, Revision 4 STP-254-1603, Division 1 and 2 Hydrogen Igniter Current, Voltage and Temperature Check, Revision 4 STP-254-4203, Post Accident Monitoring - Drywell and Containment Hydrogen Analyzer Channel Calibration CMS-PNL10A, CMS-PNL12A, CMS-AR25A (Point 1), Revision 22A STP-403-0301, Containment Unit Cooler HVR-UC1A Flow Rate Verification, Revision 10 STP-403-0303, Containment Unit Cooler HVR-UC1B Flow Rate Verification, Revision 0A STP-403-1200, HVR-UC1A System A Timer Channel Functional Test, Revision 10 STP-403-1201, HVR-UC1B System A Timer Channel Functional Test, Revision 4
-8-TSP-0016, Loop A RHR System Leak Test, Revision 3B TSP-0017, Loop B RHR System Leak Test, Revision 4 AOP-0003, Automatic Isolations, Revision 19
ARP-RMS-DSPL230, DRMS RM-11 CRT (RMS-DSPL230) Alarm Response, Revision 3 ARP-863-73, Division I Standby Gas Treatment Heat Removal System Inoperative, Revision 5 DG-LI-101, 10 CFR 50.59 Review Program Guidelines, Revision 5 EOP-00003, Emergency Operating Procedure-Secondary Containment and Radioactive Release Control, Revision 13 LI-101, 10 CFR 50.59 Review Program, Revision 3 LI-113, Licensing Basis Document (LBD) Control Program, Revision 3 SOP-0043, Standby Gas Treatment System (System Number 257), Revision 10 SOP-0059, Containment Heat Ventilation and Air Conditioning System (System 403),
Revision 22 STP-000-0001, Daily Operating Logs, Revision 45 STP-257-3602, Inservice Testing of Division II Standby Gas Treatment Filtration System, Revision 00B STP-257-0601, Standby Gas Treatment System Train A Drawdown Test, Revision 14 STP-257-0602, Standby Gas Treatment System Train B Drawdown Test, Revision 5 STP-257-8601, Standby Gas Treatment System Laboratory Carbon Filter Analysis, Revision 11 STP-309-0601, Division I 18 Month Emergency Core Cooling System Test,. Revision 20 STP-403-0603, Division I Standby Gas Treatment System and Annulus Mixing System Functional Test, Revision 2 STP-403-0604, Division II Standby Gas Treatment System and Annulus Mixing System Functional Test, Revision 2 STP-601-6801, Reactor Water Clean Up Cold Shutdown Valve Operability Test, Revision 3 UFSAR, Revision 14 Section 6.2, Containment Systems
-9-Section 7.3, Engineered Safety Feature Systems Section 15.6, Decrease In Reactor Coolant Inventory 10 CFR 50.59 Evaluations SEN 2003-013 SEN 2003-007 SEN 2001-024 SEN 2003-010 SEN 2002-024 SEN 2003-014 SEN 2002-016 10 CFR 50.59 Screenings Procedure AOP-0020, Revision 2 Procedure AOP 0031, Revision 19 Procedure AOP 0052, Revision 12 Procedure EOP 0001, Revision 19 Procedure TSP 0010, Revision NA Procedure TSP 0019, Revision NA Procedure FHP 0001, Revision 25 Procedure FHP 0008, Revision 3 ER-RB-2001-0801-000, Revision 0 ER03-0510-000, Revision 0 10 CFR 50.50 Evaluation Exemptions LAR 2004-05 ER-00-0330, Revision 0 & LCN 09.03-259 ER-RB-2002-0333-000 & LBDC-07.03-194 LBDC N.03.04-001 LCN 10.04-191 LBDC 09.02-346 LBDC 05.02-032 LBDC 06.02-107 LBDC 09.04-175 LBDC 06.02-108 Set Point Data Sheet Numbers 12210-IA-GTS*FS24 12210-IA-GTS-PDS6 12210-IA-GTS-PDS7 12210-IA-GTS-PDS16 12210-IA-GTS-PDS17 12210-IA-GTS-PDS26
-10-Special Test Report TP-98-0002 (both tests performed on April 25 and April 18, 1998)
System Design Criteria SDC-257, Standby Gas Treatment System Number 257, Revision 1 Maintenance Action Items and Associated Test Reports:
346761 350231 351498 368735 369418 349620 369423 369424 316759 316760 369419 349620 Training Manuals R-STM-257.01, Standby Gas Treatment System, Revision 1 Work Orders 50685430 50686758 50690017 50691392 50870370 50967705 50968944 50970230 50971425 50972175 50973996 50975102 50373327-01 Work Requests WO 18895 WO 18896 Engineering Requests 98-0166 98-0215 98-0323 98-0342 97-0291 99-0885 2001-0115 Inservice Test Reports for the following Residual Heat Removal Pumps and Valves (last four quarterly tests):
E12-PC002A E12-PC002B E12-PC002C E12-MOVF003A E12-MOVF004A E12-MOVF011B E12-MOVF024A E12-MOVF027A E12-MOVF042A E12-MOVF047A E12-MOVF048A E12-MOVF053B E12-MOVF064A E12-MOVF094 E12-MOVF096
-11-Surveillance Test Reports Test Reports for Containment Purge Valves in Penetrations KJB-Z31 and KJB-Z33: HVR-AOV123, -AOV128, -AOV165, -AOV166, CPP-MOV104, -MOV105, and -SOV140, dated 6/4-5/03, 8/27-28/03, 11/18-19/03, 2/9-10/04, and 5/3-4/04, respectively Test Reports for Primary Containment Upper Airlock Inner and Outer Doors, and Lower Airlock Inner and Doors dated 1/6/04, 2/5/04, 3/2/04, 3/31/04, and 4/27/04 Section XI Safety and Relief Valve Testing for FPW-RV40 for period from 9-15-1997 to present (seven reports)
Surveillance Test Procedures STP-057-0401, Revision 15 STP-403-7301, Revision 3 STP-057-7705, Revision 7 STP-057-3900, Revision 9