IR 05000416/2004006
| ML040510344 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 02/20/2004 |
| From: | Marschall C Division of Reactor Safety IV |
| To: | Gerald Williams Entergy Operations |
| References | |
| IR-04-006 | |
| Download: ML040510344 (20) | |
Text
ary 20, 2004
SUBJECT:
GRAND GULF NUCLEAR STATION - NRC INSPECTION REPORT 05000416/2004-006
Dear Mr. Williams:
On January 30, 2004, the U.S. Nuclear Regulatory Commission (NRC) completed the onsite portion of an inspection at your Grand Gulf Nuclear Station. In-office inspection was continued through February 5, 2004, to review issues associated with preventive maintenance practices applicable to containment structures and support systems. The enclosed report documents the inspection findings, which were discussed on January 30, 2004, with you and other members of your staff, and during a subsequent telephone call on February 19, 2004, with Mr. M. Krupa, Director, Nuclear Safety Assurance.
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/
Charles S. Marschall, Chief Engineering Branch Division of Reactor Safety
Entergy Operations, Inc. -2-Docket: 50-416 License: NPF-29
Enclosure:
NRC Inspection Report 05000416/2004-006
REGION IV==
Docket Nos: 50-416 License Nos: NPF-29 Report No: 05000416/2004-006 Licensee: Entergy Operations, Inc.
Facility: Grand Gulf Nuclear Station Location: Waterloo Road Port Gibson, Mississippi 39150 Dates: January 12 through February 19, 2004 Team Leader: L. E. Ellershaw, Senior Reactor Inspector, Engineering Branch Inspectors: C. E. Johnson, Senior Reactor Inspector, Engineering Branch J. M. Mateychick, Reactor Inspector, Engineering Branch W. M. McNeill, Reactor Inspector, Engineering Branch W. C. Sifre, Reactor Inspector, Engineering Branch Approved by: Charles S. Marschall, Chief Engineering Branch Division of Reactor Safety
-2-SUMMARY OF FINDINGS IR 05000416/2004006; 01/12/2004 through 02/19/2004, Grand Gulf Nuclear 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 Grand Gulf Nuclear 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 (containment spray and shutdown cooling modes), 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 6 evaluations and 12 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
-2-NRC approval. These evaluations had been performed since the last NRC inspection of 10 CFR 50.59 activities.
The team reviewed an additional 14 licensee-performed 10 CFR 50.59 screenings, in which the licensee personnel 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 team reviewed and evaluated the most recent licensee 10 CFR 50.59 program self assessment and 16 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 as consistent with actions required to meet, prevent, and/or mitigate design basis accidents. The team also considered requirements and commitments identified in the Updated 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 (containment spray and shutdown cooling modes), 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 three 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 in-service tests.
b. Findings No findings of significance were identified.
4OA6 Management Meetings Exit Meeting Summary The inspection findings were acknowledged during an exit meeting presented by the team leader on January 30, 2004, to Mr. G. A. Williams, and other members of licensee management staff and during a subsequent telephone call on February 19, 2004, with Mr. M. Krupa, Director, Nuclear Safety Assurance. The team leader confirmed that proprietary information had not been presented to the team for review.
ATTACHMENT PARTIAL LIST OF PERSONS CONTACTED Licensee:
W. Abraham, Licensing Specialist, Nuclear Safety Assurance C. Bottemiller, Manager, Plant Licensing, Nuclear Safety Assurance R. A. Courtney, Manager, Corrective Action and Assessment L. Eaton, Engineer, System Engineering R. Fuller, Mechanical Engineer, Design Engineering M. Humphries, Supervisor, Design Engineering D. Jones, Supervisor, System Engineering M. Krupa, Director, Nuclear Safety Assurance Plant Licensing G. Lantz, Electrical, Instrumentation & Controls Engineer, Design Engineering J. Roberts, Director, Nuclear Safety Assurance D. Wiles, Director, Engineering G. Williams, Vice President, Operations D. Wilson, Supervisor, Design Engineering-Mechanical M. Withrow, Manager, Nuclear Engineering R. Wright, Engineering Supervisor, Quality Assurance H. Yeldell, Manager, Design Engineering NRC T. Hoeg, Senior Resident Inspector G. Miller, Resident Inspector DOCUMENTS REVIEWED Calculations NUMBER DESCRIPTION REVISION Q 2.3.3 Sizing of Hydrogen Recombiners B XC-Q1E61-01005 Design Basis Hydrogen Analysis with 10x10 Reload 0 Fuel 3.9.3 Standby Gas Treatment System (SGTS) Sizing 1 3.9.8 Standby Gas Treatment System Drawdown Time 1 Calculation 3.9.1 SGTS Infiltration Due to Pipe Breaks 2 XC-Q1111-98017 LOCA Dose Analysis with Revised Source Term 1
-2-NUMBER DESCRIPTION REVISION MC-Q1P41- Determination of Minimum Allowable SSW Flows 4 97020 (LOCA Lineup) to Safety-Related Heat Exchangers MC-Q1E12- Residual Heat Removal System (E12) Hydraulic 1 93008 Analysis 1.1.53-Q, and ECCS Pumps NPSH Calculation 11/20/97 Supplement 1 Drawings NUMBER DESCRIPTION REVISION M-1091 P & I Diagram Combustible Gas Control Systems 32 E-1160-049 Nuclear Steam Supply Shutoff Radwaste System Trip 9 Logic Valves Unit 1 E-1253-006 Schematic Diagram T41 Auxiliary Building Ventilation 7 System Auxiliary Building Ventilation System Isolation Valve F007 Unit 1 E-1254-001 Schematic Diagram T42 Fuel Handling Area 6 Ventilation System Isolation Valve F004 Unit 1 E-1254-002 Schematic Diagram T42 Fuel Handling Area 7 Ventilation System Isolation Valve F011 Unit 1 E-1257-001 T-48 Schematic Diagram Standby Gas Treatment 7 System A Trip Circuit Unit 1 E-1257-002 T-48 Schematic Diagram Standby Gas Treatment 7 System Charcoal Filter Train D001A Unit 1 E-1257-003 T-48 Schematic Diagram Standby Gas Treatment 7 System Inlet Valve F023 Unit 1 E-1257-015 T-48 Schematic Diagram Standby Gas Treatment 11 System Filter Train Heater 0001A Unit 1 E-1257-023 T-48 Schematic Diagram Standby Gas Treatment 7 System Flow Control Damper F001-A Unit 1
-3-NUMBER DESCRIPTION REVISION E-1257-024 T-48 Schematic Diagram Standby Gas Treatment 8 System Flow Control Damper F012-B Unit 1 E-1257-003 T-48 Schematic Diagram Standby Gas Treatment 7 System Inlet Valve F023 Unit 1 M-1102A Process & Instrument Diagram Standby Gas 22 Treatment System Unit 1 M-1102B Process & Instrument Diagram Standby Gas 6 Treatment System Unit 1 SFD-1102 System Flow Diagram Standby Gas Treatment System 3 Unit 1 C-1045A Unit 1 Containment Concrete & Misc Steel Plan at EL 8 161-10" and EL 170-0" C-1053C/D Unit 1 Containment Concrete-Intermediate Wall, 6/8 Elevations to 210-10" Miscellaneous Documents NUMBER DESCRIPTION REVISION LDC No. 2001-038 Change to Tech Spec Bases B 3.6.3.3 and UFSAR 0 Sections 1.2.2.4.9.4 and 6.2.5.2.1 MP&L Letter AECM- Transmittal of Proposed Changes to Grand Gulf 6/14/83 83/0338 Technical Specifications NRC Letter Amendment No. 8 to Facility Operating License No. 8/8/83 NPF-13 - Grand Gulf Nuclear Station, Unit 1 NPF-29 Grand Gulf Station Facility Operating License Amend. 158 Engineering Report Engineering Report for Cycle 13 Reload Summary 0 No. GGNS-02-0006 USNRC Letter No. Issuance of Amendment No. 120 to Facility 2/21/95 GNRI-95/00044 Operating License No. NPF-29 - Grand Gulf Nuclear Station, Unit 1 (TAC No. M88101)
-4-Miscellaneous Documents NUMBER DESCRIPTION REVISION Training Manual Standby Gas Treatment System (SGTS) - T48 3 GLP-OPS-T4801 9645-M-632.0 Bechtel Technical Specification for Standby Gas 15 Treatment System for Mississippi Power and Light Company Grand Gulf Nuclear Station Units 1 and 2 Grand Gulf, Mississippi Assessment Report Evaluation of GGNS 10CFR50.59 Changes, Tests, 12/30/03 or Experiments Program Operations Training Residual Heat Removal System-E12 00 Lesson Plan, GLP-OPS-E1200 Plant Operations Control and Use Of Operations Section Directives 40 Manual, 02-S-01-2 GLP-OPS-E1200 Residual Heat Removal (RHR) System-E12 00 GGNS-SDC-E61 Combustible Gas Control System (E61) 1 SDC-T48 Secondary Containment and Standby Gas 1 Treatment System (T48)
Generic Letter 96-06 Assurance of Equipment Operability and 9/30/96 and and Supplement 1 Containment Integrity During Design-Basis Accident 11/13/97 Conditions Engineering Report Generic Letter 96-06 Evaluation of Drywell and 1 GGNS-97-0002 Containment Penetrations and Safety Related Piping
-5-Condition Reports CR-GGN-1997-0403 CR-GGN-2001-1829 CR-GGN-2002-02355 CR-GGN-2004-0307 CR-GGN-2001-1884 CR-GGN-2003-00017 CR-GGN-2001-00116 CR-GGN-2002-01280 CR-GGN-2003-02822 CR-GGN-2001-01412 CR-GGN-2002-01386 CR-GGN-2003-03236 CR-GGN-2004-00185 CR-GGN-1995-00041 CR-GGN-2003-00940 CR-GGN-2004-00305 CR-GGN-1995-00089 CR-GGN-2004-00085 CR-GGN-1998-00995 CR-GGN-2003-03328 CR-GGN-2004-00123 CR-GGN-2000-0538 CR-GGN-2003-03372, CR-GGN-2003-01240 CR-GGN-2001-01681 CR-GGN-2003-03373, CR-GGN-2003-01744 CR-GGN-2001-00490 CR-GGN-2003-03490 CR-GGN-2003-02922 CR-GGN-2001-0364 CR-GGN-2002-01754 CR-GGN-1995-00397 CR-GGN-2001-0854 CR-GGN-2002-01936 CR-GGN-1997-00332 CR-GGN-2001-01011 CR-GGN-2002-01954 CR-GGN-1998-00392 CR-GGN-2001-1785 CR-GGN-2002-01960 CR-GGN-1999-01640 CR-GGN-2003-00226 CR-GGN-2002-02106 CR-GGN-1999-01939 CR-GGN-2003-00376 CR-GGN-2002-01936 CR-GGN-2000-00276 CR-GGN-2001-01019 CR-GGN-2002-02237 CR-GGN-2000-01258 CR-GGN-2003-00600 CR-GGN-2002-02246 CR-GGN-1995-00197 CR-GGN-2003-00625 CR-GGN-2003-00854 CR-GGN-1995-00221 CR-GGN-2004-00289 CR-GGN-2004-00301 CR-GGN-2001-00955 CR-GGN-2001-01216 CR-GGN-2002-01493 CR-GGN-2003-01340 CR-GGN-2003-02378 CR-GGN-1999-01147 CR-GGN-1999-01256 Procedures NUMBER DESCRIPTION REVISION 04-1-01-E61-1 System Operating Instruction Combustible Gas 39 Control System 05-S-01-EP-2 Emergency Procedure - RPV Control 34 EP-2 Flowchart Emergency Procedure Flowchart - RPV Control 4/24/01 05-S-01-EP-3 Emergency Procedure - Containment Control 26 EP-3 Flowchart Emergency Procedure Flowchart - Containment 12/17/98 Control 05-S-01-SAP-1 Severe Accident Procedure - Severe Accident 3 SAP-1 Flowchart Core Debis Has Breached The RPV 12/9/02 SAP-2 Flowchart RPV Water Level Can Be Restored and 12/9/02 Maintained Above -167 in
-6-NUMBER DESCRIPTION REVISION SAP-3 Flowchart RPV Water Level Can Be Maintained Above - 12/9/02 317 in SAP-4 Flowchart RPV Injection Can Be Restored and Maintained 12/9/02 Greater Than MDRIR SAP-5 Flowchart RPV Injection Cannot Be Restored and 12/9/02 Maintained Greater Than MDRIR- Containment Pressure Is Within PSP SAP-6 Flowchart RPV Injection Cannot Be Restored and 12/9/02 MaintainedGreater Than MDRIR- Containment Pressure Is Not Within PSP 06-OP-1E61-R-0005 Primary Containment Hydrogen Recombiner A 102 Heatup Test 06-OP-1E61-R-0006 Primary Containment Hydrogen Recombiner 102 BHeatup Test 06-OP-1E61-R-0009-01 Hydrogen Igniter System Heatup Test (System 103 A)
06-OP-1E61-R-0009-02 Hydrogen Igniter System Heatup Test (System 103 B)
06-OP-1E61-R-0009-03 Hydrogen Igniter System Heatup Test (System 103 A & B)
06-OP-1E61-R-0009-04 Hydrogen Igniter System Heatup Test (System 102 A & B)
06-OP-1E61-R-0003 Drywell Purge System Operability 106 06-IC-1E61-Q-1004-01 Containment Hydrogen Analyzer (PAM) 102 Calibration -Channel A 06-IC-1E61-Q-1004-02 Containment Hydrogen Analyzer (PAM) 102 Calibration - Channel B 06-IC-1E61-Q-1004-03 Drywell Hydrogen Analyzer (PAM) Calibration - 102 Channel A
-7-NUMBER DESCRIPTION REVISION 06-IC-1E61-Q-1004-04 Drywell Hydrogen Analyzer (PAM) Calibration - 102 Channel B 06-OP-1E61-M-0001 Post-LOCA Drywell Vacuum Breaker 102 Operability 06-OP-1E61-Q-0003-01 Drywell Purge System A Operability 109 06-OP-1E61-Q-0003-02 Drywell Purge System B Operability 109 06-IC-1E61-M-1002-02 Containment/Drywell Differential Pressure 101 Functional Test Channel B 06-IC-1E61-M-1002-04 Containment/Drywell Differential Pressure 101 FunctionalTest Channel D 06-IC-1E61-M-1002-05 Containment/Drywell Differential Pressure 101 FunctionalTest Channel E 06-IC-1E61-M-1002-05 Containment/Drywell Differential Pressure 101 FunctionalTest Channel E 06-IC-1E61-M-1002-06 Containment/Drywell Differential Pressure 101 Functional Test Channel F 06-IC-1E61-M-1002-07 Containment/Drywell Differential Pressure 101 FunctionalTest Channel G 06-IC-1E61-M-1002-08 Containment/Drywell Differential Pressure 101 Functional Test Channel H 01-S-06-2 Conduct of Operations 117 NPEAP 318 Design Engineering Criteria 6 POMSOI 04-1-01-T48-1 Standby Gas Treatment 29 E-1254-003 Schematic Diagram T42 Fuel Handling Area 5 Ventilation System Isolation Valve F019 Unit 1 POMSR 06-ME-1T48-R- In-place Testing of Standby Gas Treatment 104 0005 Filtration System
-8-NUMBER DESCRIPTION REVISION POMSR 06-OP-1T48-Q- Standby Gas Treatment System A Valve Test 102 0002 POMSR 06-OP-1T48-R- Surveillance Procedure Standby Gas 107 0002 Treatment A Logic and Vacuum Test SOI 04-1-01-E12-2 Shutdown Cooling and Alternate Decay Heat 100 Removal Operation SOI 04-1-01-E12-1 Residual Heat Removal System 122 01-S-07-39 Inservice Testing 103 17-S-05-1 Performance and System Engineering 106 Instruction - Local Leak Rate Test Program 07-S-14-395 Safety and Relief Valve Program-Safety 6 Related Engineering Requests ER-GG-2002-0095-000 ER-GG-97-0022-00, -01, -02 ER-GG-02002-0095-001 ER-GG-2001-0410-000 ER-GG-02002-0123-000 ER-GG-2000-0083-000 ER-GG-91-6166 Surveillance Procedures 06-OP-1000-D-0001 06-OP-1E12-R-0023 06-OP-1E12-M-0002 06-OP-1E12-M-0001 06-IC-1M71-R-0003 06-OP-1E12-M-0023 06-OP-1E12-M-0001 06-OP-1000-D-0001 06-ME-1000-R-0003 06-OP-1E12-Q-0023 06-OP-1000-D-0001 Inservice Test Program Documents CEP-IST-1, Inservice Testing Bases Document, Revision 3 CEP-IST-2, Inservice Testing Plan, Revision 3 CEP-IST-3, Inservice Testing Cross-Reference Document, Revision 1
-9-Plant Operations Manuals 06-ME-1M61-V-0001, Surveillance Procedure - Local Leak Rate Test Low Flow Air, Revision 107.
06-OP-1E12-Q-0005, Surveillance Procedure - LPCI/RHR Subsystem A MOV Functional Test, Revision 104.
06-OP-1E12-Q-0006, , Surveillance Procedure - LPCI/RHR Subsystem B MOV Functional Test, Revision 105.
06-OP-1E12-Q-0023, Surveillance Procedure - LPCI/RHR Subsystem A Quarterly Functional Test, Revision 112.
06-OP-1E12-Q-0024, Surveillance Procedure - LPCI/RHR Subsystem B Quarterly Functional Test, Revision 108.
06-OP-1T48-Q-0002, Surveillance Procedure - Standby Gas Treatment System A Valve Test, Revision 102.
06-OP-1T48-Q-0003, Surveillance Procedure - Standby Gas Treatment System B Valve Test, Revision 102.
06-OP-1E12-C-0015, Surveillance Procedure - LPCI B Shutdown Valve Test, Revision 105.
06-OP-1E12-C-0014, Surveillance Procedure - LPCI A Shutdown Valve Test, Revision 104.
06-OP-1E12-C-0013, Surveillance Procedure - RHR B Shutdown Cooling Mode Valve Test, Revision 104.
06-OP-1E12-C-0012, Surveillance Procedure - RHR A Shutdown Cooling Mode Valve Test, Revision 105.
Valves Reviewed:
RHR System 1E12F042A 1E12F009 1E12F003A 1E12F024A 1E12F042B 1E12F008 1E12F003B 1E12F024B 1E12F006A 1E12F053A 1E12F004A 1E12F028A 1E12F006B 1E12F053B 1E12F004B 1E12F028B 1E12F027B 1E12F027A
-10-Standby Gas Treatment System T48F005 T48F023 T48F025 T48F024 T48F026 Pumps Reviewed:
1E12C002A - RHR/LPCI Pump C 1E12C002B - RHR/LPCI Pump B 1E12C003A - RHR Loop A Jockey Pump 1E12C003B - RHR Loop B Jockey Pump Preventive Maintenance Records A Seal Cooler, Perform Thermal Performance Test for the RHR C Pump Seal Cooler, Task 28140 1E12B001B/2B, Perform Thermal Performance Test for the RHR B Pump Seal Cooler, Task 28123 A Jockey Pump, Clean, inspect and lube motor. Clean and inspect motor breaker, Task 3757 B Jockey Pump, Perform pump coupling maintenance and change oil. Vibration Engineer to perform testing six weeks prior to task schedule date to allow development of work scope, Task 6268 A Motor, Clean, inspect and meg motor Change Motor Bearing Oil, Task 3453 A Motor, Replace upper and lower oil level sight glass O-rings, Task 24303 A Motor, Perform 10 year motor inspection, Task 3452 A Pump, Replace/rework mechanical seal, Task 28738 B Pump, Replace/rework mechanical seal, Task 28739 B Motor, Perform 10 year motor inspection, Task 3742 B Motor, Calibration of velocity transducers on 1E12C002B Motor, Task 24928
-11-Work Orders 00162085 00209637 50326558 01 50488399 01 00164319 50338362 01 50767002 05 50571424 01 00183313 50339439 01 50830772 50572236 01 00182214 001752220 50964800 50573467 01 00209636 50311008 50964801 001752221 Maintenance Action Items MAI# 275010 MAI# 306648 MAI# 318493 MAI# 276030 MAI# 308059 MAI# 313914 MAI# 286566 Safety Evaluations SE-2001-0028 SE 2002-0006 SE 2003-0001 SE 2002-0002 SE 2002-0007, Revision 2 SE 2003-0002 SE 2002-0005 SE 2002-0008, Revision 1 Safety Evaluation Sceenings Calculation XC-Q1J11-98017, Revision 1 Calculation 3.9.12, Revision 2 Engineering Report GGNS-99-0026, Revision 0 Procedure 01-S-02-3, Revision 107 Procedure 04-1-01-T48-1, Revision 29 Procedure 06-ME-1T48-R-0005, Revision 104 Procedure 06-OP-1E12-Q-0023, Revision 112 Procedure 06-OP-1E12-Q-0024, Revision 108 Procedure 06-OP-1T48-R-0001, Revision 105 Procedure 318, Revision 6 Procedure 17-S05-1, Revision 106 Software Check+/CTP 2.0.1 Software SCR-GGN-2001-83 Software ER-GG-2000-0859-002