IR 05000361/2004007
| ML042390380 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 08/26/2004 |
| From: | Clark J Division of Reactor Safety IV |
| To: | Ray H Southern California Edison Co |
| References | |
| IR-04-007 | |
| Download: ML042390380 (22) | |
Text
August 26, 2004
SUBJECT:
SAN ONOFRE NUCLEAR GENERATING STATION - NRC SAFETY SYSTEM DESIGN AND PERFORMANCE CAPABILITY INSPECTION REPORT 05000361/2004007; 05000362/2004007
Dear Mr. Ray:
On July 30, 2004, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your San Onofre Nuclear Generating Station, Units 2 and 3. The enclosed Safety System Design and Performance Capability Team report documents the inspection findings, which were discussed at the conclusion of the inspection with Mr. Dwight Nunn 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.
The team reviewed selected procedures and records, observed activities and interviewed personnel.
Based on the results of this inspection, no findings of significance were identified.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be made 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/
J. Clark, P.E., Chief Engineering Branch Division of Reactor Safety
Southern California Edison Co.
-2-Dockets: 50-361; 50-362 Licenses: NPF-10; NPF-15 Enclosure:
NRC Inspection Report 05000361/2004-007; 05000362/2004-007 cc w/enclosure:
Chairman, Board of Supervisors County of San Diego 1600 Pacific Highway, Room 335 San Diego, CA 92101 Gary L. Nolff Power Projects/Contracts Manager Riverside Public Utilities 2911 Adams Street Riverside, CA 92504 Eileen M. Teichert, Esq.
Supervising Deputy City Attorney City of Riverside 3900 Main Street Riverside, CA 92522 Joseph J. Wambold, Vice President Southern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92674-0128 David Spath, Chief Division of Drinking Water and Environmental Management California Department of Health Services P.O. Box 942732 Sacramento, CA 94234-7320 Michael R. Olson San Onofre Liaison San Diego Gas & Electric Company P.O. Box 1831 San Diego, CA 92112-4150
Southern California Edison Co.
-3-Ed Bailey, Chief Radiologic Health Branch State Department of Health Services P.O. Box 997414 (MS 7610)
Sacramento, CA 95899-7414 Mayor City of San Clemente 100 Avenida Presidio San Clemente, CA 92672 James D. Boyd, Commissioner California Energy Commission 1516 Ninth Street (MS 34)
Sacramento, CA 95814 Douglas K. Porter, Esq.
Southern California Edison Company 2244 Walnut Grove Avenue Rosemead, CA 91770 Dwight E. Nunn, Vice President Southern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92674-0128 D
SUMMARY OF FINDINGS
IR 05000361/2004-07, 05000362/2004-07; 07/12 -30/2004; San Onofre Nuclear Generating
Station, Units 2 and 3; Safety System Design and Performance Capability; Permanent Plant Modifications; Evaluations of Changes, Tests, or Experiments.
The report covered a 2-week period of inspection on site by a team of four region-based engineering inspectors and one consultant. 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 200
NRC-Identified and Self-Revealing Findings
No findings of significance were identified.
REPORT DETAILS
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 San Onofre Nuclear Generating 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 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
a. Inspection Scope
The minimum sample size for this procedure is 5 evaluations and 10 screenings. The team reviewed 10 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.
The team reviewed 12 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 3 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 21 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.
1R17 Permanent Plant Modifications
.2 Biennial Review
a. Inspection Scope
The minimum sample size for this procedure is 5 to10 permanent plant modifications.
The team reviewed 10 permanent plant modifications and associated engineering change package documentation (e.g., implementation reviews, safety evaluation applicability determinations, and screenings) to verify that they were performed in accordance with regulatory requirements and plant procedures. The inspectors reviewed procedures governing plant modifications to evaluate the effectiveness of the programs for implementing modifications to risk-significant systems, structures, and components, such that these changes did not adversely affect the design and licensing basis of the facility. Procedures and permanent plant modifications reviewed are listed in the attachment to this report. The inspectors interviewed the cognizant design and system engineers for the identified modifications to gain their understanding of the modification packages.
b. Findings
No findings of significance were identified.
1R21 Safety System Design and Performance Capability
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 auxiliary feedwater system. The team also reviewed the plant protection system, which includes the reactor protection system and core protection calculator, as well as the control room emergency air clean up system. The primary review prompted parallel review and examination of support systems, such as, power, instrumentation & controls, cooling and related structures and components.
.1 System Requirements
a. Inspection Scope
The team inspected the following attributes of the selected systems:
- (1) process medium (water, steam, air, electrical signal),
- (2) energy sources,
- (3) control systems, and
- (4) equipment protection. The team examined the procedural instructions to verify that instructions were 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, a parallel review and examination of support systems 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 selected systems to verify that the capabilities of the systems were periodically verified. The team also reviewed system health reports, as well as, a sample of the governing procedures and documentation for the control of calculations that were translated into values used in plant procedures. In addition, the team performed walkdowns of the selected systems to ascertain the material condition of the systems.
The team also reviewed the operation of the systems by reviewing normal, abnormal, and emergency operating procedures. The review included the Updated Final Safety Analysis Reports, technical specifications, design calculations and drawings.
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 the selected systems that were identified by licensee personnel in the corrective action program to evaluate the effectiveness of corrective actions related to design issues and aging hardware. The sample included open and closed action requests and their disposition via maintenance orders, apparent cause evaluations, spurious action documentation, or field support notes as documented in the licensees corrective action program. The sample covered the past 3 years and the documents reviewed 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.
b. Findings
No findings of significance were identified.
.4 System Walkdowns
a. Inspection Scope
The team performed walkdowns of the accessible portions of the selected systems.
The team focused on the installation, configuration, and visible material condition of equipment and components. 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 condition and preservation of systems and equipment, and
- The conformance of the currently-installed system configuration 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 selected systems 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 original system design to perform the design basis functions during normal, accident and post-accident conditions. The review included: design basis documents; specifications; reliability calculations; instrument uncertainty/setpoint calculations; uncertainty calculations related to emergency operating instruction action levels; and schematic diagrams. The adequacy of the design and maintenance of selected support systems was also reviewed.
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 selected systems and support systems. The review included the results of surveillance tests required by the technical specifications and a selective review of inservice tests.
b. Findings
No findings of significance were identified.
4OA6 Exit Meeting Summary
The inspection findings were presented by the team leader during an exit meeting on July 30, 2004, to Mr. Dwight Nunn 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 J. Aguirre, Core Protectin Calculator Harware Engineer D. Axline, Licensing Engineer B. Bostian, CRD Lead Engineer D. Breig, Maintenance Engineering Manager S. Foglio, Reactor Protection System Engineer C. Hays, Computer Engineering Supervisor T. Herring, Configuration Engineering Manager E. Kimoto, Auxiliary Feedwater Lead Engineer M. Love, Maintenance Manger J. A. Madigan, Health Physics Manager C. McAndrews, Nuclear Oversight Manager M. McBrearty, Engineer, Nuclear Regulatory Affairs D. McBride, Supervisor, Maintenance S. Medling, Nuclear Regulatory Affairs Manager D. E. Nunn, Vice President, Engineering & Technical Services R. Osborne, 50.59 Project Manager J. Ramsdell, Systems Engineer S. Root, Special Projects Manager M. P. Short, System Engineering Manager A. Scherer, Manager, Nuclear Regulatory Affairs S. Swoope, Fuels Engineer R. Waldo, Station Manager J. Wambold, Vice President, Nuclear Generation P. Wilkens, Control Room Emergency Air Clean up System Lead Engineer C. E. Williams, Compliance Engineer T.R. Yackle, Design Engineering Manager T. Yee, Structural Engineering NRC C. Osterholtz, Senior Resident Inspector M. Sitek, Resident Inspector
DOCUMENTS REVIEWED
Calculations
00000-ICE-3652, Revision 00,Reliability Prediction Calculation for Core Protection Calculator
System (CPC), May 16, 1977.
J-BBB-019, Revision 0, CPC Temperature Input Drift Analysis, February 28, 1994.
J-BBB-031, Revision 1, TLU [Total Loop Uncertainty] for Wide Range Pressurizer Pressure
-2-
Indication & Evaluation of Wide Range Pressurizer Pressure Indications for EOIs.
J-BBB-091, Revision 0, Pressurizer Pressure Narrow Range Indication Loop Uncertainties,
9/30/98; CCN N-1 dated June 11, 2000.
NFM-2-SP-1313, Revision 00, SONGS-2 Cycle 13 CPC Reload Data Block Update Analysis,
February 2, 2004.
NFM-2-SP-1322, Revision 00, SONGS-2 Cycle 13 MSQUA Post-Processor Analysis,
February 2, 2004.
SO23-944-C50, Revision 3, 2/13/97: CE NPSD-570-P,/1370-ICE-3670/1470-ICE-3698,
Revision 7, SONGS 2 & 3 Plant Protection System Setpoint Calculation, CCN 9 dated
September 25, 1998 [CE Proprietary].
SO23-944-C90-2, Revision 2: CE NPSD-704-P/ 1370/1470-ICE-36199, SONGS Units 2 and 3
CPC Input Data Channel Uncertainty Calculation, Revision 06, 6/13/1996 [CE Proprietary].
J-SAA-001, Toxic Gas Isolation System Setpoints, Revision 1
J-SPA-099, Control Room/Fuel Handling Building Monitor Concentration Ranges, Revision 0
J-SRA-179, Control Room/Fuel Handling Building Set Points, Revision 0 with Calculation
Change Notices 4 and 5
M-0073-041, Auxiliary Building Controlled Area Elevation 30' heat Load and Equipment Sizing
Normal and Emergency, Revision 8 with Calculation Change Notices 20, and 22 - 31 and
Interim Calculation Change Notices 17 and 23 -26
M-0073-133, Control Room Envelope Volume Calculation to be Used for the Tracer Gas
Testing, Revision 0
M-0073-095, Infiltration into the Control Room Envelope from Surrounding Areas, Revision 3
with Calculation Change Notice 2
N-4072-001, Fuel Handling Accident Inside Fuel Handling Building - Control Room and Offsite
Doses, Revision 6
N-4090-012, Toxic Gas Concentrations in the Control Room for Toxic Gas Isolation System
Monitored Chemicals, Revision 0
M-8910SP-2HV9307, REV 2, ICCN C-1, Generic Letter 89-10 Setpoint Calculation for
2HV9307
M-1203-161-04A, ICCN C-1, Radwaste CCW Line 1203-161 Unit 3
M-1203-161-02A, ICCN C-1, Radwaste CCW Line 1203-161
M-37632, "EQ Document Package for Auxiliary Feedwater Pump Motor," Revision 3
-3-
M-37624, "EQ Document Package for Axivane Fan inside the Auxiliary Feedwater Pump
Rooms," Revision 9
M37706, "EQ Document Package for Motorized Valve Actuators inside the Auxiliary Feedwater
Pump Room," Revision 11
M-0056-039, "AFW System Performance with 40 F Condensate," Revision 0
M56-38, " Auxiliary Feedwater Pump System Performance," Revision 0, CCN 1, CCN 2 & N-1,
and CCNT 3
M56.16, "Auxiliary Feedwater System Analysis," Revision 1
M50-2, "Condensate Storage Tank Capacity," Revision 3
M-0056-018, "Auxiliary Feedwater Pump NPSH Requirements," Revision 3
M-DSC-248, "Auxiliary Feedwater Steam Flows for Reduced Line Capacity," Revision 0
M-0050-017, "BTP RSB 5-1 Condensate Inventory," Revision 3
M-74-06, "AFW Pump Room Heat Load Calculation," Revision 1
M-0060-008, "Restriction of AFW Turbine Trip/Throttle Valve," Revision 0
Action Requests
000900187
010100770
010500462
010501377
010600258
010600605
010600694
010800440
010800825
010801221
010900531
010900829
011000126
011000765
011001127
011200343
20100971
20200468
20200745
20301482
20400823
20500474
20501682
20602171
20700544
20700761
20800755
20800840
20801668
20901177
21000630
21000631
21100202
21100715
21101166
21200579
21201342
030100500
030101056
030300007
030300217
030300835
030400155
030401421
030401471
030600145
030600288
030600552
030600856
030601402
030700070
030700105
030700364
030701532
030801131
030900049
031000738
031000888
031100693
031100693
031100694
031100964
031101052
031200453
031200636
031201634
031201694
031201697
040100147
040101030
040101548
040200972
040201564
040201800
040202185
040300464
040301598
040400212
040400227
040401098
040501000
040501221
040501394
040501562
040501673
040600147
040600608
040601053
040700829
040700842
040700865
040700866
040700873
040701424
040701512
040701564
040701601
040701603
040701631
991101109
-4-
Design Basis Documents
DBD-SO23-710, Plant Protection System, Revision 7, December 2, 2002.
00000-ICE-3001, General Engineering Specification for a Plant Protection System,
Revision 3, May 13, 1976 [CE Proprietary].
1370-ICE-3019, Project Engineering Specification for Auxiliary Protective Cabinet for San
Onofre Nuclear Generating Station Units 2 & 3, Revision 00, December 8, 1975 [CE
Proprietary].
SO23-907-125-0, Technical Manual for Control Element Assembly Position Isolation
Assembly, November 17, 1977.
DBD-SO23-780, Design Basis Document, Auxiliary Feedwater System, Revision 6
Drawings
UFHA Figure 8-6, Revision 10, Auxiliary Building (Control Area) Unit 2 &3 Elevation 30'-0" Fire
Protection Features.
SO23-944-545, Sheet 350, Revision 28, PPS Trip Channel Bypass, Wiring Schematic.
SO23-944-545, Sheet 351, Revision 28, PPS Trip Channel Bypass, Wiring Schematic.
SO23-944-545, Sheet 366, Revision 28, Test Coils, AB Matrix, Wiring Schematic.SO23-944-
545, Sheet 415, Revision 28, Matrix Relay Card, Schematic Diagram.
SO23-944-600, Sheet 1, Revision 5, Plant Protection System Simplified Functional Diagram.
W-91X0636-D-SO3, Sheet 17, Revision 6, Wiring Diagram for Auxiliary Protective Cabinet
3L91.
W-91X0636-D-S03, Sheet 19, Revision 7, Wiring Diagram for Auxiliary Protective Cabinet
3L91.
31579, Sheet 4, Revision 9, Wiring Diagram, Control Building, NSSS Protection Area (Vault)
2LO91.
2021, Sheet 36, Revision 6, Wiring Diagram, Penetration Area, Pen. 47 Type 3-2.
2085, Sheet 8, Revision 6, Wiring Diagram, Containment Structure, Reactor Head.
10101. Auxiliary Building Floor Plan Elevation 30'-0", Revision 30
40098, Air Flow Diagram - Train A Control Building - Elevation 30'-0", Revision 10
-5-
40096, Air Flow Diagram - Train B Control Building - Elevation 30'-0", Revision 19
40173A, Process and Instrument Diagram Control Room Complex Heating, Ventilation, and Air
Conditioning System No. 1510, Revision 22
40173B, Process and Instrument Diagram Control Room Complex Heating, Ventilation, and Air
Conditioning System No. 1510, Revision 9
40173C, Process and Instrument Diagram Control Room Complex Heating, Ventilation, and Air
Conditioning System No. 1510, Revision 23
40173D, Process and Instrument Diagram Control Room Complex Heating, Ventilation, and Air
Conditioning System No. 1510, Revision 12
40173X, Process Key Plan Control Room Complex Heating, Ventilation, and Air Conditioning
System No. 1510, Revision 2
Root-Cause/Apparent Cause Evaluation Reports
ACE 010900531-02
ACE 020800755-04
ACE 020801305-04
ACE 030401421-11
Procedures
SO23-V-12.2.32, CPC/CEAC Off-Line Circuit Board and Component Testing, Revision 0,
June 5, 2002.
SO23-XV-2, Troubleshooting Plant Equipment and Systems, Revision 0.
SO23-XXXVII-4.1, CPC/CEAC/COLSS Software Control and Documentation, Revision 0,
September 10, 2001.
SO23-XXXVII-4.7, Control of CPC Addressable Constants, Revision 0, February 25, 2004.
SO23-944I-20-1, Energy Incorporated Procedure SCAL-3-902, Instructions for Making
Component Substitution for CPCS Spares, Revision 1, January 7, 1986.
SO23-1-5.1, Auxiliary Building Emergency Heating, Ventilation, and Air Conditioning,
Revision 4 with Temporary Change Notice 4-1
SO23-3-2.29, Toxic Gas Analyzer Operation, Revision 7
SO23-3-3.12 ISS 2, Integrated Engineered Safety Feature System Refueling Test,
Revision 21
-6-
SO23-3-3.20, Monthly CREACUS Test, Control Room Exercise Run and Emergency Chill
Water Systems Minimum Operability Verification, Revision 16 with Temporary Change
Notice 16-5
SO23-3-3.20.1, Control Room Emergency Air Cleanup System 18-Month Surveillance,
Revision 13 with Temporary Change Notice 13-5
SO23-3-3.21, Common Shiftly Surveillance, Revision 28
SO23-5-2.25, Plant Heating, Ventilation, and Air Conditioning 83 Alarm Response Procedure,
Revision 5 with Temporary Change Notice 5-6
SO23-5-2'7, Annunciator Panel 60A Emergency Heating, Ventilation, and Air Conditioning,
Revision 5 with Temporary Change Notice 5-52
SO23-I-2.44, CREACUS-Control Room Emergency Air Cleanup System Operation and
Operability Test Surveillance, Revision 7
SO23-II-1.15, Surveillance Requirement Toxic Gas Isolation System Train A Channel Function
Test and Channel Calibration, Revision 11 with Temporary Change Notice 11-3
SO23-II-1.15.1, Surveillance Requirement Toxic Gas Isolation System Train B Channel
Functional Test and Channel Calibration, Revision 4 with Temporary Change Notice 4-4
SO23-II-8.22, Surveillance Requirement Toxic Gas Isolation System Response Test and
Channel Functional Test, Revision 8 with Temporary Change Notice 8-4
SO23-XXV-4.3, Surveillance Requirement Control Room Isolation System Train A Loop
2/3RE7824G1 Channel Calibration, Response Time Test, and Channel Functional Test,
Revision 5
SO23-XXV-4.37, Surveillance Requirement Control Room Isolation System Train B Loop
2/3RE7824G2 Channel Calibration, Response Time Test, and Channel Functional Test,
Revision 6
SO23-XXV-4.8, Surveillance Requirement Control Room Isolation System In-Duct Radiation
Monitor Train A Loop 2/3RE7824G1 Channel Functional Test, Revision 4 with Temporary
Change Notice 4-1
SO23-XXV-4.9, Surveillance Requirement Control Room Isolation System In-Duct Radiation
Monitor Train B Loop 2/3RE7824G2 Channel Functional Test, Revision 4 with Temporary
Change Notice 4-4
SO123-XV-5, Nonconforming Material, Parts, or Components, Revision17
SO123-XX-1, Action Request/Maintenance Order Initiation and Processing, Revision 15
TCN 15-2
-7-
SO123-XV-52, Operability Assessments and Reportability Evaluations, Revision 5
SO23-5-1.5, Plant Shutdown From Hot Standby to Cold Shutdown, Revision 23
SO23-12-4, Steam Generator Tube Rupture, Revision 18
CFR 50.59 Evaluations
AR 000900187-02
AR 010100770-11
AR 010600258-02
AR 010600694-01
AR 010800825-02
AR 011000765-04
AR 011200343-07
AR 020200745-03
AR 020301482-25
AR 991101109-30
CFR 50.59 Screenings
AR 000900187-02
AR 010100770-25
AR 010100770-47
AR 010500462-03
AR 020301482-22
AR 020400823-34
AR 020400823-82
AR 021100202-13
AR 031101052-05
AR 040200972-08
AR 030300797-03
AR 030801458-05
CFR 50.59 Evaluation Exemptions
AR 010801221
AR 021000630
AR 040600608
Permanent Plant Modifications
MMP No. 2-6817.00SJ, Revise PPS and ESFAS Setpoints, Revision 0, August 17, 1990.
MMP No. 2&3-6828.00SJ, PPS and ESFAS Power Supply Replacement, Revision 0, April 26,
1993.
Engineering Change Packages
010800854-2
21000630-9
030601402-5
030601402-9
030801155-3
031201461-5
010800854-11
031201461-11
Special Test Report
MO 01101388000, Perform a one-time test of PPS bistable Relay/matrix relay LEDs for
reverse breakdown voltage of greater than 10 vdc for both Units 2 & 3 (Westinghouse
Technical Bulletin T0104)
-8-
Field Support Notes
FS 030300217-01
FS 030400155-01
FS 030600288-01
FS 030701532-01
FS 040101030-01
FS 040201800-01
FS 040300464-01
FS 011000126-02
Maintenance Orders
MO 03010989000
MO 03030035000
MO 03030707000
MO 04011364000
MO 04040267000
MO 01101388000
24-Month Control Room Positive Pressure Maintenance Orders
2072005001
11052044001
18-Month Operability Test Maintenance Orders
01031550000
01031551000
01103084000
2012285000
2080243000
2080244000
2080245000
03040760000
Corrective Maintenance Work Orders
01102831000
2010169000
2020808000
2061192000
2061254000
2061254001
2080918000
2100681000
2120946000
2120963000
03061005000
03063441000
03081022000
03081022001
03081669000
03081736000
03100858000
04010114000
04020232000
04032154000
81001172002
Maintenance Rule Evaluations
MRE 020800755-08
Spurious Actuation Reports
SA 030400155-02
SA 030700070-01
SA 040300464-02
-9-
Surveillance Test Reports
MO 04061816, Unit 2 CPC Channel D Surveillance, SO23-V-12.2.1, Surveillance Requirement,
Core Protection Calculators (CPCS) Functional Test (30 Day Staggered), performed 7/14/2004.
Surveillance Test Procedures
SO23-II-1.1.5, Revision 6, Surveillance Requirement, Reactor PPS Logic Matrix Functional
Test, Sections 6.3, 6.4, 11/6/03.
SO23-II-9.258, Revision 12, Plant Protection System Bistable Card and Variable Setpoint Card
Calibration, 11/12/03.
SO23-V-12.2.1, Revision 33, Surveillance Requirement, Core Protection Calculators (CPCS)
Functional Test (30 Day Staggered), 11/12/2003.
SO23-3-3.34, REV 15, Stop and Governor Valve On Line Test Data Documentation
SO23-3-3.34, REV 16, On Line Turbine Overspeed Protection Surveillance Low Pressure
Turbine Valves
SO23-3-3.34, REV 18, "Overspeed Protection Surveillance Low Pressure Turbine Valves
Surveillance Tests of Control Room Emergency Air Cleanup System
June 19, 2003
November 20, 2003
January 16, 2004
January 28, 2004
February 12, 2004
March 1, 2004
March 11, 2004
March 25, 2004
April 8, 2004
April 21, 2004
April 22, 2004
May 7, 2004
May 18, 2004
June 3, 2004
June 17, 2004
June 28, 2004
Miscellaneous Documents
SONGS System Health Report, Core Protection Calculator, Quarter 2003-1.
SONGS System Health Report, Core Protection Calculator, Quarter 2003-2.
SONGS System Health Report, Core Protection Calculator, Quarter 2003-3.
SONGS System Health Report, Core Protection Calculator, Quarter 2003-4.
SONGS System Health Report, Core Protection Calculator, Quarter 2004-1.
SONGS System Health Report, Core Protection Calculator, Quarter 2004-2.
-10-
SONGS System Health Report, Reactor Protection System, Quarter 2003-1.
SONGS System Health Report, Reactor Protection System, Quarter 2003-2.
SONGS System Health Report, Reactor Protection System, Quarter 2003-3.
SONGS System Health Report, Reactor Protection System, Quarter 2003-4.
SONGS System Health Report, Reactor Protection System, Quarter 2004-1.
SONGS System Health Report, Reactor Protection System, Quarter 2004-2.
SDR 01678-96143, Supplier Deviation Request, NUS Instruments, November 19, 1996.
NRC Generic Letter 96-01, Testing of Safety Related Logic Circuits Review Report,
Revision 1, Appendix
- A.
OE 021001304Property "Contact" (as page type) with input value "A.</br></br>OE 021001304" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., Operational Event Assessment, Part 21 Rosemount Model 1153 Differential
Pressure Transmitters, October 25, 2002.
SO23-306-1-44-0, Technical Report, Seismic Vibration Analysis of NMC Controls, Inc. Control
Room Relay Panels 2L-71 & 3L-71 (with field changes), May 16, 1978.
Draft Control Room Envelope Inleakage Testing at San Onofre Nuclear Generating Station
NRC Generic Letter 96-01 Testing of Safety Related Logic Circuits Review Report, Revision 0
Songs System Health Report Emergency Heating, Ventilation, and Air Conditioning 1st Quarter
2004
Songs System Health Report Emergency Heating, Ventilation, and Air Conditioning 4th Quarter
2003
Songs System Health Report Emergency Heating, Ventilation, and Air Conditioning 3d Quarter
2003
Songs System Health Report Radiation Monitor 1st Quarter 2004
Songs System Health Report Radiation Monitor 4th Quarter 2003
Songs System Health Report Radiation Monitor 3d Quarter 2003
Songs System Health Report Normal and Emergency Chilled Water 1st Quarter 2004
Songs System Health Report Normal and Emergency Chilled Water 4th Quarter 2003
Songs System Health Report Normal and Emergency Chilled Water 3d Quarter 2003
SO123-XV-44.1, REV 2, 10 CFR 50.59 Resource Manual
-11-
ALSTOM, Ref: SONGS Unit 2 LP valve on-load test interval
GEC-ALSTHOM, San Onofre Retrofit Missile Analysis Report
License Amendment Application Nos. 178 and 164
Letter S-91-077, from ABB to Southern California Edison Company, Subject: Final Report for
the Reduction in Auxiliary Feedwater Analysis, dated May 20, 1991