ML062620510
| ML062620510 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 09/19/2006 |
| From: | Kalyanam N NRC/NRR/ADRO/DORL/LPLIV |
| To: | |
| Kalyanam N, NRR/DORL, 415-1480 | |
| Shared Package | |
| ML062620367 | List: |
| References | |
| TAC MC9236, TAC MC9237 | |
| Download: ML062620510 (36) | |
Text
(3) SCE, pursuant to the Act and 10 CFR Part 70, to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4) SCE, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5) SCE, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) SCE, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of San Onofre Nuclear Generating Station, Units 1 and 2 and by the decommissioning of San Onofre Nuclear Generating Station Unit 1.
C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level Southern California Edison Company (SCE) is authorized to operate the facility at reactor core power levels not in excess of full power (3438 megawatts thermal).
(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 204, are hereby incorporated in the license. Southern California Edison Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
Amendment No. 2093 204 I
TABLE OF CONTENTS 3.3 INSTRUMENTATION (continued) 3.3.5 Engineered Safety Features Actuation System (ESFAS)
Instrumentation ................ 3.3-22 3.3.6 Engineered Safety Features Actuation System (ESFAS)
Logic and Manual Trip ............. 3.3-27 3.3.7 Diesel Generator (DG) -Undervoltage Start ..... 3.3-32 3.3.8 Containment Purge Isolation Signal (CPIS) ..... 3.3-35 3.3.9 Control Room Isolation Signal (CRIS) ....... 3.3-39 3.3.10 Fuel Handling Isolation Signal (FHIS) . 3.3-42 3.3.11 Post Accident Monitoring Instrumentation (PAMI) . . 3.3-44 3.3.12 Remote Shutdown System .............. . 3.3-48 3.3.13 Source Range Monitoring Channels ......... 3.3-51 3.4 REACTOR COOLANT SYSTEM (RCS) ..... .............. 3.4-1 3.4.1 RCS DNB Pressure, Temperature, and Flow Limits . . . . . . . . . . . . . . . . . . . . . 3.4-1 3.4.2 RCS Minimum Temperature for Criticality ...... 3.4-4 3.4.3 RCS Pressure and Temperature (P/T) Limits ..... 3.4-5 3.4.3.1 Pressurizer Heatup/Cooldown Limits ............ 3.4-13 3.4.4 RCS Loops-MODES 1 and 2 ..... .............. 3.4-15 3.4.5 RCS Loops-MODE 3 .......... . ...... 3.4-16 3.4.6 RCS Loops-MODE 4 ................ 3.4-18 3.4.7 RCS Loops-MODE 5, Loops Filled .......... 3.4-21 3.4.8 RCS Loops-MODE 5, Loops Not Filled ........ 3.4-24 3.4.9 Pressurizer .................... 3.4-26 3.4.10 Pressurizer Safety Valves ............. 3.4-28 3.4.11 Not Used 3.4.12.1 Low Temperature Overpressure Protection (LTOP)
System, RCS Temperature
System, RCS Temperature > PTLR Limit ...... 3.4-35 I 3.4.13 RCS Operational LEAKAGE .............. 3.4-37 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage ..... 3.4-39 3.4.15 RCS Leakage Detection Instrumentation ....... 3.4-44 3.4.16 RCS Specific Activity ............... 3.4-47 3.4.17 RCS Steam Generator (SG) Tube Integrity ...... 3.4-51 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5-1 3.5.1 Safety Injection Tanks (SITs) . . . 3.5-1 3.5.2 ECCS- Operating .......... 3.5-4 3.5.3 ECCS- Shutdown ........... 3.5-8 3.5.4 Refueling Water Storage Tank (RWST) 3.5-9 3.5.5 Trisodium Phosphate (TSP) ..... 3.5-11 (conti nued)
SAN ONOFRE--Unit 2 ii i Amendment No. 203 204
TABLE OF CONTENTS B 3.3 INSTRUMENTATION (continued)
B 3.3.7 Diesel Generator Undervoltage Start ....... B 3.3-126 B 3.3.8 Containment Purge Isolation Signal (CPIS) .... B 3.3-135 B 3.3.9 Control Room Isolation Signal (CRIS) ......... B 3.3-145 B 3.3.10 Fuel Handling Isolation Signal (FHIS) ...... B 3.3-152 B 3.3.11 Post Accident Monitoring Instrumentation (PAMI) B 3.3-159 B 3.3.12 Remote Shutdown System ..... .............. B 3.3-176 B 3.3.13 Source Range Monitoring Channels ............ B 3.3-181 B 3.4 REACTOR COOLANT SYSTEM (RCS) ............ B 3.4-1 B 3.4.1 RCS DNB (Pressure, Temperature, and Flow) Limits. B 3.4-1 B 3.4.2 RCS Minimum Temperature for Criticality ..... B 3.4-7 B 3.4.3 RCS Pressure and Temperature (P/T) Limits .... B 3.4-9 B 3.4.4 RCS Loops-MODES 1 and 2 ..... ............. B 3.4-23 B 3.4.5 RCS Loops-MODE 3 ................ B 3.4-27 B 3.4.6 RCS Loops- MODE 4 ................ B 3.4-31 B 3.4.7 RCS Loops-MODE 5, Loops Filled ......... B 3.4-36 B 3.4.8 RCS Loops-MODE 5, Loops Not Filled B 3.4-42 B 3.4.9 Pressurizer .............. B 3.4-46 B 3.4.10 Pressurizer Safety Valves ............ B 3.4-51 B 3.4.11 Not Used B 3.4.12.1 Low Temperature Overpressure Protection (LTOP)
System, RCS Temperature
System, RCS Temperature > PTLR Limit ..... B 3.4-65 B 3.4.13 RCS Operational LEAKAGE ............. B 3.4-70 B 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage .... B 3.4-76 B 3.4.15 RCS Leakage Detection Instrumentation ...... B 3.4-82 B 3.4.16 RCS Specific Activity .............. B 3.4-88 B 3.4.17 RCS Steam Generator (SG) Tube Integrity B 3.4-93 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) . . B 3.5-1 B 3.5.1 Safety Injection Tanks (SITs) .... B 3.5-1 B 3.5.2 ECCS -Operating ........... B 3.5-11 B 3.5.3 ECCS -Shutdown ......... B 3.5-21 B 3.5.4 Refueling Water Storage Tank (RWST) B 3.5-24 B 3.5.5 Trisodium Phosphate (TSP) ...... B 3.5-30 B 3.6 CONTAINMENT SYSTEMS .......... B 3.6-1 B 3.6.1 Containment ............ B 3.6-1 B 3.6.2 Containment Air Locks ....... B 3.6-5 B 3.6.3 Containment Isolation Valves . ... B 3.6-13 B 3.6.4 Containment Pressure ........... B 3.6-27 B 3.6.5 Containment Air Temperature . . B 3.6-30 B 3.6.6.1 Containment Spray and Cooling System B 3.6-33 B 3.6.6.2 Containment Cooling System ..... B 3.6-43 B 3.6.7 Hydrogen Recombiners ........... B 3.6-48 B 3.6.8 Containment Dome Air Circulators . . B 3.6-53 (continued)
SAN ONOFRE--Unit 2 vii Amendment No. +24 204
Definitions 1.1 1.1 Definitions ENGINEERED SAFETY measurement, response time may be verified for FEATURE (ESF) RESPONSE selected components provided that the components TIME (Continued) and methodology for verification have been previously reviewed and approved by the NRC.
LEAKAGE LEAKAGE shall be:
- a. Identified LEAKAGE
- 1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) leakoff), that is captured and conducted to collection systems or a sump or collecting tank;
- 2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or
- 3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE).
- b. Unidentified LEAKAGE All LEAKAGE that is not identified LEAKAGE.
- c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.
MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.
(continued)
SAN ONOFRE--UNIT 2 1.1-4 Amendment No. 188 204
RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:
- a. No pressure boundary LEAKAGE;
- b. 1 gpm unidentified LEAKAGE;
- c. 10 gpm identified LEAKAGE; and
- d. 150 gallons per day primary to secondary LEAKAGE through any one Steam Generator (SG).
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RCS Operational A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> LEAKAGE not within within limits.
limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.
B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A AND not met.
B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR Pressure boundary LEAKAGE exists.
OR Primary to secondary LEAKAGE not within limit.
SAN ONOFRE--UNIT 2 3.4-37 Amendment No. 40 204
RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 ------------------ NOTES ---------------------- NOTE ------
- 1. Not required to be performed in MODE 3 or 4 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of steady state Only required to be performed I
operation. during steady state
- 2. Not applicable to primary to secondary operation.
LEAKAGE. If a transient evolution is occurring 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from the last water Perform RCS water inventory balance, inventory balance, then a water inventory balance shall be performed within 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> of the last water inventory balance 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
NOTE----------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.
SR 3.4.13.2 Verify primary to secondary LEAKAGE is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
< 150 gallons per day through any one SG.
SAN ONOFRE--UNIT 2 3.4-38 Amendment No. 1-24 204
SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 SG tube integrity shall be maintained.
AND All SG tubes satisfying the tube repair criteria shall be plugged or repaired in accordance with the Steam Generator Program.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTIONS
NOTE -------------------------------
Separate Condition entry is allowed for each SG tube.
CONDITION REQUIRED ACTION COMPLETION TIME I A. One or more SG tubes A.1 Verify tube integrity 7 days satisfying the tube of the affected repair criteria and tube(s) is maintained until the next not plugged or refueling outage or repaired in accordance SG tube inspection.
with the Steam Generator Program. AND A.2 Plug or repair the Prior to affected tube(s) in entering MODE 4 accordance with the following the Steam Generator next refueling Program. outage or SG tube inspection B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A AND not met.
B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR SG tube integrity not maintained.
SAN ONOFRE--UNIT 2 3.4-51 Amendment No. 204
SG Tube Integrity 3.4.17 SURVEILLANCE REQUIREMENTS I I
SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify SG tube integrity in accordance with In accordance the Steam Generator Program. with the Steam Generator Program SR 3.4.17.2 Verify that each inspected SG tube that Prior to satisfies the tube repair criteria is entering MODE 4 plugged or repaired in accordance with the following a SG Steam Generator Program. tube inspection SAN ONOFRE--UNIT 2 3.4-52 Amendment No. 204
Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.8 Primary Coolant Sources Outside Containment Program (continued) system (post-accident sampling return piping only until such time as a modification eliminates the post-accident piping as a potential leakage path). The program shall include the following:
- a. Preventive maintenance and periodic visual inspection requirements; and
- b. Integrated leak test requirements for each system at refueling cycle intervals or less.
5.5.2.9 Pre-Stressed Concrete Containment Tendon Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stressed concrete containment, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. Program itself is relocated to the LCS.
5.5.2.10 Inservice Inspection and Testing Program This program provides controls for inservice inspection of ASME Code Class 1, 2, and 3 components and Code Class CC and MC components including applicable supports. The program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program itself is located in the LCS.
5.5.2.11 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:
- a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging or repair of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected, plugged, or repaired to confirm that the performance criteria are being met.
(continued)
SAN ONOFRE--UNIT 2 5.0-13 Amendment No.1-8, 204
Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.11 Steam Generator (SG) Program (continued)
- b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.
- 1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.
- 2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.
Leakage is not to exceed 0.5 gpm per SG and 1 gpm through both SGs.
- 3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."
(continued)
SAN ONOFRE--UNIT 2 5.0-14 Amendment No. +48, 204
Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.11 Steam Generator (SG) Program (continued)
- c. Provisions for SG tube repair criteria. The non-sleeved region of a tube found by inservice inspection to contain flaws with a depth equal to or exceeding 44% of the nominal tube wall thickness shall be plugged or repaired except if the flaws are permitted to remain in service through application of an alternate tube repair criteria discussed below.
Tubes shall be plugged if the sleeved region of a tube is found by inservice inspection to contain flaws in the (a) sleeve or (b) pressure boundary portion of the original tube wall in the sleeve tube assembly (i.e., the sleeve-to-tube joint).
- d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, d.3, and d.4 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection.. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
- 2. Inspect 100% of the tubes at sequential periods of 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate for more than 24 effective full power months or one refueling outage (whichever is less) without being inspected.
(continued)
SAN ONOFRE--UNIT 2 5.0-15 Amendment No.140, 204
Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.11 Steam Generator (SG) Program (continued)
- 3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
- 4. All sleeves shall be inspected with eddy current prior to initial operation. This includes pressure retaining portions of the parent tube in contact with the sleeve, the sleeve-to-tube weld and the pressure retaining portion of the sleeve.
- e. Provisions for monitoring operational primary to secondary LEAKAGE.
- f. Provisions for SG tube repair methods. Steam generator tube repair methods shall provide the means to re-establish the RCS pressure boundary integrity of SG tubes without removing the tube from service. For the purposes of these Specifications, tube plugging is not a repair. All acceptable tube repair methods are listed below.
- 1. TIG welded sleeving with heat treatment, as described in ABB/CE Topical Report, CEN-630-P, Rev. 2, is currently approved by the NRC.
Tube repair can be performed on certain tubes that have been previously plugged as a corrective or preventive measure. A tube inspection of the entire length of the tube shall be performed on a previously plugged tube prior to returning the tube to service.
(continued)
SAN ONOFRE--UNIT 2 5.0-16 Amendment No. 148, 204
Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.12 Ventilation Filter Testing Program (VFTP)
This Program establishes the required testing of the Engineered Safety Feature filter ventilation systems, "Control Room Emergency Air Cleanup System" and "Fuel Handling Building Post-accident Cleanup Filter System." The frequency of testing shall be in accordance with Regulatory Guide 1.52, Revision 2. As a minimum the VFTP program shall include the following:
- a. Inplace testing of the high efficiency particulate air (HEPA) filters to demonstrate acceptable penetration and system bypass when tested at the appropriate system flowrate in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1975 (see Note 1); and
- b. Inplace testing of the charcoal adsorber to demonstrate acceptable penetration and system bypass when-tested at the appropriate system flowrate in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1975 (see Note 1); and
- c. Laboratory testing of charcoal adsorber samples obtained in accordance with Regulatory Guide 1.52, Revision 2 and tested per the methodology of ASTM D3803-1989 at 30'C and 70%
relative humidity to show acceptable methyl iodide penetration; and
- d. Testing to demonstrate the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers, when tested at the appropriate system flowrate.
Note 1: Sample and injection points shall be qualified per ANSI N510-1975 unless manifolds have been qualified per ASME N510-1989.
HEPA testing will be conducted with DOP aerosol or suitable alternate.
(continued)
SAN ONOFRE--UNIT 2 5.0-17 Amendment No. -97, 204
Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.12 Ventilation Filter Testing Program (VFTP) (continued)
The provisions of Technical Specification Surveillance Requirement 3.0.2 and Technical Specification Surveillance Requirement 3.0.3 are applicable to the VFTP test frequencies.
5.5.2.13 Diesel Fuel Oil Testing Program This program implements required testing of both new fuel oil and stored fuel oil. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM standards. The purpose of the program is to establish the following:
- a. Acceptability of new fuel oil use prior to addition to storage tanks by determining that the fuel oil has:
- 1. an API gravity or an absolute specific gravity within limits,
- 2. a flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and
- 3. a water and sediment content within limits.
- b. Other properties for ASTM 2D fuel oil are within limits within 31 days following sampling and addition to the storage tanks, with exceptions noted in the Bases for Surveillance Requirement 3.8.3.3; and,
- c. Total particulate concentration of fuel oil is
- 10 mg/l when tested every 92 days in accordance with ASTM D-2276, Method A.
5.5.2.14 Deleted (continued)
SAN ONOFRE--UNIT 2 5.0-18 Amendment No. +89, 204
Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (conti nued)
This Page Intentionally Left Blank (continued)
SAN ONOFRE--UNIT 2 5.0-19 Amendment No. +40, 204
Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued)
This Page Intentionally Left Blank (continued)
SAN ONOFRE--UNIT 2 5.0-20 Amendment No. +97, 204
Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.2 Special Reports Special Reports may be required covering inspection, test, and maintenance activities. These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Technical Specifications.
Special Reports shall be submitted to the U. S. Nuclear Regulatory Commission, Attention: Document Control Desk, Washington, D. C.
20555, with a copy to the Regional Administrator of the Regional Office of the NRC, in accordance with 10 CFR 50.4 within the time period specified for each report.
The following Special Reports shall be submitted:
- a. When a pre-planned alternate method of monitoring post-accident instrumentation functions is required by Condition B or Condition G of LCO 3.3.11, a report shall be submitted within 30 days from the time the action is required. The report shall outline the action taken, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the function to OPERABLE status.
- b. Any abnormal degradation of the containment structure detected during the tests required by the Pre-Stressed Concrete Containment Tendon Surveillance Program shall be reported to the NRC within 30 days. The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken.
- c. A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.2.11, Steam Generator (SG) Program. The report shall include:
(continued)
SAN ONOFRE--UNIT 2 5.0-28 Amendment No. +9 204
Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.2 Special Reports (continued)
- 1. The scope of inspections performed on each SG,
- 2. Active degradation mechanisms found,
- 3. Nondestructive examination techniques utilized for each degradation mechanism,
- 4. Location, orientation (if linear), and measured sizes (if available) of service induced indications,
- 5. Number of tubes plugged or repaired during the inspection outage for each active degradation mechanism,
- 6. Total number and percentage of tubes plugged or repaired to date,
- 7. The results of condition monitoring, including the results of tube pulls and in-situ testing,
- 8. The effective plugging percentage for all plugging and tube repairs in each SG, and
- 9. Repair method utilized and the number of tubes repaired by each repair method.
SAN ONOFRE--UNIT 2 5.0-29 Amendment No. +9 204
(3) SCE, pursuant to the Act and 10 CFR Part 70, to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4) SCE, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source and special nuclear materials as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5) SCE, pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) SCE, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of San Onofre Nuclear Generating Station, Units I and 3 and by the decommissioning of San Onofre Nuclear Generating Station Unit 1.
C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level Southern California Edison Company (SCE) is authorized to operate the facility at reactor core power levels not in excess of full power (3438 megawatts thermal).
(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 196, are hereby incorporated in the license.
Southern California Edison Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
Amendment No. +% 196 I
TABLE OF CONTENTS 3.3 INSTRUMENTATION (continued) 3.3.5 Engineered Safety Features Actuation System (ESFAS)
Instrumentation ................ 3.3-22 3.3.6 Engineered Safety Features Actuation System (ESFAS)
Logic and Manual Trip ............. 3.3-27 3.3.7 Diesel Generator (DG) -Undervoltage Start ..... 3.3-32 3.3.8 Containment Purge Isolation Signal (CPIS) ..... 3.3-35 3.3.9 Control Room Isolation Signal (CRIS) ....... 3.3-39 3.3.10 Fuel Handling Isolation Signal (FHIS) ....... 3.3-42 3.3.11 Post Accident Monitoring Instrumentation (PAMI) 3.3-44 3.3.12 Remote Shutdown System .............. 3.3-48 3.3.13 Source Range Monitoring Channels ......... 3.3-51 3.4 REACTOR COOLANT SYSTEM (RCS) ..... ............... 3.4-1 3.4.1 RCS DNB Pressure, Temperature, and Flow Limits . .. 3.4-1 3.4.2 RCS Minimum Temperature for Criticality . 3.4-4 3.4.3 RCS Pressure and Temperature (P/T) Limits ..... 3.4-5 3.4.3.1 Pressurizer Heatup/Cooldown Limits ............. 3.4-13 3.4.4 RCS Loops-MODES 1 and 2 ..... ............... 3.4-15 3.4.5 RCS Loops-MODE 3 ................. 3.4-16 3.4.6 RCS Loops-MODE 4 ................. 3.4-18 3.4.7 RCS Loops-MODE 5, Loops Filled .......... 3.4-21 3.4.8 RCS Loops-MODE 5, Loops Not Filled . ....... 3.4-24 3.4.9 Pressurizer ... .. . . . . . . . . . . . . 3.4-26 3.4.10 Pressurizer Safety Valves ............. 3.4-28 3.4.11 Not Used 3.4.12.1 Low Temperature Overpressure Protection (LTOP)
System, RCS Temperature
System, RCS Temperature > PTLR Limit ...... 3.4-35 3.4.13 RCS Operational LEAKAGE .............. 3.4-37 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage ..... 3.4-39 3.4.15 RCS Leakage Detection Instrumentation ....... 3.4-44 3.4.16 RCS Specific Activity ............... 3.4-47 3.4.17 RCS Steam Generator (SG) Tube Integrity ...... 3.4-51 I 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) . . . 3.5-1 3.5.1 Safety Injection Tanks (SITs) ..... 3.5-1 3.5.2 ECCS -Operating ............ 3.5-4 3.5.3 ECCS -Shutdown .......... 3.5-8 3.5.4 Refueling Water Storage Tank (RWST) . . 3.5-9 3.5.5 Trisodium Phosphate (TSP) ....... 3.5-11 (continued)
SAN ONOFRE--Unit 3 ii i Amendment No. +9-5 196
TABLE OF CONTENTS B 3.3 INSTRUMENTATION (continued)
B 3.3.7 Diesel Generator Undervoltage Start ...... B 3.3-126 B 3.3.8 Containment Purge Isolation Signal (CPIS) . . . B 3.3-135 B 3.3.9 Control Room Isolation Signal (CRIS) ......... B 3.3-145 B 3.3.10 Fuel Handling Isolation Signal (FHIS) ..... B 3.3-152 B 3.3.11 Post Accident Monitoring Instrumentation (PAMI) B 3.3-159 B 3.3.12 Remote Shutdown System ..... .............. B 3.3-176 B 3.3.13 Source Range Monitoring Channels ............ B 3.3-181 B 3.4 REACTOR COOLANT SYSTEM (RCS) ..... ............ B 3.4-1 B 3.4.1 RCS DNB (Pressure, Temperature, and Flow) Limits B 3.4-1 B 3.4.2 RCS Minimum Temperature for Criticality .... B 3.4-7 B 3.4.3 RCS Pressure and Temperature (P/T) Limits . . . B 3.4-9 B 3.4.4 RCS Loops-MODES 1 and 2 ...... ............ B 3.4-23 B 3.4.5 RCS Loops-MODE 3 ............... B 3.4-27 B 3.4.6 RCS Loops-MODE 4 ..... .......... B 3.4-31 B 3.4.7 RCS Loops-MODE 5, Loops Filled ........ B 3.4-36 B 3.4.8 RCS Loops-MODE 5, Loops Not Filled ...... B 3.4-42 B 3.4.9 Pressurizer . . . . . . . . . . . . . . . . . . B 3.4-46 B 3.4.10 Pressurizer Safety Valves ........... B 3.4-51 B 3.4.11 Not Used B 3.4.12.1 Low Temperature Overpressure Protection (LTOP)
System, RCS Temperature
System, RCS Temperature > PTLR Limit .... B 3.4-65 B 3.4.13 RCS Operational LEAKAGE ............ B 3.4-70 B 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage . . . B 3.4-76 B 3.4.15 RCS Leakage Detection Instrumentation ..... B 3.4-82 B 3.4.16 RCS Specific Activity ............. B 3.4-88 B 3.4.17 RCS Steam Generator (SG) Tube Integrity B 3.4-93 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) B 3.5-1 B 3.5.1 Safety Injection Tanks (SITs) . . . B 3.5-1 B 3.5.2 ECCS- Operating .......... B 3.5-11 B 3.5.3 ECCS -Shutdown .......... B 3.5-21 B 3.5.4 Refueling Water Storage Tank (RWST) B 3.5-24 B 3.5.5 Trisodium Phosphate (TSP) ..... B 3.5-30 B 3.6 CONTAINMENT SYSTEMS .......... B 3.6-1 B 3.6.1 Containment ............ B 3.6-1 B 3.6.2 Containment Air Locks ....... B 3.6-5 B 3.6.3 Containment Isolation Valves . ... B 3.6-13 B 3.6.4 Containment Pressure ............ B 3.6-27 B 3.6.5 Containment Air Temperature . ... B 3.6-30 B 3.6.6.1 Containment Spray and Cooling System B 3.6-33 B 3.6.6.2 Containment Cooling System ....... B 3.6-43 B 3.6.7 Hydrogen Recombiners .... ........ B 3.6-48 B 3.6.8 Containment Dome Air Circulators . . B 3.6-53 (continued)
SAN ONOFRE--Unit 3 vii Amendment No. 14-6 196
Definitions 1.1 1.1 Definitions ENGINEERED SAFETY measurement, response time may be verified for FEATURE (ESF) RESPONSE selected components provided that the components TIME (Continued) and methodology for verification have been previously reviewed and approved by the NRC.
LEAKAGE LEAKAGE shall be:
- a. Identified LEAKAGE
- 1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) leakoff), that is captured and conducted to collection systems or a sump or collecting tank;
- 2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or
- 3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE). I
- b. Unidentified LEAKAGE All LEAKAGE that is not identified LEAKAGE.
- c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.
MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.
(continued)
SAN ONOFRE--UNIT 3 1.1-4 Amendment No. +/-79 196
RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:
- a. No pressure boundary LEAKAGE;
- b. 1 gpm unidentified LEAKAGE;
- c. 10 gpm identified LEAKAGE; and I
- d. 150 gallons per day primary to secondary LEAKAGE through any one Steam Generator (SG).
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RCS Operational A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> LEAKAGE not within within limits.
limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.
B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A AND not met.
B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR Pressure boundary LEAKAGE exists.
OR Primary to secondary LEAKAGE not within limit.
SAN ONOFRE--UNIT 3 3.4-37 Amendment No. +32 196
RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 ----------------- NOTES ------------------------ NOTE ------
- 1. Not required to be performed in MODE 3 Only required I or 4 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of steady state to be performed operation. during steady state
- 2. Not applicable to primary to secondary operation.
LEAKAGE. If a transient evolution is occurring 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from the last water Perform RCS water inventory balance. inventory balance, then a water inventory balance shall be performed within 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> of the last water inventory balance 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
NOTE-----------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.
SR 3.4.13.2 Verify primary to secondary LEAKAGE is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
< 150 gallons per day through any one SG.
SAN ONOFRE--UNIT 3 3.4-38 Amendment No. 1-76 196
SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 SG tube integrity shall be maintained.
AND All SG tubes satisfying the tube repair criteria shall be plugged or repaired in accordance with the Steam Generator Program.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTIONS
NOTE----------------------------------
Separate Condition entry is allowed for each SG tube.
CONDITION REQUIRED ACTION COMPLETION TIME I A. One or more SG tubes A.1 Verify tube integrity 7 days satisfying the tube of the affected repair criteria and tube(s) is maintained not plugged or until the next repaired in accordance refueling outage or with the Steam SG tube inspection.
Generator Program.
AND A.2 Plug or repair the Prior to affected tube(s) in entering MODE 4 accordance with the following the Steam Generator next refueling Program. outage or SG tube inspection B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A AND not met.
B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR SG tube integrity not maintained.
SAN ONOFRE--UNIT 3 3.4-51 Amendment No. 196 I
SG Tube Integrity 3.4.17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify SG tube integrity in accordance with In accordance the Steam Generator Program. with the Steam Generator Program SR 3.4.17.2 Verify that each inspected SG tube that Prior to satisfies the tube repair criteria is entering MODE 4 plugged or repaired in accordance with the following a SG Steam Generator Program. tube inspection SAN ONOFRE--UNIT 3 3.4-52 Amendment No. 196
Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.8 Primary Coolant Sources Outside Containment Program (continued) system (post-accident sampling return piping only until such time as a modification eliminates the post-accident piping as a potential leakage path). The program shall include the following:
- a. Preventive maintenance and periodic visual inspection requirements; and
- b. Integrated leak test requirements for each system at refueling cycle intervals or less.
5.5.2.9 Pre-Stressed Concrete Containment Tendon Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stressed concrete containment, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. Program itself is relocated to the LCS.
5.5.2.10 Inservice Inspection and Testing Program This program provides controls for inservice inspection of ASME Code Class 1, 2, and 3 components and Code Class CC and MC components including applicable supports. The program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components.
Program itself is located in the LCS.
5.5.2.11 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:
- a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging or repair of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected, plugged, or repaired to confirm that the performance criteria are being met.
(continued)
SAN ONOFRE--UNIT 3 5.0-13 Amendment No. +69, 196
Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.11 Steam Generator (SG) Program (continued)
- b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.
- 1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be .determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.
- 2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.
Leakage is not to exceed 0.5 gpm per SG and 1 gpm through both SGs.
- 3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."
(continued)
SAN ONOFRE--UNIT 3 5.0-14 Amendment No. 3, 196
Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.11 Steam Generator (SG) Program (continued)
- c. Provisions for SG tube repair criteria. The non-sleeved region of a tube found by inservice inspection to contain flaws with a depth equal to or exceeding 44% of the nominal tube wall thickness shall be plugged or repaired except if the flaws are permitted to remain in service through application of an alternate tube repair criteria discussed below.
Tubes shall be plugged if the sleeved region of a tube is found by inservice inspection to contain flaws in the (a) sleeve or (b) pressure boundary portion of the original tube wall in the sleeve tube assembly (i.e., the sleeve-to-tube joint).
- d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, d.3, and d.4 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
- 2. Inspect 100% of the tubes at sequential periods of 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate for more than 24 effective full power months or one refueling outage (whichever is less) without being inspected.
(continued)
Amendment No.+3~, 196 5.0-15 SAN ONOFRE--UNIT 33 5.0-15 Amendment No. +3-2, 196
Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.11 Steam Generator (SG) Program (continued)
- 3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
- 4. All sleeves shall be inspected with eddy current prior to initial operation. This includes pressure retaining portions of the parent tube in contact with the sleeve, the sleeve-to-tube weld and the pressure retaining portion of the sleeve.
- e. Provisions for monitoring operational primary to secondary LEAKAGE.
- f. Provisions for SG tube repair methods. Steam generator tube repair methods shall provide the means to re-establish the RCS pressure boundary integrity of SG tubes without removing the tube from service. For the purposes of these Specifications, tube plugging is not a repair. All acceptable tube repair methods are listed below.
- 1. TIG welded sleeving with heat treatment, as described in ABB/CE Topical Report, CEN-630-P, Rev. 2, is currently approved by the NRC.
Tube repair can be performed on certain tubes that have been previously plugged as a corrective or preventive measure. A tube inspection of the entire length of the tube shall be performed on a previously plugged tube prior to returning the tube to service.
(continued)
SAN ONOFRE--UNIT 3 5.0-16 Amendment No. +32, 196
Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.5.2.12 Ventilation Filter Testing Program (VFTP)
This Program establishes the required testing of the Engineered Safety Feature filter ventilation systems, "Control Room Emergency Air Cleanup System" and "Fuel Handling Building Post-accident Cleanup Filter System." The frequency of testing shall be in accordance with Regulatory Guide 1.52, Revision 2. As a minimum the VFTP program shall include the following:
- a. Inplace testing of the high efficiency particulate air (HEPA) filters to demonstrate acceptable penetration and system bypass when tested at the appropriate system flowrate in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1975 (see Note 1); and
- b. Inplace testing of the charcoal adsorber to demonstrate acceptable penetration and system bypass when tested at the appropriate system flowrate in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1975 (see Note 1); and
- c. Laboratory testing of charcoal adsorber samples obtained in accordance with Regulatory Guide 1.52, Revision 2 and tested per the methodology of ASTM D3803-1989 at 30'C and 70%
relative humidity to show acceptable methyl iodide penetration; and
- d. Testing to demonstrate the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers, when tested at the appropriate system flowrate.
Note 1: Sample and injection points shall be qualified per ANSI N510-1975 unless manifolds have been qualified per ASME N510-1989.
HEPA testing will be conducted with DOP aerosol or suitable alternate.
(continued)
SAN ONOFRE--UNIT 3 5.0-17 Amendment No. +88, 196
Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued) 5.2.12 Ventilation Filter Testing Program (VFTP) (continued)
The provisions of Technical Specification Surveillance Requirement 3.0.2 and Technical Specification Surveillance Requirement 3.0.3 are applicable to the VFTP test frequencies.
5.5.2.13 Diesel Fuel Oil Testing Program This program implements required testing of both new fuel oil and stored fuel oil. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM standards. The purpose of the program is to establish the following:
- a. Acceptability of new fuel oil use prior to addition to storage tanks by determining that the fuel oil has:
- 1. an API gravity or an absolute specific gravity within limits,
- 2. a flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and
- 3. a water and sediment content within limits.
- b. Other properties for ASTM 2D fuel oil are within limits within 31 days following sampling and addition to the storage tanks, with exceptions noted in the Bases for Surveillance Requirement 3.8.3.3; and,
- c. Total particulate concentration of fuel oil is
- 10 mg/l when tested every 92 days in accordance with ASTM D-2276, Method A.
5.5.2.14 Deleted (conti nued)
SAN ONOFRE--UNIT 3 5.0-18 Amendment No.8e, 196
Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued)
This Page Intentionally Left Blank (continued)
SAN ONOFRE--UNIT 3 5.0-19 Amendment No. ÷-3,196
Procedures, Programs, and Manuals 5.5 5.5 Procedures, Programs, and Manuals (continued)
This Page Intentionally Left Blank (continued)
SAN ONOFRE--UNIT 3 5.0-20 Amendment No.+88, 196
Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.2 Special Reports Special Reports may be required covering inspection, test, and maintenance activities. These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Technical Specifications.
Special Reports shall be submitted to the U. S. Nuclear Regulatory Commission, Attention: Document Control Desk, Washington, D. C.
20555, with a copy to the Regional Administrator of the Regional Office of the NRC, in accordance with 10 CFR 50.4 within the time period specified for each report.
The following Special Reports shall be submitted:
- a. When a pre-planned alternate method of monitoring post-accident instrumentation functions is required by Condition B or Condition G of LCO 3.3.11, a report shall be submitted within 30 days from the time the action is required. The report shall outline the action taken, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the function to OPERABLE status.
- b. Any abnormal degradation of the containment structure detected during the tests required by the Pre-Stressed Concrete Containment Tendon Surveillance Program shall be reported to the NRC within 30 days. The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken.
- c. A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.2.11, Steam Generator (SG) Program. The report shall include:
(continued)
SAN ONOFRE--UNIT 3 5.0-28 Amendment No. 4-88 196
Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.2 Special Reports (continued)
- 1. The scope of inspections performed on each SG,
- 2. Active degradation mechanisms found,
- 3. Nondestructive examination techniques utilized for each degradation mechanism,
- 4. Location, orientation (if linear), and measured sizes (if available) of service induced indications,
- 5. Number of tubes plugged or repaired during the inspection outage for each active degradation mechanism,
- 6. Total number and percentage of tubes plugged or repaired to date,
- 7. The results of condition monitoring, including the results of tube pulls and in-situ testing,
- 8. The effective plugging percentage for all plugging and tube repairs in each SG, and
- 9. Repair method utilized and the number of tubes repaired by each repair method.
SAN ONOFRE--UNIT 3 5.0-29 Amendment No. +/-88 196