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| number = ML062640464 | | number = ML062640464 | ||
| issue date = 09/15/2006 | | issue date = 09/15/2006 | ||
| title = | | title = Technical Specification, Incorporate a Full-Scope Application of an Alternate Source Term Methodology | ||
| author name = | | author name = | ||
| author affiliation = NRC/NRR/ADRO/DORL/LPLB | | author affiliation = NRC/NRR/ADRO/DORL/LPLB | ||
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=Text= | =Text= | ||
{{#Wiki_filter:INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.7-3 STEAM LINE SAFETY VALVES PER LOOP ................................... | {{#Wiki_filter:INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.7-3 STEAM LINE SAFETY VALVES PER LOOP ................................... 3/4 7-3 Auxiliary Feedwater System ................................................................ 3/4 7-4 Demineralized Water Storage Tank ....................................................... 3/4 7-6 Specific A ctivity .................................................................................... 3/4 7-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM ................... 3/4 7-8 Main Steam Line Isolation Valves ........................................................ 3/4 7-9 Steam Generator Atmospheric Relief Bypass Lines ............................. 3/4 7-9a 3/4.7.2 DELETED ............................................................................................. 3/4 7-10 3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM ..3/4 7-11 3/4.7.4 SERVICE WATER SYSTEM ........................... 3/4 7-12 3/4.7.5 ULTIMATE HEAT SINK ...................................................................... 3/4 7-13 3/4.7.6 D ELETED ............................................................................................. 3/4 7-14 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM ......... 3/4 7-15 3/4.7.8 DELETED ............................................................................................. 3/4 7-18 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM ..................................... 3/4 7-20 3/4.7.10 SN UBBERS .......................................................................................... 3/4 7-22 TABLE 4.7-2 SNUBBER VISUAL INSPECTION INTERVAL ................................ 3/4 7-27 FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST .............. 3/4 7-29 3/4.7.11 DELETED ............................................................................................. 3/4 7-30 3/4.7.12 DELETED TABLE 3.7-4 DELETED TABLE 3.7-5 DELETED 3/4.7.13 DELETED 3/4.7.14 AREA TEMPERATURE MONITORING ........................................... 3/4 7-32 TABLE 3.7-6 AREA TEMPERATURE MONITORING ........................................... 3/4 7-33 MILLSTONE - UNIT 3 x Amendment No. 6-2, 84, 400, 4-60, 244, 232 | ||
3/4 7-3 Auxiliary Feedwater System ................................................................ | |||
3/4 7-4 Demineralized Water Storage Tank ....................................................... | |||
3/4 7-6 Specific A ctivity .................................................................................... | |||
3/4 7-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM ................... | |||
3/4 7-8 Main Steam Line Isolation Valves ........................................................ | |||
3/4 7-9 Steam Generator Atmospheric Relief Bypass Lines ............................. | |||
3/4 7-9a 3/4.7.2 DELETED ............................................................................................. | |||
3/4 7-10 3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM ..3/4 7-11 3/4.7.4 SERVICE WATER SYSTEM ........................... | |||
3/4 7-12 3/4.7.5 ULTIMATE HEAT SINK ...................................................................... | |||
3/4 7-13 3/4.7.6 D ELETED ............................................................................................. | |||
3/4 7-14 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM ......... | |||
3/4 7-15 3/4.7.8 DELETED ............................................................................................. | |||
3/4 7-18 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM ..................................... | |||
3/4 7-20 3/4.7.10 SN | |||
3/4 7-22 TABLE 4.7-2 SNUBBER VISUAL INSPECTION INTERVAL ................................ | |||
3/4 7-27 FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST .............. | |||
3/4 7-29 3/4.7.11 DELETED ............................................................................................. | |||
3/4 7-30 3/4.7.12 DELETED TABLE 3.7-4 DELETED TABLE 3.7-5 DELETED 3/4.7.13 DELETED 3/4.7.14 AREA TEMPERATURE MONITORING | |||
........................................... | |||
3/4 7-32 TABLE 3.7-6 AREA TEMPERATURE MONITORING | |||
........................................... | |||
3/4 7-33 MILLSTONE | |||
-UNIT 3 x Amendment No. 6-2, 84, 400, 4-60, 244, 232 | |||
INDEX BASES SECTION PAGE TABLE B 3/4.4-1 REACTOR VESSEL FRACTURE TOUGHNESS PROPERTIES .... B 3/4 4-9 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>IMeV) AS A FUNCTION OF FULL POWER SERVICE LIFE ......................................................... B 3/4 4-10 3/4.4.10 DELETED ................................................................................................... B 3/4 4-15 3/4.4.11 DELETED ................................................................................................... B 3/4 4-15 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUM ULATORS ..................................................................................... B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS ........................................................................ B 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK .................................................. B 3/4 5-2 3/4.5.5 pH TRISODIUM PHOSPHATE STORAGE BASKETS ............................ B 3/4 5-3 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT ..................................................................... B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS ............................... B 3/4 6-2 3/4.6.3 CONTAINMENT ISOLATION VALVES ................................................... B 3/4 6-3 3/4.6.4 DELETED 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM ........................ B 3/4 6-3d 3/4.6.6 SECONDARY CONTAINMENT ............................................................... B 3/4 6-4 | |||
'1/4-7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE ...................................................................................... B 3/4 7-1 3/4.7.2 DELETED .............................. .............................. B 3/4 7-7 3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM ........ B 3/4 7-7 3/4.7.4 SERVICE WATER SYSTEM ...................................................................... B 3/4 7-7 3/4.7.5 ULTIMATE HEAT SINK ............................................................................ B 3/4 7-8 3/4.7.6 DELETED ................................................................................................... B 3/4 7-10 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM ............... B 3/1 7-10 3/4.7.8 DELETED ................................................................................................... B 3/4 7-17 I 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM ............................................ B 3/4 7-23 3/4.7.10 SNUBBERS ................................................................................................. B 3/4 7-23 MILLSTONE - UNIT 3 xiv Amendment No. 49, 89,44-5,449, 4-36, 2"M, *-26,244,24-6, 224, 232 | |||
INTEGRITY shall exist when: a. All penetrations required to be closed during accident conditions are either: 1. Capable of being closed by an OPERABLE containment automatic isolation valve system*, or 2. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are opened under administrative control as permitted by Specification 3.6.3. b. All equipment hatches are closed and sealed, c. Each air lock is in compliance with the requirements of Specification 3.6.1.3, d. The containment leakage rates are within the limits of the Containment Leakage Rate Testing Program, and e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or O-rings) is OPERABLE. | DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when: | ||
CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals. CORE ALTERATIONS 1.9 CORE ALTERATIONS shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position. | : a. All penetrations required to be closed during accident conditions are either: | ||
DOSE EOUIVALENT 1-131 1.10 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microCurie/gram) which alone would produce the same CDE-thyroid dose as the quantity and isotopic mixture of 1- 131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed under "Inhalation" in Federal Guidance Report No. 11 (FGR 11), "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion." | : 1. Capable of being closed by an OPERABLE containment automatic isolation valve system*, or | ||
: 2. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are opened under administrative control as permitted by Specification 3.6.3. | |||
: b. All equipment hatches are closed and sealed, | |||
: c. Each air lock is in compliance with the requirements of Specification 3.6.1.3, | |||
: d. The containment leakage rates are within the limits of the Containment Leakage Rate Testing Program, and | |||
: e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or O-rings) is OPERABLE. | |||
CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals. | |||
CORE ALTERATIONS 1.9 CORE ALTERATIONS shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position. | |||
DOSE EOUIVALENT 1-131 1.10 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microCurie/gram) which alone would produce the same CDE-thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed under "Inhalation" in Federal Guidance Report No. 11 (FGR 11), | |||
"Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion." | |||
* In MODE 4, the requirement for an OPERABLE containment isolation valve system is satisfied by use of the containment isolation actuation pushbuttons. | * In MODE 4, the requirement for an OPERABLE containment isolation valve system is satisfied by use of the containment isolation actuation pushbuttons. | ||
MILLSTONE | MILLSTONE - UNIT 3 1-2 Amendment No. 2-, -44, 4-6, 246, 232 | ||
-UNIT 3 1-2 Amendment No. 2-, -44, 4-6, 246, 232 THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE | |||
-UNIT 3 3/4 7-18 Amendment No. 48+-, 20a3, 2-9, 232 THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE | THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE - UNIT 3 3/4 7-18 Amendment No. 48+-, 20a3, 2-9, 232 | ||
-UNIT 3 3/4 7-19 Amendment No. 4243, 24, 206, 232 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) | |||
: 2) Pre-planned operating procedures and backup instrumentation to be used if one or more monitoring instruments become inoperable, and 3) Administrative procedures for returning inoperable instruments to OPERABLE status as soon as practicable. | THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE - UNIT 3 3/4 7-19 Amendment No. 4243, 24, 206, 232 | ||
: f. Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testingof the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions*. | |||
This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995. The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 38.57 psig. The maximum allowable containment leakage rate La, at Pa, shall be 0.3 percent by weight of the containment air per 24 hours. Leakage rate acceptance criteria are: 1) Containment overall leakage rate acceptance criterion is < 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the combined Type B and Type C tests, and <0.06 La for all penetrations that are Secondary Containment bypass leakage paths, and < 0.75 La for Type A tests; 2) Air lock testing acceptance criteria are: a. Overall air lock leakage rate is | ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) | ||
: 2) Pre-planned operating procedures and backup instrumentation to be used if one or more monitoring instruments become inoperable, and | |||
: 3) Administrative procedures for returning inoperable instruments to OPERABLE status as soon as practicable. | |||
: f. Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testingof the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions*. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995. | |||
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 38.57 psig. | |||
The maximum allowable containment leakage rate La, at Pa, shall be 0.3 percent by weight of the containment air per 24 hours. | |||
Leakage rate acceptance criteria are: | |||
: 1) Containment overall leakage rate acceptance criterion is < 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the combined Type B and Type C tests, and <0.06 La for all penetrations that are Secondary Containment bypass leakage paths, and < 0.75 La for Type A tests; | |||
: 2) Air lock testing acceptance criteria are: | |||
: a. Overall air lock leakage rate is | |||
* 0.05 La when tested at 2!Pa" | |||
: b. For each door, seal leakage rate is < 0.01 La when pressurized to | |||
> Pa The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program. | |||
The provisions of Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program. | The provisions of Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program. | ||
* An exemption to Appendix J, Option A, paragraph III.D.2(b)(ii), of 10 CFR Part 50, as approved by the NRC on December 6, 1985.MILLSTONE | * An exemption to Appendix J, Option A, paragraph III.D.2(b)(ii), of 10 CFR Part 50, as approved by the NRC on December 6, 1985. | ||
-UNIT 3 6-17 Amendment No. 69, -86,}} | MILLSTONE - UNIT 3 6-17 Amendment No. 69, -86,}} |
Latest revision as of 01:34, 23 March 2020
ML062640464 | |
Person / Time | |
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Site: | Millstone ![]() |
Issue date: | 09/15/2006 |
From: | NRC/NRR/ADRO/DORL/LPLB |
To: | |
Nerses V | |
Shared Package | |
ML061990025 | List: |
References | |
Download: ML062640464 (6) | |
Text
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.7-3 STEAM LINE SAFETY VALVES PER LOOP ................................... 3/4 7-3 Auxiliary Feedwater System ................................................................ 3/4 7-4 Demineralized Water Storage Tank ....................................................... 3/4 7-6 Specific A ctivity .................................................................................... 3/4 7-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM ................... 3/4 7-8 Main Steam Line Isolation Valves ........................................................ 3/4 7-9 Steam Generator Atmospheric Relief Bypass Lines ............................. 3/4 7-9a 3/4.7.2 DELETED ............................................................................................. 3/4 7-10 3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM ..3/4 7-11 3/4.7.4 SERVICE WATER SYSTEM ........................... 3/4 7-12 3/4.7.5 ULTIMATE HEAT SINK ...................................................................... 3/4 7-13 3/4.7.6 D ELETED ............................................................................................. 3/4 7-14 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM ......... 3/4 7-15 3/4.7.8 DELETED ............................................................................................. 3/4 7-18 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM ..................................... 3/4 7-20 3/4.7.10 SN UBBERS .......................................................................................... 3/4 7-22 TABLE 4.7-2 SNUBBER VISUAL INSPECTION INTERVAL ................................ 3/4 7-27 FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST .............. 3/4 7-29 3/4.7.11 DELETED ............................................................................................. 3/4 7-30 3/4.7.12 DELETED TABLE 3.7-4 DELETED TABLE 3.7-5 DELETED 3/4.7.13 DELETED 3/4.7.14 AREA TEMPERATURE MONITORING ........................................... 3/4 7-32 TABLE 3.7-6 AREA TEMPERATURE MONITORING ........................................... 3/4 7-33 MILLSTONE - UNIT 3 x Amendment No. 6-2, 84, 400, 4-60, 244, 232
INDEX BASES SECTION PAGE TABLE B 3/4.4-1 REACTOR VESSEL FRACTURE TOUGHNESS PROPERTIES .... B 3/4 4-9 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>IMeV) AS A FUNCTION OF FULL POWER SERVICE LIFE ......................................................... B 3/4 4-10 3/4.4.10 DELETED ................................................................................................... B 3/4 4-15 3/4.4.11 DELETED ................................................................................................... B 3/4 4-15 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUM ULATORS ..................................................................................... B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS ........................................................................ B 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK .................................................. B 3/4 5-2 3/4.5.5 pH TRISODIUM PHOSPHATE STORAGE BASKETS ............................ B 3/4 5-3 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT ..................................................................... B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS ............................... B 3/4 6-2 3/4.6.3 CONTAINMENT ISOLATION VALVES ................................................... B 3/4 6-3 3/4.6.4 DELETED 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM ........................ B 3/4 6-3d 3/4.6.6 SECONDARY CONTAINMENT ............................................................... B 3/4 6-4
'1/4-7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE ...................................................................................... B 3/4 7-1 3/4.7.2 DELETED .............................. .............................. B 3/4 7-7 3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM ........ B 3/4 7-7 3/4.7.4 SERVICE WATER SYSTEM ...................................................................... B 3/4 7-7 3/4.7.5 ULTIMATE HEAT SINK ............................................................................ B 3/4 7-8 3/4.7.6 DELETED ................................................................................................... B 3/4 7-10 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM ............... B 3/1 7-10 3/4.7.8 DELETED ................................................................................................... B 3/4 7-17 I 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM ............................................ B 3/4 7-23 3/4.7.10 SNUBBERS ................................................................................................. B 3/4 7-23 MILLSTONE - UNIT 3 xiv Amendment No. 49, 89,44-5,449, 4-36, 2"M, *-26,244,24-6, 224, 232
DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:
- a. All penetrations required to be closed during accident conditions are either:
- 1. Capable of being closed by an OPERABLE containment automatic isolation valve system*, or
- 2. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are opened under administrative control as permitted by Specification 3.6.3.
- b. All equipment hatches are closed and sealed,
- c. Each air lock is in compliance with the requirements of Specification 3.6.1.3,
- d. The containment leakage rates are within the limits of the Containment Leakage Rate Testing Program, and
- e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or O-rings) is OPERABLE.
CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.
CORE ALTERATIONS 1.9 CORE ALTERATIONS shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
DOSE EOUIVALENT 1-131 1.10 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microCurie/gram) which alone would produce the same CDE-thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed under "Inhalation" in Federal Guidance Report No. 11 (FGR 11),
"Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion."
- In MODE 4, the requirement for an OPERABLE containment isolation valve system is satisfied by use of the containment isolation actuation pushbuttons.
MILLSTONE - UNIT 3 1-2 Amendment No. 2-, -44, 4-6, 246, 232
THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE - UNIT 3 3/4 7-18 Amendment No. 48+-, 20a3, 2-9, 232
THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE - UNIT 3 3/4 7-19 Amendment No. 4243, 24, 206, 232
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
- 2) Pre-planned operating procedures and backup instrumentation to be used if one or more monitoring instruments become inoperable, and
- 3) Administrative procedures for returning inoperable instruments to OPERABLE status as soon as practicable.
- f. Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testingof the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions*. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 38.57 psig.
The maximum allowable containment leakage rate La, at Pa, shall be 0.3 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Leakage rate acceptance criteria are:
- 1) Containment overall leakage rate acceptance criterion is < 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the combined Type B and Type C tests, and <0.06 La for all penetrations that are Secondary Containment bypass leakage paths, and < 0.75 La for Type A tests;
- 2) Air lock testing acceptance criteria are:
- a. Overall air lock leakage rate is
- 0.05 La when tested at 2!Pa"
- b. For each door, seal leakage rate is < 0.01 La when pressurized to
> Pa The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.
The provisions of Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program.
- An exemption to Appendix J, Option A, paragraph III.D.2(b)(ii), of 10 CFR Part 50, as approved by the NRC on December 6, 1985.
MILLSTONE - UNIT 3 6-17 Amendment No. 69, -86,