ML20045C835
| ML20045C835 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 06/18/1993 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20045C834 | List: |
| References | |
| NUDOCS 9306240455 | |
| Download: ML20045C835 (85) | |
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ABWR oesign Docum:nt Table of Contents 1.0 Introduction 1.1 Definitions ,
1.2 General Provisions 2.0 Certified Design for ABWR Systems 2.1 Nuclear Steam Supply. Systems 2.1.1 Reactor Pressure Vessel System 2.1.2 Nuclear Boiler System 2.1.3 Reactor Recirculation System ;
2.2 Control and Instrument Systems 2.2.1 Rod Control and Information System 2.2.2 Control Rod Drive System ,
2.2.3 Feedwater Control System 2.2.4 Standby Liquid Control System 2.2.5 Neutron Monitoring System 2.2.6 Remote Shutdown System 2.2.7 Reactor Protection System 1 2.2.8 Recirculation Flow Control System 2.2.9 Automatic Power Regulator System O 2.2.10 2.2-11 Steam Bypass and Pressure Control System Process Computer
- 2.2.12 Refueling Platform Control Computer
- 2.2.13 CRD Removal Machine Control Computer 2.3 Radiation Monitoring Systems 2.3.1 Process Radiation Monitoring System 2.3.2 Area Radiation Monitoring System 2.3.3 Containment Atmospheric Monitoring System 2.4 Core Cooling Systems 2.4.1 Residual Heat Removal System i 2.4.2 High Pressure Core Flooder System 4 2.4.3 Leak Detection and Isolation System ;
2.4.4 Reactor Core Isolation Cooling System l
2.5 Reactor Senicing Equipment 2.5.1 Fuel Servicing Equipment ,
2.5.2 Miscellaneous Senicine Eauinment 2.5.3 Reactor Pressure Vessei Senicing Equipment 2.5.4 RPV Internal Senicing Equipment 2.5.5 Refueling Equipment :
2.5.6 Fuel Storage Facility j 2.5.7 Under-Vessel Servicing Eauipment i 2.5.8 CRD Maintenance Facility 2.5.9 Intemal Pump Maintenance Facility 6/17/93 9306240455 DR 930618 Y -i-
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ABWR D: sign Docum:nt 2.5.10 Fuel Cask Cleanine Facility O
v 2.5.11 2.5.12 Plant Start-up Test Eauipment Insenice Inspection Eauipment 2.6 Reactor Auxiliary Systems 2.6.1 Reactor Water Cleanup System 2.6.2 Fuel Pool Cooling and Cleanup System 2.6.3 Suppression Pool Cleanup System 2.7 Control Panels 2.7.1 Main Control Room Panels 2.7.2 Radioactive Waste Control Panels (2.9.1)**
2.7.3 Local Control Panels 2.7.4 Instrument Racks (2.7.3) 2.7.5 Multiplexing System 2.7.6 Local Control Boxes (2.7.3) 2.8 Nuclear Fuel 2.8.1 Nuclear Fuel 2.8.2 Fuel Channel 2.8.3 Control Rod 2.8.4 Loose Parts Monitoring System 2.9 Radioactive Waste System n 2.9.1 Radwaste System 2.9.2 Floor Drain Collection System
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2.10 Power Cycle Systems ,
2.10.1 Turbine Main Steam System ;
2.10.2 Condensate Feedwater and Condensate Air Extraction System i 2.10.3 Heater Drain and Vent System 2.10.4 Condensate Purification System !
2.10.5 Condensate Filter Facility (2.10.4) 2.10.6 Condensate Demineralizer (2.10.4) 2.10.7 Main Turbine 2.10.8 Turbine Control System (2.10.7) 2.10.9 Turbine Gland Seal System 2.10.10 Turbine Lubricating Oil System )
2.10.11 Moisture Separator Heater 2.10.12 Extraction System 2.10.13 Turbine Bypass System 2.10.14 Reactor Feedwater Pump Driver (2.10.2) l 2.10.15 Turbine Auxiliary Steam System l 2.10.16 Generator l 2.10.17 Hvdrocen Gas Cooling Svstem 2.10.18 Generator Cooling System 2.10.19 Generator Sealing Oil System
("' / 2.10.20 Exciter 2.10.21 Main Condenser 2.10.22 Off-Gas System 6/17/93 -ii-
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i 2.10.23 Circulating Water System 2.10.24 Condenser Cleanup Facility 2.11 Station Auxilian Systems 2.11.1 Makeup Water (Purified) System 2.11.2 Makeup Water (Condensate) System ,
2.11.3 Reactor Building Cooling Water System 2.11.4 Turbine Building Cooling Water System 2.11.5 HVAC Normal Cooling Water System '
l 2.11.6 HVAC Emergency Cooling Water System 2.11.7 Oxvgen Injection System 2.11.8 Ultimate Heat Sink (4.1) 2.11.9 Reactor Service Water System 2.11.10 Turbine Service Water System 2.11.11 Station Service Air System 2.11.12 Instrument Air System 2.11.13 High Pressure Nitrogen Gas Supply System 2.11.14 Heatine Steam and Condensate Water Retum System
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2.11.15 House Boiler 2.11.16 Hot Water Heating System 2.11.17 Hydrogen Water Chemistry System 2.11.18 Zinc Injection System 2.11.19 Breathing Air System O 2.11.20 2.11.21 Sampling System Freeze Protection System 2.11.22 Iron Injection System 2.12 Station Electrical Systems 2.12.1 Electrical Power Distribution System 2.12.2 Unit Auxilian Transformer (2.12.1) 2.12.3 Isolated Phase Bus 2.12.4 Nonsegrecated Phase Bus ,
2.12.5 Metal Clad Switchgear (2.12.1) 2.12.6 Power Center (2.12.1) 2.12.7 Motor Control Center (2.12.1) 2.12.8 Raceway System (2.12.1) 2.12.9 Grounding Wire (2.12.1) 2.12.10 Electrical Wiring Penetration 2.12.11 Combustion Turbine Generator 2.12.12 Direct Current Power Supply 2.12.13 Emergency Diesel Generator System (Standby AC Power Supply) 2.12.14 Vital AC Power Supply and AC Instrument and Control Power Supply Systems 2.12.15 Instrument and Control Power Supply (2.12.14) 2.12.16 Communication System 2.12.17 Lighting and Servicing Power Supply 2.13 Power Transmission 2.13.1 Reserve Auxillag Transformer (2.12.1) 6/17/93 -iii-
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ABWR a: sign Document 2.14 Containment and Emironmental Control Systems O
V 2.14.1 2.14.2 Primary Containment System Containment Internal Structures (2.14.1) 2.14.3 Reactor Pressure Vessel Pedestal (2.14.1) 2.14.4 Standby Gas Treatment System 2.14.5 PCV Pressure and Leak Testine Facility 2.14.6 Atmospheric Control System 2.14.7 Drywell Cooling System 2.14.8 Flammability Control System 2.14.9 Suppression Pool Temperature Monitoring System 2.15 Structures and Senicing Systems 2.15.1 Foundation Work (2.15.10) 2.15.2 Turbine Pedestal (2.15.11) 2.15.3 Cranes and Hoists 2.15.4 Elevator 2.15.5 Heating, Ventilating and Air Conditioning 2.15.6 Fire Protection System 2.15.7 Floor Leakage Detection System 2.15.8 Vacuum Sweep System 2.15.9 Decontamination Svstem 2.15.10 Reactor Building 2.15.11 Turbine Building 9 Control Building (d 2.15.12 2.15.13 Radwaste Building 2.15.14 Senice Building 2.16 Yard Structures and Equipment 2.16.1 Stack (2.15.10) 2.16.2 Oil Storage and Transfer System 2.16.3 Site Security 3.0 Non-System Based Material 3.1 Human Factors Engineering 3.2 Radiation Protection 3.3 Piping Design 3.4 Instmmentation and Control 3.5 Reliability Assurance Program 3.6 Initial Test Program 4.0 Interface Requirements 4.1 Ultimate Heat Sink 4.2 Offsite Power System (2.12.1)
C) 4.3 Potable and Sanitary Water System 4.4 Turbine Senice Water System (2.11.10) 6/17/93 -iv-
l ABWR oesign onum:nt 45 Reactor Service Water System (2.11.9) .
g 4.6 Makeup Water Preparation System 4.7 Communicadon System (2.12.16) 4.8 Airborne Particulate Radiation Monitoring 4.9 Heating, Ventilating and Air Conditioning 5.0 Site Parameters Appendices :
Appendix A Legend For Figures Appendix B Abbreviations and Acronyms l
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- Underlined sections -Title only, no entry for design certification'
' ** Section number in parentheses - Section, under which the subject is covered.
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ABWR D:signDocum:nt l
2.7 Control Panels 2.7.1 Main Control Room Panels Dedan Description The Main Control Room Panels (MCRP) is comprised of four major components. These are main control console, large display panel, the supervisor's console, the auxiliary or back panels, and their respective internal l
wiring. '
The MCRP locates and configures the alarms displays and controls for plant systems that contain Class IE equipment is classified as Seismic Category I.
Non< lass 1E and divisional Class 1E control and instrument power .., provided )
for the MCRP. Independence is provided between Class IE divisions and also i between the Class IE divisions and non41 ass 1E equipment.
The MCRP has the fixed alarms, displays, and controls shown on Table 2.7.la, inspections, Tests, Analyses and Acceptance Criteria Table 2.7.lb provides a definition of the inspections, tests and/or analyses, together with associated acceptance criteria, which will be undertaken for the MCRP.
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O O 3' Table 2.7.1a Main Control Room Panels Fixed Position Alarms, Displays and Controls 5
A. Fixed Position Controls Manual Scram initiation Switch (A) DG (A) Start Switch Div. I Manual / Auto Main Steamline isolation Reset Switch Manual Scram Initiation Switch (B) DG (B) Start Switch Div.11 Manual / Auto Main Steamline Isolation Reset Switch Reactor Mode Switch DG (C) Start Switch Div. til Manual / Auto Main Steamline Isolation Reset Switch Div. I Main Steamline Manual iso!ation RCIC System Standby Mode initiation Switch Div. IV Manual / Auto Main Steamline isolation Switch Reset Switch Div.11 Main Steamline Manual isolation Condensate Pump Standby Mode initiation Primary Containment Div. I isolation Reset !
Switch Switches Switch Div. til Main Steamline Manual isolation Reactor Feedpump Standby Mode initiation Primary Containment Div. Il isolation Reset Switch Switches Switch Div. IV Main Steamline Manual isolation Condensate Pump Startup Mode Initiation Primary Containment Div.111 Isolation Reset Switch Switches Switch g
Primary Containment Div. I Manual isolation Reactor Feedpump Startup Mode Initiation RHR (A) Shutdown Cooling Mode initiation Switch - Switches Switch
- Primary Containment Div. Il Manual isolation SLC (A) Pump Control Switch RHR (B) Shutdown Cooling Mode Initiation Switch Switch L
Primary Containment Div.111 Manual isolation SLC (B) Pump Control Switch RHR (C) Shutdown Cooling Mode initiation Switch Switch i RCIC Initiation Switch ADS (A) Inhibit Switch ARI (A) Manual Initiation Switch HPCF (B) Initiation Switch ADS (B) Inhibit Switch ARI (B) Manual initiation Switch HPCF (C) Initiation Switch RHR (A) Standby Mode Switch Recirculation Runnack initiation Switch (A)
RHR (A) Initiation Switch RHR (B) Standby Mode Switch Recirculation Runback initiation Switch (B)
' RIP Start /Stop Control Switch (10)
RHR (B) Initiation Switch RHR (C) Standby Mode Switch RHR (C) Initiation Switch Main Steam isolation Valve Control Switch (8) ARI (A) Logic Reset Switch
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n Om O, U g Taiale 2.7.1a Main Control Room Panels Fixed Position Alarms, Displays and Controls (Continued) e 8 A. Fixed Position Controls (Continued)
ARI (B) Logic Reset Switch RHR (A) Suppression Pool Cooling Mode Div 11 ADS Manual ADS Channel 2 Initiation initiation Switch Switch CRD Charging Water Pressure Low Scram RHR (B) Suppression Pool Cooling Mode RCIC Div. I isolation Logic Reset Switch Byoass Switch (A) Initiation Switch CRD Charging Water Pressure Low Scram RHR (C) Suppression Pool Cooling Mode RCIC Div. Il isolation Logic Reset Switch Bypass Switch (B) Initiation Switch CRD Charging Water Pressure Low Lam RHR (B) Drywell Spray Mode initiation Switch ' RCIC Inboard Isolation Control Switch Bypass Switch (C)
CRD Charging Water Pressure Low Scram RHR (C) Drywell Spray Mode initiation Switch RCIC Outboard Isolation Control Switch Bypass Switch (D)
Manual Scram Reset Switch SGTS (A) Initiation Switch Fire Protection System Motor Pump Control Switch RPS Div.ITrip Reset Switch SGTS (B) Initiation Switch Fire Protection System Diesel Pump Control ^
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- FCS (A) Control Switch RPS Div. ll Trip Reset Switch Div. I Maaual ADS Channel 1 Initiation Switch RPS Div. ll1 Trip Reset Switch Div 1 Manual ADS Channel 2 Initiation Switch FCS (B) Control Switch RPS Div. IV Trip Reset Switch Div 11 Manual ADS Channel 1 Initiation Switch.
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O O O 3 Table 2.7.1a Main Control Room Panels Fixed Position Alarms, Displays and Controls (Continued)
B. Fixed Position Displays RPV Water Level "tCIC Flow Main Condenser Pressure RCIC Turbine Speed RCIC Injection Valve Status SRV Positions Wetwell Pressure HPCF (B) Injection Valve Status Suppression Pool Level Suppression Pool Bulk Average Temperature HPCF (C) Injection valve status Main Steamline Flow HPCF (B) Flow RHR (A) Flow SLC Boron Tank Water Level HPCF (C) Flow RHR (A) injection Valve Status Recirculation Pump Speeds RPV Pressure RHR (B) Flow Average Drywell Temperature Drywell Pressure RHR (B) injection Valve Status Wetwell Hydrogen Concentration Level Reactor Power Level,(Neutron Flux, APRM) RHR (C) Flow Drywell Hydrogen Concentration Level Reactor Power Level (SRNM) RHR (C) Injection Valve Status Drywell Oxygen Concentration Reactor Thermal Power Emergency Diesel Generator (A) Operating Wetwell Oxygen Concentration
? Status MSIV Position Status (Inboard And Outboard Emergency Diesel Generator (B) Operating FCS (A) Operating Status Valves) Status Reactor Mode Switch Mode Indications Emergency Diesel Generator (C) Operating FCS (B) Operating Status Status Main Steam Line Radiation Primary Containment Water Level Main Stack Radiation Level Scram Solenoid Lights (8) Status Condensate Storage Tank Water Level Time Manual Scram Switch (A) Indicating Light SLC Pump (A) Discharge Pressure Drywell Radiation Level Status Manual Scram Switch (B) Indicating Light SLC Pump (B) Discharge Pressure Wetwell Radiation Level Status RPV Isolation Status Display Main Condenser Pressure w
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b' (d V 3 Table 2.7.1a Main Control Room Panels Fixed Position Alarms, Displays 9nd Controls (Continued)
$" C. Fixed Position Alarms
- Indicated RPV Water Level Abnormat RPV Water Level Low (ECCS Initiation) CAN. h Level High RPV Water Level Low (Scram Level) Control Rod Not inserted To/Beyond MSBWP CAMS (A) System Abnormal RPV Pressure High RPV Water Level High CAMS (B) System Abnormal Drywell Pressure High Fire Protection Systern Status Re setor Building AP Low Neutron Flux High-High ADS (A) Logic Initiated Area Temperature High Neutron Monitoring System inoperative ADS (B) Logic Initiated Area HVAC AT High MSIV Closure SRV Open RBHVAC Exhaust Radiation High CRD Charging Water Pressure Low Main Steam Line Flow High Reactor Building Area Radiatior: High Rapid Core Flow Decrease HPIN (A) System Status Reactor Building Floor Drain Sump Water Level High-High Main Turbine Trip HPIN (B) System Status RBHVAC System Status i' Main Generator Trip Leak Detection Isolation Stack Radioactivity High Main Steam Line Radiation High RWCU System Status RCW Radioactivity High Reactor Scram Reactor Period Short Radwaste Effluent Radioactivity High RPV Low Level isolation incomplete (Scram ADS Div. Iinhibited/ Auto Out Of Service Turbine Building Ventilation System (TBVS)
Water Level) Status RPV Low Level isolation incomplete (ATWS ADS Div. Ilinhibited/ Auto Out Of Service Radiation Monitor High Scram Level)
RPV Low Level /Drywell Pressure High Isolation Suppression Pool Bulk Average Temperature RCIC System Status incomplete High RPV Water Level Low (ATWS Scram Level) Drywell Average Temperature High HPCF (B) System Status RPV Water Level Low (HPCF Initiation Level) Suppression Pool Water Level High/ Low HPCF (C) System Status
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" Inspections, Tests, Analyses and Acceptance Criteria 1 Design Commitment inspections. Tests, Analyses Acceptance Criteria ,
- 1. Equipment comprising the MCRP is 1. Inspections of the as-built system will be 1. The as-built MCRP conforms with the defined in Section 2.7.1. conducted. description in Section 2.7.1.
In the MCRP, independence is provided 2a. Tests will be conducted on the MCRP by 2a. The test signal exists only in Class 1E 2.
between Class 1E divisions, and between providing a test signal to only one Class 1E division under test in the MCRP.
Class 1E divisions and non-Class 1E division at a time, equipment.
2b. Inspections of the as-built Class 1E 2b. In the MCRP, physical separation exists divisions in the MCRP will be conducted. between Class 1E divisions. Physical separation exists between these Class 1E divisions and non-Class 1E equipment.
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ABWR 0: sign cocum:nt 2.8 Nuclear Fuel :
2.8.1 Fuel Bundle Design Description It is intended that the specific fuel to be utilized in any facility which has adopted' i the certified design be in compliance with U.S. NRC approved fuel design criteria. This strategy is intended to permit future use of enhanced / improved fuel designs as they become available.
The following is a summary of the principal requirements which must be met by the fuel supplied to any facility utilizing the certified design.
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l The fuel bundle is evaluated using methods and criteria to assure that:
(1) Fuel rod failure does not occur as a result of normal operation and anticipated operational occurrences.
(2) Control rod insertion will not be prevented as a result of normal operation, anticipated operational occurrences or postulated accident. {
(3) The number of fuel rod failures will not be underestimated for
- postulated accidents.
(4) Coolability will be maintained. ]
(5) Specified acceptable fuel design limits (thermal and mechanical design - l limits) will not b :xceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
l (6) In the power operating ranges, the prompt inherent nuclear feedback characteristics will tend to compensate for a rapid increase in reactivity.
(7) The reactor core and associated coolant, control and protection systems will be designed to assure that power oscillations which can result in 1 conditions exceeding specified acceptable fuel design limits are not l possible or can be reliably and readily detected and suppressed. l, 1
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ABWR c: sign occum:nt ,
2.8.2 Fuel Channel Design Description The fuel channels are zirconium-based _(or equivalent) and preclude cross-flow in the core region.
The following is a summary of the principal design criteria which are met by the fuel channels: .,
(1) During any design basis event, fuel channel damage will not be so severe as to prevent control rod insertion when it is required.
(2) Coolability will be maintained for all design basis events. ;
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ABWR D: sign Docum:nt 2.8.3 Control Rod
'J Design Description Control rods in the reactor perform dual functions of power distribution shaping and reactivity control and have the following design features:
(1) A cruciform cross-sectional envelope shape.
(2) A coupling at the bottom for attachment to the control rod drive.
(3) Contain neutron absorbing materials.
The following is a summary of the principal design criteria which are met by the control rods.
(1) The control rod stresses, strains, and cmamative fatigue will be evaluated to not exceed the ultimate stress or strain of the material.
(2) The control rod will be evaluated to be capable ofinsertion into the core during design basis modes of operation.
(3) The material of the control rod will be compatible with the reactor
( emironment.
(4) The reactivity worth of the control rod will be included in the plant core analyses.
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ABWR D sign Documint . l q 2.9 Radioactive Waste Systems !
V 2.9.1 Radwaste System Design Description The Radwaste (RW) System consists of a liquid waste system and a solid waste system. The liquid waste system includes primary containment penetrations, and inboard and outboard motor-operated isolation valves for the high conductivity and low conductivity waste drains from the lower drywell. The liquid waste system collects, treats, monitors, and either recycles or discharges radioactive liquid wastes within the plant. The solid waste system collects, sorts, monitors and either recycles or packages radioactive solid wastes within the plant.
The RW System is classified as non-safety-related with the exception of the primary containment isolation function.
The primary containment penetrations and isolation valves are classified as Seismic Category I, and ASME Code Class 2.
The inboard containment isolation valves are powered from Class 1E Division II, and the outboard isolation valve are powered from Class 1E Division I. In the RW p System, independence is provided is between Class 1E divisions, and between the j () Class 1E divisions and non-Class 1E equipment.
l l The main control room has control and open/close status indication for the ,
primary contamment isolation valves.
The safety-related electrical equipment that provides containment isolation and is located in the containment and Reactor Building is qualified for a harsh I environment. l The primary containment isolation motoroperated valves (MOVs) have active safety-related functions and perform these functions under differential pressure, .
fluid flow, and temperature conditions.
The liquid waste system has one discharge line which has a radiation monitor.
Discharge flow is terminated on receipt of a high radiation signal from this monitor.
I Inspections, Tests, Analyses and Acceptance Criteria '
I l Table 2.9.1 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be used for the Radwaste i f System.
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i 3 Table 2.9.1 Radwaste System i l
inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The basic configuration for the RW System 1. Inspection of the as-built system will be 1. The as-built RW System conforms with the is described in Section 2.9.1. conducted. basic configuration described in Section 2.9.1.
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- 2. The ASME Code components of the RW 2. A hydrostatic test will be conducted on 2. The results of the hydrostatic test of the !
System retain their pressure boundary those Code components of the RW System ASME Code components of the RW i integrity under internal pressures that will required to be hydrostatically tested by the System conform with the requirements in !
be experienced during service. ASME Code. the ASME Code, Section 111.
. ' 3. In the RW System, independence is 3a. Tests will be performed on the RW System 3a. The test signal exists only in the Class 1E provided between Class 1E divisions, and by providing a test signal in only one Class division under test in the RW System. ,
between Class 1E divisions and betweer. 1E division at a time.
Class 1E divisions and non-Class 1E :
equipment.
J 3b. Inspection of the as-installed Class 1E 3b. In the RW System, physical separation divisions in the RW System will be - exists between Class 1E divisions. Physical l Y performed. separation exists between these Class 1E divisions and non-Class 1E equipment.
- 4. Main control room displays and controls 4. Inspections will be performed on the main 4. Displays and controls exist or can be ,
- provided for RW System are as defined in control room displays and controls for the retrieved in main control room as defined Section 2.9.1. RW System. in Section 2.9.1.
! 5. MOVs designated in Section 2.9.1 as 5. Closing tests of installed valves will be 5. Each MOV closes.
having an active safety-related function conducted under pre-operational !
close under differential pressure fluid flow, differential pressure, fluid flow, and and temperature conditions. temperature conditions.
. 6. The liquid waste system has one discharge 6. Tests will be conducted on the as-built 6. The discharge flow terminates upon line which has a radiation monitor. liquid waste system using a simulated high receipt of a simulated high radiation signal.
1 Discharge flow is terminated on receipt of radiation signal. !
! a high radiation signal from this monitor. t l
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ABWR oesign Document 2.9.2 Floor Drain Collection System O Design Description Non Radioactive Floor Drain Collection System The non-radioactive floor drain collection system collects non-radioactive waste liquids for processing. The non-radioactive floor drain collection system consists of passive drainage piping. Flow is by gravity and the system has no valves, pumps, or other active components in the drainage paths.
Figure 2.9.2a shows the basic system configuration and scope.
The non-radioactive floor drain collection system is classified as non-safety-rclated. The system is located in the Control Building and in the Reactor Building outside the secondary containment.
The non-radioactive floor drain collection system in each divisional area of the Control Building and Reactor Building is physically separated from drains in the 3 other divisions, except for the sump for outside secondary containment Divisions A and C. Independence is provided between the non-radioactive floor drain collection system and radioactive floor drain collection system.
l Radioactive Floor Drain Collection System The radioactive floor drain collection system collects waste liquids from floor !
l drains, located in potentially radioactive contaminated areas, for processing.
The radioactive floor drain collection system consists of passive drainage piping.
Flow is by gravity and the system has no valves, pumps, or other active components in the drainage paths.
Figure 2.9.2b shows the basic configuration and scope.
The radioactive floor drain collection system is classified as non-safety-related.
The radioacthe floor drain collection system is located in the Reactor Building inside the secondary containment.
The radioactive floor drain collection system in each divisional area is physically separated from drains in the other divisions except for the sumps associated with the drains outside the emergency core cooling system (ECCS) pump rooms.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.9.2 provides a definition of the inspections, tests, and/or analyses, .
L together with associated acceptance criteria, which will be undertaken for the l
Floor Drain Collection System.
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ABWR c: sign Document DIVISION A DIVISION B DIVISION C ROOMS WITH ROOMS WITH ROOMS WITH FLOOR DRAINS FLOOR DRAINS FLOOR DRAINS V V V r-----l DIVISION A r-----l DIVISION B r-----lDIVISION C l g l l ,
l l SUMP SUMP l SUMP l l CONTROL BUILDING DIVISION B DIVISION A DIVISION C ROOMS WITH ROOMS WITH ROOMS WITH FLOOR DRAINS FLOOR DRAINS FLOOR DRAINS l
V V V r-----l ----------------1 l DIVISION B ! DIVISION A AND C l l
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i REACTOR BUILDING - OUTSIDE SECONDARY CONTAINMENT NOTES:
- 1. THE SYSTEM HAS NO VALVES, PUMPS, OR OTHER ACTIVE COMPONENTS IN THE DRAINAGE PATHS.
l Figure 2.9.2a Non Radioactive Floor Drain Collection System 6/17/93 -2 2.9.2 l
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DIVISION A DIVISION B DIVISION C l
ECCS PUMP ECCS PUMP ECCS PUMP ROOMS ROOMS ROOMS V V V '
F - --' - - ; r-----l r-----l DIVISION C l DIVISION A l DIVISION B g l l g
SUMP SUMP SUMP l l l EMERGENCY CORE COOLING SYSTEM (ECCS) AREAS i
DIVISION B DIVISION C DIVISION A '
ROOMS WITH ROOMS WITH ROOMS WITH FLOOR DRAINS FLOOR DRAINS FLOOR DRAINS V V V r-----l r----; i
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SECONDARY CONTAINMENT-OTHER AREAS NOTES:
- 1. THE SYSTEM HAS NO VALVES, PUMPS, OR OTHER ACTIVE COMPONENTS IN THE DRAINAGE PATHS.
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O Figure 2.9.2b . Radioactive Floor Drain Collection System 6/17/93 2.9.2
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Table 2.9.2 Floor Drain Collection System is Inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The basic configuration for the Floor Drain 1. Inspection of the as-built system will be - 1. The as-built Floor Drain Collection System Collection System is as shown on Figures conducted. conforms with the basic configuration 2.9.2a and 2.9.2b. shown on Figures 2.9.2a and 2.9.2b.
- 2. The nonradioactive floor drain collection 2. Tests will be conducted on the as-built 2. No interconnections exist (i.e. no water system in each divisional area of the system by individually pressurizing each leakage in to other divisions not being Control Building and Reactor Building is divisional area drains with water and tested), except for the sump for outside physically separated from drains in the observing other divisional area drains for secondary containment Divisions A and C.
other divisions, except for the sump for interdivisonal leakage.
outside secondary containment Divisions A and C.
- 3. Independence is provided between the 3. Tests will be conducted on the as-built 2. No interconnections exist (i.e. no water in-non-radioactive floor drain collection Floor Drain Collection System by leakage from the radioactive floor drains is system and the radioactive floor drain pressurizing all radioactive floor drains observed).
collection system. with water and observing the non-
- . radioactive floor drains for evidence of in-leakage from the radioactive floor drains.
- 4. The radioactive floor drain collection 4. Tests will be conducted on the as-built 4. No interconnections exist (i.e. no water system in each divisional area is physically system by individually pressurizing each leakage in to other divisions not being separated from drains in the other divisional area drains with water and tested), except for the sumps associated divisions, except for the sumps associated observing other divisional area drains for with the drains outside the ECCS pump interdivisonal leakage. rooms.
with the drains outside the ECCS pump rooms. ,
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ABWR 0: sign Documsnt 2.11 Station Auxiliary Systems 2.11.1 Makeup Water (Purified) System Design Description The Makeup Water (Purified) (MUWP) System provides demineralized makeup water to the condensate storage tank, the surge tanks which are shared by the Reactor Building Cooling Water System and Headng, Ventilation, and Air Conditioning Emergency Cooling Water System and other plant systems.
The MUWP System consists of distribution piping and valves. Makeup water is supplied to the system by the Makeup Water Preparation System.
The MUWP System is classified as non-safety-related.
Inspections, Tests, Analyses and Acceptance Criteria None required for this system.
O O
' 6/15/93 -1 2.11-
l ABWR D: sign Documint 2.11.2 Makeup Water (Condensate) System O~ Design Description i The Makeup Water (Condensate) (MUWC) System provides water to various plant systems. Figure 2.11.2 shows the basic system configuration and scope, f Except for the level sensors and associated piping, the MUWC System is classified )
I as non-safety-related.
i The level sensors and associated piping are classified as Seismic Category I.
4 Figure 2.11.2 shows the ASME Code class for the MUWC System piping and i components, i
The level instruments are located in the Reactor Building; the condensate
)
j storage tank (CST) and pump (s) are located outside.
l Each of the four MUWC System water level sensors is powered from the 4
respective divisional Class 1E power supply. In the MUWC System, independence is provided betweer 'he Class 1E divisions, and also between the Class 1E divisions and nonClass IE equipment.
i The MUWC System has parameter displays for CST water level in the main control room.
MUWC System components with display interfaces with the Remote Shutdown
? System (RSS) are shown on Figure 2.11.2.
i l Inspections, Tests, Analyses and Acceptance Criteria i
Table 2.11.2 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the MUWC System.
-0 v
6/18/93 1 *4.11.2
O O O
-3 3w MUWC MUWP NNS
~~ ~
MUWC RW/B NNS
~
U MUWC M NOTE 2 R l - "_"S p.
CST -- - -1 NOTE 1 ! MUWC R/B (RCIC, b 2l NNS I NNS HPCF & RHR) 4--- -
I 2
CRD MUWC 1 I I i e- -.
NNS HPCF MUWC
' NOTES:
- 1. ONE SENSOR ASSIGNED TO EACH OF FOUR CLASS 1E DIVISIONS.
"2. RSS INTERFACE IS WITH DIVISION ll LEVEL SENSOR ONLY.
l 1
w 3 Figure 2.11.2 Makeup Water (Condensate) System
- e eim e w -.m - e, w ~ w W w a
3 Table 2.11.2 Makeup Water (Condensate) (MUWC) System t 8
inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria l
- 1. The basic configuration of the MUWC 1. Inspections of the as-built system will be 1. The as-built MUWC System conforms with ;
System is as shown on Figure 2.11.2. conducted. the basic configuration on Figure 2.11.2. l
- 2. In the MUWC System, independence is 2a. Tests will be performed on the MUWC 2a. The test signal exists only in the Class 1E provided between Class 1E divisions, and System by providing a test signalin only division under test in the MUWC System. l between Class 1E divisions and non-Class one Class 1E division at a timc.
l 1E equipment.
2b. Inspection of the as-built Class 1E divisions 2b. In the MUWC System, physical separation ;
in the MUWC System will be performed. exists between Class 1E divisions. Physical i separation exists between these Class 1E divisions and non-Class 1E equipment. l
- 3. Main control room displays provided for 3. Inspections will be performed on the main 3. Displays exist or can be retrieved in the ;
the MUWC System are as defined in control room displays for the MUWC ' main control room as defined in Section Section 2.11.2 System. 2.11.2.
- 4. Remote Shutdown System (RSS) displays 4. Inspections will be performed on the RSS 4. Displays exist on the RSS as defined in Y provided for the MUWC System are as displays for the MUWC System. Section 2.11.2.
defined in Section 2.11.2.
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ABWR Desi.qn Docum:nt 2.11.5 HVAC Normal Cooling Water System Design Description The Heating Ventilating and Air Conditioning.(HVAC) Normal Cooling Water (HNCW) System delivers chilled water to the Drywell Cooling System and to non-safety-related fan coil units of building HVAC systems. Figure 2.11.5 shows -
the basic system configuration and scope.
The HNCW System is classified as non-safety-related with the exception of the primary containment isolation function. .
The HNCW System pumps and refrigerators are located in the Turbine Building.
The priman containment penetrations and isolation valves are classified as - '
Seismic Category I, and AShiE Code Class 2.
The inboard containment isolation valves is powered from Class IE Division II, ;
and the outboard isolation valves are powered from Class IE Division I. In the HNCW System, independence is provided is between Class IE divisions, and also '
between the Class 1E divisions and non-Class 1E equipment.
The main control room has control and open/close status indication for the primary containment isolation valves.
The safety-related electrical equipment that provides containment isolation and is located in the Reactor Building is qualified for a harsh environment. ,
The primary containment isolation motor-operated valves (hiOVs) have active 1 i safety-related functions and perform these functions under differential pressure, I
fluid flow, and temperature conditions.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.5 provides a definition of the inspections, tests, and/o: analyses, together with associated acceptance criteria, which will be undertaken for the HNCW System. i i
6/18/93 ~ 2.11.5 l- - - _
, __- , _ _ . . _ ,- _ .__ ,_...-c,....- __ - .. .
ABWR oesign Docum:nt O
HNCW HVAC HVAC HNCW TO SURGE TANK TCW JL COOLING LOADS HNCW NNS - REACTOR BUILDING I HVAC SYSTEM t - CONTROL BUILDING '
I SAFETY-RELATED l
EQUIPMENT AREA
{ HVAC SYSTEM g HNCW DWC DWC HNCW NNS NNS l NOTE 1 7 NOTE 1 , , 7 NOTE 1 i j / DRYWELL f/
l i -
/
COOLING SYSTEM W /
j y
u
-l NNS 2 f 2 NNS _ . NNS 2
% 2 NNS, I d PRIMARY PRIMARY I I CONTAINMENT CONTAINMENT I I l 1 I
l 1
l
_ _ REFRIGERATOR (S) . _ __
u_ _
_o NNS HNCW PUMP (S) JL TCW V
NOTES:
- 1. THE INB'OARD ISOLATION VALVE IS POWERED FROM CLASS 1E DIVISION 11, AND
.f THE OUTBOARD ISOLATION VALVES ARE POWERED FROM CLASS 1E DIVISION 1.
i
(
Figure 2.11.5 HVAC Normal Cooling Water System 6/18/93 -2 2.11.5
)
Table 2.11.5 HVAC Normal Cooling Water (HNCW) System
-l is Inspections, Tests, Analyses and Acceptance Critoria Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The basic configuration of the HNCW 1. Inspections of the as-built system will be 1. The as-built HNCW System conforms with System is as shown on Figure 2.11.5. conducted. the basic configuration shown in Figure 2.11.5.
- 2. The ASME Code components of the HNCW 2. A hydrostatic test will be conducted on 2. The results of the hydrostatic test of the retain their pressure boundary integrity . those code components of the HNCW ASME Code components of the HNCW System conform with the requirements in under internal pressures that will be System required to be hydrostatically experienced during service. tested by the ASME Code. the ASME Code, Section Ill.
- 3. In the HNCW System, independence is 3a. Tests will be performed on the HNCW - 3a. The test signal exists only in the Class 1E provided between Class 1E divisions, and System by providing a test signal in only division under test in the HNCW System.
between Class 1E divisions and non-Class one Class 1E division at a time.
1E equipment.
3b. Inspection of the as-installed Class 1E 3b. In the HNCW System, physical separation divisions in the HNCW System will be exists between Class 1E divisions. Physical performed. separation exists between these Class 1E Y divisions and non-Class 1E equipment.
A
- 4. Main control room displays and controls 4. Inspections will be performed on the main 4. Displays and controls exist or can be provided for HNCW System are as defined control room displays and controls for the retrieved in main control room as defined in Section 2.11.5. HNCW System. in Section 2.11.5.
, 5. Motor-operated valves (MOV) designated 5. Closing tests of installed valves will be 5. Each MOV closes.
in Section 2.11.5 as'having an active safety- conducted under preoperational related function, close under differential differential pressure, fluid flow, and pressure, fluid flow, and temperature temperature conditions.
conditions.
i 1
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ABWR Design Documsnt i 2.11.6 HVAC Emergency Cooling Water System Design Description ;
The Heating Ventilating and Air Conditioning (HVAC) Emergency Cooling )
1 Water (HECW) System delivers chilled water to the:
(1) Control Room Habitability Area HVAC System.
I (2) Control Building Safety-Related Equipment Area HVAC System. {
(3) Reactor Building HVAC System (safety-related electrical equipment I HVAC).
Figures 2.11.6a and 2.11.6b show the basic system configuration and scope.
I The HECW System is classified as safety-related except for the chemical addition tank and associated piping and valves.
The HECW System is manually inidated.
Each HECW System refrigerator unit has a capacity of not less than 5.8 x'10 5 kctl/hr. In Divisions B and C, any refrigerator unit on standby automatically l st if any of the other refrigerator units in Divisions B or C are stopped.
All safety-related portions of the HECW System are classified as Seismic Category I. Figures 2.11.6a and 2.11.6b show the ASME Code class for the HECW System ,
piping and components. l The HECW System pumps and refrigerator units are located in the Control Building.
Each of the three HECW System divisions is powered from the respective Class 1E divisions as shown on Figures 2.11.6a and 2.11.6b. In the HECW System, -
l independence is provided between Class IE divisions, and also between Class 1E l divisions and non-Class 1E equipment.
l Except for the connections to the chemical addition tanks, each mechanical !
l division of the HECW System (Divisions A, B, C) is physically separated from the other divisions.
l r l X
6/18/93 -1 2.11.6
ABWR D: sign Documsnt The HECW System has the following main control room (MCR) displays and O controls:
(1) Control and status indications for the refrigerator units and pumps shown on Figure 2.11.6a and 2.11.6b.
(2) Parameter displays for instruments shown on Figures 2.11.6a and
- 2.11.6b. 1 The pneurnatic-operated valves shown in Figures 2.11.6a and 2.11.6b fail as 1 follows in the event that either electric power to the valve-actuating solenoid is l lost or pneumatic pressure to the valve is lost. The differential pressure control valves fail close, and the flow control valves to the cooling coils fail open.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.6 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the HECW System.
1 O
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'6/18/93 .2 2.11.6 l- i
. - - . .. ., . . . . -_ a
- ABWR Design Document G
V HECW HkAC HVAC HECW TO SURGE TANK RCW Jk REACTOR BU!LDING HVAC HECW 3 SYSTEM (SAFETY-RELATED ---
ELECTRICAL EQUIPMENT HVAC)
/
l I l CONTROL BUILDING SAFETY-RELATED EQUIPMENT -l AREA HVAC SYSTEM j M
3 NNS CHEMICAL NNS 3 4 O-- ADDITION TANK
---W- O l N REFRIGERATOR O 3 HECW JL RCW V
l NOTES:
- 1. DIVISION A IS POWERED FROM CLASS 1E, DIVISION 1.
O Figure 2.11.6a HVAC Emergency Cooling Water System (Division A) 6/18/93 -3 2.11.6 l
l ABWR Design Docum:nt i
l l
l HECW HVAC HVAC HECW 3 3 l' TO SURGE TANK RCW Jk REACTOR BUILDING HVAC j l HECW 3 SYSTEM (SAFETY-RELATED :_: '
r ELECTRICAL EQUIPMENT HVAC) ;
I CONTROL ROOM ,
HABITABILITY AREA --, .
HVAC SYSTEM I l
( CONTROL BUILDING
! SAFETY-RELATED EQUIPMENT _'
AREA HVAC SYSTEM O
O 3 NNS CHEMICAL NNS 3 4 O-- ADDITION TANK
---W- O V
(o N REFRIGERATOR 3 HECW JL RCW V
l l
O N REFRIGERATOR 3 HECW JL RCW V
l NOTES:
- 1. THIS FIGURE SHOWS ONE OF TWO IDENTICAL DIVISIONS.
r~g DIVISIONS B AND C ARE POWERED FROM CLASS 1E,
( DIVISION 11 AND 111, RESPECTIVELY.
Figure 2.11.6b HVAC Emergency Cooling Water System (Divisions B and C) 6/18/93 4 2.11.6
O 3 Table 2.11.6 HVAC Emergency Cooling Water System
> .g Inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The basic configuration for the HECW 1. Visualinspections of the as-built system 1. The as-built configuration of the HECW System is shown on Figures 2.11.6a and configuration will be conducted. System is in accordance with 2.11.6b. Figures 2.11.6a and 2.11.6b.
- 2. The ASME Code components of the HECW 2. A hydrostatic test will be conducted on 2. The results of the hydrostatic test of the System retain their integrity under intemal those Code components of the HECW ASME Code components of the HECW pressures that will be experienced during System required to be hydrostatically System conform with the requirements in service. tested by the ASME Code the ASME Code, Section Ill.
- 3. Each HEWC System refrigerator unit has a 3. Tests will be conducted on an as-built 3. Each HEWC System refrigerator unit has a e capacity of not less than 5.8 x 105 kcal/hr. HECW System refrigerator units at a test capacity of not less than 5.8 x 105 kcal/hr.
facility.
- 4. In Divisions B and C, any refrigerator unit 4. Tests will be conducted on each as-built 4. In Divisions B and C, any refrigerator unit on standby automatically starts if any of HECW System refrigerator unit in Divisions on standby automatically starts if any of the other refrigerator units in Divisions B or B and C, using simulated signals indicating the other refrigerator units in Divisions B or
. C are stopped. another refrigerator unit is stopped. C are stopped.
T
- 5. In the HECW System, independence is Sa. Tests will be performed on the HECW Sa. The test signal exists only in the Class 1E provided between Class 1E divisions, and System by providing a test signalin only division under test in the HECW System.
. between Class 1E divisions and non-Class one Class 1E division at a time.
1E equipment. .
..Sb. In the HECW System, physical separation Sb. Inspection of the as-built Class 1E divisions in the HECW System will be performed. exists between Class 1E divisions. Physical separation exists between these Class 1E
' divisions and non-Class 1E equipment.
- 6. Except for the connections to the chemical 6. Inspections of the as-built HECW System 6. Each mechanical division of the HECW addition tank, each mechanical division of - will be conducted. System is physically separated from the the HECW System (Divisions A, B, C) is other mechanical divisions of the HECW physically separated from the other System by structural and/or fire barriers, divisions. with the exception connections to the chemical addition tank.
- 7. Main control room displays and controls 7. Inspections will be performed on the main 7. Displays and controls exist or can be provided for the HECW System are as control room displays and controls for the retrieved in the main control room as
~
defined in Section 2.11.6. HECW System. defined in Section 2.11.6.
to 2
' bi
O O J ,
3_ Table 2.11.6 HVAC Emergency Cooling Water System (Continued) w inspections, Tests, Analyses and Acceptance Criteria i
Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 8. The pneumatic-operated valves shown in 8. Tests will be performed on the as-built 8. The pneumatic actuated valves listed ;
' Fig ,res 2.11.3a and 2.11.3b fail as follows valves by initiating loss of pneumatic below fail as desired when either electric in the event that either electric power to the pressure and power to the actuating power to the valve actuating solenoid is valve actuating solenoid is lost or solenoids. lost or pneumatic pressure to the valve is pneumatic pressure to the valve is lost: The lost: The differential pressure control
- differential pressure control valves fail valves fait close, and the flow control valves to the cooling coils fail open. l, close, and the flow control valves to the cooling coils fail open.
I i
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d
. 6
ABWR D: sign Docum:nt 214 Containment and Environmental Control Systems
(%
V 2.14.1 Primary Containment System Design Description The Primay Containment System (PCS) encornpasses:
(1) A reinforced concrete containment stmcture with an internal steel liner. The structure includes various penetrations, equipment hatches and personnel access locks.
l (2) Structures inside the primay containment which partition the
! containment into dowell and wetwell regions, provide equipment support, radiation protection, and components for operation of the ABWR pressure suppression containment.
Figure 2.14.1 shows the basic configuration and scope.
The steel lined reinforced concrete containment structure attached to a l reinforced concrete basemat provides the primary containment pressure barrier i and is classified as ASME Code Section III, Division 2. The Reactor Pressure l Vessel (RPV) support pedestal and a diaphragm floor partition the containment volume into dowell and wetwell regions. The RPV support pedestal is a steel O' structure with concrete fill material. The diaphragm floor is a reinforced concrete structure. Other major internal structures within the containment are the reactor shield wall, lower dnwell personnel and equipment access tunnels l and the drywell equipment and piping support structure. These internal structures are steel fabrications.
l Penetrations through the containment pressure boundary include; the dnwell head closure, equipment hatches to both upper and lower dowell regions, l personnel locks into upper and lower dnwells, a combined personnel access and equipment hatch into the wetwell and piping and electrical penetrations. These pressure boundary appurtenances are steel structures classified as ASME Code l Section III, Division 1, Class MC.
l The containment design pressure is 3.16 kg/cm2 g.The design temperatures for the dnwell and the wetwell are 171 C and 104 'C respectively. The maximum calculated pressures and temperatures for the design basis accident are less than these design conditions. The primary containment pressure boundary including penetrations and isolation valves, has a leak rate equal to or less than 0.5% per day (excluding main steamline isolation valves (MSIV) leakage) of the containment gas mass at the maximum calculated containment pressure for the
,O V design basis accident.
6/18/93 -1 2.14 A
ABWR oesign occum:nt The reinforced concrete diaphragm floor separating the upper dnwell and the wetwell gas spaces, has a steel linear plate on the underside. The design differential pressure of the diaphragm floor between dnwell and wetwell is 1.76 kg/cm 2in the downward direction.
The RPV pedestal forms the lower dnwell region and consists of a cylindrical composite steel structure. It is anchored to'the basemat and supports the RPV ,
through a support ring girder.The pedestal also supports the reactor shield wall.
The pedestal consists of two concentric steel cylindersjoined together radially by vertical steel diaphragms and filled with concrete. The pressure suppression venting paths are an integral part of the pedestal structure. This includes: a) the ducts which interconnect the lower and upper drywell regions, b) the vertical downcomers from the interconnecting ducts to the horizontal vents, and c) the horizontal vents that direct steam into the suppression pool.
Vacuum relief between the dnwell volumes and the wetwell gas space is provided i l
by vacuum breaker valves on piping sleeves penetrating the pedestal wall. Eight normally closed swing check valves with nominal diameter of 500 mm are ;
provided.
The water volume in the suppression pool including the vents is equal to or - ;
greater than 3,580 m3 . The hori7ontal center line of the SRVDL quencher arms l are located at or below the elevation of the center row of horizontal vents in the suppression pool. 1 Water return paths connect the region within the pedestal to the vertical downcomers and horizontal vent paths. At least one meter of corium protection ,
fill concrete is provided on the lower drywell floor. Thermally activated flooding valves are also located in this region.
The following PCS components are classified as Seismic Categon I; the reinforced concrete containment structure, the dnwell head, equipment hatches to both upper and lower dowell regions, personnellocks into upper and l lower dnwells, the combined personnel access and equipment hatch into the wetwell, the basemat, the reactor pedestal, and containment isolation valves j together with their penetrations.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.14.1 provides a definition of the inspections, tests, and/or analyses, together with the associated acceptance criteria, which will be undertaken for the Primag Containment System.
i O
6/18/93 -2 2.14.1-l l
ABWR 0: sign Docum:nt I .
4
) ^
Lu STEEL
^ DRyutLL MEAD ^
^ e%
a rs a r%
- - g g g g n
-- 3r Nott 4 vg;# sa
>;; g p FILLCD
, ggr J
~ TERC Tree
[5 3 TOTAL fRYOT**" .
4 J ,
I EIGHT= " " g h "
_\dI [fg "
WETwtLL 70 ACCESS-VA U "" " " NQt "J -
( >
.tS#in:
m ffD IV lj% ,
B _
00 I WW $ f,ygkg ACCtgg at1 M TOTAL ,7w PERSONNCL TWFL hon WATER RETURN PATHS # WATER EO ' EVE' MATCM f.
-Jin5V4H1'N c , ur Tc. T.Rou, ADasFORCt g / sfgk w \ ACTIVATED FL000ER g twenty "t="
COc "lau 1A STRucTune coa'u" q;!c,t'oa ru ,
c n
'!S TOTAL
".;".Sw f $3r TE S
& .ASc T $
AREASI NOTES:
A. UPPER ORYWELL t CORIUW PROTECTION FLL DEPTM fB 3 9 m S. stTwtLL GAS SPACE 3. RPV PEDESTAL IN TME LowCR ORYWELL REGeoes MAS 31.Som N TO T AL DowN COWERSI
- 0. Lom MYwta 3. PER$0sett AND EQUIPesENT R C ERCNTIAL LOCAfloNS Hif'o*f RcrtRcmCc 0 v
,*Ci.'Sw'CPa%""Rt "U Figure 2.14.1 Primsry Containment System
-3 ' 2.14.1 6/18/93 a
rs n o b b 3 Table 2.14.1 Primary Containment System 8"
Inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The basic configuration of the PCS is as 1. Inspections of the as-built system will be 1. ine as-built PCS conforms with the basic shown on Figure 2.14.1. conducted. configuration shown on Figure 2.14.1.
- 2. The ASME Code pressure boundary 2. A structural integrity test (SIT) will be 2. The results of the SIT of the pressure components of the PCS retain their conducted on the pressure boundary boundary components conform with the integrity under internal pressures that will components of the PCS per ASME Code requiremonts of the ASME Code.
be experienced during service. requirements.
- 3. The maximum calculated pres.sures and 3. Analyses of the design basis accident will 3. The maximum calculated pressures and temperatures for the design basis accident be performed using as-built PCS data. temperatures are less than design are less than design conditions. conditions
- 4. The primary containment pressure 4. An integrated leak rate test of the as-built 4. The primary containment pressure boundary including penetrations and primary containment will be conducted. boundary including penetrations and isolation valves has a leak rate equal to or isolation valves has a leak rate equal to or less than 0.5% per day (excluding MSIV less than 0.5% per day (excluding MSIV
, leakage) of the containment gas mass at leakage) of the containment gas mass at t the maximum calculated containment the maximum calculated containment pressure for the design basis accident. pressure for the design basis accident.
- 5. The design differential pressure of the 5. An SIT will be conducted of the as-built 5. An SIT report exists concluding that the as-diaphragm floor between the drywell and diaphragm floor with the drywell pressure built diaphragm floor is able to withstand wetwell is 1.76 kg/cm2 in the downward greater than wetwell pressures by 1.15 the design differential pressure.
direction. times the design differential pressure
- 6. The water volume in the suppression pool 6. Analyses of the as-built PCS will be 6. The water volume in the suppression pool including the vents is equal to or greater performed. including the vents is equal to or greater than 3580 m3. than 3580 m3.
7 The horizontal center line of the SRVDL 7. Inspections of the as-built SRVDL 7. The horizontal center line of the SRVDL quencher arms are located at or below the quenchers will be conducted. quencher arms are located at or below the elevation of the center row of horizontal elevation of the center row of horizontal vents in the suppression pool. vents in the suppression pool.
N
ABWR D: sign Document 2.14.9 Suppression Pool Temperature Monitoring System O
Q Design Description The Suppression Pool Temperature Monitoring (SPTM) System monitors the suppression pool water temperature and provides signals for initiation of automatic scram on high suppression pool temperature. Figure 2.14.9 shows the SPTM System controlinterfaces. 1 1
i The SPTM System is classified as a Class 1E safety-related system and consists of i
four Class 1E divisions (Division I, II, III, and IV) of temperature sensors and their respective logic processors. l The SPTM System temperature sensors are located in the suppression pool.
There are four divisions of temperature sensors in each quadrant of the suppression pool.
In each SPTM System division, the average suppression pool temperature is calculated by the logic processors using output signals from the temperature :
sensors. In each SPTM System division, a suppression pool temperature trip signal is generated by the logic processor and sent to the Reactor Protection l
System (RPS) when its respective divisional average temperature signal exceeds .
the high average suppression pool temperature set point.
Each of the four SPTM System divisional logics is powered from its respective l divisional Class IE power supply. Independence is provided between Class IE divisions, and also between Class 1E divisions and non-Class IE equipment.
The SPTM System temperature sensors are located in the suppression pool; the SPTM System logic processors are located in the Control Building.
The SPTM System has parameter displays for suppression pool temperatures in the main control room (MCR). ]
l The SPTM System provides Division I and II suppression pool temperature j displays to the Remote Shutdown System (RSS).
Inspections, Tests, Analyses and Acceptance Criteria l
Table 2.14.9 provides a definition of the inspections, tests, and/or analyses, l together with associated acceptance criteria, which will be undertaken for the SPTM System.
p)
\.
6/16/93 . 2.14.9 1
I I
O O O u s L g w
LOCAL AREA PLANT SENSORS OUTPUT SIGNALS SPTM SYSTEM LOGIC PROCESSORS
- Suppression Pool "
Suppresse Poot Average Temperatures ? RPS Average Temperature
, T@ 2 SPTM Suppression Pool Temperature M EMS Q SSW - sensor eypass
, , g ,,,
t ;
processing ewipment -Calibration Self-Diagnosis 9
i sst_c togic7----- ,l
. g Processing I g for Other l g Safety Systems g L
NOTES:
- 1. Diagram represents one of four divisions.
^
N
[ Figure 2.14.9 Suppression Pool Temperature Monitoring System Control Interface Diagram
O O Tchla 2.14.9 Supprmi:n Paci Temparcturo Manitoring System O
j ;
la ,
- Inspections, Tests, Analyses and Acceptance Criteria .
Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The basic configuration of the SPTM 1. Inspections of the as-built system will be 1. The as-built SPTM System conforms with System is described in Section 2.14.9. conducted. the basic configuration described in Section 2.14.9.
- 2. In each SPTM System division, the average 2. Tests will be conducted in each division of 2. In each SPTM System division, a suppression pool temperature is calculated the SPTM System using simulated suppression pool temperature trip signal is by the logic processors using output temperature sensor signals. generated by the logic processor and sent ,
signals from the temperature sensors. In to the Reactor Protection System (RPS) each SPTM System division, a suppression when its respective divisional average i
. pool temperature trip signalis generated temperature signal exceeds the high by the logic processor and sent to the average suppression pool temperature set l Reactor Protection System (RPS) when its point.
respective divisional average temperature .
signal exceeds the high average ,
suppression pool temperature set point. !
43 3. In the SPTM System, independence is 3a. Tests will be performed on the SPTM 3a. A test signal exists only in the Class 1E ,
provided between Class 1E divisions, and . System by providing a test signal in only division under test in the SPTM System.
between Class 1E divisions and non-Class one Class 1E division at a time.
1E equipment. . .
3b. Inspections of the as-built Class 1E 3b. In the SPTM System, physical separation divisions in the SPTM System will be exists between Class 1E divisions. Physical performed. separation exists between these Class 1E r divisions and non-Class 1E equipment.
- 4. MCR displays provided for the SPTM 4. Inspections will be conducted on the MCR 4. Displays exist or can be retrieved in the System are as defined in Section 2.14.9. displays for the SPTM System. . MCR as defined in Section 2.14.9.
- 5. RSS displays provided for the SPTM - 5. Inspections will be conducted on the RSS 7. Displays exist on the RSS as defined in . !
System are as defined in Section 2.14.9. displays for the SPTM System. Section 2.14.9. -
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l l ABWR D: sign Document 2.15.5 Heating, Ventilating and Air Conditioning Systems
% Design Description Control Room Habitability Area HVAC System The Control Room Habitaleility Area (CRHA) Heating, Ventilating and Air Conditioning (HVAC) System provides a controlled environment for personnel comfort and safety, and for the operation of equipment in the main control area envelope (MCAE). The _ system consists of two (redundant) divisions. Each division consists of an air conditioning unit with two supply fans, two exhaust fans, and an emergency filtration unit with two circulating fans.
Toxic gas monitors may be required in the outside air intakes of the CRHA HVAC System; these sencors are not in the Certified Design. ;
f Figure 2.15.5a shows the basic configuration and scope for the CRHA HVAC l System.
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The CRHA HVAC System is classified as safety-related. ,
The CRHA HVAC System operates in the following modes:
(1) Normal operating.
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(2) High radiation.
(3) Outside smoke.
(4) Smoke removal. ,
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[ Normal Operating Mode l In the normal operating mode, one air conditioning unit, one supply fan, and one exhaust fan operate in each division. The exhaust fan automatically starts when the supply fan is started.
The MCAE is maintained at a minimum pressure of 3.2 mm water gauge l
above the outside atmosphere.
High Radiation Mode On receipt of a Process Radiation Monitoring (PRM) System signal for high radiation in the outside air intake of the operating division, the normal outside air intake dampers close, the exhaust air dampers close, the exhaust fan stops, the minimum outside air intake dampers open, and one fan of the emergency filtration unit starts.
In the high radiation mode, a positive pressure of at least 3.2 mm water gauge is maintained in the MCAE relative to the outside atmosphere. Each .
6/18/93 -1 2.15.5
ABWR D: sign Docum:;nt emergency filtration unit treats a mixture of MCAE recirculated air and outside makeup air to maintain the positive pressure with not more than 360 (J m3per hour (@ 760 mm Hg, O C) of outside air.
The redundant division of the CRHA HVAC System starts on a low flow signal from the operating emergency filtration unit. The redundant division is connected to an outside air intake which is separated from the other by a minimum of 50m.
4 Outside Smoke Mode When smoke detection sensors in the operating outside air intake detect smoke, a signal will initiate MCAE air recirculation by isolating the outside air intake, closing the exhaust damper and stopping the exhaust fan.
Smoke Removal Mode The smoke removal mode is manually initiated by closing the recirculation damper, stopping the exhaust fan, and opening the exhaust fan bypass damper to allow outside air purging of the MCAE.
The remaining discussion in this section is not mode-specific and applies (unless stated otherwise) to the entire CRHA HVAC System.
O, q' MCAE temperature is maintained between 21 C and 26 C, with a relative humidity between 10% and 60% except when in the smoke removal mode.
The CRHA HVAC System is classified as Seismic Category I. The CRHA HVAC System is located in the Control Building.
Each of the two CRHA HVAC System divisions is powered from the respective Class 1E division as shown on Figure 2.15.5a. In the CRHA HVAC System, independence is provided between Class 1E divisions, and also between the Class 1E divisions, and non-Class 1E equipment.
Each mechanical division of the CRHA HVAC System (Divisions B and C) is physically separated from the other division, except for the common ducts in the MCAE.
The CRHA HVAC System has the following displays and controls in the main control room:
(1) Controls and status indication for the active safety-related components shown on Figure 2.15.5a.
n (2) Parameter displays for the instruments shown on Figure 2.15.5a, except
() for the smoke detectors.
6/18/93 2.15.5
ABWR Design Document Interface Requirements
\. Toxic gas monitors will be located in the outside air intakes of the CRHA HVAC System, if the site is adjacent to toxic gas sources with the potential for releases of significance to plant operadng personnel in the MCAE. These monitors should have the following requirements:
(1) Be located in the outside air intakes of each division'of the CRHA HVAC System.
(2) Be capable of detecting gas concentrations at which personnel ]
protective actions must be initiated. )
Control Building Safety Related Equipment Area HVAC System The Control Building Safety-Related Equipment Area (CBSREA)HVAC System provides a controlled temperature environment for the operadon of equipment j in the Control Building, excluding the MCAE. The system also limits hydrogen l concentration in the battery rooms.The CBSREA HVAC System consists of three i independent safety-related divisions, each serving a designated area. Each division consists of an air condidoning unit with two supply fans, and two exhaust fans.
The CBSREA HVAC System also vendlates rooms that contain non-safety-related O equipment and provides supplemental cooling in these rooms using non-safety-related fan coil units (FCUs).
The basic system configuration and scope for the CBSREA HVAC System is shown on Figures 2.15.5b,2.15.5c and 2.15.5d.
The CBSREA HVAC System is classified as safety-related except for the FCUs. !
The CBSREA HVAC System operates in the following modes:
(1) Normal operating mode including accident conditions.
(2) Smoke removal mode.
Normal Operating Mode In the normal operadng mode, one air conditioning unit, one supply fan, l
and one exhaust fan of each division operate. The exhaust fan automatically l starts when the supply fan is started.
In the areas served by the CBSREA HVAC System, the temperature is maintained below 40 C.
- p. Hydrogen concentration is maintained at less than 2% by volume in the
(
- - battery rooms.
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l ABWR Disign Docum2nt l
l Smoke Removal Mode The smoke removal mode is manually initiated by closing the recirculation damper, stopping the exhaust fan, and opening the exhaust fan bypass damper to allow outside air purging of the affected Control Building area.
The normal operating mode is used to remove smoke from the battery l rooms.
The remaining discussion in this section is not mode-specific and applies (unless
! stated otherwise) to the entire CBSREA HVAC System. l J
The CBSREA HVAC System is classified as Seismic Category 1, except for the non-safety-related fan coil units. The CBSREA HVAC System is located in the Control Building.
Each of the three CBSREA HVAC System divisions is powered from the respective Class 1E division as shown on Figures 2.15.5b,2.15.5c and 2.15.5d. In the CBSREA HVAC System, independence is provided between Class IE divisions, and also between the Class 1E divisions and non-Class IE equipment.
Each mechanical division of the CBSREA HVAC System (Divisions A, B and C) is physically separated from the other divisions. CBSREA HVAC System Division B duct penetrations of Division IV firewalls are provided with fire dampers. ;
\ The CBSREA HVAC System has the following displays and controls in the main i
control room l ,
(1) Controls and status indication for the active safety-related componerits l l shown on Figures 2.15.5b,2.15.5c and 2.15.5d.
(2) Parameter displays for the instruments shown on Figures 2.15.5b, 2.15.5c and 2.15.5d.
. Reactor Building HVAC System l The Reactor Building (R/B) HVAC System provides a controlled emironment for the operation of equipment in the Reactor Building.
! The Reactor Building HVAC System consists of three independent safety-related divisions. Each division is composed of the following systems:
(1) R/B safety-related equipment HVAC system.
(2) R/B safety-related electrical equipment HVAC system.
(3) R/B safety-related diesel generator HVAC system. <
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l l _ _ - _ _ _ - , - ~.. .-
ABWR Design Docum:nt l
l n The Reactor Building HVAC System includes the following non-safety-related
( ) systems:
(1) R/B secondary containment HVAC system.
(2) R/B containment purge supply / exhaust system.
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l (3) R/B main steam tunnel HVAC system.
R/B Safety-related Equipment HVAC System The R/B safety-related equipment HVAC system provides cooling of safety- l related equipment areas, and consists ofindependent fan coil units. Figure 2.15.5e shows the basic system configuration and scope.
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The R/B safety-related equipment HVAC system is classified as safety-related.
The Residual Heat Removal (RHR) System, High Pressure Core Flooder (HPCF) System and Reactor Core Isolation Cooling (RCIC) System pump ,
room FCUs are automatically initiated upon start-up of their respective room process pump. The safety-related FCUs shown on Figure 2.15.5e are i automatically initiated upon isolation of the Reactor Building secondary containment HVAC system.The Flarnmability Control System (FCS) room l(7 FCUs are also initiated upon a remote manual FCS start signal.
lV The temperature in the safety-related equipment areas is maintained below 40 C, except for the RHR, HPCF, and RCIC pump rooms which are maintained below 66 C during pump operation.
The R/B safety-related equipment HVAC system is classified as Seismic Category I. The R/B safety-related equipment HVAC system is located in the Reactor Building.
Each of the three divisions of the R/B safety-related equipment HVAC system is powered from the respective Class IE division as shown on Figure 2.15.5e.
In the R/B safety-related equipment HVAC system, independence is provided between Class 1E divisions, and also between the Class IE divisions and non-Class lE equipment.
1 Each mechanical division (Divisions A, B, C) of the R/B safety-related equipment HVAC system is physically separated from the other divisions. .
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O 6/18/93 2.15.5
ABWR Design Docum:nt The R/B safety-related equipment HVAC system has the following displays and controls in the main control room:
(1) Controls and status indication for the FCUs shown on Figure 2.15.5e.
The safety-related electrical equipment shown on Figure 2.15.5e located in the Reactor Building is qualified for a harsh environment.
R/B Safety-Related Electrical Equipment HVAC System The R/B safety-related electrical equipment HVAC system provides cooling cf safety-related electrical equipment areas, and consists of three independent divisions. Each division consists of an ' air conditioning unit with -
two supply fans, and two exhaust fans. Figures 2.15.5f,2.15.5g, and 2.15.5h show the basic system configuration and scope.
The R/B safety-related electrical equipment HVAC system is classified as safety-related.
Normal Ooeratina Mode In the normal operating mode, the air conditioning unit, one supply fan, and one exhaust fan of each division operate. The exhaust fan automatically starts when the supply fan is started.
b V In the areas served by the R/B safety-related electrical equipment HVAC.
system temperature is maintained below 40 C, exceptin the dieselgenerator (DG) engine rooms during DG operation. l l
Smoke Removal Mode The smoke removal mode is manually initiated by closing the recirculation damper, stopping the exhaust fan, and opening the exhaust fan bypass damper to allow outside air purging of the affected area. The normal operating mode is used to remove smoke from the DG day tank rooms. )
The R/B safety-related electrical equipment HVAC system is classified as Seismic Category I. The safety-related electrical equipment HVAC system is located in the Reactor Building.
Each of the three divisions of the R/B safety-related electrical equipment HVAC system is powered from the respective Class 1E division as shown on figures 2.15.5f,2.15.5g, and 2.15.5h. In the R/B safety-related electrical equipment HVAC system, independence is provided between Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment.
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U 6/18/93 2.15.5
ABWR D: sign Docum nt l
Each mechanical division of the R/B safety-related electrical equipment i
HVAC system (Divisions A, B, C) is physically separated from the other disisions. ,
The R/B safety-related electrical equipment HVAC system has the following displays and controls in the main control rooms:
(1) Controls and status indication for the active safety-related components ;
shown on Figures 2.15.5f,2.15.5g, and 2.15.5h.
(2) Parameter displays for the instruments shown on Figures 2.15.5f, 2.15.5g and 2.15.5h.
R/B Safety Related Diesel Generator HVAC System The R/B safety-related DG HVAC system provides ventilation for the DG rooms when the DGs operate, and consists of three independent divisions.
Each division consists of a filter unit and two supply fans. Figure 2.15.51 shows the basic system configuration and scope.
t The R/B safety-related DG HVAC system is classified as safety-related.
On receipt of a DG start signal, at least one DG supply fan starts. When the ,
DG is operating, the R/B safety-related DG HVAC system and the R/B safety- l related electrical equipment HVAC system maintain the temperature below 45 C.
The R/B safety-related DG HVAC system is classified as Seismic Category I.
The R/B safety-related DG HVAC system is located in the Reactor Building.
l l Each of the three divisions of the R/B safety-related DG HVAC system is -
i powered from the respective Class 1E division as shown on Figure 2.15.5i. In l the R/B safety-related DG HVAC system, independence is provided between l Class IE divisions, and also between the Class 1E divisions and non-Class 1E equipment.
Each mechanical division of the R/B safety-related DG HVAC system l (Divisions A, B, C) is physically separated from the other divisions.
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lO 6/18/93 -7 2.15.5 l
._. __ - , , _ _. _ ~ _ . . - . _
4 l ABWR D: sign Docunnat The R/B safety-related DG HVAC system has the follmving displays and controls in the main control room:
l l (1) Controls and status indication for the active safety-related components
! shown on Figure 2.15.5i.
R/B Secondary Containment HVAC System The R/B secondary containment HVAC system provides heating and cooling for the secondary containment. Figure 2.15.5j shows the basic system configuration and scope.
Except for the seconda y containment isolation dampers, the R/B secondary containment HVAC system is classified as non-safety-related.
Normal Operatina Mode In the normal operating mode, two supply fans and two exhaust fans operate.
l The supply fans operate only when the exhaust fans are operating.
The R/B secondary containment HVAC system maintains a negative pressure !
in the secondary containment relative to the outside atmosphere.
The R/B secondary containment HVAC system isolation dampers are closed upon receipt of an isolation signal from the Leak Detection System (LDS)' or a signalindicating loss of secondary containment exhaust fans.
Smoke Removal Mode The smoke removal mode is manually initiated by starting the standby exhaust and supply fans, opening the exhaust filter unit bypass dampers,'and partially closing exhaust dampers for divisions not affected by fire.
l l The R/B secondary containment HVAC system penetrations of secondary containment and isolation dampers are classified as Seismic Category I. The R/B secondary containment HVAC system is located in the Reactor Building, except for some of the R/B secondary containment HVAC supply and exhaust air components which are located in the Turbine Building.
l Each R/B secondary containment HVAC system isolation damper requiring electrical power is powered from the Class IE division, as shown on Figure 2.15.5j. In the R/B secondary containment HVAC system, independence is provided between Class 1E divisions, and also between Class 1E divisions and non-Class 1E equipment.
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6/18/93 2.15.5 .
i
ABWR ossign Docum:nt I. The R/B secondary containment HVAC system has the following displays and controls in the main control room:
(1) Control and status indication for the active components shown on Figure 2.15.5j.
(2) Parameter displays for the instruments shown on Figure 2.15.5j.
l The exhaust duct secondary containment isolation dampers are located in the secondag containment and qualified for a harsh environment.
l The pneumatically-operated secondary containment isolation dampers, l shown on Figure 2.15.5j, fail to the closed position in the event ofloss of l pneumatic pressure or loss of electrical power to the valve actuating l solenoids.
1 R/B Containment Purge Supply / Exhaust System The R/B containment puige supply / exhaust system provides air for containment purging prior to personnel entry to the primaq containment and consists of a purge supply fan, a filter unit, and a purge exhaust fan as shown on Figure 2.15.5j.
/ The R/B containment purge supply / exhaust system is classified as non-
-- safety-related.
R/B Main Steam Tunnel HVAC System The R/B main steam tunnel HVAC system provides cooling to the main steam tunnel and consists of two FCUs. Each FCU has two fans. The FCUs are l started manually.
The R/B main steam tunnel HVAC ' system is classified as non-safety-related.
1 Inspections, Tests, Analyses and Acceptance Criteria . .
For portions of the CRHA HVAC System within the Certified Design, Table i 2.15.5a provides a definition of the inspections, tests, and/or analyses, together I
with associated acceptance criteria, which will be undertaken for the CRHA HVAC Systems.
Table 2.15.5b provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the CBSREA HVAC System.
l Table 2.15.5c provides a definition of the inspections, tests, and/or analyses, l
s together with associated acceptance criteria, which will be undertaken for the Reactor Building safety-related equipment HVAC system.
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6/18/93 -9 2.15.5 l
I
l l ABWR D: sign Docum:nt gs Table 2.15.5d provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the !
Reactor Building safety-related electrical equipment HVAC system.
Table 2.15.5e provides a definition of the inspections, tests, and/or analyses, j together with associated acceptance criteria, which will be undertaken for the !
Reactor Building safety-related DG HVAC system.
l Table 2.15.5f provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Reactor Building secondary containment HVAC system.
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6/18/93 2.15.5 l
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M EMERGENCY \
TD TD
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- HECW i AIR CONDITIONING UNIT SUPPLY FANS '
i NOTES:
L THIS FIGURE SHOWS ONE OF TWO IDENTICAL DIVISIONS ALL ELECTRICAL POWER LOADS FOR THE COMPONENTS OF DIVISION B ARE POWERED
, FROM DIVISION 11. ALL ELECTRICAL POWER LOADS FOR THE COMPONENTS OF DIVISION C ARE POWERED FROM DIVISION l!!.
I to j Figure 2.15.5a Control Room Habitability Area HVAC System y
r---- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -- - - - - - - - - - -
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NOTES:
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.1. ALL CLASS 1E ELECTRICAL LOADS SHOWN EXHAUST FANS BARRIER ARE POWERED FROM DIVISION I.
- 2. FCU COOLING WATER SUPPUED BY THE HNCW SYSTEM.
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$ Figure 2.15.5b Control Building Safety-Related Equipment Area HVAC System (Division A)
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I NOTES: -
- 1. ALL CLASS 1E ELECTRICAL LOADS SHOWN ARE POWERED FROM DIVISION ll. TORNADO MISSILE ,
EXHAUST FANS BARfMER
- 2. DIVISION B DUCT PENETRATIONS OF DIVISION IV FIREWALLS ARE PROVIDED WITH FIRE DAMPERS.
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.$ Figure 2.15.Sc Control Building Safety-Related Equipment Area HVAC System (Division B) l
._ _ _____m_____.__ _ _ __ _ ___ _ _ - _ _ _ _ .. . - . , - . . . . , . - , 4 , _ _ . _ _ _ _ _ _ _ _ _ . _ _ . _ . . _ _ _ _ . . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _
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~ BATTERY DIY 811 NOTES: ^
- 1. ALL CLASS 1E ELECTRICAL LOADS SHOWN ARE POWERED FROM DIVISION 111.
. 2. FCU COOLING WATER SUPPLIED BY . [ S$^E THE HNCW SYSTEM. EXHAUST FANS BARRIER
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@. Figure 2.15.5d Control Building Safety-Related Equipment Area HVAC System (Division C) 4 .
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DIVISION A DIVISION B DIVISION C RHR-A RHR-B RHR C l FCU l l FCU l l FCU l RCIC-A HPCF-B HPCF-C FCC lFCU l lFCU l CAMS-A CAMS-B FCS-C I FCU l l FCU__j lFCU !
SGTS-B l FCU l h FCS-B lFCU l NOTES:
- 1. FCU COOLING WATER IS SUPPLIED BY THE RCW SYSTEM.
- 2. NORMAL VENTILATION AND SMOKE REMOVALIS PROVIDED BY THE R/B SECONDARY CONTAINMENT HVAC SYSTEM.
- 3. ELECTRICAL POWER LOADS FROM DIVISIONS A, B, AND C ARE POWERED FROM DIVISIONS I,11, AND lit, RESPECTIVELY.
N Figure 2.15.5e Reactor Building Safety-Related Equipment HVAC System
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4 NOTES: ,
, 1. ALL CLASS 1E ELECTRICAL LOADS SHOWN ARE POWERED BY DMSION L j Figure 2.15.5f Reactor Building Safety-Related Electrical Equipment HVAC System (Division A) i
_ . _ _ _ _ _ _ . _ _ . _ _ . _ _ _ _ _ . _ . _ _ . _ _ _ .___________m__ . , .--s -.
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N TORNADO MISSILE BARRIER EXHAUST FANS NOTES:
- 1. ALL CLASS 1E ELECTRICAL LOADS SHOWN
' ARE POWERED BY DIVISION 11.
-p Figure 2.15.5g ' Reactor Building Safety-Related Electrical Equipment HVAC System (Division B)
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r TORNADO MISStLE DARRIER EXHAUST FANS NOTES:
- 1. ALL CLASS 1E ELECTRICAL LOADS SHOWN ARE POWERED BY DIVISION 111.
to
.@ Figure 2.15.5h Reactor Building Safety-Related Electrical Equipment HVAC System (Division C)
O O O TORNADO TORNADO BARRIER \
[)l y 3 EXHAUST / BARRIER
- - DIESEL GENERATOR AIR l -F ROOM lg _
h U TD ~
TD SUPPLY FANS 1
5 l
i' NOTES:
- 1. THIS FIGURE SHOWS ONE OF THREE IDENTICAL DIVISIONS.
ELECTRICAL POWER LOADS FOR DIVISIONS A, B, AND C ARE POWERED FROM DIVISIONS 1,11, AND 111, RESPECTIVELY.
4-sa h Figure 2.15.Si Reactor Building Safety-Related Diesel Generator HVAC System L --- - _ _ _ _ - - _ _ - _ _ _ - . .- .. - . . . . - - .
-._m_m... _ . . . _ _ . _ , - . . -
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NOTES: H I i
- 1. THE OUTBOARD ISOLATION - T,s ns REACTOR BUILDING / SECONDARY CONTAINMENT f DAMPER SOLENOID VALVES l
ARE POWERED BY DIVISION L
.i THE INBOARD ISOLATION '
DAMPER SOLENOlO VALVES g ARE POWERED BY DIVISION IL l I
h Figure 2.15.5j Reactor Building Secondary Containment HVAC System
__ _ _ . __ - - - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ - . . _ _ _ . _ , - _ _ _ _ . . . . . ~ - . . - . _ _ _ _ __
L O O 3 Table 2.15.5a Control Room Habitability Area HVAC System
.E "
Inspections, Tests, Analyses and Acceptance Criteria Design Commitment Inspections Tests, Analyses Acceptance Criteria
- 1. The basic configuration of the CRHA HVAC 1. Inspections of the as-built system will be 1. The as-built CRHA HVAC System conforms System is as shown on Figure 2.15.5a. conducted. with the basic configuration shown on Figure 2.15.5a.
- 2. The exhaust fan automatically starts when 2. Tests will be conducted on each division of 2. The exhaust fan automatically starts when
. the supply fan is started. the CRHA HVAC System by starting the the supply fan is started.
supply fan.
- 3. The MCAE is maintained at a minimum 3. Tests will be conducted on the as-built 3. The MCAE is maintained at a minimum pressure of 3.2 mm water gauge above the CRHA HVAC System in the normal mode of pressure of 3.2 mm water gauge above the outside atmosphere. operation. outside atmosphere.
4a. On receipt of a PRM System signal for high 4a. Tests will be conducted on each CRHA 4a. Upon receipt of a simulated initiation radiation in the outside air intake of the HVAC System division using a simulated signal the following occurs:
operating division, the normal outside air initiation signal. a) Normal outside air intake dampers are intake dampers close, the exhaust air closed.
dampers close, the exhaust fan stops, the b) Exhaust air dampers are closed.
4 c) Exhaust fan is stoppad.
f minimum outside air intake dampers open, and one fan of the emergency filtration unit .d) Minimum outside air intake dampers starts. are opened.
e) Emergency filtration unit fan is started.
4b. In the high radiation mode, pod'% 4b. Tests will be conducted on each division of 4b.The MCAE is maintained at a positive pressure of at least 3.2 mm water gauge is the as-built CRHA HVAC System in the high pressure of at least 3.2 mm water gauge maintained in the MCAE relative to the radiation mode. relative to the outside atmosphere with outside atmosphere. Each emergency outside makeup air of not more than filtration unit treats a mixture of MCAE 360 m3per hour (@ 760 mm Hg,0 'C).
recirculated air and outside makeup air to maintain the positive pressure with not more than 360 m3per hour (@ 760 mm Hg, 0*C) of outside air.
4c. The redundant division of the CRHA HVAC 4c. Tests will be conducted on each division of 4c. The redundant division of the CRHA HVAC System starts on a low flow signal from the the as-built CRHA HVAC System using System starts on a low flow signal from the operating emergency filtration unit. simulated low flow signals. operating emergency filtration unit.
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h 3 Tcbla 2.15.5m Contrcl R:om Habit:bility Area HVAC Sy:t:m (Continund) w inspections Tests Analyses and Acceptance Criteria ,
Design Commitment inspections. Tests, A:.alyses Acceptance Criteria 4d. The redundant division of the CRHA HVAC 4d. Inspections will be conActed on the CRHA 4d. The CRHA HVAC System outside air intakes System is connected to an outside air HVAC System. are at least 50m apart.
intake which is separated from the other by a minimum of 50m.
- 5. When smoke detection sensors in the - 5. Tests will be conducted on each CRHA 5. Upon receipt of a simulated initiation operating outside air intake detects smoke, HVAC System division using a simulated signal the following occurs:
a signal will initiate MCAE air recirculation smoke signal. a) Outside air intake dampers are closed.
by isolating the outside air intake, closing b) Exhaust air dampers are closed.
the exhaust damper, and stopping the c) Exhaust fan is stopped.
exhaust fan.
- 6. In the CRHA HVAC System, independence 6a. Tests will be performed on the CRHA HVAC 6a. The test signal exists only in the Class 1E '
is provided between Class 1E divisions, System by providing a test signal in only division under test in the CRHA HVAC and between Class 1E divisions and non. one Class 1E division at a time. System.
Class 1E equipment.
6b. Inspection of the as-built Class 1E divisions sb. In the CRHA HVAC System, physical 4 separation exists between Class 1E Y in the CRHA HVAC System will be performed. divisions. Physical separation exists '
between these Class 1E divisions and non-Class equipment.
- 7. Each mechanicai division of the CRHA 7. Inspections of the as-built CRHA HVAC 7. Each mechanical division of the CRHA HVAC System (Division 8 and C) is System will be performed. HVAC System is physically separated from physically separated from the other the other mechanical division of the CRHA division, except for the common ducts in HVAC System by structural and/or fire the MCAE. barriers.
- 8. Main control room displays and controls 8. Inspections will be performed on the main 8. Displays and controls ~ exist or can be provided for CRHA HVAC System are as control room displays and controls for the retrieved in the main control room as
. defined in Section 2.15.5. CRHA HVAC System. defined in Section 2.15.5.
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3 Table 2.15.5b Control Building Safety-Related Equipment Area HVAC System E !
Inspections, Tests, Analyses and Acceptance Criteria s Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The basic configuration of the CBSREA 1. Inspections of the as-built system will be 1. The as built CBSREA HVAC System HVAC System is as shown on Figures conducted. conforms with the basic configuration !
2.15.5b,2.15.5c and 2.15.5d. shown on Figures 2.15.5b,2.15.5c and 2.15.5d.
- 2. The exhaust fan automatically starts when 2. Tests will be conducted on each division of 2. The exhaust fan automatically starts when the supply fan is started. the as-built CBSREA HVAC System by the supply fan is started.
starting the supply fan.
- 3. Hydrogen concentration is maintained at 3. Flow tests will be conducted on each 3. Hydrogen concentration is maintained at k less than 2% by volume in the battery battery room served by the CBSREA HVAC less than 2% by volume in the battery ?
rooms. System. Hydrogen concentration analyses rooms. i i will be performed for each battery room l
- using measured flow rates and maximum expected battery hydrogen evolution rates.
- 4. In the CBSREA HVAC System, 4a. Tests will be performed on the CBSREA 4a. The test signal exists only in the Class 1E .
4 HVAC System by providing a test signal in division under test in the CBSREA HVAC t Y independence is provided between Class '
1E divisions, and between Class 1E: only one Class 1E division at a time. System.
divisions and non-Class 1E equipment, j 4b. Inspection of the as-built Class 1E divisions 4b. In the CBSREA HVAC System, physical in the CBSREA HVAC System will be separation exists between class 1E performed. divisions. Physical separation exists '
between these Class 1E divisions and non-Class 1E equipment.
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- 5. Each mechanical division of the CBSREA 5. Inspections of the as-built CBSREA HVAC 5. Each mechanical division of the CBSREA HVAC System (Divisions A, B and C)is System will be conducted. HVAC System is physically separated from l
! physically separated from the other the other mechanical divisions of the divisions. CBSREA HVAC System by structural and/or fire barriers.
- 6. Main control room displays and controls 6. Inspections will be performed on the main 6 Displays and controls exist or can be
! provided for CBSREA HVAC System are as control room displays and controls for the retrieved in the main control room as ;
defined in Section 2.15.5. CBSREA HVAC System. defined in Section 2.15.5.
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j Table 2.15.5c Reactor Building Safety-Related Equipment HVAC System is Inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The basic configuration of the R/B safety- 1. Inspections of the as-built system will be 1. The as-built R/B safety-related equipment related equipment HVAC system is as conducted. HVAC system conforms with the basic shown on Figure 2.15.5e. configuration as shown on Figure 2.15.5e.
- 2. The RHR, HPCF, and RCIC pump room 2. Tests will be conducted on each pump 2. Each pump room FCU starts when a signal FCUs are automatically initiated upon start- room FCU using simulated signals indicates start-up of their respective room up of their respective room process indicating pump start-up. process pumo.
pumps.
- 3. The safety-related FCUs shown on Figure 3. Tests will be conducted on each as-built 3. The safety-related FCUs shown on Figure 2.15.5e are automatically initiated upon safety-related FCUs using simulated 2.15.5e are automatically initiated upon isolation of the R/B secondary containment signals indicative isolation of the R/B . isolation of the R/B secondary containment HVAC system. secondary containment HVAC system. HVAC system.
- 4. The FCS room FCUs are initiated upon a 4. Tests will be conducted on each as-built - 4. The FCS room FCU starts upon receipt of a ,
FCS start signal. FCS room FCU using a simulated initiation signal indicating FCS start.
4 signal.
- 5. In the R/B safety-related equipment HVAC Sa. Tests will be performed on the R/B safety- Sa. The test signal exists only in the C! ass 1E system, independence is provide d between related equipment HVAC system by division under test in the in the R/B safety-Class 1E divisions, and between Claes 1E providing a test signal in only one Class 1E related equipment HVAC system. '
divisions and non-Class 1E equipment. division at a time.
Sb. Inspection of the as-built Class 1E divisions 5b. In the R/B safety-related equipment HVAC [
in the R/B safety-related equipment HVAC system, physical separation exists between system will be performed. Class 1E divisions. Physical separation exists between these Class 1E divisions and non-class 1E equipment.
- 6. Each mechanical division (Divisions A, B, 6. Inspections of the as-built R/B safety- 6. Each mechanical division of the R/B safety- 1
- C) of the R/B safety-related equipment related equipment HVAC system will be related equipment HVAC system is HVAC system is physically separated from conducted. physically separated from the other the other divisions. mechanical divisions of the R/B safety-
' related equipment HVAC system by structural and/or fire barriers.
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O O O i Table 2.15.5c Reactor Cuilding Safety-R; lated Equipment HVAC System (Ccntinu d)
" Inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 7. Main control room displays and controls 7. Inspections will be performed on the main 7. Displays and controls exist or can be provided for the R/B safety-related control room displays and controls for the retrieved in the main control room as equipment HVAC system are as defined in R/B safety-related equipment HVAC defined in Section 2.15.5.
Section 2.15.5. system.
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O O a j Table 2.15.5d Reactor Building Safety-Related Electrical Equipment HVAC System is Inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria
- 1. The basic configuration of the R/B safety- 1. Inspections of the as-built system will be 1. The as-built R/B safety-related electrical related electrical equipment HVAC system conducted. equipment HVAC system conforms with is as shown on Figures 2.15.5f,2.15.5g, and the basic configuration shown on Figures 2.15.5h. 2.15.5f,2.15.5g, and 2.15.Sh.
- 2. The exhaust fan automatically starts when 2. Tests will be conducted on each division of 2. The exhaust fan automatically starts when the supply fan is started. the as-built R/B safety-related electrical the supply fan is started.
equipment HVAC system by starting the supply fan.
- 3. In the R/B safety-related electrical 3a. Tests will be performed on the R/B safety- 3a. The test signal exists only in the Class 1E equipment HVAC system, independence is related electrical equipment HVAC system division under test in the safety-related provided between Class 1E divisions, and by providing a test signalin only one Class electrical equipment HVAC system.
between Class 1E divisions and non-Class 1E division at a time.
1E equipment.
3b. Inspection of the as-built Class 1E divisions 3b. In the R/B safety-related electrical eb . in the R/B safety-related electrical equipment HVAC system, physical P equipment HVAC system will be separation exists between Class 1E performed- divisions. Physical separation exists between these Class 1E divisions and non-Class 1E equipment.
- 4. Each mechanical division of the R/B safety- 4. Inspections of the as-built R/B safety- 4. Each mechanical division of the R/B safety-related electrical equipment HVAC system related electrical equipment HVAC system related electrical equipment HVAC system (Divisions A, B, and C) is physically will be conducted. is physically separated from the other separated from the other divisions. mechanical divisions of the R/B safety-related electrical equipment HVAC system by structural and/or fire barriers.
- 5. Main control room displays and controls 5. Inspections will be performed on the main 5. Displays and controls exist or can be provided for R/B safety-related electrical control room displays and controls for the retrieved in the main control room as equipment HVAC system are as defined in R/B safety-related electrical equipment defined in Section 2.15.5.
Section 2.15.5. HVAC system.
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O j Table 2.15.5e Reactor Building Safety-Related Diesel Generator HVAC System
. 4 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria
. 1. The basic configuration of the R/B safety- 1. Inspections of the as-built system will be 1. The as-built R/B safety-related DG HVAC 1 related DG HVAC system is as shown on conducted. system conforms with the basic Figure 2.15.5i. configuration shown on Figure 2.15.5i.
- 2. On receipt of a DG start signal, at least one 2. Tests will be conducted on each division of 2. On receipt of a DG start signal, at least one DG supply fan starts. the as-built R/B safety-related DG HVAC DG supply fan starts. ;
system using a simulated DG start signal.
- 3. In the R/B safety-related DG HVAC system, 3a. Tests will be performed on the R/B safety- 3a. The test signal exists only in the Class 1E independence is provided between Class related DG HVAC system by providing a division under test in the R/B safety-related 1E divisions, and between Class 1E test signalin only one Class 1E division at a DG HVAC system.
divisions and non-Class 1E equipment. time.
3b. Inspection of the as-built Class 1E divisions 3b. In the R/B safety-related DG HVAC system, in the R/B safety-related DG HVAC system physical separation exists between Class ,
will be performed. 1E divisions. Physical separation exists A, between these Class 1E divisions and non-
? Class 1E equipment
. 4. Each mechanical division of the R/B safety . 4. Inspections of the as-built R/B safety- 4. Each mechanical division of the R/B safety-related DG HVAC system (Divisions A, E related DG HVAC system will be related DG HVAC system is physically i
and C)is physically separated from the conducted. separated from the other mechanical .
other divisions. divisions of the R/B safety-related DG HVAC system by structural and/or fire barriers.
- 5. Main control room displays and controls 5. Inspections will be performed on the main 5. Displays and controls exist or can be provided for R/B safety-re ated DG HVAC control room displays and controls for the retrieved in the main control room as -
system are as defined in Section 2.15.5. R/B safety-related DG HVAC system. defined in Section 2.15.5.
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- - _ _ . _ . _ _ _ _ _ - - _ - . - - _ - - - _ _ _ - - _ _ _ _ _ - _ - - - - . _ _ . - _ _ _ _ - - - _ _ - . _ ~ _ _ _ - - - _ _ - _ _ _
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.3' Table 2.15.5f Reactor Building Secondary Containment HVAC System 3"
Inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptarce Criteria
- 1. The basic configuration of the R/B 1. Inspections of the as-built system will be 1. The as-built R/B secondary containment secondary containment HVAC system is as conducted. HVAC system conforms with the basic shown on Figure 2.15.5j. configuration shown on Figure 2.15.5j.
4
- 2. The R/B secondary containment HVAC 2. Tests will be conducted on the R/B 2. The R/B secondary containment HVAC system maintains a negative pressure in secondary containment HVAC system in system maintains a negative pressure in i the secondary containment relative to the the normal mode of operation. the secondary containment relative to the outside atmosphere. outside atmosphere.
- 3. The R/B secondary containment HVAC 3. Tests will be conducted on the R/B 3. Upon receipt of a simulated signal, system isolation dampers are closed upon secondary containment HVAC system isolation dampers are automatically receipt of an isolation signal from the LDS, using simulated LDS isolation and loss of closed.
or signalindicating loss of secondary secondary containment supply and containment exhaust fans. exhaust fans signals.
- 4. In the R/B secondary containment HVAC 4a. Tests will be performed on the R/d 4a. The test signal exists only in the Class 1E g, system, independence is provided between secondary containment HVAC system by division under test in the R/B secondary P Class 1E divisions, and between Class 1E providing a test signalin only one Class 1E containment HVAC system.
divisions and non-Class 1E equipment. division at a time.
4b. Inspection of the as-built Class 1E divisions 4b. In the R/B secondary containment HVAC in the secondary containment HVAC system, physical separation exists between system will be performed. Class 1E divisions. Physical separation exists between these Class 1E divisions and non-Class 1E equipment.
- 5. Main control room displays and controls 5. Inspections will be performed on the main 5. Displays and controls exist or can be
- provided for the R/B secondary control room displays and controls for the retrieved in the main control room as containment HVAC system are as defined R/B secondary containment HVAC system. defined in Section 2.15.5.
in Section 2.15.5.
- 6. The pneumatically-operated secondary 6. Tests will be conducted on the as-built R/B 6. The secondary containment isolation containment isolation dampers, shown on secondary containment HVAC system dampers shown on Figure 2.15.5j fail to the Figure 2.15.5j, fail to the closed position in pneumatic isolation dampers. closed position on loss of pneumatic the event of loss of pneumatic pressure or pressure or loss of electrical power to the loss of electrical power to the valve valve actuating solenoids. >
actuating solenoids.
to in tn 1
ABWR D: sign Docum:nt 4.3 Potable and Sanitary Water System Interface Requirements The Potable and Sanitary Water (PSW) System provides potable water to all plant buildings and collects liquid wastes and entrained solids and conveys them to a sewage treatment facility. The PSW System is not within the Certified Design. A site specific PSW System that contains no interconnections with systems having the potential for containing radioactive materials will be provided.
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ABWR Design Docum:nt 4.6 Makeup Water Preparation System Interface Requirements The Makeup Water Preparation (MWP) System provides makeup water to the plant via the Makeup Water (purified) (MUWP) System. The MWP System is not within the Certified Design. A site specific MWP System will be designed for any facilitywhich has adopted the Certified Design to provide demineralized water to the MUWP System.
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ABWR D: sign Docum:nt j I
4 .8 Airborne Particulate Radiation Monitoring lO Monitoring of radiation areas in the plant are performed with a variety of instruments to measure both contained (in equipment) and uncontained (external contamination and airborne species) radiation sources. This i
equipment is within the scope of the certified design and is addressed in Sections 2.3.1 and 2.3.2. However, instrumentation to measure the airborne l concentration of particulate radionuclides in personnel access areas is not l within the scope of the current design. Such instrumentation will be provided on i a site-specific basis and shall to meet the interface requirements set forth in Section 3.2b, item 2.
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4.8 6/10/93
ABWR Design Document 4 .9 Heating, Ventilating and Air Conditioning i
O Covered in Section 2.15.5. (Control Room Habitability Area HVAC System),
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ABWR c::ign occummt APPENDIX B ABBREVIATIONS AND ACRONYMS AC Alternating Current CV Control Valve AC Atmospheric Control CVCF Constant Voltage Constant ADS Automatic Depressurization Frequency System CW Circulating Water AFPC Augmented Fuel Pool Cooling ALTM Automated Thermal Limit DC Direct Current Monitor DG Diesel Generator, Emergency APR Automatic Power Regulator DIV Division APRM Average Power Range Monitor D/S Dryer and Separator ARD Anti-Rotation Device DTM Digital Trip Module ARI Altemate Rod Insertion DWC Drywell Cooling ARM Area Radiation Monitoring System ECCS Emergency Core Cooling System ASD Adjustable Speed Drive EDG Emergency Diesel Generator ASME American Society of Mechanical EMS Essential Multiplexing System Code Engineers, Boiler and Pressure EPD Electrical Power Distribution Vessel Code ATWS Anticipated Transient Without FCS Flammability Control System Scram FDWC Feedwater Control FIV Flow-Induced Vibration BLDG Building FMCRD Fine Motion Control Rod Drive FP Fire Protection C&I Control and Instrumentation FPC Fuel Pool Cooling C/B Control Building FPS Fire Protection System CAMS Containment Atmospheric l Monitoring System GL Grade Level
- CBSREA Control Building Safety-Related GSC - Gland Seal Condenser Equipment Area l CF&CAE Condensate, Feedwater and HAZ Heat-Affected Zone
! Condensate Air Extraction .HCU Hydraulic Control Unit l CFDWA Condensate, Feedwater, and Air HCW- - High Conductivity Waste l Extraction HECW HVAC Emergency Cooling CFS Condensate and Feedwater Water System . HEPA High Efficiency Particulate Air CID ControlInterface Diagram HFE Human Factors Engineering CIV Combined Intercept Valve HNCW HVAC Normal Cooling Water CMU Control Room Multiplexing Unit HPCF High Pressure Core Flooder l CPS Condensate Purification System HPIN High Pressure Nitrogen Gas l CRD Control Rod Drive Supply l CRDHS Control Rod Drive Hydraulic HPSH Net Positive Suction Head l System HSI Human-System Interfaces CRGT Control Rod Guide Tube HVAC Heating, Ventilating, and Air l CRHA Control Room Habitability Area Conditioning (r CS CST Containment Spray Condensate Storage Tank HWH HX Hot Water Heating Heat Exchanger CTG Combustion Turbine Generator CUW Reactor Water Cleanup 6-18-93 Appendix B
ABWR D=lgn Docum:nt IA Instmment Air System O
b ICGT In-Core Guide Tube NMS Neutron Monitoring System I&C Instrument and Control NPSH Net Positive Suction Head INST Instrumentation NRHX Non-Regeneration HX ISLOCA Intersystem Loss-of-Coolant NSD Non-Radiation Storm Drain Accident OGS Off-Gas System ISI In-Service Inspection OLU Output Logic Unit ITAAC Inspection, Tests, Analyses, and P/C Power Center Acceptance Criteria ITP InitialTest Program PASS Post-Accident Sampling System PCS Primary Containment System LCP Local Control Panels PIP Plant Investment Protection LCW Low Conductivity Waste PMG Plant Main Generator LD Load Driver PRM Process Radiation Monitoring LDIS Leak Detection and Isolation PS Pipe Space System LDS Leak Detection System R/B Reactor Building LOCA Loss-of-Coolant Accident RAT Reserve Auxiliary Transformer LOOP Loss of Offsite Power RCCV Reinforced Concrete LOPP Loss of Preferred Power Containment Vessel LPFL Low Pressure Core Flooder RCIC Reactor Core Isolation Cooling LPFL Low Pressure Flooding RCIS Rod Control and Information LPMS Loose Parts Monitoring System System LPRM Local Power Range Monitor RCPB Reactor Coolant Pressure
(~3 Boundary
(./ LSPS Lighting and Servicing Power l Supply RCW Reactor Building Cooling Water l
RFC Recirculation Flow Control MCAE Main Control Area Envelope RHR Residual Heat Removal M/C Metal-Clad RHX Regeneration HX MCC Motor Control Center RIP Reactor Internal Pump MCES Main Condenser Evacuation RMU Remote Multiplexing Unit System RPS Reactor Protection System MCPR Main Control Room Panel RPV Reactor Pressure Vessel MCR Main Control Room RRS Reactor Recirculation System MG Motor Generator RSS Remote Shutdown System MOV Motor-Operated Valve RSW Reactor Service Water MPT Mean Power Transformer RW Radwaste MRBM Multi-Channel Rod Block RW/B Radwaste Building Monitor RX Reactor MS Main Steam MSIV Main Steamline Isolation Valve S/B Service Building MSL Main Steam Line S/P Suppression Pool MSV Main Stop Valve SA Service Air MT Main Turbine SAM Sampling System MUWC Makeup Water (Condensate) SB&PC Steam Bypass and Pressure MUWP Makeup Water (Purified) Control (q
j SC Shutdown Cooling NBS Nuclear Boiler System SCRRI Selected Control Rods Run-In NEMS Non-Essential Multiplexing SD Storm Drain 6-18 Appendix B
ABWR oo:rgn occumsnt SDC Shutdown Cooling TBS Turbine Bypass System SGTS Standby Gas Treatment System TBVS Turbine Building Ventilation U SJAE Steam Jet Air Ejector System l SLC Standby Liquid Control TCW Turbine Building Cooling Water S/P Suppression Pool TGS Turbine Gland Seal l SPC Suppression Pool Cooling TLU Trip Logic Unit SPCU Suppression Pool Cleanup TMSL Typical Mean Sea Level SPTM Suppression Pool Temperature Elevation Monitoring TS Transmission System SRNM Startup Range Neutron Monitor TSW Turbine Service Water SRV Safety / Relief Valve SSE Safe Shutdown Earthquake UAT Unit Auxiliary Transfonners SSLC Safety System Logic and Control UHS Ultimate Heat Sink SSPV Scram Solenoid Pilot Valve USE Upper-Shelf Energy j STP Simulated Thermal Power l
T/B Turbine Building TB Turbine Bypass TD Tornado Damper l
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l 6-18-93 -3 Appendix B
i ABWR DESIGN CERTIFICATION O MEMORMOUM ROAD MAPS
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THE RELATIONSHIP BETWEEN PLANT SAFETY ANALYSES AND OTHER SAFETY-RELATED ISSUES -
AND TIER 1 ENTRIES l
O l
1 May 21,1993 GE Nuclear Energy 1
Table 11 O ea ^ ^# iv i-Verifying SSAR Entrv Parameter Ma[yg ITAAC Important Features from Level 1 Internal Events Analyses RCIC System Period of Time RCIC System Able to j Operate Without AC Power (hrs) 8 none Able to Override Switchover of RCIC -
Makeup Water Source from CST to Suppression Pool ---
none .
Period of Time Station Battery Able to Provide RCIC Control Power Without AC Power (hrs) 8 none Combustion Turbine Generator Able to Power Any One of the Three Safety Divsions --
2.12.11 HPCF Logic and Control l
, Number of HPCF Subsystems that can be Operated independent of Essential Mutiplexing System 1 2.2.6 AC Independent Water Addition System Diesel Driven Pump -- 2.15.6 l
- independent Water Supply --
2.15.6-Manually Operable Valves --- 2.15.6 I
O 1
i Table 11 PRA Analysis (Cont.)
Verifying i SSAR Entry Parameter Valug ITAAC Important Features from Level 1 Internal .
Events Analyses (Cont.)-
RCW and RSW Systems-' 1 Number of Parallel Loops in Each Division 'i
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RCW -2' 2.11.3 RSW 2 none (Not in ;
Certified
- Design): ;
A Division of RCW/RSW (With Only :
One Loop in Each System Operating) . '
is Capable of Providing Sufficient -
Cooling to the ECCS Pump O Components in that Division _
[ Miller - Ultimate system capability not Tier 1) none ,
A Division of RCW/RSW (With Both ,
Loops in Each System Operating) l Is Capable of Providing (Through !
l the RHR Heat Exchanger) Sufficient Suppression Pool Cooling- ---- none
[ Miller -- Ultimate system capability not Tier 1]
Design Pressure of Some Low Pressure
- Components Upgraded to 28.8 atg l [Paradiso .4*1025 psig '- Ultimate Rupture Strength.can withstand 1025 psig) .
l (Design Description Only)
! RHR System --
2.4.1 HPCF System --
2.4.2 -
- RCIC System - -
2.4.4 L CRD System - - . 2.2.2 .
l SLC System --
2.2.4 (later) _
l CUW System ? ----
2.6.1. (later) ' 1 l FPC Systern ----
none -
l Nuclear Boiler. System ---
none-l Reactor Recirculation System ---
none l_
MUWC System --
none-MUWP System _ - - --
none Radwaste System (LCW Receiving Tank and HCW Receiving Tank) --
none i
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Table 11 O PRA Analysis (Cont.)
1 Verifying SSAR Entry Parameter Value LIMQ Important Features from Level 1 Internal Events Analyses (Cont.)
l CRD Scram Capability Consists of Both Hydraulic and Electric Run-in Capability ----
2.2.7 Initiation Logic 2/4 2.2.7 ARI ----
2.2.8 Initiation Logic 2/3 2.2.8 On ATWS Signal
- , Automatic SLC Initiation ----
2.2.4 i
l \ Automatic Recirculation Pump Trip ---
2.2.8 Capabilty to Power Condensate Pumps with Emergency Diesel Generator ----
none Independent and Separate ESF Divsions Combination of:
Number 3 2.4.1 2.4.2 Number of High Pressure ECCS 2.4.4 ,
per Divslon 1 2.1.2 !
2.11.3 i Number of Low Pressure ECCS 2.11.9 i per Divsion 1 2.12.1 1 2.12.12 Number of RHR Systems per Divsion 1 2.12.13 )'
i 2.12.14 ABWR has an ADS ---- 2.1.2 Diesel Generator Provides AC Power Backup for Each ESF Division Following ;
Loss of Normal AC Power ----
2.12.13 3
. . . _ _ = - _ - . _ . - - _-
Table 11 -
PRA Analysis (Cont.)
SSAR Entry Parameter verifying i V.ahHt ITAAC j
important Features from Level 1 Internal Events Analyses (Cont.)
i Safety System Logic and Control '
instrumentaion Numberof Divisions i 4 3.4 '
Basic Actuation Logic 2/4- 3.4
- Self-Testing ---
3.4
-i Essential Multiplexing System and SSLC System Tested Quarterly - - - -
COL Req.
important Features from Seismic Analyses Seismic Categoy I Structures O Centeinmeet --
2.14.1 Reactor Building . ---
i 2.15.10.- t Seismic Category I Safety Related '
Components Divisional Station Batteries ---
2.12.12 Battery Racks Battery Chargers
.2.12.12 l 2.12.12 ;
Emergency Diesel Generators --- ..
2.12.13' s
480 Volt Transformers ---
2.12.1-Circuit Breakers '
AC ---
DC 2.12.1 2.12.12'- !
Motor Control Centers AC_ ----
2.12.1 DC ---
2.12.12. t L
A 1 ,
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. = . - _ . _
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} Table 11 j PRA Analysis (Cont.)
1 i Verifying SSAR Entry Parameter Valug
] ITAAC-i important Features from Seismic Analyses l (Cont.)
): SLC System Components Required . "
for RPV Injection are Seismic Category 1 ---
2.2.4 l
- RCW is Seismic Category 1 ---
2.11 3 j AC Independent Water Addition System i
i Seismically Oualified ---
2.15.6 i
. Capable of Vessel injection or
) Drywell Spray ---
2.15.6 1
2.4.1
. Collapse of AClWA Building Does
- not Prevent Pumps from Operating ---
2.15.6 All Valves Needed for System j Operation can be Accessed and 2.15.6 ,
j Operated Manually -- . 2.4.1 l l The RHR is Seismic Category 1 ----
2.4.1 l
? :
j important Features From Fire Protection Analyses
[L A Means of Fire Dectection, Alarming and
} Suppression is Provided and Accessible ---
2.15.6-
, Fire Detection Methods include Infrared
' Sensors and Product-of-Combustion Type Smoke Detectors ---- . none Fire Suppression System 4
Automatic
, Foam ---
2.15.6 Sprinkler ----
2.15.6 i
I Manual
- _ Water Hoses . ----
2.15.6 Hand Held Fire Extinguishers ----
2.15.6-i .
5
.. . - - - , -. - , . . - - . - . ~ . . .. . .
l l
\
Table 11 '
PRA Analysis (Cont.)
i SSAR Entry Verifying Parameter l Malug ITAAC !
Important Features From Fire Protection Analyses (Cont.) :
a Actuation of the FPS is Alarmedin the Control Room ---
2.15.6 Divisional Separation is Provided by Three Hour Fire Barriers ---
2.15.10 '
Remote Shutdown Panel 2.15.12 Physically and Electrically independent from the Control Room ---
2.2.6 '
, Able to Control 1 HPCF,4 SRVs and 1 2 Divisions of RHR ---
! 2.2.6 Able to Operate the RCIC from Outside the Control Room !
none '
)
ESF Systems (including Their Sup Systems, Piping and Cable Trays) portHaving Similar Safety-Related or Shutdown Functions are Divisionally Separated by Three Hour Fire Barriers ----
2.15.10 Features to Prevent Spread of Smoke 2.15.12 I Automatic Closure of Zone Fire Doors if Fire Present ---
none Zone isolation Dampers in HVAC System 2.15.5c i Fire Detection. Alarmed in Control Room -- 2.15.6 l
.O 6
l 1-I , . . - , _ -
Table 11 PRA Analysis (Cont.)
Verifying SSAR Entry Parameter Value ITAAC Important Features from Suppression Pool Bypass and Ex-Containment LOCA Analyses DW-WW Vacuum Breakers Number 8 2.14.1 Mounted on Pedestal Wall ----
2.14.1 l Main Steam isolation Valves l
Redundant Valves in Each Main Steamline ---
2.1.2 Pneumatically Operated ---
2.1.2 Failed-Closed Design --
2.1.2 Actuation Logic 2/4 2.4.3 SRV Discharge Lines Design and Fabricated to Quality-Group C Requirements ---
2.1.2 Welds in Wetwell Airspace are -
Non-Destructively Examined to the l Requirements of ASME Section lil, Class 2 ---
2.1.2 Sample and Drywell Purge Unes are Closed During Normal Operation ---
none
! Blowout Panels in Secondary Containment ---
none i
O 7
- l l
Table 11 O PRA Analysis (Cont.) I SSAR Entrv Parameter Verifying h ITAAC 1 I
important Features from Flooding Analyses 1
The Three Safety Divisions are Physically Separated Reactor Building --- l Control Building 2.15.10 1 2.15.12 '
Entrances to Rooms Containing Safety Related Equipment on the First Floor Have Water Tight Doors j Reactor Building Control Building 2.15.10 2.15.12 Water Tight Doors on All Below Grade Entrances from Service Building Reactor Building Control Building 2.15.10 2.15.12
( Cable Penetrations of Divisional Rooms are Sealed Reactor Building Control Building 2.15.10 2.15.12 Floor Drains in All Upper Floors in Reactor and Control Buildings -
2.9.2 Water Level Sensors in Turbine Building Condenser Pit Trip CWS Pumps ---
2.10.23 l Close CWS isolation Valves -- I 2.10.23 i Alarmed in Control Room 2.10.23 '
Water Level Sensors in RCW Rooms in Control Building Number of Sets 2 Trip RSW Pumps none none t
(Interface i Req.)
Close RSW lsolation Valves ---
2.11.9 1'
Alarmed in Control Room --
none O
'8 l
Table 11 PRA Analysis (Cont.)
SSAR Entry verifying Parameter Value ITAAC important Features from Flooding Analyses (Cont.)
Reactor Building Corridor on floor B3F is Large Enough to Contain the Largest Flood Sources (i.e. CST and Suppression Pool) ---
2.15.10 One Anti-Siphon Valve at the Discharge of Each RSW Pump ---
none (Interface Req. 2.11.9) 1 Reactor Building Sumps on Floor B1F Have
! Overfill Line to the B3F Corridor ---
none i
l Loop Seals on Overfill Lines ..--
none l An Opened Water Tight Dooris Alarmed
,o in the Control Room Reactor Building
() Control Building 2.15.10 2.15.12 Operator Check on Each Shift that Water Tight Doors are Closed and Dogged --
COL Req.
High Pressure or High Termperature Unes not Routed Through Floors or Walls Separating Two Different Safety Divisions ---
2.15.10 (Divisional Important Features from Shutdown Events Separation)
Analyses RHR System Number of Independent Divisions with Shutdown Cooling Mode 3 2.4.1 RSW System l
Numberof Independent Divisions 3 2.11.9 Number of Pumps per Division 2 none (Not in Certified Design) 9 l
j e
l '
l l
I' Table 11 PRA Analysis (Cont.)
IO verifying SSAR Entry Parameter Ltig ITAAC important Features from Shutdown Events. 4 Analyses (Cont.) '
UHS Able to Absorb All RSW Heat Loads ---
none
' (Interface - -
Req. 4.1) j RHR System i
Number of Independent Divisions with '
VesselInjection Mode 3- 2.4.1 Can inject Waterinto the Vessel at Low Pressures ---
2.4.1 >
CRD System
- Number of Pumps that Can Supply C . Water to the Core Through the CRD Purge Unes 2 2.2.2 l
Can inject Waterinto the Vessel at 1
High and Low Pressures ---
none- >
HPCF System -
Initiated on Low RPV Water Level- ---
2.4.2 Number of Subsystems that Can inject Water into the Vessel 2 2.4.2 -
-Can inject Waterinto the Vessel at .
High and Low Pressures ---
2.4.2 -
ACIWA System Number of Pumps with Dedicated..
Diesel 1- 2.15.6 Can Supply Makeup Water to the Reactor at Low Pressure ---
2.15.6 2.4.1 10
~
l Table 11 I PRA Analysis (Cont.) !
Verifying SSAR Entry Parameter Value ITAAC important Features from Shutdown Events Analyses (Cont.)
Unes Connected to RPV isolated on Low Water Level Main Steam Lines ----
2.4.3 RHR Shutdown Cooling Lines ---
2.4.3 CUW Lines ----
2.4.3 RHR Mode Switch has Pressure Permissives and inhibits Associated with l Each Mode of Operation ----
2.7.1 RHR Valve Pressure Interlocks Shutdown Suction Une Valve ---
2.4.1 Vessel injection Valve ---
2.4.1 O RPV Water Level Indication in the Control Room ---
2.1.2 RPS High Flux Scram (set Down) ---
2.2.7 CRD Brake Prevents CRD Ejection ---
2.2.2 ;
Reactor Mode Switch in Refueling Position No More Than Two CRD Blades can be Withdrawn at a Time ---
2.2.1 No Fuel Assembly can be Holsted Over i
the RPV it a CRD Blade has been Removed --
2.2.1 Isolation of Secondary Containment in Modes 3 and 4 on High Radiation Signal ----
2.4.3 SGTS Processes Gases Before Release to the Atmosphere ----
2.14.4
("N G
l 11 i
l
1 4
i a
1 Table 11 j PRA Analysis (Cont.)
1 1
- Verifying i SSAR Entrv Parameter Va!ue ITAAC 1 i j Important Features from Shutdown Events Analyses (Cont.)
l Number of Physically and Electrically j independent Divisions of Safety Related Electrical Power 3 2.12.1 i 2.12.12 i 2.12.14 j Sources of AC Power
- Onsite
) Number of Emergency Diesel Generators 3 2.12.13 Number of Combustion Turbine Generators 1 2.12.11 >
- Offsite
? Number of Independent Sources 2 none (Interface Req. 2.12.1) i important Features to Mitigate Severe Accidents .
Lower Drywell Flooder -
l Actuated by Melting of Fusible Plug ---
2.14.1 2
4 Consists of Unes from Suppression j Pool to Drywell Number of Lines _ 10 2.14.1 l
{ Location of Suction (meters none below suppression poollevel) 4 i
12
Table 11 PRA Analysis (Cont.)
SSAR Entry Verifying Parameter Value ITAAC l l
Important Features to Mitigate Severe Accidents (Cont.)
Centainment Overpressure Protection i
Number of Lines Connecting Wetwell Airspace to Stack 1 2.14.6 l Number of Reclosable Valves in Line 2 2.14.6 Normally Open ----
none Fail Open ---
2.14.6 Vessel Depressurization Valves l
ADS Actuation Logic
(] Redundant ---
2.1.2 Divisionally Separate ---
2.1.2 l
' Accumulator on Each Depressurization Valve for Backup to Normal Pneumatic )
Supply ---
2.1.2 Lower Drywell Design Pedestal Formed by Two Concentric Steal Shells Filled with Concrete ---
2.14.1 Lower Drywell Sumps Designed to Minimized the Amount of Core Debris Entering Them ----
none Path from the Lower to Upper Drywell includes 90 Degree Turns ---
none i
Vessel Skirt Does not Have Any Penetrations Connecting Upper Drywell and Lower Drywell ---
none O
l 13 i
l O
4 l
l Table 11 !
(O ,/ PRA Analysis (Cont.) l Verifying i SSAR Entry Parameter Value ITAAC i
important Features to Mitigate Severe Accidents .
(Cont.) l Containment Normally inerted --
2.14.6 Unes Penetrating Containment have Redundant Isolation Valves --- 2.14.1 Non-Essential Unes in Mitigating Accidents Automatically isolated by Diverse Isolation Signals Main Steam Unes ---
2.4.3 Reactor Water Cleanup System ---- 2.4.3 ;
HVAC System --- 2.4.3 l Containment Purge and Vent Unes --- 2.4.3
-- Reactor Building Cooling Water System --- 2.4.3
()'
HVAC Normal Cooling Water System RHR Shutdown Cooling Lines 2.4.3 2.4.3 Suppression Pool Cleanup System --- 2.4.3 Flammability Control System --- 2.4.3 l Drywell Sump Discharge Unes --- 2.4.3 i I
Fission Products Monitor Drywell Sampling Unes --- 2.4.3 l
Unes Usefulin Mitigating Accidents are Automatically isolated on Signal Indicating l Break in Une Main Steam Line --- 2.4.3 '
Reactor Water Cleanup System --- 2.4.3 RCIC Steam Une --- 2.4.3 RHR Shutdown Cooling Line --- 2.4.3 (3
L) 14
I p Table 11 PRA Analysis (Cont.)
V Verifying SSAR Entry Parameter Yalug ITAAC Important Features to Mitigate Severe Accidents (Cont.)
Design Pressure of Some Low Pressure Components Upgraded to 28.8 atg (Paradiso .4*1025 psig - Ultimate Rupture Strength can withstand 1025 psig)
(Design Description Only)-
RHR System ---
2.4.1 HPCF System --
2.4.2 RCIC System ---
2.4.4
- CRD System ----
2.2.2 SLC Sydem ---
2.2.4 (later)
CUW System ---
2.6.1 (later)
FPC System --
none Nuclear Boiler System --
none Reactor Recirculation System --
none MUWC System ---
none
/ MUWP System - - -
none Radwaste System (LCW Receiving l Tank and HCW Receiving Tank) ---
none Drywell-Wetwell Vacuum Breakers Number 8 2.14.1 Diameter (inches) 20 2.14.1 Mounted on Pedestal Wall ----
2.14.1 i
l Positive Position Indication in Control Room 2.14 1
! Tested During Plant Outages ---
COL Req. ;
l l
i i
'q .
v l
15 l 4
m.m.m
Table 11 O PRA Analysis (Cont.)
SSAR Entrv Verifying
- Parameter Malug ITAAC i
-Important Features to Mitigate Severe Accidents (Cont.)
j Residual Heat Removal Sytem Number of Independent Loops 3 2.4.1 Able to Pump Saturated Water ---
2.4.1 Shutoff Head (kg/cm2d -- Vesselto Drywell) 15.8 2.4.1 Number of RHR Loops Required for Longterm Heat Removal 1 none Number of Loops with Drywell Spray Capability 2 2.4.1 :
l
\ Number of Loops with Wetwell Spray.
Capability 2 2.4.1 Table 1 Key Severe Accident Parameters !
initial Core Thermal Power (MW t) 3926~ none
.I Elevation of TAF (m -- relative to vessel"0") 9.05 none Elevation of Vessel Normal Water Level l
(m - relative to vessel"0") 13.26 none l
ADS Total Flow Area (m2) 0.07 none !
l Containment Leak Rate (% per day) 0.5 '2.14.1 1
Containment Capability - Service Level C j (psig) 97 none Containment Ultimate Strength (psig) 134 none Total Zr in Core (kg) 72550 none '
i . 16 l
l l
Table 11 PRA Analysis (Cont.)
Verifying SSAR Entrv Parameter Value ITAAC Important Features to Mitigate Severe Accidents (Cont.)
Table 1 Key Severe Accident Parameters (Cont.)
Sacrificial Concrete Layer Above
- Containment Liner -
l t
Type of Concrete Basaltic ?.14.1 Thickness (m) 1.5_ 2.14.1
, Thickness of Pedestal (m) 1.7 2.14.1 Compartment Volumes l Lower Drywell(m3)- 1860 none
'f k Upper Drywell(m3) 5490 none Wetwell(m3) 9585 none Floor Areas l Lower Drywell(m2) 88 none .
l Upper Drywell(m2) 610 none Wetwell(m2) 507 none Tolerance of Vacuum Breaker Position Switch (cm) 0.9 none Overflow Elevation (relative to vessel "0")
L Lower Drywell to Wetwell(m) -4.55 none Upper Drywell to Wetwell(m) 7.35 none 17 r
Table 11 (q> PRA Analysis (Cont.) -
SSAR Entry verifying Parameter Yalue. ITAAC Important Features to Mitigate Severe Accidents (Cont.)
Table 1 Key Severe Accident Parameters (Cont.)
Lower Drywell to Upper Drywell Vent Area (m2)- 11.25 2.14.1 t_ower Drywell Flooder Elevation of Discharge (m - relative to vessel "0") -10.5 none Area Per Valve (m2) - 0.0081 none Plug Melting Temperature (OK) 533 none Suppression Pool Water Mass (kg x 106 ) 3.6 - none Containment Overpressure Protection Diameter of Disk (m) 0.2 none Diameter of Piping (m) 0.36 none Setpoint (MPa) 0.72 none Tolerance at Nominal Termperature (%) 5 none Effect of Temperature on Setpoint
(% per 1000F) 2 none r
18
j l
3 1
Table 11 1 PRA Analysis (Cont.)
SSAR Entry verifying Parametet yg[gg ITAAC l important Features to Mitigate Severs Accidents
}
(Cont.)
Table 1 Key Severe Accident Parameters (Cont.)
1
- Firewater Addition System injection Locations j Through LPFL Header 9
2.15.6 2.4.1 ;
{
Through Drywell Spray Header ---
2.15.6- 4 2,4.1 Nominal Runnout Flowrate (m3/sec) 0.055 none Flowrate (m3/sec) 0.044 none at Containment Pressure (MPa) 0.72- none l
\
I i
O 19
i l
I Table 13
()
- I TMIissues i
SSAR Entry Verifying Parameter Value ITAAC i
l 19A.2.41 i.C.5 Procedures for Feedback of Operating l Experience to Plant Staff Procedure to evaluate design and construction experience --
COL Req. !
1 A.2.3 1.D.2 Plant Safety Parameter Display Console l
Safety Parameter Display Console Integrated into Control Room Design ---
COL Req.
19A.2.17 1.D.3 Safety System Status Monitoring Automatic Indication of Bypassed and inoperable Status of Safety Systems ---
3.4 19B.2.65 1.D.5(2) Plant Status and Post-Accident Monitoring
\ -)
Post-Accident Information Available to Refer to the Operator is in Compliance with RG 1.97 --- II.F.1 198.2.66 1.D.5(3) On-Line Reactor Surveillance System ABWR Design incorporates a Reactor Vessel Loose Parts Monitoring System :
2.8.4 t 1 A.2.4 1.G.1 Training Requirements I i
Operator Training Requirements ----
COL Req. ;
19B.2.67 1.G.2: Scope of Test Program I ABWR will have a Test Program to Evaluate Plant Operating Procedures ---
COL Req.
'w) 1
m Table 13 TMI issues (Cont.)
Verifying SSA.R Entrv Parameter Value ITAAC 1 A.2.5 ll.B.1 Reactor Coolant System Vents Reactor Vessel Head Vent Valves 2 none Steam-Driven RCIC 1 2.4.4 Power-Operated Relief Valves Number 18 2.1.2 Dual Position Indication Position Sensors ----
2.1.2 SRV Discharge Temperature Elements ----
2.1.2 gm Remotely Operable from the
() Control Room ---
2.1.2 Plant Specife Procedures for Venting Reactor Pressure ----
COL Req.
1 A.2.6 ll.B.2 Plant Shielding to Provide Access to Vital Areas and Protect Safety Equipment for Post-Accident Operation Vital Areas as per NUREG-0737 Accessible Post-LOCA Continuous Occupancy ---
3.2 Non-Continuous Occupancy ---
3.2
(~h V
2
Table 13 TMIissues (Cont.)
SSAR Entry Verifying Parameter Value ITAAC 1 A.2.7 II.B.3 Post-Accident Sampling Able to Obtain Samples Under-Accident Conditions --
2.11.20 Able to Quantify Radionuclides including Noble Gases,- lodines and Cesiums, '
Nonvolatile isotopes --
none '
Able to Perform Boron and Choride Chemical Analysis ----
none 19A.2.21 II.B.8 Rulemaking Proceeding on Degraded Core Accidents inerted Primary Containment - - -
2.14.6 ;
Permanently-installed Recombiners O 2.14.8 1 A.2.9 II.D.1 Testing Requirements SRVs Qualified for Steam Discharge --
2.1.2 3 Redundant Logic to Respond to High Water Level Conditions ---
3.4 RHR Shutdown Cooling Systems Number 3 2.4.1 Separate Vessel Penetration and Suction Lines ---
2.4.1 l
1 A.2.10 ll.D.3 Relief and Safety Valve Position Indication i
Dual Position Indication Position Sensors ---
2.1.2 SRV Discharge Temperature O
w/
Elements ---
2.1.2 3
l i
'- Table 13 TMI issues (Cont.)
SSAR Entry Verifying Parameter Maly.g ITAAC 1 A.2.13 II.E.4.1 Decated Penetrations Recombiners in Secondary Containment Number 2 2.14.8 Permanently Installed ----
~ 2.14.8 ,
1 A.2.14 !
li.E.4.2 Isolation Dependability !
l Diverse Containment isolation i Signals ----
2.4.3 .)
Non-Essential Systems Automatically isolated On l Containment Isolation Signal ---
2.4.3 i Redundant Isolation Valves ---
2.14.1
! Resetting isolation Signal Does I Not Automatically Reopen i
isolation Valves ----
2.4.3 '
Containment Purge and Vent Valves Close on isolation Signals. ---
2.4.3 Fail Closed ---
2.14.6 Remote Pneumatically Operated ---
none !
Close on High Radiation ---
2.4.3 O
4
Table 13
[]_ TMI issues (Cont.)
Verifying SSAR Entrv Parameter Value ITAAC 198.2.68 II.E.6.1 Test Adequacy Study in-service inspection aitd Testing of Safety l Related Valves will be in Accordance with ASME/ ANSI OMa-1988 Addenda to ASME/ ANSI OM-1987, Parts 1,6, and 10 ---
COL Req.
1 A.2.15 ll.F.1 Additional Accident Monitoring instrumentation Plant Post Accident Monitoring Variables Neutron Flux ---
2.7.1 Control Rod Position ---
2.7.1 Boron Concentration ---
2.11.20 (Sampling Only)
ReactorCoolant System Pressure ---
2.7.1 Drywell Pressure ---
2.7.1 (o) Drywell Sump Level --
2.7.1 Coolant Levelin Reactor ---
2.7.1 Suppression Pool Water Level ---
2.7.1 Containment Area Radiation ---
2.7.1 Primary Containment Pressure ---
2.7.1 Primary Containment isolation Valve Position ----
2.7.1 Coolant Gama ---
2.11.20 (Sampling Only)
Coolant Radiation ---
2.3.1 RHR Flow ---
2.7.1 HPCF Flow ---
2.7.1 RHR Heat Exchanger Outlet Temp ---
2.4.1 RCIC Flow ---
2.7.1 SLC Pressure ---
2.7.1 SLCS Storage Tank Level ---
2.7.1 SRV Position ---
2.7.1 Feedwater Flow ---
2.2.3 High Racioactivity Liquid Tank Level ---
none Standby Energy Status ---
2.7.1 l n
( .)
5
i Table 13 D TMIissues (Cont.)
G SSAR Entrv Parameter Verifying )
Value ITAAC 1 A.2.15 t li.F.1 Additional Accident Monitoring l instrumentation (Cont.)
l Plant Post Accident Monitoring Variables l Suppression Pool Water Temp ---
Drywell Air Temperature 2.7.1 I ---
Drywell/ Containment 2.7.1 Hydrogen Concentration ----
Drywell/ Containment 2.7.1 Oxygen Concentration --
2.7.1 Primary Containment Air Temp ----
_2.7.1 Secondary Containment Airspace (effluent) Radiation Noble Gas ---
2.3.1
, Containment Effluent Radioactivity
- Noble Gas 2.11.20 Condensate Storage Tank Level (Sampling Only)
O Cooling Water Temperature to ESF 2.7.1 -
System Components ----
-2.11.3 Cooling Water Flow to ESF System Components ----
2.11.3 Emergency Ventilation Damper Position ---
'none '
Service Area Radiation Exposure Rate ---
2.3.2 Purge Flows - Noble Gases and Vent l l Flow Rate ---
2.3.1 Identified Release points - Particulates and Halogens --
2.3.1
) Airborn Radio Halogens and Particulars ---
! 2.3.1 Pant and Environs Radiation t
/ Radioactivity (Portable Instruments) --- COL Req.
Meterological Data (Wind Speed, Wind I Direction, and Atmospheric Stability) ---- COL Req.
On Site Analysis Capability (Primary )
Coolant, Sump and Containment i Air Grab Sampling) ----
COL Req.
O 6
l
i i
Table 13
,q TMI issues (Cont.)
t NJ Verifying SSAR Entrv Parameter Value ITAAC 1 A.2.16 ll.F.2 Identification of and Recovery from Conditions Leading to inadequate Core Cooling Reactor Wide Range Water Level 4
Numberof Divisions 4 2.1.2 Number of Sensors per Division 2 2.1.2 Number of Sets of Sensing Lines per Division 1 2.1.2 Trip Logic per Set of Sensors 2/4 3.4 Number of Sets of Sensors 2 3.4 O
'd 1 A.2.17 II.F.3 instrumentation for Monitoring Accident Conditions ---
Refer to 1 A.2.15 1 A.2.19 II.K.1(10) Review and Modify Procedures for Removing Safety-Related Systems from Service -- COL Req.
1 198.2.69 II.K.1(13) Propose Technical Specification Changes Reflecting implementation of All Bulletin items ABWR Technical Specifications are in Accordance with NUREG 1433 and 1434 with ABWR Specific Features ---
none a
7
I 1
l Table 13
! TMI issues (Cont.)
Verifying SSAR Entrv Parameter yahnt ITAAC <
1 A.2.20 Describe Automatic and Manual Actions for Proper Functioning of Auxiliary Heat Removal :
Systems when FW System not Operable Reactor Scram on Low Water Level ---
2.2.7 RCIC System initiates on Low Water Level ---
2.4.4 Terminates injection on High Water Level ---
2.4.4 Restarts on Low Water Level ---
2.4.4 RPV Pressure Controlled by Main Turbine Bypass Valves --
2.10.13 '
2.1.2 l Discharge to Suppression Pool ---
2.1.2 :
l p. RHR Systems has Manual Pool Cooling l s Mode --
2.4.1 l HPCF Systems l Initiates on Low. Water Level ----
2.4.2 i ADS Initiates on Low Water Level ---
2.1.2 RHR - LPFL Mode initiates on Low Water Level ---
2.4.1 1 A.2.21 II.K1(23) Describe Uses and Types of RV ,
Level Indication for Automatic and Manual !
Initiation of Safety Systems Shutdown Water-Level Measurement Range Top of RPV ---
2.1.2 Bottom of Dryer Skirt ---
2.1.2 8
l l
m Table 13 i TMI issues (Cont.) l Verifying SSAR Entrv Parameter Value ITAAC 1 A.2.21 II.K1(23) Describe Uses and Types of RV l Level Indication for Automatic and Manual ;
Initiation of Safety Systems (Cont.)
Narrow Water-Level Measurement Range Above Main Steam Outlet Nozzle ---
2.1.2 Bottom of Dryer Skirt ----
2.1.2 Low Water Level 3 Automatic initiation Reactor Scram ---
2.2.7 i RHR Shutdown Cooling isolation - ----
2.4.3 j Containment isolation ----
2.4.3 i l
l Wide Water-Level Measurement l
Range l (_
O) Above Main Steam Outlet Nozzle ---
2.1.2 .
Top of Active Fuel i
2.1.2 I
Low Water Level 2 Automatic Initiation RCIC ---
2.4.4 CUW isolation ---
2.4.3 Low Water Level 1.5 Automatic Initiation HPCF - - -
2.4.2 MSIV Closure ---
2.4.3 Drywell Cooling System Isolation ----
2.4.3 Low Water Level 1 Automatic initiation ADS ---
2.1.2 RHR-LPFL ---
2.4.1 Fuel-Zone Water-Level Measurement Range l Above Main Steam Outlet Nozzle ---
2.1.2 l Above RIP Deck ---
2.1.2 Uniform Drop for Narrow and Wide Range Water-LevelInstrumentation Unes from RPV 3 to Drywell Penetration ---
none (O
9
Table 13
(-)
Q)
TMlissues (Cont.)
SSAR Entrv Verifying Parameter Value ITAAC 1 A.2.21.1 II.K.3(3) Report Safety and Relief Valve Failures Promptly and Challenges Annually ----
COL Req.
! 19A.2.70 ll.K.3(11) Control Use of PORV Supplied by Control Components, Inc. Until Further Review Complete l ABWR does not Have PORVs Supplied Not by Control Components, Inc. ----
Applicable 1 A.2.22 II.K.3(13) Separation of HPCS and RCIC System initiaion Lovels l
RCIC System j
h initiates on Low Water Level ---
2.4.4 Setpoint (Level 2) ---
none 1
' Terminates injection on High Water '
. ('/)
~
Level Setpoint (Level 8) 2.4.4 none Restarts on Low Water Level ---
2.4.4 Setpoint (Level 2) ----
none HPCF System initiates on Low Water Level ---
2.4.2 Setpoint (Level 1.5) - - -
none Terminates injection on High Water Level ---
2.4.2 Setpoint (Level 8) ---
none Restarts on Low Water Level ---
2.4.2 Setpoint (Level 1.5) ---
none
- V.
l 10
\ -
I i
i Table 13 TMI issues (Cont.) :
SSAR Entry Verifying Parameter Value ITAAC 1 A.2.23 II.K.3(15) Modify Break Detection Logic to Prevent Spurious isolation of HPCI and RCIC 1 Systems '
- RCIC has a Bypass Start System ----
2.4.4 1 A.2.24 ll.K.3(16) Reduction of Challenges and Failures of Safety Relief Valves - Feasibility Study and System Modification Elimination of Pilot Operated Relief Valves ---
2.1.2 (Design Des. Only)
Redundant Solid State Logic ----
3.4 Pressure Relief Mode Operation is Direct Opening Against Spring Force O ----
2.1.2 (Design Des. Only) l I
1 A.2.25 ll.K.3(17) Report on Outage of ECC Systems-Licensee Report and Technical Specification Changes ---
COL Req.
1 A.2.26 ll.K.3(18) Modification of ADS Logic-Feasibility Study and Modification for increased Diversity -
of Some Event Sequences High Drywell Pressure Bypass Timer (minutes) 8 2.1.2 initiates on Low Water Level ----
2.1.2 Setpoint (Level 1) ----
Not Tier 1 l 1 A.2.27 . II.K.3(21) Restart of Core Spray and LPCI No Systems on Low Level- Design and Modification Modification
~
Required 11
c' Table 13 l
(v) TMI issues (Cont.) '
Verifying SSAR Entry Parameter Value ITAAC l
, 1 A.2.28 II.K.3(22) Automatic Switchover of RCIC l System Suction - Verify Procedures and Modify l Design RCIC Automtically Swtiches Pump Suction Source From CSP toSuppression Pool ---
2.4.4 Switchover Signals Low CSP Water Level ---
2.4.4 or High Suppression Pool Level ---
2.4.4 l 1 A.2.29 lI.K.3(24) Confirm Adequacy of Space Cooling Study for HPCI and RCIC Systems individual Room Safety Grade q Cooiing Units V RCIC ----
2.15.5c HPCF ---
2.15.5c Separate Essential Electrical Power Supples RCIC ---
2.4.4 HPCF ---
2.4.2 1 A.2.30 ll.K.3(25) Effect of Loss of AC Power on Pump Seals RIPS do not Require Shaft Seals ---
none RCW and RSW Pumps Automatically 2.11.3 Loaded to D / Gs Following LOPP ----
2.11.9 2.12.13 RCW Primary Containment Isolation Valves Do Not Close on LOPP ----
none l
/ \
12
i Table 13 TMIissues (Cont.)
- SSAR Entrv Parameter Verifying 1
Value JTAAC j 1 A.2.30 i ll.K.3(25) Effect of Loss of AC Power on Pump Seals (Cont.)
i 4
RIP Motors in Stopped Hot Standby Condition Will Not Be Damaged with RCW Cooling Available none CRD Pumps Automatically Loaded to
{
. Diesel Generators Following LOPP ----
none
' 19B.2.71 II.K.3(27) Provide Common Reference Level for VesselInstrumentation For ABWR the Common Reference for the Reactor Vessel Water Level is at the Top of the Active Fuel ---
2.1.2 (Design 1 A.2.31 II.K.3 Des. Only) j Accum(28) Study and Verify Qualification of ulators on ADS Valves Accumulator Sized to Provide One ADS Actuation with Drywell at Design Pressure ---
2.1.2 Seismic Category l Pneumatic Piping within Primary Containment -----
2.11.13 Two Redundant 7-day Supplies of Bottled Air ---
none Loss of Pneumatic Supply Pressure to the Accumulators is Alarmed ---
none 1 A.2.32 II.K.3(30) Revised Small-Break LOCA Methods to Show Compliance with 10CFR50, Appendix K ABWR LOCA Evaluations Performed with the GE SAFER Evaluation Modelin Accordance with 10CFR50.46, Appendix K ---
none O
13
I i
i Table 13 -
TMI issues (Cont.)
SSAR Entry Verifying l Parameter _%%1g ITAAC. l 1 A.2.33 II.K.3(31) Plant-Specific Calculations to Show- Refer to Compliance with 10CFR50.46 1 A.2.32 1 A.2.33.1 II.K.3(44) Evaluation of Anticipated Transients with Single Failure to Verify No Siignificant '
Fuel Failure ABWR Transient Evaluation Performed Assuming Worst Single Failure ---
none
)
1 A.2.33.2 II.K.3(45) Evaluate Depressurization with other No than Full ADS ,
Modification Required 1 A.2.33.3 II.K.3(46) Response to List of Concems from ACRS Consultant High Pressure injection ECCS O RCIC HPCF 2 1 2.4.4 2.4.2 1
Automatic Depressurization on Low Vessel Water Level -- . 2.1.2 ECCS injection Directly into Vessel HPCF 2 2.4.2.
RHR-LPFL 2 2.4.1 ECCS Injection into Feedwater Lines RCIC 1 2.4.4 RHR-LPFL 1 2.4.1 ECCS injection Lines Maintained Filled with Water '
RCIC --
2.4.4 HPCF ----
2.4.2 RHR-LPFL ---
2.4.1 l
O 14 !
I l
1 Tabie13 TMlissues (Cont.)
SSAR Entrv Verifying Parameter Value ITAAC 1 A.2.33.3 II.K.3(46) Response to List of Concerns from ACRS Consultant (Cont.)
High Pressure ECCS Designed to Take Suction from Suppression Pool RCIC ---
2.4.4 HPCF ---
2.4.2 High Pressure ECCS have a Designed Test Mode which Takes Suction from and Discharges to the Suppression Pool -
RCIC ----
2.4.4 HPCF ---
2.4.2 High Pressure ECCS have a Designed Low Flow Bypass Mode which Dicharges to the Suppression Pool RCIC 2.4.4 G
HPCF V ---
2.4.2 RCIC and HPCF Do not Share Any Common Suction Piping with RHR RCIC --
2.4.4 HPCF -
2.4.2 LPFL ---
2.4.1 '
Suction Piping of RCIC and HPCF Sized i
to insure Adequate NPSH if All Pumps Running Simultaneously --
none Pre-op Tests are Conducted to Demonstrate NPSH Requirements are Met ----
none ECCS Have Minimum Flow Protection for All Operating Modes RCIC ---
2.4.4 HPCF ---
2.4.2 RHR ---
2.4.1 15 O
, , , ,_, . .n- , *=~ *-
Table 13 D
TMI lssues (Cont.)
b Verifying SSAR Entrv Parameter Value ITAAC i 1
1 A.2.33.3 II.K.3(46) Response to List of Concerns from ACRS Consultant (Cont.)
RHR and HPCF Seals Cooled by HCW ----
none l Number of RCW Divisions 3 2.11.3 i High Level Alarms in ECCS Room Sumps ----
none '
[ James -- General agreement not to cover alarms] '
individual ECCS Pumps Can be Isolated Without Affecting Other ECCS Pumps RCIC --
2.4.4 HPCF ---
2.4.2 RHR ---
2.4.1 ABWR has Water Level Measurement Directly on the Vessel ---
2.1.2 Containment Sprays are Manually initiated ---
2.4.1 Essential Equipment Inside the Containment is Qualified for Harsh Environment ---
2.14.1 ADS Automatically Depressurizes the Vessel on Low Water Level ----
2.1.2 ABWR has Manual Vessel Depressurization Capability ---
2.1.2 l 19B.2.72 Ill.D.3.3(1) Issue Letter Requiring improved Refer to Radiation Sampling Instrumentation ---
lil.D.3.3(2) 19B.2.73 Ill.D.3.3(2) Set Criteria Requiring Licensees to Evaluate Need for Additional Survey Equipment '
Portable Instruments to Measure Radio-lodine Concentration in Plant Areas Under Accident Conditions ----
Col Req.
C)v 16 I
A Table 13 U TMI issues (Cont.)
Verifying SSAR Entry Parameter Value ITAAC 1 A.2.36 Ill.D.3.4 Control Room Habitability HVAC System Redundant Safety Grade Systems with Outdoor Air intakes ---
2.15.5c Charcoal Filters (inches) 2 none Able to Maintain 3.2 mm WG Positive Pressure in Habitable Control Room ----
2.15.5c Radiation and Smoke Sensors in intake Lines to isolate Outdoor Air intake ---
2.15.Sc Habitable Control Room Shielding
~3 Min. Thickness of Concrete Between (Q Habitable Control Room Area and Steam Lines (meters) 1.6 2.15.12 Control Room Constructed Below Grade Level ---
2.15.12 l
l 17 i l