ML20117H356
| ML20117H356 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 08/31/1996 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20117H333 | List: |
| References | |
| 23A6100, 23A6100-R09, 23A6100-R9, NUDOCS 9609090230 | |
| Download: ML20117H356 (250) | |
Text
{{#Wiki_filter:-_--- __ _ __ _ _ _ - _. . _ _ _ l @ 0 GENuclearEnergy 23A6100 Rev. 9 Aug.1996 , s
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l l 1 l Standard ! Safety lO Analysis l Report : i i O
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23A6100 Rsv. 9 ABWR standardsafety Analysis neport n 1.2.2.12.14 Heating Steam and Condensate Water Return System The Heating Steam and Condensate Water Return System supplies heating steam from the House Boiler for general plant use and recovers the condensate return to the boiler feedwater tanks. The system consists of piping, valves, condensate recovery set and l associated controls and instrumentation. l l 1.2.2.12.15 House Boiler System l The House Boiler System consists of the house boilers, reboilers, feedwater components, boiler water treatment and control devices. The House Boiler System supplies turbine gland steam and heating steam, including the concentrating tanks and devices of the high conductivity waste equipment. 1.2.2.12.16 Hot Water Heating System The Hot Water Heating System is a closed-loop hot water supply to the various heating i l coils of the HVAC systems. The system includes two heat exchangers, surge and chemical addition tanks and associated equipment, controls and instrumentation. g 1.2.2.12.17 Hydrogen Water Chemistry System \ The Hydrogen Water Chemistry System is summarized in Subsection 9.3.9.2. 1.2.2.12.18 Zinc injection System l The Zinc Injection System is summarized in Subsection 9.3.11.1. 1.2.2.12.19 Breathing Air System l The Breathing Air System includes air compressors, dryers, purifiers and a distribution network. This network makes breathing air available in all plant areas where operations or maintenance must be performed and high radioactivity could occur in the ambient air. Special connections are provided to assure that this air is used only for breathing apparatus. 1.2.2.12.20 Sampling System (includes PASS) ! l The Process Sampling System is furnished to provide process information that is required to monitor plant and equipment performance and changes to operating - parameters. Representative liquid and gas samples are taken automatically and/or I manually during plant operation for laboratory or online analyses. l l 1.2.2.12.21 Freeze Protection System ; O q The Freeze Protection System provides insulation, steam and electrical heating for all l external tanks and piping that may freeze during winter weather. l l l General Plant Description - Amendment 37 1.2 31
23A6100 R1v. 4 ABWR StandardSafety Analysis Report 9 1.2.2.12.22 Iron injection System The Iron Injection System consists of an electrolytic iron ion solution generator and means to inject the iron solution into the feedwater system in controlled amounts. 1.2.2.13 Station Electrical Systems 1.2.2.13.1 Electrical Power Distribution System The unit Class 1E AC power system supplies power to the unit Class 1E loads. The offsite power sources converge at the system. The system includes diesel generators that serve as standby power sources, independent of any onsite or offsite source. Therefore, the system has multiple sources. Furthermore, the system is divided into three divisions, each with its own independent distribution network, diesel generator, and redundant load group. A fourth division battery for the safety logic and control system bus receives charger power from the Division II source. 1.2.2.13.2 Unit Auxiliary Transformer The unit auxiliary AC power system supplies power to unit loads that are non-safety-related and uses the main generator as the normal power source with the resene auxiliary transformer as a backup source. The unit auxiliary transformer steps down the AC power to the 6900V station bus voltage. 1.2.2.13.3 Isolated Phase Bus The isolated phase bus duct system provides electrical interconnection from the main generator output terminals to the generator breaker and from the generator breaker to the low voltage terminals of the main transformer, and the high voltage terminals of the unit auxiliary transformers. During the time the main generator is offline, the generator breaker is open and power is fed to the unit auxiliary transformers by backfeeding from the main transformer. During startup, the generator breaker is closed at about 7% power to provide power to the main and the unit auxiliary transformers for normal operation of the plant. A package cooling unit is supplied with the isolated bus duct system. 1 2.2.13.4 Non Segregated Phase Bus The non-segregated phase bus provides the elecuicalinterconnection between the unit auxiliary transformers and their associated 6.9 kV metal-clad switchgear. 1.2.2.13.5 Metal-clad Switchgear The metal < lad switchgear distributes the 6.9 kV power. Circuit breakers are drawout type, stored energy vacuum breakers. The switchgear interrupting rating shall be determined in accordance with requirements of ANSI C37.10. 1.2 32 General Plant Description - Amendment 34
23A6100 Rev. S j A r3wg se,,g,,g g,,,,y a,,1,;, 7 g,,,,, 1 l w l Table 1.91 Summary of ABWR Standard Plant COL License information (Continued) Item No. Subject Subsection 15.5 Mislocated Fuel Bundle Accident 15.4.11.1 15.6 Misoriented Fuel Bundle Accident 15.4.11.2 15.7 lodine Removal Credit 15.6.7.1 15.8 Deleted 15.9 Radiological Consequences of Non-Line Break Accidents 15.7.6.1 l 16.1 COL Information Required for Plant Specific Technical 16.1.1 I l Specifications
, l 17.1 QA Programs For Construction And Operation 17.0.1.1 l
17.2 Policy and Implementation Procedures for D-RAP 17.3.13.1 l 17.3 D-RAP Organization 17.3.13.2 17.4 Provision for O-RAP 17.3.13.3 18.1 HSI Design Implementation Process 18.8.1 18.2 Number of Operators Needing Controls Access 18.8.2 /* 18.3 Automation Strategies and Their Effects on Operator Reliability 18.8.3 (w 18.4 SPDS Integration With Related Emergency Response 18.8.4 ' Capabilities 18.5 Standard Design Features Design Validation 18.8.5 18.6 Remote Shutdown System Design Evaluation 18.8.6 ; 18.7 Local Valve Position Indication 18.8.7 l Operator Training 18.8 18.8.8 18.9 Safety System Status Monitoring 18.8.9 18.10 PGCS Malfunction 18.8.10 18.11 Local Control Stations 18.8.11 18.12 As-Built Evaluation of MCR and RSS 18.8.12 18.13 Accident Monitoring Instrumentation 18.8.13 18.14 In-Core Cooling instrumentation 18.8.14 18.15 Performance of CriticalTasks 18.8.15 18.16 Plant Status and Post-Accident Monitoring 18.8.16 19.1 Post Accident Recovery Procedure for Unisolated CUW 19.9.1 Line Break 19.2 Confirmation of CUW Operation Beyond Design Bases 19.9.2
.O \
19.3 Event Specific Procedures for Severe External Flooding 19.9.3
\ 19.4 Confirmation of Seismic Capacities Beyond the Plant 19.9.4 Design Bases COL License Information - Arrendment 3S 1.9 11
23A6100 Rsv. 9 ABWR standardsafety Analysis neport O Table 1.9-1 Summary of ABWR Standard Plant ! COL License information (Continued) Item No. Subject Subsection 19.5 Plant Walkdowns 19.9.5 19.6 Confirmation of Loss of AC Power Event 19.9.6 19.7 Procedures and Training for Use of AC-Independent 19.9.7 Water Addition System 19.8 Action to Avoid Common-Cause Failures in the Essential 19.9.8 Multiplexing System (EMUX) and Other Common-Cause Failures 19.9 Action to Mitigate Station Blackout Events 19.9.9 19.10 Actions to Reduce Risk of internal Flooding 19.9.10 19.11 Actions to Avoid Loss of Decay Heat Removal and 19.9.11 Minimize Shutdown Risk 19.12 Procedures for Operation of RCIC from Outside the 19.9.12 Control Room 19.13 ECCS Test and Surveillance Intervals 19.9.13 I 19.14 Accident Management 19.9.14 19.15 Manual Operation of MOVs 19.9.15 19.16 High Pressure Core Flooder Discharge Valve 19.9.16 19.17 Capability of Containment isolation Valves 19.9.17 19.18 Procedures to Ensure Sample Lines and Drywell Purge 19.9.18 Lines Remain Closed During Operation l 19.19 Procedures for Combustion Turbine Generator to Supply 19.9.19 Power to Condensate Pumps 19.19a Actions to Assure Reliability of the Supporting RCW and 19.9.20 Service Water Systems 19.19b Housing of AICWA Equipment 19.9.21 19.19c Procedures to Assure SRV Operability During Station 19.9.22 Blackout 19.19d Procedures for Ensuring Integrit'yof Freeze Seals 19.923 19.19e Procedures for Controlling Combustibles During 19.9.24 Shutdown 19.19f Outage Planning and Control 19.9.25 19.199 Reactor Service Water Systems Definition 19.9.26 l 19.19h Capability of Vacuum Breakers 19.9.27 1.9 12 COL License Information . Amendment 37
I 23A6100 Rsv. 9 ABWR standardsatory Analysis aeport O Table 3.2-1 Classification Summary (Continued) Quality Quality Assur-Group ance Safetg Classi- Require- Seismic Principal Component
- Class Location
- ficationd ment
- Category' Notes
- 6. Other non-safety- N SC,RZ,X -
E - related electrical components P14 Heating Steam and N T,SC,W - E - Condensate Water Return System P15 House Boiler N T - E - l P16 Hot Water Hetting System N T - E - l P17 Hydrogen Water Chemistry N T - E - System P18 Zincinjection System N T - E - P19 Breathing Air System N C,SC,T - E - P20 Sampling System (includes N SC,RZ,T - E - l PASS) l i 1 P21 Freeze Protection System N O - E - ! P22 Iron Injection System N T - E - R1 Electrical Power Distribution System
- 1. 120 VAC safety-related 3 SC,X, -
B I distribution equipment RZ,U including inverters p 2. Safety-related Motors 3 SC,C,X, - B l l
' RZ,U Notes and footnotes are listed on pages 3.2-54 through 3.2-61 l Classification of Structures. Components, and Systems - Amendment 37 3.243 i
, 23A6100 Rsv. 4 ABWR Standard Safety Analysis Report O
Table 3.2-1 Classification Summary (Continued) Quality Quality Assu r-Group ance Safetg Classi- Require- Seismic Principal Component
- Class Location
- ficationd ment' Category' Notes
- 3. Safety-related 3 SC,X,RZ, -
B l Protective relays and U control panels
- 4. Safety-related Valve 3 SC,C, X, -
B I l operators RZ,U R2 Unit AuxiliaryTransformers
- 1. Unit Auxiliary N O -
E - Transformers l 2. Safety-related 3 RZ - B i Transformers R3 Isolated Phase Bus N O,T - E - R4 Non Segregated Phase Bus N O,T - E - R5 Metalciad Switchgear l 1. Safety-related 6900 3 RZ - B l Velt switchgear R6 PowerCenter l 1. Safety-relawd 480 Volt 3 RZ,U - B I power centera R7 Motor Control Center l 1. Safety-related 480 Voit 3 X,RZ,U - B i motor control centers Notes and footnotes are listed on pages 3.2-54 through 3.2-61 0 3.2M Classification of Structures, Components, and Systems - Amendment 34
23A6100 Rtv. 9 ABWR standard safety Analysis Report b Those beams and columns supporting pipe supports will carry piping dynamic loads without buckling and while remaining clastic. Those beams and columns supporting pipe whip restraints allow inelastic deformations due to pipe rupture loads. All safety-related items which the inelastic beam deformations may effect are evaluated to verify that no required safety function would be compromised. 3.8.3.4.5 Other Internal Structures The desip and analysis procedures used for other internal structures are similar to those used for the drywell equipment and pipe support structure as described in Subsection 3.8.3.4.4. 3.8.3.5 Structural Acceptance Criteria 3.8.3.5.1 Drywell Equipment and Pipe Support Structure The structural acceptance criteria for the DEPSS are in accordance with ANSI /AISC-N690. 3.8.3.5.2 Other Internal Structures g The structural acceptance criteria for other internal concrete or steel structures are in accordance with ACI-349 and ANSI /AISC-N690, respectively. 3.8.3.6 Materials, Quality Control, and Special Construction Techniques 3.8.3.6.1 Diaphragm Floor The materials, quality control, and construction techniques used for the diaphragm floor and liner plate are the same as those used for the containment wall and liner plate in Subsection 3.8.1.6. 3.8.3.6.2 Reactor Pedestal The materials conform to all applicable requirements ofANSI/AISC N690 and ACI 349 and complywith the following: Item Specification Inner and outer shells ASThi A441 or A572 (excluding the portions submerged in the suppression pool) Internal stiffeners ASThi A441 or A572 Concrete fill f J= 27.56 hiPa Outer shell submerged in the ASThi A533, Type B, Class 2 with suppression pool SA-240 Type 304 L clad Seismic Category I Structures - Amendment 37 3.8-29
23A6100 Rsv. 9 ABWR standardsafety Analysis Report O 3.8.3.6.3 Reactor Shield Wall The materials conform to all applicable requirements of ANSI /ASIC N690 and ACI 349 and comply with the following: Item Specification Inner and outer shells ASTM A441 or A572 Internal stiffeners ASTM A441or A572 Concrete fill f c'= 27.56 MPa minimum 3.8.3.6.4 Drywell Equipment and Pipe Support Structure The materials conform to all applicable requirements of ANSI /AISC N690 and comply with the following: l Item Specification j Structural steel and connections ASTM A36 High strength structural steel plates ASTM A572 or A441 Bolts, studs, and nuts (dia. > 19 mm) ASTM A325 l Bolts, studs, and nuts (dia. 519 mm) ASTM A307 3.8.3.6.5 Other Internal Structures The materials conform to all applicable requirements ofANSI/AISC N690 and comply with the following: Item Specification Miscellaneous platforms Same as Subsection 3.8.3.6.4 Lower drywell equipment tunnel ASTM A533, Type B, Class 2 with SA-240 Type 304 L clad Lower dnwell personnel tunnel ASTM A533, Type B, Class 2 with SA-240 Type 304 L clad Lower dowell floor fill material A material other than limestone concrete 3.8-30 Seismic Category I Structures - Amendment 37
l 23A6100 Rsv. 4 ABWR standantsaretyAastysisneport CT U Table 3.9-8 Inservice Testing Safety-Related Pumps and Valves 1 Valve Page MPL System Pump Page No. No. l B21 Nuclear Boller 3.9-99 l B31 Reactor Recirculation 3.9-102 l C12 Control Rod Drive 3.9-103 l C41 Standby Liquid Control 3.9-99 3.0-103 l C51 Neutron Monitoring (ATIP) 3.9-104 l D23 Containment Atmospheric Monitoring 3 Slo 4 l E11 Residual Heat Removal 3.9-99 3.9-105 l E22 High Pressure Core Flooder 3.9-99 3.9-110 l E31 Leak Detection & Isolation 3.9-112 l E51 Reactor Core Isolation Cooling 3.9-99 3.9-113 l l G31 Reactor Water Cleanup 3.9-118 j l G41 Fuel Pool Cooling & Cleanup 3.9-119 l G51 Suppression Pool Cleanup 3.9-121 l K17 Radwaste 3.9-121 l P11 Makeup Water (Purified) 3.9-121 l P21 Reactor Building Cooling Water 3.9-99 3.9-121 l P24 HVAC Normal Cooling Water 3.9-127 l P25 HVAC Emergency Cooling Water 3.9-99 3.9-127 l P41 Reactor Service Water 3.9-99 3.9-130 l P51 Service Air 3.9-131 l PS2 Instrument Air 3.9-131 l P54 High Pressure Nitrogen Gas Supply 3.9-131 l T22 Standby Gas Treatment 3.9-132 l T31 Atmospheric Control 3.9-134 l T49 Flammability Control 3.9-137 l U41 Heating, Ventilating and Air Conditioning 3.9-138 Y52 Oil Storage and Transfer 3.9-99 3.9-139 See page 3.9-140 for notes. 1 This table responds to NRC Questions 210.47,210.48 and 210.49 regarding provisions for (^~) inservice testing of safety-related pumps and valves within the scope of the ABWR Standard Plant in accordance with the ASME Code. The information is presented separately for each system for the MPL number. Mechanical Systems and Components - Amendment 34 3.9-97
I 23A6100 Rev. 4 l ABWR standardsafety Analysis Report l O Table 3.9 8 Inservice Testing Safety-Related Pumps and Valves' (Continued) Safety Test Test Class Param Freq. SSAR Fig. No. Oty Description (h) (i) (a) (b) (f) (g) System Pumps C41-C001 2 Standby Liquid Control System Pump 2 Pd,Vd, 3 mo 9.3-1 Q E11-C001 3 Residual Heat Removal System Pump 2 Pd, Pi, 3 mo 5.4-10 Q, Vv (Sh. 3, 4, 6) E11-C002 3 Residual Heat Removal System fill 2 Pd,Pi, E10 5.4-10 pump (i1) Vv (Sh. 3, 4, 6) E22-C001 2 High Pressure Core Flooder pump 2 Pd,Pi, 3 mo 6.3-7(Sh. 2) Q,Vv E51-C001 1 Reactor Core Isolation Cooling pump 2 N,Pd,Pi 3 mo 5.4-8(Sh.1) ; Q,Vv l l P21-C001 6 Reactor Building Cooling Water pump 3 Pd, Pi, E10 9.2-1 l Q, Vv (Sh.1, 4, 7) l P25-C001 6 HVAC Emergency Cooling Water 3 Pd, Pi, E10 9.2-3 System pump Q, Vv (Sh.1, 2, 3) l l P41-C001 6 Reactor Service Water System pump 3 Pd, Pi, E10 9.2-7 Q, Vv (Sh.1, 2, 3) YS2-C001 6 Standby D/G Fuel Oil Transfer Pump 3 Pd, Pi, 3 mo 9.5-6 Q, Vv 1 This table responds to NRC Questions 210.47,210.48 and 210.49 regarding provisions for inservice testing of safety-related pumps and valves within the scope of the ABWR Standard Plant in accordance with the ASME Code.The information is presented separately for each system for the MPL number. I O 3.9-98 Mechanical Systems and Components - Amendment 34
23A6100 Rsv. 4 ABWR StandardSafety Analysis Report m 1 y 1 1 Table 3.9-8 inservice Testing Safety-Related Pumps and Valves (Continued) ' Safety Code Valve Test Test Class Cat. Func Para Freq SSAR i No. Oty Description (h) (l) (a) (c) (d) (e) (f) Fig. (g) F705 1 Pump discharge line 2 B P E1 5.4-8 pressure instrumentation sh.1 instrument root valve F706 1 Pump discharge line flow 2 B P E1 5.4-8 instrument root valve sh.1 F707 1 Pump discharge line flow 2 B P E1 5.4-8 instrument root valve sh.1 F708 1 Pump discharge line flow 2 B P E1 5.4-8 instrument root valve sh.1 F709 1 Pump discharge line flow 2 B P E1 5.4-8 instrument root valve sh.1 F710 1 Pump discharge line 2 B P E1 5.4-8 pressure instrument root sh.1 valve (O F711 1 Pump discharge line pressure instrument root 2 B P E1 5.4-8 sh.1 valve F712 1 Turbine accessories cooling 2 8 P E1 5.4-8 water line instrument root sh.3 valve F713 1 Turbine accessories cooling 2 B P E1 5.4-8 water line instrument root sh.3 valve F714 1 Turbine accessories cooling 2 B P E1 5.4-8 water line instrument root sh.3 valve I F716 1 Steam supply line pressure 2 B P E1 5.4-8 l instrument root valve sh.2 F717 1 Steam supply line pressure 2 B P E1 5.4-8 i instrument root valve sh.2 I F718 1 Steam supply line drain pot 2 B P E1 5.4-8 instrument root valve sh.2 i 1 F719 1 Steam supply line drain pot 2 B P E1 5.4-8 I instrument root valve sh.2
)
F720 1 Steam supply line drain pot 2 B P E1 5.4-8 ! p instrument root valve sh 2 I k F721 1 Steam supply line drain pot 2 B P E1 5.4-8 instrument root valve sh.2 Mechanical Systems and Components- Amendment 34 3.9-117
23A6100 R1v. 9 j ABWR StandardSafety Analysis Report O Table 3.9-8 Inservice Testing Safety-Related Pumps and Valves (Continued) Safety Code Valve Test Test Class Cat. Func Para Freq SSAR l No. Oty Description th) (i) (a) (c) (d) (e) (f) Fig. (g) l F722 1 Turbine exhaust pressure 2 B P E1 5.4-8 instrument root valve sh.3 F723 1 Turbine exhaust pressure 2 B P E1 5.4-8 l instrument root valve sh.3 F724 1 Turbine exhaust pressure 2 B P El 5.4-8 between rupture disk sh.3 instrument root valve F725 1 Turbine exhaust pressure 2 B P E1 5.4-8 between rupture disk sh.3 instrument root valve G31 Reactor Water Cleanup System Valves F001 1 Line inside containment from 1 B P E1 5.4-12 RHR system maintenance sh.1 valve F002 1 CUW System suction line 1 A I,A L,P RO 5.4-12 ; inboard isolation valve S 3 mo sh.1 F003 1 CUW System suction line 1 A I,A L,P RO 5.4-12 l outboard isolation valve S 3mo sh.1 i F017 1 CUW System RPV head 1 A 1,A L,P RO 5.4-12 spray line outboard isolation S CS sh.1 valve (h3) F018 1 CUW System RPV head 1 A, C 1,A L, S RO 5.4-12 spray line inboard check sh.1 valve (h1) F019 1 CUW System bottom head 1 B P E1 5.4-12 drain line maintenance valve sh.1 F026 1 CUW System suction line 1 B P P,S RO 5.4-12 shutoff valve sh.1 F050 1 Test line off the suction line 2 B P E1 5.4-12 outboard isolation valve sh.1 G31-F003 F058 1 Test line off RPV head spray 2 B P E1 5.4-12 line outboard isolation valve sh.1 3.9 118 Mechanical Systems and Components - Amendment 37
23A6100 Rtv. 9 ABWR StandardSafetyAnalysisReport (D U Table 3.9 8 Inservice Testing Safety-Related Pumps and Valves (Continued) Safety Code Valve Test Test Class Cat. Func Para Freq SSAR No. Oty Description (a) (c) (d) (e) (f) Fig. l F025 6 Cooling water supply line to 3 B A S E2 9.2-1 HECW refrigerator PCV sh. 2,5,8 l F026 6 Cooling water supply line to 3 B P E1 9.2-1 HECW refrigerator maintenance sh. 2,5,8 valve l F027 6 Cooling water line to HECW 3 B P E1 9.2-1 refrigerator bypass line sh. 2,5,8 l F028 6 Cooling water return line from 3 B P E1 9.2-1 HECW refrigerator sh. 2,5,8 F029 2 Cooling water supply line to FPC 3 B P E1 9.2-1 Hx sh. 2,5 F030 2 Cooling water return line from FPC 3 B P E1 9.2-1 Hx sh. 2,5 F031 2 Cooling water supply line to FPC 3 B P E1 9.2-1 pump room air conditioner sh. 2,5 'v F032 2 Cooling water return line from FPC 3 B P E1 9.2-1 pump room air conditioner sh. 2,5 F033 2 Cooling water line to PCV 3 B P E1 9.2-1 Atmospheric Monitoring System sh. 2,5 clr F034 2 Return line from PCV Atmospheric 3 B P E1 9.2-1 Monitoring System c!r sh. 2,5 F035 2 Cooling water supply line to SGTS 3 B P E1 9.2-1 room air conditioner sh. 2,5 F036 2 Cooling water return line from 3 B P E1 9.2-1 SGTS room air conditioner sh. 2,5 F037 2 Cooling water supply line to FCS 3 B P E1 9.2-1 room air conditioner sh. 2,5 F038 2 Cooling water return line from 3 B P E1 9.2-1 FCS room air conditioner sh. 2,5 F039 3 Cooling water supply line to RHR 3 B P E1 9.2-1 equipment room air conditioner sh. 2,5,8 F040 3 Cooling water return line from 3 8 P E1 9.2-1 RHR equipment room air sh. 2,5,8 p conditioner F041 3 Cooling water supply line to RHR 3 B P E1 9.2-1 pump motor sh. 2.5,8 MechanicalSystems and Components - Amendment 37 3.9 123
1 23A6100 Rsv. 4 ABWR StandardSafetyAnalysis Report O Table 3.9-8 Inservice Testing Safety-Related Pumps and Valves (Continued) Safety Code Valve Test Test Class Cat. Func Para Freq SSAR No. Oty Description (a) (c) (d) (e) (f) Fig. F042 3 Cooling water return line from 3 B P E1 9.2-1 RHR pump motor sh. 2,5,8 F043 3 Cooling water supply line to RHR 3 B P E1 9.2-1 pump mechanical seals sh. 2,5,8 I F044 3 Cooling water return line from 3 8 P E1 9.2-1 RHR pump mechanical seals sh. 2,5,8 F045 1 Cooling water supply line to RCIC 3 B P E1 9.2-1 equipment room air conditioner sh.2 F046 1 Cooling water supply line from 3 B P E1 9.2-1 l RCIC equipment room air sh.2 conditioner F047 2 Cooling water supply line to HPCF 3 8 P E1 9.2-1 equipment room air conditioner sh. 5,8 F048 2 Cooling water supply line from 3 B P E1 9.2-1 ) HPCF equipment room air sh. 5,8 conditioner l F049 2 Cooling water supply line to HPCF 3 B P E1 9.2-1 pump motor bearing sh. 5,8 l F050 2 Cooling water return line from 3 8 P E1 9.2-1 l HPCF pump motor bearing sh. 5,8 l F051 2 Cooling water supply line to HPCF 3 B P E1 9.2-1 l pump mechanical seals sh. 5,8 l F052 2 Cooling water return from HPCF 3 B P E1 9.2-1 pump mechanical seals sh. 5,8 F053 2 Surge tank outlet line to HECW 3 8 P E1 9.2-1 System sh.2,5,8 l F055 6 Cooling water return line from 3 B A P 2 yr 9.2-1 Emergency Diesel Generator S 3mo sh. 5,8 F056 3 Cooling water return line from 3 B P E1 9.2-1 Emergency Diesel Generator sh.2,$,8 maintenance valve F057 2 Cooling water line to PCV 3 B P E1 9.2-1 Atmospheric Monitoring System sh. 2,5 air conditioner F058 2 Return line from PCV Atmospheric 3 B P E1 9.2-1 Monitoring System air conditioner sh. 2,5 F061 3 Cooling water line Emergency 3 B P E1 9.2-1 Diesel Generators sh. 2,5,8 3.9-124 Mechanical Systems and Components - Amendment 34
l l l 23A6100 Rsv. 9 ABWR standardsafety Analysis Report O
%J Table 3.9-8 Inservice Testing Safety Related Pumps and Valves (Continued)
Safety Code Valve Test Test h Class Cat. Func Para Freq SSAR No. Oty Description (a) (c) (d) (e) (f) Fig. F719 3 Cooling water line to DG 3 B P E1 9.2-1 instrument line sh. 2,5,8 F720 3 Return water line from DG 3 B P E1 9.2-1 instrument line sh. 2,5,8 a F721 3 Cooling water supply line to non- 3 8 .P E1 9.2-1 essential coolers FT instrument sh. 2,5,8 root valve F722 3 Cooling water supply line to non- 3 B P E1 9.2-1 essential coolers FT instrument sh. 2,5,8 1 root valve l P24 HVAC Normal Cooling Water System Valves F053 1 HNCW supply line outboard 2 A I,A L,P RO 9.2-2 isolation valve S 3mo n F054 1 HNCW supply line inboard 2 A, C I,A L,S RO 9.2-2 isolation check valve (h1)
/ ]\ \ \
F141 1 HNCW return inboard isolation 2 A 1,A L,P,S RO 9.2-2 valve (h1) F142 1 HNCW return outboard isolation 2 A i,A L,P RO 9.2-2 valve S 3mo P25 HVAC Emergency Cooling Water System Valves l F001 6 Pump discharge line check valve 3 C P S E2 9.2-3 sh.1,2,3 l F002 6 Pump discharge line maintenance 3 B P E1 9.2-3 valve sh.1,2,3 l F003 6 Refrigerator outlet line 3 B P E1 9.2-3 maintenance valve sh.1,2,3 F004 2 Maintenance valve at HECW 3 8 P E1 9.2-3 supply to MCR cooler TCV ' sh.1,2,3 F005 2 HECW supply to MCR cooler 3 B A S E2 9.2-3 Temperature Control Valve (TCV) sh.1,2,3 F006 2 Maintenance valve at HECW 3 B P E1 9.2-3
)
supply to MCR cooler TCV sh.1,2,3 F007 6 Maintenance valve at HECW 3 B P E1 9.2-3 supp!y to MCR cooler sh.1,2,3
,f\ 6 Maintenance valve at HECW D) F008 return from MCR cooler 3 8 P E1 9.2-3 sh.1,2,3 Mechanical Systems and Components - Amendment 37 3.9-12 7 \
l
23A6100 R1v. 9 ABWR standardsafety Analysis Report O Table 3.9-8 Inservice Testing Safety-Related Pumps and Valves (Continued) Safety Code Valve Test Test Class Cat. Func Para Freq SSAR No. Oty Description (a) (c) (d) (e) (f) Fig. l F009 6 Pump suction line maintenance 3 B P E1 9.2-3 valve sh.1,2,3 F010 2 TCV bypass at HECW discharge to 3 B P E1 9.2-3 MCR cooler sh.1,2,3 F011 3 Pump suction line/ discharge line 3 B P E1 9.2-3 PCV maintenance valve sh.1,2,3 F012 3 Pump suction line/ discharge line 3 B A S E2 9.2-3 PCV sh.1,2,3 F013 3 Pump suction line/ discharge line 3 B P E1 9.2-3 PCV maintenance valve sh.1,2,3 F014 3 Pump suction line/ discharge line 3 B P E1 9.2-3 PCV bypass line sh.1,2,3 F015 3 Maintenance valve at HECW 3 B P E1 9.2-3 supply to C/B Essential Electrical s h.1,2,3 Equipment Room Cooler TCV F016 3 HECW supply to C/B Essential 3 8 A S E2 9.2-3 1 Electrical Equipment Room cooler sh.1,2,3 TCV F017 3 Maintenance valve at HECW 3 B P E1 9.2-3 supply to C/B Essential Electrical sh.1,2,3 Equipment Room Cooler TCV F018 6 HECW supply to C/B Essential 3 B P E1 9.2-3 Electrical Equipment Room cooler sh.1,2,3 maintenance valve F019 6 Maintenance valve at HEC .' 3 B P E1 9.2-3 return from C/B Essential Electrical s h.1,2,3 Equipment Room Cooler F020 3 TCV bypass valve at HECW supply 3 B P E1 9.2-3 to C/B Essential Electrical sh.1,2,3 Equipment Room cooler F021 3 Maintenanco valve at HECW 3 B P E1 9.2-3 supply to DG zone cooler TCV sh.1,2,3 F022 3 HECW supply to DG zone cooler 3 B A S E2 9.2-3 TCV s h.1,2,3 F023 3 Maintenance valve at HECW 3 B P E1 9.2-3 supply to DG zone cooler TCV s h.1,2,3 F024 6 Mainter:ance valve at HECW 3 B P E1 9.2-3 supply to DG zone cooler sh.1,2,3 3.9-128 Mechanical Systems and Components - Amendment 37
23A6100 Rsv. 9 ABWR standardSafety Analysis Report V Table 3.9-8 Inservice Testing Safety-Related Pumps and Valves (Continued) Safety Code Valve Test Test Class Cat. Func Para Freq SSAR No. Oty Description (a) (c) (d) (e) (f) Fig. F025 6 Maintenance valve at HECW 3 B P E1 9.2-3 return from DG zone cooler sh.1,2,3 F026 3 TCV bypass valve at HECW supply 3 B P E1 9.2-3 to DG zone cooler sh.1,2,3 F030 3 Chemical addition tank return 3 B P E1 9.2-3 valve from HECW sh.1,2,3 F031 3 Chemical addition tank feed valve 3 B P E1 9.2-3 to HECW sh.1,2,3 F050 2 Make-up Water Purified (MUWP) 3 C A S E2 9.2-3 line to pump suction check valve sh.1,2,3 l F070 6 Pump discharge line drain valve 3 B P E1 9.2-3 sh.1,2,3 l F400 6 Pump drain line valve 3 8 P E1 9.2-3 sh.1,2,3 l F401 6 Pump bearing cooling water 3 B P E1 9.2-3 1 needle valve sh.1,2,3 F402 3 Refrigerator outlet line sample line 3 8 P E1 9.2-3 valve sh.1,2,3 l l F700 6 Pump discharge line pressure 3 B P E1 9.2-3 l Instrument line root valve sh.1,2,3 l l F701 6 FE P25-FE003 upstream 3 B P E1 9.2-3 instrument line root valve sh.1,2,3 l F702 6 FE P25-FE003 downstream 3 B P E1 9.2-3 instrument line root valve sh.1,2,3 l F703 6 Pump suction pressure instrument 3 B P E1 9.2-3 line root valve sh.1,2,3 F704 6 Pump suction / discharge line op 3 B P E1 9.2-3 instrument line root valve sh.1,2,3 Mechanical Systems and Components Amendment 37 3.9-129
l 23A6100 Rsv. 4 1 ABWR standardsafetyAnalysis neport Table 3.9-8 Inservice Testing Safety-Related Pumps and Valves (Continued) Safety Code Valve Test Test Class Cat. Func Para Freq SSAR No. Oty Description (h)(i) (a) (c) (d) (e) (f) Fig. (g) P41 Reactor Service Water System Valves F001 6 Pump discharge line check valve 3 C A S E2 9.2-7 sh.1,2,3 F002 6 Pump discharge line maintenance 3 B P E1 9.2-7 valve sh.1,2,3 F003 9 Service water intet valve to RCW 3 A A P 2 yr 9.2-7 System heat exchanger S E2 sh.1,2,3 l F004 6 Service water inlet valve to service 3 B P P 2 yr 9.2-7 water strainer sh.1,2,3 F005 9 Service water outlet valve from 3 A A P 2 yr 9.2-7 RCW heat exchanger S E2 sh.1,2,3 l F006 6 Service water strainer blowout 3 B P P 2 yr 9.2-7 valve sh.1,2,3 F007 9 Supply line from Potable Water 3 C P E1 9.2-7 check valve sh.1,2,3 l F008 9 Supply line from Potable Water 3 C P E1 9.2-7 check valve sh.1,2,3 F009 9 Supply valve from Potable Water 3 B A P 2 yr 9.2-7 System S E2 sh.1,2,3 l F010 9 RCW Hx tube side (service water 3 C P R 5 yr 9.2-7 side) relief valve sh.1,2,3 l F011 9 Bypass line around RCW Hx outlet 3 8 P E1 9.2-7 line outlet valve MOV P41-F005 sh.1,2,3 F012 9 Service water sampling valve 3 B P E1 9.2-7 sh.1,2,3 F013 6 Service water strainer outlet valve 3 A A P 2 yr 9.2-7 , S E2 sh.1,2,3 ! F014 3 Common service water strainer 3 A A P 2 yr 9.2-7 outlet valve S E2 sh.1,2,3 F015 3 Discharge line to discharge canal 3 A A P E1 9.2-7 MOV S E2 sh.1,2,3 F501 9 RCW Hx shcIl side drain valve to 3 B P E1 9.2-7 SWSD sh.1,2,3 F502 9 RCW Hx shell side vent valve to 3 B P E1 9.2-7 SWSD sh.1,2,3 i F503 9 RCW Hx shell side drain valve to 3 B P E1 9.2-7 SWSD sh.1,2,3 3.9-130 Mechanical Systems and Components Amendment 34
i i 1 23A6100 Rtv. 9 l ABWR standardSafetyAnalysis Report l [
\
J l Table 31-18 Radiation Environment Conditions inside Reactor Building Design Basis Accident Conditions (Outside Secondary Containment) l l LOCA Dose Rate Integrated Dose l l l Gamma Gamma l Plant Zone / Typical Equipment Accident (Gy/h) Beta (Gy/h) (Gy) Beta (Gy) t i Clean zone outside secondary 15.6.5 8E-5 2E-3 2E-2 3E-1 i containment area (not otherwise noted) [ Figs. 6.2-26/6.7-1] 4 Monitor room [ Figs. 1.2-8/6.5-11 15.6.5 8E-5 2E-3 2E-2 3E-1 2
- 1. Integrated dose is summed over a six month period for Accident Case 15.6.5.
I i l Table 31-19 Radiation Environment Conditions inside Control Building Design Basis Accident Conditions 4 b-.
\
l LOCA Dose Rate Integrated Dose 1 Gamma Gamma Plant Zone / Typical Equipment Accident (Gy/h) Beta (Gy/h) (Gy) Beta (Gy) Main Control Room and Process 15.6.5 1.0E-3 1.0E-2 8.0E-2 2.0E+0 i computer Room 4 l l HVAC Rooms, Level 17150mm 4 15.6.5 5.0E-3 to 5.0E-2 3.0E+1 to 3.0E+2 1.0E-12 1.0E +22 4 All Other Areas 15.6.5 5.0E-3 to 5.0E-2 3.0E+1 to 3.0E+2 5.0E-23 3.0E+2 3 1
- 1. Integration dose is summed over a six month period for Accident Case 15.6.5.
- 2. Highest dose rates closer to the CR HVAC Filter Units. i
- 3. Highest dose rates closer to the RBCW Units and HVAC intake units. l
- 4. Refer to Figure 1.2-15.
O l Equipment Qualification Environmental Design Criteria Amendment 37 3117
i l I 23A6100 Rsv. 5 ABWR StandardSafetyAnalysis Report O l l TMSL 21300 l a-1 l b-1 l 1 TMSL 1230 I l 1 l a-3 b-4 a-1 I b-2 l a-2 ~~~.............. ,,, TMSL -1100 b-3 TMSL -6600 l TMSL -8200 l a Thermodynamic Environment Zone l b Radiation Environment Zone TMSL - In minimeters O Figure 31-1 Zones in Primary Containment Vessel 31-1 8 Equipment Qualification Environmental Design Criteria - Amendment 3S
23A6100 Rsv. 4 ABWR StandardSafetyAnsfysis Report O V a Makeup Water (Condensate) System upstream of the injection valve for the purpose of providing the piping keep-fill water source and a filling and flushing water source. The MUWC System is discussed in Section SMA.11. m High Conductivity Waste System for drainage is located downstream of CST suction check valve. HCW is discussed in Section SMA.13. m Reactor Core Isolation Cooling System shares common CST suction. The CST suction has open communication to the CST, and the CST is vented to atmosphere; this line cannot be pressurized and was not practical to upgrade to the URS design pressure. ; a Suppression Pool Cleanup System shares common CST suction. The CST suction has open communication to the CST, and the CST is vented to atmosphere; this line cannot be pressurized and was not practical to upgrade to the URS design pressure. 3MA.4.3 Upgraded Components - RCIC System A detailed listing of the components upgraded for the RCIC System follows, including ) , identification of those interfacing system components not requiring upgrade. l REACTOR CORE ISOLATION COOLING SYSTEM, SSAR Figure 5.4-8, Sheets 1 & 3. RCIC pump suction piping Press./ Temp./ Design / Reference Components Scismic Class Remarks i Sheet 1 200A RCIG001-W Pipe 2.82 MPaG,77 C,3B,As Was 1.37 MPaG l 20A-RCIG703-W Pipe 2.82 MPaG,77 C,3B,As Was 1.37 MPaG 20A-RCIC-F701 Valve 2.82 MPaG,77 C,3B,As Was 1.37 MPaG 20A-RCIGPX015 P. Test 2.82 MPaG,77 C,3B,As Was 1.37 MPaG 200A-RCIGD001 Strainer 2.82 MPaG,77*C,3B,As Was 1.37 MPaG 20A-RCIC 700-W Pipe 2.82 MPaG,77 C,3B,As Was 1.37 MPaG 20A-RCIC-F700 Valve 2.82 MPaG,77 C,3B,As Was 1.37 MPaG 20A-RCIC-PT001 P.Trans 2.82 MPaG,77 C,3B,As Was 1.37 MPaG 20A-RCIG701-W Pipe 2.82 MPaG,77 C,3B,As Was 1.37 MPaG 20A-RCIC-702-W Pipe 2.82 MPaG,77 C,3B,As Was 1.37 MPaG 20A-RCIGPI003 P.Ind. 2.82 MPaG,77 C,3B,As Was 1.37 MPaG 20A-RCIGPT002 P.Trans 2.82 MPaG,77 C,3B,As Was 1.37 MPaG 50A-RCIG018-W Pi e 2.82 MPaG,77 C,3B.As Was 1.37 MPaG 50A-RCIGF017 R. alve 2.82 MPaG,104 C,3B,As Was 1.37 MPaG 200A-RCIGF002 T. Check 2.82 MPaG,77 C,3B,As Was 1.37 MPaG 200A-RCIGF060 Valve 2.82 MPaG,77 C,3B,As Was 1.37 MPaG
- 200A-RCIGF001 MO Valve 2.82 MPaG,77 C,3B,As Was 1.37 MPaG
** 200A-RCIG005-W Pipe 2.82 MPaG,77 C,3B,As Was 1.37 MPaG
** 200A-RCIGF007 Check V. 2.82 MPaG,77 C,3B,As Was 1.37 MPaG
** 20A-RCIG025-W Pipe 2.82 MPaG,77 C,3B,As Was 1.37 MPaG G Sheet 1
** 20A-RCIGF027 T.\alve
** 200A-RCIGF006 MO Valve 2.82 MPaG,77 C,3B,As 2.82 MPaG,104 C,3B,As Was 1.37 MPaG Was 1.37 MPaG I
System Evaluation For ISLOCA - Amendment 34 3MA 13
23A6100 R:v. 9 ABWR standard sainy Anstysis soport O'l
- IIPCF Interface Piping 200A-HPCF-015-S,1.37 MPaG,66 C,B (S1,S2), As (open pathway to Con-densate Storage Tank mth LO valves).
- Suction Piping from Suppression Pool Interface 200A-RCIC-004-W,0.310 MPaG,104*C, SB, As.
RCIC discharge from relief valves and test line valve direct to the suppression pool without restric-tion. Press./ Temp./ Design / Reference Components Seismic Class Remarks Sheet 1 50A-RCIG009-W Pipe 0.310 MPaG, 104*C,3B,As No change 50A-RCIC-019-W Pipe 0.310 MPaG,104 C,3B,As No change 100A-RCIC-007-W Pipe 0.310 MPaG,104*C,3B,As No change 250A-RHR-008 Pipe 0.310 MPaG,104*C,3B As No change Sheet 1 Suppression Pool ABWR High Press. Core Flooder System, SSAR Figure 6.3-7, components interfacing with RCIC System are not upgraded because this is the open pathway to the condensate storage tank vented to the atmosphere. Press./ Temp./ Design / Reference Components Seismic Class Remarks Sheet 1 200A-HPCF-015-W Pipe 1.37 MPaG, 66 C,B (S1,S2), As No change 400A-HPCF-105-W Pipe 1.37 MPaG, 66 C,B (S1,S2), As No change ' 500A-IIPCF-004-W Pipe 1.37 MPaG, 66 C,B (S1,S2), As No change 300A-HPCF-001-W Pipe 1.37 MPaG, 66 C,B (S1,S2), As No change 300A-HPCF-002-W Pipe 1.37 MPaG, 66 C,B (S1,S2), As No change 300A-HPCF-003-W Pipe 1.37 MPaG, 66 C,B (S1,S2), As No change ABWR Makeup Water System (Condensate), SSAR Figure 9.2-4, components interfacing with HPCF System are not upgraded due to the open pathway to the condensate storage tank vented to the atmosphere. Press./ Temp./ Design / Reference Components Seismic Class Remarks Sheet 1 300A-MMVCF100 Valve 1.37 MPaG, 66 C,B (S1,S2), As No change 300A-MmVGF101 Valve 1.37 MPaG, 66 C,B (S1,S2), As No change 300A-MmVC-F102 Valve 1.37 MPaG, 66 C,B (S1,S2), As No change 300A-MmVC-100 Pipe Static Hd, 66 C,B (S1,S2), As No change 300A-MUWC101 Pipe Static Hd, 66 C,B (S1,S2), As No change 300A-MmVC-102 Pipe Static Hd, 66 C,B (S1,S2), As No change Condensate Storage Tank, 66'C,4D, Non-seismic No change RCIC turbine condensate piping to the suppression pool Press./ Temp./ Design / Reference Components Seismic Class Remarks Sheet 3 250A-RCIC 037-S Pipe 8.62 MPaG,302*C,3B,As Was 0.981 MPaG 20A-RCIC-720-S Pipe 8.62 MPaG,302*C,3B,As Was 0.981 MPaG 20A-RCIC-F722 Valve 8.62 MPaG,302*C,3B,As Was 0.981 MPaG 20A-RCIGPIO12 P.Ind. 8.62 MPaG,302 C,3B,As Was 0.981 MPaG 3MA-14 System Evaluatior: For ISLOCA - Amendment 37
23A6100 Rxv. 9 ABWR standedsarny Analysis neport O b l 350A-RCIC-Cond. Pot 8.62 MPaG,302 C,3B,As Was 0.981 MPaG Press./ Temp./ Design / Reference Components Seismic Class Remarks 350A-RCIG038-S Pipe 8.62 MPaG,302 C,3B,As Was 0.981 MPaG l 20A-RCIC-721-S Pipe 8.62 MPaG,302*C,3B,As Was 0.981 MPaG 20A-RCIGF723 Valve 8.62 MPaG,302 C,3B,As Was 0.981 MPaG 20A-RCIG722-S Pipe 8.62 MPaG,302 C,3B,As Was 0.981 MPaG 20A-RCIC PT013A P.Trans 8.62 MPaG,302*C,3B,As Was 1.37 MPaG 20A-RCIGPT013E P.Trans 8.62 MPaG,302 C,3B,As Was 1.37 MPaG
** 25A-RCIC-051-S Pipe 8.62 MPaG,302'C,3B,As Was 0.981 MPaG
** 25A-RCIC-F051 Valve 8.62 MPaG,302 C,3B,As Was 0.981 MPaG
** 25A-RCIC-D012 Strainer 8.62 MPaG,302'C,3B,As Was 0.981 MPaG
** 25A-RCIGD013 S. Trap 8.62 MPaG,302*C,3B,As Was 0.981 MPaG
** 25A-RCIGF052 Valve 8.62 MPaG,302*C,3B,As Was 0.981 MPaG Sheet 3 ** 25A-RCIG052-S Pipe 2.82 MPaG,184*C,4D,As Was 0.981 MPaG Sheet 1 350A-RCIGF038 Check 8.62 MPaG,302 C,3B,As Was 1.37 MPaG 20A-RCIC053-S Pipe 8.62 MPaG,302 C,3B,As Was 0.981 MPaG f't N 20A-RCIGF053 T. Valve 8.62 MPaG,302 C,3B,As Was 0.981 MPaG 350A-RCIGF0f9 Valve 8.62 MPaG,302 C,3B,As Was 0.981 MPaG
' \ A-RCIGF069 T. Valve 2.82 MPaG,184*C,3B,As Was 10.981 MPaG 350A-RCIC-039-S Pipe 0.981 MPaG,184 C,3B,As No change Sheet 1 Suppression Pool
- Vent via Rupture Disks.
** RCIC Turbine Condensate Piping to the Barometric Condenser.
RCIC vacuum tank condensate piping to the suppression pool. Press./ Temp./ Design / Reference Components Seismic Class Remarks j Sheet 3 50A-RCIC-Vacuum Pump 2.82 MPaG,121*C,4D,As Was 0.755 MPaG ; 50A-RCIG044-S Pipe 2.82 MPaG, 88'C,4D,As Was 0.310 MPaG l 50A-RCIG067-S Pipe 2.82 MPaG, 88 C,4D,As Was 0.310 MPaG , 50A-RCIC PCV Valve 2.82 MPaG,121*C,4D,As Was 0.755 MPaG l Sheet 3 20A-RCIG068-S Pipe 2.82 MPaG,121 C,4D,As Was 0.981 MPaG l Sheet 1 50A-RCIC-F046 Check V. 2.82 MPaG,104*C,3B,As Was 0.310 MPaG 20A-RCIC-057-S Pipe 2.82 MPaG,104 C,3B,As Was 0.310 MPaG 20A-RCIGF059 T. Valve 2.82 MPaG,104 C,3B,As Was 0.310 MPaG 50A-RCIGF047 MOValve 2.82 MPaG,104*C,3B,As Was 0.310 MPaG 4 50A-RCIG0454 Pipe 0.981 MPaG,104 C,3B As No change Sheet 1 Suppression Pool RCIC steam drains from trip and throttle valve piping and turbine to condensate chamber. O Press./ Temp./ Design / Reference Components (l Sheet 3
- 20A-RCIG063-S Pipe Seismic Class 8.62 MPaG,302 C,3B,As Remarks Was 0.981 MPaG System Evaluation ForISLOCA - Amendment 37 3MA 15
23A6100 R1v. 9 ABWR standard Safety Anslysis Report O
- 20A-RCIG061-S Pipe 8.62 MPaG,302*C,3B,As Was 0.981 MPaG
** 20A-RCIG064-S Pipe 8.62 MPaG,302*C,3B,As Was 0.981 MPaG RCIC Trip and Throttle Valve leakoffs are piped to Condensing Chamber.
RCIC Turbine Condensate Drain connects to the Condensing Chamber RCIC turbine valve leakoffs are piped to the barometric condenser Press./ Temp./ Design / Reference Components Seismic Class Remarks Sheet 3
- 25A-RCIG0584 Pipe 2.82 MPaG,184 C,4D,As Was 0.981 MPaG
** 25A-RCIG059-S Pipe 2.82 MPaG,184*C,4D,As Was 0.981 MPaG Barometric Condenser 2.82 MPaG,184*C,4D,As Was 0.755 MPaG
*** 25A-RCIG065-S Pipe 2.82 MPaG,184 C,4D,As Was 0.755 MPaG 25A-RCIGReliefValve 2.82 MPaG,121 C,4D,As Was 0.755 MPaG 25A-RCIG066-S Pipe 0 MPaG,121*C,4D,As No change RCIC Trip and Throttle Valve Stem leakoffis piped to the Barometric RCIC Turbine Governor Valve Stem is piped to the to Barometdc Condenser. l
- " Barometric Condenser Press. relief and piping.
RCIC pump cooling water piping for pump and turbine lube oil coolers l Press./ Temp./ Design / l Reference Components Seismic Class Remarks l Sheet 3 50A-RCIG011-W Pipe 2.82 MPaG,77 C,3B,As Was 0.863 MPaG ; 50A-RCIG028-W Pipe 2.82 MPaG,77 C,3B,As Was 0.863 MPaG l 50A-RCIC-F030 Relief V. 2.82 MPaG,77'C,3B,As Was 0.863 MPaG l 50A-RCIG029-W Pipe 2.82 MPaG,77 C,3B,As Was 0.863 MPaG l 20A-RCIC-713-W Pipe 2.82 MPaG,77*C,3B,As Was 0.863 MPaG l 20A-RCIC-PX018 Press 2.82 MPaG,77 C,3B,As Was 0.863 MPaG l 50A-RCIGTurb.LO Cooler 2.82 MPaG, 77'C,3B,As Was 0.863 MPaG ! 50A-RCIGPump LO Cooler 2.82 MPaG,77 C,3B,As Was 0.863 MPaG 15A-RCIGTX019 Temp.Pt. 2.82 MPaG,77'C,3B,As Was 0.863 MPaG 20A-RCIC-714-W Pipe 2.82 MPaG,77 C,3B,As Was 0.863 MPaG 20A-RCIGF714 Valve 2.82 MPaG,77"C,3B As Was 0.863 MPaG 20A-RCIGPX020 Press.Pt. 2.82 MPaG,77'C,3B,As Was 0.863 MPaG 15A-RCIG012-W Pipe 2.82 MPaG,77*C,3B,As Was 0.863 MPaG 15A-RCIC-013-W Pipe 2.82 MPaG,77 C,3B,As Was 0.863 MPaG 15A-RCIC-014-W Pipe 2.82 MPaG,77 C,3B,As Was 0.863 MPaG 15A-RCIC-015-W Pipe 2.82 MPaG,77 C,3B,As Was 0.863 MPaG Sheet 3 Barometdc Condenser 2.82 MPaG,121 C,4D,As Was 0.755 MPaG RCIC vacuum tank and condensate pump piped to RCIC pump suction pipe. Press./ Temp./ Design / Reference Components Seismic Class Remark Sheet 3 RCIC Vacuum Tank 2.82 MPaG,77 C,4D,As Was 0.755 MPaG RCIC Press. Switch H 2.82 MPaG,121*C,4D,As Was 0.755 MPaG RCIC Level Switch H 2.82 MPaG,121 C,4D,As Was 0.755 MPaG RCIC Level Switch L. 2.82 MPaG,121*C,4D,As Was 0.755 MPaG RCIC Cond. Pump 2.82 MPaG, 88 C,4D,As Was 1.37 MPaG 50A-RCIC-F014 Check V. 2.82 MPaG, 88 C,4D,As Was 1.37 MPaG 50A-RCIC-016-W Pipe 2.82 MPaG, 88 C,4D,As Was 1.37 MPaG 20A-RCIC-715-W Pipe 2.82 MPaG, 88*C,4D,As Was 1.37 MPaG 3MA 16 System Evaluation For ISLOCA - Amendment 37
23A6100 Rsv. 9 ABWR standardsaretyAnalysis neport I 5.4.6.1.1.2 Isolation Isolation valve arrangements include the following: (1) Two RCIC lines penetrate the reactor coolant pressure boundary (RCPB). The first is the RCIC steamline, which branches off one of the main steamlines between the reactor vessel and the MSIVs.This line has two automatic motor-operated isolation valves, one located inside and the other outside the drywell. An automade motor-operated inboard RCIC isolation bypass valve is used. The isolation signals noted earlier close these valves. (2) The RCIC pump discharge line is the other line that penetrates the RCPB, which directs flow into a feedwater linejust outboard of the primary containment. This line has a testable check valve and an automatic motor-operated valve located outside primary containment. (3) The RCIC turbine exhaust line also penetrates the containment. Containment penetration is located about a meter above the suppression pool maximum water level. A vacuum brea king line with two vacuum breakers in series runs ( ( in the suppression pool air space and connects to the RCIC turbine exhaust line inside the containmeat. Located outside the containment in the turbine exhaust line is a remote-manually controlled motor operated isolation valve. 1 i (4) The RCIC pump suction line, minimum flow pump discharge line, and } turbine exhaust line penetrate the containment and are submerged in the l suppression pool. The isolation valves for these lines are outside the ' { containment and require automatic isolation operation, except for the i turbine exhaust line which has remote manual operation. The RCIC System design includes interfaces with redundant leak detection devices, monitoring: (1) A high pressure drop across a flow device in the steam supply line equivalent to 300% of the steady-state steam flow at 8.22 MPaA pressure. (2) A high area temperature utilizing temperature switches as described in the leak detection system (high area temperature shall be alarmed in the control room). (3) A low reactor pressure of 0.34 MPaG minimum. sl (4) A high pressure in the RCIC turbine exhaust line. Component and Subsystem Design - Amendment 37 5.4 21
83A6100 Rev. 5 ABWR stanitard satery Analysisneport These devices, activated by the redundant power supplies, automatically isolate the O steam supply to the RCIC turbine and trip the turbine. The HPCF System provides redundancy for the RCIC System should it become isolated. 5.4.6.1.2 Reliability, Operability, and Manual Operation 5.4.6.1.2.1 Reliability and Operability The RCIC System (Table 3.2-1) is designed commensurate with the safety importance of the system and its equipment. Each component is individually tested to confinn compliance with system requirements. The system as a whole is tested during both the startup and pre-operational phases of.the plant to set a base mark for system reliability. To confirm that the system maintains this mark, functional and operability testing is performed at predetermined intervals throughout the life of the plant. A design flow functional test of the RCIC System may be performed during normal plant operation by drawing suction from the suppression pool and discharging through a full flow test return line to the suppression pool. All components of the RCIC System are capable ofindividual functional testing during normal plant operation. System control provides automatic return from test to operating mode if system initiation is required, and the flow is automatically directed to the vessel (Subsection 5.4.6.2.4). l See Subsection 5.4.15.2 for COL license information. 5.4.6.1.2.2 Manual Operation In addition to the automatic operational features, provisions are included for manual startup, operation, and shutdown of the RCIC System in the event initiation or shutdown signals do not exist or the control room is inaccessible. 5.4.6.1.3 Loss of Offsite Power The RCIC System power is derived from a reliable source that is maintained by either onsite or offsite power. 5.4.6.1.4 Physical Damage The system is designed to the requirements presented in Table 3.2-1 commensurate with the safety importance of the system and its equipment. The RCIC System is physically located in a different quadrant of the reactor building and utilizes different divisional power and separate electrical routings than its redundant system (Subsections 5.4.6.1.1.1 and 5.4.6.2.4). 5.4.6.1.5 Environment The RCIC System operates for the time intervals and the emironmental conditions specified in Section 3.11. 5.4-22 Component and Subsystem Design - Amendment 35
23A6100 Rsv. 9 i ABWR stuhntsafetyAulysisReport b value, controls transient acceleration, and positions the turbine governor valve as required to maintain constant pump discharge flow over the pressure range of the system. Input to the turbine governor is from the flow controller monitoring the pump discharge flow. During standby conditions, the flow controller output is saturated at its maximum value. When the RCIC System is shut down, the low signal select feature of the turbine control system selects the idle setting of a speed ramp generator. The ramp generator output signal during shutdown corresponds to the low limit step and a turbine speed demand of 73.3 to 104.7 rad /s. 1 On RCIC System startup, bypass valve F045 (provided to reduce the frequency of ' turbine overspeed trips) opens to accelerate the turbine to an initial peak speed of .) approximately 157 rad /s; now under governor control, turbine speed is returned to the l low limit turbine speed demand of 73.3 rad /s to 104.7 rad /s. After a predetermined ' delay (5 to 10 s), the steam supply valve leaves the full closed position and the ramp l generator is released. The low signal select feature selects and sends this increasing l
- ramp signal to the governor. The turbine increases in speed until the pump flow satisfies the controller setpoint. Then the controller leaves saturation, responds to the input error, and integrates the output signal to satisfy the input demand.
The operator has the capability to select manual control of the governor, and adjust speed and flow (within hardware limitations) to match decay heat steam generation during the period of RCIC operation. The RCIC pump delivers the makeup water to the reactor vessel through the feedwater line, which distributes it to obtain mixing with the hot water or steam within the reactor - vessel. , The RCIC turbine will trip automatically upon receipt of any signalindicating turbine i overspeed, low pump suction pressure, high turbine exhaust pressure, or an aut& isolation signal. Automatic isolation occurs upon receipt of any signal indicating: (1) A high pressure drop across a flow device in the steam supply line equivalent to 300% of the steady-state steam flow at 8.22 MPaA. i (2) A high area temperature. (3) A low reactor pressure of 0.34 MPaG minimum. l (4) A high pressure in the turbine exhaust line. ( The steam supply valve F037, steam supply bypass valve F045 and cooling water supply valve F012 will close upon receipt of signal indicating high water level (Level 8) in the reactor vessel. These valves will reopen (auto restart) should an indication oflow water Component and Subsystem Design - Amendment 37 5 4-27
23A6100 R:v. 8 ABWR StandardSafetyAnalysis Report l l Icvel (Level 2) in the reactor vessel occur. Water Level 2 automatically resets the water level trip signal. The RCIC System can also be started, operated, and shut down remote-manually provided inidation or shutdown signals do not exist. I 5.4.6.2.5.3 Test Mode A design functional test of the RCIC System may be performed during normal plant operation by drawing suction from the suppression pool and discharging through a full flow test return line back to the suppression pool. The discharge valve to the vessel remains closed during test mode operation. The system will automadcally return from test to operating mode if system initiation is required and the flow will be automatically directed to the vessel. 5.4.6.2.5.4 Limiting Single Failure The most limiting single failure with the RCIC System and its HPCF system backup is the failure of HPCF. With an HPCF failure,if the capacity of the RCIC System is adequate to maintain reactor water level, the operator shall follow Subsection l 5.4.6.2.5.2. However, if the RCIC capacity is inadequate, Subsection 5.4.6.2.5.2 still applies. Additionally, the operator may initiate the ADS described in Subsection 6.3.2.2.2. 5.4.6.3 Performance Evaluation The analytical methods and assumptions in evaluating the RCIC System are presented in Chapter 15 and Appendix 15A. The RCIC System provides the flows required from the analysis (Figure 5.4-9) within a 30 second interval based upon considerations noted in Subsection 5.4.6.2.4. 5.4.6.4 Preoperational Testing The preoperational and initial startup test program for the RCIC System is presented in Chapter 14. 5.4.7 Residual Heat Removal System Evaluations of the Residual Heat Removal (RHR) System against the applicable General Design Criteria (GDC) are provided in Subsections 3.1.2 and 5.4.7.1.4. 5.4.7.1 Design Basis l The RHR System is composed of three electrical and mechanical independent disisions designated A, B, and C. Each division contains the necessary piping, pumps, valves and heat exchangers. In the low pressure flooder mode, suction is taken from the suppression pool and iniected into the vessel outside the core shroud (sia the feedwater line on Division A and sia the low pressure flooder subsystem discharge return line on Divisions B and C). S.4-28 Component and Subsystem Design - Amendment 36
l 22A6100 Rtv. 9 ABWR standussafety Analysis Report l 1 V) Table 5.4-2 Design Parameters for RCIC System Components (1) RCIC Pump Operation (C001) Flow rate Injection flow-182 m 3/h Cooling water flow-4 to 6 m3/h Total pump discharge - 188 m3/h (includes no margin for pump wear) Water temperature range 10' to 60*C, continuous duty j 40' to 77'C, short duty l NPSH 7.3m minimum ) Developed head 900m at 8.22 MPaA reactor pressure 186 m at 1.14 MPaA reactor pressure I Maximum pump 675 kW at 900m developed head , shaft power 125 kW at 186m developed head l Design pressure 11.77 MPaG l (2) RCIC Turbine Operation (C002) High Pressure Condition Low Pressure Condition l Reactor pressure 8.19 MPaA 1.14 MPaA N (saturated temperature) Steam inlet pressure 8.12 MPaA, 1.03 MPaA, minimum minimum Turbine exhaust 0.11 to 0.18 MPaA, 0.11 to 0.18 MPaA, pressure maximum maximum Design inlet pressure 8.62 MPaG at saturated temperature l Design exhaust pressure 8.62 MPaG at saturated temperature (3) RCIC leakoff orifices Sized for 3.2 mm diameter minimum to 4.8 mm (D017, D018) diameter maximum Flow element (FE007) Flow at full meter 250 m3/h differential pressure Normal temperature 10 to 77'C System design 8.62 MPaG/302'C pressure / temperature Maximum unrecoverable 0.031 MPa loss at normal flow Installed combined accuracy 2.5% at normal flow and normal (Flow element, Flow s
\j transmitter and Flow indicator)
Component and Subsystem Design - Amendment 37 S.4-61
23A6100 Rov. 4 ABWR standardSafetyAnalysis Report O Table 5.4-2 Design Parameters for RCIC System Components (Continued) (4) Valve Operation Requirements Steam supply valve (F037) Open and/or close against full differential pressure l of 8.12 MPa within 15 seconds Pump discharge valve Open and/or close against full differential pressure l (F004) of 9.65 MPa within 15 seconds Pump minimum flow bypass Open and/or close against full differential pressure l valve (F011) of 9.65 MPa within 5 seconds RCIC steam isolation valve Open and/cr close against full differential pressure l (F035&F036) of 8.12 MPa within 30 seconds Cooling water pressure control Self-contained downstream sensing control valve j valve (F013) capable of maintaining constant downstream l pressure of 0.52 MPa l Pump suction relief 1.48 MPaA setting; . ' h at 10% accumulation i valve (F017) Cooling water relief Sized to prevent , ssuring piping, valves, and valve (F030) equipment in the . loop in the event of failure of pressure controi v ..a F313 Pump test return valve Capable of throttling control against differential (F008) pressures up to 7.58 MPa and closure against differential pressure at 9.65 MPa j Pump suction valve, Capable of opening and closing against 1.37 MPa suppression pool (F006) differential pressure j Testable check valve Open and/or close against full differential pressure l , l equalizing valve (F026) of 8.12 MPa j Outboard check valve Accessible during plant operation and capable of j (F005) local testing l Turbine exhaust isolation Opens and/or closes against 1.10 MPa differential valve (F039) pressure at a temperature of 170 C, physically located in the line on a horizontal run as close to the containment as practical ; Isolation valve, steam warmup Opens and/or closes against differential pressure of line (F048) 8.12 MPa Barometric condenser These valves operate only when RCIC System is ! condensate drain Line shutdown, allowing drainage to CUW System and isolation valves they must operate against a differential pressure of (F031 & F032) 0.52 MPa Condensate storage tank This valve isolates the condensate storage tank so isolation valve (F001) that suction may be drawn from the suppression pool; valve must operate against a differential l pressure of 1.37 MPa 5.4 62 Com;>onent and Subsystem Design - Amendment 34 l
23A6100 Rxv. 4 ABWR standardsafetyAnalysisReport from the drywell is completed, the wetwell and, subsequently, the dowell pressure peak occurs as the volumetric compression is completed and the pool volume begins to decrease due to the drawdown effects of the ECCS flow. Since the suppression pool volume continues to decrease as the ECCS flow continues, the short-term pressure peak is the peak pressure for the transient. 6.2.1.1.3.3.1.4 Long-Term Accident Responses In order to assess the adequacy of the containment system following the initial blowdown transient, an analysis was made of the long-term temperature and pressure response following the accident. The analysis assumptions are those discussed in Subsection 6.2.1.1.3.3.1.2. l The short-term pressure peak (268.7 kPaG) of Figure 6.2-6 is the peak pressure for the whole transient. Figure 6.2-8 shows temperature time histories for the suppression pool, wetwell, and drywell temperatures. The peak pool temperature (96.9 C) is reached at 15,350 seconds (4.264 hours) and remains below the 97.2*C limit. 6.2.1.1.3.3.2 Main Steamline Break A schematic of the ABWR main steamlines, with a postulated break in one of the main steamlines, is shown in Figure 6.2-9. The main steamline (hiSL) break is a double-ended l j break with one end fed by the RPV directly through the broken line, and the other fed by the RPV through the unbroken main steamlines until the hfSIVs are closed. Once the hiSIVs are closed, the break flow is only from the RPV through the broken line. The effective break area used for the hiSL is shown in Figure 6.210. hfore detailed descriptions of the hiSL break model are provided in the following: (1) Each MSL contains a flow limiter built into the MSL nozzle on the RPV with a throat area of 0.0983m 2, as shown in Figure 6.2-9. (2) The break is located in one MSL at the inboard hfSIV. (3) During the inventory depletion period, the flow multiplier of 0.75 is applied (Reference 6.2-1). (4) The flow resistance of open MSIVs is considered. A conservative value of 2.062 for pressure loss coefficient for two open MSIVs was taken. The nominal value is approximately 3.0. When the open MSIV resistance is considered, the flow chokes at the MSIV on the piping side as soon as the inventory depletion period ends. The effective flow area on the piping side reduces to 70% of a ( frictionless piping area. The value of 70% applies to flow of steam and two- \ phase mixture with greater than 15% quality. Containment Systems - Amendment 34 6.2 11
23A6100 Rtv. 9 ABWR StandardSafety Analysis Report O This assumption is quite consenative because all other resistances in piping are ignored and the flow in the steamline within a one to two second period is either all steam or a two-phase mixture of much greater than 15% quality. (5) MSIVs are completely closed at a conservative closing dme of 5.5 seconds (0.5 seconds greater than the maximum closing time plus instrument delay), in order to maximize the break flow. 6.2.1.1.3.3.2.1 Assumptions for Short-Term Response Analysis The response of the reactor coolant system and the containment system during the short-term blowdown period of the MSLB accident is analyzed using the assumptions listed in the above subsection and Subsection 6.2.1.1.3.3.1.1 for the feedwater line break, with the following exceptions: (1) The vessel depressurization flow rates are calculated using the Moody's HEM for the critical break flow. (2) The turbine stop valve closes at 0.2 second. This determines how much steam flows out of the RPV, but does not affect the inventory depletion time on the piping side. (3) The break flow is saturated steam if the RPV collapsed water level is below the MSL elevadon; othenvise, the flow quality is the vessel average quality. This case provides the limiting dnwell temperature. Another case was evaluated with the assumption that the two-phase level swell would reach the main steam nozzle in one second, thereby changing the flow quality to the RPV average quality after one second. This case prosides a higher drywell pressure but a lower dqwell temperature than the first assumption. (4) The feedwater mass flow rate for a MSL break was assumed to be 130% NBR for 120 seconds. This is a standard MSL break containment analysis assumption based on a conservative estimate of the total available feedwater inventory and the maximum flow available from the feedwater pumps with discharge pressure equal to the RPV pressure. The feedwater enthalpy was calculated as described for the FWL break (Subsection 6.2.1.1.3.3.1.1) for 130% NBR flow, and is shown in Figure 6.2-11. (5) The SRVs are not actuated. O 6.2 12 Containment Systems - Amendment 37
23A6100 Rsv. 9 ABWR standardsarery Analysis Report Table 6.2 9 Secondary Containment Penetration List (Continued) Penetration Name Elevation Diameter Number (mm) (mm) 32 RCW (C) -1700 100 33 RCW (C) 1700 200 34 RCW (C) -1700 200 35 HS 4800 150 36 MS 4800 80 37 LCW (FPC) 4800 150 38 LCW (CUW) 4800 150 39 RCIC 4800 50 40 MS (4) 16191 700 41 FDW (2) 13810 600 42 HVAC Exhaust 27200 t 43 HVAC Supply 31700 i O
- 44 Controlled Access (2) 12300 45 Equipment Lock 12300
- 46 Railroad Car Door 12300
- 47 HS 12300 150 l 48 Deleted l 49 Deleted 50 HNCW 12300 200 51 HNCW 12300 200 52 MUWP 4800 150 53 AC 4800 50 54 AC 4800 250 l 55 Deleted l 56 Deleted 57 Cabletrays 23500 58 Cabletrays 12300 59 Cabletrays 4800
' This table is provided in response to Question 430.34.
t These HVAC openings have safety related isolation valves with both local monitoring and remote (in control room) monitoring.
- These doors are monitored in the control room as per Subsection 13.6.3.4.
Containment Systems - Amendment 37 6.2-183
23A6100 Rtv. 4 ABWR standardsafetyAns/ysisReport O Table 6.2-10 Potential Bypass Leakage Paths' Penetration Diameter Termination Leakage Potential Number Name (mm) Regior, t Barriers
- Bypass Path X-1 U/D Equipment Hatch 2600 S C/M-J No X-2 U/D Personnel Hatch 2400 S C/M-J No X-3 ISI Hatch 200 S C/M-J No X-4 Wetwell Access Hatch 2000 S C/M-J No X-5 L/D Personnel Hatch 2400/5000 S C/M-J No X-6 L/D Equipment Hatch 2400/5000 S No C/M-J X-10A Main Steamline 1200 E E/D/G Yes X-10B Main Steamline 1200 E E/D/G Yes X 10C Main Steamline 1200 E E/D/G Yes X-10D Main Steamline 1200 E E/D/G Yes X-11 Main Steamline Drain 500 E E/D/G Yes X-12A Feedwater Line 950 E E/D/L Yes X-12B Feedwater Line 950 E E/D/L Yes X-22 Borated Water injection 450 S E/C/L No X-308 Drywell Spray 200 S E/C/L No X 30C Drywell Spray 200 S E/C/L No l X-31 A HPCF (B) 600 S E/C/L No l X-31B HPCF (C) 600 S E/C/L No X-328 LPFL (B) RHR (B) 650 S E/C/L No X-32C LPFL (C) RHR (C) 650 S E/C/L No X 33A RHR Suction (A) 750 S C/L No X-33B RHR Suction (B) 750 S C/L No X-33C RHR Suction (C) 750 S C/L No X-37 RCIC Turbine Steam 550 S C/G No !
X-38 RPV Head Spray 550 S E/C/L No X-50 CUW Pump Feed 600 S E/C/L No X-60 MUWP Suction 200 S C/L No l X-61 RCW Suction (A) 200 E E/D/H No X-62 RCW Return (A) 200 E E/D/H No X-63 RCW Suction (B) 200 E E/D/H No X-64 RCW Return (B) 200 E E/D/H No 6.2 184 Containment Systems - Amendment 34
- .~ _.
l l 23A6100 R:v. 8 ABWR standardsafety Analysis Report (~' N (3) A minor fraction of the fission products released from the containment will be taken into the Control Building via one of the CRHA HVAC System dual air intakes. The majority of the iodine taken in will be absorbed on a charcoal bed, which will then become a concentrated source within the building. Also, solid daughters of noble gases collect on the filters. Personnel on the control l room level, as well as the equipment room and HVAC room levels, will be I shielded from this source. (4) Fission products that pass through and evolve from the filters become a source of radiation exposure to Control Building personnel.This source determines a portion of the whole body dose, as well as the thyroid and beta skin doses. See Subsection 15.6.5 for these dose analyses. The DBA analysis is structured on the conservative NRC assumptions. The percentages of 102% rated power fission product equilibrium inventories released from the reactor I vessel and available for release from the containment are given below: Released from Available for Release Fission product Reactor Vessel from Containment Noble Gases 100 % 100 % l Halogens 50 % 25 % Solids 1% Negligible The primary containment leak rate assumed for the design analyses is 0.5% of the containment volume per day. Radioactive decay during transport through the containment is taken into account. The leaked radioactivity goes into the Reactor Building secondary containment and then to the SGTS, from which it is vented to the atmosphere. The SGTS charcoal filter is assumed to be 99% efficient for filtering radioiodines, and none of the vented gas is assumed to bypass the filter. 6.4.2.5.2 Source Terms and Results Containment sources " shining" on the Control Building are listed in Section 12.2. Source terms for the cloud and filter are consistent with the activity releases given in Subsection 15.6.5. Concentration of each isotope is calculated as the product of the release rate (Bq/s) times the appropriate relative concentration, or Chi /Q (s/m3 ). These values and the corresponding result are provided in Subsection 15.6.5. a 6.4.3 System Operation Procedures G During normal operation, the control room habitability area HVAC System operates with mixed recirculated and outdoor air, which pressurizes the subject spaces. Habitability Systems . Amendment 36 6.47
23A6100 Riv. 9 ABWR StandardSafetyAnalysis Report O Emergency conditions such as a LOCA or high radiadon cause an automatic changeover reducing outside air intake and to start charcoal filtering all outside air and a portion of the return air. This effectively isolates operating personnel from the emironment and from airborne contamination. Protection from direct radiation is discussed in Subsection 6.4.2.5. Detection of radioacdvity is instrumented, and changeover to reduced circulation and charcoal filtering is automatic. Redundancy ofinstrumentation and air handling systems ensures against system failure due to single component failure. The above operational description is brief. For a more detailed descripdon of normal and emergency operation of the control room habitability systems, see l Subsections 9.4.1, 9.5.1, 9.5.3,12.3.4, 6.5.1, and Chapter 8. 6.4.4 Design Evaluations 6.4.4.1 Radiological Protection l The Chi /Qs used for evaluation of the control room operator dose to meet General Design Criterion 19 are presented in Subsectiou 15.6.5. 6.4.4.2 Smoke and Toxic Gas Protection As discussed and evaluated in Subsection 9.5.1, the use of non-combustible construction and heat and flame-resistant materials throughout the plant minimizes the likelihood of fire and consequential fouling of the main control area envelope atmosphere with smoke or noxious vapor introduced in to the control room air. In the smoke removal mode, the purge flow through the Control Building provides three air changes per hour in order to sweep atmospheric contaminants out of the area. The main control area envelope is normally exhausted from the recirculation plenum by one of the exhaust fans. Smoke removal is accomplished by starting both exhaust fans at high speed in conjunction with a supply fan and realigning the dampers for exhausting directly to the exhaust vent. The above changeover is under manual control from the main control room. Operating personnel in the control room exercise this opdon in response to signals from the smoke detection sensors located in the subject spaces and in the associated ductwork. Transfer of the system to the isolation mode for exterior smoke may also be initiated manually from the control room. Local, audible alarms warn the operators to shut the self-closing doors,if, for some reason, they are held open after the receipt of a transfer signal. Isolation mode makeup air flow, required after approximately 72 hours of isolation (based on the buildup of carbon dioxide to 1% by volume in the space due to the respiration of 12 persons), must be initiated manually by the operator after tests 6.4-8 Habitability Systems - Amendment 37
23A6100 R1v. 9 ABWR standardsafetyAnalysis Report r0 l V system. Refer to IBD Figure 7.3-7 for specific indication of equipment status in the control room. See Chapter 16 for setpoints and margin. (k) Parts of System Not Required for Safety The non-safety-related portions of the RGV System include the annunciators and the computer. Other instrumentadon considered non-safety-related are those indicators that are provided for operator information, but are not essential to correct operator action. 7.3.1.1.8 Essential HVAC Systems-Instrumentation and Controls See Subsections 9.4.1 and 9.4.5. 7.3.1.1.9 HVAC Emergency Cooling Water System-Instrumentation and Control (1) System Identification The HVAC Emergency Cooling Water System (HEQV) supplies demineralized chilled water to the cooling coils of the control building safety-related electrical equipment rooms and main control room coolers, and the ( diesel generator zone air conditioning systems. The system is composed of l three divisions, each containing two refrigerators and chilled water pumps . The Control Building Chilled Water System instrumentation and controls are shown on P&lD Figure 9.2-3 and the corresponding logic on Figure 7.3-9. (2) Support Systems (Power Source) The instrumentation and controls of the HEQV System are supplied with [ 120 VAC and 125 VDC electric power from Division I, II, and III power buses. (3) Equipment Design The HEQV System consists of three mechanically (and electrically) separate systems-Divisions A, B, and C. The system is designed to provide chilled water to the cooling coils of the Control Building Control Room Habitability Area HVAC and Safety-Related Equipment Area HVAC and Reactor Building Safety-Related Electrical Equipment HVAC Systems. The HECW System is designed to operate during both accident conditions and normal plant operation and during all modes of operation for the cooling systems it serves. (V Each division of the HEQV System consists two chilled water pump and refrigerator units; each refrigerator unit includes the condenser, evaporator, l Engineered Safety Feature Systems, Instrumentation and Control- Amendment 37 7.3 57 l
I l 23A6100 Rov. 9 l l ABWR StandardSafety Analysis Repott centrifugal compressor, refrigerant pipings and package chiller controls. The system condenser is cooled by the RCW System. l Lack of flow of Reactor Building cooling water to the refrigerant condenser l automatically stops the refrigerator. Supply flow is controlled by the [ condensing pressure of the refrigerant. A flow switch provided at the chilled water line shuts down the refrigerator and chilled water pump indication of low flow in the chilled water line. (a) Initiating Circuits The HECW System operation is initiated automatically when the controls in the main control room are set for automatic operation and l any of the HVAC systems it serves are started. The HECW System can also be started manually from the main control room. (b) Logic and Sequencing l The standby unit (refrigentor and chilled water pump) in Division A is l automatically initiated when the operating unit is shut down. In I Divisions B and C, any unit on standby is automatically initiated when I any of the other operating units in Division B or C are stopped. (c) Bypass and Interlocks Low and high surge tank level switches actuate the demineralized water makeup or supply valves. Low-low or high-high surge tank level initiates an alarm in the control room to indicate a leak or a failure in the level control loop. Flow switches provided on the chilled water line are interlocked to automatically shut down the refrigerator in the event oflow flow in the chilled water line. A common trouble alarm for each refrigerator unit is annunciated in the control room upon detection of any refrigerator unit alarm or trip. A running signal from each RCW pump in each division is interlocked to trip the refrigerators if at least one RCW pump is not operating. Each refrigerator unit when on standby is interlocked to automatically start as described in (b). The running refrigerator is interlocked to trip on abnormal operating conditions such as lack of flow of chilled water and chiller package trouble. (d) Redundancy and Diversity 7.3-58 Engineered Safety Feature Systems. Instrumentation and Control- Amendment 37
h
- 23Assoo mv. s ABWR studentsafetyAulysis Report d
- The Control Room Habitability Area Chilled Water System is divided into two completely independent and functionally redundant systems.- ,
Physical and electrical separation is maintained between the two i redundant systems. (e) Actuated Devices , One refrigerator and chilled water pump in each division is running at all times during all modes of plant operation. The chilled water pumps and refrigerator units are started automatically or by remote manual switch. Status lights in the control room are also provided for this equipment. High and low surge tank level switches actuate the opening and closing , of the demineralized water makeup valve and high-high and low-low tank level switches annunciate an alarm in the control room. l The refrigerator capacity is controlled to maintain the chilled water temperature at the refrigerator outlet constant. This is done by adjusting the suction valve and hot-gas bypass within the refrigerator. (f) Separation The instrumentation, controls, and sensors of each operating division
- have sufficient physical and electrical separation to prevent l
- environmental, electrical, or physical accident consequences from
, inhibiting the systems from performing each protective action. Physical separation is maintained by use of separate cabinets and racks for each
' division, and by housing redundant chiller equipment in separate cubicles.
Electrical separation is maintained by separate independent sensors and l circuiuy. (g) Testability Manual initiation of the HECW System is possible from the control room. Redundant standby components can be periodically tested, manually, to ensure system reliability while the other system is operating. 5
- Su:ge tank operation can be checked by varying the tank level and
. observing the level at which the demineralized water makeup valve starts i- to open and close and when the level alarm annunciates. Automatic initiation of the standbysystem can be tested by simulating the trip action of the operating refrigerator system.
Engineered Safety Feature Systems. Instrumentation and Control- Amendment 37 7.3-59
23A6100 Rev. 4 ABWR Standard SafetyAnalysis Report 9 initiation of the standby system can be tested by simulating the trip action of the operating refrigerator system. s All motor-operated valves can be independently checked by operating the respective manual switch in the control room and obsening the corresponding position indicator. System chilled water flow rate and temperature can be checked by readout oflocally mounted pressure and temperature gauges at the main control panel. (h) Environmental Consideration All components of the HECW System are selected in consideration of the normal and accident emironment in which it must operate. The control equipment is seismically qualified and emironmentally classified, as discussed in Sections 3.10 and 3.11. (i) Operational Consideration The HECW System operation is initiated in the control room by a manual master control switch. Once the system is started, it will continuously operate under all modes of plant operation to supply chilled water to the cooling coils. Running lights, alarms, flow and temperature indicators, and valve position indicators are available in the control room for the operator to accurately monitor the HECW System operation. Chilled water pumps have running lights. A common trouble alarm is provided for each chiller unit. Surge tank high-high and low-low levels are alarmed. Motor-operated valves have position indicators. Chilled water flows have I position indicators. l 7.3.1.1.10 High Pressure Nitrogen Gas Supply System--instrumentation and Controls (1) System Identification l l The High Pressure Nitrogen Gas Supply (HPIN) System provides compressed l nitrogen of the required pressure to the ADS SRVs, the MSIVs (for testing , only), instruments and pneumatically operated valves in the PCV and other l nitrogen using components in the reactor building (see P&ID in Figure 6.7-1 and the interconnection block diagram in Figure 7.3-10). (2) Support Systems (Power Source) 7.3-60 Engineered Safety Feature Systems. Instrumentation and Control- Amendment 34
23A6100 Rn. 9 ABWR standardsafety Analysis Report /x In accordance with the Standard Review Plan for Section 7.3, and with Table 7.12, only BTPs 21 and 22 are considered applicable for the HVAC System. They are addressed as follows: 1 (a) BTP 1CSB 21 -Guidancefor Applicalionfor Regulatory Guide 1.47 The ABWR design is a single unit. Therefore, item B-2 of the BTP is not applicable. Othenvise, the HVAC System is in full compliance with this BTP. i l (b) BTP ICSB 22-Guidancefor Application ofRegulatory Guide 1.22 All actuated equipment within the HVAC System can be fully tested during reactor operation. (5) TMI Action Plan Requirements (TMI) In accordance with the Standard Resiew Plan for Section 7.3, and with Table 7.1-2, there are no TMI action plan requirements applicable to the l HVAC System. (O V) 7.3.2.9 HVAC Emergency Cooling Water System-Instrumentation and Control 7.3.2.9.1 Conformance to General Functional Requirements The HVAC Emergency Cooling Water (HECW) System provides chilled water to the Control Building Safety-related Equipment Area HVAC and to the Control Room Habitability Area HVAC and Reactor Building Safety-related Electrical Equipment HVAC Systems. It is designed to function under all operating, emergency and accident conditions. 7.3.2.9.2 Specific Regulatory Requirements Conformance l Table 7.1-2 identifies the HECWSystem and the associated codes and standards applied in accordance with the Standard Review Plan. The following analysis lists the applicable criteria in order of the listing on the table, and discusses the. degree of conformance for each. Any exceptions or clarifications are so noted. (1) 10CFR50.55a (IEEE-279) The HVAC Emergency Cooling Water (HECW) System has three independent electrical divisions and is redundantly designed so that failure of any single electrical component will not interfere with the required safety action of the system. The HECW System is manually actuated, but is designed to run continuously during reactor operation. Should a loss of station power or a LOCA event Engineered Safety Feature Systems. Instrumentation and Control- Amendment 37 7.3-91
23A6100Rtv 9 ABWR StandardSafety Analysis Report 9 occur, the system power sources will automatically switch over to the emergency diesels. Thus, condnuous operation is assured for all plant conditions. All components used for the safety functions are qualified for the emironments in which they are located (Sections 3.10 and 3.11). The HECW System utilizes mechanical Divisions A, B and C corresponding with electrical Divisions I, II, and III, respectively. Electrical separation is maintained between the redundant divisions. The HECW System is designed to meet all applicable requirements ofIEEE-279. Detailed system design descriptions are given in Subsection 7.3.1.1.9 and in Chapter 9. (2) General Design Criteria (GDC) In accordance with the Standard Review Plan for Section 7.3, and with Table 7.1-2, the following GDCs are addressed for the HVAC System: (a) Criteria: GDCs 2,4,13,19,20,21,22,23,24,29, and 44. (b) Conformance: The HECW System is in compliance as a whole, or in part as applicable, with all GDCs identified in (a), as discussed in l Subsection 3.1.2. l 1 1 (3) Regulatory Guides (RGs) { 1 In accordance with the Standard Review Plan for Section 7.3, and with Table ' 7.1-2, the following RGs are addressed for the HECW System: (a) RG 1.22-Periodic Testing ofProtection System Actuation Functions (b) RG 1.47-Bypassed and inoperable Status indicationfor Nuclear Poutr Plant Safety Systems (c) RG 1.53-Application ofIhe Single-Failure Cnterion to NuclearPowerProtection Systems (d) RG 1.62-ManualInitiation ofProtective Actions (c) RG 1.75-PhysicalIndependence ofElectric Systems (0 RG 1.118-Periodic Testing ofElectric Power and Protection Systems The HECW System conforms with all the above listed RGs, assuming the same interpretations and clarifications identified in Subsections 7.3.2.1.2 and 7.1.2.10. 7.3-92 Engineered Safety Feature Systems, Instrumentation and Control- Amendment 37
23A6100 Rev. 9 ABWR standardsateoyAnalysis neport 1 List of Tables (Continued) l Table 9.4-4g HVAC System Component Descriptions-Non-Safety-Related Fans .. ... ... .9.4-47 Table 9.4-4h HVAC System Component Descriptions-Non-Safety-Related Filters......... .. 9.4-47 Table 9.4 4i HVAC System Component Descriptions-Non-Safety-Related l Air Handiing Units.. . . . . . . ............................................. .... 9.4-48 Table 9.+5 Turbine Building and Electrical Building HVAC System-Non-Safety-Related Equipment List . . .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . .. . . . . 9. 4-4 9 Table 9.5-1 Normal and/or Standby Lighting (Non-Class 1E AC Power Supply) .. . .. . 9.5-78 Table 9.5-2 Lighting and Power Scurces ......... ....... . . . . . . . . . . . . . . . . . . . . ... . . . 9.5-79 Table 9.5-3 Stan d by Ligh ting .... ...... ........... . ... . .. .. .... . ... ... . . . . . . . . . . . . . . . . . . . . . . . . . 9. 5-7 9 Table 9.5-4 DC Emergency Lighting (Class 1E DC Power Supply) .. . ... . ..... .. .... ... . .. 9.5-80 Table 9.5-5 Summary of Automatic Fire Suppression Systems.. .... .... . . . ... .. ... .... ... . . 9.5-81 Table 9A.2-1 Core Coolin g... ........... ...... .. . ... ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... . .. 9A.2-8 Table 9A.5-1 Redundant Instmmentation or Equipment in Same Fire Area .. ....... .... . 9A.5-18 Table 9A.5-2 Summary of the Reactor Building Special Cases.. .... ... ......... ... ... .. . .... . 9A.5-19 Table 9A.61 Fire Hazard Analysis, Equipment Data Base - Sorted by MPL Number (Superceded by Table 9A.62) .. . ........... ................ ... .. . ... . .. ...... . ....... .... 9A.65 1 Table 9A.6-2 Fire Hazard Analysis Equipment Database Sorted by Room -Reactor Building......................................................................... .. ... . . ..... . 9A.6-6 Table 9A.6-3 Fire Hazard Analysis Equipment Data Base -Sorted by Room - Control Building........... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .... .. . ... ... 9A.6-9 6 Table 9A.64 Fire Hazard Analysis Equipment Database-Sorted by Room-Turbine Building..................................................................... . ... . . . .. .. ... .. . 9A.6-10 6 Table 941 Estimated Fire Severity for OfIices and Light Commercial Occupancies.... .. . 9&l2 Table 9&2 Fire Severity Expected by Occupancy . ......... .... . ...... .... . ....... ... ..9&13 l Table 9&S - Cable Type and Configuration for Ul Tests.. . .. . . . . . . . . . . . . . . . . . . . . . . . . . ...... 9&l 4 Table 944 Summary of Burning Rate Calculations.. . . . . . . . . . . . . . . . . . . . . . . . . . . . ... .. . . . 9 41 4 List of Tables - Amendment 37 9.0-v/vi
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23A6100 R1v. 2 ABWR standardsafetyAnalysis Report t U O) sludge from the final clarifiers is aerated, sent to the aeration tanks and mixed with incoming sewage. l 9.2.4.2.5 Evaluation of Potable and Sanitary Water System Performance (Interface Requirements) The COL applicant shall analyze the PSW System to assure that the system meets all applicable regulatory requirements and is compatible with site conditions. 0.2.4.2.6 Safety Evaluation (Interface Requirements) The PSW System has no interconnection with systems having the potential for containing radioactive materials. Protection includes, where necessary, the use of air gaps. 9.2.4.2.7 Instrumentation and Alarms (Interface Requirements) The subsystems of the PSW System are provided with control panels located in the Control Building which are designed for remote manual and automatic control of the processes. O A flow proportioning controller is used to operate the hypochlorinator pump as water V enters the PSW System. Pressure and level switches are provided to start and stop the potable water pumps and the air compressor. Low hydropneumatic tank pressure is alarmed. Low level in the hypochlorite feed tank is alarmed. The minimum instrumentation requirements for the STS are a treated eflluent sewage flow meter and a common air blower discharge pressure gauge. 9.2.4.2.8 Tests and inspections (Interface Requirements) Drainage piping is hydrostatically tested to the equivalent of a 3-meter head of water for a minimum of 15 minutes. The operability of all other parts of the PSW System is demonstrated by use during normal system operation. 9.2.5 Ultimate Heat Sink This subsection provides a conceptual design of the ultimate heat sink (UHS) as required by 10CFR52. The interface requirements for the UHS are part of the design certification. 9.2.5.1 Safety Design Bases (Interface Requirements)
! (1) The UHS is designed to provide sufIicient cooling water to the Reactor Senice Water (RSW) System to permit safe shutdown and cooldown of the unit and Water Systems - Amendment 32 9.2-5
23A6100 Rtv. 9 ABWR standardsafety Analysis Report 1 1 l O l maintain the unit in a safe shutdown condition. The RSW water temperature at the inlet to the RCW/RSW heat exchangers is not to exceed 35 C during a LOCA. (2) In the event of an accident, the UHS is designed to provide sufficient coohng water to the RSW System to safely dissipate the heat for that accident. The amount of heat to be removed is provided in Tables 9.2-4a,9.2-4b and 9.2-4c. (3) The UHS is sized so that makeup water is not required for at least 30 days following an accident and design basis temperature and chemisuylimits for safety-related equipment are not exceeded. (4) The UHS is designed to perform its safety function during periods of adverse site conditions, resulting in maximum ws.ter consumption and minimum cooling capability. (5) The UHS is designed to withstand the most severe natural phenomenon or site-related event (e. g., SSE, tornado, hurricane, flood, freezing, spraying, pipe whip, jet forces, missiles, fire, failure of non-Seismic Category I equipment, fic.oding as a result of pipe failures or transportation accident), and reasonably probable combinations ofless severe phenomena and/or events, withouti mpairing its safety function. (6) The safety-related portion of the UHS shall be designed to perform its required cooling function assuming a single active failure in any mechanical or electrical system. (7) The UHS is designed to withstand any credible single failure of man-made structural features without impairing its safety function. (8) All safety-related heat rejection systems shall be redundant so that the essential cooling function can be performed even with the complete loss of one division. Single failures of passive components in electrical systems may lead to the loss of the affected pump, valve or other components and the partial or complete loss of cooling capability of that division but not of other divisions. (9) The UHS and any pumps, valves, structures or other components that remove heat from safety systems shall be designed to Seismic Category I and ASME Code, Section III, Class 3, Quality Assurance B, Quality Group C, IEEE-279 and IEEE-308 requirements. l j (10) The safety-related portions of the UHS shall be mechanically and electrically separated. (11) The UHS is designed to include the capability for full operational inspection and testing. 9.2-6 Water Systems - Amendment 37
23A6100 Rsv. 6 ABWR senaderdseteryAutysis neport l 9.2.10.2 System Description i The MUWP System P&ID is shown in Figure 9.2-5. This system includes the following: (1) Distribution piping, valves, instmments and controls shall be prosided. (2) Any outdoor piping shall be protected from freezing. 1 I (3) All surfaces coming in contact with the purified water shall be made of corrosion-resistant materials. l (4) Continuous analyzers are located at the MUWP System. These are ! supplemented as needed by grab samples. Allewance is made in the water I quality specifications for some pickup of carbon dioxide and air in any demineralized water storage tank. The pickup of corrosion products should be minimal because the MUWP piping is stainless steel. (5) Intrusion of radioactivity into the MUWP System from other potentially radioactive systems are prevented by one or more of the following: (a) Check valves in the MUWP lines. L - (b) Air (or siphon) breaks in the MUWP lines. (c) The MUWP System lines are pressudzed while the receiving system is at essentialh atmospheric pressure. (d) Piping to the useris dead ended. (6) Th.:re are no automatic valves in the MUWP System. Dudng a LOCA, the safety-related systems are isolated from the MUWP System by automatic valves in the safety-related system. (7) The outboard primary containment isolation valve is locked closed during standby, hot standby and power operation. 9.2.10.3 Safety Evaluation Operation of the MUWP System is not required to assure any of the following conditions: (1) Integrity of the reactor coolant pressure boundary. (2) Capability to shut down the reactor and maintain it in a safe shutdowm condition. % (3) Ability to prevent or mitigate the consequences of events which could result in potential offsite exposures. Water Systems - Amendment 35 9.2-19
23A6100 Rev. 9 ABWR StandardSafety Analysis Report O The MUWP System is not safety-related. However, the systems incorporate features that assure reliable operation over the full range of normal plant operations. 9.2.10.4 Tests and Inspections The MUWP System is proved operable by its use during normal plant operation. Portions of the system normally closed to flow can be tested to ensure operability and integrity of the system. Flow to the various systems is balanced by means of manual valves at the individual takeoff points. 9.2.11 Reactor Building Cooling Water System 9.2.11.1 Design Bases 9.2.11.1.1 Safety Design Bases (1) The Reactor Building Cooling Water (RCW) System shall be designed to remove heat from plant auxiliaries which are required for a safe reactor shutdown, as well as those auxiliaries whose operation is desired following a LOCA, but not essential to safe shutdown. The heat removal capacity is based on the heat removal requirement during a ; LOCA with the maximum RSW water temperature at the inlet to the RCW/RSW heat exchangers of 35'C. As shown in Table 9.2-4a, the heat I removal requirement is higher during other plant operation modes, such as shutdown at 4 hours. However, the RCW System is designed to remove this larger amount of heat to meet the requirements in Subsection 5.4.7.1.1.7. l (2) The RCW System shall be designed to perform its required cooling functions following a LOCA, assuming a single active or passive failure. (3) The safety-related portions and valves isolating the non-safety-related portions of the RCW System shall be designed to Seismic Category I and the ASME Code, Section III, Class 3, Quality Assurance B, Quality Group C, IEEE-279 and IEEE-308 requirements. (4) The RCW System shall be designed to limit leakage to the emironment of radioactive contamination that may enter the RCW System from the RHR System. (5) Safety-related portions of the RCW System shall be protected from flooding, spraying, steam impingement, pipe whip, jet forces, missiles, fire, and the effect of failure of any non-Seismic Category I equipment, as required. 9.2-20 Water Systems - Amer.dment 37
23A6100 Rsv. 9 ABWR standadsaktyAnstyskneret v l (4) The system shall be powered from Class 1E buses. Power shall be available from the Alternate AC (AAC) power source when required. l (5) The HECW System shall be protected from missiles in accordance with Subsection 3.5.1. (6) Design features to preclude the adverse effects of water hammer are in accordance with the SPJ section addressing the resolution of USI A-1 discussed in NUREG-0927, l These features shallinclude: (a) An elevated surge tank to keep the system filled. (b) Vents provided at all high points in the system. l (c) After any system drainage, venting is assured by personnel training and procedures. ' , (d) System valves are slow acting. lO lh (7) The HECW System shall be protected from failures of high and medium energy lines as discussed in Section 3.6. (8) The design operation of the HECW compressors will take into account power or operational perturbations which could result in a) frequent immediate or L elongated restarts, b) in unacceptable compressor coolant and lubrication oil j interactions, and c) compressor coolant leaks or releases. (9) The system piping design will take into account unacceptable nil-ductility-temperature conditions associated with normal and transient operation. 9.2.13.2 System Description l The HECW System consists of subsystems in three divisions. Divisions A, B and C have two refrigerator units, two pumps, instrumentation and distribution piping and valves to corresponding cooling coils. A chemical addition tank is shared by all HECW divisions. Each HECW division shares a surge tank with the corresponding division of the RCW System. The refrigerator capacity is designed to cool the Reactor Building safety-related electrical equipment HVAC Systems and Control Building safety-related
- equipment area HVAC Systems.
1
- The system is shown in Figure 9.2-3. The refrigerators are located in the Control
{ Building as shown in Figures 1.2-20 and 1.2-21. Each refrigerator unit consists of a j evaporator, a compressor, refrigerant, piping, and package chiller controls. This system i shares the RCW surge tanks which are in the Reactor Building (Figure 1.2-12). Equipment is listed in Table 0.2-8. Each cooling coil is controlled by a room thermostat. Water Systems - Amendment 37 9.2 31 l
23A6100 Rav. 9 ABWR Standard Safety Analysis Report O I l Alternately, flow may be controlled by a temperature control valve. Condenser coolmg j is from the corresponding division of the RCW System. Piping and valves for the HECW System, as well as the cooling water lines from the RCW j System, designed entirely to ASME Code, Section 111, Class 3, Quality Group C, Quality ' Assurance B requirements. The extent of this classification is up to and including drainage block valves. There are no primary or secondary containment penetrations within the system.The HECW System is not expected to contain radioactivity. High temperature of the returned cooling water causes the standby refrigerator unit to start automatically. Makeup water is supplied from the MUWP System, at the surge tank. Each surge tank has the capacity to replace system water losses for more than 100 days i during an emergency. The only non-safety-related portions of the HECW divisions are the chemical addition tank and the piping from the tank to the safety-related valves which isolate the safety-related portions of the system. Also, see Subsection 9.2.17.1 for COL license information requirements. j i 9.2.13.3 Safety Evaluation i The HECW System is a Seismic Category I system, protected from flooding and tornado . missiles. All components of the system are designed to be operable during a loss of l normal power by connection to the ESF buses (Tables 8.3-1 and 8.3-2). Redundant l components are provided to ensure that any single component failure does not l l preclude system operation. The system is designed to meet the requirements of Criterion 19 of 10CFR50. The refrigerators of each division are in separate rooms. During a Station Blackout (SBO), the HECW refrigerators, pumps and instrumentation will be powered by the AAC System which will become available in ten minutes. Provisions will be made to ensure prompt and reliable restart of the chiller units. COL license information requirements are provided in Subsection 9.2.17.1. The response to SBO is discussed in Chapter 1, Appendix IC. During the SBO, little heat will be generated in the areas cooled by HECW because only battery powered equipment will be operating. These areas are the main control room, the Control l Building essential electrical equipment rooms and the Reactor Building essential j electrical equipment rooms. The HVAC fans in these areas are powered by Class 1E j buses. When AAC power becomes available, these fans will be powered and will start supplying outside air and exhausting any hot air from these areas. When chilled water l becomes available, cooled air will be circulated in these areas to restore normal i temperature. l If a LOPP event occurs, there are provisions for a stop signal to the HECW pumps to i trip the breakers or for sequencing the HECW pumps back onto the emergency bus 9.2-32 Water Systems - Amendment 37
23A6100 Rev. 9 ABWR standardsafety Analysis Report l /^'\ V during the allotted dme frame (load block 3), which is 15 seconds after the emergency buses are picked up by the diesel generators. Once the pumps are reconnected to the emergency bus, they are prevented from cycling on and off until the remaining LOPP sequence loads are connected to the emergency bus. If a LOCA follows a LOPP, there are provisions for resetting the start timers and connecting the HECW pumps to the emergency busses at the proper time if they are not already connected when the LOCA appears. l l Power is provided to the HECW refrigerators thirty seconds after it is provided to the HECW pumps. The HECW refrigerators will then begin a programmed startup process. The HECW system air operated valves will upon loss ofinstrument air or power assume configurations or positions that assure continued system cooling service. l 9.2.13.4 Tests and inspection l Initial testing of the system includes performance testing of the refrigerators, pumps and coils for conformance with design capacity water flows and heat transfer capabilides. An integrity test is performed on the system upon completion. l The HECW System is designed for periodic pressure and functional testing to assure: V t (1) the structural and leaktight integrity by visual inspection of the components; l (2) the operability and the performance of the acdve components of the system; and (3) the operability of the system as a whole. l Local display devices are provided to indicate all vital parameters required in testing and inspecdons. Standby features are periodically tested by initiating the transfer sequence during normal operation. The refrigerators are tested in accordance with ASHRAE Standard 30. The pumps are tested in accordance with standards of the Hydraulic Institute. ASME Section VIII and TEMA C standards apply to the heat exchangers. The cooling coils are tested in l accordance with ASHRAE Standard 33. 9.2.13.5 Instrumentation and Alarms i A regulated supply of makeup water is provided to add purified water to the surge tanks by water level controls. The chilled water pumps are controlled from the main control panel. The standby refrigerator has an interlock which automatically starts the standby refrigerator and l pump upon failure of the operating unit. I Water Systems - Amendment 37 9 2-33
l l l l 23A6100 Rtv. 8 1 ABWR StandardSafety Analysis Report The refrigerator units can be controlled individually from the main control room by a l remote manual switch. Chilled water temperature is controlled by inlet guide vanes on ) each chiller refrigerant circuit. Condenser water flow is controlled by a two-way olve { based on refrigerant compressor discharge pressure. l l A temperature controller and flow switch continuously monitor the discharge of each l evaporator. If the temperature of the chilled water drops below a specified level, the controller automatically adjusts the position of the compressor inlet guide vanes. Flow switches prohibit the chiller from operating unless there is water flow through both evaporator and condenser. 9.2.14 Turbine Building Cooling Water System . l 9.2.14.1 Design Bases l 9.2.14.1.1 Safety Design Bases l The Turbine Building Cooling Water (TCW) System (Figure 9.2-6) serves no safety function and has no safety design basis. There are no connections between the TCW System and any other safety-related I systems. 9.2.14.1.2 Power Generation Design Bases (1) The TCW System provides corrosion-inhibited, demineralized cooling water to all Turbine Island auxiliary equipment listed in Table 9.2-11. (2) During power operation, the TCW System operates to provide a continuous supply of cooling water, at a maximum temperature of 41 C, to the Turbine Island auxiliary equipment, with a service water inlet temperature not exceeding 37.8 C. (3) The TCW System is designed to permit the maintenance of any single active component without interruption of the cooling function. (4) Makeup to the TCW System is designed to permit continuous system l operation with design failure leakage and to permit expeditious post-maintenance system refill. l (5) The TCW System is designed to have an atmospheric surge tank located at the highest point in the system. (6) The TCW System is designed to have a higher pressure than the power cycle heat sink water to ensure leakage is from the TCW System to the power cycle l heat sink in the event a tube leak occurs in the TCW System heat exchanger. 9.2 34 Water Systems Amendment 36 l l r I
a p em t U U V . t k N
= =
?
5 9 Table 9.2-3 Capacity Requirements for Condensate Storage Tank lEl0 Dead space--top of pool 29,901L' l} ly Normal operation variation and receiving volume for plant startup return water 999,240L U Minimum storage volume 247,500L I 129,901L' g Dead space-middle of pool E Water source for station blackout 569,567Lt Dead space-bottom of pool 129,901L* , Total 2,108,321L
- These values are based on a bottom area of 130m3.
t Water for operation of RCIC is taken from the condensate storage tank and the suppression pool as described in the EPGs of Appendix 18A. g S 8
- n
.I
=
l 1 E. I. ' 4 e. a-
$ 2@
b a
$ Table 9.2-4a Reactor Building Cooling Water Division A h a 0 13 Emergency Normal (LOCA)
Operating Shutdown at 4 Shutdown at Hot Standby Hot Standby (Suppression Operating Mode / Components Conditions Hours 20 Hours (No Loss of AC) (Loss of AC) Pool at 97*C Heat" Flow
- Heat Flo w Heat Flow Heat Flo w Heat Flow Heat Flo w Essential Emergency Diesel Generator A - - - - - - - -
13.40 229 13.40 229 RHR Heat Exchanger A - - 108.02 1,199 34.75 1,199 - - 25.54 1,199 89.18 1,199 Others (essential)* 3.18 205 3.60 205 3.81 205 3.39 205 4.10 205 4.19 205 Non-Essential CUW Heat Exchanger' 20.10 159 - 159 159 20.10 159 20.93 159 - - FPC Heat Exchanger Af 7.12 279 7.12 279 7.12 279 7.12 279 7.12 279 9.63 279 Inside Drywell** 5.86 320 5.86 020 5.86 320 5.86 320 3.39 320 - - Others (non-essential)" 2.64 160 2.64 160 2.64 160 2.64 160 0.84 59 0.75 59 Total Load 38.94 1,123 127.24 2,322 54.01 2,32? 38.94 1,123 75.36 2,450 117.23 1,971 ! 2
- Heat in GJ/h; flow in m 3/h, sums may not be equal due to rounding.
t HECW refrigerator, CAMS coolers, room coolers (RHR, RCIC, CAMS), RHR motor and seal coolers.
- The heat transferred from the CUW heat exchanger at the start of cooldown is appreciable, but during the critical last part of a cooldown, the heat removed is very little because the temperature difference between the reactor water and the RCW System is small. Sometimes, the operators may remove the CUW heat exchangers from service during cooldown. Thus, the heat removed varies from about that during normal operation at the start of cooldown to very little at the end of cooldown.
g f includes FPC room cooler. E i ** Drywell (A & C) and RIP coolers. m @ m. l} t t instruments and service air coolers; CUW pump cooler, CRD pump oil, and RIP MG sets. p e 2 a 5-
! 3
?, P
~ 2 a
e O - O
,,m .
u 5 Table 9.2-4b Reactor Building Cooling Water Division B b E El m Emergency
} Normal Operating Shutdown at 4 Shutdown at Hot Standby Hot Standby (LOCA)
(Suppression 3 i Operating Mode / Components Conditions Hours 20 Hours (No Loss of AC) (Loss of AC) Poolat 97 C l Heat Flow
- Heat Flo w Heat Flow Heat Flow Heat Flo w Heat Flo w
{ Essential {" Emergency Diesel Generator B - - - - - - - - 13.40 229 13.40 229 RHR Heat Exchanger B - - 108.02 1,199 34.75 1,199 - - 25.54 1,199 89.18 1,199 Others (essential)t 6.28 360 6.70 360 6.70 360 6.28 360 7.12 360 7.95 360 Non-Essential CUW Heat Exchanger
- 20.10 159 -
159 - 159 20.10 159 20.93 159 - - FPC Heat Exchanger BI 7.12 279 7.12 279 7.12 279 7.12 279 7.12 279 9.63 279 Inside Drywell" 5.44 279 6.28 279 5.40 279 5.40 279 2.51 279 - - g Others (non-essential)" 2.93 159 1.47 159 1.47 159 1.47 159 0.33 9.1 - 9.1 e Total Load 41.87 1,236 129.79 2,435 55.27 2,435 40.19 1,236 77.04 2,514 120.16 2,076 k E
- Heat x GJ/h; flow x m 3/h, sums may not be equal due to rounding.
t HECW refrigerator, room coolers (RHR, HPCF, SGTS, FCS, CAMS), CAMS cooler, HPCF and RHR motor and mechanical seat coolers.
- The heat transferred from the CUW heat exchanger at the start of cooldown is appreciable, but during the critical last part of a cooldown, the heat removed is very little because the temperature difference between the reactor water and the RCW System is small. Sometimes, the operators may remove the CUW heat exchangers from service during cooldown.Thus, the heat removed g varies from about that during normal operation at the start of cooldown to very little at the end of cooldown. g f includes FPC room cooler. &
** Drywell (B) and RIP coolers, k.
l t t Recctor Building sampling coolers; LCW sump coolers (in drywell and reactor building), RIP MG sets and CUW pump coolers. { D w 2 w 2a
@ Table 9.2-4c Reactor Building Cooling Water Division C b g
tu Emergency Normal (LOCA) Operating Shutdown at4 Shutdown at Hot Standby Hot Standby (Suppression Operating Mode / Components Conditions Hours 20 Hours (No Loss of AC) (Loss of AC) Pool at 97 C Heat Flo w
- Heat Flow Heat Flo w Heat Flo w Heat Flo w Heat Flo w Essential Emergency Diesel Generator C - - - - - - -
13.40 229 13.40 229 l RHR Heat Exchanger C - - 108.02 1,199 34.75 1,199 - - 25.54 1,199 89.18 1199 l Others (essential)' 6.28 360 6.70 360 6.70 360 6.28 360 6.70 360 7.12 360 l Non-Essential Others (non-essential)* 20.51 422 19.26 422 7.54 422 20.51 422 0.54 50 0.75 50 l l Total Load 26.80 782 133.98 1,981 48.57 1,981 26.80 782 46.05 1838 110.53 1838 l
- Heat x GJ/h; flow x m3 /h, sums may not be equal due to rounding.
t HECW refrigerator, room coolers, motor coolers, and mechanical seal coolers for RHR and HPCF, FCS room cooler, SGTS room cooler. 8 t Instrument and service air coolers, CRD pump oil cooler, radwaste components, HSCR condenser, and turbine building sampling coolers. s tr E :.
? E I ? ! e Y
I h I a
- Ta-g =
A $ 2 E O O O
23A6100 Rw. 4 ABWR senaduesaktyAnalysis neport O < U ! l Table 9.2-7 HVAC Normal Cooling Water Loads ; During Refueling , During Normal Operation Shutdown ) l l Name of Area or Unit Capacity GJ/h Flow m3 /h Capacity GJ/h Flow m /h 3
)
l Reactor Building I Drywell Coolers 0.96 69.5 0.80 69.5 j RIP Coolers 1.59 20.9 3.06 104 Others (Note 1) 10.05 131 18.84 636 Turbine Building (Note 2) l l 2.26 43.5 1.13 39 i i l Radwaste Building 5.69 81.2 6.70 232 j l (Note 4) l Service Building 3.64 175 3.64 175 l Others 4.61 151 3.56 151 (Note 5) i l Total 28.89 672 37.68 1,407 l (Note 6) i 1 lt 0 NOTES: j (1) Loads include reactor / turbine building supply units, HVH, FCU and room coolers. l (2) Loads are the offgas cooler condenser (normal operation only) and the electrical equipment supply j unit. (3) Deleted l 1 (4) Loads included are the radwaste building supply unit and the radwaste building electrical l equipment room supply unit. (5) Loads include HVH units not previously included. l l (6)The HNCW chillers are 9.38 GJ/heach and the pumps 449m3 /h each. Thus, four HNCW pumps have total capacity in excess of the amount required as shown in the last column of the table. l , j l l 1 1 i V l Water Systems - Amendment 34 9.2-59 l l
l 23A6100 Rzv. 9 ABWR standardsareryAnalysis neport O Table 9.2-8 HECW System Component Description
- HECW Chillers l Quantity 6 l Capacity (Refrigerator) six 2.51 GJ/h l Chilled water pump flow six 57 m3/h Supply temperature 7'C l Condenser water flow six 128 m3/h Supply temperature (max.) 45'C Condenser Shell and tube Evaporator Shell and tube HECW Water Pumps l Quantity 6 (57 m3/h each)
Type Centrifugal, horizontal l
- Each of Divisions A, B and C have two parallel pump-refrigerator units.
O O 9.2-60 Water Systems - Amendment 37
i 23A6100 R1v. 4 ABWR Standard Safety Analysis Report 1 ( Table 9.2-9 HVAC Emergency Cooling Water System Heat Loads Normal Emergency Chilled Chilled ) Water Heat Water j Heat Load Flow Load Flow l Division System 3 3 l (GJ/h) (m /h) (GJ/h) (m /h) l A Reactor Building 0.88 14.3 0.88 14.3 Electrical Equipment Room (A) l l Control Building 1.26 20.2 1.26 20.2 Electrical Equipment Room (A) l l Total 2.14 34.5 2.14 34.5 l B Main Control Room 1.42 26 1.30 24 l / \ Reactor Building 0.92 15 0.92 'i 5 \. /l Electrical Equipment Room (B) I l Control Building 1.26 20.2 1.26 20.2 Electrical Equipment Room (B) l Total 3.60 61.2 3.48 59.2 l C Main Control Room 1.42 26 1.30 24 l Reactor Building 0.92 15 0.92 15 Electrical Equipment Room (C) l Control Building 1.26 20.2 1.26 20.2 Electrical Equipment Room (C) l Total 3.6 61.2 3.48 59.2 O G Water Systems - Amendment 34 9.2-61
23A6100 Rsv. 9 ABWR StandardSafety Analysis Report \ l Table 9.2-10 HVAC Emergency Cooling Water System Active Failure Analysis Ol Failure of diesel generator to start or failure Loss of one refrigerator and pump in of all power to a single Class 1E power Division B or C would not permit sending system bus, chilled water to the Control Room Habitability Area HVAC System from the l affected division. The other HECW division would send chilled water to the Control Room Habitability Area HVAC System which would maintain adequate cooling. in l Division A, loss of both of the refrigerators l or the pumps would result in loss of cooling water flow to Division A Control Building safety-related Equipment Area HVAC System and Reactor Building safety-related Electrical Equipment HVAC System. Cooling i of Control Room Habitability Area HVAC System not affected. Failure of auto pump or refrigerator signal. Same analysis as above. Failure of a single HECW refrigerator. Same analysis as above. Failure of a single HECW pump. Same analysis as above. Failure of HECW pump and refrigerator Same analysis as above. room cooling. Table 9.2-11 Turbine Island Auxiliary Equipment The TCW System removes heat from the following components: HVAC normal cooling water chillers
+
Generator stator coolers, hydrogen coolers, seal oil coolers, exciter coolers and breaker coolers Turbine lube coolers
+
Mechanical vacuum pump coolers a Isophase bus coolers Electro-hydraulic control coolers
+
Reactor feed pump and auxiliary coolers
+
Standby reactor feed pump motor coolers Condensate pump motor coolers Heater drain pump motor coolers O 9.2 62 Water Systems - Amendment 37
' * '" Rev.4 ABWR S!*ndard Safety Analysis Report O
l The following figures are located in Chapter 21 : Figure 9.2-1 Reactor Building Cooling Water System P&lD (Sheets 1-9) Figure 9.2-1a Deleted Figure 9.2-2 HVAC Normal Cooling Water System P&lD Figure 9.2-3 HVAC Emergency Cooling Water System P&lD (Sheets 1-3) Figure 9.2-4 Makeup Water (Condensate) System P&lD Figure 9.2-5 Makeup Water (Purified) System P&lD (Sheets 1-3) O Water Systems - Amendment 34 9.2-69
E b I [A f suncE TANK *
< RETURN FROu FIG. 92-6B W
CHEMICAL ADDITION / . , RETURN FROM TANK L '
-lg l ' FIG. 92-6C
>4 k >I: < MAKE-UP WATER >
IG 2 B
-l K F V V I
_' , sueety To FIG. 92-6C T i i i ISO PHASE BUS COOLERS O lXl GEN. STATOR COOLERS iX F 1 I TCW PUMPS
+'* % :x: ng :u
!Xl g
~
TCW HEAT EXCHANGERS - 4XF M -l k t-
-l k t M -I X f-
-_jg; I I - jgp 1Af M -l X f-IK Q M I Ilkl GEN. C/B COOLERS
- lXi] l M j likt
' I
-ix : _
-- lwF-i w f-a %
g EXCITER COOLERS I I E C L b !! TCW HXS IICVI i ix: af3 '4- t i
- The surge tank is shared with the HNCW system.
l
)
)
1 H2 SEAL OIL COOLERS -l5 Y i;r i Figure 9.2-6a Turbine Building Cooling Water System Diagram e g j O O O _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ ___________________________r _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _
_ - _ = _ _ . _ 23A6100 R;v. 9 ABWR standed safety Analysis negat a f (w An ATWS condition exists when either of the following occurs: (a) IIigh RPV pressure (7.76 MPaG) and startup range neutron monitor (SRNM) not downscale for 3 minutes, or (b) Low RPV level (Level 2) and SRNM not downscale for 3 minutes. A light in the control room indicates that power is available to the pump motor contactor and that the contactor is deenergized (pump not mnning). Another light indicates that the contactor is energized (pump running). Storage tank liquid level, tank outlet valve position, pump discharge pressure, and injection valve position indicate that the system is functioning. If any of these items indicates that the liquid may not be flowing, the operator shall immediately change the other switch to the START mode, thereby activating the redundant train of the SLCS. The local switch cannot prevent the operation of the pump from the control room. Pump discharge pressure and valve status are indicated in the control room. Equipment drains and tank overflow are not piped to the Radwaste System but to separate containers (such as 208L drums) that can be removed and disposed of p independently to prevent any trace of boron from inadvertently reaching the reactor. O Instrumentation consisting of solution temperature indication and control, solution level and heater system status is provided locally at the storage tank. Table 9.3-1 contains the process data for the various modes of operation of the SLCS. Seismic category and quality class are included in Table 3.24. Principals of system testing are discussed in Subsection 9.3.5.4. 9.3.5.3 Safety Evaluation The SLCS is a reactivity control system and is maintained in an operable status whenever the reactor is critical. The system is never expected to be needed for safety reasons because of the large number ofindependent control rods available to shut down the reactor. To assure the availability of the SLCS, two sets of the components required to actuate the system (pumps and injection valves) are provided in parallel redundancy. The system is designed to bring the reactor from rated power to a cold shutdown at any time in core life. The reactivity compensation provided will reduce reactor power from rated to zero level and allow cooling of the nuclear system to room temperature, with the control rods remaining withdrawn in the rated power pattern. It includes the g reactisity gains that result from complete decay of the rated power xenon inventory. It i also includes the positive reactivity effects from eliminating steam voids, changing water density from hot to cold, reduced Doppler effect in uranium, reducing neutron leakage from boiling to cold, and decreasing control rod worth as the moderator cools. Process Auxiliaries . Amendment 37 9.343
l l 1 23A6100 Rev 4 ABWR standard sarery Analysis Report To meet this objective, it is necessary to inject a quantity of boron which produces a minimum concentration of 850 parts per million (ppm) by weight of natural boron in l the reactor core at 20*C. To allow for potential leakage and imperfect mixing in the I reactor system, an additional approximately 25% (220 ppm) is added to the above requirement, resulting in a total requirement ofgreater than or equal to 1079 ppm. The required concentration is thus achieved in a mass of water equal to the sum of the mass of water in the RPV at normal water level (equal to or less than 455 x 108 kg) plus the mass of water in the RPV shutdown cooling piping (equal to or less than 130 x 103 kg). The quantity of boron solution contained in the storage tank above the pump suction i shutoff level provides the required concentration of 1070 ppm when injected into the reactor, and this concentration will be achieved if the solution is prepared as defined in Subsection 9.3.5.2 and maintained above saturation temperature. 1 Cooldown of the nuclear system will require a minimum of several hours to remove the l thermal energy stored in the reactor, cooling water, and associated equipment. The controlled limit for the reactor vessel cooldown is 56'C/hr, and normal operating l temperature is approximately 288 C. Use of the main condenser and various shutdown cooling systems requires 10 to 24 hours to lower the reactor vessel to room temperature (21 C); this is the condition of maximum reactivity and, therefore,is the condition that l requires the maximum concentration of boron. The specified boron injection rate is limited to the range of 8 to 20 ppm / min. The lower rate assures that the boron is injected into the reactor in approximately two-and-one- l half hours. This resulting reactivity insertion is considerably quicker than that covered l by the cooldown. The upper limit injection rate assures that there is sufficient mixing ; so that boron does not recirculate through the core in uneven concentrations that I could possibly cause reactor power to rise and fall cyclically. The SLCS equipment essential for injection of neutron absorber solution into the reactor is designed as Seismic Category I for withstanding the specified earthquake loadings (Chapter 3). The system piping and equipment are designed, installed, and tested in accordance with the requirements stated in Section 3.6. The SLCS is required to be operable in the event of a plant offsite power failure; there-fore, the pumps, heater, valves, and controls are powered from the standby AC power supply. The pumps and valves are powered and controlled from separate buses and circuits so that a single active failure will not prevent system operadon. l The SLCS and pumps have sufficient pressure margin, up to the system relief valve l l setting of approximately 10.79 MPaG, to assure solution injection into the reactor above l the normal pressure in the bottom of the reactor.The nuclear system safety /reliefvalves [ l begin to relieve pressure above approximately 7.58 MPaG. Therefore, the SLCS positive i displacement pumps cannot overpressurize the nuclear system. 9.3-14 Process Auxiliaries- Amendment 34
I I 23A6100 Rev. 2 ABWR standsidsohrty Analysis neport O ( l 9.4 Air Conditioning, Heating, Cooling and Ventilating Systems J 9.4.1 Control Building HVAC l The Control Building (C/B) Heating, Ventilating and Air-Conditioning (HVAC) System is divided into two separate systems: (1) an HVAC System for the main control l area envelope within two floors, and (2) an HVAC System for safety-related elecuical l and RCW heat exchange equipment. l 9.4.1.1 Control Room Habitability Area HVAC 9.4.1.1.1 Design Basis (1) The control room habitability area (CRHA)HVAC System is designed with sufficient redundancy to ensure operation under emergency conditions assuming the single failure of any one active component. Independence is provided between Class IE divisions and also between Class 1E divisions and non-class IE equipment. (2) Provisions are made in the system to detect and limit the introduction of pl airborne radioactive materialin the main control area envelope (MCAE). V (3) Provisions are made in the system to detect and remove smoke and radioactive l material from the MCAE. (4) The control room habitability area HVAC System is designed to proside a controlled temperature environment to ensure the continued operation of safety-related equipment under accident conditions. (5) The control room habitability area HVAC System and components are located l in the Seismic Category I Control Building, a structure that is tornado-missile, and flood protected. (6) Tornado missile barriers and tornado dampers are provided at each intake and exhaust structure. (7) Protection from exterior smoke, toxic chemical and chlorine releases are discussed in Section 6.4. 9.4.1.1.2 Power Generation Design Basis (1) The control room habitability area HVAC System is designed to provide an environment with controlled temperature and humidity to ensure both the ( comfort and safety of the operators. The range of design conditions for the \ l control room environment are 21 C to 26 C and 10% to 60% reladve humidity. Air Conditioning, Heating, Cooling and Ventilating Systems - Amendment 32 9.4-1
I I 23A6100 R1v. 9 ABWR standard safety Analysis Report O (2) The system is designed to permit periodic inspection of the principal system components. (3) The outside design conditions for the control room habitability area HVAC System are 46*C during the summer and -40*C during the winter. 9.4.1.1.3 System Description The CRHA HVAC System consists of redundant divisions. Each division consists of an air conditioning unit (ACU) with two supply fans, two exhaust fans, and an emergency filtration unit with two circulating fans. The main control area envelope is heated, cooled and pressurized with filtered outdoor air mixed with recirculated air for ventilation and pressurization purposes. Under nc,rmal conditions, sufIicient air is supplied to pressurize the main control area envelope and the exfiltrate pressurizes the remainder of the Control Building. The control room habitability area ACU consists of two independent divisions, each with a medium efficiency bag filter, an electric heating coil, chilled water cooling coil, and humidifier. Two 100% capacity fans draw air from the instrument panel areas, corridors, main control room, computer room, office areas, and the switch and tag room. hiodulating dampers in the exhaust duct to the exhaust fans are controlled by a pressure controller to maintain the required 3.2 mm of water gauge positive pressure with respect to the atmosphere. The controller is located in the instrument panel area of the main control room. Normally one air conditioning unit, one supply fan and one exhaust fan are in operation. Redundant emergency air filtration divisions each consist of an electric heating coil, a prefilter, HEPA filter, charcoal adsorber, a HEPA filter, and two 100% capacity circulating fans treat mixed outdoor and return air before discharging it into the main l control area envelope. The charcoal adsorber will be 100 mm deep as a minimum. The emergency filtration unit supply fans are normally on standby for use only during high radiation conditions. A Process Radiation hionitoring System monitors two CRHA air intakes for radiation. The radiation monitors allow the control room operator to select manually one of the air intakes which are 50m apart. On receipt of a high radiation signal from the radiation monitors, only the corresponding emergency filtration unit of the operating division starts. The makeup air for pressurization can be treated by the HEPA and charcoal adsorbing system before distribution in the main control area envelope. The control room habitability area HVAC P&ID is shown in Figure 9.4-1. Flow rates are given in Table 9.4-3, and the system component descriptions are given in Table 9.4-4. Smoke detectors in the main control area envelope actuate an alarm on indication of smoke so the operators can place the system in the smoke removal mode manually. Air 9.42 Air Conditioning. Heating, Cooling and Ventilating Systems - Amendment 37
}
23A6100 Rsv. 4 ABWR studadsareryAutysisneput V ducts and air intakes are sized for 100% outdoor airflow. Dual smoke detectors in each CRHA HVAC system air intake detect and alarm on smoke originating outside the air intake. The CRHA IWAC system automaticallyisolates and is placed in full recirculation mode. Fire dampers with fusible links in the HVAC ductwork will close under air flow conditions after fusible link melts. 9.4.1.1.4 Safety Evaluation l The control room habitability area HVAC System is designed to maintain a habitable . emironment and to ensure the operability of components in the control room. All CRHA HVAC equipment and surrounding structures are of Seismic Category I design l and operable during loss of the offsite power supply. The ductwork which serves these safety functions is termed ESF ductwork, and is of ) Seismic Category I design. ESF ducting is high-pressure safety grade ductwork designed j to withstand the maximum positive and/or negative pressure to which it can be l subjected under normal or abnormal conditions. Galvanized steel (ASTM A526 or ASTM A527) is used for outdoor air intake and exhaust ducts. All other ducts are ( welded black steel ASTM A570, Grade A or Grade D. Ductwork and hangers are Seismic Category I. Bolted flange and weldedjoints are qualified per ERDA 76-21. Dhisions B and C equipment and ducts are separated except the common supply and common exhaust ducts sening the main control area envelope. Each emergency filtration
- division will utilize all welded construction for their charcoal trays and charcoal tray j screen to preclude the possible loss of charcoal from absorber cells per IE Bulletin 80-03.
l j Redundant and independent components are provided where necessary to ensure that a single failure will not preclude adequate main control area envelope ventilation.
- A four channel radiation monitoring system is provided to detect high radiation in the outside air intake ducts. A radiation monitoris provided in the control room to monitor control room area radiation levels. These monitors alarm in the control room upon
] detection of high radiation conditions. Isolation of the normal outdoor air intake j sening the emergency habitability area HVAC system control room and initiation of j outdoor air intake and the operating division emergency filtration unit and fan are accomplished by the following signals: (1) High radiation inside the air intake duct
; p (2) Manualisolation i
(~ Under normal conditions, sufficient air is supplied to pressurize the main control area
- envelope and exfiltrate to pressurize the remaining areas of the Control Building.
Air Conditioning, Heating, Cooling and Ventilating Systems - Amendment 34 9.4-3
l 23A6100 Rsv. 9 ABWR standard sarery Analysis Report The safety-related isolation valves at the outside air intakes are protected from O\ becoming inoperable due to freezing, icing, or other emironmental conditions. Upon detection of smoke in the CRIIA, the operating division of the HVAC System is l put into smoke removal mode by the main control room operators. For smoke removal, l both exhaust fans are started at high speed in conjunction with a supply fan, the l recirculation damper is closed. Either division of the CRHA HVAC System can be used l as a smoke removal system. 9.4.1.1.5 Inspection and Testing Requirements Provisions are made for periodic tests of the emergency filtration unit fans and filters. l These tests include measurement of differential pressure across the filter and of filter l efficiency. Connections for testing, such as injection, sampling and monitoring, are properly located so that test results are indicative of performance. , l The high-efficiency particulate air (HEPA) filters of the CRHA HVAC System shall be l tested periodically with dioctyi phthalate smoke (DOP). The charcoal filters will be , periodically tested with an acceptable gas for bypasses. Removal efficiency shall be at l l least 99% for all forms ofiodine (elemental, organic, particulate and HI, hydrogen iodide in the influent system). Each emergency filtration division duct work outside MCAE shall be periodically tested I for unfiltered inleakage in accordance with ASME N510. 1 Each emergency filtration division shall be periodically inspected for open maintenance access doors or deteriorated seals that could lead to charcoal filter bypass. The balance of the system is proven operable byits use during normal plant operation. Portions of the system normally closed to flow can be tested to ensure operability and integrity of the system. 9.4.1.1.6 Instrumentation Application One of two air conditioning unit supply fans is started manually. A high radiation signal automatically starts the emergency air filtration fan, closes the normal CRHA HVAC System air inlet dampers and closes the exhaust air dampers and stops the exhaust fan. A temperature indicating controller senses the temperature of the air leaving the emergency filtration system. The controller then modulates an electric heating coil to maintain the leaving air temperature at a preset limit. A limit switch will cause an alarm to be actuated on high air temperature. A moisture-sensing element, working in conjunction with the temperature controller, measures the relative humidity of the air entering the charcoal adsorber. 9.44 Air Conditioning. Heating, Cooling and Ventilating Systems - Amendment 37
3 I 23A6100 Rav. 9 l ABWR standardsafety Analysis Report ! ( \ ! Differential pressure indicators show the pressure drop across the prefilters and the
)
HEPA filters. The switch causes an alarm to be actuated if the pressure drop exceeds a I preset limit. A flow switch in the emergency filtration system fan discharge duct automatically starts the standby system and initiates an alarm on low flow or operating fan failure. ! 1 The main control area envelope exhaust fans start automatically when the air-conditioning unit supply fan is started. Each fan inlet damper is opened automatically. The return air dampers to the air-conditioning unit are opened automatically, j i Differential pressure-indicating controllers modulate dampers in the exhaust air ducts to maintain positive pressure of at least 3.2 mm water gauge. Manual start of an air conditioning unit supply fan provides a start signal to the HECW l pump and an interlock signal to open the cooling coil chilled water valve. A { temperature indicating controller installed in the MCR modulates the chilled water valve to maintain space temperatures. A moisture sensor controls the operation of a humidifier. The exhaust fan starts automatically when the supply fan starts. [ During winter, the electric unit heaters in the equipment rooms are cycled by i temperature-indicating control switches, located within the filter rooms and the air-conditioner rooms. j The supply, return and exhaust air ducts have manual balancing dampers provided in the branch ducts for balancing purposes. The dampers are locked in place after the system is balanced. 9.4.1.1.7 Regulatory Guide 1.52 Compliance Status j l The control room habitability area emergency filter units comply with all applicable 4 provisions of Regulatory Guide 1.52, Section C, except as noted below. ! The revisions of ANSI N509 and ANSI /ASME AG-1 listed in Table 1.8-21 are used for ABWR ESF filter train design; the Regulatory Guide references older revisions of these standards. 9.4.1.1.8 Standard Review Plan 6.5.1 Compliance Status ! l The control room habitability area emergency filtration units comply with SRP 6.5.1, I Table 6.5.1-1. I v Air Conditioning. Heating. Cooling and Ventilating Systems . Amendment 37 9.4-5 i
23AB100 Rsv. S ABWR standardsafety Analysis Report O 9.4.1.2 C/B Safety-Related Equipment Area HVAC 9.4.1.2.1 Design Basis (1) The C/B safety-related equipment area C/B SREA HVAC System is designed with sufficient redundancy to ensure operation under emergency conditions, assuming the failure of any one active component. (2) The C/B SREA HVAC System is designed to provide a controlled temperature I environment to ensure the continued operation of safety-related equipment I under accident conditions. (3) The C/B SREA HVAC System and components are Seismic Category I and are located in a Seismic Category I control building structure that is tornado-missile, and flood protected. (4) Tornado missile barriers and tornado dampers are provided at each intake and exhaust structure. (5) The rooms cooled by the C/B SREA HVAC System are maintained at positive pressure relative to atmosphere during normal and accident conditions. This is achieved by sizing intake fans larger than exhaust fans, i 9.4.1.2.2 Power Generation Design Basis (1) The C/B SREA HVAC System is designed to provide an environment with controlled temperature during normal operation to ensure the comfort and safety of plant personnel and the integrity of the safety-related electrical and RCW equipment, j (2) The system is designed to facilitate periodic inspection of the principal system components. (3) Design outside air temperature for the C/B HVAC System are 46 C during the summer and -40 C during winter. (4) Design inside air temperatures for the C/B safety-related equipment areas are 40 C maximum in the summer and a minimum of10 C in the winter. Battery rooms shall have suflicient air supply to keep the temperature between 10 C and 40*C. 9.4.1.2.3 Systern Description The C/B SREA HVAC System is divided into three independent subsystems with each l subsystem sening a designated divisional area for Divisions A, B, and C. Non-safety-related equipment is cooled by non-safety-related FCUs. 9.46 Air Conditioning, Heating, Cooling and Ventilating Systems - Amendment 35
I l l 23A6100 Rsv. 9 ABWR stadudserreyAnalysis neport i l One of the safety-related electrical equipment area exhaust fans starts automadcally , when the air-conditioning unit supply fan is started. I On a smoke alarm in a division of the Control Building safety-related electrical equipment area IWAC System, that disision of the IWAC System shall be put into smoke removal mode. No other division is affected by this action. For smoke removal, the recirculation duct damper is closed, and both exhaust fans are started in conjunction with a supply fan. Normal once through ventilation of the battery rooms also removes smoke from the battery rooms. Fire dampers separating electrical divisions II and IV rooms that use fusible links in HVAC ductwork will close under airflow conditions after fusible link melts. 9.4.2 Spent Fuel Pool Area HVAC System The Spent Fuel Pool Area IWAC System is part of the Reactor Building secondary containment HVAC System described in Subsection 9.4.5.1. 9.4.3 Auxilary Area HVAC System O The Auxilary Area INAC System is also part of the Reactor Building Secondary Containment HVAC System described in Subsection 9.4.5.1. 9.4.4 Turbine Island HVAC System The Turbine Island heating, ventilating, and air conditioning system consists of the Turbine Building (T/B) IIVAC System and the Electrical Building (E/B) HVAC System. 9.4.4.1 Design Bases 9.4.4.1.1 Safety Design Bases The T/B HVAC and E/B HVAC Systems do not serve or support any safety function and have no safety design bases. 9.4.4.1.2 Power Generation Design Bases (1) The T/B IWAC and E/B HVAC are designed to supplyfiltered and tempered air to all Turbine Island spaces during all modes of normal plant operation, including plant startup and shutdown. The systems are also designed to maintain inside air temperatures above 15 C and below the following upper design limits:
,C o Ceneral Turbine Building areas: 40 C s
u Condenser compartment: 43 C Air Conditioning, Heating, Cooling and Ventilating Systems - Amendment 37 9.4-9
1 23A6100 Riv. 4 ABWR StandardSafetyAnalysisReport O a Resin tank room: 43 C a Steam tunnel: 49 C a Moisture separator compartments: 49 C , i e Electrical Building areas: 40 C l (2) The E/B HVAC is designed to provide independent supply and exhaust i ventilation to the electrical switchgear, chillers and air compressor rooms, and i independent exhaust for the combustion turbine generator and auxiliarv ) boiler rooms. The ventilation exhaust for these areas is discharged directly to I the atmosphere. Recirculation from clean areas is provided. (3) The T/B HVAC is designed to direct airflow from areas oflow potential radioactivity to areas of high potential radioactivity. The T/B HVAC design is based on supplying air from the Turbine Building periphery (outer walls) both above and below the operating floor and ventilating areas radially inwards towards the return / exhaust air inlet points located below the operating floor in equipment rooms, the condenser area and under the building roof. The main stairwells that are designed for personnel evacuation routes are pressurized to prevent infiltration of smoke from other Turbine Building areas, during a fire. (4) The T/B HVAC is designed to minimize exfiltration by maintaining a slightly negstive pressure by exhausting 10% more air than is supplied to the Turbine Building. (5) Exhaust air from potentially high airborne concentrations in turbine building areas or component vents is collected, filtered and discharged to the atmosphere through the Turbine Building Compartment Exhaust (TBCE) System. (6) Exhaust air from other (low potential airborne concentrations) Turbine Building areas and component vents, except lube oil areas, is either exhausted to the atmosphere through a medium efficiency filter, or is returned to the supply air unit and mixed with outside air. (7) Exhaust air from the tube oil areas is exhausted to the atmosphere without filtration. (8) All Turbine Building exhaust air is directed tc the plant stack, where it is monitored for radiation prior to being discharged to the atmosphere. 9.4-10 Air Conditioning, Heating, Cooling and Ventilating Systems - Amendment 34 I l l
23A6100 Rcv. 9 ABWR standardsarety Analysis nepour t Design inside air temperatures for the secondary containment dudng normal operation is 40'C maximum in the summer and 10'C minimum in the winter. 9.4.5.1.2 System Description i The Reactor Building secondary containment HVAC System P&ID is shown m Figure 9.4-3. The system flow rates are given in Table 9.4-3, and the system component i thermal capacities are given in Table 9.44. The HVAC System is a once-through type. Outdoor air is filtered, tempered and delivered to the secondary containment. The l supply air system consists of filters, heating coils, cooling coils, and three 50% supply
- fans located in the Turbine Building. Two are normally operating and the other is on l
standby.The supply fan delivers conditioned air through ductwork and registers to the secondary containment equipment rooms and passages. The exhaust air system consists of 3 filter ands 50% capacity fans to be located in the Turbine Building. The exhaust fans pull air from the seconday containment rooms through du:twork, and filters. Monitors measure radioactivity before it is exhausted from the plant stack. HVAC air supply and exhaust used by the ACS for primag containment deinerting is discussed in Subsection 6.2.5.2.1(14) and the shutdown mode of operation in Subsection 6.2.5.2(3). Electric unit heaters are located in the large component entrance building. Supply air f N is directed into the space when the interior doors are open. 9.4.5.1.3 Safety Evaluation Operation of the Secondary Containment HVAC System is not a prerequisite to assurance of either of the following: (1) Integdty of the reactor coolant pressure boundary. 1 (2) Capability to safely shut down the reactor and to maintain a safe shutdown condition. However, the system does incorporate features that provide reliability over the full range of normal plant operation. The following signals automatically isolate the Secondary Containment HVAC System: (1) Seconday containment high radiation signal (LDS) (2) Refueling floor high radiation signal (LDS) (3) Drywell pressure high signal (LDS) (4) Reactor water level low signal (LDS) 7 i
.] (5) Secondary containment HVAC supply / exhaust fans stop Air Conditioning, Heating, Cooling and Ventilating Systems - Amendment 37 9.4 17
23A6100 Rzv. 8 ABWR standardsafety Analysis Report On a smoke alarm in a division of the secondary containment HVAC System, the HVAC 9 System shall be put into smoke removal mode. To remove smoke from the secondary containment, the exhaust filter by-pass dampers are opened, standby exhaust and supply fans are started to provide an increase in airflow through the secondary containment. The divisions that are not on fire shall have their exhaust dampers closed to a partially closed posidon. This position shall be set during system setup. When the ) exhaust dampers are partially closed, the non-fire divisions' pressure will be maintained at a positive pressure. The division experiencing the fire will be maintained more ! I negative with respect to the non-fire divisions. Fire zone dampers can isolate the division with the fire undt smoke removal is required. I When fire doors are opened between divisions, the air pressure in the non-fire zones will limit smoke intmsion. Fire dampers with fusible links in HVAC ductwork will close ; under airflow conditions after fusible link melts. ! 9.4.5.1.4 Inspection and Testing Requirements l l The system is designed to permit periodic inspection ofimportant components, such as l fans, motors, belts, coils, filters, ductwork, dampers, piping and valves, to assure the ! integrity and capability of the system. Standby components can be tested periodically to i ensure system availability. All major components are tested and inspected as separate components prior to installation and as integrated systems after installation, to ensure design performance. l The system is preoperationally tested in accordance with the requirements of j Chapter 14. 9.4.5.1.5 Instrumentation Application The Secondary Containment FIVAC System is started manually. Fan inlet dampers are interlocked to open before the fan is started. A flow switch installed in the operating fans discharge ductwork automadcally starts the standby fan on indication of any operating fan failure due to a reduction in air. The pneumatically-operated secondary containment inboard and outboard isolation dampers fail to the closed position in the event ofloss of pneumatic pressure or loss of electrical power to the valve actuating solenoids. Upon receiving a leak detection system signal (Subsection 9.4.5.1.3), the isolation dampers automatically close, supply and exhaust fans stop, and a start signal calls for automatic SGTS operation. The supply fans ! and exhaust fans are interlocked to prevent operation of the supply fans when the j exhaust fans are shut down. e 9.4-18 Air Conditioning, Heating. Cooling and Ventilating Systems - Amendment 36
23A6100 Rsv. 9 ABWR standardSafetyAnalysis Report U 9.4.5.2.3 Safety Evaluation ! 1 All equipment is located completely in a Seismic Categog I structure that is l tomado-missile, and flood protected. All equipment is designed to Engineered Safety Feature requirements. 9.4.5.2.4 Inspection and Testing Requirements All major components are tested and inspected as separate components prior to installation to ensure design performance. The system is preoperationally tested in j accordance with the requirements of Chapter 14. Each HVAC System is periodically tested to assure availability upon demand. Equipment layout provides easy access for inspection and testing. 9.4.5.2.5 Instrumentation Application l Instrumentation and controls for the Secondary Containment Safety-Related Equipment HVAC System are designed for manual or automatic operation when safety-related equipment starts. Also, manual override from pushbutton stations in the main e control room or at the MCC sening the unit. (
'A 9.4.5.3 Reactor Building Non-Safety-Related Equipment HVAC System 9.4.5.3.1 Design Bases l
9.4.5.3.1.1 Safety Design Bases The Non-safety-related Equipment HVAC System has no safety-related function as defined in Section 3.2. Failure of the system does not compromise any safety-related l component and does not prevent safe reactor shutdown. 9.4.5.3.1.2 Power Generation Design Bases The Non-safety-related Equipment HVAC System is designed to proside an emironment with controlled temperature and humidity to insure both the comfort and safety of plant personnel and the integrity of equipment and components. ! 9.4.5.3.2 System Description l The R/B Non-Safety Related HVAC System consists of six air handling units. The
- following rooms are cooled by the HVAC System
l l
/N (1) ISI room Air Conditioning, Heating, Cooling and Ventilating Systems - Amendment 37 9.4-21
23 ASS 00 Rsv. 9 ABWR standardSafetyAnalysis Report O (2) CRD control room I
)
(3) SPCU pump room i (4) Refueling machine control room (5) R/B Fuel pool cooling unit A l (6) R/B Fuel pool cooling unit B These rooms are cooled by the Secondary Containment HVAC System during normal conditions. The units are open ended and recirculate cooling air within the space served. Space heat is removed by cooling water passing through the coil section. HVAC i normal cooling water or divisional RCW is used as the cooling medium. The units are fed from the non-divisional power source. Humidity is not specifically maintained at a set range, but is automatically determined by the surface temperature of the cooling coil. Drain pan discharge (condensate) is routed to a drain sump located within the room.
)
9.4.5.3.3 Safety Evaluation Operation of the R/B Non-safety-related Equipment HVAC System is not a prerequisite to assurance of either of the following-(1) Integrity of the reactor coolant pressure boundary , (2) Capability to safely shut down the reactor and to maintain a safe shutdown Condition i However, the system does incorporate features that provide reliability over the full j range of normal plant operations. j 9.4.5.3.4 Inspection j i The system is designed to permit periodic inspection ofimportant components, such as fans, motors, belts, coils, and valves, to assure the integrity and capability of the system. Ol! l j 9.4-22 Air Conditioning, Heating, Cooling and Ventilating Systems - Amendment 37 l
23A6100 Rsv. 9 ABWR standardsafety Analysis Report O All major components are tested and inspected as separate components prior to installation to ensure design performance. The system is preoperationally tested in accordance with the requirements of Chapter 14. 9.4.5.3.5 Instrumentation Application l The R/B Non-safety-related Equipment HVAC System starts manually. 9.4.5.4 R/B Safety Related Electrical Equipment HVAC System 9.4.5.4.1 Design Bases 9.4.5.4.1.1 Safety Design Bases 1 The R/B Safety-Related Electrical Equipment HVAC System is designed to proside a l controlled temperature environment to ensure the continued operation of safety- l related equipment under accident conditions. The rooms cooled by the R/B Safety- l Related Electrical Equipment HVAC System are maintained at positive pressure relative to atmosphere during normal and accident conditions. This is achieved by sizing intake fans larger than exhaust fans. [ The power supplies to the HVAC systems for the R/B safety-related electrical D equipment rooms allow uninterrupted operation in the event ofloss of normal offsite power. The system and components are located in a Seismic Category I structure that are tornado-missile, and flood protected, including tornado missile barriers on intake and exhaust stmctures. For compliance with code standards and regulatory guides, see Sections 3.2 and 1.8. On a smoke alarm in a d'. vision of the Reactor Building Safety-Related Electrical Equipment HVAC System, that division of the HVAC System shall be put into smoke removal mode manually. No other division is affected by this action. For smoke removal, the recirculation damper is closed, the exhaust fan bypass damper opened, the exhaust fan is stopped, and the smoke removal fan is started in conjunction with the supply fan.- Normal once through ventilation of the day tank rooms also removes smoke from the day tank rooms. The intake louvers are located at 15.2m above grade. The exhaust louvers are located at 13.3m above grade. (See general arrangement layout, Figures 1.2-10 and 1.2-11.) 9.4.5.4.1.2 Power Generation Design Bases A The system is designed to provide an emironment with controlled temperature and humidity to ensure both the comfort and safety of plant personnel and the integrity of safety-related electrical equipment. The system is designed to facilitate periodic inspection of the principal system components. Air Conditioning. Heating. Cooling and Ventilating Systems - Amendment 37 9k23
23A6100 Rtv. 8 ABWR standardsarety Analysis neport O l The system design is based on outdoor summer conditions of 46.1*C and outdoor winter conditions of-40 C. The indoor design temperature in the safety-related electrical equipment areas is 40 C maximum in the summer and a minimum of 10 C in l the winter except 50 C in the diesel generator (DG) engine rooms during DG operation. The system along with the DG supply fan maintain DG room temperature l below 50*C. 9.4.5.4.2 System Description Divisions A, B, and C Safety-Related Electrical Equipment HVAC Systems are independent, physically separated, and functionally identical except for their power bus designations and divisional source of cooling water. The HVAC System for each disision of safety-related electrical equipment consists of two 100% capacitysupply fans, two 100% capacity exhaust fans, and one air conditioning unit. Each air conditioning unit consists of a medium grade filter and a cooling coil. (See Figure 9.4-4 for the system P&ID. See Table 9.4-4 for the component descriptions.) The following divisional rooms are cooled by the Safety-Related Electrical Equipment HVAC System : (1) Day tank room, Dinsions A, B, C (2) Diesel generator engine room, Divisions A, B, C (3) Non-safety-related reactor internal pump ASD rooms (4) Electrical equipment room, Divisions I, II, III, IV (5) HVAC equipment room, Divisions A, B, C (6) Remote shutdown panel room, Divisions A, B 1 (7) Diesel generator MCC area, Divisions A, B, C l l (8) Non-Safety-Related FMCRD control panel rooms l HVAC system Division A serves electrical Division I, Division B serves electrical Disisions II and IV, and Division C serves electrical Division III of the electrical equipment rooms. Also, non-safety-related reactor internal pumps ASD rooms are cooled by the Electrical Equipment HVAC system. l 9.4.5.4.3 Safety Evaluation All safety-related equipment is located in a Seismic Category I structure that is tornado-missile, and flood protected. All IIVAC equipment is designed to Engineered Safety Feature requirements. 9.4.5.4.4 Inspection and Testing Requirements The systems are designed to permit periodic inspection ofimportant components, such as fans, motors, coils, filters, ductwork, dampers, piping, and valves to assure the mtegrity and capability of the system. Standby components can be tested periodically to ensure system availability. 9.4 24 Air Conditioning, Heating, Cooling and Ventilating Systems - Amendment 36
23A6100 R1v. 4 ABWR standantsaferyAnalysisneport G minimize release of radioactive substances to the atmosphere and to prevent operator exposure. The Radwaste Building HVAC System P&ID is shown in Figure 9.4-10. 9.4.6.1.2 Power Generation Design Bases The Radwaste Building HVAC System is designed to provide an environment with controlled temperature and airflow patterns to insure both the comfort and safety of plant personnel and the integrity of equipment and components. The Radwaste Building is divided into two zones for air conditioning and ventilation purposes. These zones are the radwaste control room and the balance of the Radwaste Building. A positive static pressure with respect to the balance of the building and to the atmosphere is maintained in the radwaste control room. The balance of the Radwaste Building is maintained at a negative static pressure with respect to the atmosphere. The system design is based on an outdoor summer maximum of 46 C. Summer indoor temperatures include 24 C in the radwaste control room,32'C in operating areas and corridors, a maximum temperature of 40 C in areas that may be occupied and 43'C in the equipment cells. Winter indoor design temperatures include 16 C in occupied ( (, areas,21'C in the radwaste control room and 16 C in the equipment cells, based on an outdoor design temperature of-40*C. w 9.4.6.2 System Description The COL applicant will provide an equipment list and system flow rates including l RG 1.140 compliance for NRC resiew (Subsection 9.4.10.2). 9.4.6.2.1 Radwaste Building Control Room Heating, cooling and pressurization of the control room are accomplished by an air-conditioning system. The air-conditioning system is a unit air-conditioner consisting of : a water-cooled condenser, compressor, cooling coil, heating coil, filters and fan. Outdoor air and recirculating air are mixed and drawn through a prefilter, a high efficiency filter, a heating coil, a cooling coil, and two 100% supply fans. One fan is normally operating and the other fan is on standby. A pressure differential controller regulates the exfiltration from the control room to maintain it at a positive static pressure, preventing airborne radioactive contamination from entering. No separate exhaust fan system is required. The Radwaste Control Room HVAC Smoke Removal System consists of one 100% fan. l This fan is operated manually. Smoke from the control room is released directly to the atmosphere. C\ h An area radiation monitor is provided in the radwaste control room and will alarm on high radiation to alert personnelin the area. Air Conditioning, Heating, Cooling and Ventilating Systems - Amendment 34 9.4-31
23A6100 Rtv. 9 ABWR StandardSafety Analysis Report , O 9.4.6.2.2 Radwaste Building Process Area HVAC System i The Radwaste Building Process Area HVAC System is a once-through type. Outdoor air is filtered, tempered and delivered to the non-contaminated areas of the building. The supply air system consists of a prefilter, a high efficiency filter, heating coil, cooling coil, l and two 100% supply fans. One fan is normally operating and the other fan is on j standby. The supply fan furnishes conditioned air through ductwork and diffusers, or registers to the work areas of the building. Electric unit heaters are prosided in the I trailer bays and the sorting table area. Air from the work areas is exhausted through the I tank and pump rooms. Thus, the overall airflow pattern is from the least potentially contaminated areas to the most contaminated areas. The exhaust air system consists of three 50% exhaust fans, two normally operating and one on standby. Exhaust air from the Radwaste Building is filtered through a prefilter and a high efficiency filter before release to the plant stack and it is monitored for airborne radioactivity. A high level of radioacthity activates an alarm in the main control i room, simultaneously isolating the process area. The exhaust air is monitored before it is released to the main plant stack. 9.4.6.3 Safety Evaluation Although the IWAC System is not safety-related as defined in Section 3.2, several features are provided to insure safe operation. A completely separate HVAC System is l provided for the radwaste control room. Pressure control fans for radwaste areas are redundant, with provision for automatic start of the standby unit. Area and process j exhaust radiation detectors and isolation dampers are provided to permit isolation of the radwaste process areas. 9.4.6.4 Tests and inspections The system is designed to permit periodic inspection ofimportant components, such as fans, motors, belts, coils, filters, ductwork, piping and valves, to assure the integrity and capability of the system. Local display and/or indicating devices are provided for periodic inspection of vital parameters such as room temperature, and test connections are provided in exhaust filter trains and piping for periodic checking of air and water flows for conformance to the design requirements. All major components are tested and inspected as separate components prior to installation to ensure design performance. The system is preoperationally tested in accordance with the requirements of Chapter 14. 9.4.6.5 Instrumentation Application 9.4.6.5.1 Radwaste Building Control Room The air <onditioning unit for the radwaste control room IWAC is started manually. A l temperature indicating controller modulates the air-conditioning system via chilled 9.4-32 Air Conditioning, Heating, Cooling and Ventilating Systems - Amendment 37
l I l 23A6100 R1v. 9 l ABWR StandardSafety Analysis Report f i l water valves and an electric heating coil to maintain space conditions. A differential pressure indicating controller modulates inlet vanes in the supply fan air inlets to maintain the positive static room pressure. Differential pressure indicators measure the pressure drop across the filter bank. 9.4.6.5.2 Radwaste Building Process Area HVAC The air exhaust and supply fans for the Radwaste Building Process Area HVAC are started manually. The fan inlet dampers open when the fan is started. A flow switch installed in the exhaust fan discharge duct actuates an alarm on indication of fan failure in the main and radwaste control rooms and automatically starts the standby fan. The exhaust fan is interlocked with the supply fan to prevent the supply fan from operating if the exhaust fan is shut down. Two pressure-indicating controllers modulate variable inlet vanes in the supply fan to maintain the area at a negative static pressure with respect to the atmosphere. The switch causes an alarm to be actuated if the negative pressure falls below the preset limit. Differential pressure indicators measure the pressure drop across the filter section. The 3 switch causes an alarm to be actuated if the pressure drop exceeds the preset limit. Radiation monitors are installed in the radwaste process area exhaust duct to the main plant stack. A high radiation signalin the duct causes alarms to annunciate in the main control room and the radwaste control room. If the radwaste process area exhaust radiation alarm continues to annunciate, the work area branch ducts are manually isolated selectively to locar.e the affected building area. Should this technique fail, because the airborne radiation bas generally spread throughout the building, control room air conditioning continuenperating. However, the air conditioning for the balance of the building is shut down. Th- operators, using approved plant health physics procedures, then enter the work areas to locate and isolate the leakage source. The supply and exhaust air ductwork have manual balancing dampers provided in the branch ducts for balancing purposes. The dampers are locked in place after the system is balanced. 9.4.6.5.3 incinerator Exhaust Stack Radiation monitors are installed in the incinerator exhaust stack. A high radiation signalin the stack causes alarms to annunciate in the main control room and the radwaste control room. See Subsection 11.5.2.2.11 and Table 11.5-2.
%)
Air Conditioning, Heating, Cooling and Ventilating Systems Amendment 37 9.4-33
23A6100 Rett. 4 AMWR Standard SafetyAnalysis Report O 9.4.7 R/B Safety-Related Diesel Generator HVAC System l The safety-related Diesel Generator HVAC System is part of the Reactor Building HVAC System described in Subsection 9.4.5.5. 9.4.8 Service Building HVAC System This system serves all areas within the Senice Building, including locker rooms, men and women's change rooms, laundry, lunch room, instrument repair room, HVAC eqtepment rooms, and the Technical Support Center (TSC). This system operates during all normal station conditions. The Senice Building HVAC System consists of two subsystems; the Clean Area HVAC System and the Controlled Area HVAC System. 9.4.8.1 Design Basis 9.4.8.1.1 Safety Design Basis l The Senice Building HVAC System is not required to function in any but the normal station operating conditions and. therefore, has no safety bases. 9.4.8.1.2 Power Generation Design Bases (1) The Clean Area HVAC System is designed to maintain a quality emironment l suitable for personnel health and safety in the Senice Building. It is designed to limit the maximum temperature in the Senice Building to 29 C. The temperature in each area conforms to the equipment requirements in that area. l (2) The Clean Area HVAC System prosides a quantity of filtered outdoor air to purge any possible contamination. l (3) Both the Clean Area HVAC System and the Controlled Area HVAC System operate manually and continuously. Isolation dampers at each supply fan, each exhaust fan, and each filter package close when the respective equipment is not operating. There is an additional isolation damper at the supply air inlet which closes when the supply air system is not operating. An automatic damper in the supply system ductwork regulates the flow of air to l maintain the Senice Building clean areas at a positive pressure with respect to the atmosphere. (4) In the event of a loss of offsite electric power, the Senice Building HVAC System is shut down. 9.4-34 Air Conditioning, Heating. Cooling and Ventilating Systems - Amendment 34
1 i 23A6100 Rtv. 9 ABWR StandardSafetyAnalysisReport m
- U Table 9.4-4 HVAC System Component Descriptions - Safety-Related
, Heating / Cooling Coils (Response to Question 430.243) 1 Cooling Heating l j Heating / Cooling Coils Quantity (MJ/h) (MJ/h) R/B Electrical HVAC Division A 1 675.25 No Coil Required R/B Electrical HVAC Division B 1 675.25 No Coll Required R/B Electrical HVAC Division C 1 675.25 No Coil Required C/B Electrical HVAC Division A 1 886.26 No Coil Required C/B Electrical HVAC Division B 1 886.26 No Coil Required a j C/B Electrical HVAC Division C 1 886.26 No Coil Required 4 CRHA HVAC Division B 1 662.61 591.59 ! CRHA HVAC Division C 1 662.61 591.59 l CRHA Emergency HVAC Division B 1 - 252 l j l CHRA Emergency HVAC Division C 1 - 252 d i i l j I e i 1 l Air Conditioning Heating, Cooling and Ventitating Systems - Amendment 37 9.4-43
l l l 23A6100 Rsv. 9 ABWR standardsateryAnalysn, port l O l l Table 9.4-4a HVAC System Component Descriptions - Safety-Related Fans l (Response to Question 430.243) Capacity Rated i Fans Quantity (m3 /h) Power (kW) R/B Electrical Div A Supply Fans 2 (1 on standby) 30,000 75 l R/B Electrical Div B Supply Fans 2 (1 on standby) 30,000 75 R/B Electrical Div C Supply Fans 2 (1 on standby) 30,000 75 R/B Electrical Div A Exhaust Fans 2 (1 on standby) 6,000 4 R/B Electrical Div B Exhaust Fans 2 (1 on standby) 6,000 4 l R/B Electrical Div C Exhaust Fans 2 (1 on standby) 6,000 4 DG Div A Supply Fans 2 80,000 22 DG Div B Supply Fans 2 80,000 22 DG Div C Supply Fans 2 80,000 22 C/B Electrical Div A Supply Fans 2 (1 on standby) 35,000 75 C/B Electrical Div B Supply Fans 2 (1 on standby) 35,000 75 l C/B Electrical Div C Supply Fans 2 (1 on standby) 35,000 75 l C/B Electrical Div A Exhaust Fans 2 (1 on standby) 4,000 4 C/B Electrical Div B Exhaust Fans 2 (1 on standby) 4,000 4 C/B Electrical Div C Exhaust Fans 2 (1 on standby) 4,000 4 CRHA Div B Supply Fans 2 (1 on standby) 80,000 2': CRHA Div C Supply Fans 2 (1 on standby) 80,000 22 CRHA Div B Exhaust Fans 2 (1 on standby) 5,000 4 CRHA Div C Exhaust Fans 2 (1 on standby) 5,000 4 l CRHA Emergency Div B Filter Supply Fan 2 (1 on standby) 5,100 7.5 l CRHA Emergency Div C Filter Supply Fan 2 (1 on standby) 5,100 7.5 9,4-44 Air Conditioning, Heating. Cooling and Ventilating Systems - Amendment 37
l 23A6100 R&v. 9 ABWR standardsafety Analysis Report O (G Table 9.4-4b HVAC System Component Descriptions - Safety Related Filter (Response to Question 430.243) Filters Quantity Capacity (m3 /h) R/B Electrical Div A filter 1 35,000 l R/B Electrical Div B Filter 1 35,000 R/B Electrical Div C Filter 1 35,000 l DG Div A Filter 1 200,000 - DG Div B Filter 1 200,000 DG Div C Filter 1 200,000 C/B Electrical Div A Filter 1 40,000 C/B Electrical Div B Filter 1 40,000 l g C/B Electrical Div C Filter 1 40,000
\
CRHA Div B Filter 1 80,000 CRHA Div C Filter 1 80,000 Table 9.4-4c HVAC System Component Descriptions - Emergency Use Adsorption Units (Safety Related) (Response to Question 430.243) Emergency Use Adsorption Unit Quantity Capacity (m3 /h) l CRHA Emergency Div B Filter 1 5,100 l CRHA Emergency Div C Filter 1 5,100 fh b Air Conditioning. Heating. Cooling and Ventilating Systems Amendment 37 9.4-45
l l 23A6100 Rev. S ABWR Standard SafetyAnalysisReport O l Table 9.4-4d Deleted l l 1 l l l Table 9.4-4e HVAC System Component Descriptions - Safety-Related
- Fan Coil Units (Response to Question 430.243)
I l Safety-Related Fan Coll Units Capacity (MJ/h) HPCF Pump Room Div 8 460.55 HPCF Pump Room Div C 460.55 RHR Pump Room Div A 307.73 RHR Pump Room Div B 307.73 RHR Pump Room Div C 307.73 l RCIC Pump Room Div A 69.08 i l FCS Room Div B 54.85 FCS Room Div C 54.85 CAMS Room Div A 83.74 CAMS Room Div B 83.74 SGTS Room Div 8 16.75 SGTS Room Div C 16.75 O 9.4 46 Air Conditioning. Heating, Cooling and Ventilating Systems - Amendment 35 l
23A6100 Rev. 9 ABWR standard safety Analysis Report , l Table 9.4-4f HVAC System Component Descriptions-Non-Safety-Related Heating Cooling Coils (Response to Question 430.243) Cooling Heating Heating / Cooling Colis Quantity (MJ/h) (MJ/h) l R/B Secondary Containment HVAC 3 (1 on standby) 6435.95 9601.17 RIP ASD HVAC Division A 1 2110.15 RIP ASD HVAC Division B 2110.15 1
)
Table 9.4-4g HVAC System Component Descriptions-Non-Safety-Related Fans (Response to Question 430.243) I I Fans Quantity Capacity (m3 /h) )
. R/B Secondary Containment Supply Fans 3 (1 on standby) 84,250 t j R/B Secondary Containment Exhaust Fans 3 (1 on standby) 86,250 R/B Primary Containment Supply Fan 1 22,000 R/B Primary Containment Exhaust Fan 1 22,000 RIP ASD Division A Suppiy Fans 2 (1 on standby) 50,000 RIP ASD Division B Supply Fans 2 (1 on standby) 50,000 I
Table 9.4-4h HVAC System Component Descriptions-Non Safety-Related Filters l (Response to Question 430.243) l Filters Quantity Capacity (m3/h) l R/B Secondary Containment HVAC 3 (1 on standby) 86,250 ! R/B Primary Containment Intake HEPA Filter 1 22,000 l n R/B Secondary Containment Exhaust Fans 3 57,500 (each) Air Conditioning, Heating, Cooling and Ventilating Systems - Amendment 37 9.4~17
23A6100 Rsv. 9 ABWR StandardSafety Analysis Report l Table 9.4-4i HVAC System Component Descriptions-Non-Safety Related Air Handling Units (Response to Question 430.243)
- l Non-Safety Related Air Handling Units Quantity Capacity (MJ/h)
Main Staam Tunnel 2 628.02
- Refueling Machine Control Room 1 83.74 l ISI Room 1 54.43 l MG Set Room 2 1047.96 l C/B Non-Safety-Related Electric Room 1 211.01 l R/B FPC Room 2 28.47 l CRD Control Room 1 18.42 I
I I l SPCU Pump Room 1 42.29 4
- The COL applicant shall supply equipment lists for the Service Building HVAC and the Radwaste Building HVAC System. See Subsection 9.4.10.1 for the Service Building, and 9.4.10.2 for the Radwaste Building.
4 i O 9.4M Air Conditioning, Heating, Cooling and Ventilating Systems - Amendment 37
23A6100 R1v. 9 ABWR standardsafetyAnalysis Report n U (2) Fire barrier walls which are of the special construction described in Subsection 9A.S.6 or of other approved construction types bearing a UL (or equal) label for a three-hour rating. 1 (3) Fire doors, which are required to have a UL (or equal) label certifying that they have been tested for a three hour-rating per ASTM E152, including a hose stream test. l (4) Both ends of all electrical and piping penetrations between the divisions and I between a division and a non-division should be qualified to the same standard and tested to ASTM E119. (5) Deleted (6) Fire dampers, which are required for any HVAC duct penetrating a fire. barrier, must have a rating of three hours. The plant arrangement minimias fire dampers. l p (7) Columns and support beams, which are required to be of reinforced concrete ( constniction or enclosed or coated to provide a three-hour rating if of steel construction. (8) Backup of the fire barrier penetration seals by the HVAC Systems when the HVAC Systems are operating in the smoke removal mode. This backup feature is accomplished in the Reactor and Control Buildings by maintaining a positive static pressure for the redundant divisional fire areas with respect to the fire area with the fire. Leakage is into the fire impacted area under sufficient static pressure to confine smoke and heat to the fire area experiencing the fire, even if there is a major mechanical failure of the penetration seal. (9) AC independent water addition (ACIWA) can be connected to the reactor building fire protection system header. Sufficient Fire Water pressure and flow should be available to perform the intended function. Refer to Subsection 5.4.7.1.1.10, AC-Independent Water Addition. 9.5.1.1.4 Combustible Loading Allowable combustible loadings for the plant were established as follows (see Appendix 9B, Subsection 9B.2.3 for additional details): A ( (1) 1454 MJ/m 2of room area, maximum allowable average exposed combustible loading without an automatic fire suppression system. This is termed the normal combustible loading limit (NCLL). Other Auxiliary Systems - Amendment 37 9.5-9
23M100 Rw. 4 ABWR standard saretyAnalysis neport 2 O l (2) 2908 MJ/m of room area, maximum average allowable exposed combustible loading (cable insulation) for electrical equipment rooms. This is termed the electrical room combustible loading limit (ECLL). Transient combustible loadings other than minor amounts required for maintenance of the equipment in the electrical equipment rooms are not allowed. (3) Deleted (4) Deleted (5) Deleted (6) Transient combustibles such as lubricating oils and grease, cleaning solvent, etc. in small quantities and in approved containers are permitted. (7) Transient combustibles such as bags of protective clothing are permitted within the constraints ofitems (1) and (2). Ifit can be shown by analysis or testing that flames from the burning transient combustibles will not likely impinge directly on cables within trays or risers, the contribution from the cable insulation need not be considered in calculating the total combustibles in the area with the transient combustible load. Combustible loading due to cable insulation has been minimized bylocating the power sources adjacent to the loads sened and multiplexing the control signals to and from the control room. This has allowed the elimination of cable spreading rooms and most of the cables to and from the control room. Multiplexing is also used within the control room so that the cables between panels have been reduced to mostly power cables. 9.5.1.1.5 HVAC Systems The HVAC Systems have been matched to the divisional areas which they sene. For example, there are three divisions of power supply outside of secondary containment in the Reactor Building. The divisions cre in separate fire areas and each fire area is sen ed byits corresponding dhision of the HVAC System. This same philosophy is carried into the Control Building, where there is a divisional HVAC System for each of the three divisions of core cooling systems and a fourth HVAC System for the control room complex which contains four dhisions of control equipment. Division 4 contains a battery as the source of power but there is not a Division 4 diesel generator. Division 4 loads are comprised ofinstrumentation and controls to provide two-out-of-four logic. Loss of Division 4 reverts the logic back to two-outef-three, which is acceptable on a permanent basis from a safety standpoint. The main function sened by the Division 4 control and logic is one ofimproving plant availability for power production. On this basis Dhision 4 is supplied cooling from Division 2 in both the reactor and control buildings. 9.5-10 Other Auxiliary Systems - Amendment 34
4 23A6100 Rtv. 9 ABWR standardsafetyAnalysis Report l j A single non-safety-related HVAC System supplies normal cooling for secondary containment in the Reactor Building. Within the Reactor Building the system is
; branched into three separate systems with valves and fire dampers for each branch (Subsection 9.5.5). Required emergency cooling for safety-related systems is provided j by room coolers on a divisional basis.
)
, 9.5.1.1.6 Smoke Control System ] The smoke control system for the plant provides major features as follows: 1 (1) Venting of fire areas to prevent undue buildup of pressure due to a fire. l 1 (2) Pressure control across the fire barriers to assure that any leakage is into the j fire area experiencing the fire. , (3) Pressure control and purge air supply to prevent back-flow of smoke and hot gases when fire barrier doors are maintained open for access for manual fire j suppression activities.
- (4) Augmented and directed clean air supply to provide a clean air path to the fire i for fire suppression personnel.
j (5) Smoke control by fans and systems external to the fire area experiencing the fire. (6) Removal of smoke and heat from the fire by exhaust fans and operating supply j fans to provide clean, cool air. f l (7) Manually reset position of fire dampers in the smoke removal path. These features are provided by designing the HVAC Systems for the dual purpose of HVAC and smoke control. ASHRAE's " Design of Smoke Control Systems for Buildings" and NFPA's " Recommended Practice for Smoke Control Systems" (References 9.5-3
- and 9.5-4) were used as the basis for the design of the smoke control features of the combined systems.
i The normal operating modes of the HVAC Systems are shown in Figures 9.4-1 through , 9.4-6. The pressun. at the input of an air handling unit is held at atmospheric pressure by a ducted, direct anpply from outside through a bag type filter. The systems are designed so the di 4sion of the air flow to the rooms within an HVAC/ fire area is determined by the supply and exhaust ducting and the adjustable O volume dampers. i Other Auxiliary Systems - Amendment 37 9.5 11 y
I 23A6100 Rav. 9 ABWR standard sauty Analysis Report l l The IWAC Systems in the fire areas not experiencing a fire continue to operate in their normal fashion so that the pressure in the other fire areas remains at a positive value. This assures that air leakage through any openings in the fire barriers surrounding the fire is to the fire. The magnitude of the differential pressure which must be maintained across a fire barrier to provide adequate smoke control varies with the intensity of the fire and the room height. For this reason, it is a COL license information requirment (Subsection 9.5.13.10) that the required differential pressure value for each barrier be calculated j during the detailed design phase, the HVAC Systems be designed to proside the required pressure, and that the capability be confirmed during premperational testing. Normally the differential pressure woulld not have to be more than about 6A mm of water, and it most likely would be less. Entry to a fire is gained from an adjacent fire area which by design is at a positive pressure with respect to the area experiencing the fire. The pressure differential is sufficient to provide adequate velocity through the open door to carry the combustion products back into the zone of the fire. The flow through the open door into the area of the fire and out the area of the fire's exhaust duct system is maintained by the positive pressure of the non-fire area and by the operation of smoke removal mode in the fire area. It gives the fire fighting squad a tenable emironment from which to work. 9512 Other Auxiliary Systems - Amendment 37
l i l i 23A6100 Rw. 9 ABWR standantsafery Analysis nopet A i There are fire dampers in the HVAC penetrations of building internal walls between safety-related fire areas. l Upon manual initiation of the smoke removal mode, the recirculation damper is closed, i the exhaust fans are stopped, and the smoke removal fan is started in conjunction with the supply fan for 100% outside air purging. In the Control Building, the recirculation l damper is closed and both the exhaust fans are operated in conjunction with the supply fan for smoke removal. l In order to maintain the objective of smoke and heat removal during a fire situation, C ! the HVAC supply and exhaust duct openings in the exterior walls of the Reactor building do not have fire dampers. The walls are designated as three-hour fire barriers and would normally require fire dampers for HVAC duct penetrations. Fire dampers could close due to heat from an internal fire, however. Internal fires are a more serious threat to the plant than external fires. Omission of the fire dampers in the supply ducts is deemed acceptable because: 0 (1) Each HVAC/ fire area has a separate intake structure. (2) The intake structures are dispersed around the perimeters of the buildings. (3) Deleted 4 (4) Isolation valves are provided and could be manually closed should there be a challenge due to an ext.crnal fire. (5) Each intake serves one fire area and, therefore, one division only except for the control room. The two redundant divisions are in separate fire areas. The control room fire area is separate from all other fire areas and the safe shutdown function is backed up by the remote shutdown panel. 2 Omission of the fire dampers in the exhaust ducts is deemed acceptable because: t
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(1) Each HVAC/ fire area has a separate exhaust. Other Auxiliary Systems - Amendment 37 9.5 13
23A6100Rw 4 ABWR Standard SafetyAnalysisReport O (2) Flow is normally out of the building so that external combustion products l would not be drawn into the building. l l (3) Isolation valves are provided and could be manually closed ifit becomes l necessary to shut a HVAC System down during an external fire situation. l (4) The run of metal duct from the exhaust structure to the isolation valves will l contain stagnant air which will protect the isolation valves from high temperatures due to external fires if the valves are closed. The isolation valves
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will have far less leakage than normal fire dampers. It is an interface l requirement that the exhaust duct between the exhaust structure and the l isolation valves be insulated for high temperatures. ' The combination of these three things assures that the equivalent of a three-hour fire damper is provided. l (5) It is extremely important to be able to vent internal fires without intermption i from a fire damper that has closed due to the temperature of the exhaust I gases. The HVAC/ smoke control system for the Reactor Building secondary containment differs slightly from the other systems in the Reactor and Control Buildings, since because a common supply and exhaust system is used for all three divisional areas within l secondary containment. The systems for each division are branched from the common I system. A dual purpose isolation / fire damper valve is provided for each supply and exhaust branch. A two position motor-operated volume damper is also provided in each exhaust branch. Upon detection of a fire, a normally non operating exhaust fan is i started to increase the negative pressure of the exhaust system. The motormperated dampers in the exhaust ducts for the divisional HVAC/ fire areas without the fire reposition to their predetermined fire settings to maintain normal negative pressure in their zones. The pressure in the HVAC/ fire area experiencing the fire moves negative with the change in exhaust pressure. This establishes a pressure differential to the , adjacent fire areas to provide smoke control by the differential pressure across the fire l barriers surrounding the fire. See Subsection 9.4.4 for a description of the smoke control system for the Turbine Building. 9.5.1.1.7 Spurious Control Actions As stated above, the systems are separated by fire areas on a divisional basis. The multiplexing system is a dual channel system. Two simultaneous, identical digitized control signals are required at the de-multiplexer for control action to be taken at the field device. The probability of two spurious signals matching is essentially zero. 9.S.14 Other Auxiliary Systems - Amendment 34
23A6100 Rev. 5 ABWR standardsafetyAnalysis Report b . V Table of Contents (Continued) 9A.4.1.5.17 Deleted. . . . . . . . . . . . . . . . . . . . . . 9A.4-242 9A.4.1.5.18 Deleted ... .. .. . . . .. . . . . . . . . . . . 9A.4-242 9A.4.1.5.19 Electrical Penetration Room (Rm 543) . ... .. . .... . . . 9A.4-242 9A.4.1.5.20 FPC Valve Room (Rm No. 542). . . .. . . . . 9A.4-244 9A.4.1.5.21 FPC Pump Room (Rm No. 546)... . . . . . .. 9A.4-245 9A.4.1.5.22 Deleted. ... . . . .. . . . . . .. 9A.4-247 9A.4.1.5.23 Deleted. . . . .. . .. .. . . . . 9A.4-247 9A.4.1.5.24 Deleted. .. .. . . . . . . . . 9A.4-247 9A.4.1.5.25 Deleted. . . . . . . . . . .. . 9A.4-247 9A.4.1.5.26 FPC Heat Exchanger Room (Rm No. 544)... . .. . . . . . 9A.4-247 9A.4.1.5.27 Instrument Piping Penetration Room (Rm No. 511) .. . . 9A.4-249 9A.4.1.5.28 Clean Area Access Room (Rm No. 517) . . . . . . . 9A.4-250 9A.4.1.5.29 Division 1 Electrical Penetration Room (Rm No. 518). . 9A.4-252 9A.4.1.5.30 Service Corridor B (Rm No. 527). . .. .. .. . . . 9A.4-253 9A.4.1.5.31 Division 2 Electrical Penetration Room (Rm No. 528). . 9A.4-255 9A.4.1.5.32 Division 3 Electrical Penetration Room (Rm No. 532). .. 9A.4-257 9A.4.1.5.33 Pits and Pools (Rm No. 538,539).. . .. . . . . . . 9A.4-258 9A.4.1.5.34 RIP Transformer Room (Rm No. 541) . . .. .. .. ..9A.+258 9A.4.1.5.35 FPC Heat Exchanger Room (Rm No. 545).. .. .. . . 9A.4-260 9A.4.1.5.36 Corridor D (Rm No. 547) . . . .. . . . . 9A.4-262 9A.4.1.5.37 Upper Drywell (Rm No.591) .. . . . . .. . .. 9A.4-263 g ) 9A.4.1.6 Building-Reactor Bldg El 23500 mm and 27200 mm. ... 9A.4-264 L/ 9A.4.1.6.1 Cross Corridor A (Rm No. 614) . . ....... . . 9A.4-264 9A.4.1.6.2 D/G Fuel Day Tank A Room (Rm No. 610) . . . . . . . . . . . 9A.4-265 9A.4.1.6.3 AC Filter / Fan Area (Rm No. 615) . . . . . . . . . . . . .... 9A.4-267 9A.4.1.6.4 D/G (A) Equipment Room (Rm No. 613) . .... . . . 9A.4-269 9A.4.1.6.5 D/G (A)/Z HVAC Room (Rm No. 612).. . . . . . . . . . 9A.4-271 9A.4.1.6.6 SRV/MSIV Maintenance Room (Rm No. 616) . . . . 9A.4-273 9A.4.1.6.7 ISI Test Room (Rm No. 617). . . .... . . . . .. . . 9A.4-274 9A.4.1.6.8 D/G (C) Equipment Room (Rm 633) .. . . . . 9A.4-276 9A.4.1.6.9 D/G (C)/Z HVAC Room (Rm No. 632) . .. .. . . .. . 9A.4-278 9A.4.1.6.10 D/G Fuel Day Tank C Room (Rm No. 630) . . . .. . . 9A.4-279 9A.4.1.6.11 Hatch and Corridor B/C Room (Rm No. 634). . .. . .. 9A.4-281 9A.4.1.6.12 Corridor B SLC Area (Rm No. 622). . . . . . . . . . . .. 9A.4-283 9A.4.1.6.13 D/G Fuel Day Tank Room B (Rm No. 620) .. .. . . . 9A.4-285 9A.4.1.6.14 D/G (B) Equipment Room (Rm No. 625) . . . . . . 9A.4-287 9A.4.1.6.15 D/G (B)/Z HVAC Room (Rm No. 624).. . ... 9A.4-288 9A.4.1.6.16 ISl lnspection (Rm No. 639) .. ... . .. . . . . . 9A.4-290 9A.4.1.6.17 Deleted ... .. . .. . .. . . . . . .. . . . 9A.4-292 9A.4.1.6.18 Deleted. . . . . . . . . . .. . . . . . . . 9A.4-292 9A.4.1.6.19 Corridor D (Rm No. 643) . . . . .. . . . . . . ... ... 9A.4-292 9A.4.1.6.20 SGTS B Division 2 Room (Rm No. 641) . . . . . . 9A.4-293 9A.4.1.6.21 SGTS A Division 3 Room (Rm No. 642) . . . . 9A.4-295 9A.4.1.6.22 Deleted .. .. ... . . . .. 9A.4-297 9A.4.1.6.23 Deleted. . . . . . . . . . . . . . - . . . . . 9A.4-297
) 9A.4.1.6.24 Upper D/G A HVAC Room (Rm No. 653).. .. .. . 9A.4-297 9A.4.1.6.25 FMCRD A/C Panel Room (Rm No. 654) . . . . . . 9A.4-299 Table of Contents - Amendment 35 9A.0-v
1 23A6100 Rsv. 9 l ABWR Standard Safety Analysis Report i Table of Contents (Continued) 9A.4.1.6.26 Deleted. .. .. .. . . . . .. .. . . 9A.4-301 9A.4.1.6.27 Deleted. . . . . . . . . . . . . . . . . 9A.4-301 9A.4.1.6.28 Deleted . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . 9A.4-301 9A.4.1.6.29 Deleted . . . . . . . .. . .. . . . . . . . . . . . . . . . . . 9A.4-301 9A.4.1.6.30 Upper D/G C HVAC Room (Rm No. 673) .. . . . .. . . 9A.4-301 9A.4.1.6.31 Deleted. .. . . . . . . ... ... .. . . . .. 9A.4-304 9A.4.1.6.32 Upper D/G B HVAC Room (Rm No. 663). . .. . . . 9A.4-304 9A.4.1.6.33 Upper Corridor B (Rm No. 626). . . . . . . . . . 9A.4-305 9A.4.1.6.34 Deleted ... . . .. .. . . . . . . . .. . . . . 9A.4-306 9A.4.1.6.35 FMCRD D/B Panel Room (Rm No. 681) . .. . . . . . . . 9A.4-306 9A.4.1.6.36 Deleted.. . . ... . . . . . ... . . . . 9A.4-308 9A.4.1.6.37 Deleted. . . . . . . .. . . . . . . . . 9A.4-308 9A.4.1.6.38 MS Tunnel HVH Room (Rm No. 685) . . . . 9A.4-308 9A.4.1.6.39 Pits and Pools.... . . . . . . . .. . .9A.+310 9A.4.1.6.40 PVC Purge Exhaust Fan (Rm No. 623) . . . . . . 9A.4-310 9A.4.1.6.41 D/G C Corridor Room (Rm No. 635) . .. . . .. . . 9A.4-312 9A.4.1.6.42 RIP Power Supply Room (Rm No. 638). . . . . . . . . . . . 9A.4-313 l 9A.4.1.6.43 Electrical Equipment Room (Rm No. 640) . ........ . . . . . 9A.4-315 9A.4.1.6.44 Leak Detection Dust Radiation Monitoring Room (Rm No. 657) .. . .. . . . . . . . . . . ... . .. . 9A.4-316 9A.4.1.6.45 Room No. 658. .. . . . . . . . . . . . 9A.4-318 9A.4.1.6.46 Containment Atmosphere Monitor System CAMS A Rack , Room (Rm No. 659). . . . . . . . . . . . . . . . . . . . . 9A.4-319 l 9A.4.1.6.47 Electrical Room (Rm No. 680) . . . . .. . . . . . . .. . 9A.4-321 9A.4.1.6.48 Deleted. . . . . . . .. ... . . . . . . . . . . . ... 9A.4-323 l 9A.4.1.6.49 Containment Atmosphere (CAM) Monitor System Rack B Room (Rm 621). . . . . .. . . . .. . 9A.4-323 9A.4.1.6.50 Deleted. . . . . . . . . . . . . . . . . . . . 9A.4-325 i 1 9A.4.1.7 Building-Reactor Building El 31700 mm .. . . . . . . . 9A.4-325 9A.4.1.7.1 Reactor Building Operating Deck (Rm No. 716) . . . ... . . 9A.4-325 9A.4.1.7.2 RIP (A) Supply Fan and RCW (C) Surge Tank (Rm No. 715) 9A.4-327 9A.4.1.7.3 Deleted. . .. . . . . . . . . .. . . . . . . . . . .. . 9A.4-329 9A.4.1.7.4 DG (C) Exhaust Fan Room (Rm No. 730) . .. . . . . . . 9A.4-329 9A.4.1.7.5 Dele ted .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9A.4-330 9A.4.1.7.6 RIP (B) Supply Fan and RCW (B) Surge Tank (Room No. 740). . . . . . . . . . . . . . . . . . . . . ... 9A.4-330 9A.4.1.7.7 Access Service Area (Rm No. 764) ... . . . . . . . . 9A.4-332 9A.4.1.7.8 Refueling Machine Control Room (Pan No. 760). . . . .. .. 9A.+334 9A.4.1.7.9 Gallery (Rm No. 762) .. . .. . ... .. .. . . 9A.4-335 9A.4.1.7.10 Mezzanine Corridor (Rm No. 761) .. ... . . . .. . 9A.4-337 9A.4.1.7.11 Roof A/C Area (Rm No. 810 and 830) . .. . .. . 9A.4-338 9A.4.1.7.12 Roof B/D Area (Rm No. 820 and 840) . . . . . . . . . . . . 9A 4-340 9A.4.1.7.13 RCW (A) Surge Tank (Rm No. 710) . . .... . . . . 9A.4-341 9A.4.1.7.14 Periodic Inspection Room (Rm No. 720) . . . . . . . . . 9A.4-343 9A.4.1.7.15 RIP Repair Room (Rm No. 723) . . . . . 9A.4-344 9A.4.1.7.16 Refuel Machine Control Room HVH (Rm No. 722) . .. . . 9A.4-346 9A.0-vi Table of Contents . Amendment 37
l 23A6100 Rev. 9 ABWR standardsafety Analysis Report ( l Table of Contents (Continued) 9A.4.1.7.17 Standby Gas Treatment System Pipe Space Room (Rm No. 741 ) .. .......... .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... 9A.+347 9A 4.1.7.18 HVAC Supply Duct Room (Rm No. 711) . ... ...... . . . . . . . 9A.+349 l 9A.4.1.7.19 Elevator Equipment Room (Rm No. 811) .... . . . . ......... .. 9A.4-351 l 9A.4.1.7.20 Elevator Equipment Room (Rm No. 821) ....... ..... .. ... . .. 9A.+352 9A.4.2 Con trol Building...... ..... .. . . .. ... . . . . . . . . . .. . ... . .. .. 9A.+354 9A.4.2.1 Floor One El -8200 mm and -2150 mm . . . .... ... . .. .. ... 9A.4-354 9A.4.2.1.1 RCW "A" (Rm No. I11).. .... .. . . . . . . . . . . . . . . . . . . . . . . . . ... . 9A.4-354 9A.4.2.1.2 Passageway (Rm No.112) . . .. . . . . . . . . . . . . . . . . ... . .. . . 9A.4-356 9A.4.2.1.3 RCU "B" (Rm No.121) .. ... . . . .. . . . . . .. . . . . ... . . . 9A. 4-3 58 9A.4.2.1.4 Passageway (Rm No.122) ... ........ ... .... . . . . . . . . . . . . . . . 9A.4-360 9A.4.2.1.5 RCW "C," (Rm No.131).. . . . . . . . . . . . . . . ... . ..... .. . 9A.4-362 9A.4.2.1.6 Passageway (Rm No.132) . ... . . . . . . . . . . . . . .. . .. . . .. 9A.4-365 9A.4.2.2 Floor Two El -2150 mm.... ..... ... . . .... ... .. ... . . . . . .. 9A.4-367 9A.4.2.2.1 Passageway (Rm No. 211) . . ........ ... . . . . . . . . . . . . . . . . ..... 9A.4-367 9A.4.2.2.2 Passageway (Rm No. 221) .... . . . . . . . . . . . . .. ... . . . 9A.+368 9A.4.2.2.3 Passageway (Rm No.231) . .. .. . .. . . . . . . . ... . . . . . . 9A.4-370 9A.4.2.3 Floor Three El 3500 mm.. .. . . . . . . . . . . . . . . . . . . ... .... . . . 9A.4-372 9A.4.2.3.1 250 VDC Batteg Room (Rm No. 313) . ........ ... .. .... .. . 9A.4-372 9A.4.2.3.2 Passageway (Rm No. 312) .... . .... ... ..... ... .. ... ... .. ... . . . . 9A.4-374 s 9A.4.2.3.3 Non-Divisional Electrical Equipment Room (Rm No. 311) ... 9A.4-376 9A.4.2.3.4 Passageway (Rm No. 314) ..... ..... . . . . . . . . . . . . . . . . . .. . . . . . . 9A.4-378 9A.4.2.3.5 Battey Room Div 2 (Rm No. 322) .. ...... . .. . . . .. . .. .. .. . . . . . 9A.4-380 9A.4.2.3.6 Division 2 Electrical Equipment Room (Rm No. 323). .. ... .. 9A.4-381 9A.4.2.3.7 Division 4 Electrical Equipment Room (Rm No. 342).... ....... 9A.4-383 9A.4.2.3.8 Battery Room Division 4 (Rm No. 341) ... ..... .. ........ . ....... . 9A.4-385 9A.4.2.3.9 Passageway (Rm No. 343) ... . . . . . . . . . . . . . . . . .. . .. .. . . . . .... 9A.4-38 7 9A.4.2.3.10 Passageway (Rm No. 321) .... . . . . . . . . . . . . . . .. . . . .. . . . . . ... . . . . 9A.4-389 9A.4.2.3.11 Division 2 HVAC Chase (Rm No. 324) ..... . ... . .. ... . 9A.4-391 9A.4.2.3.12 Battey Room Division 1 (Rm No. 316) . .. .... ... .. . . .... . 9A.4-393 9A.4.2.3.13 Division 1 Electrical Equipment Room (Rm No. 317).. . . . 9A.4-395 9A.4.2.3.14 Division 1 HVAC Chase (Rm No. 319) .... ...... .... .. ............ . 9A.4-397 9A.4.2.3.15 Batteg Room Division 3 (Rm No. 332) ....... . ... . . . . ... . . . .... 9A.4-398 9A.4.2.3.16 Division 3 Elect. Equipment Room (Rm No. 331) ... .. .. ........ 9A.4-400 9A.4.2.3.17 Division 3 HVAC Chase (Rm No. 335) . . . .. ... ... . ... . ... . 9A.4-402 9A.4.2.3.18 Passageway (Rm No. 333) ... ..... .. .. . . . . . . . . . . . . . . . .. 9A.4-404 9A.4.2.3.19 Passageway (Rm No. 318) .. ...... . . . . . . . . . . . . . . . . . . . . .. . 9A.4-406 9A.4.2.3.20 Passageway (Rm No. 315) ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9A.4-407 ) 9A.4.2.3.21 Stairwell (Rm No. 325) .. . .... .. .. ...... . . ...... ..... ........... .. . .. 9A.4-409 9A.4.2.3.22 Stairwell (Rm No. 336).. . . .. . . . . . . . . . . . .. . . . . . . . . 9A.4-41 1 9A.4.2.3.23 Elevator (Rm No. S37) . ... .. ... .. ... ..... .. . . . . . .. 9A.4-413 9A.4.2.4 Floor Four El 7900 mm . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... 9A.4-414 9A.4.2.4.1 Control Room Complex..... .. . . . . . . . . . . . . . . . . . . . . . . .. 9A.4-414 w 9A.4.2.5 Floor Five El 12300 mm .... .. ... . . . . . . . . . . . . . . . . . . . . . . . . . . 9A.4 418 l 9A.4.2.5.1 Control Room HVAC "B" Exhaust Duct Chase (Rm No. 522) 9A.4-418 9A.4.2.5.2 HVAC "A" Supply (Rm Nos. 511,512 and 513) . . . .. . ...... 9A.4-420 9A.4.2.5.3 HVAC "C" Supply (Rm Nos. 531,532 and 533) . . .. . 9A.4-422 ] l l Table of Contents - Amendment 37 9A.0-vii l l
23A6100 R1v. 9 ABWR StandardSafety Analysis Rep nt Table of Contents (Continued) l 9A.4.2.5.4 Stairwell Landing (Rm No. 505) . . . . . .. . . .. 9A.4-424 l 9A.4.2.5.5 Chiller Unit "C," (Rm No. 534) . . .. .. . . .. . 9A.4-426 9A.4.2.5.6 Recirc Internal Pump MG Sets and Control Panels (Rm Nos. 501,502,503 and 504) . . . .. . . . . . . . . . . .. 9A.4-428 9A.4.2.5.7 Computer Room (Rm No. 591) . . . .. . 9A.4430 9A.4.2.5.8 Passageway (Rm No. 521) . . . . . . . .. . . 9A.4-432 9A.4.2.5.9 Deleted. . .. . . . . . . . . . . . . . 9A.4434 l 9A.4.2.5.10 Passageway (Rm No. 592) . . ... .. . .. . . 9A.4-434 9A.4.2.5.11 Passageways (Rm No. 593) . . . . .. . . . . . 9A.4-436 l 9A.4.2.5.12 Control Room HVAC "C," Exhaust Duct Chase (Rm No. 595) . . . ... .. . .. .. . 9A.4-438 9A.4.2.5.13 Passageway (Rm No. 506) . .. . 9A.4-439 9A.4.2.6 Floor Six El 17150 mm.. .. .. . . 9A.4-441 9A.4.2.6.1 Control Room HVAC Supply "B" (Rm No. 621) . . . . 9A.4-441 9A.4.2.6.2 Passageway and Room (Rm No. 622 and 662) . . 9A.4443 i 9A.4.2.6.3 Chiller Unit "B" (Rm No.623) . . . . .. .. . .. . . . . . .. 9A.4-445 9A.4.2.6.4 HVAC "B" Supply and Exhaust (Rm Nos. 624,625,627, 661 and 664) . . . . ... .. . .. ... . 9A.4447 9A.4.2.6.5 HVAC "A" Intake Duct and Exhaust (Rm Nos. 613,617, 618 and 619) . . . . ... . . . . . . 9A.4-449 l 9A.4.2.6.6 Control Room HVAC Exhaust "B" (Rm No. 626) . .. . . 9A.4-451 ! 9A.4.2.6.7 Chiller Unit "A" (Rm No. 612) . . . . . . .9A.4453 l 9A.4.2.6.8 Control Room HVAC Supply "C" (Rm No. 615) .. . . 9A.4-455 l 9A.4.2.6.9 Passageway and Room (Rm Nos. 611 and 652) . .. . .. 9A.4-457 l l 9A.4.2.6.10 Control Room HVAC Exhaust "C" (Rm Nos. 614) . .... ... . 9A.4-459 9A.4.2.6. l l HVAC "C" Intake Duct and Exhaust (Rm Nos. 631,632, 633,634,651, and 653) . . . .. . ... . 9A.4-461 9A.4.3 Turbine Building.. . . . . . . . . . . . . . . . . . . .. . 9A.4-463 1 l 9A.4.3.1 Floor One El 5.3 m. . . . . . . . . . . . . . . . . . . .. . 9A.4-463 l 9A.4.3.1.1 Floor One (Except Fire Areas FT1501-FT1503) . . . . . 9A.4-463 9A.4.3.1.2 Air Compressors and Dryer Area (Rm No. I11) . .. . .. 9A.4-466 9A.4.3.1.3 Stair Tower # 1 (Rm No. I14)... . . . .. . . . .. . 9A.4-468 9A.4.3.1.4 Stair Tower # 2 (Rm No.122)...... . . .. . . . . . . . 9A.4-469 9A.4.3.2 Floor Two El 12.3m ..... . . . . . . . . . . . . . .. .. . . 9A.4-470 9A.4.3.2.1 Floor Two (Except Fire Areas FT-1501, FT2500-FT2505). . 9A.4-470 9A.4.3.2.2 Switchgear "A" Area (Rm No. 210), and Chillers Area (Rm No. 248) . . . . . . . . . . . . . . . . . . . . . . . . . . 9A.4-473 9A.4.3.2.3 Lube Oil Conditioning Area (Rm No. 230) .. . . 9A.4-475 9A.4.3.2.4 Stair Tower # 5 (Rm No. 236).. . . . . . . . . . . . 9A.4-477 9A.4.3.2.5 Stair Tower # 3 (Rm No. 212).. .. .. .. . . . ... .. . 9A.4-478 9A.4.3.2.6 House Boiler Area (Rm No. 247) .. . .. . . 9A.4-480 9A.4.3.2.7 Stair Tower # 4 (Rm No. 249).. . . . 9A.4-482 9A.4.3.2.8 Steam Tunnel Area (Rm No. 219) .. . . 9A.4-484 9A.4.3.3 Floor Three El 20.3m . . 9A.4-485 0 9A.0-viii Table of Contents . Amendment 37 l
23A6100 Rev. 4 ABWR stendedsaferyAassysisseret "O (a) The function is located in a separate fire-resistive enclosure. (b) Fire detection and suppression capability is provided and accessible. (c) Fire stops are provided for cable tray and piping penetrations through rated fire barriers. (9) Consequences of Fire-The postulated fire assumes the loss of the function. Access to the corridor is from one end only. The corridor does not provide access to any area containing equipment required for save shutdown of the plant. Smoke from a fire wil! *.e removed by the normal HVAC System operating in its smoke removal mode. (10) Consequences of Fire Suppression-Suppression extinguishes the fire. Refer to Section 3.4, WaterLevel(Flood) Design, for the drain system. (11) Design Criteria Used for Protection Against Inadvertent Operation, Careless l Operation or Rupture of the Suppression System:
/N i
/ ) (a) Location of the manual suppression system in the corridor, external to ' the rooms containing safety-related equipment (b) Provision of raised supports for the equipment (c) Refer to Section 3.4, WaterLevel(Rood) Design, for the drain system. l (d) ANSI B31.1 standpipe (rupture unlikely) (12) Fire Containment or Inhibiting Methods Employed: (a) The functions are located in a separate fire-resistive enclosure. (b) The means of fire detection, suppression and alarming are provided and accessible. (13) Remarks-None. 9A.4.1.5.37 Upper Drywell (Rm No.591) (1) Fire Area-F4901 (2) Equipment: See Table 9A.6-2 for this elevation. Devices within the upper drywell are also listed at floor elevation 12300 mm. O ( Note: Section 9A.4.1.4.1 applies for the remainder of the information for the upper drywell. See that section for additional information. Analysis -4nendment 34 9A.4-263
23A6100 Rev. 9 ABWR StandardSafety Analysis Report O 9A.4.1.6 Building-Reactor Bldg El 23500 mm and 27200 mm 9A.4.1.6.1 Cross Corridor A (Rm No. 614) (1) Fire Area-F4100 (2) Equipment: See Table 9A.fr2 Safety-Related Provides Core Cooling Yes, D1 Yes, D1 (3) Radioactive hfaterial Present-None that can be released as a result of fire. (4) Qualifications of Fire Barriers-The exterior wall, inside wall, ceiling and floor of this corridor are of 3 h fire-resistive construction. This corridor extends across the reactor building. At the south end of the corridor, a 3 h fire-l resistive door opens to the electrical equipment room (Rm 640). At the other end of the corridor, a nonrated door opens into D/G (A) exhaust fan area (Rm 613). (5) Combustibles Present: Fire Loading Total Heat of Combustion (hy) None 727 hy/m2 NCLL (727 hy/m 2 maximum average) applies. (6) Detection Provided-Class A supervised POC in the room and manual alarm pull station at Col.1.0-B.2 and 6.2-B.0. (7) Suppression Available: Type Location / Actuation Standpipe and hose reel Col.1.0-B.2 & 6.2-B.0/hlanual ABC hand extinguishers Col.1.0-B.2 & 6.2-B.0/hfanual (8) Fire Protection Design Criteria Employed: (a) The function is located in a separate fire-resistive enclosure. (b) Fire detection and suppression capability is provided and accessible. 9A4-264 Analysis Amendment 37
l l 23A6100RW 9 ABWR standardsafery Analysis neport i p (12) Fire Containment or Inhibiting Methods Employed: 1 (a) The functions are located in a separate fire-resistive enclosure. I l (b) The means offire detection, suppression and alarming are provided and l accessible. l l l (13) Remarks-The sunken volume of the room is adequate to hold the entire l contents of the day tank if an uncontrolled leak should occur. 9A.4.1.6.14 D/G (B) Equipment Room (Rm No. 625) l (1) Fire Area-F4202 (2) Equipment: See Table 9A.f> 2 l Safety-Related Provides Core Cooling Yes, D1 Yes, D1 (3) Radioactive Material Present-None that can be released as a result of fire. l (4) Qualifications of Fire Barriers-All walls are of 3 h fire-resistive construcdon. l A section of the ceiling below the FMCRD panel room (Rm 681, fire, F7200) is of 3 h fire-resisdve concrete construction. Sections of the floor above the D/G Fan and HVAC room B (Rm 524), and the senice corridor B (Rm 527) are of 3 h fire-resisdve concrete construction. Access to the area is prosided from the stairs and elevator (areas 329 and 328 respecdvely), from corridor Rm 635 (via corridor Rm 626) and from the~ electrical equipment room l l (Rm 640). Each access route is through a 3 h fire-resistive door. (5) Combustibles Present: Fire Loading Total Heat of Combustion (MJ) l 2 2 Cable Tray 727 MJ/m NCU (727 MJ/m l l 1 maximum average) applies (6) Detection Provided-Class A supervised POC in the room and manual alarm pull stations at 1.4 D.7,1.0-B.2. c [t
\
Analysis . Amendment 37 ' 9A.4-287
I l l l 23A6100 Rtv. 4 ABWR standardsafety Analysis neport l (7) Suppression Available: l l Type Location / Actuation Standpipe and hose reel Col.1.4-D.7, and 1.0-B.2/ Manual ABC hand extinguishers Col.1.4-D.7,and 1.0-B.2/ Manual (8) Fire Protection Design Criteria Employed: (a) The functicn is located in a separate fire-resistive enclosure. (b) Fire detection and suppression capability is provided and accessible. (c) Fire stops are provided for cable tray and piping penetrations through rated fire barriers. (9) Consequences of Fire-The postulated fire assumes the loss of the function. Diesel generators A and C would not be affected. Smoke from a fire will be removed by the EHVAC (B) system operating in its smoke removal mode. (10) Consequences of Fire Suppression-Suppression extinguishes the fire. Refer to Section 3.4, WaterLevel(Rood) Design, for the drain system. (11) Design Criteria Used for Protection Against Inadvertent Operation, Careless Operation or Rupture of the Suppression System: (a) Provision of raised supports for the equipment (b) Refer to Section 3.4, Water Level (Rood) Design, for the drain system. l (c) ANSI B31.1 standpipe (nipture unlikely) (12) Fire Containment or Inhibiting Methods Employed: (a) The functions are located in a separate fire-resistive enclosure. (b) The means of fire detection, suppression and alarming are provided and accessible. (13) Remarks-None. 9A.4.1.6.15 D/G (B)/Z HVAC Room (Rm No. 624) (1) Fire Area-F4200 t 9A.4-288 Analysis - Amendment 34
l 23A6100 Rsv. 4 ABWR standantsareryAaaiysisneport (4) Qualification of Fire Barriers-All four walls, the floor and ceiling are internal to fire area F4301 and therefore are not fire rated. (5) Combustibles Present-No significant amount of exposed combustibles. 727 2 MJ/m NCLL (727 MJ/m2 maximum average) applies. (6) Detection Provided-Class A supenised POC detection system in the room and alarm pull station at 5.2-D.8 and 5.2-B.6. l (7) Suppression Available: Type Location / Actuation l Standpipe and hose reel 5.2-D.8 & 5.2-B.6/ Manual l ABC hand extinguishers 5.2-D.8 & 5.2-B.6/ Manual (8) Fire Protection Design Criteria Employed: (a) Fire detection and suppression capability is provided and accessible. ( (b) Fire stops are prosided for cable tray and piping penetration through rated fire barriers. (9) Consequences of Fire-The postulated fire assumes the loss of the function. Smoke from a fire will be removed by the normal HVAC System operating in its smoke removal mode. (10) Consequences of Fire Suppression-Suppression extinguishes the fire. Refer to Section 3.4, WaterLevel(Flood) Design, for the drain system. (11) Design Criteria Used for Protection Against Inadvertent Operation, Careless Operation or Rupture of the Suppression System: (a) Location of the manual hose suppression system external to the room (b) Provision of raised supports for the equipment (c) Refer to Section 3.4, WaterLevel(Flood) Design, for the drain system. l (d) ANSI B31.1 standpipe (rupture unlikely) (12) Fire Containment or Inhibiting Methods Employed: 'O Q (a) The functions are located in a separate fire-resistive enclosure. Analysis - Amendment 34 9A.4 291
23A6100 Rxv. 9 l l ABWR Standard Safety Analysis Report O (b) The means of fire detection, suppression and alarming are provided and accessible. (13) Remarks-None. l 9A.4.1.6.17 Deleted I ! 9A.4.1.6.18 Deleted 9A.4.1.6.19 Corridor D (Rm No. 643) (1) Fire Area-F4201 (2) Equipment-See Table 9A.6-2 Safety-Related Provides Core Cooling No No (3) Radioactive Material Present-None. l (4) Qualificadons of Fire Barriers-The walls common with the electrical equipment room (Rm 640), the SGTS A filter train room (Rm 642), corridor room (Rm 614), the floor above the steam tunnel and the ceiling serve as fire j l barriers between adjacent fire areas and are of S h fire-resistive concrete construction. A S h fire rated door provides access from the AC filter / fan area ! (Rm 615). Room 643 connects directlyinto room 622. (5) Combusdbles Present-l Fire Loading Total Heat of Combustion (MJ) 1 2 2 Cable Tray 727 MJ/m NCLL (727 MJ/m maximum average) applies (6) Detection Provided-Class A supervised POC in the room and manual alarm pull stadons at 2.7-C.0 and 2.8-F.1. (7) Suppression Available: Type Location / Actuation Standpipe and hose reel Col. 2.7CO,& 2.8-F.1/ Manual ABC hand extinguishers Col. 2.7CO,& 2.8-F.1/ Manual 9A.4-292 Analysis - Amendment 37 l 1
l l 23A6100 R1v. 4 ABWR standardsareryAurysisneport r 1 (8) Fire Protection Design Criteria Employed: 1 (a) The function is located in a separate fire-resistive enclosure.
)
1 (b) Fire detection and suppression capability is provided and accessible. l (c) Fire stops are provided for cable tray and piping penetrations through i rated fire barriers. i l (9) Consequences of Fire-The postulated fire assumes the loss of the function. l Loss of the SGTS by an exposure fire is acceptable. Smoke from a fire will be removed by the normal HVAC System operating in l its smoke removal mode. (10) Consequences of Fire Suppression-Suppression extinguishes the fire. Refer to Section 3.4, WaterLevel(&d) Design, for the drain system. (11) Design Criteria Used for Protection Against Inadvertent Operation, Careless l Operation or Rupture of the Suppression System: 1 C (a) Provision of raised supports for the equipment l k' (b) Refer to Section 3.4, WaterLevel(&d) Design, for the drain system. l (c) ANSI B31.1 standpipe (rupture unlikely) (12) Fire Containment or Inhibiting Methods Employed: (a) The functions are located in a separate fire-resistive enclosure. ; (b) The means of fire detection, suppression and alarming are provided and accessible. 1 (13) Remarks-None. 9A.4.1.6.20 SGTS B Division 2 Room (Rm No. 641) (1) Fire Area-F4201 (2) Equipment: See Table 9A.6-2 Safety-Related Provides Core Cooling Yes, D2 No i t (~ Analysis - Amendment 34 9A.4 293
23A6100 Rrv. 9 ABWR standardsafety Analysis Report O (3) Radioactive Material Present-Filters within their housing may become contaminated with use. Releases up the stack could occur as a result of fire. However, the system is capable of being isolated in case of any fire, and burn itself out by cutting the oxygen to the fire. (4) Qualifications of Fire Barriers-The walls common with the electrical equipment room (Rm 640), the SGTS A division 3 room (Rm 642), the ceiling, and a section of the floor common to fire area F3400 (Rm 543) below serve as fire barriers between adjacent fire areas and are of 3 h fire-resistive concrete construction. The remainder of the floor (not common to F3400), the wall common with SLC Area and corridor B room 622 are not rated as they are internal to fire area F4201. A non-fire rated door provides access from corridor D (Rm 643). l (5) Combustibles Present: I I 1 Fire Loading Total Heat of Combustion (MJ) ; 1 2 2 Cable Tray ~27 MJ/m NCLL (727 MJ/m ) I maximum average) applies (6) Detection Provided-Class A supervised POC in the room and manual alarm pull station at Col. 2.7-C.0 and 2.8-F.1. (7) Suppression Available: Type Location / Actuation Standpipe and hose reel Col. 2.7-C.0 & 2.8-F.1/ Manual ABC hand extinguishers Col. 2.7-C.0 & 2.8-F.2/ Manual (8) Fire Protection Design Criteria Employed: (a) The function is located in a separate fire-resistive enclosure. (b) Fire detection and suppression capability is provided and accessible. (c) Fire stops are provided for cable tray and piping penetrations through rated fire barriers. (9) Consequences of Fire-The postulated fire assumes loss of function. The complete loss of the SGTS B as a consequence of a single fire is acceptable. Functional backup is provided by SGTS A (Div. III). 9A.4 294 Analysis - Amendment 37
1 23A6100 Rsv. 9 O StandardSafety Analysis Report
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(4) Qualifications of Fire Barriers-All walls, the ceiling, and the floor are of 3 h ' fire-resistive concrete constniction. Access to room 681 is from stair well l (Rm 329) and elevator (Rm 328) via 3 h rated fire-resistive doors. The room l l provides access to D/G B upper fan room (Rm 663) and to the electrical room (Rm 680) through 3 h rated fire-resistive doors. (5) Combustibles Present: Fire L~*ag Total Heat of Combustion (MJ) Cable Tray 2 727 MJ/m2 NCLL (727 MJ/m maximum average) applies j , (6) Detection Provided-Class A supenised POC in the room and manual alarm pull stations at 1.4-E.0,1.7-C.0. (7) Suppression Available: l 1 Type Location / Actuation l Standpipe and hose reel Col.1.4-E.0, and 1.7-C.0/ Manual 1 ABC hand extinguishers Col.1.4-E.0, and 1.7-C.0/ Manual (8) Fire Protection Design Criteria Employed: (a) The function is located in a separate fire-resistive enclosure. (b) Fire detection and suppression capability is provided and accessible. (c) Fire stops are provided for cable tray and piping penetrations through l rated fire barriers. (9) Consequences of Fire-The postulated fire assumes the loss of the function. The effects of fire on the FMCRD system are discussed in Section 9A.5. Smoke from a fire will be removed by the EHVAC(B) system operating in its smoke removal mode. (10) Consequences of Fire Suppression-Suppression extinguishes the fire. Refer to Section 3.4, Water Level (Flood) Design, for the drain system. l (11) Design Criteria Used for Protection Against Inadvertent Operation, Careless I ' Operation or Rupture of the Suppression System: (a) Provision of raised supports for the equipment Analysis Amendment 37 9A.4-307 {
23A6100 Rtv. 4 ABWR Standard SafetyAnalysis Report O (b) Refer to Section 3.4, WaterLevel(Flood) Design, for the drain system. l (c) ANSI B31.1 standpipe (rupture unlikely) (12) Fire Containment or Inhibiting Methods Employed: (a) The functions are located in a separate fire-resistive enclosure. (b) The means of fire detection, suppression and alarming are provided and accessible. j (13) Remarks-None. 9A.4.1.6.36 Deleted 9A.4.1.6.37 Deleted 9A.4.1.6.38 MS Tunnel HVH Room (Rm No. 685) (1) Fire Area-F4201 l
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(2) Equipment: See Table 9A.6 2 l Safety-Related Provides Core Cooling i No No j l (3) Radioactive Material Present-None that can be released as a result of fire. l l (4) Qualifications of Fire Barriers-All walls and the ceiling serve as fire barriers l and are of Shr fire-resistive concrete. The floor is concrete but is non rated as it is internal to fire area F4201. Access to this room is a stairway leading form l a lower ares, internal to fire area F4201. l (5) Combustibles Present: 1 l Fue Leading Total Heat of Combustion (MJ) l 2 2 Cable Tray 727 MJ/m NCLL (727 MJ/m ; maximum average) applies i (6) Detection Provided-Class A supenised POC detection system in the room and manual alarm pull stations at Col. 2.7-C.0. (El. 23500 mm). O 9A.4-308 Analysis - Amendment 34
r 1 23A6100 Rev. 9 i ABWR standard safetyAnalysis Report h
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(11) Design Criteria Used for Protection Against Inadvertent Operation, Careless l Operadon or Rupture of the Suppression System: (a) Location of the manual hose suppression system external to the room i (b) Provision of raised supports for the equipment (c) Refer to Section 3.4, WaterLevel(Flood) Design, for the drain system. j (d) ANSI B31.1 standpipe (rupture unlikely) l . (12) Fire Containment or Inhibiting hiethods Employed: l l I (a) The functions are located in a separate fire-resistive enclosure. (b) The means of fire detection, suppression and alarming are provided and accessible. l (13) Remarks-None. l 9A.4.1.6.43 Electrical Equipment Room (Rm No. 640) t (1) Fire Area-F6200
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(2) Equipment: See Table 9A.6-2 Safety-Related Provides Core Cooling No No (3) Radioactive Material Present-None that can be released as a result of fire. (4) Qualifications of Fire Barriers-All walls and the floor are of 3 h fire-resistive concrete construction. A section of the ceiling is common to the FMCRD room (Rm 681) above and is of 3 h fire-resistive concrete construction. The remainder of the ceiling is internal to fire area F6200 and is not fire rated. Access is provided from rooms 625 and 614 through 3 h fire-resistive doors. (5) Combustibles Present: Fire Loading Total Heat of Combustion (MJ) 2 2 Cable Tray 727 MJ/m NCLL (727 MJ/m , p maximum average) applies (6) Detection Provided-Class A supenised POC in the room and manual alarm pull stations at 1.0-B.2 and 1.4-D.7. Analysis Amendment 37 9A.4-315
23A6100 Rsv. 4 ABWR standardsateryAnalysis aeport O (7) Suppression Available: Type Location / Actuation Standpipe and hose reel Col.1.0-B.2 & l.4-D.7/ Manual ABC hand extinguishers Col.1.0-B.2 & l.4-D.7/ Manual (8) Fire Protection Design Criteria Employed: (a) The function is located in a room separate from the rooms which contain safety-related equipment. (b) Fire detection and suppression capability is provided and accessible. (c) Fire stops are provided for cable tray and piping penetrations through rated fire barriers. (9) Consequences of Fire-The postulated fire assumes the loss of the function. Smoke from a fire will be removed by the EHVAC(B) system operating in its smoke removal mode. (10) Consequences of Fire Suppression-Suppression extinguishes the fire. Refer to Section 3.4, WaterLevel(Flood) Design, for the drain system. (11) Design Criteria Used for Protection AgainstInadvertent Operation, Careless Operation or Rupture of the Suppression System: (a) Provision of raised supports for the equipment (b) Refer to Section 3.4, WaterLevel(Flood) Design, for the drain system. l (c) ANSI B31.1 standpipe (rupture unlikely) (12) Fire Containment or Inhibiting Methods Employed: (a) The functions are located in a separate fire-resistive enclosure. (b) The means of fire detection, suppression and alarming are provided and accessible. 1 (13) Remarks-None. 9A.4.1.6.44 Fission Product Monitoring (Rm No. 657) i (1) Fire Area-F4301 9A4-316 Analysis - Amendment 34
4 l 23A6100 Rw. 9 ABWR standedsafetyAnotysis neper 'b V (10) Consequences of Fire Suppression-Suppression extinguishes the fire. Refer to Section 3.4, WaterLevel(&d) Design, for the drain system. (11) Design Criteria Used for Protection Against Inadvertent Operation, Careless Operation or Rupture of die Suppression System: (a) Location of the manual hose suppression system external to die room (b) Provision of raised supports for the equipment . (c) Refer to Section 3.4, WaterLevel(&d) Design, for the drain system. (d) ANSI B31.1 standpipe (rupture unlikely) (12) Fire Containment or Inhibiting Methods Employed: (a) The functions are located in a separate fire-resistive enclosure. (b) The means of fire detection, suppression and alarming are prosided and 4 accessible. (13) Remarks-None.
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^ l 9A.4.1.6.47 Electrical Room (Rm No. 680) (1) Fire Area-F6400 (2) Equipment: See Table 9A.6-2 i Safety-Related Provides Core Cooling No No (3) Radioactive Material Present-None. (4) Qualifications of Fire Barriers-The walls in common with the FMCRD room l (Rm 681), the SGTS A filter train room (Rm 642), corridor B room (Rm 643), both exterior walls and the ceiling are of 3 h fire-resistive concrete construction. The floor is common to room 640 below and is not fire rated. Access to room 680 is provided from the FMCRD room via a 3 h fire-resistive door and directly from room 640 below via a stainvell. O
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Analysis - Amendment 37 9A.4-321
l 23A6100 R5v. 4 ABWR standard Safety AnalysisReport l O (5) Combustibles Present: l Pse Loading Total Heat of Combustion (MJ) Cable Tray 727 hij/m 2NCLL (727 hfJ/m2 maximum average) applies. (6)' Detection Prosided-Class A supenised POC in the room and manual alarm pull stations at 1.7-C.0 and 1.4-E.0. (7) Suppression Available: Type Location / Actuation Standpipe and hose reel Col.1.7-C.0 & 1.4-E.0/hfanual ABC hand extinguishers Col.1.7-C.0 & 1.4-E.0/h1anual (8) Fire Protection Design Criteria Employed: (a) The function is located in a room separate from the rooms which contain safety-related equipment. (b) Fire detection and suppression capability is provided and accessible. l (c) Fire stops are provided for cable tray and piping penetrations through j rated fire barriers. l (9) Consequences of Fire-The postulated fire assumes the loss of the function. Smoke from a fire will be removed by the EHVAC(B) system operating in its smoke removal mode. (10) Consequences of Fire Suppression-Suppression extinguishes the fire. Refer to Section 3.4, WaterLevel(Flood) Design, for the drain system. (11) Design Criteria Used for Protection Against Inadvertent Operation, Careless l Operation or Rupture of the Suppression System: (a) Location of the manual suppression system external to this non safety-related room (b) Refer to Section 3.4, WaterLevel(Flood) Design, for the drain system. l (c) ANSI B31.1 standpipe (rupture unlikely) (12) Fire Containment or Inhibiting hiethods Employed: 9A.4 322 Analysis - Amendment 34
23A6100 Rsv. 4 ABWR SinaderdSonsty Analysis Report b i A. (b) The means of fire detection, suppression and alarming are provided and accessible. (c) Fire stops are provided for cable tray and piping penetrations through rated fire barriers. (13) Remarks-The Main Control Room includes a raised floor which is considered part of the room. The raised floor area will be used to route cable to and from the Safety System and Logic Control (SSLC) cabinets, the operator bench boards and displays, and the divisional electrical equipment rooms. The control room area are raised floor are considered to be non-hazard areas per IEEE 384. Section 8.3.3.6.2.2.3 discusses at length the separation criteria applies to divisional electrical cabling in the control room. It was determined that fire suppression equipment is not needed in the raised floor area. The justification for this position is based on the following: (a) The amount of cabling in this area is substantially reduced over current designs. N (b) The control room is continuously manned so that the presence of a fire will be quickly detected. (c) The types of cables located in the raised floor area smolder for a long time and are usually self extinguishing. (d) There has never been a fire in d.e operating plant that has required the evacuation of the control room. (e) In the unlikely event that the conttui room were to require evacuation the Remote Shutdown Panels provide the necessary controls to bring the plant to cold shutdown. The cabling that will be located in the raised floor area will be one of three types: (a) Fiber Optic Cables (b) Control and Signal Cables (c) Low Voltage Power Cables (<480 Volts) Divisional separation of these cables will maintained per requirements ofIEEE 384, Reg Guide 1.75, and GDC 17 (SSAR 8.3.3.1). For the raised floor area this effectively means that divisional cable trays will be separated by a minimum of 0.91 m horizontal or will be enclosed with at least 3 cm clearance. ; Furthermore, all low voltage power cables will be contained in flexible or rigid Analysis - Amendment 34 9A4-417
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23A6100 R;v. 9 ABWR Standard Salery Analysis Report O conduit in the raised floor areas. Cable contained in conduit or enclosed trays are not considered to contribute to the combustible loading for the room. The divisional panels are physically separated as much as practical and located above the divisional electrical equipment rooms. The cabling from the divisional electrical equipment rooms will be routed to the Safety System Logic Control (SSLC) cabinets with Divisions I and III on one side of the operator area and Divisions II and IV located on the opposite side of the operator area. There is a suspended ceiling but only cables associated with lighting and the fire alarm system are routed above the false ceiling. The cables are in conduit. Paper within the control room complex is required to be stored in approved containers (file cabinets, cabinets, waste baskets) except when in use. 9A.4.2.5 Floor Five El 12300 mm 9A.4.2.5.1 Control Room HVAC "B" Exhaust Duct Chase (Rm No. 522) (1) Fire Area-FC4220 (2) Equipment: See Table 9A.6-3 Safety-Related Provides Core Cooling l Yes, D2 No (3) Radioactive Material Present-None. (4) Qualification of Fire Barriers-Rm No. 522 is defined as a vertical section of HVAC chase extending from the ceiling of the control room, formed by the floor located at the 12300 mm elevation, to the floor of Rm No. 629 located at the 17150 mm elevation. All four walls are designated as fire barriers and are of three hour fire-resistive concrete construction. Access to Rm No. 522 from the 12300 mm level is provided by a three hour, fire-resistive removable panel. (5) Combustibles Present-(NCLL Applies) Fire Loading Total Heat of Type Combustion (MJ) 2 2 Cable in trays 727 MJ/m NCLL (727 MJ/m maximum average) applies 9A.4418 Analysis Amendment 37
23A6100 Rw. 9 l ABWR standardsateryAnalysis neport l O 1 (6) Detection Prosided-Class A Supenised POC detection system in the room l and manual pull alarm station at 1.62-J.60. l (7) Suppression Available: Type Location / Actuation l l Standpipe and hose reel 4.00-K.95 & 1.6 -J.5/ Manual l ABC hand extinguishers 4.0 - K.95 & 1.6 -J.5/ Manual (8) Fire Protection Design Criteria Employed: (a) The function is located in a fire area which is separate from fire areas providing alternate means of performing the safety or shutdown function. (b) Fire detection and suppression capability is provided and accessible. (c) Fire stops are provided for cable tray and piping penetrations through designated barriers. ; s 1 (9) Consequences of Fire-Postulated fire assumes loss of the function. Alternate j l means is provided by control room HVAC "C". i Smoke control is by the normal HVAC System functioning in the smoke I control mode. Refer to 9.5.1.1.6 for additional information. , 1 i (10) Consequences of Fire Suppreuion-Suppression extinguishes the fire. Refer i to Section 3.4, WaterLevel(Flood) Design, for the drain system. j i (11) Design Criteria Used for Protection Against Inadvertent Operation, Careless l Operation or Rupture of the Suppression System: I (a) Refer to Section 3.4, WaterLevel(Rood) Design, for the drain system. (b) Location of the manual suppression system in an area external to the room containing the safety-related equipment . 1 (12) Fire Containment or Inhibiting Methods Employed: i (a) The functions are located in a separate fire-resistive enclosure. p (b) The means of fire detection, suppression and alarming are provided and ( accessible. 1 Analysis Amendment 37 9A.4-419
23A6100 R5v. 4 i ABWR standardsafetyAnalysis Report 1 i O (c) Fire stops are provided for cable tray and piping penetrations through , rated fire barriers. l (13) Remarks-Quantities of cable may be so small that they will be in conduit rather than cable tray. , l 9A.4.2.5.2 HVAC "A" Supply (Rm Nos. 511,512 and 513) (1) Fire Area-FC1110 (2) Equipment: See Table 9A.6 3 Safety-Related Provides Core Cooling Yes, D1 Yes, DI See Remarks (3) Radioactive Material Present-None. (4) Qualification of Fire Barriers-The exterior walls of the space consisting of Rm Nos. 511,512,513 are common to adjacent fire areas FC1310 and FC5110 on the 12300 mm level. Therefore all space exterior walls are designated as fire barriers and are of three hour fire-resistive concrete construction. The ceiling is common to fire area FC1110 above and is not a fire barrier. The floor is common to fire area FC4910 below and is of three hour fire-resistive concrete construction. Access to this area is from Rm No. 593 through a three hour fire-resistive door. Access to Rm No. 512 and 513 from Rm No. 511 is via removable panels. (5) Combustibles Present-(NCLL Applies)) l Fire Loading Total Heat of l Type Combustion (MJ) 2 2 Cable in trays 727 MJ/m NCLL (727 MJ/m Bag filters maximum average) applies (6) Detection Provided-Class A Supervised POC detection system in the room and manual pull alarm station at 6.50-J75. O l 9A.4420 Analysis - Amendment 34
l 23A6100 Rsv. 4 ABWR studerdsafetyAnlysis Report ~O i (7) Suppression Available: Type Location / Actuation Standpipe and hose reel 6.60J.75/ Manual ] ABC hand extinguishers 6.50J.75 and 5.45J.50/ Manual
- (8) Fire Protection Design Criteria Employed
1 (a) The function is located in a fire-resistive enclosure.
- (b) Fire detection and suppression capability is provi . ed and accessible.
- (c) Fire stops are provided for cable tray and piping penetrations through
- designated fire barriers.
1
- (9) Consequences of Fire-Postulated fire assumes loss of the function:
Equipment (not in FC1310) on all elevations would remain operational, , HVAC A and B remain operational and will not be affected by the fire. Smoke control is by the nonnal HVAC System functioning in the smoke
. control mode. Refer to 9.5.1.1.6 for additional information.
1
- (10) Consequences of Fire Suppression--Suppression extinguishes the fire. Refer 4
to Section 3.4, WaterLevel(Flood) Design, for the drain system. l (11) Design Criteria Used for Protection Against Inadvertent Operation, Careless ! Operation or Rupture of the Suppression System: j (a) Refer to Section 3.4, WaterLevel(Rood) Design, for the drain system. (b) ANSI B31.1 standpipe (rupture unlikely) { l l (12) Fire Containment or Inhibiting Methods Employed: (a) The functions are located in a fire-resistive enclosure. 1 (b) The means of detection, suppression and alarming are provided and accessible. } (c) Fire stops are provided for cable tray and piping penetrations through rated fire barriers. (13) Remarks-None.
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i i Analysis - Amendtnent 34 9A.4-437
23A6100 R2v. 9 ABWR StandardSafety Analysis Report O 1 9A.4.2.5.12 Control Room HVAC "C," Exhaust Duct Chase (Rm No. 595) (1) Fire Area-FC4310 l (2) Equipment: See Table 9A.6-3 Safety-Related Provides Core Cooling l l Yes, D3 No, See Remarks. l l l (3) Radioactive Material Present-None. j l (4) Qualification of Fire Barriers-Rm No. 595 is defined as a vertical section of I HVAC chase extending from the ceiling of the control room, formed by the floor at the 12300 mm elevation, to the 17150 mm elevation. Walls common j to Rm No. 512 (FCll10), Rm No. 532 (FCl310) Rm No. 593 (FCl310) and Rm No. 506 (FC5110) are designated fire barriers and are of three hour fire-resistive concrete construction. Access to Rm No. 595 is provided by a l removable panel. { (5) Combustibles Present-(NCLL Applies) Fire Loading Total Heat of l Type Combustion (MJ) None 2 2 727 MJ/m NCLL (727 MJ/m maximum average) applies (6) Detection Provided-Class A Supervised POC detection system in the room and manual pull alarm station at 6.50-J.75. (7) Suppression Available: Type Location / Actuation Standpipe and hose reel 4.0 -J.1 & 6.60-J.67 on the 17150 level / Manual ABC hand extinguishers 4.0 -J.1 & 6.60 -J.67/ Manual (8) Fire Protection Design Criteria Employed: (a) The functica is located in a fire-resistive enclosure. (b) Fire detection and suppression capability is provided and accessible. 9A4438 Analysis - Amendment 37
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] 23A6100 Rsv. 9 l ABWR stondedsaforyanalysis neport 'i 1 i (9) Consequences of Fire-Postulated fire assumes loss of the function, alternate { l means is provided by control room HVAC "B". 3 Smoke control is by the normal HVAC System functioning in the smoke control mode. Refer to 9.5.1.1.6 for additional information, i j (10) Consequences of Fire Suppression-Suppression extinguishes the fire. Refer to Section S.4, WaterLevel(Flood) Design, for the drain system. j 2 ; j (11) Design Criteria Used for Protection Against Inadvertent Operation, Careless ! Operation or Rupture of the Suppression System. } l (a) Location of the manual suppression system in an area external to the room containing the safety-related equipment { (b) ANSI B31.1 standpipe (rupture unlikely) l l (12) Fire Containment or Inhibiting Methods Employed: l' (a) The functions are located in a separate fire-resistive enclosure. O (b) The means of fire detection, suppression and alarming are provided and j accessible. - 1
- (c) Fire stops are provided for cable tray and piping penetrations through i
rated fire barriers. l { (13) Remarks-Quantities of cable may be so small that they will be in conduit rather than cable tray. e 9A.4.2.5.13 Passageway (Rm No. 506) (1) Fire Area-FC5110 (2) Equipment: See Table 9A.fr3 j Safety-Related Provides Core Cooling j No No 4 j 1 (3) Radioactive Material Present-None. (4) Qualification of Fire Barriers-All walls of this passageway are designated fire barriers and are of three hour fire-resistive concrete construction.The ceiling ( is common to fire areas FC4310, FCl210, FC1110, FCl310, FC4220 and the steam tunnel above and is of three hour fire-resistive concrete construction. The floor is common to fire area FC4910 below and is also of three hour fire-Analysis - Amendment 37 9A.4-439
23A6100 Riv. 4 ABWR StandardSafetyAnalysis Report O resistive concrete construction. This corridor provides controlled access between the reactor and senice buildings. It serves no purpose for the control building and is therefore separated from the remainder of the control building by fire barriers. (5) Combustibles Present-(NCLL Applies) Fire Loading Total Heat of l Type Combustion (MJ) i None 2 2 ) 727 hf]/m NCLL (727 MJ/m maximum average) applies (6) Detection Provided-Class A Supenised POC detection system in the fire area and manual pull alarm station at 4.00-K95 (7) Suppression Available: Type Location / Actuation Standpipe and hose reel 4.00-K95/ Manual (8) Fire Protection Design Criteria Employed: (a) The function is located in a separate fire-resistive enclosure. i (b) Fire detection capability is provided. Fire suppression capability is provided. A backup manual hose is provided from the senice building. (c) Fire stops are provided for cable tray and piping penetrations through ! designated fire barriers. (9) Consequences of Fire-All systems would continue to function normally. Access between the senice building and reactor building would not be possible while a fire was in progress. Smoke control is by the normal HVAC System functioning in the smoke control mode. Refer to 9.5.1.1.6 for additional information. (10) Consequences of Fire Suppression-Suppression extinguishes the fire. Refer to Section 3.4, WaterLevel(Mood) Design, for the drain system. (11) Design Criteria Used for Protection Against Inadvertent Operation, Careless Operation or Rupture of the Suppression System: 9A440 Analysis - Amendment 34
23A6100 Rav. 9 ABWR standardsafetyAnslysisReport s I (10) Consequences of Fire Suppression-Suppression extinguishes the fire. Refer to Section 3.4, WaterLevel(&d) Design, for the drain system. (11) Design Criteria Used for Protection Against Inadvertent Operation, Careless Operation or Rupture of the Suppression System: (a) Refer to Section 3.4, WaterLevel(&d) Design, for the drain system. (b) Provision of raised supports for the equipment (c) Location of manual suppression system in an area external to the room containing the safety-related equipment (d) ANSI B31.1 standpipe (rupture unlikely) (12) Fire Containment or Inhibiting Methods Employed: (a) The functions are located in a separate fire-resistive enclosure. (b) The means of detection, suppression and alarming are prosided and accessible. (c) Fire stops are provided for cable tray and piping penetrations through rated-fire barriers. (13) Remarks-This equipment is also required to function to support equipment required for remote shutdown and therefore is in a fire area separate from the control room and its HVAC equipment. The exhaust fans do not provide any cooling function. They only serve a purge function which is not necessary to the cooling function of the IWAC System. 9A.4.2.6.6 Control Room HVAC Exhaust "B" (Rm No. 626) (1) Fire Area-FC4220 (2) Equipment: See Table 9A.f> 3 Safety-Related Provides Core Cooling l Yes, D2 No (3) Radioactive Material Present-None. (4) Qualification of Fire Barriers-The building exterior wall and the steam g tunnel wall are fire barriers of three hour fire-resistive concrete construction. The common interior walls between Rm Nos. 627 and 663 are in the same fire area and are not fire barriers. The ediing of this fire area forms a building Analysis Amendment 37 9A.4451 {
23AS100 REv. 9 ABWR standardsatery Analysis Report O exterior boundary and is of three hour fire-resistive concrete constmction. l The exhaust duct through the ceiling in Rm No. 626 does not have a fire damper. See Subsection 9.5.1.1.6 for a discussion of this design feature. A section of the floor common to fire area FCl210 and FC5010 below, is also of three hour fire-resistive concrete construction. Access to the CR HVAC Exhaust area is provided from Rm No. 622 through Rm No. 627. 4 (5) Combustibles Present-(NCLL Applies) Fire Loading Total Heat of Type Combustion (MJ) 2 2 Cable in trays 727 MJ/m NCLL (727 MJ/m ) maximum average) applies (6) Detection Provided-Class A Supervised POC detection system in the fire area and manual pull alarm station at 1.42-J.67. (7) Suppression Available: Type Location / Actuation Standpipe and hose reel 1.37-J.67/ Manual ABC hand extinguishers 1.42-J.67 and 1.30-K.55/ Manual (8) Fire Protection Design Criteria Employed: (a) The function is located in a fire area which is separate from fire areas providing alternate means of performing the safety or shutdown function. (b) Fire detection and suppression capability is provided and accessible. (c) Fire stops are provided for cable tray and piping penetrations through designated fire barriers. (9) Consequences of Fire-Postulated fire assumes loss of the function, but continued operation of the exhaust fans are not required for the equipment and systems served. If the CR HVAC "B" or "C" are placed in the smoke remowil mode the control room should remain habitable. 9A.4-452 Analysis - Amendment 37
l 23A6100 R:v. 9 l ABWR standedsarnyAulysis neport i i ( (9) Consequences of Fire-Postulated fire assumes loss of function. Even though j access to rooms 612,636,631,634 and 651 are not possible, the equipment in l ! these rooms are functional (they are in a different fire area). Alternate means j is provided by CRHVAC "B". (10) Consequences of Fire Suppression-Suppression extinguishes the fire. Refer l to Section 3.4, WaterLevel(bd) Design, for the drain system. l (11) Design Criteria Used for Protection Against Inadvertent Operation, Careless Operation or Rupture of the Suppression System: (a) Refer to Section 3.4, WaterIevel(&d) Design, for the drain system. (b) ANSI B31.1 standpipe (rupture unlikely) (12) Fire Containment or Inhibiting Methods Employed: (a) The functions are located in a fire-resisdve enclosure. (b) The means of detection, suppression and alarming are prosided and accessible. l (j (c) Fire stops are prosided for cable tray and piping penetrations through rated fire barriers. (13) Remarks-safety-related cooling for multiple divisions is prosided by redundant systems. The equipment on level 17150 in this fire area provides one division of cooling for the multi-divisional control room. 9A.4.2.6.10 Control Room HVAC Exhaust "C" (Rm No. 614) (1) Fire Area-FC4310 (2) Equipment:See Table 9A.6-3 Safety-Related Provides Core Cooling l Yes, D3 No, See Remarks. (3) Radioactive Material Present-None. (4) Qualification of Fire Barriers-All walls in this area are interior w alls. The w alls l common to fire area FC1110 are designated as fire barriers and are of three hour fire-resistive concrete construction. The rennining interior walls are not g fire barriers. The ceilingis a building exterior wall and is also of three hour s Analysis - Amendment 37 9A.4469
23A6100 RIV. 9 ABWR StandardSafety Analysis Report O fire-resistive concrete construction. The floor is common to adjacent fire area j FCl310 below, is of three hour fire-resistive concrete construction. Access to l l the CR HVAC "C" exhaust area is provided from Rm No. 631. l l (5) Combustibles Present-(NCLL Applies) l I Fire Loading Total Heat of Type Combustion (hfJ) ' Cable in trays 2 2 727 MJ/m NCLL (727 MJ/m maximum average) applies (6) Detection Provided-Class A Supervised POC detection system in the fire area and manual pull alarm station at 6.70-J.67. (7) Suppression Available: Type Location / Actuation Standpipe and hose reel 6.60-J.67/ Manual ABC hand extinguishers 6.60-J.67 and 6.70-K.55/ Manual (8) Fire Protection Design Criteria Employed: 1 (a) The function is located in a fire area which is separate from fire areas l providing alternate means of performing the safety or shutdown l function. (b) Fire detection and suppression capability is provided and accessible. (c) Fire stops are provided for cable tray and piping penetrations through designated fire barriers. h (9) Consequences of Fire-Postulated fire assumes loss of the function, but j continued operation of the exhaust fans is not required for the equipment and systems served. If the control room HVAC is manually switched to the smoke removal mode the control room should remain habitable. Smoke control is by the normal HVAC System functioning in the smoke I control mode. Refer to 9.5.1.1.6 for additional information. (10) Consequences of Fire Suppression-Suppression extinguishes the fire. Refer to Section 3.4, WaterLevel(Flood) Design, for the drain system. ! 9AA460 Analysis - Amendment 37 1
23A6100 R:v.1 ABWR standardsatoryAnstysis asport e O i ) V l Table 9A.6-2 Fire Hazard Analysis Equipment Database Sorted by Room - Reactor Building (Continued) Location Location hem Elect Elev. Number Alpha System Room No. MPL No Div. Location Coord. Coord. Description Drawing No. 2297 G41-F005B N 18100 1.8 A.3 MO GATE VALVE (ISOL) 107E6042/0 546 2298 G41-F021A 1 18100 1.8 A.3 MO GLOBE VALVE 107E6042/0 546 2299 G41-F0218 2 18100 1.8 A.3 MO GLOBE VALVE 107E6042/0 546 2300 G41 TE002 N 18100 1.4 A.3 TEMP ELEMENT 107E6042/0 546 2301 H23-P027' N 18100 1.9 B.2 MULTIPLEXER 547 2302 H23-P028' N 18100 1.9 B.3 MULTIPLEXER 547 2303 U41-0109 1 18100 1.4 A.7 FPC PUMP (A) ROOM HVH 107E5189/0 547 2304 U41 D110 2 18100 1.8 A.7 FPC PUMP (B) ROOM HVH 107E5189/0 547 2305 G41-FT006A N 18100 2.1 A.7 FLOW TRANSMITTER 107E6042/0 547 2306 G41-FT006B N 18100 2.1 A.7 FLOW TRANSMITTER 107E6042/0 547 2307 G41-PT003A N 18100 2.1 A.7 PRESS TRANSMITTER 107E6042/0 547 2308 G41-PT0038 N 18100 2.1 A.7 PRESS TRANSMITTER 107E6042/0 547 2309 H22-PO42' [s N 18100 2.1 A.7 FPC CU SYS INST RACK 100273-284 547 ( 2310 B21-TE012A 1 23000 4.0 C.3 HMP ELEMENT 103E1791/1 591 2311 821-TE012C 2 23000 3.5 D.5 TEMP ELEMENT 103E1791/1 591 2312 821 TE013A 1 23000 4.0 C.3 HMP ELEMENT 103E1791/1 591 2313 B21-TE013C 2 23000 3.5 D.5 TEMP ELEMENT 103E1791/1 591 2314 B21-TE019A 1 18100 5.0 C.0 TEMP ELEMENT 103E1791/1 591 2315 B21 TE019B 2 18100 3.0 E.0 TEMP ELEMENT 103E1791/1 591 2316 E31-TE002 N 18500 4.0 E.0 DW AMB TEMP ELEMENT 103E1792/1 591 2317 E31-TE003 N 22000 4.0 E.0 DW AMB TEMP ELEMENT 103E1792/1 591 2318 T31 TE051D N 18100 3.5 C.8 TEMP ELEMENT 107E6043/0 591 2319 T31-TE051E N 18100 3.3 C.2 TEMP ELEMENT 107E6043/0 591 2320 T31 TE051F N 18100 3.5 C.0 TEMP ELEMENT 107E6043/0 591 2321 T31-TE051G N 18100 4.5 C.0 TEMP ELEMENT 107E6043/0 591 2322 T31-TE051H N 18100 4.7 C.2 TEMP ELEMENT 107E6043/0 591 2323 T31-TE052A N 18100 4.8 C.5 TEMP ELEMENT 107E6043/0 591 2324 T31-TE052B N 18100 3.2 C.9 TEMP ELEMENT 107E6043/0 591 2325 T31 TE052C N 18100 4.2 D.9 TEMP ELEMENT 107E6043/0 591 2326 T31 TE052D N 18100 4.8 C.5 TEMP ELEMENT 107E6043/0 591 2327 T31-TE052E N 18100 3.2 C.9 TEMP ELEMENT 107E6043/0 591 2328 T31-TE052F N 18100 4.2 D.9 TEMP ELEMENT 107E6043/0 591
% j 2329 R43-A005A' 1 23500 6.5 A.2 FUEL OIL DAY TANK SSAR FIG 9.5-6 610 2330 R43-LS395A' 1 23500 6.8 B.O LEVEL SWITCH SSAR FIG 9.5-6 612 Fire Hazard Analysis Database - Amer:dment 31 9A.6-77
23A6100 Riv. 9 ABWR standardsatory Analysis neport O Table 9A.6-2 Fire Hazard Analysis Equipment Database Sorted by Room - Reactor Building (Continued) Location Location item Elect Elev. Number Alpha System Room No. MPL No Div. Location Coord. Coord. Description Drawing No. 2331 U41-82028 1 27200 6.8 A.8 COU. COIL ELEC EQ (A) 107E5189/0 612 2332 U41-B202A 1 27200 6.5 A.8 COOL COIL ELEC EQ (A) 107E5189/0 612 2333 U41-F005A 1 27200 6.4 A.5 MOVALVE 107E5189/0 613 2334 P54-A001A 1 23500 6.2 B.2 N2 STORAGE BOTTLE 107E5128/0 613 2335 P54-A001C 1 23500 6.2 B.2 N2 STORAGE BOTTLE 107E5128/0 613 2336 P54.A001E 1 23500 6.2 B.2 N2 STORAGE BOTTLE 107E5128/0 613 2337 P54 A001G 1 23500 6.2 B.2 N2 STORAGE BOTTLE 107E5128/0 613 2338 P54-A001J 1 23500 6.2 B.2 N2 STORAGE BOTTLE 107E5128/0 613 2339 P54.A001L 1 23500 6.2 B.2 N2 STORAGE BOTTLE 107E5128/0 613 2340 P54-A001N 1 23500 6.2 B.2 N2 STORAGE BOTTLE 107E5128/0 613 2341 P54 A0010 1 23500 6.2 B.2 N2 STORAGE BOTTLE 107E5128/0 613 2342 P54-A001S 1 23500 6.2 B.2 N2 STORAGE BOTTLE 107E5128/0 613 2343 P54.A001U 1 23500 6.2 B.2 N2 STORAGE BOTTLE 107E5128/0 613 2344 P54-F003A 1 23500 6.2 B.2 MO GLOBE VALVE 107E5128/0 613 2345 P54-F012A 1 23500 6.2 B.2 MO GLOBE VALVE 107E5128/0 613 2346 P54-PIS001A 1 23500 6.2 B.2 PRESS IND SWITCH 107E5128/0 613 2347 U41-C202A 1 23500 6.4 B.2 DG(A) HVAC EXH FAN A 107E5189/0 613 2348 U41 C202E 1 23500 6.4 B.5 DG(A) HVAC EXH FAN E 107E5189/0 613 2348a U41-C210A N 23500 6.4 B.5 SREE HVAC Smoke 107E5189/0 613 1 Removal Fan A 2349 U41-F006A 1 23500 6.4 B.5 MO VALVE 107E5189/0 613 2350 R24 MCC A310 N 23500 6.3 C.0 MCC A310 - R/B 107E5072/0 613 2351 R43-C201 A* 1 23500 6.6 B.8 DG AIR COMPRESSOR A SSAR FIG 9.54 613 2352 R434202A' 1 23500 6.9 B.8 DG AIR COMPRESSOR A SSAR FIG 9.5 8 613 2353 D21-RE007 N 23 7.0 5.2 B.0 AREA RAD DETECTOR 299X701-171/0 615 2354 T31.SS A051 N 23500 5.3 C.3 SELECT SWITCH 107E6043/0 615 2355 T31 SS A053 N 23500 5.3 C.3 SELECT SWITCH 107E6043/0 615 2356 T31-T1051 N 23500 5.3 C.4 TEMP INDICATOR 107E6043/0 615 2357 T31 T1053 N 23500 5.3 C.3 TEMP INDICATOR 107E6043/0 615 2358 T31 TT051 N 23500 5.3 C.3 TEMP TRANSMITTER 107E6043/0 615 2359 T31-TT053 N 23500 5.3 C.3 TEMP TRANSMITTER 107E6043/0 615 2360 U41-C103 N 23500 5.4 C.1 PCV PURGE SUPPLY FAN 107E5189/0 615 2361 U41-F004C 3 29000 5.8 B.8 MO VALVE 107E5189/0 616 2362 U41-F101C 3 29000 5.8 B.8 MO VALVE 107E5189/0 616 2363 T31-F731 1 23500 5.8 C.8 SO VALVE 107E6043/0 616 9A.6-78 Fire Hazard Analysis Database - Amendment 37
i 23A6100 Rhv. 9 StandardSafety Analysis Report
/'^
i V) f Table 9A.6-2 Fire Hazard Analysis l Equipment Database Sorted by Room - Reactor Building (Continued) I Location Location ltem Elect Elev. Number Alpha System Room No. MPL No Div. Location Coord. Coord. Description Drawing No. 2364 T31-PT054 N 23500 5.8 C.8 PRESSURE TRANSMITTER 107E6043/0 616 i 2365 D21-RE013 N 23500 5.8 D.0 AREA RAD DETECTOR 299X701 171/0 616 2366 T31-PT055A N 23500 5.5 D.5 PRESSURE TRANSMITTER 107E6043/0 617 2367 U41-0134A N 23500 5.3 D.8 ISI ROOM FCU A 107E5189/0 617 2368 R43-A0058' 2 23500 1.5 F.8 FUEL OIL DAY TANK SSAR FIG 9.5-6 620 2369 D11-0302 N 23500 2.7 F.4 FILTER DEVICE 107E6071/0 621 2370 D23-C0018 2 23500 2.7 F.4 ACCIDENT SMPL PUMP 107E5139/1 621 2371 D23-C002B 2 23500 2.7 F.4 NORM SMPLBOOSTER 107E5139/1 621 PUMP 2372 D23-C003B 2 23500 2.7 F.4 ACC. SMPL. BOOSTER 107E5139/1 621 PUMP 2373 D23-C004B 2 23500 2.7 F.4 NORM. SMPL PUMP 107E5139/1 621 2374 D23 0010B 2 23500 2.7 F.4 STEAM SEPARATOR 107E5139/1 621 2375 D23-D0128 2 23500 2.7 F.4 DEHUMIDIFIER 107E5139/1 621
\'
2376 D23-D0228 2 23500 2.7 F.4 DRAIN MEAS VESSEL 107E5139/1 621 2377 D23-F105B 2 23500 2.7 F.4 SO VALVE 107E5139/1 621 2378 D23-F108B 2 23500 2.7 F.4 SO VALVE 107E5139/1 621 2379 023-F1188 2 23500 2.7 F.4 SO VALVE 107E5139/1 621 2380 D23-F121B 2 23500 2.7 F.4 SO VALVE 107E5139/1 621 2381 D23-F123B 2 33500 2.7 F.4 SO VALVE 107E5139/1 621 2382 D23-F1278 2 23500 2.7 F.4 SO VALVE 107E5139/1 621 2383 D23-F130B 2 23500 2.7 F.4 SO VALVE 107E5139/1 621 2384 D23 F132B 2 23500 2.7 F.4 SO VALVE 107E5139/1 621 2385 D23-F190B 2 23500 2.7 F.4 SO VALV;' 107E5139/1 621 2386 023-F191B 2 23500 2.7 F.4 SO VALVE 107E5139/1 621 2387 D23 F193B 2 23500 2.7 F.4 SO VALVE 107E5139/1 621 2388 D23-F1958 2 23500 2.7 F.4 SO VALVE 107E5139/1 621 2389 D23-F1978 2 23500 2.7 F.4 SO VALVE 107E5139/1 621 2390 D23f2018 2 23500 2.7 F.4 AO VALVE 107E5139/1 621 l 2391 D23-F202B 2 23500 2.7 F.4 AO VALVE 107E5139/1 621 2392 D23-F5108 2 23500 2.7 F.4 AO VALVE 107E5139/1 621 2393 023-F513B 2 23500 2.7 F.4 AO VALVE 107E5139/1 621 'f% ' 2394 D23-F515B 2 23500 2.7 F.4 AO VALVE 107E5139/1 621 b 2395 D23-FIT 019B 2 23500 2.7 F.4 FLOW IND TRANSMITTER 107E5139/1 621 2396 023-H2AM001B 2 23500 2.7 F.4 HYDROGEN ANALYZER 107E5139/1 621 Fire Hazard Analysis Database Amendment 37 9A.6-79
23A6100 R1v. 9 \ ABWR StandardSafetyAnalysis Report l 9 Table 9A.6-2 Fire Hazard Analysis Equipment Database Sorted by Room - Reactor Building (Continued) l l Location Location l Item Elect Elev. Number Alpha System Room l No. MPL No Div. Location Coord. Coord. Description Drawing No. l 2397 D23-H2E0018 2 23500 2.7 F.4 HYDROGEN ANAL ELEM 107t5139/1 621 2398 D23-LIT 031B 2 23500 2.7 F.4 LEVEL IND TRANS 107E5139/1 621 l 2399 D23-02AM003B 2 23500 2.7 F.4 OXYGEN ANALYZER 107E5139/1 621 2400 D23-02E0038 2 23500 2.7 F.4 OXYGEN ANAL ELEM 107E5139/1 621 l 2401 D23-PIS0178 2 23500 2.7 F.4 PRESS IND SWITCH 107E5139/1 621 j 2402 D23-PIT 0218 2 23500 2.7 F.4 PRESS IND TRANSMITTER 107E5139/1 621 1 2403 D23-PS0248 2 23500 2.7 F.4 PRESS SWITCH 107E5139/1 621 2404 D23-PS0268 2 23500 2.7 F.4 PRESS SWITCH 107E5139/1 621 2405 D23-PS027B 2 23500 2.7 F.4 PRESS SWITCH 107E5139/1 621 2406 D23-SC0338 2 23500 2.7 F.4 STEAM CONDENSER 107E5139/1 621 2407 D23-TE0208 2 23500 2.7 F.4 TEMP ELEMENT 107E5139/1 621 2408 D23 TS0168 2 23500 2.7 F.4 TEMP SWITCH 107E5139/1 621 2409 H22-P0538' 2 23500 2.7 F.4 023, CAMS RACK B 107E5139/1 621 2410 H22-P0548' 2 23500 2.7 F.4 D23, CAMS CAllB RACK B 107E5139/1 621 2411 U41-F003B 2 23500 2.7 F.4 MO VALVE 107E5189/0 621 2412 021-RE008 N 23500 2.8 F.1 AREA RAD DETECTOR 299X701 171/0 622 2413 C41-C001A 1 23500 2.4 E.2 SLC INJECTION PUMP A 107E6016/0 622 2414 C41-C001B 2 23500 2.2 E.2 SLC INJECTION PUMP B 107E6016/0 622 2415 C41-A001 N 23500 2.3 D.6 SLC STORAGE TANK 107E6016/0 622 2416 C41-A002 N 23500 2.1 D.2 SLC TEST TANK 107E6016/0 622 2417 C41-B001 N 23500 2.3 D.6 SLC MIXING HEATER 107E6016/0 622 2418 C41-8002 N 23500 2.3 D.6 SLC OPERATING HEATER 107E6016/0 622 2419 C41-F001A 1 23500 2.5 E.0 MO GLOBE VALVE (SUCT) 107E6016/0 622 2420 C41-F0018 2 23500 2.2 E.0 MO GLOBE VALVE (SUCT) 107E6016/0 622 2421 C41-F010 N 23500 2.1 D.2 MAN OPER GLOBE VALVE 107E6016/0 622 2422 C41 F012 N 23500 2.1 D.2 MAN OPER GATE VALVE 107E6016/0 622 2423 C41-LE001 N 23500 2.3 D.6 LEVEL SENSOR 107E6016/0 622 2424 C41-L1001 N 23500 2.3 D.6 LEVEL INDICATOR 107E6016/0 622 l 2425 C41 LT001 N 23500 2.3 D.6 LEVEL TRANSMITTER 107E6016/0 622 2426 C41-PT005 N 23500 2.1 D.2 PRESS TRANSMITTER 107E6016/0 622 2427 C41 TE002 N 23500 2.3 D.6 TEMP ELEMENT 107E6016/0 622 2428 C41-TE003 N 23500 2.3 0.6 TEMP ELEMENT 107E6016/0 622 2429 C41 TE006 N 23500 2.3 D.6 TEMP ELEMENT 107E6016/0 622 2430 C41 TIS 002 N 23500 2.3 0.6 TEMP SWITCH 107E6016/0 622 I 9A.6-80 Fire Hazard Analysis Database - Amendment 37
23A6100 Rw. 9 ABWR standardseteryAnalysis neport v Table 9A.6-2 Fire Hazard Analysis Equipment Database Sorted by Room - Reactor Building (Continued) Location Location , item Elect Elev. Number Alpha System Room l No. MPL No Div. Location Coord. Coord. Description Drawing No. 2431 C41-TIS 003 N 23500 2.3 D.6 TEMP SWITCH 107E6016/0 622 2432 C41 TIS 006 N 23500 2.3 0.6 TEMP SWITCH 107E6016/0 622 2433 U41-C104 N 23500 2.3 E.7 PCV PURGE EXHAUST 107Eb189/0 623 FAN 2434 R43-LS395B* 2 23500 1.3 E.9 LEVEL SWITCH SSAR FIG 9.5-6 624 2435 U4182048 2 27200 1.2 F.2 COOL COIL,ELEC EQ (B) 107E5189/0 663 2436 U41-8204F 2 27200 1.5 F.2 COOL Coll,ELEC EQ (B) 107E5189/0 663 2437 U41 F005B 2 27200 1.6 F.3 MOVALVE 107E5189/0 663 2438 U41-C205F 2 23500 1.8 E.7 0G(B) HVAC EXH FAN F 107E5189/0 625 2438a U41-C211B N 23500 1.8 E.7 SREE HVAC Smoke 107E5189/0 625 1 Removal Fan B 2439 U41-F006B 2 23500 1.6 E.5 MOVALVE 107E5189/0 625 2440 R43-C201B' 2 23500 1.1 E.2 DG AIR COMPRESSOR B SSAR FIG 9.5-8 625 [] 2441 R43-C202 B' 2 23500 1.3 E.2 DG AIR COMPRESSOR B SSAR FIG 9.5 8 625 \d 2442 U41 C205B 2 23500 1.8 E.4 DG(B) HVAC EXH FAN B 107E5189/0 625 2443 R24 MCC B310 N 23500 1.8 D.5 MCC B310 - R/B 107E5072/0 625 2444 R43-A005C* 3 23500 6.5 F.8 FUEL OIL DAY TANK SSAR FIG 9.5-6 630 2445 R43-LS395C* 3 23500 6.7 E.9 LEVEL SWITCH SSAR FIG 9.5-6 632 2446 U41-B206G 3 27200 6.8 F.2 COOL Coll,ELEC EQ (C) 107E5189/0 673 2447 U41-B206C 3 27200 6.5 F.2 COOL Coll,ELEC EQ (C) 107E5189/0 673 2448 H22-PO44A' 1 23500 6.3 F.1 CAMS GAS CYL RACX A 107E5139/1 633 2449 U41-F005C 3 27200 6.4 F.3 MO VALVE 107E5189/0 673 2449a U41-C212C N 23500 6.3 E.5 SREE HVAC Smoke 107E5189/0 633 l Removal Fan C 2450 R24 MCC C310 N 23500 6.3 E.5 MCC C310 - R/B 107E5072/0 633 2451 R43-C201C' 3 23500 6.6 E.2 DG AIR COMPRESSOR C SSAR FIG 9.5-8 633 2452 R43.C202C' ' 3 23500 6.6 E.4 DG AIR COMPRESSOR C SSAR FIG 9.5-8 633 2453 R10 C001E' N 7"l500 6.3 D.5 RIP ASD OUTPUT XFMR 638 2454 R10 C001B+ N 23500 6.3 C.7 RIP ASD OUTPUTXFMR 638 2455 U41-0134B N 23500 5.5 E.4 ISI ROOM FCU B 107E5189/0 639 2456 P54 A0018 2 23500 13 *a N2 STORAGE BOTTLE 107E5128/0 640 2457 P54.A0010 2 23500 1.s A.4 N2 STORAGE BOTTLE 107E5128/0 640 2458 P54-A001F 2 23500 1.8 A.4 N2 STORAGE BOTTLE 107E5128/0 640 2459 P54 A001H 2 23500 1.8 A.4 N2 STORAGE BOTTLE 107E5128/0 640 NJ 2460 P54 A001K 2 23500 1.8 A.4 N2 STORAGE BOTTLE 107E5128/0 640 Fire Hazard Analysis Database - Amendment 37 9A.6-81
23A6100 Rsv. 9 ABWR standardsarety Analysis neport O Table 9A.6-2 Fire Hazard Analysis Equipment Database Sorted by Room - Reactor Building (Continued) location Location item Elect Elev. Number Alpha System Room No. MPL No Div. Location Coord. Coord. Description Drawing No. 2461 P54.A001M 2 23500 1.8 A.4 N2 STORAGE BOTTLE 107E5128/0 640 2462 P54.A001P 2 23500 1.8 A.4 N2 STORAGE BOTTLE 107E5128/0 640 2463 P54-A001R 2 23500 1.8 A.4 N2 STORAGE BOTTLE 107E5128/0 640 2464 P54 A001T 2 23500 1.8 A.4 N2 STORAGE BOTTLE 107E5128/0 640 2465 P54-A001V 2 23500 1.8 A.4 N2 STORAGE BOTTLE 107E5128/0 640 2466 P54-F003B 2 23500 1.8 A.6 MO GLOBE VALVE 107E5128/0 640 2467 P54.F012B 2 23500 1.8 A.6 MO GLOBE VALVE 107E5128/0 640 2468 P54-F203 N 23500 1.8 A.6 MO GLOBE VALVE 107E5128/0 640 2469 P54-PIS0018 2 23500 1.8 A.6 PRESS IND SWITCH 107E5128/0 640 2470 P54-PT004 N 23500 1.8 A.6 PRESS TRANSMITTER 107E5128/0 640 2471 R23 P/C EN110A N1 23500 1.5 A.5 P/C EN110A - LO VOLT 107E5072/0 640 l SWTGR 2472 R23 P/C EN110B N2 23500 1.2 A.5 P/C/EN1108 - LO VOLT 107E5072/0 640 I SWTGR 2473 R23 P/C EN110C N3 23500 1.2 A.2 P/C EN110C - LO VOLT 107E5072/0 640 I SWTGR 2474 H22-PO44B' 2 23500 1.7 B.8 CAMS GAS CYL RACK B 107E5139/1 640 2475 R24 MCC N1 23500 1.3 B.5 MCC EN110A - R/D 107E5072/0 640 1 EN110A 2475a R24 MCC N2 23500 1.3 B.5 MCC EN110B - R/B 107E5072/0 640 I EN110B 2476 R24 MCC N3 23500 1.3 B.7 MCC EN110C - R/B 107E5072/0 640 I EN110C 2477 H23.P029' N 23500 1.9 C.3 MULTIPLEXER --? - 640 2478 H23-P030' N 23500 1.9 C.5 MULTIPLEXER - 640 2479 H23-P031' N 23500 1.9 C.7 MULTIPLEXER 640 2480 D11-F053 N 23500 2.4 C.9 SOLENOID VALVE 107E6071/0 641 2481 011-F054 N 23500 2.4 C.9 SOLENOID VALVE 107E6071/0 641 2482 011-RE002A N 23500 2.5 C.1 SBGTS lON CHAMBER 107E6071/0 641 2483 D11-RE0028 N 23500 2.5 C.1 SBGTS ION CHAMBER 107E6071/0 641 2484 H22-P043B 2 23500 2.2 C.5 CBGT INSTR RACK 100273 285 641 2485 P54-DPSCO3 N 23500 2.0 C.5 DIFF PRESS SWITCH 107E5128/0 641 2486 T22-B0018 2 23500 2.2 C.7 DRYER HEATER B 107E6046/1 641 2487 T22-C001B 2 23500 2.2 C.1 EXHAUST FAN B 107E6046/1 641 2488 T22-C002B 3 23500 2.2 C.7 COOUNG FAN B 107E6046/1 641 2489 T22-C003B' 2 23500 2.2 C.6 PREHTR & FAN B - FLTR 107E6046/1 641 9A.6-82 Fire Hazard Analysis Database - Amendment 37
23A6100 Rev. 9 ABWR standardsateryAnalysisReport O V Table 9A.6-2 Fire Hazard Analysis Equipment Database Sorted by Room - Reactor Building (Continued) Location Location item Elect Elev. Number Alpha System Room No. MPL No Div. Location Coord. Coord. Descrirtion Drawing No. 2490 T22.C0048* 2 23500 2.2 C.6 AFTRHTR & FAN B - FLTR 107E6046/1 641 2491 T22-D003B 2 23500 2.2 C.6 PRE HEPA FILTER B 107ESO46/1 641 2492 T22-D004B 2 23500 2.2 C.2 POST HEPA FILTER B 107E6046/1 641 2493 T22-DOO2B' 2 23500 2.2 C.7 PRE FILTER TRAIN 107E6046/1 641 2494 T22-DP10038 2 23500 2.2 C.7 DIFF PRESS INDICATOR 107E6046/1 641 2495 T22 DP1007B 2 23500 2.2 C.7 DIFF PRESS INDICATOR 107E6046/1 641 2496 T22-DPl008B 2 23500 2.2 C.6 OlFF PRESS INDICATOR 107E6046/1 641 2497 T22-DP10128 1 23500 2.2 C.6 DIFF PRESS INDICATOR 107E6046/1 641 2498 T22-DPl0178 2 23500 2.2 C.2 DIFF PRESS INDICATOR 107E6046/1 641 2499 T22-DPT003B 2 23500 2.2 C.7 DIFF PRESS XMTR 107E6046/1 641 2500 T22-DPT007B 2 23500 2.2 C.7 DIFF PRESS XMTR 107E6046/1 641 2501 T22-DPT008B 2 23500 2.2 C.6 DIFF PRESS XMTR 107E6046/1 641
% 2502 T22-DPT012B 2 23500 2.2 C.6 DIFF PRESS XMTR 107E6046/1 641
( 2503 T22-DPT0178 2 23500 2.2 C.2 DIFF PRESS XMTR 107E6046/1 641 2504 T22-F002B 2 23500 2.2 C.9 MO BUTTERFLY VALVE 107E6046/1 641 2505 T22 F002C 3 23500 2.2 B.8 MO BUTTERFLY VALVE 107E6046/1 641 2506 T22-F004B 2 23500 2.2 C.1 MO BUTTERFLY VALVE 107E6046/1 641 2507 T22-F0058 2 23500 2.3 C.7 MO VALVE 107E6046/1 641 2508 T22-F020 3 23500 2.2 B.8 AO VALVE 107E6046/1 641 2509 T22-F022 2 23500 2.2 C.9 AO VALVE 107E6046/1 641 2510 T22 F511B 2 23500 2.3 C.1 MAN OPER VALVE 107E6046/1 641 2511 T22-FT0188 2 23500 2.2 C.9 FLOW TRANSMITTER 107E6046/1 641 2512 T22-FT018C 3 23500 2.2 B.8 FLOW TRANSMITTER 107E6046/1 641 2513 T22-LS004B 2 23500 2.2 C.7 LEVEL SWITCH 107E604G/1 641 2514 T22-LS019B 2 23500 2.2 C.1 LEVEL SWITCH 107E6046/1 641 2515 T22-ME011B 2 23500 2.2 C.6 MOISTURE ELEMENT 107E6046/1 641 2516 T22-ME012B 2 23500 2.2 C.6 MOISTURE ELEMENT 107E6046/1 641 2517 T22-MT011B 2 23500 2.2 C.6 MOISTURE TRANSMITTER 107E6046/1 641 2518 T22-MT012B 2 23500 2.2 C.6 MOISTURE TRANSMITTER 107E6046/1 641 j 2519 T22 POE001B 2 23500 2.2 C.9 POSITION ELEMENT 107E6046/1 641 2520 T22.POE001C 3 23500 2.2 B.8 POSITION ELEMENT 107E6046/1 641
, 2521 T22-TE002B 2 23500 2.2 C.7 TEMP ELEMENT 107E6046/1 641 2522 T22-TE0108 2 23500 2.2 C.6 TEMP ELEMENT 107E6049/1 641 2523 T22-TE013B 2 23500 2.2 C.6 TEMP ELEMENT 107E6046/1 641 Fire Hazard Analysis Database - Amendment 37 9A.6-83
23A6100 Rev. 9 ABWR Standard Safety Analysis Report l O Table 9A.6-2 Fire Hazard Analysis Equipment Database Sorted by Room - Reactor Building (Continued) Location Location item Elect Elev. Number Alpha System Room N o. MPL No Div. Location Coord. Coord. Description Drawing No. 2524 T22-TE014B 2 23500 2.2 C.3 TEMP ELEMENT 107E6046/1 641 2525 T22 TE0168 2 23500 2.2 C.2 TEMP ELEMENT 107E6046/1 641 2526 T22 TS0058 2 23500 2.2 C.7 TEMP SWITCH 107E6046/1 641 2527 T22 TS009B 2 23500 2.2 C.6 TEMP SWITCH 107E6046/1 641 2528 T22 TS0138 2 23500 2.2 C.6 TEMP SWITCH 107E6046/1 641 2529 T22-TS0158 2 23500 2.2 C.2 TEMP ELEMENT 107E6046/1 641 2530 U41-D112 2 23500 2.5 C.9 SGTS ROOM HVH (B) 107E5189/0 641 2531 H22.PO43A 3 23500 2.4 B.5 SBGT INSTR RACK 100273-285 642 2532 T22-B001C 3 23500 2.2 B.6 DRYER HEATER C 107E6046/1 642 2533 T22-C001C 3 23500 2.2 B.2 EXHAUST FAN C 107E6046/1 642 2534 T22-C002C 2 23500 2.2 B.6 COOLING FAN C 107E6046/1 642 2535 T22-C003C' 3 23500 2.2 B.5 PREHTR & FAN C - FLTR 107E6046/1 642 2536 T22-C004C' 3 23500 2.2 B.5 AFTRHTR & FAN C - FLTR 107E604G/1 642 2537 T22-0002C' 3 23500 2.2 B.6 PRE FILTER TRAIN 107E6046/1 642 2538 T22-D003C 3 23500 2.2 B.5 PRE HEPA FILTER C 107E6046/1 642 2539 T22.D004C 3 23500 2.2 B.3 POST HEPA FILTER C 107E6046/1 642 2540 T22.DP1003C 3 23500 2.2 B.5 DIFF PRESS INDICATOR 107E6046/1 642 l 2541 T22-DPl007C 3 23500 2.2 B.6 DIFF PRESS INDICATOR 107E6046/1 642 2542 T22 DP1008C 3 23500 2.2 B.5 DIFF PRESS INDICATOR 107E6046/1 642 2543 T22-DP1012C 3 23500 2.2 B.5 DIFF PRESS INDICATOR 107E6046/1 642 l 2544 T22 DP101?C 3 23500 2.2 B.3 DIFF PRESS INDICATOR 107E6046/1 642 2545 T22-DPT003C 3 23500 2.2 B.5 DIFF PRESS XMTR 107E6046/1 642 2546 T22-DPT007C 3 23500 2.2 B.6 DIFF PRESS XMTR 107E6046/1 642 2547 T22-DPT008C 3 23500 2.2 B.5 DIFF PRESS XMTR 107E6046/1 642 2548 T22-DPT012C 3 23500 2.2 B.5 DIFF PRESS XMTR 107E6046/1 642 2549 T22-DPT017C 3 23500 2.2 B.3 OlFF PRESS XMTR 107E6046/1 642 2550 T22-F004C 3 23500 2.2 B.2 MO BUTTERFLY VALVE 107E6046/1 642 2551 T22-F005C 3 23500 2.3 B.3 MO VALVE 107E6046/1 642 2552 T22 F511C 3 23500 2.3 B.2 MAN OPER VALVE 107E6046/1 642 l 2553 T22-LS004C 3 23500 2.2 B.6 LEVEL SWITCH 107E6046/1 642 l 2554 T22-LS019C 3 23500 2.2 B.2 LEVEL SWITCH 107E6046/1 642 l 2555 T22-ME011C 3 23500 2.2 B.5 MOISTURE ELEMENT 107E6046/1 642 ! 2556 T22-ME012C 3 23500 2.2 B.5 MOISTURE ELEMENT 107E6046/1 642 l 2557 T22-MT011C 3 23500 2.2 B.5 MOISTURE TRANSMITTER 107E6046/1 642 9A.6-84 Fire Hazard Analysis Database Amendment 37
l 23A6100 Rsv. 9 \ ABWR standardSafetyAnalysis Report n
/ i
() l Table 9A.6-2 Fire Hazard Analysis Equipment Database Sorted by Room - Reactor Building (Continued) l Location Location l Item Elect Elev. Number Alpha System Room
- No. MPL No Div. Location Coord. Coord. Description Drawing No.
2558 T22-MT012C 3 23500 2.2 B.5 MOISTURE TRANSMITTER 107E6046/1 642 2559 T22-TE002C 3 23500 2.2 B.6 ,TEMD ELEMENT 107E6046/1 642 2560 T22 TE010C 3 23500 2.2 B.5 TEMP EL6:cNT 107E6046/1 642 l 2561 T22 TE013C l 3 23500 2.2 B.5 TEMP ELEMENT 107E6046/1 642 ' 2562 T22-TE014C 3 23500 2.2 B.4 TEMP ELEMENT 107E6046/1 642 2563 T22-TE016C 3 23500 2.2 B.3 TEMP ELEMENT 107E6046/1 642 2564 T22-TS005C 3 23500 2.2 B.6 TEMP SWITCH 107E6046/1 642 2565 T22 TS009C 3 23500 2.2 B.5 TEMP SWITCH 107E6046/1 642 2566 T22 TS013C 3 23500 2.2 B.5 TEMP SWITCH 107E6046/1 642 l 2567 T22 TS015C 3 23500 2.2 B.3 TEMP ELEMENT 107E6046/1 642 2568 U41-0111 3 23500 2.2 B.2 SGTS ROOM HVH (A) 107E5189/0 642 2569 R24 MCC SB110 N 23500 2.8 C.5 MCC SB110 - R/B 107E5072/0 643 f5 2570 R24 MCC SB111 N 23500 2.8 C.5 MCC SB111 - R/B 107E5072/0 643 2571 U41-D114 2 27500 2.7 C.7 CAMS (B) ROOM HVH 107E5189/0 643 2572 U41-F004B 2 29000 2.8 B.5 MO VALVE 107E5189/0 643 2573 U41-F1018 2 29000 2.8 B.8 MO VALVE 107E5189/0 643 2574 U41-OPIS013 N 27500 2.8 A.7 DIFF PESS IND SENSOR 107E5189/0 643 2575 U41-F002A 1 27500 2.8 A.7 AO VLV - R/A EXH ISO (A) 107E5189/0 643 2576 U41-F002B 2 27500 2.8 A.9 AO VLV - R/A EXH ISO (B) 107E5189/0 643 2577 U41-C201A 1 27600 6.5 B.2 DG(A) HVAC SUPP FAN A 107E5189/0 653 2578 U41-C201E 1 27600 6.8 B.2 DG(A) HVAC SUPP FAN E 107E5189/0 653 2579 U41-TE052 1 27600 6.7 B.3 TEMP ELEMENT 107E5189/0 653 2580 P25-F022A 1 27600 6.7 A.8 TCV; DG A RM CLG 107E5182/0 653 2581 H21-P009-02 N 27600 6.5 D.0 REMOTE COMM CABNET 103E1167 654 (C11) 2582 H21-P009-04 N 77500 6.5 D.0 REMOTE COMM CABNET 103E1167 654 (C11) 2583 H21-P009-06 N 27600 6.5 0.0 REMOTE COMM CABNET 103E1167 654 (C11) 2584 H21-P009-08 N 27600 6.5 D.0 REMOTE COMM CABNET 103E1167 654 (C11) 2585 H21-P009-10 N 27600 6.5 0.0 REMOTE COMM CABNET 103E1167 654 (C11) l 2586 H21-P00912 N 27600 6.5 0.0 REMOTE COMM CABNET 103E1167 654 i [) (C11) l 2587 Hi .P00914 N 27600 6.5 D.0 REMOTE COMM CABNET 103E1167 654 (C11) Fire Hazard Analysis Database - Amendment 37 9A.6-8S
23A6100 Rev. 9 i ABWR StandardSafety Analysis Report O Table 9A.6-2 Fire Hazard Analysis i Equipment Database Sorted by Room - Reactor Building (Continued) Location Location item Elect Elev. Number Alpha System Room j No. MPL No Div. Location Coord. Coord. Description Drawing No. 2588 H21-P009-16 N 27600 6.5 D.0 REMOTE COMM CABNET 103E1167 654 (C11) i 2589 H21-P00918 N 27600 6.5 D.0 REMOTE COMM CABNET 103E1167 654 l (C11) l 2590 H21-P009-20 N 27600 6.5 0.0 REMOTE COMM CABNET 103E1167 654 l (C11) 2591 H21-P009-22 N 27600 6.5 0.0 REMOTE COMM CABNET 103E1167 654 l (C11) 2592 H21-P009-24 N 27600 6.5 0.0 REMOTE COMM CABNET 103E1167 654 l (C11) l 2593 H21-P009-26 N 27600 6.5 D.0 REMOTE COMM CABNET 103E1167 654 (C11) 2594 H21 P010-02 N1 27600 6.5 0.0 BREAK CTRL CABNET 103E1167 654 (C11) 2595 H21 P010-04 N2 27600 6.5 D.0 BREAK CTRL CABNET 103E1167 654 (C11) 2596 H21-P01006 N3 27600 6.5 D.0 BREAK CTRL CABNET 103E1167 654 (C11) 2597 H21 P011-02 N1 27600 6.5 D.0 FINE MOTION DR CAB 103E1167 654 (C11) 2598 H21-P011-04 N1 27600 6.5 D.0 FINE MOTION DR CAB 103E1167 654 l (C11) 2599 H21-P011-06 N2 27600 6.5 D.0 FINE MOTION DR CAB 103E1167 654 (C11) l 2600 H21-P011-08 N2 27600 6.5 D.0 FINE MOTION DR CAB 103E1167 654 i (C11) 2601 H21-P011-10 N3 27600 6.5 0.0 FINE MOTION DR CAB 103E1167 654 (C11) 2602 H21-P011 12 N3 27600 6.5 D.0 FINE MOTION DR CAB 103E1167 654 (C11) l 2603 H21-P011-14 N1 27600 6.5 D.0 FINE MOTION DR CAB 103E1167 654 (C11) 2604 H21-P011 16 N1 27600 6.5 D.0 FINE MOTION DR CAB 103E1167 654 (C11) 2605 H21-P011-18 N2 27600 6.5 D.0 FINE MOTION OR CAB 103E1167 654 (C11) 2606 H21.P011-20 N2 27600 6.5 D.0 FINE MOTION DR CAB 103E1167 654 (C11) 2607 H21-P01122 N3 27600 6.5 D.0 FINE MOTION DR CAB 103E1167 654 (C11) 2608 H21-P01124 N3 27600 6.5 D.0 FINE MOTION DR CAB 103E1167 654 (C11) 9A.6-86 Fire Hazard Analysis Database Amendment 37
23A6100 Rev. 9 ABWR standardsatery Analysis aeport 1 s Table 9A.6 2 Fire Hazard Analysis Equipment Database Sorted by Room - Reactor Building (Continued) l l Location Locat item Elect Elev. Number Alpha System Room l No. MPL No Div. Location Coord. Coord. Description Drawing No. 2609 H21 P011-26 N1 27600 6.5 D.0 FINE MOTION DR CAB 103E1167 654 (C11) 2610 H21-P01128 N1 27600 6.5 D.0 FINE MOTION OR CAB 103E1167 654 (C11) 2611 H21-P01130 N2 27600 6.5 D.0 FINE MOTION OR CAB 103E1167 654 (C11) 2612 H21-P011-32 N2 27600 6.5 D.0 FINE MOTION DR CAB 103E1167 654 (C11) i 2613 H21-P011-34 N3 27600 6.5 0.0 FINE MOTION DR CAB 103E1167 654 (C11) 2614 H21-P011-36 N3 27600 6.5 D.0 FINE MOTION DR CAB 103E1167 654 (C11) 2615 H21 P01138 N1 27600 6.5 0.0 FINE MOTION DR CAB 103E1157 654 (C11) 2616 H21-P01140 N1 27600 6.5 D.0 FINE MOTION DR CAB 103E1167 654
\ (C11) y 2617 H21-P01142 N2 27600 6.5 D.0 FINE MOTION DR CAB 103E1167 654 (C11) 2618 H21-P01144 N2 27600 6.5 D.0 FINE MOTION DR CAB 103E1167 654 (C11) 2619 H21-P01146 N3 27600 6.5 D.0 FINE MOTION DR CAB 103E1167 654 (C11) 2620 H21 P011-48 N3 27600 6.5 0.0 FINE MOTION DR CAB 103E1167 654 (C11) 2621 H21 P01150 N1 27600 6.5 D.0 FINE MOTION DR CAB 103E1167 654 (C11) 2622 H21-P011-52 N2 27600 6.5 D.0 FINE MOTION DR CAB 103E1167 654 (C11) 2623 H21-P011-54 N3 27600 6.5 D.0 FINE MOTION DR CAB 103E1167 654 (C11) 2624 H21-P012-02 N1 27600 6.5 0.0 FMCRD DISTR CAB (C11) 103E1167 654 2625 H21-P012-04 N1 27600 6.5 D.0 FMCRD DISTR CAB (C11) 103E1167 654 2626 H21 P012-06 N2 27600 6.5 D.0 FMCRD DISTR CAB (C11) 103E1167 654 2627 H21-P012-08 N2 27600 6.5 D.0 FMCRD DISTR CAB (C11) 103E1167 654 2628 H21 P012-10 N3 27600 6.5 D.0 FMCRD DISTR CAB (C11) 103E1167 654 2629 H21 P01212 N3 27600 6.5 D.0 FMCRD DISTR CAB (C11) 103E1167 654 2630 H21 P013-02' N 27600 6.5 0.0 FMCRD SCRAM TIME CAB 103E1167 654 2631 H21-P013-04' N 27600 FMCRO SCRAM TIME CAB 103E1167 f 6.5 D.0 654 2632 H21-P013-06' N 27600 6.5 D.0 FMCRD SCRAM TIME CAB 103E1167 654 Fire Hazard Analysis Database - Amendment 37 9A.6-87
23A6100 Rsv. 9 ABWR StandardSafety Analysis Report l i O l Table 9A.6-2 Fire Hazard Analysis l Equipment Database Sorted by Room - Reactor Building (Continued) Location Location item Elect Elev. Number Alpha Systeri Room ) No. MPL No Div. Location Coord. Coord. Description Drawing No. I 2633 R23 P/C EA10B N 27600 6.5 D.0 P/C EA10 - LO VOLT 107E5072/0 654 SWTGR , 2634 H22-P010 N 27200 5.5 8.9 E31, AIR PARTl SMPL PNL 107E5015/A 657 2635 H22 P011 N 27200 5.5 B.9 E31,10 DINE / NOBEL GAS 107E5015/A 657 PNL l 2636 023-F195A 1 26900 5.3 D.9 SO VALVE 107E5139/1 659 2637 D23-C001A 1 26900 5.3 D.9 ACCIDENT SMPL PUMP 107E5139/1 659 2638 D23-02E003A 1 26900 5.3 D.9 OXYGEN ANAL ELEM 107E5139/1 659 2639 D23-F105A 1 26900 5.3 D.9 SO VALVE 107E5139/1 659 2640 D23-TS016A 1 26900 5.3 D.9 TEMP SWITCH 107E5139/1 659 2641 D23 02AM003A 1 26900 5.3 D.9 OXYGEN ANALYZER 107E5139/1 659 2642 D23-TE020A 1 26900 5.3 D.9 TEMP ELEMENT 107E5139/1 659 i 2643 D23 F108A 1 26900 5.3 D.9 SO VALVE 107E5139/1 659 l 2644 D23-SCO33A 1 26900 5.3 D.9 STEAM CONDENSER 107E5139/1 659 I 2645 D21 LIT 031A 1 26900 5.3 D.9 LEVEL IND TRANS 107E5139/1 659 2646 D23-F118A 1 26900 5.3 0.9 SO VALVE 107E5139/1 659 2647 D23-H2E001A 1 26900 5.3 D.9 HYDROGEN ANAL ELEM 107E5139/1 659 2648 D23-F121A 1 26900 5.3 D.9 SO VALVE 107E5139/1 659 2649 D23.PS027A 1 26900 5.3 D.9 PRESS SWITCH 107E5139/1 659 2650 D23-H2AM001A 1 26900 5.3 D.9 HYDROGEN ANALYZER 107E5139/1 659 l 2651 D23-PS026A 1 26900 5.3 0.9 PRESS SWITCH 107E5139/1 659 2652 D23-F123A 1 26900 5.3 D.9 SO VALVE 107E5139/1 659 1 2653 023-PS024A 1 26900 5.3 D.9 PRESS SWITCH 107E5139/1 659 l 2654 D23-FIT 019A 1 26900 5.3 D.9 FLOW IND TRANSMITTER 107E5139/1 659 2655 D23.P1T021 A 1 26900 5.3 D.9 PRESS IND TRANSMITTER 107E5139/1 659 l 2656 D23-F127A 1 26900 5.3 D.9 SO VALVE 107E5139/1 659 2657 D23-PIS017A 1 26900 5.3 D.9 PRESS IND SWITCH 107E5139/1 659 2658 D23-F515A 1 26900 5.3 D.9 AO VALVE 107E5139/1 659 2659 D23-C002A 1 27200 5.3 D.9 NORM SMPL. BOOSTER 107E5139/1 659 PUMP 2660 C23-F130A 1 26900 5.3 D.9 SO VALVE 107E5139/1 659 2661 D23-C004A 1 26900 5.3 D.9 NORM. SMPL PUMP 107E5139/1 659 2662 023-FS13A 1 26900 5.3 D.9 AO VALVE 107E5139/1 659 l 2663 D23-D012A 1 26900 5.3 D.9 DEHUMIDIFIER 107E5139/1 659 2664 D23-F132A 1 26900 5.3 D.9 SO VALVE 107E5139/1 659 9A.6-88 Fire Hazard Analysis Database - Amendment 37
l 23A6100 Rsv. 9 I ABWR standedsatory Analysis neport O Table 9A.6-2 Fire Hazard Analysis Equipment Database Sorted by Room - Reactor Building (Continued) Location Location item Elect Elev. Number Alpha System Room No. MPL No Div. Location Coord. Coord. Description Drawing No. 2665 D23-F510A 1 26900 5.3 D.9 AO VALVE 107E5139/1 659 2666 D23-F190A 1 26900 5.3 D.9 SO VALVE 107E5139/1 659 2667 D23-F202A 1 26900 5.3 D.9 AO VALVE 107E5139/1 659 2668 0234003A 1 26900 5.3 0.9 ACC. SMPLBOOSTER 107E5139/1 659 PUMP 2669 D23-F191A 1 26900 5.3 D.9 SO VALVE 107E5139/1 659 2670 D23-0022A 1 26900 5.3 D.9 DRAIN MEAS VESSEL 107E5139/1 659 2671 D23-F201A 1 26900 5.3 D.9 AO VALVE 107E5139/1 659 2672 D23-F193A 1 26900 5.3 D.9 SO VALVE 107E5139/1 659 2673 D23-0010A 1 26900 5.3 0.9 STEAM SEPARATOR 107E5139/1 659 2E74 D23-F197A 1 26900 5.3 D.9 SO VALVE 107E5139/1 659 2675 H22.P053A* 1 26900 5.3 D.9 023, CAMS RACK A 107E5139/1 659 2676 H22-P054A* 1 26900 5.2 D.7 923, CAMS CAUB RACK A 107E5139/1 659 2677 U41-D113 1 27200 5.3 0.7 CA.'3S (A) ROOM HVH 107E5189/0 659 2678 U41-C2048 2 27600 1.4 E.8 DG(B)MVAC SUPP FAN 8 107E5189/0 663 2679 U41-C204F 2 27600 1.2 E.8 DG(B) HVAt SUPP FAN F 107E5189/0 663 2680 U41 TE056 2 27600 1.4 E.8 TEMP ELEMENT 107E5189/0 663 2681 P25-F0228 2 27600 1.2 F.3 TCV; DG B RM CLG 107E5182/0 663 2682 U41 C207C 3 27600 6.8 E.8 DG(C) HVAC SUPP FAN C 107E5189/0 673 2683 U41-C207G 3 27600 6.5 E.8 DG{C) HVAC SUPP FAN G 107E5189/0 673 2684 U41-TE060 3 27600 6.7 E.8 TEMP ELEMENT 107E5189/0 673 2685 P25-F022C 3 27600 6.7 F.3 TCV: DG C RM CLG 107E5182/0 673 l 2686 DELETED l 2687 DELETED l 2688 DELETED l 2689 DELETED l 2690 DELETED 2691 H21 P009-01 N 27600 1.5 D.0 REMOTE COMM CABNET 103E1167 681 (C11) 2692 H21 P009-03 N 27600 1.5 D.0 REMOTE COMM CABNET 103E1167 681 (C11) 2693 H21 P009-05 N 27600 1.5 D.0 REMOTE COMM CABNET 103E1167 681 (C11) j 2694 H21-P009-07 N 27600 1.5 D.0 REMOTE COMM CABNET 103E1167 681 j (011) Fire Hazard Analysis Database - Amendment 37 9A.6-89
23A6100 Rev. C ABWR StandardSafety Analysis Report O Table 9A.6-2 Fire Hazard Analysis Equipment Database Sorted by Room - Reactor Building (Continued) Location Location item Elect Elev. Number Alpha System Room No. MPL No Div. Location Coord. Coord. Description Drawing
)
No. ' 2695 H21.P009 09 N 27600 1.5 D.0 REMOTE COMM CABNET 103E1167 681 (C11) 2696 H21 P009-11 N 27600 1.5 D.0 REMOTE COMM CABNET 103E1167 681 l (C11) 2697 H21 P00913 N 27600 1.5 D.0 REMOTE COMM CABNET 103E1167 681 (C11) ' 2698 H21 P009-15 N 27600 1.5 D.0 REMOTE COMM CABNET 103E1167 681 (C11) 2699 H21-P009-17 N 27600 1.5 D.0 REMOTE COMM CABNET 103E1167 681 (C11) 2700 H21 P00919 N 27600 1.5 0.0 REMOTE COMM CABNET 103E1167 681 (C11) j 2701 H21 P009-21 N 27600 1.5 D.0 REMOTE COMM CABNET 103E1167 681 (C11) ! 2702 H21 P009-23 N 27600 1.5 D.0 REMOTE COMM CABNET 103E1167 681 (C11) 2703 H21 P009 25 N 27600 1.5 D.0 REMOTE COMM CABNET 103E1167 681 (C11) 2704 H21-P010-01 N1 27600 1.5 D.0 BREAK CTRL CABNET 103E1167 681 (C11) 2705 H21-P010-03 N2 27600 1.5 D.0 BREAK CTRL CABNET 103E1167 681 (C11) 2706 H21-P010-05 N3 27600 1.5 D.0 BREAK CTRL CABNET 103E1167 681 (C11) 2707 H21 P011-01 N1 27600 1.5 D.0 FINE MOTION OR CAB 103E1167 681 (C11) 2708 H21-P011-03 N1 27600 1.5 D.0 FINE MOTION DR CAB 103E1167 681 (C11) 2709 H21-P011-05 N2 27600 1.5 D.0 FINE MOTION DR CAB 103E1167 681 (C11) 2710 H21-P011-07 N2 27600 1.5 D.0 FINE MOTION DR CAB 103E1167 681 (C11) 2711 H21-P011-09 N3 27600 1.5 D.0 FINE MOTION DR CAB 103E1167 681 (C11) 2712 H21-PC11 11 N3 27600 1.5 0.0 FINE MOTION OR CAB 103E1167 681 (C11) 2713 H21-P011 13 N1 27600 1.5 D.0 FINE MOTION DR CAB 103E1167 681 (C11) 2714 H21 P01115 N1 27600 1.5 D.0 FINE MOTION DR CAB 103E1187 681 (C11) 2715 H21-P011 17 N2 27600 1.5 0.0 FINE MOTION DR CAB 103E1167 681 (C11) 9A.6 90 Fire Hazard Analysis Database - Amendment 37
l { l 23A6100 R1v. 9 l ABWR standardsarery Analysis Report v Table 9A.6-2 Fire Hazard Analysis Equipment Database Sorted by Room - Reactor Building (Continued) Location Location item Elect Elev. Number Alpha System Room No. MPL No Div. Location Coord. Coord. Description Drawing No. 2716 H21 P011-19 N2 27600 1.5 0.0 FINE MOTION DR CAB 103E1167 681 (C11) 2717 H21-P01121 N3 27600 1.5 D.0 FINE MOTION DR CAB 103E1167 681 (C11) 2718 H21-P01123 N3 27600 1.5 D.0 FINE MOTION DR CAB 103E1167 681 (C11) 2719 H21-P01125 N1 27600 1.5 D.0 FINE MOTION DR CAB 103E1167 681 (C11) 2720 H21-P01127 N1 27600 1.5 D.0 FINE MOTION DR CAB 103E1167 681 (C11) l 2721 H21-P01129 N2 27600 1.5 D.0 FINE MOTION DR CAB 103E1167 681 l (C11) 2722 H21-P01131 N2 27600 1.5 0.0 FINE MOTION DR CAB 103E1167 681 (C11) 2723 H21-P011-33 N3 27600 1.5 D.0 FINE MOTION DR CAB 103E1167 681 j
\ (C11) 's 2724 H21-P01135 N3 27600 1.5 D.0 FINE MOTION DR CAB 103E1167 681 (C11) 2725 H21-P011-37 N1 27600 1.5 D.0 FINE MOTION DR CAB 103E1167 681 l (C11) 2726 H21-P011-39 N1 27600 1.5 D.0 FINE MOTION DR CAB 103E1167 681 l (C11) 2727 H21-P011-41 N2 27600 15 D.0 FINE MOTION DR CAB 103E1167 681 (C11) 2728 H21-P011-43 N2 27600 1.5 0.0 FINE MOTION DR CAB 103E1167 681 (C11) l 2729 H21 P011-45 N3 27600 1.5 D.0 FINE MOTION DR CAB 103E1167 681 (C11) 2730 H21-P011-47 N3 27600 1.5 0.0 FINE MOTION DR CAB 103E1167 681 (C11) 2731 H21 P011-49 N1 27600 1.5 0.0 FINE MOTION DR CAB 103E1167 681 (C11) 2732 H21-P011-51 N2 27600 1.5 D.0 FINE MOTION OR CAB 103E1167 681 l (C11) 2733 H21 P011-53 N3 27600 1.5 D.0 FINE MOTION DR CAB 103E1167 681 (C11) l 2734 H21 P012-01 N1 27600 1.5 D.0 FMCRD DISTR CAB (C11) 103E1167 681 2735 H21-P012-03 N1 27600 1.5 D.0 FMCRD DISTR CAB (C11) 103E1167 681 2736 H21 P012-05 N2 27600 1.5 0.0 FMCRD DISTR CAB (C11) 103E1167 681 \
b 2737 H21 P012-07 N2 27600 1.5 D.0 FMCRO DISTR CAB (C11) 103E1167 681 2738 H21 P012-09 N3 27600 1.5 0.0 FMCRO DISTR CAB (C11) 103E1167 681 Fire Hazard Analysis Database - Amendment 37 9A691 l l
23A6100 Rsv. 9 ABWR StandardSafety Analysis Report O Table 9A.6-2 Fire Hazard Analysis Equipment Database Sorted by . Room - Reactor Building (Continued) Location Location l i item Elect Elev. Number Alpha System Room No. MPL No Div. Location Coord. Coord. Description Drawing No. l
- 2739 H21-P01211 N3 27600 1.5 D.0 FMCRO DISTR CAB (C11) 103E1167 681 2740 H21-P01341' N 27600 1.5 D.0 FMCRD SCRAM TIME CAB 103E1167 681 l 2741 H21-P013-03' N 27600 1.5 D.0 FMCRD SCRAM TIME CAB 103E1167 681 2742 H21 P013-05' N 27600 1.5 D.0 FMCRD SCRAM TIME CAB 103E1167 681 2743 R23 P/C EA10A N 27600 1.5 D.0 P/C EA10 LO VOLT 107E5072/0 681 l
SWTGR j 2744 R23 P/C EA10C N 27600 1.5 D.0 P/C EA10 - LO VOLT 107E5072/0 681 l SWTGR 2745 U41-D121 A 1 N 26000 4.7 A.1 R/A MS TUNNEL HVH A 107E5189/0 685 2746 U41 D121A 2 N 26000 4.T A.1 R/A MS TUNNEL HVH A 107E5189/0 685 2747 U41-01218-1 N 26000 4.4 A.1 R/A MS TUNNEL HVH B 107E5189/0 685 2748 U41-D1218-2 N 26000 4.4 A.1 R/A MS TUNNEL HVH B 107E5189/0 685 2749 E31 TE035A N 26000 4.7 A.5 MSL TUN DIFF TEMP 103E1792/1 690 ELEM 2750 E31-TE036A N 26000 4.7 A.5 MSL TUN DIFF TEMP 103E1792/1 690 l ELEM 2751 T31 TE051A N 23500 4.5 C.8 TEMP ELEMENT 107E6043/0 691 2752 T31 TE051B N 23500 3.5 D.2 TEMP ELEMENT 107E6043/0 691 2753 T31 TE051C N 23500 4.5 D.2 TEMP ELEMENT 107E6043/0 691 2754 T31 TE051J N 27200 4.1 D.1 TEMP ELEMENT 107E6043/0 691 2755 T31 TE051K N 27200 3.9 C.9 TEMP ELEMENT 107E6043/0 691 2756 E31-TE004 N 26000 4.0 E.c DW AMB TEMP ELEMENT 103E1792/1 693 2757 P11-FOT106A N 31700 62 A.7 FLOW XMTR (MUWP) 107E111/1 710 < 2758 P21 A001A 1 31700 6.8 A6 RCW SURGE TANK (A) 107E5112/0 710 2759 P21-F018A 1 31700 6.8 A.6 MO GLOBE VALVE 107E5112/0 710 2760 P21 F019A N 31700 6.8 A.7 AO GLOBE VALVE 107E5112/0 710 2761 P21-LT013A 1 31700 6.8 A.7 LVL XMTR (SURGE TK A) 107E5112/0 710 2762 P21-LT014A 1 31700 6.8 A.7 LVL XMTR (SURGE TK A) 107E5112/0 710 2763 P21 LT0140 1 31700 6.8 A.7 LVL XMTR (SURGE TK A) 107E5112/0 710 2764 P21-LT014G 1 31700 6.8 A.7 LVL XMTR (SURGE TK A) 107E5112/0 710 2765 U41-F001A 1 31700 6.1 A.4 AO VLV - R/A SUP ISO VLV 107E5189/0 710 2766 U41-F0018 2 31700 6.3 A.4 AO VLV - R/A SUP ISO VLV 107E5189/0 710 2767 U41-B301A N 31700 6.6 B.1 COOLING COIL. RIP A 107E5189/0 715 2768 U41 B3018 N 31700 6.6 B.3 COOLING CCIL. RIP A 107E5189/0 715 2769 U41-C301 A N 31700 6.6 B.1 RIP ZONE (A) SUPP FAN A 107E5189/0 715 2770 U41-C3018 N 31700 6.6 B.3 RIP ZONE (A) SUPP FAN B 107E5189/0 715 9A6-92 Fire Hazard Analysis Database Amendment 37
23A6100 Ru. 9 ABWR standardsafetyAnalysis Report [ ( Table SA.6-2 Fire Hazard Analysis Equipment Database Sorted by Room - Reactor Building (Continued) ! Location Location item Elect Elev. Number Alpha System Room No. MPL No Div. Location Coord. Coord- Description Drawing No. 2771 U41 TE071 A N 31700 6.6 B.2 TEMP ELEMENT 107E5189/0 715 2772 U41 TE071B N 31700 6.6 B.2 TEMP ELEMENT 107E5189/0 715 2773 U41 TE071C N 31700 6.6 8.2 TEMP ELEMENT 107E5189/0 715 2774 H23-P032' N 31700 6.2 E.4 MULTIPLEXER -?- 715 2775 H23-P033' N 31700 6,2 E.5 MULTIPLEXER 715 2776 H23-P034' N 31700 6.2 E.7 MULTIPLEXER 715 2777 P11 FQT106C N 31700 6.2 E.3 FLOW XMTR (MUWP) 107E111/1 715 2778 P21-A001C 3 31700 6.7 E.5 RCW SURGE TANK (C) 107E5112/0 715 2779 P21-F018C 3 31700 6.8 E.5 MO GLOBE VALVE 107E5112/0 715 2780 P21.F019C N 31700 6.8 E.5 AO GLOBE VALVE 107E5112/0 715 2781 P21 LT013C 3 31700 6.8 E5 LVL XMTR (SURGE TK C) 107E5112/0 715 2782 P21 LT014C 3 31700 6.8 E.5 LVL XMTR (SUNGE TK C) 107E5112/0 715 2783 P21-LT014F 3 31700 6.8 E.5 LVL XMTR (SURGE TK C) 107E5112/0 715 ~ 2784 P21-LT014J 3 31700 6.8 E.5 LVL XMTR (SURGE TK C) 107E5112/0 715 2785 R24 MCC SA110 N 31700 6.3 D.3 MCC SA110 - R/B 107E5072/0 715 2786 R24 MCC SA111 N 31700 6.3 D.3 MCC SA111 R/B 107E5072/0 715 2787 011-RE022A 1 31700 5.1 C.9 FUEL HANDLING AREA 107E6071/0 716 EXH. 2788 D21-RE002 N 31700 5.2 C.8 AREA RAD DETECTOR 299X701 171/0 716 2789 U41-F003C 3 33000 5.2 A.5 MO VALVE 107E5189/0 716 2790 021-RE006 N 31700 2.7 F.4 AREA RAD DETECTOR 299X701 171/0 721 2791 011 RE0228 2 31700 2.8 E.8 FUEL HANDLING AREA 107E6071/0 721 EXH. 2792 U410123-1 N 31700 1.7 E.8 REFUEL MACH CR HVH 107E5189/0 722 2793 U41-D123-2 N 31*00 1.7 E.8 REFUEL MACH CR HVH 107E5189/0 722 2794 U41 TE015 N 31700 1.7 E.8 TEMP ELEMENT 107E5189/0 722 2795 U41-D131C N 31700 1.5 E.1 RIP /FMCRD CP RM FCU C 107E5189/0 723 2796 U41 C208C 3 31700 6.6 F.5 DG(C) HVAC EXH FAN C 107E5189/0 730 2797 U41 C208G 3 31700 6.6 F.3 DG(C) HVAC EXH FAN G 107E5189/0 730 2798 U41-F006C 3 31700 6.6 F.5 MO VALVE 107E5189/0 730 2799 D11-RE022C 3 31700 5.1 E.8 FUEL HANDLING AREA 107E6071/0 733 EXH. f 2800 021-RE003 N 31700 5.1 F.0 AREA RAD DETECTOR 299X701-171/0 733
, 2801 D21-RE004 N 31700 5.1 F.0 AREA RAD DETECTOR 299X701-171/0 733 2802 G41 LS001 N 31700 5.1 F.5 LEVEL SWITCH 107E6042/0 733 Fire Hazard Analysis Database -knendment 37 9A.6-93
23A6100 Rw. O ABWR Standard Safety Analysis Report O' Table 9A.6-2 Fire Hazard Analysis i Equipment Database Sorted by Roorn - Reactor Building (Continued) l Location Location l ttom Elect Elev. Number Alpha System Room i No. MPL No Div. Location Coord. Coord. Description Drawing N o. 2803 G41 TE015 N 31700 5.1 F.5 TEMP ELEMENT 107E6042/0 733 2804 P11-FQT1068 N 31700 1.7 A.7 FLOW XMTR (MUWP) 107E111/1 740 2805 P21.A0018 2 31700 1.2 A.6 RCW SURGE TANK (B) 107E5112/0 740 2806 P21-F0188 2 31700 1.2 A.6 MO GLOBE VALVE 107E5112/0 740 l 2807 P21 F019B N 31700 1.2 A.6 AO GLOBE VALVE 107E5112/0 740 2808 P21 LT0138 2 31700 1.3 A.6 LVL XMTR (SURGE TK B) 107E5112/0 740 l 2809 P21-LT014B 2 31700 1.3 A.6 LVL XMTR (SURGE TK B) 107E5112/0 740 2810 P21 LT014E 2 31700 1.3 A.6 LVL XMTR (SURGE TK B) 107E5112/0 740 2811 P21 LT014H 2 31700 1.3 A.6 LVL XMTR (SURGE TK B) 107E5112/0 740 2812 U41-B302A N 31700 1.4 B.1 COOLING COIL RIP B 107E5189/0 740 l 2813 U41-8302B N 31700 1.4 B.3 COOLING COIL, RIP B 107E5189/0 740 l l 2814 U41-C302A N 31700 1.4 B.1 RIP ZONE (B) SUPP FAN A 107E5189/0 740 2815 U41-C3028 N 31700 1.4 B.3 RIP ZONE (B) SUPP FAN 8 107E5189/0 740 2816 U41-TE072A N 31700 1.4 B.2 TEMP ELEMENT 107E5189/0 740 2817 U41-TE072B N 31700 1.4 B.2 TEMP ELEMENT 107E5189/0 740 2818 U41 TE072C N 31700 1.4 B.2 TEMP ELEMENT 107E5189/0 740 2819 D11-D304 N 31700 1.1 B.8 FILTER DEVICE 107E6071/0 740 2820 011-D305 N 31700 1.1 B.8 FILTER DEVICE 107E6071/0 740 2821 D11-RE041 A N 31700 1.1 B.8 STACK RAD MON SCIN 107E6071/0 740 DET. 2822 D11 RE041B N 31700 1.1 B.8 STACK RAD MON SCIN 107E6071/0 740 DET. 2823 D11 RE042 N 31700 1.1 B.8 STACK RAD MON Ge DET. 107E6071/0 740 2824 D11-RE043A N 31700 1.1 B.8 STACK RAD ION 107E6071/0 740 CHAMBER 2825 D11-RE043B N 31700 1.1 B.8 STACK RAD lON 107E6071/0 740 CHAMBER 2826 D11 RSM041 A N 31700 1.1 B.8 STACK GAS SAMPLER 107E6071/0 740 2827 011-RSM0418 N 31700 1.1 B.8 STACK GAS SAMPLER 107E6071/0 740 2828 D11 RSM042 N 31700 1.1 B.8 STACK GAS SAMPLER 107E6071/0 740 2829 D11 SV301 N 31700 1.1 B.8 SOLENOID VALVE 107E6071/0 740 2830 011 SV302 N 31700 1.1 B.8 SOLENOID VALVE 107E6071/0 740 2831 H21 P301 N 31700 1.1 B.8 D11, STACK RAD SIG 107E6071/0 740 CONV. 2832 H21.P331 N 31700 1.1 B.8 D11, CONTROL SAMPL 107E6071/0 740 PNL 9A6-94 Fire Hazard Analysis Database - Amendment 37
23A6100 Rev. 9 ABWR standardsafety Analysis Report p b Table 9A.6-2 Fire Hazard Analysis Equipment Database Sorted by Room - Reactor Building (Continued) Location Location item Elect Elev. Number Alpha System Room No. MPL No Div. Location Coord. Coord. Description Drawing No. 2833 H22-P251 N 31700 1.1 B.8 D11, STACK RAD SMPL 107ES071/0 740 RACK 2834 H.22.P252 N 31700 1.1 B.8 D11, STACK RAD SMPL 107E6071/0 740 RACK 2835 011 RE022D 4 31700 2.8 0.0 FUEL HANDLING AREA 107E6071/0 742 EXH. l 2836 D21-RE001 N 31700 2.2 D.0 AREA RAD DETECTOR 299X701 171/0 742 2837 D21 RE005 N 31700 2.2 D.0 AREA RAD DETECTOR 299X701 171/0 742 2838 G41-F038 N 32700 3.3 D.6 MO GATE VALVE 107E6042/0 743 ; 2839 G41-LT020A N 31700 4.7 D.8 LEVEL TRANSMITTER 107E6042/0 743 2840 G41.LT020B N 31700 3.3 0.8 LEVEL TRANSMITTER 107E6042/0 743 2841 P11-FQT104 N 31700 3.2 C.5 FLOW XMTR (MUWP) 107E111/1 743 2842 011 0041 N 38200 1.5 C.5 SAMPLING PROBE 107E6071/0 840 s 2843 D11-RE003A 1 38200 1.5 C.5 REA BLOG EX DETECTOR 107E6071/0 840 2844 D11 RE003B 2 38200 1.5 C.5 REA BLOG EX DETECTOR 107E6071/0 840 2845 011-RE003C 3 38200 1.5 C.5 REA BLOG EX DETECTOR 107E6071/0 840 2846 D11-RE0030 4 38200 1.5 C.5 REA BLDG EX DETECTOR 107E6071/0 840 2847 U41-FT008 N 38200 1.5 C.5 FLOW TRANS 107E5189/0 840 O Fire Hazard Analysis Database Amendment 37 9A.6-95
1 23A6100 R1v. 2 ABWR standardsafety Analysis Report O Table 9A.6-3 Fire Hazard Analysis Equipment Data Base - Sorted by Room - Control Building LOCATION LOCATION ITEM MPL ELECT ELEV. NUMBER ALPHA SYSTEM ROOM NO. NO. DIV. LOCATION COORD. COORD. DESCRIPTION DRAWING NO. 1 P21-C001A 1 -8200 3.70 K.6 RCW PUMP A 107E5112/0 111 2 P21-C0010 1 -8200 4.20 K.6 RCW PUMP D 107E5112/0 111 3 P21-F072A 1 -8200 3.80 K.2 AO BUTTERFLY VALVE 107E5112/0 111 4 P21-F0720 1 -8200 3.80 K.2 AO BUTTERFLY VALVE 10)E5112/0 111 l 5 P21-F074A 1 -8200 3.80 K.2 MO GATE VA'.VE 107E5112/0 111 l 6 P21-F082A 1 -8200 3.80 K.2 MO GATE VALVE 107E5112/0 til j 7 P21-F084A 1 -8200 3.80 K.2 MAN OPER GATE VALVE 107E5112/0 111 l 8 P21-F171A N -8200 3.80 K.2 MAN OPER GATE VALVE 107E5112/0 111 9 P21-F172A N -8200 3.00 K.2 MAN OPER GATE VALVE 107E5112/0 111 10 P21-FT042A 1 -8200 3.80 K.2 FLOW XMTR (C/B SUPPLY) 107E5112/0 111 11 P21 A002A 1 -8200 4.70 J.5 RCW CHEM ADDTANK A 107E5112/0 111 12 P21-F006A 1 -8200 4.80 K.8 AO BUTTERFLY VALVE 107E5112/0 111 13 P21-F010A 1 -8200 4.80 K.8 AO BUTTERFLY VALVE 107E5112/0 111 14 P21-FT006A 1 -8200 4.80 K.8 FLOW XMTR (RCW SUPPLY) 107E5112/0 111 15 P21-PT004A 1 -8200 4.80 K.8 PRESS XMTR (RCW 107E5112/0 111 SUPPLY) 16 P21 TE005A 1 -8200 4.80 K.8 TEMP ELEM (RCW SUPPLY) 107E5112/0 111 i 17 P218001A 1 -8200 4.70 J.9 RCW/RSW HX A 107E5112/0 - 111 18 P21-8001D 1 -8200 4.70 K.2 RCW/RSW HX 0 107E5112/0 til 19 P21-8001G 1 -8200 4.70 K.5 RCW/RSW HX G 107E5112/0 111 20 P21 E/P105A 1 -8200 4.50 K.5 E/P CONVERT (TCV-RCW) 107E5112M 111 21 P21 F004A 1 -8200 4.60 J.9 MO GATE VALVE 107E5112M 111 22 P21-F004D 1 -8200 4.60 K.2 MO GATE VALVE 107E5112/0 111 23 P21-F004G 1 -8200 4.60 K.5 MO GATE VALVE 107E5112/0 111 [ 24 K17 C001AC 1 -8200 4.80 K.8 C/B RSW/RCW RM-SUMP A 107E5112/0 111 l 25 K17 LS401A 1 -8200 4.80 K.8 LEVEL SWITCH 107E5112/0 111 l 26 K17-LS401E 1 -8200 4.80 K.8 LEVEL SWITCH 107E5112/0 111 l 27 K17-LS4011 1 -8200 4.80 K.8 LEVEL SWITCH 107E5112/0 111 l 28 K17-LS401M 1 -8200 4.80 K.8 LEVEL SWITCH 107E5112/0 111 l 29 K17-LS402A 1 -8200 4.80 K.8 LEVEL SWITCH 107E5112/0 111 l 30 K17-LS402E 1 -8200 4.80 K.8 LEVEL SWITCH 107E5112/0 111 l 31 K17-LS4021 1 -8200 4.80 K.8 LEVEL SWITCH 107E5112/0 111 l 32 K17-LS402M 1 -8200 4.80 K.8 LEVEL SWITCH 107E5112/0 111 l 33 K17-LS403A 1 8200 4.80 K.8 LEVEL SWITCH 107E5112/0 111 9A 6-96 Fire Hazard Analysis Database - Amendment 32
23A8100 Rw. 2 ABWR standard sarety Analysis neport G Table 9A.6-3 Fire Hazard Analysis Equipment Data Base - Sorted by Room - Control Building (Continued) LOCATION LOCATION ITEM MPL ELECT ELEV. NUMBER ALPHA SYSTEM ROOM NO. NO. DIV. LOCATION COORD. COORD. DESCRIPTION DRAWING NO. l 166 H12-P003A' 1 7900 5.50 K.0 OlV 1 CONTROL PNLS -?- 497 l 167 H12.P004A N 7900 5.80 K.0 NON DIV CONT.PNLS 497 l 168 H12 P003C* 3 7900 6.30 J.5 DIV 3 CONTROL PNLS ~7- 497 l 169 C81-D001A' N 12300 2.50 J.3 RIP MG SET A 107E5072 501 l l 170 U41-D132A N 12300 1.80 J.4 MG SET ROOM FCU A 107E5189/0 501 l 171 C81-P001A N 12300 2.50 J.7 RIP MG A CONTROL PNL 502 l 172 C81-D001B+ N 12300 2.50 K.1 RIP MG SET B 107E5072 503 l 173 U41-0132B N 12300 1.80 K.4 MG SET ROOM FCU B 107E5189/0 503 l 174 C81-P001B N 12300 2.50 K.4 RIP MG B CONTROL PNL 504 l 175 B21-PT028A 1 12300 2.10 K.9 PRESS TRANSMITTER 795E877 506 l 176 B21-PT028B 2 12300 2.10 K.9 PRESS TRANSMITTER 795E877 506 l 177 B21-PT028C 3 12300 2.10 K.9 PRESS TRANSMITTER 795E877 506 , % l 178 B21-PT0280 4 12300 2.10 K.9 PRESS TRANSMITTER 795E877 506 ( l 179 B21 PT301A 180 B21 PT301B 1 2 12300 2.10 2.10 K.9 PRESS TRANSMITTER PRESS TRANSMITTER 795E877 795E877 506 506 l 12300 K.9
] 181 821-PT301C 3 12300 2.10 K.9 PRESS TRANSMITTER 795E877 506 l 182 B21-PT301D 4 12300 2.10 K.9 PRESS TRANSMITTER 795E877 506 l 183 P25-F016A 1 12300 5.20 K.6 TCV: C/B ELEC RM A 107E5182/0 511 l 184 U41-B603A 1 12300 5.20 K.5 ESS EQUIP RM COOL COIL 107E5189/0 511 l 185 U41-B603E 1 12300 5.20 K.5 ESS EQUIP RM COOL COIL 107E5189/0 511 l 186 U41-C604A 1 12300 5.20 K.1 EM ELEC (A) SUPP FAN A 107E5189/0 511 l 187 U41-C604E 1 12300 5.20 K.2 EM ELEC (A) SUPP FAN E 107E5189/0 ' 511 l 188 U41 TE112A 1 12300 5.40 K.5 TEMP ELEMENT 107E5189/0 511 l 189 U41 TE113A 1 12300 5.40 K.5 TEMP ELEMENT 107E5189/0 511 l 190 P25-F016C 3 12300 6.10 K.6 TCV: C/B ELEC RM C 107E5182/0 531 l 191 U41-B605C 3 12300 5.80 K.5 ESS EQUIP RM COOL COIL 107E5189/0 531 l 192 U41-B605G 3 12300 5.80 K.5 ESS EQUIP RM COOL COIL 107E5189/0 531 l 193 U41-C608C 3 12300 6.00 K.1 EM ELEC (C) SUPP FAN C 107E5189/0 531 l 194 U41-C608G 3 12300 6.00 K.2 EM ELEC (C) SUPP FAN G 107E5189/0 531 l 195 U41 TE112C 3 12300 6.10 K.5 TEMP ELEMENT 107E5189/0 531 l 196 U41-TE113C 3 12300 6.10 K.5 TEMP ELEMENT 107E5189/0 531 l 197 P25 C001C 3 12300 6.00 J.2 HECW PUMP C 107E5182/0 534 (jj 198 P25-C001F 3 12300 6.70 J.2 HECW PUMP F 107E5182/0 534 l 199 P25-0001C 3 12300 6.00 J.2 HECW REFRIGERATOR C 107E5182/0 534 Fire Hazard Analysis Database - Amendment 32 9A.6-101
l 23AG100 R1v. 9
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ABWR StandardSafetyAnalysis Report O Table 9A.6-3 Fire Hazard Analysis Equipment Data Base - Sorted by Room - Control Building (Continued) LOCATION LOCATION ITEM MPL ELECT ELEV. NUMBER ALPHA SYSTEM ROOM NO. NO. DIV. LOCATION COORD COORD. DESCRIPTION DRAWING NO. 200 P25-D001G 3 12300 6.70 J.2 HECW REFRIGERATOR F 107E5182/0 534 1 201 P25-DPT007C 3 12300 5.70 J.6 DP XMTR (FLO CONT C/F) 107E5182!0 534 202 P25-F005C 3 12300 6.90 J.2 TCV: MCR CLG 107E5182/0 534 203 P25-F012C 3 12300 5.70 J.4 PCV: HECW UNITS C/F 107E5182/0 534 204 P25-FIS003C 3 12300 6.00 J.2 FLOW IND SWITCH C 107E5182/0 534 205 P25-FIS003F 3 12300 6.70 J.2 FLOW IND SWITCH F 107E5182/0 534 206 P25-TE005C 3 12300 6.00 J.2 TEMP ELEM (UNIT C/F) 107E5182M 534 207 U41-C623C 3 12300 6.20 J.1 MCR RECIRC SUPP FAN C 107E5189/0 534 208 U41-C623G 3 12300 6.20 J.1 MCR RECIRC SUPP FAN G 107E5189/0 534 1 209 H11-P001* N 12300 4.00 K.0 COMPUTER PANELS 591 210 P25-A002 N 12300 l 5.30 J.5 CHEMICAL FEED TANK 107E5182/0 593 ; j 211 P25-DPT007A 1 12300 5.30 J.2 DP XMTR (FLO CONT A/D) 107E5182/0 593 l l 212 P25 TE005A 1 12300 6.30 J.2 TEMP ELEM (UN:T A/D) 107E5182/0 593 213 P25-C001A 1 17150 5.30 J.4 HECW PUMP A 107E5182/0 612 I l 213a P25-C0010 1 17150 5.30 J.4 HECW PUMP D 107E5182/0 612 214 P25-D001A 1 17150 5.30 J.4 HECW REFRIGERATOR A 107E5182/0 612 l l 214a P25-D001D 1 17150 5.30 J.4 HECW REFRIGERATOR D 107E5182/0 612 l 215 P25-F012A 1 17150 5.50 J.2 PCV: HECW UNIT A 107E5182/0 612 1 216 P25-FIS003A 1 17150 5.30 J.2 FLOWIND SWITCH A 107E5182/0 612 I l 216a P25-FIS0030 1 17150 5.30 J.2 FLOW IND SWITCH D 107E5182/0 612 217 U41-C605A 1 17150 5.20 K.5 EM ELEC (A) EXH FAN A 107E5189/0 613 218 U41 C60SE i 1 17150 5.20 K.6 EM ELEC (A) EXH FAN E 107E5189/0 613 l 219 DELETED 220 U41-C622C 3 17150 5.70 K.5 MCR HVAC EXH FAN C 107E5189/0 614 221 U41 C622G 3 17150 5.70 K.6 MCR HVAC EXH FAN G 107E5189/0 614 222 U41 DP1106C 3 17150 5.70 K.5 DIFF PRESS INDICATOR 107E5189/0 614 223 U41-DP1107C 3 17150 5.70 K.5 DIFF PRESS INDICATOR 107E5189/0 614 224 U41-DP1108C 3 17150 5.70 K.5 DIFF PRESS INDICATOR 107E5189/0 614 225 U41 DP1109C 3 17150 5.70 K.5 DIFF PRESS INDICATOR 107E5189/0 614 l 226 U41-F009C 3 17150 5.70 K.5 MO VALVE 107E5189/0 615 l 227 U41-F009F 2 17150 5.70 K.6 MO VALVE 107E5189/0 615 228 U41.F010C 3 17150 5.70 K.5 MO VALVE 107E5189/0 614 l 229 U41-F010F 2 17150 5.70 K.6 MO VALVE 107E5189/0 614 l ,230 DELETED 9A.6-102 Fire Hazard Analysis Database - Amendment 37
23A6100 Rsv. 9 ABWR standardsafety Analysis Report ( Table 9A.6-3 Fire Hazard Analysis Equipment Data Base - Sorted by Room - Control Building (Continued) LOCATION LOCATION ITEM MPL ELECT ELEV. NUMBER ALPHA SYSTEM ROOM NO. NO. DIV. LOCATION COORD. COORD. DESCRIPTION DRAWING NO. 231 U41-POT 105C 3 17150 5.70 K.5 POSITION TRANSMITTER 107E5189/0 614 232 L.'41-POT 105G 3 17150 5.70 K.6 POSITION TRANSMITTER 107E5189/0 614 233 U41 B601C 3 17150 6.80 J.3 MCR COOLING COIL 107E5189/0 615 234 U41-8601E 3 17150 6.80 J.3 MCR COOLING COIL 107E5189/0 615 235 U41-8601G 3 17150 6.80 J.3 MCR COOLING COIL 107E5189/0 6*5 236 U41-C621C 3 17150 6.70 J.5 MCR HVAC SUPP FAN C 107E5189/0 615 237 U41 C621G 3 17150 6.70 J.6 MCR HVAC SUPP FAN G 107E5189/0 615 238 U41-DP1101C 3 17150 6.80 J.3 DIFF PRESS INDICT. TOR 107E5189/0 615 239 U41.F007C 3 17150 6.80 J.1 MO VALVE 107E5189/0 615 l 240 U41 F007F 2 17150 6.90 J.1 MO VALVE 107E5189/0 615 241 U41-F008C 3 17150 6.80 J.1 MO VALVE 107E5189/0 615 242 U41-F011C 3 17150 6.60 J.1 MO VALVE 107E5189/0 615 l 243 U41-F008F 2 17150 6.90 J.1 MO VALVE 107E5189/0 615 244 U41-ME104C 3 17150 6.60 J.1 MOISTURE ELEMENT 107E5189/0 615 245 U41 TE103C 3 17150 6.60 J.1 TEMP ELEMENT 107E5189/0 615 246 U41 DPl111 A 1 17150 5.20 K.8 DIFF PRESS INDICATOR 107E5189/0 619 247 U41-F104A 1 17150 5.20 K.8 MO VALVE 107E5189/0 619 248 U41 TE110A 1 17150 5.20 L1 TEMP ELEMENT 107E5189/0 619 249 U41-F010F 2 17150 2.70 K.7 MO VALVE 107E5189/0 620 250 U41-POT 105F 2 17150 2.70 K.7 POSITION TRANSMITTER 107E5189/0 620 251 U41-C601B 2 17150 1.30 J.5 MCR HVAC SUPP FAN 8 107E5189/0 621 252 U41-C601F 2 17150 1.30 J.6 MCR HVAC SUPP FAN F 107E5189/0 621 253 P25-F0058 2 17150 1.10 J.2 TCV: MCR CLG 107E5182/0 621 254 U41 8601B 2 17150 1.10 J.3 MCR COOLING COIL 107E5189/0 621 255 U418601D 2 17150 1.10 J.3 MCR COOLING COLL 107E5189/0 621 256 U41 8601F 2 17150 1.10 J.3 MCR COOLING COIL 107E5189/0 621 257 U41-C603B 2 17150 1.80 J.1 MCR RECIRC SUPP FAN B 107E5189/0 621 258 U41-C603F 2 17150 1.80 J.1 MCR REC!RC SUPP FAN F 107E5189/0 621 259 U41-DP1101B 2 17150 1.10 J.3 DIFF PRESS INDICATOR 107E5189/0 621 260 U41-DP11068 2 17150 1.80 J.1 DiFF PRESS INDICATOR 107E5189/0 621 261 U41-OP11078 2 17150 1.80 J.1 DIFF PRESS INDICATOR 107E5189/0 621 [ 5 262 U41-DP1108B 2 17150 1.80 J.1 DIFF PRESS INDICATOR 107E5189/0 621
\ 263 U41 OPl109B 2 17150 1.80 J.1 DIFF PRESS INDICATOR 107E5189/0 621 264 U41-F0078 2 17150 1.10 J.1 MO VALVE 107E5189/0 621 l
Fire Hazard Analysis Database Amendment 37 9A.6- 103 j
l l 23A6100 Rsv. 9 ABWR standardSafetyAnslysis Repod O Table 9A.6-3 Fire Hazard Analysis Equipment Data Base - Sorted by Room - Control Building (Continued) i LOCATION LOCATION ITEM MPL ELECT ELEV. NUMBER ALPHA SYSTEM ROOM I NO. NO. DIV. LOCATION COORD COORD. DESCRIPTION DRAWING NO. l 265 U41-F007G 3 17150 1.10 J.1 MO VALVE 107E5189/0 621 266 U41-F008B 2 17150 1.10 J.1 MO VALVE 107E5189/0 621 l 267 l U41-F008G 3 17150 1.10 J.1 MO VALVE 107E5189/0 621 268 U41-F009B 2 17150 1.80 J.1 MO VALVE 107E5189/0 621 l 269 U41-F009G 3 17150 1.80 J.1 MO VALVE 107E5189/0 621 270 U 41-F0118 2 17150 1.50 J.1 MO VALVE 107E5189/0 621
- 271 U41-ME1048 2 17150 1.50 J.1 MOISTURE ELEMENT 107E5189/0 621 l I 272 U41 TE1038 2 17150 1.50 J.1 TEMP ELEMENT 107E5189/0 621 273 P25-C0018 2 17150 2.80 J.4 HECW PUMP B 107E5182/0 623 1
274 P25-C001E 2 17150 2.80 J.8 HECW PUMP E 107E5182/0 623 275 P25-D001B 2 17150 2.80 J.4 HECW REFRIGERATOR B 107E5182/0 623 i
; 276 P25-D001E 2 17150 2.80 J.8 HECW REFRIGERATOR E 107E5182/0 623 277 P25-DPT007B 2 17150 2.30 J.2 DP XMTR (FLO CONT B/E) 107E5182/0 623 278 P25-F0128 2 17150 2.50 J.2 PCV: HECW UNITS B/E 107E5182/0 623 279 P25-FIS0038 2 17150 2.80 J.4 FLOWIND SWITCH B 107E5182/0 623 280 P25-FIS003E 2 17150 2.80 J.8 FLOWIND SWITCH E 107E5182/0 623 281 P25-TE0058 2 17150 2.80 J.4 TEMP ELEM (UNIT BE) 107E5182/0 623 282 P25-F0168 2 17150 1.70 K.8 TCV:C/B ELEC RM B 107E5182/0 624 283 U41 B6048 2 17150 1.70 K.8 ESS EQUIP RM COOL COIL 107E5189/0 624 284 U41-8604F 2 17150 1.70 K.8
)
ESS EQUIP RM COOL COLL 107E5189/0 624 ; 285 U41-C606B 2 17150 1.80 K.5 EM ELEC (B) SUPP FAN b 107E5189/0 624 l 288 U41-C606F 2 17150 1.80 K.6 EM ELEC (B) SUPP FAN F 107E5189/0 624 287 U41 DPl111B 2 17150 1.60 K.8 DIFF PRESS INDICATOR 107E5189/0 624 288 U41-F1048 2 17150 1.60 K.8 MO VALVE 107E5189/0 624 289 U41 TE110B 2 17150 1.60 L.1 TEMP ELEMENT. 107E5189/0 624 l 290 U41 TE1128 2 17150 2.00 K.6 TEMP ELEMENT 107E5189/0 624 291 U41 C6078 2 17150 2.20 K.5 EM ELEC (B) EXH FAN 8 107E5189/0 625 292 U41-C607F 2 17150 2.20 K.6 EM ELEC (B) EXH FAN F 107E5189/0 625 l 293 DELETED 294 U41 TE1138 2 17150 2.20 K.5 TEMP ELEMENT 107E5189/0 625 295 U41-C6028 2 17150 2.70 K.5 MCR HVAC EXH FAN B 107E5189/0 626 296 U41 C602F 2 17150 2.70 K.6 MCR HVAC EXH FAN F 107E5189/0 626 l 297 U41 F010B 2 17150 2.70 K.5 MO VALVE 107E5189/0 626 l 297a U41-F010G 3 17150 2.70 K.5 MO VALVE 107E5189/0 626 9A.6 104 Fire Hazard Analysis Database - Amendment 37
I 23A6100 Rev. 9 ABWR StandardSafety Analysis Report i Table 9A.6-3 Fire Hazard Analysis Equipment Data Base-Sorted by Room - Control Building (Continued) LOCATION LOCATION ITEM MPL ELECT ELEV. NUMBER ALPHA SYSTEM ROOM NO. NO. DIV. LOCATION COORD. COORD. DESCRIPTION DRAWING NO. l 298 DELETED 299 U41-POT 105B 2 17150 2.70 K.5 POSITION TRAt;; NUTTER 107E5189/0 626 300 U41-C609C 3 17150 6.20 K.5 EM ELEC (C) EXH FAN C 107E5189/0 631 301 U41-C609G 3 17150 6.20 K.6 EM ELEC (C) EXH FAN G 107E5189/0 631 l l 302 DELETED 303 U41-DP1111C 3 17150 5.80 K.8 DIFF PRESS INDICATOR 107E5189,V 653 l \ 304 U41-F104C 3 17150 5.80 K.8 MO VALVE 107E5189/0 653 j l 305 U41 TE110C 3 17150 5.90 L1 TEMP ELEMENT 107E5189,V 653 I b U 1 l l l V' l Fire Hazard Analysis Database Amendment 37 9AG-105
23A6100 Rsv. 2 ABWR StandardSafety Analysis Report O Table 9A.6-4 Fire Hazard Analysis Equiprnent Database-Sorted by Room-Turbine Building Loc Loc MPL Elec Elev. No. Alpha System Room l No. Div. Loc. Coord Coord Description Drawing No. [ D11-C001 A* N 350 2.4 G.3 PUMP NP-1005256 110 l D17-C001R* N 350 2.8 G.3 PUMP NP-1005256 110 l D11-F402 N 350 2.4 G.3 MO VALVE NP-1005256 110 l D11-LIS064 N 350 2.6 G.2 LEVEL INDICATOR NP-1005256 110 SWITCH l D11-RE061 N 350 2.3 F.3 RADIATION DETECTOR NP-1005256 112 l D11-RE111A N 350 2.6 F.6 RADIATION DETECTOR NP-1005256 112 l D11 RE111C N 350 2.9 F.3 RADIATION DETECTOR NP-1005256 112 l D11-RE111D N 350 2.4 F.9 RADIATION DETECTOR NP-1005256 112 l N22-C0018 N 350 3.5 G.3 HEATER DRAIN PUMP B -?- 113 l 821-PT301B 2 1500 6.2 D.8 PRESS TRANSMI' ITER 795E877 120 1 D21 RE026 N 3350 4.0 F.0 AREA RAD. DETECTOR NP 1005274 120 l l 821-PT301D 4 1500 6.6 D.8 PRESS TRANSMITTER 795E877 120 l B21-PT301A 1 1500 6.2 D.0 PRESS TRANSMITTER 795E877 120 l 821 PT301C 3 1500 6.6 D.0 PRESS TRANSMITTER 795E877 120 l D21 RE030 N 3350 6.2 G.5 AREA RAD DETECTOR NP-1005274 130 l N22 C001 A N 350 6.5 G.3 HEATER DRAIN PUMP A -?- 130 l N21-C001A N 350 2.4 D.4 CONDENSATE PUMP A -?- 140 l N21-C001B N 350 2.4 D.8 CONDENSATE PUMP B -?- 140 l N21-C001C N 350 2.4 E.3 CONDENSATE PUMP C -?- 140 l N21-C001D N 350 2.4 E.6 CONDENSATE PUMP D -?- 140 l K11-C051 A N 380 2.1 C.6 LCW PUMP FOR SUMP NT-5000339 142 (A) l K11-C051B N 380 2.2 C.7 LCW PUMP FOR SUMP NT-5000339 142 (A) l K11-C051C N 380 2.3 C.6 LCW PUMP FOR SUMP NT-5000339 142 (B) l K11-C051D N 380 2.4 C.7 LCW PUMP FOR SUMP NT-5000339 142 (B) l K11-C151 A N 350 2.1 C.6 HCW SUMP PUMP NT-5000339 142 l K11-C151B N 350 2.2 C.7 HCW SUMP PUMP NT-5000339 142 9A6-106 Fire Hazard Analysis Database - Amendment 32
23A6100 Rtv. 9 ABWR standardsafetyAnalysisRepon in r s 9C Regulatory Guide 1.52, Section C, Compliance Assessment This Appendix provides the compliance status of the ABWR Control Room Habitability Area (CRHA) HVAC System design with each of the regulatory positions specified under Section C of Regulatory Guide 1.52, and the revision cited in Table 1.8-20. Following each prosision of Regulatory Guide 1.52 is an evaluation of the ABWR compliance with that position. If the ABWR deviates from die Regulatory Guide 1.52 position. justification is prosided. Note that the similarly numbered sections from the revisions cited in Table 1.8-21 for ANSI N509 and N510 are used for ABMR CRHA HVAC System design except as othenvise noted; Regulatory Guide 1.52 references older revisions (1976) of these standards. Compliance as described in the remainder of this response is measured against the applicable section of the standards referenced in Table 1.8-21. The Habitability systems provide the capability to detect and limit the introduction of 1 radioactive material and smoke into the control room from the sources external to the control building. 9C.1 ABWR Compliance with RG 1.52, Revision 2, Section C g ( (1) Emironmental Design Criteria (a) "The design of an engineered safety feature atmosphere cleanup system should be based on the maximum pressure differential, radiation dose rate, relative humidity, maximum and minimum temperature, and other conditions resulting from the postulated DBA and on the duration of such conditions." 1 The design is in compliance with this position. l l (b) "The design of each ESF system should be based on the radiation dose l to essential services in the vicinity of the adsorber section, integrated I over the 30-day period following the postulated DBA. The radiation , source term should be consistent with the assumptions found in i Regulatory Guides 1.3,1.4 and 1.25. Other engineered safety features, ; including pertinent components of essential senices such as power, air, i and control cables, should be adequately shielded from the ESF atmosphere cleanup systems." l The design is in compliance with this position. Table 31-19 provides the radiation emironmental conditions inside control room for plant abnormal and accident conditions. Note that integrated doses for six months, not 30 l days, are prosided in Table 31-19. ABWR Compliance with RG 1.52. Revision 2, Section C Amendment 37 9C 1
l 23A6100 Rvu 4 ABWR StandardSafety Analysis Report l l l (c) "The design of each adsorber should be based on the concentration and I relative abundance of the iodine species (elemental, particulate, and organic), which should be consistent with the assumptions found in Regulatory Guides 1.3,1.4 and 1.25." The design is in compliance with this position. (d) "The operation of any ESF atmosphere cleanup system should not deleteriously affect the operation of other engineered safety features such as a containment spray system, nor should the operation of other engineered safety features such as a containment spray system deleteriously affect the operation of any ESF atmosphere cleanup l system." 1 The design is in compliance with this position. I (e) Components of systems connected to compartments that are unheated during a postulated accident should be designed for post-accident effects of both the lowest and highest predicted temperatures. The design is in compliance with this position. (2) System Design Criteria l (a) "ESF atmosphere cleanup systems designed and installed for the purpose of mitigating accident doses should be redundant. The systems should consist of the following sequential components: (1) prefilters (2) HEPA filters before the adsorbers, (3) iodine adsorbers (impregnated activated carbon or equhalent adsorbent such as metal zeolites), (4) HEPA filters after the adsorbers, (5) ducts and valves, (6) fans, and (7) related instrumentation. Heaters or cooling coils used in conjunction with heaters should be used when the humidity is to be controlled before filtration." The design is in compliance with this position (b) "The redundant ESF atmosphere cleanup systems should be physically separated so that damage to one system does not also cause damage to the second system. The generation of missiles from high-pressure equipment rupture, rotating machinery failure, or natural phenomena should be considered in the design for separation and protection." The design is in compliance with this position. (c) "All components of an engineered-safety-feature atmosphere cleanup system should be designated as Seismic Category I (see Regulatory 9C 2 Regulatory Guide 1.52. Section C. Compliance Assessment- Amendment 34
1 23A6100 R:v. 9 ABWR senadudsawAnsarsisneper ,i The carbon bed is 100 mm deep. Per Table 2, the decontamination efIiciency for bed depths 100 mm is 99%." (i) "The adsorber section meets the conditions given in regulatory i) a Position C.5.d of this guide." As stated previously, the ABWR CRHA HVAC System complies with [ Position C.5.d. 1, j (ii) "New activated carbon meets the physical property specifications
- given in Table 5.1 of ANSI N509-1976, and" 1 Activated carbon installed in the CRHA HVAC System will be covered by
[ purchase requirements to meet the physical properties specified in j Table 5-1 of ANSI N509. 1 j (iii) " Representative samples of used activated carbon pass the
- laboratory tests given in Table 2."
a l Surveillance testing is prosided to comply with this position. This i position is interpreted as follows. Representative samples of used lg jp activated carbon will be laboratory tested with a frequency defined in Footnote c of Table 2 and as reflected in the technical specifications. } Also per Footnote c of Table 2, a representative sample is defined in
- Position C.6.b. Testing will be performed at a relative humidity of 70%
! per ASTM D3803. The test acceptance criterion will be a methyl iodide l penetration ofless than 0.175%. ASTM D3803 is cited in Table 5-1 of l ANSI N509-1980 for tests equivalent to those specified in Test 5.b of j ANSI N5091976. p If the activated carbon fails to meet any of the above conditions,it should not be used in engineered-safety-feature adsorbers. s j "The acuvated carbon for the CRHA HVAC System will meet the conditions of i Position 6.a(i), (ii) and (iii)." I (b) "The efficiency of the activated carbon adsorber section should be , determined by laboratory testing of representative samples of the
- actinted carbon exposed simultaneously to the same service conditions as the adsorber section. Each representative sample should be not less than two inches in both length and diameter, and each sample should l have the same qualification and batch test characteristics as the system adsorbent. There should be a sufficient number of representative
) samples located in parallel with the adsorber section to estimate the
- {s amount of penetration of the system adsorbent throughout its senice
- life. The design of the samplers should be in accordance with the i
ABWR Compliance with RG 1.52, Revision 2, Section C- Amendment 37 9C-11 i
23A6100 R3v. 5 ABWR standard safetyAnalysis Report provisions of Appendix A of ANSI N509-1976. Where the system 91 i activated carbon is greater than two inches deep, each representative sampling station should consist of enough two. inch samples in series to equal the thickness of the system adsorbent. Once representative samples are removed for laboratory test, their positions in the sampling , array should be blocked off." The detailed design will be in compliance with this position.
"l2boratory tests of representative samples should be conducted, as indicated in Table 2 of this guide, with the test gas flow in the same direction as the flow during senice conditions. Similar laboratory tests should be performed on an adsorbent sample before loading into the adsorbers to establish an initial point for comparison of future test results. The activated carbon adsorber section should be replaced with new unused activated carbon meeting the physical property specifications of Table 5.1 of ANSI N509-1976 if (1) testing in l accordance with the frequency specified in Footnote c ofTable 2 results in a representative sample failing to pass the applicable test in Table 2 o(2) no representative sample is available for testing."
The CRHA HVAC System design and testing will comply with this position. Physical property testing is addressed in the response to Position C.6.a(2). O 9C-12 Regulatory Guide 1.52. Section C, Compliance Assessment - Amendment 35
l l i 234ts100 Rtv. 2 ABWR StandardSafetyAnalysisReport i l'D
\v) \
i (d) Proper operating conditions (flow, vibration, bearing temperature) of i the RCW pumps in design mode of operations.
]
)
(e) Acceptable pump NPSH under the most limiting design flow conditions. l (f) Proper operating conditions and system performance capability during f the following system operational tests: 1 (i) System operation tests at various operating modes (normal, l shu.tdown cooling, hot standby without offsite power source, l LOCA, and refueling outage) l (ii) Switching capability test of RCW pumps and heat exchangers l' between 1-unit and 2-unit operations (iii) Operation mode transfer test from normal mode to LOCA mode by LOCA signal (ii, Transferability test to the hot standby mode operation upon loss of offsite power ; (v) System operation capability from the remote shutdown panel j c.si (vi) Flow balancing for all modes of operation ] ( ( ,) (g) Proper RCW pump motor start sequence and actuation of protective desices. (h) Proper operation ofinterlocks functions including operations of all components subject to interlocking. This test can be performed by using simulated signal, if actual initiation is not practical. (i) Proper operation cf permissive, prohibit, and bypass functions. (j) Proper system operation while powered from primag ana ernate sources, including transfers, and in degraded modes for which the system is expected to remain operational. This includes isolation / shedding of nonessential loads and divisional interties when a
- LOCA signal is present.
j (k) Acceptable pump / motor vibration and noise levels and system piping l movements during both trmsient and steady-state operation. This test can be perforu.ed in conjuction with expansion, vibration and dynamic effects preoperational test (Subsection 14.2.12.1.51). (1) Proper operation of system surge tank and chemical addition tank and their associated functions during system operational test. (m) Acceptable performance capability of RCW heat exchangers, to the f']/ (, extent practical. Otherwise, RCW heat exchanger confirmatog test can be performed in startup test stage. Specific information to be included in Final Safety Analysis Reports - Amendment 32 14.2-59
23A6100 Rev. 9 ABWR standardSafety Analysis Report O 14.2.12.1.30 Deleted 14.2.12.1.31 Hot Water Heating System Preoperational Test (1) Purpose Verify the ability of the Hot Water Heating System (HWHS) to proside hot water to the appropriate HVAC systems and the operation of HWH pump, heat exchanger, surge tank and chemical addition tank. (2) Prerequisites The construction tests have been completed, and the SCG has reviewed the test procedure and approved the initiation of testing. Electrical power, SA System, TCW System, Heat Steam System, HVAC System, HNCW System and other required interfacing systems shall f.e available, as needed, to support the specified tesdng. Additionally, a temporary strainer shall be installed at the suction side of the HWH pump. (3) GeneralTest Methods and Acceptance Criteria Performance shall be observed and recorded during a series ofindividual component and integrated system tests.These tests shall demonstrate that the HWHS operates properly as specified in appropriate HWHS design specification and manufacturer's technicalinstruction manual through the following testing: 1 (a) Proper operation ofinstrumentadon and system controls in all combinations oflogic and instrument channel trip I l (b) Verification ofvarious component alarms, for correct system response to process variable, and provides alarms at the prescribed value (c) Proper operation of system valves, including open/ closure cycling and position indicator verification,if applicable (d) Proper operating conditions (flow, vibration, bearing temperature) of the HWH pumps during continuous pump run test (e) Acceptable pump NPSH under the most limiting design flow conditions. (f) Proper operating conditions and system performance capabili' during the following operadon mode tests: (i) Plant normal operation mode (ii) Plant shutdown and inspecticn mode (g) Proper pump motor start sequence and actuation of protective devices 14.2 60 Speciric Information to be included in Final Safeny Analysis Reports - Amendment 37
23A6100 REv. 9 ABWR standers safety Analysis neport / k (h) Proper operation ofinterlock functions, including operation of all l components subject to interlocking (e.g., HWHS pump trip on low l surge tank level, and system water temperature control, etc.) (i) Proper operation of permissive, prohibit, and bypass functions (j) Proper operation of system surge tank and chemical addition tank and i their associated functions during system operation mode tests 1 14.2.12.1.32 HVAC Emergency Cocling Water System Prooperational Test I (1) Purpose l j To verify the ability of the HVAC Emergency Cooling Water (HECW) System ; to supply the design quantities of chilled water at the specified temperatures
)
to the various cooling coils of the HVAC systems serving rooms and areas containing essential systems and equipment. (2) Prerequisites
.' The construction tests have been successfully completed, and the SCG has reviewed the test procedure and approved the initiation of testing. Normal x
and auxiliary electrical power, IA, MUWP, RCW, applicable HVAC System cooling coils, and other required sy.< terr. interfaces shall be available, as l needed, to support the specified system testing. I l (3) General Test Methods and Acceptance Cntena l Performance shall be observed and recorded during a series ofindividual component and integrated system tests. These tests shall demonstrate that the HECW System and its auxiliary equipment operate properly as specified in Subsections 9.2.13 and 7.3.1.1.9 and applicable HECW System design specification through the following testing: (a) Proper operation ofinstrumentation and system control functions including flow switch, surge tank level controller, and chilled water temperature controller (b) Verification of various component alarms, for correct alarm actuation and reset, alarm set value, alarm indication and operating logic (c) Proper operation of system motor-operated and air-operated valves, including operability and position indication verifications, if applicable b (d) Proper operadon of HECW pumps and motors during continuous run U tests ; Specific Information to be included in Final Safety Analysis Reports - Amendment 37 14.2-61
23A6100 R2. 2 ABWR standardsafetyAnalysisReport O (e) Acceptable pump NPSH under the most limiting design flow conditions (f) Proper operating conditions (flow rate, pressure, and temperature) and system performance capability in conformity with the design during the following system operational tests: (i) System flow rate test to confirm that system flow rate is prescribed value under the system design operating conditions.
~'
(ii) Single operational test of HECW pumps to verify that each HECW pump can be individually operated continuously at rated flow rate without any abnormalities. (iii) Operational test of all HECW pumps to confirm that all HECW pumps can be continuously operated without any problems in HECW System. (iv) Flow rate to each load shall be verified and adjusted (if necessary) to be consistent with the prescribed value. This test shall also confirm that each coil in supply units has adequate cooling capacity and each room temperature is under the design temperature. (v) Chemical addition test to confirm that the concentration of inhibitor in circulating water in HECW System is within prescribed limits. (g) Proper pump motor start sequence and actuation of protective desices (h) Proper operation ofinterlocks including confirmation that all components are operated in conformity with IBD and Sequence Diagram (i) Proper operation of permissive, prohibit, and bypass functions (j) Proper system operation while powered from primary and alternate sources, including transfers, and in degraded modes for which the system is expected to remain operational (k) Acceptable pump / motor vibration and nois: levels and system piping movements during both transient and steady-state operation 14.2.12.1.33 HVAC Narmal Cooling Water System Preoperational Test (1) Purpose To verify the ability of the HVAC Normal Cooling Water (HNCW) System to supply the design quantities of chilled water at the specified temperatures to the various cooling coils of the HVAC systems sening rooms and areas containing nonessential equipment and systems. 14 2-62 Specific Information to be Included in Final Safety Analysis Reports - Amendment 32
7 23A6100 Rw. 2 l ABWR sondedsecuryAasysisnever i J Ievel 2 l During full closure testing ofindividual turbine control, stop, and bypass valves, the transient peak values of reactor vessel pressure, neutron flux, i simulated fuel surface heat flux, and main steamline flow must have adequate scram avoidance margins as required by the GE Startup Test Specifications. The measured total bypass valve capacity shall be equal to or greater than that used for the nuclear safety operational analysis (NSOA) as shown in Table 15.0-1. l 14.2.12.2.26 MSIV Performance (1) Purpose To demonstrate proper operation of and to verify closure times for main steamline isolation valves, including branch steamline isolation valves, during power operation. (2) Prerequisites The preoperational tests have been completed and plant management has reviewed the test procedure (s) and approved the initiation of testing. For each scheduled testing iteration, the plant shall be in the appropriate operational configuration with the specified prerequisite testing complete. Additionally, applicable instrumentation shall have been checked or calibrated as appropriate.
)
(3) Description At rated temperature and pressure, and then again at an intermediate power level, each MSIV will be individually stroked in the fast closure mode. Vahe operability and closure time will be verified and overall plant response j obsen>ed. Closure times will be evaluated consistent with Technical Specification and safety analysis requirements. The MSIV closure time equals l the interval from de-energization of the valve solenoids until the valve is 100% l closed. If appropriate, it may also be desirable to determine the maximum ! power level at which such tests can safely be performed by extrapolating plant l response during such tests at successimly power levels during power j ascension. In addition, at rated temperature and pressure, proper functioning and stroke timing ofbranch steamline isolation valves (e.g., on common drain f line) will be demonstrated as part of the IST program (Table 3.9-8), i \ l 3 Specific Information to be included in Final Safety Analysis Reports - Amondment 32 142-169 1 1
l 23As100 R1v. 9 ABWR standardsatoryAnalysis neport 1 O l (4) Criteria Level 1 l l MSIV closure time (exclusive of electrical delay), for any indisidual value, shall be within the upper and lower closure time limits specified in the plant Technical Specifications. The reactor shall not scram or isolate during full trip closure ofindisidual MSIV at power levels up to the maximum allowable power level for conducting such tests as specified by the applicable plant surveillance procedure. Level 2 During full trip closure testing ofindividual MSIV, the transient peak values of reactor vessel pressure, neutron flux, simulated fuel surface heat flux, and main steamline flow must have adequate scram avoidance margins as required by the GE Startup Test Specifications. l < 14.2.12.2.27 SRV Performance (1) Purpose To demonstrate that there are no major blockages in the relief valve discharge i piping, and that each safety / relief valve can be opened and closed properlyin , the manual actuation mode, and will reseat properly after operation, during l l reactor power operation. l , l (2) Prerequisites l l l l The preoperational tests have been completed and plant management has l reviewed the test procedure (s) and approved the initiation of testing. For each l scheduled testing iteration, the plant shall be in the appropriate operational l configuration with the specified prerequisite testing complete. Additionally, l applicable instrumentation shall have been checked or calibrated as appropriate. (3) Description A functional and flow demonstration test of each SRV shall be made when adequate reactor steam dome pressure is available to avoid damaging the valve. Adequate pressure at which this test is to be performed is 6.55 MPaG, as recommended by the valve manufacturer. Opening and closing of each valve, as well as evidence of steam discharge, will be verified by response of SRV l discharge tailpipe temperature sensors and by observed changes in steamflow 14.2 170 Specific Information to be includedin Final Safety Analysis Reports - Amendment 37
23A6100 Rsv. 9 ABWR standardsafety Analysis Report O V 14.3 Certified Design Material This secdon of the SSAR provides the selecdon criteria and processes used to develop the ABWR Certified Design Material (CDM) that is presented in the GE document 25A5447,"ABWR Certified Design Material." This document provides the principal design bases and design characteristics that are certified by the 10 CFR Part 52 rulemaking process and included in the formal certification Rule. This top-level design information in the CDM is extracted direcdy from the more detailed ABWR design information presented in the SSAR (which is part of the certification application). Limiting the certified design contents to top-level information reflects the tiered approach to design certification endorsed by the Commission (Staff Review Memorandum 2/15/1991 regarding SECY-90-377; 10 CFR Part 52 Statement of Consideradons 54 Fed. Reg. 15372,154377, (1989). See also SECY-90-241,90-377 and SECY-91-178.) The objective of this SSAR section is to define the bases and methods that were used to develop the CDM document for the ABWR. This SSAR section contains no new technicalinformation regarding the ABWR design. ( The ABWR CDM consists of the following: (1) An introduction section which defines terms used in the CDM as well as listing general provisions that are applicable to all CDM entries. The intent of these entries is to avoid ambiguities and misinterpretations by providing front-end guidance to users of the CDM. (2) Design descriptions for: a) systems that are fully within the scope of the ABWR design certification, and b) the in-scope portion of those systems that are only partially within the scope of the ABWR design certification. The intent of the CDM design descriptions is to delineate the principal design bases and principal design characteristics that are referenced in the design certification Rule. The design descriptions are accompanied by the inspections, tests, analyses and acceptance criteria (ITAAC) required by 10 CFR 52.47(a) (1) (vi) to be part of the design certification application. The ITAAC define verification activities that are to be perfcrmed for a facility with the objective of confirming that the plant is built and will operate in accordance with the design certification. Successful completion of these certified design ITAAC, together with the combined license (COL) applicant's ITAAC for the site-specific portions of the plant, will be the basis for the NRC finding under 10 f CFR Part 52.103 (g). \ Certified Design Material- Amendment 37 14.3 1
23A6100 Rev. 3 ABWR StandardSafetyAnalysisReport O I (3) Design descriptions and their associated ITAAC for design and construction activities that are applicable to more than one system. Design related processes have been included in the CDM for: (a) Aspects of the ABWR design likely to undergo rapid, beneficial technological developments in the lifetime of the design certification. Certifying the design processes associated with these areas of the design rather than specific design details permits future license applicants referencing the ABWR design certification to take advantage of the best technology available at the time of COL application and facility construction. Example: design of programmable, microprocessor-based instrumentation and control systems. (b) Aspects of the design which are dependant upon characteristics of as-procured, as-installed systems, structures and components. These characteristics are not available at the time of certification and therefore cannot be used to develop and certify design details. Example: design of piping systems which are dependent upon detailed routing and equipment information. (4) Interface requirements as defined by 10 CFR Part 52.47(a) (1) (vii). Interface requirements are those requirements which must be met by the site-specific portions of the complete nuclear power plant that are not within the scope of , the certified design. These requirements define characteristics of the site- l specific features which must be provided in order for the certified design to comply with certification commitments. Interface requirements are defined Ior: a) systems entirely outside the scope of the design certification and b) the out<>f-scope portions of those systems that are only partially within the scope of the design certification. The COL applicant will provide ITAAC for the site-specific design features that implement the interface requiremen ts; therefore, the CDM does not include ITAAC for interface requirements. l (5) Site parameters used as the basis for ABWR design presented in the SSAR. These parameters represent a bounding envelope of site conditions for any license application referencing the ABWR design certification. No ITAAC are necessary for the site parameters entries because compliance with site parameters will be verified as part ofissuance of a license for a plant that references the ABWR design certification. (6) Appendices listing acronyms and legends used in the body of the CDM. (This material is self-explanatory and is not discussed any further in this SSAR section.) 14.3-2 Certdied Design Material- Amendment 33
I i 23A6100 Rsv. 9 ABWR standard safety Analysis Report O Table 14.3-3 Transient Analysis SSAR Entry Parameter SSAR Value Tab le 15.0-1 Reactor Internal Recirculation Pumps Number of Pumps 10 j Pump Trip Inertia (kg m2) Trip Mitigation (maximum) 26.5 Accident (minimum) 17.5 I Relief Valve (Relief Function) Capacity (% NBR Steam Flow at 7.89 MPaG) 91.3 Number of Valves 18 Opening Time (s) 0.15 (valve stroke timo only. Does not include 0.1 s delay to energize solenoid) High Flux Trip Scram -- APRM Simulated Therroal Power Trip Scram -- [~) Tota: Steamline Volume (m3) 113.2 Table 15.0-6 FMCRD Scram Times 10% Rod insertion (s) 0.46 40% Rod Insertion (s) 1.208 60% Rod Insertion (s) 1.727
\
100% Rod Insertion (s) 3.719 j 15.1.1.2.2 High Simulated Thermal Power Trip Scram -- l 1 Table 15.1-5 High Water Level 8 initiates l Feedwater Pump Trip -- Table 15.1-5 Turbine Stop Valve Position Switches initiate Reactor Scram - - - ! Trip of 4 RIPS -- Table 15.1-7 Low Water Level 2 Initiates I l Trip of 6 RIPS -- RCIC System -- Maximum Startup Time [(s)-includes 1.0 s for 30 instrument delay) [~ MSIV Closure on Low Turbine inlet Pressure h] 15.1.3.3.1 MSIV Closure Time (s - maximum isolation valve closing 5.0 ) time plus 0.5 s for instrument delay) j i l Certified Design Material- Amendment 37 14.3 29 l
23A6100 Rsv. 9 ABWR Standard Safety Annlysis Report O Table 14.3-3 Transient Analysis (Continued) SSAR Entry Parameter SSAR Value [ Table 15.1-9 SRNM High Neutron Flux Scram 15.2.1.3.1 TCV Full Stroke Servo Closure (s) 2.5 Table 15.2-1a Low Water Level 3 Initiates Trip of 4 RIPS -- Table 15.2-2 High Dome Pressure initiates Trip of 4 RIPS --- l Table 15.2-3 T/G Load Rejection initiates Turbine Control Valve Fast Closure --- Turbine Bypass System Operation on High Pressure --- Fast Control Valve Closure Initiates Scram --- l Trip of 4 RIPS ---- 15.2.2.3.1 TCV Full Stroke Fast Closure (s -- from normal operating position) 0.08 Table 15.2-6 Turbine Trip initiates Turbine Control Valve Fast Closure --- Turbine Bypass System Operation on High Pressure --- 15.2.3.3.1 Turbine Stop Valve Full Stroke Closure (s) 0.10 Table 15.2-9 MSIV Position Switches Initiate Scram --- 15.2.4.3.1 Minimum MSIV Closure Time (s) 3.0 Table 15.2-14 Low Condenser Vacuum initiates MSIV Closure -- 15.2.6.1.1.2 RIP M/G Set Number of RIPS 6 Length of Time Hold Original Speed (s) 1.0 RIP Coastdown Rate (% per s) 10 Length of Time (s) 2.0 Time of RIPTrip(s) 3.0 ! Table 15.2-17 Low Water Level 3 initiates Reactor Scram - 15.2.7.2.21 Meets Single-failure Criterion -- 15.2.9 RHR System has 3 Independent Divisions - 15.3.1.1.1 No More Than 3 RIPS on One Electrical Power Bus -- I 14.3-30 Certified Design Material- Amendment 37
23A6100 R:v. 9 ABWR standardsareryAnalysisnepon C (.)\ expected during the course of the event. Once isolation occurs, the pressure will increase to a point where the SRVs open. The operator should: (1) Monitor that all rods are in (2) Monitor reactor water level and pressure (3) Observe turbine coastdown and break vacuum before the loss of steam seals. Check turbine auxiliaries (4) Observe that the reactor pressure relief valves open at their setpoint (5) Observe that RCIC initiated on lowwater level (6) Secure RCIC when reactor pressure and level are under control (7) Monitor reactor water level and continue cooldown per the normal procedure (8) Complete the scram report and initiate a maintenance survey of the SB&PCS before reactor restart 15.1.3.2.2 Systems Operation w/ 15.1.3.2.2.1 Inadvertent Opening of One Turbine Bypass Valve This event does not require any protection system or safeguard system operation. This analysis assumes normal functioning of plant instnsmentation and controls. 15.1.3.2.2.2 Inadvertent Opening of All Turbine Control Valves and Bypass Valves To properlysimulate the expected sequence of events, the analysis of this event assumes normal functioning of plant instrumentation and controls, plant protection and reactor protection systems, except as otherwise noted. Initiation of RCIC System functions occurs when the vessel water level reaches the L2 setpoint. Normal startup and actuation can take up to 30 seconds before effects are realized. If these events occur, they will follow sometime after the primary concerns of fuel thermal margin and overpressure effects have occurred, and are expected to be less severe than those already experienced by the system. 15.1.3.3 Core and System Performance 15.1.3.3.1 Input Parameters and initial Conditions e C A five-second isolation valve closure (maximum isolation valve closing time plus instrument delay ) instead of a three second closure is assumed when the turbi ie Decrease in Reactor Cooiant Temperature - Amendment 37 15.1 9
i l l 23A6100 Riv. 9 ' l ABWR standus satory Analysis neport i l 1 pressure decreases below the turbine inlet low pressure setpoint for main steamline isolation initiation. This is within the specification limits of the valve and represents a conservative assumption. 15.1.3.3.2 Results 15.1.3.3.2.1 Inadvertent Opening of One Turbine Bypass Valve l The simulated inadvertent opening of one turbine bypass valve is presented in j Figure 15.14. When the decrease in reactor pressure is sensed, the pressure control l system starts immediately to command turbine control valves to close to maintain the l reactor pressure. The vessel water level increases slightly (about 10 cm) and then settles back to its norrnal level. Reactor pressure decreases by about 0.069 MPaD. AfCPR remains above the safetylimit. l 15.1.3.3.2.2 Inadvertent Opening of All Turbine Control Valves and Bypass Valves l Figure 15.1-5 presents graphically how the high water level turbine trip and the isolation l valve closure stops vessel depressurization and produces a normal shutdown of the ! isolated reactor I Depressarization results in formation of voids in the reactor coolant and causes a decrease in reactor power almost immediately. The depressurization rate is large enough such that water level swells to the sensed level trip setpoint (L8), initiating main turbine and feedwater pump trips. Position switches on the turbine stop valves initiate reactor scram and a trip of four RIPS. 1 After a pressurization resulting from the turbine stop valve closure, pressure again drops and continues to drop until turbine inlet pressure is below the low turbine pressure isolation setpoint when main steamline isolation finally terminates the depressurization. The turbine trip and isolation limit the duration and severity of the depressurization so that no significant thermal stresses are imposed on the reactor coolant pressure boundary. No significant reduction in fuel thermal margins occur; therefore, this event does not have to be analyzed for specific core configurations. i 15.1.3.4 Barrier Performance Barrier performance analyses were not required because the consequences of this event do not result in any temperature or pressure transient in excess of the criteria for which fuel, pressure vessel or containment are designed. Dudog the event of inadvertent opening of all turbine control and bypass valves, peak pressure in the bottom of the vessel reaches 8.02 MPaG, which is below the AShiE code limit of 9.48 AffaGfor the reactor coolant pressure boundary. Vessel dome pressure reaches 7.88 hfPaG, below the setpoint of the second pressure relief group. Niinimum vessel dome pressure of 4.96 l MPaG occurs at about 40 seconds. 15.1-10 Decrease in Reactor Coolant Temperature ~ Amendment 37
1 23A6100 Rsv. 9 t ABWR standardsafety Analysis Report n
/ T O l 15.2.4.3 Core and System Performance l 15.2.4.3.1 Input Parameters and initial Conditions The main steam isolation valves close in 3 to 4.5 seconds. The worst case (the 3 second closure time) is assumed in this analysis. No credit was taken for instrument delay.
1 Position switches on the valves initiate a reactor scram when the valves are less than 85% open. Closure of these valves causes the dome pressure to increase. Four RIPS are tripped when the high pressure setpoint is reached. ABWR has motor-driven feedwater pumps. However, a conservative feedwater flow ; coastdown model was used in order to bound both the motor-driven and steam turbine I driven feedwater pump designs.
)
15.2.4.3.2 Results l 15.2.4.3.2.1 Closure of All Main Steamline Isolation Valves Figure 15.2-9 shows the changes m important nuclear system variations for the l simultaneous isolation of all main steamlines while the reactor is operating at 102% of 1 NBR power. Neutron flux increases slightly, and fuel surface heat flux shows no (x increase. 1 Four RIPS are tripped due to high pressure. Water level decreases sufIiciently to cause a trip of remaining 6 RIPS and the initiation of the RCIC system on the Level 2 (L2) trip at some time greater than 10 seconds. However, there is a delay up to 30 seconds before ) the water supply enters the vessel. Nevertheless, there is no change in the thermal margins. Therefore, this event does not have to be reanalyzed for specific core configurations. 15.2.4.3.2.2 Closure of One Main Steamline Isolation Valve Only one isolation valve is permitted to be closed at a time for testing purposes to prevent scram. Normal test procedure requires an initial power reduction to approximately 75 to 80% of design conditions in order to avoid high flux scram, high pressure scram, or full isolation from high steam flow in the " live" lines. With a 3 second closure of one MSIV during 102% rated power conditions, the steam flow disturbance may raise vessel pressure and reactor power enough to initiate a high ncucron flux scram. This transient is considerably milder than closure of all MSIVs at full power. No quantitative analysis is furnished for this event. However, no significant change in thermal margins is experienced and no fuel damage occurs. Peak pressure remains m below SRV setpoints. Therefore, this event does not have to be reanalyzed for specific f) core configurations. V increase in Reactor Pressure - Amendment 37 15215
23A6100 Rsv. 4 ABWR StandardSafetyAnalysisReport 15.2.4.4 Barrier Performance 9 t 15.2.4.4.1 Closure of All Main Steamline Isolation Valves The nuclear system relief valves begin to open at approximately 2.9 seconds after the start ofisolation. The valves close sequentially as the stored heat is dissipated but l continue to discharge the decay heat intermittently Peak pressure at the vessel bottom reaches 8.47 MPaG, below the pressure limits of the reactor coolant pressure boundary. Peak pressure in the main steamline is 8.25 MPaG. I 15.2.4.4.2 Closure of One Main Steamline isolation Valve l No signil m effect is imposed on the RCPB, since, if closure of the valve occurs at an unacceptable high operating power level, a flux or pressure scram may result. The main turbine bypass system continues to regulate system pressure via the other three open steamlines. 15.2.4.5 Radiological Consequences I 15.2.4.5.1 General Observations . The radiological impact of transients involves consequences which do not lead to fuel ! rod damage as a direct result of the event itself. Additionally, many events do not lead to the depressudzation of the pdmary system but only the venting of sensible heat and energy via fluids at coolant loop activity through relief valves to the suppression pool. In the case of previously defective fuel rods, a depressurization transient will result in considerably more fission product carryover to the suppression pool than hot-standby ! transients. The time duration of the transient varies from several minutes to more than four hours. These observations lead to the realization that radiological aspects can involve a broad spectrum of results. For example: (1) Transients where appropriate operator action (seconds) results in quick return (minutes) to planned operation, little radiological impact results. (2) Where major RCPB equipment failure requires immediate plant shutdown and its attendant depressurization under controlled shutdown timetables (4 hours), the radiological impact is greater. To envelope the potential radiological impact, a worst case like example No. 2 is described below. However,it should be noted that most transients are like example (1) and the radiological envelope conservatively overpredicts the actual radiological impact by a factor greater than 100. 15.2-16 Increase in Reactor Pressure - Amendment 34
23A6100 REv. 9 ABWR standardsatory Anatvsis Report O I U break. This level of activity is consistent with an offgas release rate of 3.7 GBq/s for Case 1 and 14.8 GBq/s for Case 2 referenced to a 30 minute decay.The iodine concentradon in the reactor coolant is: MBq/g d Case 1 Case 2 ' I-131 0.001739 0.03515 I-132 0.01536 0.30747 I-133 0.01206 0.24161 1-134 0.02634 0.52688 I-135 0.01647 0.3293 Other isotopes of high intrinsic acdvity such as N-16 have been precluded due to their (' extremely short halflives. 15.6.4.5.1.2 Fission Product Transport to the Environment The transport pathway is a direct unfiltered release to the environment. The MSIV detecdon and closure dme of 5.0 s (maximum MSIV closing time and instrument delay) results in a discharge of 12,870 Kg of steam and 21,953 Kg ofliquid from the break. Assuming all the activity in this discharge becomes airborne, the release of acthity to the environment is presented in Table 15.6-6. 15.6.4.5.1.3 Results The calculated exposures for the design basis analysis are presented in Table 15.6 7 and are less than the guidelines of 10CFR100. COL applicants need to update the calculadons to conform to the as<lesigned plant and site-specific parameters (see Subsecdon 15.6.7.2 for COL license information.). 15.6.5 Loss-of-Coolant Accident (Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary)-Inside Containment This event postulates a spectrum of piping breaks inside containment varying in size, type, and locanon. The break type includes steam and/or liquid process system lines. This event is also assumed to be coincident with a safe shutdown earthquake (SSE) for b] / the mechanical design of components. The event has been analyzed quantitativelyin Sections 6.3 (Emergency Core Cooling Systems),6.2 (Containment Systems),7.3 and 7.1 (Instrumentation and Controls), and Decrease in Reactor Coolant inventory- Amendment 37 15.6 7
23A6100 Rect. S ABWR standardsafety Analysis arport O information not presented in the subject sections. All other information is cross-referenced. The postulated event represents the envelope evaluation for liquid or steamline failures inside containment. 15.6.5.1 Identification of Causes and Frequency Classification 15.6.5.1.1 Identification of Causes There are no realistic, identifiable events which would result in a pipe break inside the ; containment of the magnitude required to cause a loss-of-coolant accident coincident I with an SSE. The subject piping is of high quality, designed to construction industry I codes and standards, and for seismic and environmental conditions. However, because such an accident provides an upper limit estimate for the resultant effects for this category of pipe breaks, it is evaluated without the causes being identified. 15.6.5.1.2 Frequency Classification This event is categorized as a limiting fault. 15.6.5.2 Sequence of Events and Systems Operation 15.6.5.2.1 Sequence of Events The sequence of events associated with this accident is presented in Table 6.3-2 for core l system performance. Following the pipe break and scram, the MSIV begins closing on the low level 1.5 signal. The low water level or high drywell pressure signal initiates RCIC, HPCF and RHR flooding systems. 15.6.5.2.2 Identification of Operator Actions Because automatic actuation and operation of the ECCS is a system design basis, no operator actions are required. However, the operator, after assuring that all rods have been inserted, should perform the following: (1) Determine plant conditions by obsening the annunciators. (2) After obsening that the ECCS flows are initiated, check that the diesel generators have started and are on standby condition and confirm that the Senice Water System is operating in the LOCA mode. (3) After the RHR System and other auxiliary systems are in proper operation, the operator should periodically monitor the oxygen concentration in the drywell - and wetwell. 15.6-8 Decrease in Reactor Coolant inventory - Amendment 35
23A6100 Rsv. 2 ABWR standardsatoryAnalysis anort O through dynamic analysis to sunive such events, whereas in the case of the actual experience database, the lines shown to sunive were designed to lesser standards to meet only normally expected loads. Therefore, based upon the facts above, the main steamlines and drain lines in the l ABWR are used as mitigative components in the analysis ofleakage from the MSIVs. l The analysis ofleakage from the MSIVs follows the procedures and conditions specified l in Reference 15.6-4. Two flow paths are analyzed for dose contribations. The first pathway through the drain lines is expected to dominate because of the incorporation of a safety grade isolation valve on the outboard drain line which will open the line for flow down the drain line under LOCA conditions. The second pathway through the main steamlines into the turbine is expected to carryless than 0.3% of the flow based i upon a determination that the maximum leakage past the turbine stop valves with an open drain line would permit only 0.3% flow for the valves to operate within !
- specification. Specific values used and results of the main steamline leakage analysis are given in Table 15.6-8.
j The COL applicant will recalculate iodine removal credit on the basis ofits design characteristics of main steamlines, drain, and main condenser. See Subsection 15.6.7.1 ' V for COL license information requirements. J 15.6.5.5.1.3 Condenser and Turbine Modeling l The condenser and turbine are modeled as detailed in Reference 15.6-4 with specific l i values used given in Table 15.6-8. Both volumes are modeled primarily as stagnant volumes, assuming the shutdown of all active components. Both turbine and condenser are used as mitigative volumes based upon the determination that such components designed to standard engineering practice are sufficiently strong to withstand SSE l conditions due wholly to their design (Reference 15.6-4). The only requirement in the
- design of the condenser is that it be bolted to the building basemat to prevent walking during an earthquake. The turbine has no such restriction and may possibly move. The requirement on these components for purposes of mitigation is only that they sunive as a volume and not that they provide functionality or leaktightness following an earthquake.
Release from the condenser / Turbine Building pathway is assumed via diffuse sources in the Turbine Building. The two major points of release in the Turbine Building are expected to be the truck doors at the far end of the Turbine Building and the maintenance panels located midway on the Turbine Building on the side opposite the service building. Releases are assumed to be ground level releases. See t i Subsection 15.6.5.5.3 for applicable meteorology. b Decrease in Reactor Coolant Inventory- Am . Iment32 15.6-15
l 23A6100 Rtv. 9 ABWR StandardSafety Analysis Report O The COL applicant will recalculate iodine removal credit on the basis ofits design characteristics of main steamlines, drain, and main condenser. See Subsection 15.6.7.1 for COL license information requirements. 15.6.5.5.2 Contr01 Room The ABWR control room is physicallyintegrated with the Reactor Building and Turbine Buildings and is located between these structures (Figure 15.64). During a LOCA, exposure to the operators will consist of contributions from airborne fission products entrained into the control room ventilation system and gamma shine from the Reactor Building and airborne fission products external to the Control Building. Of these contributions, the last two invohing gamma shine are negligible, since the inhabited portions of the ABWR control room are physically located underground with sufficient shielding overhead (a minimum of1.6 meters of concrete) and in the side walls (1.2 meters) to protect the operators from the normal steamline gamma shine. Such l shielding is more than sufficient to protect the operators given any amount of airborne fission products. l Therefore, exposure to the operators will consist almost entirely of fission products entrained into the control room emironment from the atmosphere.The ABWR control room uses a redundant safety grade HVAC System with 100 mm (four inch) charcoal l filters for removal ofiodines and two wall-mounted automatically controlled intake I vents. The locations of the vents are given in Figure 15.6-4. Because of the location of l these vents, it cannot be assumed that at least one vent will be uncontaminated, given most conditions of meteorology. Therefore, no credit for dual intakes was taken. In l addition, the location of these vents with respect to the potential release points shows that, given any wind flow condition, the vents may be contaminated only by a release from the Reactor Building orTurbine Building but not both. Nevertheless, for purposes of consenative calculations, it was arbitrarily assumed that for 30% of the time stagnant meteorological conditions were assumed such that the primary intake vent was contaminated by both sources.1 or the remaining 70% of the time, only the more significant source was assumed to contaminate the primary intake vent. Infiltration of airborne contamination to the control room was considered negligible, owing to the pathway for access to the control room complex. Entry into the control room is via the Senice Building and a labyrinth doorway entry system through double doors into the clean portions of the Service Building. From the Senice Building, l additional controlled access through double doors provides entry into the control room. In each of these enny/ access door systems, positive pressure is maintained to vent infiltrated air to the outside and away from the control room complex. As such, no j contamination is anticipated beyond the initial access entry way from which infiltrating air is purged to the emironment. l l 15.6-16 Decrease in Reactor Coolant Inventory- Amendment 37
I l 23A6100 Rtv. 4 i ABWR senadardsaresyAnalysis separt I Control room dose is based upon fission product releases modeled as described in Subsection 15.6.5.5.1 and the values presented in Table 15.6-8. Operator exposure was l based upon those conditions given in Table 15.6-8 and occupancy factors as shown j below derived from SRP 6.4. Meteorology was derived as is specified in l Subsection 15.6.5.5.3.2. l I Time Occupancy Factor 0-1 day 1.0 ! 1-4 days 0.6 i
>4 days 0.4 l i
15.6.5.5.3 Meteorology j 15.6.5.5.3.1 Offsite Meteorology j i The SSAR involves the use of a generic U.S. site which does not specifically identify l meteorological parameters adequate to define dispenion conditions for accident O evaluation. Therefore, for the evaluation of offsite accident conditions, recourse was k made to Regulatog Guides 1.145 and 1.3 for meteorological definitions. Specifically, the table found in Section C.2.g(4) of Regulatog Guide 1.3 was used to define the meteorological parameters for use with the models found in Regulatory Guide 1.145. All releases were defined as ground level incorporating building wake conditions using the minimum ABWR building cross section. Unlike the other design basis accidents found in Chapter 15, the LOCA accident analysis requires the development of meteorological conditions over a 30 day period. To develop a bounding 30 day set of four x/Q dispersion parameters, recourse was made to Regulatog Guid 1.3 and the metrological prescription found under Subsection 2.g. From this prescription, the x/Qvalues for 30 days were " walked"in from a 4828 m LPZ to approximately 1140 meters where the 30 day thyroid dose became 3 Sv. By plotting these resulting four X/Qvalues on log-log paper a straight line curve was established from which a 2-hour 95% LPZ X/Qand an nual average X/Qvalue were back fitted with a small factor of conservatism in the derivation so that the resultant integrated dose was less than 300 Rem. The resultant straight line plot and x/Qvalues are shown in Figure 15.6-6. The end points are the 95% 2-hour ISZ X/Q of 4.11E-04 and annual average (8760 hour) X/Qof1.17E-06 from which the intermediate values given in Table 15.613 (shown as Chp 2 values) were derived as specified in Regulatog Guide 1.145. Decrease in Reactor Coolant Inventory - Amendment 34 15.6 17
l l 23A6900 Rev. 9 l ABWR StandardSafetyAnalysis Report 15.6.5.5.3.2 Control Room Meteorology No specific acceptable method exists to calculate the meteorology for standard plant application for control room dose analysis. Unlike the offsite dose methodology, which is a relatively straight forward application of Regulatory Guides 1.3 and 1.145, the parameters and rnethods by which the control room intake concentrations can be calculated are poorly characterized and currently not codified in a usable form. I Therefore, for application to the ABWR, a back<alculation was used to provide an l estimate of the meteorological x/Qdispersion parameters which would provide for the maximum acceptalite dose under SRP 6.4. Since the calculation covers a period of 30 days, a variation in meteorological x/Qwas assumed for variations in wind direction I and wind speed. The variation factors chosen were taken from Table 1 of Reference 15.63 and are shown below. 1 Murphy-Campe X/Q Time Period Improvement Factor 0-8 hours 1.0 l 8-24 hours 0.59 l-4 days 0.375
> 4 days 0.165 Also, since the control room may be contaminated from two physically separated sources, the Reactor Building stack base or the Turbine Building truck doors, reference was made to the most recently published work of Ramsdell to evaluate the differences in x/Q for releases from each source to the control building. Using the methodology l given in References 15.65 and 15.6-6, it was determined that ieleases from the Turbme l Building at 108 meters from the control room intake would be a factor of six lower in concentration for an equal release than releases from the Reactor Building stack base at 41 meters from the nearest Control Building intake. Therefore, a factor of six improvement in x/Qwas assumed for releases from the Turbine Building.
For application to specific site analysis, two methods exist for determination of control room dose. The first method is a one-on-one comparison of the x/Q values in Table 15.414 to the site x/Qs. If the site x/Qs are for all values less than the values in Table 15.414, then the control room doses are less than regulatory requirements. If this is not true, then a site specific calculation needs to be performed for the site. For this l purpose, an isotope-by-isotope release rate table is given in Tables 15.610 and 15.412, from which actual calculations can be made. l 15.6 18 Decrease in Reactor Coolant Inventory - Amendment 37
23A6100 Rev. 4 ABWR standard safetyAnalysis Report Table 15.6-2 Instrument Line Break Accident Isotopic Inventory l Reactor Building Inventory (Megabecquerel) l lsotope 1- min 10-min 1 hour 2-hour 4-hour 8-hour l l l-131 3.77E+01 3.27E+02 2.60E+04 1.73E+04 1.38E+04 4.59E+00 l l-132 3.68E+02 3.11 E+03 2.31 E+05 1.44E+05 1.17E+05 1.17E+01 l l-133 2.59E+02 2.24E+03 1.75E+05 1.16E+05 9.29E+04 2.72E+01 ) l l-134 7.22E+02 5.92E+03 3.89E+05 2.26E+05 1.86E+05 2.65E+00 l-135 3.77E+02 3.25E+03 2.52E+05 1.64E+05 1.32E+05 2.90E+01 l l l [ Total 1.76E+03 1.48E+04 1.07E+06 6.68E+05 5.41 E+05 7.52E+01 l l Isotopic Release to Environment (Megabecquerel) l l Isotope 1 min 10-min 1-hour 2-hour 4-hour 8-hour l l l-131 6.36E-01 5.77E+01 2.77E+04 6.81 E+04 1.27E+05 1.41E+05 l l l-132 6.18E+00 5.51E+02 2.52E+05 5.96D05 1.09E+06 1.19E+06 l l-133 4.37E+00 3.96E+02 1.87E+05 4.59E+05 8.51 E +05 9.44E+05 l l-134 1.21E+01 1.06E+03 4.44E+05 9.92E+05 1.76E+06 1.90E+06 l l-135 6.36E+00 5.74E+02 2.71E+05 6.59E+05 1.21 E+06 1.34E+06 l Total 2.97E+01 2.64E+03 1.18E+06 2.77E+06 5.04E+06 5.51E+06 O Decrease in Reactor Coolant Inventory - Amendment 34 15.6-25
23A6100 Riv. 9 ABWR Standard Safety Analysis Report O Table 15.6-3 Instrument Line Break Accident Results Meteorology
- and Dose Results Meteorology Distance Thyroid Dose Whole Body Dose (s/m3) (m) (Sv) (Sv) 8.59E-03 max 3.0E-01 6.0E-03 1.37E-03 Chp2 4.8E-02 9.4E-04 2.19E-04 800 7.6E-03 1.5E-04 1.11E-04 1600 3.9E-03 7.9E-05 5.61E-05 3200 2.0E-03 4.0E-05 3.73E-05 4800 1.3E-03 2.6E-05
- Meteorology calculated using Regulatory Guide 1.145 for a ground level 1.0 m/s, F stability release.
" Max" = maximum meteorology to meet 10% of 10CFR100 limits.
Table 15.6-4 Sequence of Events for Steamline Break Outside Containment Time (s) Event 0 Guillotine break of one main steamline outside primary containment.
~0.5 High steamline flow signalinitiates closure of main steamline isolation valve
<1.0 Reactor begins scram.
l 505 Main steamline isolation valves fully closed. 38 Safety / relief valves open on high vessel pressure. The valves open and close to maintain vessel pressure at approximately 7.58 MPa. 30 RCIC initiates on vessel low water Level 2. 50 RCIC begins injection. 199 HPCF initiates on low suasr level. 236 One HPCF begins injecdon (the other HPCF is unavailable due to the single failure assumption). 1-2 hours Normal reactor cooldown procedure established. 1S.6-26 Decrease in Reactor Coolant Inventory - Amendment 37 ~.
. . ~ . . . - _ . - . _ -. .- --
23A6100 Rev. 9 ABWR StandardSafetyAnalysis Report 1" t x Table 15.6-5 Steamline Break Accident Parameters l Data and assumptions used to estimate source terms. A. Power Level 4005 MWt B. Fuel damage none C. Reactor coolant activity Subsection 15.6.4.5 D. Steam mass released 12,870 kg E. Water mass released 21,953 kg ll Data and assumptions used to estimate activity released l A. MISV closure time (break until 5.0 s fully closed) B. Maximum release time 2h 111 Dispersion and Dose Data A. Meteorology Table 15.6-7 B. Boundary and LPZ distances Table 15.6-7 C. Method of Dose Calculation Reference 15.6-2 D. Dose conversion Assumptions Reference 15.6 2, RG 1.109, and ICRP 30 ' p E. Activityinventory/ release Table 15.6-6 t, F. Dose Evnluations Table 15.6-7 l l l Decrease in Reactor Coolant inventory - Amendment 37 15.6 27
23A6100 Rev. 4 ABWR StandardSafetyAnalysis Report O Table 15.6-6 Main Steamline Break Accident Activity l Released to Environment (Megabecquerel) Isotope Case 1 Case 2 l l-131 7.29E+04 1.46E+06 l l-132 7.10E+05 1.42E+07 . l l-133 5.00E+05 9.99E+06 l l-134 1.40E+06 2.79E+07 l l-135 7.29E+05 1.46E+07 l l Total Halogens 3.41E+06 6.81E+07 l KR-83M 4.07E+02 2.44E+03 l l KR-85M 7.18E+02 4.2SE+03 l KR-85 2.26E+00 1.36E+01 l KR-87 2.44E+03 1.47E+04 l KR-88 2.46E+03 1.48E+04 l KR-89 9.88E+03 5.92E+04 l KR-90 2.55E+03 1.55E+04 l XE-131M 1.76E+00 1.06E+01 l XE-133M 3.39E+01 2.04E+02 l XE-133 9.47E+02 5.70E+03 l XE-135M 2.89E+03 1.74E+04 l XE-135 2.70E+03 1.62E+04 l XE-137 1.23E+04 7.40E+04 l XE-138 9.44E+03 5.66E+04 l XE-139 4.33E+03 2.59E+04 l Total Noble Gases 5.11 E+04 3.07E +05 0 15.6-28 Decrease in Reactor Coolant Inventory - Amendment 34
23A6100 Rsv. 9 ABWR standardSafety Analysis Report n Table 15.6-8 Loss of Coolant Accident Parameters (Continued) E. Condenser data Free Air Volume 6230 m3 Fraction of Volume involved 20% Leakage Rate 11.6%/d lodine Removal Factor Elemental 0.993 Particulate 0.993 Organic 0 111 Control Room Data A. Control Room Volumes Total Free Air Volume 5,509 m 3 Gamma Room Volume (room size) 1,400 m 3 3 circulation Rates l Filtered intake 0.944 m3/s Unfiltered Intake 0.0 Filtered Recirculation 0.47 m3/s Filter Efficiency (100mm) 99% f" IV Dispersion artd Dose Datn v A. Meteorology Sec 15.6.5.5.3 Tbis 15.6-13,15.6-14 B. Dose Calculation Method (semi-infinite) Ref 15.6-2 & 15.6-3, RG 1.109 C. Dose Conversion Assumptions Ref 15.6-2,15.6-3 D. Activity / Releases Tbis 15.6-9,15.6-10,15.6-11,15.6-12 Appendix 15F E. Dose Evaluation Tbis 15.6-13,15.6-14
/
I.
\
Decrease in Reactor Coolant Inventory - Amendment 37 15.6-3 1
5 Table 15.6-9 lodine Activities b 2 03 l" Isotope 1 min 10 min 1h l2 h 4h 8h 12 h 1 day 4 days 30 days l A. Primary Containment Inventory (megabecquerel) 1-131 5.2E+11 5.2E+11 5.2E+11 5.2E+11 4.8E+11 4.8E+11 4.8E+11 4.4E+11 3.4E+11 2.7E+10 1-132 7.4E+11 7.0E+11 5.6E+11 4.1 E+11 2.2E+12 6.7E+10 1.9E+10 5.2E+8 1.6E+11 0 1-133 1.1 E+12 1.0E+12 1.0E+12 1.0E+12 9.3E+11 8.1 E+11 7.0E+11 4.8E+11 4.1 E+10 2.8E+1 1-134 1.1 E+12 1.0E+12 5.2E+11 2.4E+11 4.8E+10 2.1E+9 8.9E+7 6.7E+3 0 0 1-135 1.0E +12 1.0E+12 8.9E +11 8.1E+11 6.7E+11 4.4E+11 2.8E+11 7.8E+10 4.1E+7 0 l Total 4.5E+12 4.3E+12 3.5E+12 3.0E+12 4.3E+12 1.8E+ 12 1.5E+12 1.0E+12 5.4E+11 2.7E-.10 l B. Reactor Building inventory (megabecquerel) 1-131 1.7E+6 1.5E+7 9.3E+7 1.9E+8 3.7E+8 7.0E+8 9.6E+8 1.4E+9 1.7E+9 1.3E+8 l-132 2.5E+6 2.1 E+7 1.0E+8 1.6E &8 1.7E+8 9.3E+7 3.7E+7 1.6E+6 7.8E--4 0 l-133 3.6E+ 6 3.1E+7 1.9E+8 3.7E+8 7.0E+8 1.1 E+9 1.4E+9 1.5E+9 2.0E+8 1.4E-1 1-134 4.1E4 6 3.( E+7 1.0E+8 9.3E+7 3.7E+7 2.9E+6 1.7E+ 5 2.1E+1 0 0 1-135 3.4E+6 2.9 3+7 1.7E+8 3.1E+8 4.8E+8 5.9E+8 5.6E+8 2.5E+ 8 1.9E+5 0 $ m l Total 1.5E+7 1.31 +8 6.5E+8 1.1 E+9 1.8E+9 2.5E+9 2.9E+9 3.2E+9 1.9E+9 1.3E+8 @ [o C.1 MSIV Pathway-Condenser inventory (megabecquerel)-Elemental lodine f
! l-131 0 0 7.8E+ 6 8.5E+6 4.8E+7 2.0E+8 8.9E+7 6.3E+8 8.5E+7 1.1 E+9 2.4E+9 2.6E+6 4.1 E+9 3.1E+7 y 1-132 0 0 4.1E+7 4.4E+7 1.9E-3 0 s- I-133 0 0 1.6E+7 9.6E+7 3.7E+8 1.0E+0 1.6E+9 2.4E+9 4.8E+8 3.3E-2 & I-134 0 0 8.1 E+ 6 2.4E+7 2.0E+7 2.7E+6 2.0E +5 3.5E+ 1 0 0 '
E l-135 0 0 1.4E+7 8.1 E+7 2.7E+8 5.6E+8 6.3E+8 4.1E+8 4.8E+5 0 4 lg Total 0 0 5.4E+7 2.9E+ 8 9.5E+8 2.3E+9 3.4E+9 5.2E+9 4.6E+9 3.1 E+7 w (( C.2 MSIV Pathway-Condenser Inventory (megabecquerel)-Organic lodine (Primary Containment) { [g I-131 I-132 0 0 0 0 6.7E+5 7.0E+5 4.1E+6 3.4E+6 1.7E+7 7.8E+6 5.6E+7 7.0E+6 9.3E+7 3.7E+6 2.1E+8 2.3E+5 5.9E+8 2.8E-4 1.3E+8 0 E. Q I-133 0 0 1.4E+ 6 8.1E+6 3.2E+7 8.9E+7 1.4E+8 2.1E+10 7.0E+7 1.4E-1 l 1-134 0 0 7.0E+5 2.0E+6 1.7E+6 2.3E+5 1.7E+4 3.0E+0 0 0 { 5 I l-135 0 0 1.2E+6 6.7E+6 2.3E+7 4.8E+7 5.6E+7 3.6E+8 6.7E+4 0 $ 8 % lu2 Total 0 0 4.6E+ 6 2.4E+7 8.1 E+7 2.0E+8 2.9E+8 2.2E+10 6.6E+8 1.3E+8 W m a e t E O O O
p Table 15.6-9 lodine Activities (Continued) b : ID j' isotope 1 min 10 min 1h 2h 4h 8h 12 h 1 day 4 days 30 days g g C.3 MSIV Pathway-Condenser inventory in Curies-Resuspended Organic 23 k l-131 0 0 2.8E+3 5.6E+3 3.4E+4 8.9E+4 2.7E+5 9.3E +5 4.8E+7 9.6E+7 St 1-132 0 0 2.2E+3 3.7E+3 8.5E+3 1.2E+4 5.9E+3 1.7E+3 0 0
& l-133 0 0 5.6E+3 1.1 E+4 5.9E+4 1.5E+ 5 3.6E+5 9.3E+5 5.9E+6 5.2E-1 0 1-134 0 0 1.4E+3 2.1E+3 1.3E+3 1.0E+3 5.6E+1 1.2E+ 0 0 0 h l-135 0 0 4.4E+3 8.5E+3 3.6E+4 7.4E+4 1.1 E+5 1.6E+5 9.6E+3 0 Total 0 0 1.6E+4 3.1E+4 1.4E+5 3.2E+5 7.5E+ 5 2.0E + 6 5.4E+7 9.6E+7 ;
j C.4 Condenser Inventory (megabecquerel)-Combined ! g 1-131 0 0 8.5E+6 5.6E+7 2.2E+8 7.0E+8 1.2E+9 2.6E+9 4.8E+9 2.6E+8 l g 1-132 0 0 9.3E+6 4.4E+7 9.6E+7 9.3E+7 4.8E+7 2.8E+6 2.2E-3 0 y 1-133 0 0 1.7E+7 1.1 E+8 4.1 E+8 1.1 E+9 1.7E+9 2.7E+9 5.6E+8 7.0E-1 3, 1-134 0 0 8.9E+6 2.6E+7 2.2E+7 2.9E+6 2.2E+5 4.1E+1 0 0 ; y 1-135 0 0 1.5E+7 8.5E+7 2.9E+8 5.9E+8 7.0E+8 4.4E+8 5.6E+5 0 g Total 0 0 5.9E+7 3.2E+8 1.0E+9 2.5E+9 3.7E+9 5.7E+9 5.4E+9 2.6E+8 D.1 Control Room Inventory (megabecquerel) {
?
l-131 8.4E-1 7.3E+1 1.4E+2 5.8E+ 1 2.0E+ 1 2.6E+1 2.2E+1 3.6E+1 3.2E+1 1.9E+0 1-132 1.2E+0 1.0E+2 1.5E+2 4.6E+1 8.7E+0 3.4E + 0 8.9E-1 4.0E-2 0 0 l I-133 1.8E +0 1.5E+2 2.8E+2 1.1 E+2 3.7E+1 4.2E+1 3.2E+1 3.7E+1 3.8E+0 3.8E-9 l-134 1.9E+0 1.5E+2 1.4E+2 2.8E +1 2.0E+0 1.1E-1 4.0E-3 5.4E-7 0 0 1-135 1.7E+0 1.4E+2 2.4E+2 9.3E+1 2.6E+1 2.2E+1 1.3E+1 6.3E+0 3.6E-3 0 l Total 7.4E+ 0 6.2E+2 9.5E+2 3.4E+ 2 9.3E+12 9.3E+1 6.8E+1 8.0E+1 3.5E+1 1.9E+0 l D.2 Control Room Integrated Activity (megabecquerel-seconds) f 1-131 1.7E+1 1.5E+4 5.4E+5 3.3E+5 2.3E+ 5 3.0E+ 5 l 3.1E+5 1.3E+ 6 8.0E+6 1.5E+7 k l-132 2.5E+1 2.1E+4 6.7E+5 3.1E+5 1.5E+5 7.4E+4 2.4E+4 1.2E+4 0 0 I-133 3.5E+1 3.2E+4 1.1E+6 6.5E+ 5 4.4E+5 5.2E+5 4.8E+5 1.6E+ 6 3.2E +6 2.0E+ 5 h iig-l-134 3.9E+1 3.2E+4 8.1 E +5 2.5E + 5 6.4E+4 8.4E+3 4.1 E +2 2.0E+1 0 0 l-135 3.3E+1 3.0E+4 1.0E +6 5.6E+5 3.4E+5 3.2E +5 2.2E +5 4.1 E+5 1.7E+ 5 0 {g l Total 1.5E+2 1.3E+ 5 4.1 E+ 6 2.1E+ 6 1.2E+6 1.2E+ 6 1.0E+6 3.3E+6 1.1 E+7 1.5E+7 f G P 8 2 u a
s Table 15.6-10 lodine Activity Release to the Environment b T, tti isotope 1 min 10 min 1h 2h 4h 8h 12 h 1 day 4 days 30 days A. Release from Reactor Building to Environment (megabecquerel) 1-131 2.9E+4 2.6E+ 6 9.6E+6 9.6E+6 1.0E+7 1.3E+7 1.7E+7 3.6E+7 1.9E+8 6.7E+8 l-132 4.1 E+4 3.7E+6 1.3E+7 1.3E +7 1.4E +7 1.4E+7 1.5E+7 1.5E+7 1.5E+7 1.5E+7 l-133 5.9E +4 5.6E+6 2.0E +7 2.0E+7 2.1 E+7 2.6E+7 3.3E+7 5.6E+7 1.2E+8 1.3E+8 l-134 6.7E+4 5.6E+ 6 1.9E+7 1.9E+7 1.9E+7 1.9E+7 1.9E +7 1.9E+7 1.9E +7 1.9E+7 l l-135 5.6E+4 5.2E+6 1.9E+7 1.9E+7 2.0E+7 2.3E+7 2.6E+7 3.1 E+7 3.5E +7 3.5E+7 Total 2.5E+ 5 2.3E+7 8.0E+7 8.1E+7 8.4E+7 9.5E+7 1.1 E+8 1.6E+8 7.0E+8 8.6E+8 B.1 MSIV Pathway Release to Environment-Elemental (megabecquerel) 1-131 0 0 5.6E+1 9.3E +2 9.3E+3 6.3E+4 1.8E+ 5 8.9E + 5 1.0E +7 3.0E +7 l-132 0 0 6.3E+ 1 8.5E+ 2 5.6E+3 1.8E+4 2.7E +4 3.3E + 4 3.4E + 4 3.4E+4 l-133 0 0 1.1E+2 1.9E+3 1.7E+ 4 1.1 E+5 3.0E+ 5 1.1 E +6 4.8E+ 6 5.6E+6 l-134 0 0 6.3E+ 1 6.3E+2 2.3E+3 3.6E+3 3.7E +3 3.7E + 3 1.3E+ 2 3.7E+3 I-135 0 0 1.0E+2 1.6E+3 1.3E+4 7.0E +4 1.6E+ 5 3.7E+ 5 5.6E+5 5.6E+5 c Total 0 0 3.9E+ 2 5.9E+3 4.8E+4 2.7E+ 5 6.6E+ 5 2.4E+ 6 1.6E +7 3.6E +7 5 g B.2 MSIV Pathway Release to Environment-Organic (megabecquerel) [
$ l-131 0 0 6.7E+2 1.1E+4 1.1E+ 5 7.8E +5 2.2E+6 1.1 E+7 1.6E + 8 1.4E +9 y I-132 0 0 7.4E + 2 1.0E+4 6.7E +4 2.2E+ 5 3.3E + 5 4.1 E+ 5 4.1 E + 5 4.1E+ 5 s- l-133 0 0 1.3E+3 2.3E+4 2.1E+5 1.4E +6 3.6E + 6 1.4E +7 7.0E+7 8.1 E +7 i? l-134 0 0 7.8E+2 7.8E +3 2.8E+4 4.4E +4 4.4E +4 4.4E+4 4.4E +4 4.4E +4 E l-135 0 0 1.2E+3 1.9E+4 1.6E+ 5 8.5E+ 5 1.9E +6 4.4E + 6 7.0E + 6 7.0E + 6 R -
n Total 0 0 4.7E+3 7.1 E+4 5.8E+ 5 3.3E+6 8.0E+6 3.0E +7 2.4E+ 8 1.5E+ 9 U3 MSIV Pathway Release to Environment-Resuspended Organic (megabecquerel) [ A l-131 l-132 0 0 0 0 6.7E+0 5.6E+0 2.7E+1 2.1E+1 2.2E+2 8.1 E+1 1.4E+3 2.9E+2 4.8E+3 4.4E +2 3.7E+4 6.7E+2 6.3E+ 6 7.0E+2 5.2E +8 7.0E+2 It. M l-133 0 0 1.3E+1 5.2E+1 4.1E+2 2.4E+3 7.4E+3 4.4E+4 1.4E+6 3.3E+6 l 1-134 0 0 4.1E+ 0 1.3E+1 2.8E+1 4.8E+1 5.6E+1 5.6E+1 5.6E+1 5.6E+1 {5
$ l-135 0 0 1.1E+1 4.1 E+1 2.6E +2 1.4E+3 3.2E+3 1.1E +4 3.4E+4 3.5E +4 E 8
a Total 0 0 4.1E+1 1.5E+2 1.0E+3 5.5E+3 1.6E +4 9.4E +4 7.7E+ 6 5.2E+8
$Cr $. 2=
0 0 0
t O O P Table 1S.6-11 Noble Gas Activities (Continued) b isotope 1 min 10 min 1h 2h 4h 8h 12 h 1 day 4 days 30 days
. E
[ C. Condenser inventory (megabecquerel) 23
$ Kr-83m 0 0 5.6E+6 2.4E +7 4.8E+7 3.3E+7 1.3E +7 3.5E+ 5 0 0 R Kr-85 0 0 7.8E +5 4.8E+6 2.0E+7 6.3E+7 1.1 E+8 2.6E+8 9.6E+8 2.0E+9 9 Kr-85m 0 0 1.4E+7 7.8E+7 2.4E+8 4.1 E+ 8 3.7E+ 8 1.3E+8 5.6E+3 0 E Kr-87 0 0 1.9E+7 7.0E +7 9.6E+7 3.4E+7 6.7E+6 2.2E+4 0 0 '
i h Kr-88 0 0 3.6E+7 1.8E+ 8 4.4E+8 5.2E+8 3.5E +8 4.1 E+7 2.7E+0 0
;5 Kr-89 0 0 1.2E+2 1.5E-3 0 0 0 0 0 0 ; $ Xe-131m 0 0 4.1E+5 2.6E +6 1.0E+7 3.3E+7 5.6E+7 1.3E+8 4.1 E+8 1.9E+8 4 Xe-133 0 0 1.4E+ 8 8.9E+8 3.6E+9 1.1E410 1.9E+10 4.1 E+10 1.0E +11 7.4E+9
[ Xe-133m Xe-135 0 0 0 0 5.9E+6 1.7E+7 3.6E+7 1.0E+8 1.4E+8
, 3.5E +8 4.4E+8 8.1E+8 7.4E+ 8 1.1 E+9 1.4E+9 1.0E+9 2.1 E+9 1.6E+7 1.6E+ 6 1.0E-13 3 & Xe-135m 0 0 1.9E4 6 8.5E+ 5 1.7E+4 1.4E+ 0 6.3E-5 0 0 0 2 Xe-137 0 0 2.3E+3 2.7E-1 0 0 0 0 0 0 E Xe-138 0 0 6.3E+6 2.1E+6 2.5E+4 6.3E-1 0 0 0 0 %
Total 0 0.0E+0 2.5E+8 1.4E+9 4.9E +9 1.3E+10 2.2E+10 4.4E+ 10 1.1E+11 9.6E+9 D.1 Control Room Inventory (megabecquerel) Kr-83m 7.7E+1 6.4E+3 1.3E+4 1.1E+4 1.0E+4 5.4E+3 1.1 E+3 2.2E+1 0 0 $ Kr-85 7.5E+0 6.6E+2 1.9E+3 2.2E+3 4.5E+3 1.0E+4 9.6E+3 1.6E +4 1.7E+4 6.4E+3 , Kr-85m 1.7E+2 1.4E+4 3.6E+4 3.6E+4 5.3E+4 6.5E+4 3.2E+4 8.1 E+3 9.8E-2 0 Kr-87 3.2E+2 2.6E+4 4.7E+4 3.2E+4 2.2E+4 5.6E+3 5.8E+2 1.4E+0 0 0 Kr-88 4.5E+2 3.8E+4 8.9E+4 8.2E+4 1.0E+ 5 8.6E +4 3.0E +4 2.6E+3 4.7E-5 0 Kr-89 4.5E+2 5.6E+3 2.9E-1 7.0E-7 0 0 0 0 0 0 t Xe-131m 3.9E+0 3.4E+2 9.9E+2 1.2E+3 2.3E+3 5.3E+3 4.9E+3 8.0E+3 7.1 E+3 5.9E+2 Xe-133 1.4E+3 1.2E+5 3.4E+5 4.0E+5 8.0E+5 1.8E+6 1.6E+ 6 2.6E+ 6 1.8E+ 6 2.3E+4 g Xe-133m S.7E+1 5.0E+3 1.4E+4 1.7E+4 3.3E+4 7.1E+4 6.3E+4 9.1E+4 3.8E+4 4.9E+0 = Xe-135 Xe-135m 1.8E + 2 2.5E4 2 1.5E+4 1.5E+4 4.1E+4 4.6E+3 4.5E+4 3.8E+2 7.8E+4 3.9E +0 1.3E+5 2.2E-4 9.2E+4 5.1 E-9 6.2E+4 0 2.8E+2 0 0 0 fp Xe-137 1.0E+3 1.7E+4 5.6E+0 1.2E-4 0 0 0 0 0 0 g Xe-138 1.1 E+3 6.2E +4 1.5E +4 9.7E +2 5.6E+0 1.0E-4 7.8E-10 0 0 0 4 D l Total 5.4E+3 3.2E+5 6.1E+5 6.3E+5 1.1 E+6 2.2E+6 1.9E+6 2.8E+ 6 1.9E+6 3.0E+4 g W 5 @ 8
~
ltu
;;; Table 15.6-11 Noble Gas Activities (Continued) ? b M isotope 1 min 10 min 1h 2h 4h 8h 12 h 1 day 4 days tO D.2 Control Room Integrated Inventory (megabecquerel-seconds) 30 days g
l 2ll Kr-83m 1.6E+3 1.3E +6 4.8E +7 4.2E +7 7.8E+7 1.2E + 8 3.7E +7 1.3E+7 0 0 Kr-85 1.5E+2 1.4E+5 5.7E+6 7.1E+ 6 2.3E+7 1.1E+8 1.3E +8 5.6E+ 8 3.9E+9 1.6E+10 Kr-85m 3.3E+3 3.0E +6 1.2E+8 1.3E+8 3.2E+8 9.0E+8 6.3E+8 7.8E+8 1.5E+8 0 Kr-87 6.4E+3 5.4E+6 1.8E+8 1.4 E+8 1.9E+8 1.8E +8 3.0E+7 4.2E+6 0 0 Kr-88 9.1 E+3 8.0E +6 3.0E+8 3.0E+8 6.6E+8 1.4E+9 7.1 E+8 4.9E+8 3.0E+7 0 Kr-89 9.6E+3 2.2E+6 2.9E+6 8.0E+1 0 0 0 0 0 0 Xe-131m 7.9E+1 7.1E+4 3.0E+ 6 3.7E+6 1.2E+7 5.5E+7 6.8E+7 2.8E+ 8 1.8E+ 9 3.6E+9 Xe-133 2.7E+4 2.5E+7 1.0E + 9 1.3E+9 4.2E+9 1.9E+10 2.3E+10 9.3E+10 5.2E+11 5.2E+11 Xe-133m 1.1E+3 1.0E+ 6 4.3E +7 5.3E+7 1.7E+8 7.5E + 8 9.0E+8 3.4E+S 1.4E+10 4.9E4 9 Xe-135 3.5E+3 3.2E+6 1.3E+8 1.5E+8 4.4E+8 1.6E +9 1.5E +9 3.4E+9 2.4E+9 0 Xe-135m 5.0E+3 3.4E+6 4.8E+7 5.9E+6 5.9E+5 5.9E+3 2.9E-1 0 0 0 Xe-137 2.1 E+4 6.0E+6 1.2E+7 1.8E+3 0 0 0 0 0 0 Xe-138 2.2E+4 1.4E+7 1.9E+8 1.8E+7 1.3E+ 6 7.6E+3 1.2E-1 0 0 0 % l Total 1.1E+5 7.3E+7 2.1E+9 2.1E+9 6.1E+9 2.4E+10 2.7E+10 1.0E+11 5.4E+11 5.5E+11 k 5 a & s 5-D a R 9 E E 5-4 a. 0 U a w l 5 b k 2, 2 g n. e b
~ Y=
0 0 0
i v v fd
? Table 15.6-12 Noble Gas Activity Release to Environment Ih a cs j Isotope l1 min l10 min l1 h l2 h l4 h 8h l12 h 1 day l4 days 30 days g A. Reactor Building Release to Environment (megabecquerel) N k Kr-83m 2.7E+4 2.3E+6 9.3E+6 1.2E+7 1.9E+7 2.8E+7 3.2E+7 3.3E+7 3.3E+7 3.3E+7 2.6E +3 2.3E +5 1.0E+ 6 1.5E+ 6 3.6E+ 6 1.2E+7 2.4E+7 8.1E+7 6.7E+8 5.6E+9 A Kr-85 Kr-85m 5.6E+4 5.2E+6 2.1 E+7 3.1E+7 5.9E+7 1.3E+8 1.9E+8 2.7E+8 2.9E+8 2.9E+8 l{8.
l Kr-87 m Kr-88 1.1E+ 5 1.6E+ 5 9.3E+6 1.4E +7 3.6E+7 5.6E+7 4.4E+7 7.8E+7 6.3E+7 1.4E+8 7.8E+7 2.5E+8 8.1E +7 3.1 E + 8 8.1 E+7 3.6E+8 8.1 E+7 3.7E+8 8.1E+7 3.7E+8 B Kr-89 1.7E+5 4.8E+ 6 6.7E+6 6.7E+6 7.8E+5 6.7E+6 1.9E+6 6.7E+6 5.9E +6 6.7E+6 1.3E+7 6.7E+6 4.1E +7 6.7E+6 3.0E+8 6.7E+ 6 1.4E+9 Xe-131m 1.3E+3 1.2E+ 5 5.2E+ 5 Xe-133 4.8E+ 5 4.1 E+7 1.8E+8 2.8E+8 6.7E+8 2.1E+9 4.4E+9 1.4E+10 8.9E+10 2.5E+11 2.0E+4 7.4E+6 1.1E+7 2.7E+7 8.5E+7 1.7E+ 8 5.2E+8 2.6E+ 9 4.1E+9 g Xe-133m 1.8E+ 6
. Xe-135 5.9E+4 5.6E + 6 2.3E+7 3.4E+7 7.4E+7 1.9E+8 3.3E+8 6.7E+ 8 1.0E+9 1.0E+9 8.5E+4 5.9E+6 1.7E+7 1.8E+7 1.8E+7 1.8E+7 1.8E+7 1.8E+7 1.8E+7 1.8E+ 7 h Xe-135m 1.9E +7 1.9E+7 1.9E+7 1.9E + 7 1.9E+7 1.9E +7 g Xe-137 3.7E +5 1.3E+7 1.9E+7 1.9E +7 3.7E+5 2.6E+7 7.4E+7 7.4E+7 7.4E+7 7.4E +7 7.4E+7 7.4E+7 7.4E+7 7.4E +7 5 Xe-138 g Total 1.9E + 6 1.3E + 8 4.3E+ 8 5.7E+8 1.2E +9 3.0E +9 5.7E+ 9 1.6E+10 9.4E + 10 2.6E + 11 S 8
B. Condenser Release to Environment (megabecquerel) g Kr-83m 0 0 5.9E+3 7.8E+4 4.4E+5 1.3E+ 6 1.7E+6 1.9E+6 1.9E+6 1.9E+ 6 Kr-85 0 0 7.4E+2 1.3E+ 4 1.3E+ 5 9.3E + 5 2.6E + 6 1.3E +7 2.3E+ 8 5.9E+9 Kr-85m 0 0 1.5E+4 2.3E+5 1.8E+ 6 8.5E+6 1.6E+7 3.0E +'7 3.6E+7 3.6E+7 Kr-87 0 0 2.0E+4 2.4E+5 1.1E+ 6 2.4E +6 2.7E+ 6 rE+ 6 2.8E+ 6 2.8E+6 Kr-88 0 0 3.7E+4 5.6E+5 3.7E+ 6 1.4E +7 2.3E+7 '+7 3.2E+7 3.2E+7 Kr-89 0 0 4.1 E+0 4.1 E+0 4.1 E+ 0 4.1E+0 4.1 E+0 +0 4.1 E+0 4.1E+0 Xe-131m 0 0 4.1 E+2 6.7E+3 6.7E +4 4.8E+ 5 1.3E+ 6 6.7E + 6 1.0E+ 8 1.3E + 9 Xe-133 0 0 1.4E+5 2.4E+ 6 2.3E+7 1.6E+8 4.4E+8 2.2E+9 3.0E+10 1.8E+11 E Xe-133m 0 0 5.6E+3 1.0E+5 9.3E+5 6.7E+ 6 1.8E+7 8.1E+7 8.1E+8 2.0E+9 E. Xe-135 0 0 1.7E+4 2.7E+5 2.4E+ 6 1.4E+7 3.3E+7 9.6E+7 2.0E + 8 2.0E+8 E. Xe-135m 0 0 2.9E+3 1.0E+4 1.3E+4 1.3E+4 1.3E+4 1.3E+4 1.3E+4 1.3E+4 & Xe-137 0 0 3.4E+1 3.5E+1 3.5E+1 3.5E+1 3.5E+1 3.5E+1 3.5E+1 3.5E+1 Xe-138 0 0 1.0E+ 4 3.2E+ 4 3.7E+4 3.7E+4 3.7E+4 3.7E+4 3.7E+4 3.7E+4 g Total 0 0 2.5E+5 3.9E+ 6 3.4E+7 2.1 E+8 5.4E+ 8 2.5E+9 3.2E+10 1.9E+11 hW tt 8 '8 a
23A6100 Riv. 9 ABWR StandardSafety Analysis Report O Table 15.6-13 Loss of Coolant Accident Meteorology and Offsite Dose Results Site Boundary Dose Results Meteorology
- Dist Thyroid Dose Whole Body Dose (s/m3) (m) (Sv) (Sv) 2.18E-03 max 3 6.4E-02 1.37E-03 Chp 2 1.9E+00 4.1 E-02 1.18E-03 300 1.6E+00 3.5E-02 2.19E-04 800 3.0E-01 6.5E-02
* " Max" = maximum meteorology to meet 10CFR100 limitation.
Low Population Zone Boundary Dose Results Time Meteorology Dist Thyroid Dose Whole Body Dose (h) (s/m3) (m) (Sv) (Sv) 8 3.73E-05 4828 7.3E-02 2.5E-03 24 1.21E-05 9.9E-02 3.5E-03 96 4.27E-06 2.0E-01 4.9E-03 720 9.09E-07 3.4E-01 5.9E-03 8 1.56E-04 Chp 2 3.1 E-01 1.0E-02 24 9.61 E-05 5.1 E-01 1.8E-02 96 3.36E-05 1.3E+00 2.9E-02 I 720 7.42E-06 2.4E+00 3.8E-02 Table 15.6-14 Loss of Coolant Accident Meteorology q and Control Room Dose Results l l Time Meteorology
- Thyroid t Whole Body t Betat 3
(h) (s/m ) (Sv) (Sv) (Sv) l 0-8 h 3.10E-03 3.60E-02 3.50E-03 4.20E-02 l 8-24 h 1.83E-03 7.20E-02 9.00E-03 1.33E-01 l 1-4 days 1.16E-03 1.66E-01 1.96E-02 3.20E-01 l 4-30 days 5.12E-04 2.76E-01 2.67E-02 4.47E-01
- See Subsection 15.6.5.5.3.2 for description of meteorology. Values are for dispersion from Reactor Building. Dispersion values for releases from Turbine Building are a factor of six less than Reactor Building dispersion values.
l t These values are cumulative from the beginning to the end of period in the first colurnn. 15.6-40 Decrease in Reactor Coolant inventory- Amendment 37
23A6100 R1v. 9 ABWR standutsarety Analysis neport a 1.0 0.9 w 0.8 0.7 - 0.6 - U E 0.5 - B f 0.4 - E W 0.3 - d 0.2 - TOTAL R 0.1 N ll
, DRYWELL i ;
l O 20 40 60 80 100 TIME (s) Figure 15.6-3 Airborne lodine in Primary Containment During Blowdown Phase O - PLANT REACTOR TURBINE BUILDNG STACK BUILDING CONTROL CONTROL ROOM BUILDING INTAKE VENTS h $ REACTOR
/ TURBINE BUILDNG f
SERVICE O BUILDING U Figure 15.6-4 ABWR Plant Layout Decrease in Reactor CoolantInventory- Amendment 37 15.645
G b i # I BREAK FLOW / x
/
REACTOR BUILDING TO TURBINE STEAM TUNNEL / X E >@ <
' ~ - ~ ~ ~ ~ ~ ,
TURBINE CONTAINMENT / v nX f ><l- BUILDING
/ E O
I CONTROL $ l GROUND LEVEL I I _l I i E a I , i, h $ REGENERATIVE h HEAT EXCHANGER BREAK d g B NON-REGENERATIVE HEAT EXCHANGER R t : a ir $ Figure 15.6-5 Leakage Path for Clean Up Water Line Break j O O O
_. _. _ . _ . ~ ._. _ . _ . . . _ . . .- .__ _ _. _ 23A6100 Rev. 9 $($ 2.0 G 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs l 2.1.1.1 With the reactor steam dome pressure < 5.41 MPaG or core flow < 10% rated core flow: THERMAL POWER shall be s 25% RTP. 2.1.1.2 With the reactor steam dome pressure 2 5.41 MPaG and core flow 210% rated core flow: MCPR shall be 21.07. 2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel. 2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be s 9.13 MPaG. 2.2 SL Violations With any SL violation, the following actions shall be completed: 2.2.1 Within I hour, notify the NRC Operations Center, in accordance with 10 CFR 50.72. 2.2.2 Within 2 hours: 2.2.2.1 Restore compliance with all SLs; and 2.2.2.2 Insert all insertable control rods. 2.2.3 Within 24 hours, notify the [ General Manager-Nuclear Plant and Vice President-Nuclear Operations] and the [offsite reviewers specified in Specification 5.5.2, "[0ffsite) Review and Audit"). l i I (continued) ABWR TS 2.0-1 Amendment 37
23A6100 Rev. 9 SSLC Sensor Instrumentation 3.3.1.1 Table 3.3.1.1-1 (Page 3 of 7)
# SSLC Sensor Instrumentation
( t APPLICABLE CONDITIONS MODES OR REFERENCED J OTHER FROM SPECIFIED REQUIRED REQUIRED SURVE!LLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS ACTIONS REQUIREMENTS VALUE I
- 7. Reactor Vessel Water Level. Low, Level 2 7a. ESF Initiation 1,2,3 4 N SR 3.3.1.1.1 m ( ) em SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13 7b. Isolation Initiation. 1,2,3 4 K SR 3.3.1.1.1 m t 3 cm SR 3.3.1.1.5 l 4
' SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.14 (f) 4 L SR 3.3.1.1.1 SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10 i SR 3.3.1.1.14 l 7c. SLCS and FWRB Initiation 1,2 4 G SR 3.3.1.1.1 a [ ] cm l SR 3.3.1.1.6 i SR 3.3.1.1.11 '
- 8. Reactor vessel Water Level -Low, l f3 Level 1.5 t i
(' bj Ba. ESF Initiation. 1,2,3, 4 N SR 3.3.1.1.1 m I 3 cm SR 3.3.1.1.5 4('),5(*3 SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13
- 86. Isolation Initiation. 1,2,3 4 o SR 3.3.1.1.1 2 [ ] cm SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10 i SR 3.3.1.1.14 8c. ATWS ADS Inhibit. 1, 2 4 H SR 3.3.1.1.1 m [ 3 cm SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.10
- 9. Reactor Vessel Water Level-Low, Level 1 9a. ADS A, CAMS A, LPFL A & 1,2,3, 4 N SR 3.3.1.1.1 t [] cm LPFL C Initiation SR 3.3.1.1.5 4(e)' $(e) SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13 9b. ADS B, Diesel Generator, 1,2,3, 4 N SR 3.3.1.1.1 t t 3 cm RCW, CAMS B, & LPFL B SR 3.3.1.1.5 Initiation 4(e)' 5(e) SR 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.13
\
L -- (Continued) ABWR TS 3.3-12 Amendment 37
23A6100 Rev. 9 S/RVs 3.4.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.2.1. Verify the safety function lift setpoints In accordance of the required S/RVs are as follows: with the Inservice Number of Setpoint Testing Program S/RVs (MPaG) 2 7.92 1 0.0792 4 7.99 1 0.0799 4 8.06 1 0.0806 4 8.13 1 0.0813 4 8.19 1 0.0819 , l Following testing, lift settings shall be , within 1%. I SR 3.4.2.2 -------------------NOTE-------------------- Not required to be performed until 12 hours , after reactor steam dome pressure is N 2 6.55 MPaG. l Verify each required S/RV opens when 18 months manually actuated. , i 1 ABWR TS 3.4-3 Amendment 37
23A6100 Rev. 9 ECCS-Operating 3.5.1 /"T SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.5.1.6 NOTE Not required to be performed until 12 hours after reactor steam dome pressure is l 2 1.03 MPaG. Verify, with RCIC steam supply pressure 18 months s 1.14 MPaG, the RCIC pump can develop a flow rate 2182 m3/h against a system head corresponding to reactor pressure. SR 3.5.1.7 - NOTE Vessel injection may be excluded. Verify each ECCS subsystem actuates on an 18 months actual or simulated automatic initiation signal. A SR 3.5.1.8 NOTE Valve actuation may be excluded. Verify the ADS actuates on an actual or 18 months simulated automatic initiation signal. SR 3.5.1.9 NOTE Not required to be performed until 12 ' hours after reactor steam dome pressure is 2 6.55 MPaG. l Verify each ADS valve opens when manually 18 months actuated. l l i l O l b I 1 ABWR TS 3.5-6 Amendment 37 l.
l 23A6100 Rev. 9 RCW/RSW System and UHS-Operating 3.7.1 l ( SURVEILLANCE REQUIREMENTS j\ SURVEILLANCE FREQUENCY 4 i SR 3.7.1.1 Verify the water level of each UHS 24 hours ~ [ spray pond] is 2 [ ] m. } j SR 3.7.1.2 Verify the water level in each RSW pump 24 hours ! well of the intake structure is 2 [ ] m. i l SR 3.7.1.3 Verify the RSW water temperature at the 24 hours
- inlet to the RCW/RSW heat exchangers is 2
s [33.3]*C. ) SR 3.7.1.4 -------------------NOTE-------------------- Isolation of flow to individual components does not render RCW/RSW System inoperable. O Verify each RCW/RSW division and associated 31 days UHS [ spray network] division manual, power operated, and automatic valve in the flow path servicing safety related systems or components, that is not locked, sealed, or otherwise secured ir, position, is in the correct position. SR 3.7.1.5 Verify each RCW/RSW division and associated 18 months VHS [ spray network] division actuate on an actual or simulated initiation signal. ABWR TS 3.7-3 Amendment 37
23A6100 Rev. 9 RCW/RSW System and UHS-Shutdown 3.7.2
/~
i] SURVEILLANCE REQUIREMENTS
~
SURVEILLANCE FRE0VENCY SR 3.7.2.1 Verify the water level of each UHS 24 hours [ spray pond] is 2 [ ] m. SR 3.7.2.2 Verify the water level in each RSW pump 24 hours well of the intake structure is 2 [ ] m. SR 3.7.2.3 Verify the RSW water temperature at the 24 hours inlet to the RCW/RSW heat exchangers is s [33.3]'C. SR 3.7.2.4 -------------------NOTE-------------------- Isolation of flow to individual components does not render RCW/RSW System inoperable. l
\s) Verify each RCW/RSW division and associated UHS [ spray network] division manual, power 31 days operated, and automatic valve in the flow path servicing safety related systems or components, that is not locked, sealed, or otherwise secured in position, is in the correct position.
SR 3.7.2.5 Verify each RCW/RSW division and associated 18 months VHS [ spray network] division actuate on an actual or simulated initiation signal. r~ (N)
%J I
i ABWR TS 3.7-6 Amendment 37
23A6100 Rev. 9 RCW/RSW System and UHS - Refueling 3.7.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY i SR 3.7.3.1 Verify the water level of each UHS [ spray 24 hours pond] is 2 [ ] m. SR 3.7.3.2 Verify the water level in each RSW pump 24 hours well of the intake structure is 2 [ ] m. SR 3.7.3.3 Verify the RSW water temperature at the 24 hours inlet to the RCW/RSW heat exchangers is s [33.3]*C. l 1 l i SR 3.7.3.4 --------------NOTE------------------------ Isolation of flow to individual components does not render RCW/RSW System inoperable. Verify RCW/RSW division and associated UHS 31 days [ spray network] division manual, power operated, and automatic valve in the flow path servicing safety related systems or components, that is not locked, sealed, or otherwise secured in position is in the correct position. SR 3.7.3.5 Verify each RCW/RSW division and assoicated 18 months i UHS [ spray network] division actuate on an actual or simulated initiation signal. O ABWR TS 3.7-8 Amendment 37
23A6100 Rev. 9 Design Features 4.0 ,O t, ) 4.0 DESIGN FEATURES (continued) v 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:
- a. Fuel assemblies having a maximum k-infinity of 1.35 in the normal reactor core configuration at cold conditions;
- b. k 5 0.95 if fully flooded with unborated water, which in##c1udes an allowance for uncertainties as described in Section 9.1 of the SSAR.
4.3.1.2 The new fuel storage racks are designed and shall be maintained with:
- a. Fuel assemblies having a maximum k-infinity of 1.35 in l the normal reactor core configuration at 20*C;
- b. k ## s 0.95 if fully flooded with unborated water, which )
[] inc1udes an allowance for uncertainties as described in ; () Section 9.1 of the SSAR;
- c. k,,, s 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties as described in Section 9.1 of the SSAR; and
- d. A nominal [approximately 16] cm center to center distance between fuel assemblies placed in storage racks.
4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to pre ent inadvertent draining of the. pool below 3.1 m above the top of the active fuel. 4.3.3 Capacity 4.3.3.1 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no less than 2354 fuel assemblies (270% of full core discharge). l'h
]
ABWR TS 4.0-2 Amendment 37
23A6100 Rev. 9 S/RVs B 3.4.2 C.^.S E S ACTIONS The 14 day Completion Time to restore the inoperable (continued) required S/RVs to OPERABLE status is based on the relief capability of the remaining S/RVs, the low probability of an event requiring S/RV actuation, and a reasonable time to complete the Required Action. B.1 and B.2 With less than the minimum number of required S/RVs OPERABLE, a transient may result in the violation of the ASME Code limit on reactor pressure. If the inoperable required S/RV cannot be restored to OPERABLE status within the associated Completion Time of Required Action A.1 or if two or more required S/RVs are inoperable, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full l power conditions in an orderly manner and without j challenging plant systems. 1 ( SURVEILLANCE SR 3.4.2.1 REQUIREMENTS This Surveillance demonstrates that the required S/RVs will ) open at the pressures assumed in the safety analysis of I Reference 2. The demonstration of the S/RV safety function lift settings must be performed during shutdown, since this is a bench test. The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures. The S/RV setpoint is i 3% for OPERABILITY; however, the valves are reset to i 1% during the Surveillance to allow for drift. The Frequency is specified in accordance with the Inservice ! Testing Program. 4
/ (continued)
ABWR TS B 3.4-8 Amendment 37 l
23A6100 Rev. 9 S/RVs B 3.4.2 O] ( BASES SURVEILLANCE SR 3.4.2.2 REQUIREMENTS (continued) A manual actuation of each required S/RV is performed to verify that, mechanically, the valve is functioning properly and no blockage exists in the valve discharge line. This can be demonstrated by the response of the turbine control valves or bypass valves, by a change in the measured steam flow, or any other method suitable to verify steam flow. Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Sufficient time is therefore allowed after the required pressure is achieved to perform this test. Adequate pressure at which this test is to be performed is [6.55] MPaG (the pressure recommended by the valve manufacturer). Plant startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME requirements, prior to valve , installation. Therefore, this SR is modified by a Note that ' states the Surveillance is not required to be performed until 12 hours after reactor steam dome pressure is 2: ([6.55] MPaG). The 12 hours allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions for testing and provides a O) ( reasonable time to complete the SR. If the valve fails to actuate due only to the failure of the solenoid but is i l capable of opening on overpressure, the safety function of l the S/RV is considered OPERABLE. l l The Frequency is consistent with SR 3.4.2.1 to ensure that , the S/RVs are manually actuated following removal for l refurbishment or lift setpoint testing. REFERENCES 1. ASME, Boiler and Pressure Vessel Code, Section III. )
- 2. ABWR SSAR, Section 5.2.2.
- 3. ABWR SSAR, Chapter 15.
l l l l l t v ABWR TS B 3.4-9 Amendment 37
23A6100 Rsv. 9 ECCS-0perating B 3.5.1 i BASES V BACKGROUND The RCIC System is designed to provide core cooling for a l ( Continued ) wide range of reactor pressures, 1.03 MP M to 8.12 MPaG. Upon receipt of an initiation signal, the RCIC turbine accelerates to a specified speed. As the RCIC flow increases, the turbine control valve is automatically adjusted to maintain design flow. Exhaust steam from the RCIC turbine is discharged to the suppression pool. A full flow test line is provided to route water from and to the suppression pool to allow testing of the RCIC System during normal operation without injecting water into the RPV. For the station black out scenario, where all AC power from the offsite AC circuits and from the standby diesel generators are assumed to be lost, RCIC is designed to provide makeup water to the RPV. Diverse alternatives to RCIC are provided by the Combustion Turbine Generator (CTG) and the AC-Independent Water Addition (ACIWA) mode of RHR(C) (References 13 and 14). If RCIC is inoperable, water can be injected into the RPV either by powering other ECCS subsystems from the CTG or by the Fire Protection System (FPS) using the ACIWA mode of RHR(C). The ECCS pumps are provided with minimum flow bypass lines, which discharge to the suppression pool. The valves in (o N these lines automatically open to prevent pump damage due to overheating when other discharge line valves are closed or l RPV pressure is greater than the LPFL pump discharge i pressures following system initiation. To ensure rapid I delivery of water to the RPV and to minimize water hammer effects, the ECCS discharge line " keep fill" systems are designed to maintain all pump discharge lines filled with l water. ,i The ADS (Ref.1) consists of 8 of the 18 S/RVs. It is designed to provide depressurization of the primary system during a small break LOCA if RCIC and HPCF fail or are unable to maintain required water level in the RPV. ADS operation reduces the RPV pressure to within the operating pressure range of the low pressure ECCS subsystems (LPFL), so that these subsystems can provide core cooling. Each ADS valve is supplied with pneumatic power from either its own dedicated accumulator located in the drywell, or from the atmospheric control system (ACS) directly when pneumatic power from the accumulators is not needed. The ACS also supplies the nitrogen (at pressure) necessary to assure the ADS accumulators remain charged for use in emergency actuation. (continued) ABWR TS B 3.5-4 Amensent 37
23A6100 Rev. 9 ECCS-Operating B 3.5.1 i BASES LC0 With less than the required number of ECCS subsystems i (continued) OPERABLE during a limiting design basis LOCA concurrent with the worst case single failure, the margins to the limits specified in 10 CFR 50.46 (Ref. 7) would be reduced. Furthermore, all ECCS subsystems are assumed to be initially 4 available in the comprehensive set of analyses performed to satisfy the single failure criterion required by 10 CFR 1' 50.46 (Ref. 7) . Thus all ECCS subsystems must be OPERABLE. The ECCS is supported by other systems that provide automatic ECCS initiation signals (LC0 3.3.1.1, "SSLC Sensor 4 Instrumentation" and LC0 3.3.1.4, "ESF Actuation
- Instrumentation"), cooling and service water to cool rooms containing ECCS equipment (LC0 3.7.1, " Reactor Building Cooling Water (RCW) System, Reactor Service Water (RSW)
System and Ultimate Heat Sink (VHS)-Operating", LC0 3.7.2, I
"RCW/RSW and UHS-Shutdown" and LC0 3.7.3 "RCW/RSW and UHS-Refueling"), electrical power (LC0 3.8.1, "AC Sources-Operating," and LC0 3.8.4, "DC Sources-0perating").
A LPFL subsystem may be considered OPERABLE during alignment and operation for decay heat removal when below the actual RHR cut in permissive pressure in MODE 3, if capable of being manually realigned (remote or local) to the LPFL mode s and not otherwise inoperable. At these low pressures and decay heat levels, a reduced complement of ECCS subsystems can provide the required core cooling, thereby allowing operation of an RHR shutdown cooling loop when necessary. APPLICABILITY All ECCS subsystems are required to be OPERABLE during MODES 1, 2, and 3 when there is considerable energy in the reactor core and core cooling would be required to prevent fuel damage in the event of a break in the primary system piping. In MODES 2 and 3, the RCIC System is not required l to be OPERABLE when pressure is s 1.03 MPaG since other ECCS subsystems can provide sufficient flow to the vessel. In MODES 2 and 3, the ADS functirn is not required when pressure is s 0.343 MPaG because the low pressure ECCS subsystems (LPFL) are capable of providing flow into the RPV below this pressure. ECCS requirement.i for MODES 4 and 5 are specified in LC0 3.5.2, "ECCS Shatdown." (continued) ABWR TS B 3.5-7 Amendment 37
23A61oo anv. 9 ECCS -0perating B 3.5.1 D BASES SURVEILLANCE SR 3.5.1.9 REQUIREMENTS (continued) LC0 3.3.1.1 and LC0 3.3.1.4 overlap this Surveillance to provide complete testing of the assumed safety function. l The Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. REFERENCES 1. ABWR SSAR, Section 6.3.2.
- 2. ABWR SSAR, Section 15.6.4.
- 3. ABWR SSAR, Section 15.6.5.
- 4. ABWR SSAR, Section 15.6.6.
- 5. 10 CFR 50, Appendix K.
- 6. ABWR SSAR, Section 6.3.3.
- 7. 10 CFR 50.46.
- 8. ABWR SSAR, Section 6.3.3.9.
- 9. ABWR SSAR, Section 19D.9.
- 10. ABWR SSAR, Section 7.3.1.1.1.2.
- 11. 10 CFR 50, Appendix A, GDC 33. I
- 12. ABWR SSAR, Section 6.7.
- 13. ABWR SSAR, Section 9.5.11.
l i l (continued) ABWR TS B 3.5-27 Amendment 37
23A6100 Rev. 9 RCW/RWS System and UHS-Operating B 3.7.1 BASES LC0 (continued) d. The associated piping, valves, instrumentation, and controls required to perform the safety related function are OPERABLE. OPERABILITY of the VHS is based on a maximum RSW water temperature ~of [33.3]'C at the inlet to the RCW/RSW heat exchangers with OPERABILITY of each division requiring a minimum water level at or above elevation [mean sea level (equivalent to an indicated level of ;t [ ] m) and six OPERABLE spray networks]. The maximum RSW water temperature of [33.3]'C will insure that the peak temperature at the inlet to the RCW/RSW heat exchangers will not exceed the designed value of 35'C during a LOCA. The isolation of the RCW/RSW System to components or systems i may render those components or systems inoperable, but does ' not affect the OPERABILITY of the RCW/RSW System. l APPLICABILITY In MODES 1, 2, and 3, the RCW/RSW System and UHS are required to be OPERABLE to support OPERABILITY of the Q equipment serviced by the RCW/RSW System and UHS, and are required to bc OPERABLE in these MODES. In MODES 4 and 5, the OPERABILITY requirements of the i RCW/RSW System and UHS are specified in LCOs 3.7.2, "RCW/RSW l and UHS-Shutdown" and 3.7.3, "RCW/RSW and UHS-Refueling". l ACTIONS A.1 If one RCW pump and/or one RSW pump and/or one RCW/RSW heat exchanger and/or one [ spray network] in the UHS in the same division is inoperable, action must be taken to restore the inoperable component (s), and thus the division affected, to OPERABLE status within 14 days. In this condition sufficient equipment is still available to provide cooling water to the required safety related components and sufficient heat removal capacity is still available to adequately cool safety related loads, even assuming the worst case single failure. Therefore, continued operation for a limited time is justified. (continued) ABWR TS B 3.7-4 Amenerent 37
i 2346100 nev. 9 RCW/RWS System and UHS-Operating i B 3.7.1 !l ! BASES L j ACTIONS A.1 (continued) l (continued) i The 14-day Completion Time is reasonable, based on the low ! probability of an accident occurring during the 14 days that one or more components are inoperable in one division, the number of available redundant divisions, the substantial l j cooling capability still remaining in a division in this ! Condition, and the expected high division availability :
- afforded by a system where most of the equipment, including i the minimum required for most functions, is normally )
j operating. This Completion Time is also based on PRA l sensitivity studies (Ref. 8). 4 4
- B.1 and B.2 1
! If one RCW/RSW division or both [ spray networks] in one UHS i division is inoperable for reasons other than Condition A, I then, immediately, those required feature (s) supported by j the inoperable RCW/RSW division must be declared inoperable ! (e.g., Emergency Diesel Generatory RHR heat exchanger, etc.) t and the applicable Canditions andSRequired Actions of, the j appropriate LCOs for the inoperable required feature (s) must I i be entered. For example, applicable Conditions of LC0 ! 3.8.1, "AC Sources-0perating," LC0 3.4.7, " Residual Heat l Removal (RHR) Shutdown Cooling System-Hot Shutdown," LC0 l 3.4.1, " Reactor Internal Pumps (RIP) Operating," LC0 l 3.6.1.5, "Drywell Air Temperature", LC0 3.6.2.3,
" Suppression Pool Cooling," and LC0 3.6.2.4, " Containment
- Spray" be entered and the Required Actions taken if the l inoperable RCW/RSW division results in an inoperable DG, RHR
' shutdown cooling, RIPS, drywell temperature increase due to inoperable drywell coolers, RHR suppression pool cooling, and RHR containment spray, respectively. This is in accordance with LC0 3.0.6 and ensures the proper actions are taken for these components. Additionally, immediate action must be taken to restore the inoperable RCW/RSW division or UHS [ spray networks] to OPERABLE status. This is consistent with the Required Actions of the applicable LCOs for those support feature (s) declared inoperable as a result of the inoperable RCW/RSW division. O V (continued) ABWR TS B 3.7-5 Amendment 37
23A6100 Rsv. 9 RCW/RWS System and UHS-Operating B 3.7.1
'/9 i BASES N_)-
ACTIONS C.1 and C.2 (continued) If one RCW pump and/or one RSW pump and/or one RCW/RSW heat exchanger and/or one UHS [ spray network) in the same division is inoperable in two or more separate divisions, one RCW/RSW or UHS [ spray network] division must be restored to OPERABLE status within 7 days and two RCW/RSW or UHS [ spray network] divsions must be restored to OPERABLE status in 14 days. In this condition sufficient equipment is still available to provide cooling water to the required safety related components and sufficient heat removal capacity is still available to adequately cool safety related loads. Therefore, continued operation for a limited time is justified. However, in the degraded mode of this Condition, overall reliability and heat removal capability is reduced from that of Condition A, and thus a more restrictive Completion Time is imposed. The 7 and 14 day Completion Times are reasonable, based on the low probability of an accident occurring during the period that one or more redundant components are inoperable in one or more divisions, the number of available redundant
/ divisions, the substantial cooling capability still Q remaining in divisions in this Condition, and the expected high division availability afforded by a system where most of the equipment, including the minimum required for most functions, is normally operating. These Completion Times 1 are also based on PRA sensitivity studies (Ref. 8).
! D.1 and 0.2 If the RCW/RSW division cannot be restored to OPERABLE status within the associated Completion Time, or two or more RCW/RSW divisions are inoperable for reasons other than Condition C, or the UHS is determined inoperable, or two or more VHS [ spray network] divisions are inoperable for reasons other than Condition C, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours and in MODE 4 within 36 hours. The allowed , Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. O) g NJ (continued) ABWR TS B 3.7-6 Amendment 37
l 23A6100 Rev. 9 RCW/RWS System and UHS-Operating B 3.7.1 1 ) BASES v l SURVEILLANCE SR 3.7.1.1 REQUIREMENTS This SR ensures adequate long term (30 days) cooling can be maintained. With the UHS water source below the minimum level, the affected RCW/RSW division must be declared I inoperable. The 24 hour Frequency is based on operating l experience related to trending of the parameter variations during the applicable MODES. SR 3.7.1.2 This SR verifies the water level in each RSW pump well of the intake structure to be sufficient for the proper operation of the RSW pumps (net positive suction head and pump vortexing are considered in determining this limit). The 24 hour Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES. SR 3.7.1.3 () Verification of the RSW water temperature at the inlet to the RCW/RSW heat exchanger ensures that the heat removal capability of the RCW/RSW System is within the assumptions of the DBA analysis. The 24 hour Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES. SR 3.7.1.4 Verifying the correct alignment for each manual, power operated, and automatic valve in each RCW/RSW and associated UHS [ spray network] division flow path provides assurance that the proper flow paths will exist for RCW/RSW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position and yet considered in the correct position, provided it can be automatically realigned to its accident position. This SR does not require any testing or valve manipulation; rather, it l involves verification that those valves capable of potentially being mispositioned are in the correct position.
/m 1 (continued) 'Y ABWR TS B 3.7-7 Amendment 37 l
23 moo nev. 9 RCW/RWS System and UHS-Shutdown B 3.7.2 f% (x BASES SURVEILLANCE SR 3.7.2.1 REQUIREMENTS This SR ensures adequate long term (30 days) cooling can be maintained. With the UHS water source below the minimum level, the affected RCW/RSW division must be declared inoperable. The 24 hour Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES. SR 3.7.2.2 This SR verifies the water level in each RSW pump well of the intake structure to be sufficient for the proper operation of the RSW pumps (net positive suction head and pump vortexing are considered in determining this limit). The 24 hour Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES. SR 3.7.2.3 () Verification of the RSW water temperature at the inlet to the RCW/RSW heat exchangers ensures that the heat removal capability of the RCW/RSW System is within the assumptions of the DBA analysis. The 24 hour Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES. SR 3.7.2.t Verifying the correct alignment for each manual, power operated, and automatic valve in each RCW/RSW and associated UHS [ spray network] division flow path provides assurance that the proper flow paths will exist for RCW/RSW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position and yet considered in the correct position, provided it can be automatically realigned to its accident position. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves
's.) (continued)
ABWR TS B 3.7-13 Amendment 37
23A6100 Rev. 9 RCW/RSW System and UHS-Refueling B 3.7.3 BASES 5 ACTIONS A.1 and (continued) diesel $ or made inoperable and LC0 3.9.7, " Residual Heat F 1HR)-High Water Level" for RHR shutdown cooling , , , moperable. This is in accordance with LCO 3.0.6 and c.nsures the proper actions are taken for these componentL SURVEILLANCE SR 3.7.3.1 REQUIREMENTS This SR ensures adequate long term (30 days) cooling can be maintained. With the UHS water source below the minimum level, the affected RCW/RSW division must be declared inoperable. The 24 hour Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES. SR 3.7.3.2 This SR verifies the water level in each RSW pump well of
\O the intake structure to be sufficient for the proper operation of the RSW pumps (net positive suction head and pump vortexing are considered in determining this limit).
The 24 hour Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES. SR 3.7.3.3 Verification of the RSW water temperature at the inlet to the RCW/RSW heat exchangers ensures that the heat removal capability of the RCW/RSW System is within the assumptions of the DBA analysis. The 24 hour Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES. SR 3.7.3.4 Verifying the correct alignment for each manual, power operated, and automatic valve in each RCW/RSW and associated UHS [ spray network) division flow path provides assurance (continued) ABWR TS B 3.7-17 Amen *=nt 37
23A6100 Rzv. 5 ABWR standantsaretykarysis neport O Chapter 19 V Table of Contents List of Tables . .. .... . .. . . ... . . . . . . . . . . . . . . . . . . . . . . ... .. . . . . . . . . . . . . . . . . . . . . . . . . 19-ix List of Figures.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... . 19-xix 19.0 Response to Severe Accident Policy Statement.......... ..... . . . . . . . . . . . . . . .. . 19.1-1 19.1 Purpose and Summary. . .. .. ..... ...... . . . . . . . . . . . . . . . . . . . .. . .. . . . . . . . . . 19.1-1 19.1.1 Purpose . .. .... . .. .... ............ .. . .. ... . . . . . . . . . . . . . ... .. . 19.1-1 19.1.2 Summary . ..... ... . . . . . . . . . . . . . . . . . . ... . . . . . . . . . . . . . . .. .. .... 19.1-1 19.1.3 References.. .. .......... . . . . . ... . . . . . . . . . . . . . . . . . . . .. . . .. . .... . 19.1 -2 19.2 Introduction .... ... . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .19.2-1 19.2.1 Definitions. . . . . . . . . . . . . . . . . . . ... .. .. . .. . .. . . . . . .. .. . . . 19.2- 1 19.2.2 Objective and Scope... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ....... . 19.2-1 19.2.3 PRA Basis.. .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.2-2 19.2.4 Methodology........ .. . . .. . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . 19. 2-5 19.2.5 References.... .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . 19.2-8 19.3 Internal Event Analysis ........ . .. . ....... . . ... . . . . . . . . . . . . . . . . . .. . . . . . . .. . .. . . 19. S 1 19.3.1 Frequency of Core Damage . . ... .. . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . 19. S 1 19.3.2 Frequency of Radioactive Release ... . . . . . . . . . . . . . . . . . . . . .... .. .. 19.S14 , 19.3.3 Magnitude and Timing of Radioactive Release.. . . . . . . . . . . ..... 19.S21
....... 19.S22 (Q'j l 19.3.4 19.3.5
'bsequence of Radioactive Release.
References... . ..... . . .
. ....... ....... 19 S23 19.4 External Event Analysis and Shutdown Risk Analysis ...... . . . . . . . . . . . . . . . . . . . . . . . . . . 19.4- 1 19.4.1 External Event Review.... . ... . . . . . .. . . . . . . . . . . . . . . . . . . . 19. 4 1 19.4.2 Tornado Strike Analysis ...... . .. . . . . . . . . . . . . . . . .. . . . . . . . . . . . . 19.42 19.4.3 Seismic Margins Analysis... ... ... ... . ... . . . . . . . . . . . . . .... . 19.4-4 19.4.4 Fire Protection Probabilistic Risk Assessment..... . ... . . . . . . . . . . .19.4-11 19.4.5 ABWR Probabilistic Flooding Analysis.. ... .. . . . . . . . . . . . . . . . . . . . . . . 19. 4-12 19.4.6 ABWR Shutdown Risk.. . . .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... ... 19.4-13 19.4.7 References... . . . . . ...............................................19.4-14 19.5 Source Term Sensitivity Studies. ... . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19. 5- 1 19.5.1 Core Melt Progression and Hydrogen Generation. .. ........ . . . . . . . . .. . . . . . 19.5 1 19.5.2 Effect of Overpressure Relief Rupture Disk on Fission Product Release.. . 19.5-2 19.5.3 Alternate Definition of Containment Failure... . . . .... .19.5-3 19.6 Measurement Against Goals. . . .. .. . . . . . ... . . . . . . . . . . . . . . . . . . . . 19. 61 19.6.1 Goals... . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . .. . .. . . . .. . .. 19.6-1 19.6.2 Prevention of Core Damage . . . . . .. . . . . . .... .. . . . . . . .. . 19.6-1 19.6.3 Prevention of Early Containment Failure For Dominant Accident Sequences . . .. . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .19.61 19.6.4 Hydrogen from 100% of Active Zirconium. . . . . . . . . 19.6 2 N 19.6.5 Reliable Heat Removal to Reduce Probability of Containment Failure ... . .19.42
) 19.6.6 Prevention of Hydrogen Deflagration and Detonation.. .. .. . . . .... . 19.6 3 s0 19.6.7 Offsite Dose /Large Release. . . . . . . . . ... . .. . . . 19.6-4 19.6.8 Containment Conditional Failure Probability.. . .. . ..... . .. . 19.64 Table of Contents - Amendment 3S 19-1
l l 23A6100 REv. 9 l ABWR standardsarery Ansiysis neport Table of Contents (Continued) 9!l i 1 19.6.9 Safety Goal Policy Statement .. . . . . . .. . . . . . . . .19.6-6 l 19.6.10 Deleted. .. .. . . .. . . . . .. . 19.6-6 1 1 19.6.11 Conclusion. .. . . . . .. ... .. . . . . . 19.6-6 ; i 19.6.12 References.. .. .. . . . . . . . .. . . . . . 19.6-6 - 1 19.7 PRA as a Design Tool.. ... . . . . . . . . . . . . . . . 19.7-1 l 19.7.1 ABWR Design and Operating Experience..... . . .. . .19.7-1 l 19.7.2 Early PRA Studies ... . . .. . . . . . . . . . . . .19.7-1 19.7.3 PRA Studies During the Certification Effort . . .. . ... . .. . . .19.7-3 19.7.4 Conduct of the PRA Evaluations. . . . . . . . .19.7-10 l 19.7.5 Evaluation of Potential Design Improvements.. . . 19.7-10 l 1 19.8 Important Features Identified by the ABWR PRA. ... .. 19.8-1 l 19.8.1 Important Features from Level 1 Internal Events Analyses. . . . . . 19.8-2 19.8.2 Important Features from Seismic Analyses... . . . 19.8-8 19.8.3 Important Features from Fire Analyses.. . . . . . . . . . . . . .19.8-10 19.8.4 Important Features from Suppression Pool Bypass and Ex-Containment LOCA Analyses. . . . .. .. . . . .. .. . . . .. . . .. . 19.8-13 l 19.8.5 Important Features from Flooding Analyses . . . . . . . .. .. .. 19.8-16 19.8.6 l Important Features from Shutdown Events Analyses.. . . . . .. .19.8-19 { 19.8.7 ABWR Features to Mitigate Severe Accidents... . ...... .. .. .19.8-22 ! 19.9 COL License Information .. . . . . ... .. .. .. . . .. .. .19.9 1 19.9.1 Post Accident Recovery Procedure for Unisolated CUW Line Break.. .. .19.9-1 19.9.2 Confirmation of CUW Operation Beyond Design Bases. . ...... . . 19.9-2 19.9.3 Event Specific Procedures for Severe External Flooding . . . . . .19.9-2 19.9.4 Confirmation of Seismic Capacities Beyond the Plant Design Bases.. .. . 19.9-3 l 19.9.5 Plant Walkdowns . .. . .. . . . . .. .. .19.9-3 19.9.6 Confirmation of Loss of AC Power Event.. ... . . . . . . . . . .. .19.9-4 19.9.7 Procedures and Training for Use of AC-Independent Water Addition System.. . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . 19.9-4 19.9.8 Actions to Avoid Common-Cause Failures in the Essential Multiplexing System (EMUX) and Other Common-Cause Failures. .. . . . .. . . 19.9-5 19.9.9 Actions to Midgate Station Blackout Events. . . . . . . . . ... . . . . . . . . .. 19.9-5 19.9.10 Actions to Reduce Risk ofInternal Flooding .. .. . .. . . . . . . . . 19.9-6 19.9.11 Actions to Avoid Loss of Decay Heat Removal and Minimize Shutdown Risk. . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . .. . . .. . 19.9-7 19.9.12 Procedures for Operation of RCIC from Outside the Control Room. .. 19.9-9 19.9.13 ECCS Test and Surveillance Intenals. . . . . . . . . . . . . . . . . .19.9-10 19.9.14 Accident Management. . ... . . . . . . . . . . . . . . . . ... .. .. 19.9-10 19.9.15 Manual Operation of MOVs. . . . . . . . . . . .. . . . . . . . .19.9-11 19.9.16 High Pressure Core Flooder Discharge Valve .. .. . .. . . . . . .19.9-11 19.9.17 Capability of Containment isolation Valves.. . .. . . . . . .. .. 19.9-11 19.9.18 Procedures to Insure Sample Lines and Drywell Purge Lines Remain Closed During Operation . . . . . . . . . . . . . 19.9-11 19.9.19 Procedures for Combustion Turbine Generator to Supply Power to Condensate Pumps.. . . .. . . . .19.9-11 19-il Table of Contents - Amendment 37
23A6100 R1v. 9 ABWR standedsafety Analysis Reput
~%
d Table of Contents (Continued) 19M.8 References..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . 19 M.1 2 19N Analysis of Common-Cause Failure of Multiplex Equipment...... . . . .. . .. . . . .. .. 19N-1 19N.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19N-1 19N.2 Results and Conclusions . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . .. . .. . 19N-1 19N.S Basis for the Analysis . ...... .. . ....... .. .... . .. .. ... . .. . . . . .. . .. . . . . . . . . 19N-4 19N.4 Potential Causes of and Defenses Against EMUX CCF.. . . ... . ..... ... . 19N-6 19N.5 Discussion of the Effect on Core Damage Frequency... .. ..... .... ....... .. . 19N-10 19N.6 Discussion of the Effect on Isolation Capability. . ... .. .. .... . ... 19N-17 19N.7 S u m m aq . . . . . . .. .. . . . . .. ..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. ....... . 19N-18 19N.8 References...... ....... .... .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .19N-19 19 0 Not Used 19P Evaluation of Potential Modifications to the ABWR Design. . . . . . . . . . . .. . .. 19 P-1 /m \ 19Q ABWR Shutdown Risk Assessment... . .... . . . . . . . . . . . . ..
. . . . . . . . . . . . . . . . . . . . . . . 19 Q 1 19Q.1 Intro d uction . .... .. . .. . . . .. .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
.19Q1 19Q2 Evaluation Scope...... . . .. .. ... . . . . . . . . . . . . . ..
.19Q1 19Q.3 Summary of Results. . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... .
.... . . .... 19Q2 19Q4 Features to Minimize Shutdown Risk.......... ... . . . . . . . . . . . . . . . . . . ..
.... 19Q4 19Q.5 I ns trum e n tatio n.. ..... . . . . . . . . . . .. . . .. . .. .. .. . . . . . .. . .. . .. . . . . . . . . ..19Q15 19Q.6 Flooding and Fire Protection . . . . ..... .. . . . . . . . . . . . . . . . . .
. . .. ... . . . . 19Q17 19Q.7 Decay Heat Removal Reliability Study ... .... . .. .. . . . . . . . . . ..
. ...... 19Q22 19Q8 Use of Freeze Seals in ABWR.. .. .. . ..... .... .... . . . . . . . . . . . . . . .
. . . . .. 19QS4 19Q.9 Shutdown Vulnerability Resulting from New Features. . . .. . . . . . . . . . .
.19Q35 19Q.10 Procedures . ... ......... .. . . . . . . . . . . . . . . . . . . . . . . . . . ...... ...... 19 QS5 19Q.11 Summary of Review of Significant Shutdown Events: Electrical Power and Decay Heat Removal. . . . . . . . . . . . . . ........ . . . . . . .... .19QS9 19Q.12 Results and Interface Requirements.. . . . . . . . . . . . .. . . . . ..... .19Q42 19QA Fault Trees .. . . .. ... .. ... . . . . . . . . . . . . . . . . . . .... . . . . . . . . . . . . ..
. . . . 19QA-1 19QB DHR Reliability Study. ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
.19QB-1 19QB.1 Offsite Dose and Operator Recovey Calculations..... . . . . . . . . . . . . . .
. 19QB-1 19QB.2 Time to Reach Boiling..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . 19QB-2 19QB.3 Time for RPV Water Level to Reach Top of Active Fuel.... ... . .
.19QB-2 q 19QB.4 Human Reliability Analysis (HRA).. .. . . .. . . . . . . . . . . . . . . . . . . . . . .
.19QB-2 j 19QB.5 Decay Heat Removal Capability of CUW and FPC.. .. . .
. . . .19QB-4 19QB.6 References.. . . ... ... .. . ... . ... . . . . .
.19QB-4 l
l Table of Contents - Amendment 37 19-vil
23A61C3 Rett. 9 ABWR StandardSafety Analysis Report O 19QC Review of Significant Shutdown Events: Electrical Power and Decay Heat Removal.19QG1 19QC.1 Review of Significant Shutdown Events . .. . .. . . ..
.19QG1 19R Probabilistic Flooding Analysis.. .. . . . . . . . . .. .19R-1 19R.1 Introduction and Summary. . . . . . . . .. .19R-1 19R.2 Scope of Analysis . . . . . . . .19R-3 19R.3 Screening Analysis (Water Sources and Buildings) . . . . .19R-4 19R.4 Deterministic Flood Analysis. . . . .19R-4 19R.5 Probabilistic Flood Assessment... . . .. . .19R-17 19R.6 Results and Interface Requirements. .. . .. .. .. .19R-29 O
O' 19 viii Table of Contents - Amendment 37 e i
23A6100 R:v.1 ABWR StandardSafetyAnalysis Report
)
, ,U without the inclusion of random failures. As seen in the event trees and the table, the HCLPF values for all accident sequences are greater than 0.60g, which is twice the safe shutdown earthquake (SSE = 0.30g). The results of the convolution analysis in terms of accident classes are shown in Table 191-3. The HCLPF value of accident sequences obtained from the min-max analysis are printed on the event trees next to the column of accident classes. The combination of HCLPF and random failure probabilities of accident sequences are described in Table 191-4. As can be seen, no accident sequence has a HCLPF lower than 0.60g. For most accident sequences, the min-max method of analysis provided lower (more conservative) HCLPF values. However, the use of either method of analysis produced l HCLPFs greater than twice the safe shutdown earthquake for all potential accident I sequences. The seismic margins analysis has provided confidence that the ABWR design
)
will withstand an earthquake of at least 0.6g intensity-twice the design SSE-and achieve safe shutdown without damage to the reactor core. 19.4.4 Fire Protection Probabilistic Risk Assessment
,f - A fire screening analysis was performed to assess the minerability of ABWR to fires within the plant. Mutual agreement was reached earlier with the NRC that a fire screening approach was appropriate and that the Fire Induced Vulnerability Evaluation (FIVE) methodology developed by EPRI provided a proper vehicle for performing this analysis. The methodology is based on conservative assumptions using industrial and plant specific databases for evaluating fire event sequences while making maximum use of existing plant fire analysis and documentation.
The FIVE methodology provides procedures for identifying fire compartments for evaluation purposes, defining fire ignition frequencies, and performing quantitative l screening analysis of fire risk. The criterion for screening acceptability is that the risk of core damage from any postulated fire be less than 1.0E-06 per year. Any fire scenarios not meeting this criterion require more detailed consideration. Five bounding fire scenarios and corresponding fire ignition frequencies were developed on the basis of the FIVE methodology. The first three of these considered the impact of fires which incapacitate each of the three safety divisions (separated by three-four fire barriers, and each encompassing several fire areas) and, thus, the ECCS equipment which is dependent on each for successful performance. Any fire in a divisional area was assumed to result in the immediate and complete loss of function of the division. The fourth scenario considered the impact of a fire in the control room with the conservative assumption that the only ECCS functions available are those that C
; can be controlled and operated from the remote shutdown panel, and the RCIC system, k which can be manually operated outside the control room. The fifth scenario examined the consequences of a fire in the turbine building, based on the assumption that External Event Analysis and Shutdown Risk Analysis - Amendment 31 19.4 11
23A6100 Rsv. 9 ABWR Standard Safety Anslysis Report O 1 resulting loss of offsite power bounds the possible outcomes of this initiator. Considering these composite bounding scenarios is an added conservatism to the already consenative FIVE methodology. Fire ignition frequencies were developed for each of the above scenarios by directly l applying the prescriptive steps documented in the FIVE methodology. Bounding core damage frequency estimates were developed by applying these initiating event frequencies to appropriately modified ABWR Level 1 fault and event tree models and reevaluating them. The final bounding core damage frequency for each of the five scenarios was calculated i to be less than 1.0E-06. These sum of all five was calculated to be 1.3E-06. These results reflect the inherently conservative nature of the FIVE methodology itself, compounded by its additional conservative application in evaluating fire impact at the divisional fire area, control room complex, and turbine building fire levels. Addressing ABWR fire risk at the fire compartment level, considering ignition sources, fire progression, and 3 suppression in more detail will reduce this value by at least an order of magnitude. It is therefore estimated that the core damage frequency resulting from fire is less than 1E-07 per year. 19.4.5 ABWR Probabilistic Flooding Analysis The results of the ABWR Probabilistic Flooding Analysis show that the turbine, control, and reactor buildings are the only stmctures that required evaluations for potential flooding. The other buildings do not contain any equipment that could be used for safe shutdown or potential flooding would not result in a plant transient. Flooding in the turbine building could result in a turbine trip due to loss of circulating water or feedwater. Automatic pump uips and valve closure on high water level should terminate the flooding. But if these were to fail, a non-watertight door at grade level in the turbine building should allow water to exit the building. If this door retained water, watertight doors would prevent water entering the control and reactor buildings. The ! core damage frequency (CDF) for turbine building flooding is less than 4.0E-9 per year j for a plant with low power cycle heat sink (PCHS) and less than 1.0E-8 per year for a l plant with high PCIIS. ) The worst case flood in the control building is a break in the reactor senice water system ! (RSW) which is an unlimited source. Floor drains and other openings in the floor would i direct all flood water to the first floor where the reactor component cooling water (RCCW) rooms are located. The RCCW rooms contain sump pumps. Water level i sensors in the RCCW rooms should actuate alarms in the control room and send signals I to uip the RSW pumps and close isolation valves in the RSW system. If these sensors were to fail, watertight doors on each room should limit flood damage to only one of l l the three RCCW divisions. Breaks in the fire water system could result in interdivisional 19.4 12 ExternalEvent Analysis and Shutdown Risk Analysis- Amendment 37
23A6100 R1v. 9 ABWR standardSafetyAnalysis Report a flooding on the upper floors but floor drains would limit water height to below installed equipment for the first hour. To prevent damage to safety-related equipment after diis time requires operator actions to limit the depth ofwater. The CDF for control building flooding is less than 1.0E-8 per year. Reactor building flooding could occur either inside or outside secondary containment. In either case, the flooding sources are finite with the suppression pool and condensate storage tank being the largest sources. Inside secondary containment flooding cannot cause damage to equipment in more than one of the three safety divisions on the first floor because of watertight doors on each safety division room. As was the case in die control building, water from breaks in lines on upper floors will be directed by floor drains to sump pumps on the first floor. The available volume of rooms on the first floor can contain all potential flood sources. Outside secondary containment, floor drains direct all flood water to sump pumps on floor BlF (third floor). If the sump pumps fail ; or cannot keep up with the flooding rate, an overfill line in the sumps direct water to l the corridor of the first floor where it can be contained as discussed above. Interdivisional flooding may occur but the floor drains will limit the water elevation such that no damage to safety equipment will occur. The CDF for reactor building flooding is approximately 8.5E-9 per year. 5 The total CDF for internal flooding is about 2.SE-8/ year for low PCHS and approximately 2.9E-8 per year for a high PCHS. 19.4.6 ABWR Shutdown Risk The ABWR design has been evaluated for risks associated with shutdown conditions l l (i.e., Modes 3,4 and 5). The evaluation included the following shutdown risk categories: (1) Decay heat removal, (2) Inventory control, i 1 (3) Containment integrity, (4) Loss of electrical power, (5) Reactivity control. l The evaluation also included risk reduction features of the ABWR due to instrumentation, flooding and fire protection, use of freeze seals, and procedure guidelines. ABWR features that are not part of current BWR designs were evaluated to l determine if any aew vulnerabilities would be introduced. In addition, an evaluation of 1 approximately 200 events at operating BWR plants which were considered precursors to
- d loss of decay heat remcal capability showed that ABWR design features could mitigate the effects of all these events.
ExternalEvent Analysis and Shutdown Risk Analysis - Amendment 37 19.4 13
23A6100 R1v. S ABWR standardsateryAnalysis aeport O The results of the ABWR shutdown risk analysis demonstrated that the core damage frequency (CDF) for all shutdown event is less than 1.0E-7 per year. The main features that contribute to this low CDF are: (1) Three physically and electrically independent residual heat removal (RHR) and support systems. (2) Multiple makeup sources for inventory control (e.g., suppression pool, condensate storage tank, AC independent water addition system). (3) Two independent off-site sources of electric power and four on-site sources (three emergency diesel generators and a combustion turbine generator). (4) Reactor protection system (RPS) and standby liquid control system (for boron addition) and interlocks, to prevent accidental reactivity excursions. 19.4.7 References l 19.4-1 Advanced Light Water Reactor Utility Requirements Document, Volume II, Chapter 1: OverallRequirements, Electric Power Research Institute, June 1986. 19.4-2 Policy Statement on Severe Accidents, Federal Register, U.S. Nuclear Regulatory Commission, August 8,1985, p. 32138. l 19.4-3 Advanced Light WaterReactor Utility Requirements Document, VolumeII, Chapter 1; Appendix A: PRA Key Assumptions and Groundrula, Revision 5, Electric Power Research Institute, December 1992. 19.4-4 PRA Procedures Guide-A Guide to the Perfonnance ofProbabilistic Risk Assessments forNuclear Power Plants, NUREG/CR-2300, Final Report, U.S. Nuclear Regulatory Commission, January 1983. 19.4-5 Donald Gene Harrison, Interim ExtemalEvents Integrationfor the EPPI '. LWR Requirements Document (WBS 4.3.3), DOE /ID - 10227, Advanced Reast Severe Accident Program, U. S. Department of Energy, January 1989. 19.4-6 Standard Review Planfor the Review ofSafety Analysis Reportsfor NuclearPower Plants, LWR Edition, NUREG0800, U.S. Nuclear Regulatory Commission, July 1981. 19.4-7 Ricky Lynn Summit, Estimation of Core DamageFrequencyfor Advanced Light Water Reactors Due to Tornado Events (Task 4.3.2.1), Advanced Reactor Severe Accident Program, U.S. Department of Energy, December 1988. 19.4-8 Tornado Missile Simulation and Design Methodology, Volumes 1 and 2, EPRI NP-2005, Electric Power Research Institute, August 1981. 19.4 14 External Event Analysis and Shutdown Risk Analysis - Amendment 35
l 23A6100 Rsv. 9 ABWR standardsareryAnalysisaeport i l G The capability of the Reactor Water Cleanup (CUW) System to provide an additional l means of decay heat removal with the reactor at high pressure wasjudged to be less important than the features selected as "important features." The additional l redundancy prosided by this capability does not significantly reduce the calculated ABWR CDF. This is due to the high reliability of other means of decay heat removal such l as the various modes of operation of the three RHR loops and the containment j overpressure protection system which result in a very small contribution of Class 11 sequences to total CDF without the CUW capability. The degree of redundancy in SRVs to perform the ADS function was alsojudged to be less important than other features. Only three SRVs are required to open to depressurize the reactor so that low pressure pumps can provide the necessary cooling. l The eight ADS SRVs plus the remaining ten SRVs that can be manually actuated far exceed redundancy requirements for depressurization. ADS failure is dominated by common cause failure of the ADS valves. i Another featurejudged to be less important than other features is the automatic initiation of RHR on suppression pool high temperature. Many hours are available to i initiate RHR to remove heat from the suppression pool following transients that dump heat to the suppression pool.The reliability of operators to manuallyinitiate this l function when required isjudged to be very high, therefore this automatic initiation feature does not significantly reduce the calculated CDF. l The capability to manually initiate scram wasjudged to be less important than the selected feature . The ability to manually initiate scram is not an important feature from the standpoint of CDF due to the highly reliable, redundant, and diverse features of the reactivity control systems. The capability to use the CRD hydraulic system to provide additional water injection into the core wasjudged to be less important than the selected features. This is the primary reason that numerical credit for this function was not taken. The valve in some l sequences for the coolant injection capability of the CRD pumps is of a lesser l importance since adequate core cooling is available from other sources to assure a very low core damage frequency. It was alsojudged that the high drywell pressure signal for ADS was less important than j the selected features. With the incorporation of the drywell high pressure signal bypass j l timer, the high drywell pressure signal for ADS is less important. i 1 O Important Features identified by the ABWR PRA - Amendment 37 19.8 7
23A6100 hv. 3 ABWR standard safety Analysis Report O 19.8.2 Important Features from Seismic Analyses 19.8.2.1 Summary of Analysis Results A seismic margins analysis has been performed for the ABWR to calculate a high I confidence low probability of failure (HCLPF) acceleration for important accident sequences and classes of accidents. The results of the analysis indicate that all hypothesized accident sequences and all accident classes had HCLPFs equal to or l greater than 0.60g. This is twice the 0.30g for SSE. Two implicit assumptions in the seismic margins analysis are that a seismic event will result in the unavailability of offsite power and the combustion turbine generator (CTG). The ceramic insulators in the switchyard are not tolerant of high seismic loads and therefore are assumed to fail. Also, the CTG is not qualified for seismic loads and is assumed to be unavailable in a seismic event. Therefore, all of the seismic analyses assume that only emergency AC power and DC power are potentially available. 19.8.2.2 Logical Process Used to Select important Design Features The seismic margins analysis did not include the calculation of minimal cutsets which contdbute to CDF. Therefore, there was no calculation ofimportance parameters such as Fussell-Vesely or Risk Achievement. Since importance parameters were not available, two alternate bases were used to select the important features. The first basis used was the identification of the functions and equipment whose failure would result in the shortest path to core damage in terms of the number of failures required and the relative seismic capacities of the components involved. The second basis used was the identification of the most sensitive functions and equipment in terms of the effect on accident sequence and accident class HCLPFs due to potential variations of component seismic capacities. Using these two bases, the seismic margins analysis was systematically reviewed to identify the "important" features. 19.8.2.3 Features Selected Table 19.8-2 lists the features selected and the rationale for selection. These features met the criteria of either the shortest path to core damage or the most sensitive components. Shortest Paths to Core Damage It is assumed that the failure of any Category I structure leads directly to core damage. The structures with lowest HCLPFs are the containment (HCLPF = 1.11g) and the reactor building (HCLPF = 1.12g). It is important that HCLPFs for Category I stmctures not be compromised by future modifications or additions that could affect safety equipment. 19.8-8 Important Features identified by the ABWR PRA - Amendment 33 1
23A6100 Rw. 9 ABWR standardsafety Analysis soport O event or as a result of control blade relocation during the recovery of a badly damaged core. A possible strategy could be a caution for the operators and/or technical support staff to monitor the power level (perhaps indirectly via the rate of containment pressurization) and enter ADVS procedures as necessary. 19.9.15 Manual Operation of MOVs As noted in Subsection 19.7.3 (3)(a), manual operation of MOVs can be used to improve the availability of decay heat removal. The COL applicant will implement procedures for such an operation. 19.9.16 High Pressure Core Flooder Discharge Valve As noted in Subsection 19D.7.7.5, the HPCF loop B pump discharge valve is in the dawell. Plant procedures should include independent verification that the valve is locked-open following maintenance. 19.9.17 Capability of Containment isolation Valves To insure that containment isolation valve capability does not reduce the containment
,m capability, the COL applicant will demonstrate that the stresses ofcontainment isolation f valves, when subjected to severe accident loadings of 0.77 MPa internal pressure and
)
260 C (500 F) temperature in combination with dead loads, do not exceed ASME l Section III senice level C limits. In addition, the ultimate pressure capability at 260 C (500*F) will be shown to be at least 1.03 MPa. I 19.9.18 Procedures to insure Sample Lines and Drywell Purge Lines Remain Closed During Operation 1 As noted in Subsection 19.8.4.3,it is important that these lines be normally closed j during plant operation. The COL applicant will develop procedures and administrative l controls to ensure the valves are normally sealed closed and that the purge valves have motive power to the valve operators removed. l 19.9.19 Procedures for Combustion Turbine Generator to Supply Power to Condensate Pumps The COL applicant willimplement procedures for manual transfer of Combustion l Turbine Generator (CTG) power to the condensate pumps. Condensate pump support ! systems (lube oil, cooling water) are also supplied power from the CTG. l
/ D
'V)
COL License Information - Amendment 37 19.9 11
I l i 23A6100 RDv. 5 ABWR Standard SafetyAnalysis Report O 19.9.20 Actions to Assure Reliability of the Supporting RCW and Service Water Systems To assure the reliability of the RCW and Senice Water Systems, the COL applicant will take the following action. At least each month, the standby pumps and heat exchangers are started and the previously running RCW and senice water equipment is placed in a standby mode. 19.9.21 Housing of ACIWA Equipment If AC-independent water addition (ACIWA) equipment is housed in a separate building, that building must be capable of withstanding site specific seismic events, flooding, and other site-specific external events such as high winds (e.g., hurricanes). The capability of the building housing the ACIWA equipment must be included in the plant-specific PRA. 19.9.22 Procedures to Assure SRV Operability During Station Blackout To assure the operability of the SRVs during station blackout, the COL applicant will develop procedures for the use of the stored nitrogen bottles as discussed in Subsection 19E.2.1.2.2.2 (b). 19.9.23 Procedures for Ensuring Integrity of Freeze Seals The COL applicant will provide administrative procedures to ensure the integrity of the l temporary boundary when freeze seals are used. Mitigative measures will be identified in advance, and appropriate back-up systems will be made available to minimize the effects of a loss of coolant inventory (See Subsection 19Q.8). 19.9.24 Procedures for Controlling Combustibles During Shutdown The COL applicant shall provide administrative procedures for controlling the combustibles and ignition sources during shutdown operations. (See Subsection 19Q.6 under " Fires During Maintenance"). 19.9.25 Outage Planning and Control The COL applicant shall provide an outage planning and control program to ensure that the safety principle is clearly defined and documented (See Subsection 19Q.10). 19.9.26 Reactor Service Water Systems Definition Senice water systems modeled in the ABWR PRA are described and fault trees presented in Subsection 19D.6.4.2. These include the Reactor Building Cooling Water (RCW) System, Reactor Senice Water (RSW) System, and the Ultimate Heat Sink (UHS).Those portions of the RSW System that are outside of the Control Building and the entire UHS are not in the scope of the ABWR Standard Plant. The COL applicant 19.9-12 COL License information - Amendment 35
23A6100 R v.1 ABWR anatantseveryAmarysieneper .C detailed investigation of the influence and scheduling of the periodic inspections..."This conclusion was also noted in the Reactor Safety Study. (2) The " leak before break" phenomenon is a key consideration whichjustifies a failure rateless than 10'7 per vessel year (Reference 19D.3-8) WASH-1318 further states that the " pressure vessel will have a considerable margin to failure by (a) brittle fracture, even with large postulated initial flaws and (b) that leak-before-break capability is maintained even after a LOCA." This means that long before a crack could propagate to the point that a disruptive failure could occur the crack would propagate through the vessel wall and be detected due to significant leakage. The leak detection system would detect the existence ofleaks and allow shutdown of the reactor to avoid propagation of the crack and vessel failure. (3) Reactor vessels are subjected to periodic inspections, in accordance with Section XI of the ASME Code. This inspection is generally more intensive than that for non-nuclear vessels, and consists of an ultrasonic inspection of weld joints before the vessel goes into service and every 10 years thereafter, supplemented by surface inspections (visual, liquid penetrant test and e magnetic particle test). (4) In recent years, a significant amount of research has been conducted in the area of pressure vesselintegrity, and the factors relating to material specifications which play a key role in material embritthment have been identified and well understood. The RPV material specifications and the RPV irradiation levels for the ABWR produce nil ductility temperature shifts that make the potential for nil ductility failures negligible. (5) Recent GE work on the ABWR design evaluated large RPV bottom head breaks. Structural evaluations showed that loads on equipment and structures were insufficient to cause loss of structural integrity. These results show that the severe accident response for RPV failures would be no worse than for a large break LOCA severe accident. (6) Reactor vessels are designed with a higher degree of protection from pressure transients and temperature events than are non-nuclear vessels. This higher degree of protection is assured by virtue of design measures, including over-pressure relief devices and operational control procedures. (7) Reactor vessels are designed and constructed in accordance with Section III of ( L the ASME Code. These rules are more elaborate than the rules of Sections I and VIII, which are used for non-nuclear vessels. Input Dets - Amendment 31 19 0.3-5
23A6100 Rov. 9 ABWR StandardSafety Analysit Report O (8) Reactor vessels are operated in accordance with the limitations specified in NRC license technical specifications and no such requirements are imposed on non-nuclear vessels. Based on the above considerations, it is concluded that while it is not possible to quantify the probability of RPV failure with great precision, the failure probability of an RPV rupture for the ABWR plant is so low that its explicit inclusion in this analysis would not significantly impact the results. Furthermore, the RPV failure modes that are mechanistically plausible would produce consequences similar to the higher probability LOCA events because of the leak-before-break phenomenon. 19D.3.1.4.2 Loss of Main Control Area Envelope HVAC The HVAC emergency cooling water (HECW) System delivers chilled water to the control building safety-related equipment area cooling coils, reactor building safety-related electrical equipment area cooling coils, and the main control area envelope served by the control room habitability area cooling coils during shutdown of the reactor, normal operating modes, and abnormal reactor conditions.The HECW System consists of three mechanically separated divisions, A, B, and C. Each HECW division provides cooling to the control building safety-related equipment area and the reactor building safety-related electrical equipment area in its division. Also, either division B or C can independently cool the main control area envelope. Power is supplied to each division from independent Class lE sources. Each division of HECW consists of two parallel pumps, and refrigeration units, instrumentation, and distribution piping and valves to the cooling coils. System configurations for each division are illustrated in Figure 19D.3-1,19D.3-2 and 19D.3-3. l The HECW system is capable of removing all heat loads with four of the six units j running and one pump and refrigerator unit from division "A" in standby mode and j one of the four pump and refrigerator units from divisions "B" and "C"in standby At , any given time the division with two pumps and refrigeration units in operation ) provides cooling to the main control area envelope. The design philosophy is that if one of the refrigerators or pumps fails in this division, the standby refrigerator will automatically start and provide main control room area envelope cooling while the l reactor building safety-related electrical equipment area and the Control Building l Safety-related Equipment Area cooling requirements will continue to be met by the remaining refrigerator in the affected division. Cooling water for the HECW refrigerators is provided by the corresponding dhision of the Reactor Building Cooling Water (RCW) System which in turn rejects heat through I the Reactor Service Water (RSW) System to the ultimate heat sink. Each division of the RCW and RSW consists of two parallel trains interfacing through three heat exchangers. RCW and RSW system design capacities are such that one RCW train, one RSW train, 19 0.3-6 Input Data - Amendment 37
23A6100 Rev.1 ABWR senader.1sareryAnalysis neert m Table 19D.3-2 ABWR Maintenance and Test Unavailabilities RCIC 0.02 HPCFB 0.02 HPCFC 0.02 RHRA 0.02 RHRB 0.02 RHRC 0.02 Table 19D.3 3 Recovery of Electric Power (Conditional Probability of Not Recovering Offsite Power or/and Not Restoring One Diesel Generator for ABWR) Failure to Recover Offsite Failure to Recover One Diesel O' Power Before Time T Generator Before Time T Vme interval Complementary Complementary T (Hr) (Hours) Cumulative Conditional Cumulative Conditional 0.5 0-0.5 0.421 0.421 0 0 2.0 0.5-2.0 0.175 0.416 0.66 0.66 8.0 2.0-8.0 0.0175 0.100 0.23 0.35 Failure to Recover Either Offsite Power or One Diesel l Generator Before Time T Time Interval Complementary T (Hr) (Hours) Cumulative Conditional 0.5 0-0.5 0.421 0.421 2.0 0.5-2.0 0.116 0.276 8.0 2.0-8.0 0.004 0.034 Input Data - Amendment 31 190.3 17
23A6100 REv. 9 ABWR standard safety Analysis Report l O RCW/HECW- A SURGETANK Ak A D/G ZONE (A) COOLING COILS (REACTOR BUILDING) M A ESSENTIAL ELECTRICAL EQUIPMENT i ROOM (A) COOLING COILS M (CONTROL BUILDING) O y l FE HECW q8 REFRIGERATOR (CONTROL BUILDING) i i I i O HECW PUMP B LDING RCW & RCW FE HECW REFRIGERATOR q8 (CONTROL BUILDING) 1 I O HECW PUMP
"" RCW &
BU LDlNG RCW O
- Figure 19D.3-1 HECW Division A l 190.3-18 Input Data - Amendment 37
23A6100 Rev. 9 ABWR standardsafety Analysis Report l l s l (ms_) 1 1 ^ Table 19D.4-1 Bases for Core Cooling and Heat Removal Function Event Tree (Branch inputs Not Derived from Subsection 19D.6 Fault Trees) Symbol Description Value
- 1. O Feedwater Unavailability Following a Transient 5.0E-02 (1 FW Pump + 1 Condensate Pump + 1 Cond. Transfer Pump)
It is estimated that 50% of the time feedwater pumps will trip on high water level. In the event of loss of feedwater, failure to manually recover at least
- one pump train is estimated to be 0.1.
- 2. V2 Failure to Recover 1 Condensate and 1 Cond. Transfer Pump 1.0E-01 Similar to the preceding estimate, but at low pressure, failure to manually recover at least one pump train is estimated to be 0.1. For loss of offsite l power events, a combustion turbine generator bus transfer is required and assumed.
4
- 3. W1 Normal Heat Removal (NHR) 1.0E-02 f .s The NHR failure probability of 1.0E-02 is taken from Section D.1.5 of
{d i GESSAR and is a conservative application due to the improved reliability expected from the use of motor-driven feedwater pumps.
- 4. W2 Reactor Water Cleanup System (CUW) 1.0E-01 The CUW System is capable of removing decay heat at high RPV pressures if return water bypasses the regenerative heat exchanger. Failure to manually activate this alternate heet removal system is taken to be 0.1.
Table 19D.4-2 Event Tree Branch Point Values
, (Not Derived From Subsection 19D.6 Fault Tree)
Symbol Description Reference Value Tm Event Frequency Table 19D.3-1 1.0579 (1.0+ ) Transfer from l Fig.19D.4-4) Q Failure to inject with feedwater Table 19D.41 5.0E-02
, U2 Failure to inject with condensate Table 19D.4-1 1.0E-01 I W1 Failure to restore normal heat removal Table 19D.4-1 1.0E-02 W2 Failure to actuate CUW Table 19D.4-1 1.0E-01 Accident Event Trees - Amendment 37 190.4 7
I 23A6100 Rtv. 4 ! ABWR standardsafetyAnalysisReport O' Table 19D.4-3 Branch Point Values for Non-Isolation Event Tree (Not Derived From Subsection 19D.6 Fault Trees) Symbol Description Reference Value Tt Event Frequency Table 19D.3-1 0.62 l PO Failure of SRVs to open GESSAR ll, Table D.1.2-1 1.0E-06 l PC Failure of SRVs to reclose GESSAR ll, Table D.1.2-1 3.0E-03 O Failure to inject with feedwater Table 19D.4-1 5.0E-02 V2 Failure to inject with condensate Table 19D.4-1 1.0E-01 W1 Failure to restore normal heat removal Table 19D.4-1 1.0E-02 W2 Failure to actuate CUW Table 19D.4-1 1.0E-01 Table 19D.4-4 Branch Point Values for Isolation / Loss of Feedwater Event Tree (Not Derived From Subsection 19D.6 Fault Trees) Symbol Description Reference Value Tis Event Frequency Table 19D.3-1 0.18 [ PO Failure of SRVs to open GESSAR 11, Table D.1.2-1 1.0E-06 l PC Failure of SRVs to reclose GESSAR 11, Table D.1.2-1 3.0E-03 QUI Failure to inject with feedwater Table 19D.4-1 0.43 V Failure to inject with condensate Table 19D.4-1 1.0E-01 W1 Failure to restore normal heat removal Table 19D.4-1 1.0E-02 W2 Failure to actuate CUW Table 19D.4-1 1.0E-01 Note: (1) 40% of the initiating event frequency (Tis) represents loss of feedwater events. By definition, O = 1.0 for such events and thus: O = 0.4(1.0) + 0.6(0.05) = 0.43 O l 19 0.4-8 Accident Event Trees - Amendment 34
l 23A6100 Rsv. 2 ABWR standardsareryAnalysisneport r , (m s l subsystem is activated to remove residual heat from the reactor vessel water to cool it to meet shutdown cooling requirements after the control rods are inserted. The subsystem then maintains or reduces this temperature. Reactor water is cooled by pumping it directly from the reactor shutdown cooling nozzles, through the heat exchangers, and back to the vessel via feedwater on one loop and the dedicated RHR return lines on the other two loops. The subsystem is initiated and shut down by operator action. The fault tree for the RHR System in the shutdown cooling mode of operation is presented in Figure 19D.67. Failure rate data used in the analysis are provided in Table 19D.6-3. This tree represents the probability that none of the three loops will be initiated and provide shutdown cooling, given a demand. 190.6.3.3 RHR-Wetwell and Drywell Spray Subsystem The wetwell and drywell Spray Subsystem is employed to remove decay heat and condense steam in both the drywell and wetwell gas columns to prevent overpressurization of the containment. Two of the RHR loops (B and C) provide wetwell and drywell spray cooling. This subsystem provides steam condensation and containment atmospheric cooling by pumping water from the suppression pool, through the heat exchangers, and into the wetwell and drywell spray spargers in the containment building. This subsystem is initiated and terminated by operator action. Drywell spray is enabled in the presence of high drywell pressure. The fault trees representing the wetwell and drywell spray modes of operation of the RHR System are provided in Figure 19D.6-8. The two trees are identical except where specifically noted on the fault tree drawings. The database for evaluation of these fault trees is presented in Table 19D.6-3. Evaluation of this tree provides the probability that neither Loop B nor Loop C will be initiated and provide drywell (or wetwell) spray cooling, given a demand. 19D.6.4 Support System Fault Trees 19D.6.4.1 Electric Power System The station electrical power distribution system is designed to provide reliable power s supply to the ABWR safety-related systems. This power is taken from two offsite sources. In the event offsite sources are lost, three emergency diesel generators and four DC
's./
) batteries are available onsite to meet the power requirements of the safety-related ;
systems. The electrical power is supplied to the safety-related loads from different AC i l i Fault Trees - Amendment 32 19 0.6-5 l 1
23A6100 Riv. 9 ABWR standardSafety Analysis Report 9 and DC buses at different voltage levels. These buses and the onsite emergency sources are arranged into three AC divisions, four DC divisions and four 120V AC unintermptable power supply (UPS) divisions designed with a high degree of independency. Fault trees were developed for each bus supplying essen tial loads. These trees are linked l to the various other safety system fault trees. Events common in different trees are designated with identical acronyms to insure proper common cause failure treatment when the electrical power fault trees are linked to the various system fault trees. The developed fault trees are presented in Figures 19D.6-9 through 19D.6-13. Failure rates used to quantify these fault trees are presented in Table 19D.6-5. 19D.6.4.2 Service Water Systems Essential equipment in the reactor building is cooled by the Reactor Building Cooling Water (RCW) System, which consists of three divisions. Each division is a closed cooling water loop which removes heat from the RHR heat exchangers, HVAC emergency cooling water system refrigerators, diesel generators, and other equipment. Heat is discharged through the RCW heat exchangers to the Reactor Service Water (RSW) System. Each RCW division has two 50% capacity motor driven pumps and three 33.5% capacity heat exchangers. The RSW also consists of three divisions, each of which removes heat from its corresponding RCW heat exchangers and releases it to the UHS. Each division has two 50% capacity motor driven pumps which send UHS cooling water through the RCW heat exchangers. During normal operation, one RCW and one RSW pump in each loop in each division and two RCW heat exchangers in each division are operating. Under these conditions, sufficient cooling capacity is available to provide seal and motor bearing cooling water for the core cooling pumps. Also, sufficient cooling capacity is available to remove heat from the RHR heat exchangers during LOCA if at least two loops are operated with all pumps and heat exchangers. The operating and standby pumps and heat exchangers are interchanged monthly. During accident conditions, the standby pumps and heat exchangers are put into operation to provide additional cooling capacity. The HVAC Emergency Cooling Water (HECW) System receives cooling from the RCW l System through six refrigerators.This system in turn provides cooling to the three reactor building safety-related electrical equipment areas, three con trol building safety-related equipment areas, as well as the main control area envelope served by the control room habitability area HVAC. 19 0.6-6 Fault Trees - Amendment 37
23A6100 Rsv. 9 \ ABWR standardsafety Analysis Report i i The HECW System is compdsed of three loops. Loop A has two pumps and two refrigerators which provide cooling to the control building Division I equipment area i and the reactor building safety-related electrical equipment area. l Loops B and C have two pumps and two refrigerators each. One of these four pumps I and its associated refrigerator is normally in standby mode with the pump / refrigerators ; rotated in and out of senice equally. Loops B and C provide cooling to the reactor building safety-related equipment areas B and C as well as the control building safety-related equipment areas B and C, respectively. The loop with both refrigerators and pumps in operation provides cooling to the main control area envelope. The standby refrigerator and pump in the other loop are available to cool the main control area envelope should one of the two pumps in the operating loop fail. Each division is designed so one pump / refrigerator is sized to provide cooling to the reactor building safety.related electrical equipment area and to the control building safety-reied , equipment area. l The combined RCW and RSW System fault tree for each of the three divisions is presented in Figure 19D.614 and applicable failure rate data are provided in m Table 19D.66. The HECW System fault tree is presented in Figure 19D.623. The HECW failure rate data are included in Table 19D.66. These support system trees are combined with the various front line system and functional fault trees to evaluate core cooling and heat removal function failures. See Subsection 19.9.25 for COL license information requirements. 19D.6.4.3 Instrumentation System Each fault tree contained in this subsection represents the overall comp;-x of instrument channels, signal logics, and transmission networks involved in generating ! either a reactor pressure, reactor level, or drywell pressure signal used to cause a reactor trip or to initiate the various ECCS Systems in the event of an emergency. Fault trees were developed for each signal in each electrical division. These trees are l linked to the various other safetysystem fault trees. Events common to a number of trees are designated with identical acronyms to insure proper common cause failure treatment when these instrumentation trees are linked to the system fault trees. The instrumentation fault trees are presented in Figure 19D.615. Failure rate data used ' to evaluate these trees are provided in Table 19D.67. 19D.6.5 Reactivity Control Fault Trees (n\ 19D.6.5.1 Reactivity Control Functional Fault Tree System fault trees developed to determine the probability of failure to control reactivity and successfully shut down the reactor, given a demand, are presented in this Fault Trees - Amendment 37 19D.6-7
23A6100 Rsv.1 ABWR standardsaretyAnalysis soport O subsection. The functional fault tree shown in Figure 19D.616 integrates each of the individual systems into the overall reactivity control function. Fault trees were developed for the Reactor Protection System, the Control Rod Drive System, the Standby Liquid Control System, the recirculation pump trip, and alternate rod insertion in vaging degrees of detail. The probability of total loss of reactivity control, given a demand, was assessed to be veg low. 19D.6.5.2 Reactor Protection System (RPS) The Reactor Protection System (RPS) is the overall complex ofinstrument channels, trip logics, manual controls and trip actuators that are involved in generating a reactor trip or scram. The system causes a reactor trip for situations which could result in unsafe reactor operating conditions. The RPS is a four-division system which is redundantly designed so that the failure of any single element will not interfere with a required trip. Any single channel or division element operating falsely will not cause a trip because it will trip only one channel or only one of the two solenoids of the scram pilot valves. It combines a very high probability of operating when needed with a very low probability of operating falsely. The Reactor Protection System is a warning and trip system implemented with software logic installed in microprocessors. The critical functions of this system are to: (1) Make the primag decisions related to warning and trip conditions of the individual instrument channels. (2) Make the decision for system trip (emergency reactor shutdown) based on coincidence ofinstrument channel trip conditions. RPS includes detectors, switches microprocessors, solid state logic circuits, relay type contactors, relays, solid state load drivers, lamps, displays, signal transmission routes, circuits and other equipment which are required to execute the functions of the system. The RPS fault tree is presented in Figure 19D.6-24 and applicable failure probabilities are provided in Table 19D.68. Evaluation of this fault tree provides the conditional probability that the RPS will fail to transmit a scram signal given the need. 19D.6.5.3 Control Rod Drive (CRD) System The Control Rod Drive (CRD) System provides rapid control rod insertion (scram) so that no fuel damage results from any abnormal operating transient. An alternative method can be used to insert all the control rods in the event of a failure of the RPS logic. ARI valves can be opened by ATWS logic to vent the air header and cause hydraulic scram. The system is composed of three major elements: (1) Electro-hydraulic fine motion control rod drive (FMCRD) mechanisms. 19 0.6-8 Fault Trees - Amendment 31
~
f \ V J e h 5
;t A HECW A(B,C) FAILS m ,, TO FIG.1 TO PROVIDE E CHILLED WATER l
WHECW1A(B,C) i 5 0 N FAILURE OF DIV LOSS OF 480 V PUMP OR BYPASS LINE A(B,C) RCW TO AC POWER FROM PIPE RUPTURED FAfLS OPEN REFRIGERATOR PROVIDE COOLING BUS C2(D2,E2) FAILURE FLOW VBYPASSA(B,C) WPRA(B,C) -. WRCWA(B.C) EAC6C1(D1.E1) WPPHEAH(BH,CH) (FROM FIG.190.6-14a) FROM FIG.19D.8-12a) 3.44E-06 @ 5 to i i FIRST PUMP OR SECOND PUMP OR PCV P25-F012A(B C) DIFF PRESSURE TEMP SENSOR REFRIGERATOR REFRIGERATOR FAILS FULL TRANSMtTTER P25-TE405A(B,C) FAILURE FAILURE OPEN (NOFO) P25-DPT-007A(B,C) FAILS g FA!LS g O O k=. VPUMPA(B,C) VPUMPD(E F) VPVF012A(B,C) VPR007A(B,C) VTE005A(B'C) (FROM FIG. 3) (FROM FIG. 3) 2.20E45 1.20E-05 2.40E47 f
.D l
g 9 r:GuRE 2 {
= % g y Figure 19D.6-23b HVAC Emergency Cooling Water Fault Tree .
E b O PUMP OR y* TO MG. 2 REFRIGERATOR FAILURE VPUMPA(B,C,D.EJ)
- LOOPS B AND C HAVE TWO PUMP /HXS.
ONE LOOP SET HEREIN AS LCOP B, PUMP HAS BOTH PUMP /HXS RUNNING REFRIGERATOR FAILURE FAILURE LOOP C HAS ITS SECOND PUMP /HX IN STANDBY. WHSPA(8,C,D EJ) U k E O PUMP OPERATOR STANDBY REFRIGERATOR FAllURE MLS TO FAILURE REFRIGERATOR IN $ START PUMP
- MAINTE NANCE
- 5 to WPMHC1A(8,C,0 E,F) VOPPERR4 , , ,-J) - WRFD1A(B.C,D,EJ) WRF MNT-( , . .-J) 1.94E-04 1.00E-03 6.00E 44 6.90E-05 MAN L STAND PUMP SA STANDBY PUMP IN P21F SA CHEC PNM ER y SIGNAL FAILS START. P25-F001F MAINTENANCE (8,C.EJ) FA!LS A(8,C EJ) d c FAILS TO OPEN. CLOSED FAILS
[ b I WPMSTRT-( .. ,-J) WCVH1-(, ,-5) WPMMNT4. , .-J) $ b EMSCONN-(- - ~ '3) (FROM FIG.19b$6-15a) 2.27E 03. 1.40E 04 9.22E45 WPV025A(B.C.D.EJ) 2.20E45 VPRXXXA(B.C,D,EJ) 1.20E-05 g 9 . m RGURE 3 y Figure 19D.6-23c HVAC Emergency Cooling Water Fault Tree j O O O
23A6100 Rty. 4 ABWR standantsareryAnalysis neport p The success or failure of emergency DC power (station batteries) (node DP), and the emergency AC power and/or senice water (node APW) are taken into consideration in Figure 191-1 to account for support system dependencies. Failure of all DC power results in a high-pressure core melt since all control is lost, the high-pressure systems fail, and the reactor cannot be depressurized. The condition ofsuccessful emergency DC and AC power and successful scram is indicated by the ET transfer and is described in detail in Figure 191-2. The condition of successful emergency DC and AC power, but with failure to scram is indicated by the ATWS transfer, and is described in Figure 191-3. The condition of successful emergency DC and failure of emergency AC continues on Figure 19I-1. The next question is whether or not there is a failure to scram (node C). Failure to scram is considered as a Class IV core melt. With successful scram, RCIC (node UR) and firewater (node FA) are the only available means of water injection into the RPV since all AC power is lost. Since station batteries will eventually discharge resulting in loss of RCIC, or if RCIC fails, the reactor must then be depressurized (node X) to allow firewater injection. The loss of emergency DC power (station batteries) results in a high-pressure core melt as shown in Figure 191-1. The firewater system has diesel driven pumps and all needed valves can be accessed and ( operated manually. No support systems are required for firewater operation. See ( Subsection 19.9.21 for COL license information pertaining to housing of'AICWA equipment. The random failure probability of firewater is dominated by operator failure to initiate the system. For the upper branch, where RCIC is successful, the operator has 8 hours before the station batteries expire and RCIC trips. The human error probability (HEP) for this case is 1E-3. For the lower branch, where RCIC fails, the operator has only 30 minutes in which to depressurize the reactor and initiate firewater injection. For this case, the HEP is 0.1. In the event that the firewater diesel fails to start, the operator could make use of a fire truck, but this was not modeled. If the RHR heat exchanger fails (node HX) due to the earthquake,it is presumed that the failure could include a pipe break that could partially drain the suppression pool into the RHR pump room. These core damage sequences are identified with a "P" (e.g., IB2-P). Fission product scrubbing would still be effective in preventing a large release. The effects of possible flooding on equipment operation beyond the RHR room were considered and found to be relativelyinsignificant because of the relatively high HCLPF of the heat exchangers (0.70), the ability of the operator to isolate the break, and the presence of the independent ACIWA (firewater) sptem. 191.3.2 LOSP with Emergency Power and Scram Event Tree In the event tree of Figure 191-2 (ET transfer), there are two similar divisions depending (' on whether or not there is a stuck-open relief valve (node PC). If there is a stuck-open valve, the reactor will eventually depressurize causing loss of RCIC steam supply. The probability of having a stuck-open valve is 2E-3, based on operating experience. If both Seismic Margins Analysis - Amendment 34 191-3
23A6100 Ratt. 9 ABWR Standard Safety Analysis Report O high-pressure injection systems fail, the reactor must be depressurized rapidly for low-pressure system use (LPFL-VI). 191.3.3 ATWS Event Tree Figure 191-3 (ATWS transfer) represents failure to scram, and requires standby liquid con trol (automatic) and operator action to con trol reactor water level with the injection system (s) that are available. The HEP for this action is 0.01. In this ATWS analysis,if high-pressure systems fail, core damage results. No credit is given to low-pressure injection. For an ATWS, the probability of a stuck-open SRV was conservatively increased to 0.1, on the basis ofincreased SRV actisity. 191.4 System Analysis The fault trees used in the seismic system analysis are shown on Figures 191-4 through 191-15. The seismic system analysis calculates the probability of seismic failure and corresponcling system HCLPFs of each of the important systems throughout the seismic l ground acc:leration spectrum. The system HCLPFs are then input to the event trees 1 and combineci with random system failure probabilities and human errors. The seismic l fault trees contain only those components that might be subject to seismic failure. l Random syste.m failure probabilities are taken from the internal events analysis and I include all other components. One of the important ground rules of the seismic margins analysis is that all like comynents in a system always fail together. The reactor protection svstem, control rod drive system, and alternate rod insertion system were not modeled cince the failure of control rods to insert is dominated by the relatively low seismic fragility of the fuel assemblies, control rod guide tubes, and housings. A seismic fault tree for reactivity control is shown on Figure 191-13. The fuel assemblies are the most fragile component. A seismic fault tree for the standby liquid control system is shown on Figure 191-14. Failure of the standby liquid control system is dominated by failure of two components: the pump and boron supply tank. Since the most fragile essential component in the plant is the ceramic insulator in the switchyard, the loss of offsite power dominates the analysis and the availability of emergency power becomes very important. The loss-of-power fault tree (Figure 191-10) is for emergency AC power. In the loss of emergency AC power fault tree, the more fragile components are the diesel generator, transformers, motor control centers, inverter and circuit breaker. The DC power fault tree (Figure 191-11) has two elements: batteries and cable tray. 191-4 Seismic Margins Analysis - Amendment 37
23A6100 R2v. 9 ABWR staaantsaretyAnalysis neport Systems and equipment which require offsite power, such as the feedwater system and condensate injection system, are not modeled since offsite power is presumed to be not available for the core damage sequences. Essential senice water is as 'mportant as emergency power, and its loss would have much the same effect as the loss of emergency AC power. The losser-senice-water fault tree is shown on Figure 191-12. The more fragile components in this system are the senice water pump, heat exchanger, and room air conditioning unit. The senice water pump house, with a HCLPF of 0.60, is also included in this fault tree. Structure failures that could contribute to seismic core damage are shown on Figure 19I-9. In this analysis, any one or more of these structural failures are consenatively presumed to result in core damage. The structures having the lowest seismic capacity are the reactor building and control building. The remainder of the fault trees are for core cooling (Figures 191-4 through 191-8). The more fragile components in these systems are the pumps, heat exchangers, and the firewater supply tank. The condensate storage tank (CST) is not modeled since the ECCS systems that take suction from the CST have automatic switchover to the suppression pool if CST level is low. Valves for the switchover are included in the fault trees. The ACIWA (firewater) system (Figure 191-8) is designed to inject water into the reactor if the ECCS systems are not available. It is also the only means of water injection in case of a station blackout beyond 8 hours. Although firewater is not a Class IE safetysystem, because of the safety function described above, the firewater diesel-driven pump, the firewater tank, valves, and related piping will be seismically designed for a HCLPF of 0.5g. Because of the importance of RCIC in station blackout sequences, differences between the seismic RCIC fault tree and the internal events fault tree are explained below: (1) The internal events fault tree contains basic events that would not be affected by an earthquake, e.g., test and maintenance unavailability. These events contribute to the random failure probability during the seismic event and are included in the random failure part of the seismic analysis. They are deleted from the RCIC seismic fault tree. Seismic Margins Analysis - Amendment 37 19b5
23A6100 R;v. 3 ABWR sundedsatory Antysis nopart O (2) The internal events fault tre: contains common-cause failure events. These are deleted from the RCIC seismic fault tree since a basic rule of the seismic analysis is that all like wmponents within a system fail together. (3) The internal events RCIC fault tree contains separate events for the turbine and for the pump. The seismic fault tree t.ses a combined event, " turbine-driven pump", since that is the assembly for which there is a seismic capacity. 191.5 Accident Sequence HCLPF Analysis Seismic fragility of a stmeture or component is defined as the conditional probability of its failure as a function of peak ground acceleration. The probability model adopted for each ccmponent fragility is the log-normal distribution. The density function for the component fragility, f(g), can be written
/ -
r g > -. 2T 1n 1 1 'Am> f(g) = En* *g exp 2 c forg > 0 c where:
/
Am = median capacity of the component, c
= logarithmic standard deviation of the fragility function, g = peak ground acceleration.
The cumulative distribution of the component fragility, F(g), will then be
- r 3 2' El in F(g) = exp' 1 *dgI
'o E *@c*gi 2 c
191.5.1 Convolution Analysis If a system, S, (or sequence) contains two components (A, B) operating in OR logic, the failure of either component will fail the system (S = A + B), and the cumulative fragility distribution of the system is one minus the product of their complementary cumulative fragility distributions: I 1914 Seismic Margins Analysis - Amendment 33
- -. .. ~~ . . - - . ~ . - _ . . . - . - . - . . . - . - . . . - . - . - . ..- . _ . - . ~ . _ _ _ . -_ .
i { 23A6100 Rrt. 9 ABWR studardsaretyAurysisneport b U HCLPF values for containment isolation for events that could cause containment bypass l as a result of an earthquake, with potential for large releases to the environment. Based on the results of the bypass analysis discussed in Subsection 19E.2.3.3 and shown l on Figures 191-16 through 191-25, the events selected for evaluation in this analysis are: ! (1) Main steam lines (Figure 19E.219a) (2) Feedwater or SLC injection lines (Figure 19E.2-19b) i (3) Reactor instrument, CUW instrument, LDS instrument / sample or j containment atmosphere monitoring lines (Figures 19E.2-19d,19E.2-19e, and - l 19E.2-19f, respectively) (4) RCIC steam supply or CUW suction lines (Figure 19E.2-19e) (5) Post accident sampling lines (Figure 19E.2-19j) (6) Dgwell sump drain line (Figure 19E.2-19j) i l (7) SRV discharge lines (Figure 19E.2-19k) (8) ECCS lines (Figure 19E.219c) (9) Drywell inerting/ purge lines (Figure 19E.2-19i) 1 (10) Wetwell/dgwell vacuum breaker lines '(Figure 19E.2-19g) l The bypass paths for atmospheric control system crosstie lines (Figure 19E.2-19h) require inadvertent opening of two normally closed motor operated valves. Since the seismic analysis does not consider a fail-open mode for normally closed valves, these ! bypass paths are not included in the analysis. In the bypass analysis of Subsection 19E.2.3.3, several potential bypass pathways were l excluded from detailed analysis on the basis of various reasons. The reasons are discussed in Subsection 19E.2.3.3.2 and Table 19E.2-1. These reasons were reviewed to determine whether they remain valid in regard to seismic events. All but one of the reasons are based on configuration details that would not be affected by an earthquake. RHR wetwell and dnwell spray lines were excluded on the basis that the pipes are designed for higher internal pressures than will be seen in actual operation and would
- thus have a vey low probability of breaking. In this case, the seismic event could j .
increase the probability of a break in these lines. However, these pipes have vey high l , seismic capacity (3.0 g) with veg low probability of breaking due to a seismic event. a \ , j An event tree was constructed for each of the above events.These event trees are shown on Figures 191-16 through 191-25. All event trees start with the earthquake as the Seismic Margins Analysis - Amendment 37 19I-9
23Ab 100 Riv. 9 ABWR Staridard Safety Analysis Report 9 initiadng event followed by a core-damaging accident. If there is no core damage there is no large release. The HCLPF and random failure probability are shown for each branch point, and the sequence HCLPFs using convolution and min-max methods are also shown on the figures. l Figure 19I-16 is for suppression pool bypass via main steam lines. Following the earthquake and accident, the question is asked whether or not there is a break in a main steam line outside containment. If there is a break, the question is asked whether or not at least one MSIV in each steam line closes to isolate the break. For the case where there is no break, there could still be a bypass release to the main condenser if a turbine bypass valve is open-unless the MSIVs are closed to isolate the break. The two bypass sequences for this event both have min-max HCLPF capacities of 0.74 g. l Figure 191-17 is an event tree for bypass via feedwater or standby liquid control lines. These lines inject into the RPV and are protected from reverse flow by redundant check valves. These check valves provide isolation of upstream breaks provided that one of the valves closes in the line with the break. The two bypass sequences for this event also have min-max HCLPF capacities of 0.74 g. l Figure 191-18 is for bypass via reactor instrument, CUW instrument, LDS instrument, LDS sample or containment atmosphere monitoring lines. These lines are also protected by check valves, a single valve in each line. The bypass sequence for this case also has a min-max HCLPF of 0.74 g. l Figure 191-19 is for bypass via either the RCIC steam supply line or the CUW suction line. Both of these lines are protected by motor operated isolation valves which require power. Since offsite power is lost due to the earthquake, emergency power is required. The two bypass sequences for this event both have min-max HCLPFs of 0.74 g. Figure 191-20 is for bypass via the post accident sampling lines. These lines are also isolated by motor operated valves. The bypass sequences for this event also have min-max HCLPFs of 0.74 g. Figure 191-21 is for bypass via the drywell sump drain line. This line is protected by a motor operated isolation valve and a check valve. Both components have HCLPF capacities of 0.74 g and the two bypass sequences have min-max HCLPFs of 0.74 g. Figure 191-22 is for bypass via the SRV discharge lines. If there is a break in an SRV discharge line during a core-damaging accident, and that SRV is open, a bypass pathway will exist. In this analysis, it is assumed that the SRV will be open during the accident. The resulting HCLPF capacity for this sequence is the capacity of the SRVelischarge line (0.74 g). 191-10 Seismic Margins Analysis - Amendment 37
23A6100 Rsv. 2 ABWR standardsafety Analysis Report Table 191-1 ABWR Systems and Components / Structures Fragilities (Continued) i
\
Syst . / component MED_CP (Au) LOG _STD ( c) HCLPF* (in g) l
- 4. Reactor Cove is. Cooling (UR) l 1
- Pump (Turbine Driven) 2.0 .45 .70 j
- Steam Gup. Valve (MO) 3.0 .60 .74
- Discharge Valve (MO) 3.0 .60 .,74
- Min Flow Valve (MO) 3.0 .60 .74
- Check Valve 3.0 .60 .74
- RCIC Piping 3.0 .60 .74
- Switchover Valve (MO) 3.0 .60 .74
- 5. Low-Press Core Flooder (V1)
- Pump (Motor Driven) 1.8 .46 .62
- Check Valve 3.0 .60 .74
- Injection Valve (MO) 3.0 .60 .74
- Discharge Valve (MO) 3.0 .60 .74
- LPCF Piping 3.0 .60 .74
- 6. RHR Heat Exchanger (HX)
- Heat Exchanger 2.0 .45 .70
- 7. Reactivity Control Sys. (C)
- Fuel Assemblies 1.4 .35 .62
- CRD Guide Tube 1.8 .36 .78
- CRD Housing 3.5 .46 1.20
- Shroud Support 2.0 .36 .87
- Hydraulic Control Unit 2.0 .60 .63
- 8. SRVs Close (PC, PC1)
- Safety Relief Valve 3.0 .60 .74
- 9. Depressurization (X)
- Safety Relief Valve 3.0 .60 .74
- 10. Level & Press. Control (LPL)
- Safety Relief Valve 3.0 .60 .74
- 11. Inhibit ADS (PA)
- Safety Relief Valve 3.0 .60 .74 V
Seismic Margins Analysis - Amendment 32 19I-13
23A6100 Rev. 9 ABWR StandardSafetyAnalysis Report O Table 191-1 ABWR Systems and Components / Structures Fragilities (Continued) System / Component MED_CP (Au) LOG _STD ( c) HCLPF" (in g)
- 12. Standby Liq. Cont. Sys. (C4)
- SLC Tank 1.8 .46 .62
- SLC Pump 1.8 46 .62
- Valve (Motor Operated) 3.0 .60 .74 l
- SLC Piping 3.0 .60 .74 l
l l l l 1 l 13. Firewater System (FW)
- FW Tank 1.5 t .46 .51
- Pump (Diesel Driven) 1.5 .46 .51
- Injection Valve (Manual) 2.0 .60 .50
- FW Piping 2.0 .60 .50
- Valve (Manual) 2.0 .60 .50
- HCLPF = A, x exp (-2.326x c) t Firewater tank may be designed and built to a lower capacity if provision is made for a pumper truck housed in such a manner that it will survive a SSE and a hose that will reach an alternate water supply.
O 191-14 Seismic Margins Analysis - Amendment 37
i 23A6100 Rev. 9 ABWR standardsafety Analysis Report a Table 1912 Seismic Margins for ABWR Accident Sequences (Convolution Method) : l Accident With Random Failure Without Random Failure i Sequence HCLPF MED_ CAP LOG _STD HCLPF MED_ CAP LOG _STD l Number * (in g) (Am) @c) (in g) (Am) @c) 1 0.64 1.14 0.25 0.64 1.14 0.25 ' 1 2 0.89 2.02 0.35 0.89 2.02 0.35 3 0.81 3.00 0.56 0.81 3.00 0.56 4 1.21 3.34 0.44 1.21 3.34 0.44 l 5 0.77 1.40 0.26 0.82 1.43 0.24 6 1.02 2.09 0.31 1.03 2.10 0.30 7 0.98 3.01 0.48 0.99 3.01 0.48 8 1.29 3.34 0.41 1.29 3.34 0.41 9 0.73 1.23 0.23 0.73 1.23 0.23 ! 10 0.94 2.01 0.33 0.94 2.01 0.33 11 0.77 2.37 0,48 0.77 2.37 0.48 12 1.21 2.79 0.36 1.21 2.79 0.36 13 1.02 2.30 0.35 1.02 2.30 0.35 14 1.33 2.65 0.30 1.33 2.65 0.30 l 15 0.99 1.80 0.26 1.01 1.81 0.25 16 1.13 3.04 0.43 1.14 3.04 0.42 l 17 1.16 3.05 0.42 1.16 3.05 0.42 18 1.46 4.16 0.45 1,46 4.16 0.45 19 0.96 1.68 0.24 0.97 1.69 0.24 20 0.89 3.00 0.52 0.90 3.00 0.52 l 21 0.95 2.82 0.47 1.08 3.04 0.44 22 1.26 4.05 0.50 1.39 4.16 0.47 23 0.87 2.98 0.53 0.90 3.00 0.52 24 0.80 1.44 0.25 0.80 1.44 0.25 25 0.96 1.69 0.24 0.97 1.69 0.24 See event trees. f G Seismic Margins Analysis - Amendment 37 19I-15
23A6100 Rsv. 9 ABWR standardSafety Analysis Report O Table 191-3 Seismic Margins for ABWR Accident Classes (Convolution Method) With Random Failure Without Random Failure Accident HCLPF MED_ CAP LOG _STD HCLPF MED_ CAP LOG _STD Class (in g) (Am) (Oct (in g) (Am) (@c) IA 0.75 1.68 0.35 0.76 1.68 0.34 182 0.64 1.14 0.25 0.64 1.14 0.25 IC 0.90 1.44 0.20 0.92 1.46 0.20 l ID 0.78 1.32 0.23 0.82 1.36 0.22 IE 1.02 2.30 0.35 1.02 2.30 0.35 IV 0.70 1.13 0.21 0.71 1.14 0.20 lA-P,IE-P 0.89 1.46 0.22 0.89 1.47 0.21 O l l l 191-16 Seismic Margins Analysis - Amendment 37
23A6100 Rtv. 9 ABWR standardsatery Analysis neport l% t 1 V Table 191-4 HCLPF Derivation for the ABWR Accident Sequences (MIN-MAX Method) Sequence 1 : APW'FA * (0.60g+1.6E-3)*(0.50g+1.0E-3) --+
- 0.60g, 0.50g
- 1.6E-3 Sequence 2 : HX*APW'FA --> 0.70g*(0.60g+1.6E-3)*(0.50g+1.0E-3) -+ 1
- 0.70g Sequence 3 : X*APW -> 0.74g *(0.60g+1.6E-3) -+
- 0.74g )
Sequence 4 : HX*X*APW -+ 0.70g *0.74g*(0.60g+1.6E-3) -+ l
- 0.74g Sequence 5 : FA*UR*APW -+ (0.50g+1.0E-1)*(0.70g+6.0E-2)*(0.60g+1.6E-3) -+
- 0.70g, 0.60g *6.0E-2 Sequence 6 : HX*FA*UR*APW -*
- 0.70g*(0.50g+1.0E-1)*(0.70g+6.0E-2)*(0.60g+1.6E-3) -+
n : 0.70g Sequence 7 : X* UR*APW -> 0.740g *(0.70g+6.0E-2)*(0.60g+1.6E-3) -+
- 0.74g Sequence 8 : HX*X'UR*APW -+ 0.70g*0.74g*(0.70g+6.0E-2)*(0.60g+1.6E-3) -+
- 0.74g Sequence 9 : C*APW -+ 0.62g *(0.60g+1.6E-3) -+
- 0.62g Sequence 10 : HX*C*APW -> 0.70g*0.62g*(0.60g+1.6E-3) -+
- 0.70g Sequence 11 : DP --+ 0.749
- 0.74g I Sequence 12 : HX*DP -> 0.70g*0.749 1
- 0.74g Sequence 13 : SI -+ l
- 1.11g l
Sequence 14 : HX*St -+ 0.70g*1.11g -+ l
- 1.119 l l Sequence 15 : V1 *UH'UR -+ l l :0.62 g * (0.62g +2.7E-3 ) * ( 0.70g + 6.0E-2) -+
- 0.70g, 0.62g *6.0E-2 Seismic Margins Analysis - Amendment 37 191-17
l 1 l 23A6100 R1v. 9 ABWR StandardSafety Analysis Report 1 Ol Table 191-4 HCLPF Derivation for the ABWR Accident Sequences (MIN MAX Method) Sequence 16 : X'UH'UR --> 0.74g*(0.62g+2.7E-3)*(0.70g+6.0E-2) -+ l l
- 0.74g '
l Sequence 17 : V1
- U H'PC -->
l : 0.62g *(0.62g +2.7E-3)*(0.74g +2.0E-3) -+ l
- 0.74g,0.62g*2.0E-3 Sequence 18 : X* UH'PC -> 0.749*(0.62g+2.7E-3)*(0.74g+2.0E-3) -->
- 0.74g Sequence 19 : UR* UH'C --+ (0.70g+6.0E-2)*(0.62g+2.7E-3)*0.629 ->
- 0.70g, 0.62g *6.0E-2 Sequence 20 : PA*C -+ (0.74g+2.4E-3)*0.629 --+
- 0.74g, 0.62g *2.4E-3 Sequence 21 l
- UH* PC1*C -+ (0.62g+2.7E-3)*(0.74g+1.0E-1)*0.62g -+ '
- 0.74g, 0.62g
- 1.0E-1 i Sequence 22 : PA*PC1 *C -+ (0.74g+2.4E-3)*(0.74g+1.0E-1)*0.62g -+
- 0.74g Sequence 23 : LPL*C --* (0.74g+1.0E-2)*0.62g -*
- 0.749,0.62g*1.0E-2 Sequence 24 : C4*C -> (0.62g+1.4E-2)*0.62g -+
- 0.62g Sequence 25 : UH*C4*C -+ (0.62g+2.7E- 3)*(0.62g+1.4E-2)*0.629 -+
- 0.62g O
191-18 Seismic Margins Analysis - Amendment 37
f% a - ? b i es e q HCLPF 1.11g 0.084g 0.749 0.60g 0.629 0.70g 0.74g 0.50g 0.70g g
. Sequence HCLPF en Random - - - 16E43 -
6 OE-02 - 1 E-03.1 E-01 - t Fad. Prob. 7. w SEISMIC STRUCTURAL OFFSITE EMERGENCY EMERGENCY SCRAM OR RCC DEPRESS. FIRE WATER RHR HEAT CLASS Min 44ax Convolugon EVENT INTEGRITY POWER OC POWER AC POWER / ARI EXCHANGER 3 EMERGENCY DOES NOT $ SERV WATER RUPTURE a , j SE St LOP DP APW C UR X FA HX i a W " ET Trans Trans f I ATWS Trans Traem Ox 4 I 18 2 0.60g 0.64g g i 2 IB2-P 070g 0 893 3
, IA 0.74g 081g I 4 .
LA-P 0.74g 1.21g S h Ox 5
, to 070g 0.77g I e to-P 0.70g 1.02g 7
, tA 0.74g 0 98g I 8 1A-P 0.74g 1.29g
- IV O 62g
, 0.73o I 'O IV-P 0.70g 0 94g l' =
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g I 14 IE-P 1.11g 1.33g Q h
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e T a-i = Figure 191-1 Seismic Support State Event Tree
}f
5 b @ HCLPF 0.74 g 0.70g 0.62g 0.74g 0.62g N Random Fail. Prob. 2E-03 6E-02 2.7E-03 - RCIC HPCF DEPRESS. LPFL CLASS Min-Max Convolution CRAi OE ET PC UR UH X V1 OK OK OK 15 ID 0.70g 0.999
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16 2 m IA 0.74g 1.13g 5 Q R s e OK OK ? W 17 3 ID 0.74g 1.16g n 5 e a e s-
- 18
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a u Tw P ~ Figure 191-2 LOOP with SCRAM Event Tree 1a O O O
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23A6100 Rev. 9 Standard Safety Analysis Report ( 19P Evaluation of Potential Modifications to the ABWR Design Refer to Attachment A of the " Technical Support Document for the ABWR" (Revision 1 - December 1994). (See letter from J.F. Quirk, GE, to R.W. Borchardt, NRC, dated December 21,1994). I l l l
\
Evaluation of Potential Modifications to the ABWR Design - Amendment 37 19P4'2
& Table 190C-1 Loss of Offsite Power Precursors 3h 5: tn
{ Event Description Applicable ABWR Features
$ Indian Point 2 and Yankee Rowe (11/9/65) " Great Northeast Blackout" ABWR has two independent offsite power $ sources.These are backed up by three y physically and electrically separate trains of a Class 1E AC power each containing an
[ emergency diesel generator. These are further
& backed up by a permanent onsite Combustion i Turbine Generator (CTG) which is capable of
{ powering any one of the three trains if all three 3 diesels were to fail. The CTG is also capable of
@ supplying power to non-safety busses such that condensate pumps can also be used to provide lg!i reactor coolant make up.
2 Point Beach 1 (2/5/71) Loss of all transmission lines, failure of See discussion of Indian Point 2 and Yankee {* three transformer differential relays, Rowe (11/9/65). N g causing transformer lockout. k P Ginna (3/4/71) Plant siding fell into 34.5 kV line ABWR has two independent transformers
$ connecting sole startup transformer. powered by two independent offsite power ll!
it supplies which reduces the probability of losing 'o g offsite power. In the event of losing offsite g power, features described under Indian Point 2 g and Yankee Rowe (11/9/65) can mitigate the y event. p Palisades (9/2/71) Transmission line fault, isolation breaker See discussion of Ginna (3/4/71). g failure. Backup relay isolated 345 kV bus.
- o. 3 a
l;, San Onofre 1 (6/7/73) 138 kV auxiliary transformer out for maintenance. Ground fault operated ABWR uses three auxiliary transformers. Each powers one of the three Class 1E and non-1E
" R. i differential relays, de-energizing other buses. In addition, a reserve auxiliary La auxiliary transformers. transformer is available to power all three Class 1E buses. CTG is also available which can
{ h power 1E and non-1E busses without using the E auxiliary transformers. k fft 5 %
~ =
;; Table 19QC-1 Loss of Offsite Power Precursors (Continued) b Event Description Applicable ABWR Features l Oconee 1 (1/4/74) 230 kV switchyard isolated,100 kV offsite The ABWR also has two sources of offsite source remained energized to supply power.
power to the plant. 3 Fort Calhoun (3/1395) Sole 161 kV backup offsite transmission See Indian Point 2 and Yankee Rowe (11/9/65) g, line out for maintenance. 345 kV output and Ginna (3/491). y breaker tripped (faulty protective relays), q opening remaining connection to offsite
,jp power. Offsite power could have been supplied from 345 kV switchyard by l(2.
opening generator disconnects. 13 Turkey Point 4 (5/16/77) Loss of Offsite Power (LOOP) See Indian Point 2 and Yankee Rowe (11/9/65). Si
& Connecticut Yankee (6/26R6) Protective relays operated when lines See Indian Point 2 and Yankee Rowe (11/9/65).
3 were re-energized after service, causing %
? LOOP. M 3 S E Indian Point 2 (7/13/77) LOOP due to lightning strikes. Emergency See Indian Point 2 and Yankee Rowe (11/9/65). 8 $ Diesel Generators (EDGs) operated. R a 5 ABWR has two offsite power sources so "
g St. Lucie 1 (5/14R8) Substation switching error. q probability of one switching error resulting in y loss of all offsite power is low. But if it were to
$ occur, mitigation features exist as discussed in & Indian Point 2 and Yankee Rowe (11/9/65).
O y Turkey Point 3 (4/4R9) Loss of all 7 transmission lines due to See Indiar. Point 2 and Yankee Rowe (11/9/65).
}a weather. g Davis Besse (4/19/80) One EDG out for maintenance. One 13.8 ABWR has two sources of non-1E power. A g. @ kV bus connected, other energized but not ground fault on one would not result in loss of B. ! connected. Ground fault on 13.8 kV bus all non-1E power. In addition,if all non-1E D k caused loss of non-nuclear instruments. power were to be lost, no valves connected to k
[ 3 Air was pulled into DHR pump, and pump the RHR System would automatically cycle and was stopped by operator. Pump vented cause loss of NPSH to any RHR pump. Also, the g 2.
& and restarted after 2-1/2 hours. ABWR has three independent (100%) RHR j.
2 Systems such that loss of one would not result go in loss of the ability to remove decay heat. 'g 4 O O O
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~r No. TITLE E ,
1 NOTES, CONTENTS 2 2 HECW PUMP (A&O) (1) 3 HECW PUMP (A&D) (2) 4 4 HECW PUMP (A&D) (3) 5 HECW REFRIGERATOR (A) COMPRESSOR 6 HECW PUMA (G&E) (1) D 7 HECW PUMP (B&E) (2) 8 HECW PUMP (8&E) (3) REF 9 HECW REFRICERATOR (A) RE ACTCR BUILDING j COOLING WATER INLET VALVE (P21) HECW CH.LLED WATER IN-CUT PRESSURE 2 10 OlFFERENCE CCNTROL VALVE y 11 ANNUNCI A TOR C B 1 A I l 137C9445\0075001 Of N
y 4 3 2 5
>MLESS OTHERW!SE NOTED. THr EQUIPMENT NUMBERS
..;HOWN ON 1HIS DI AGR AM ARE PREFIXED WITH P25 JNLESS OTHERWISE NOTED. THE POWER SUPPLY FOR
- UPPORTING EQUIPMENT IS FROM THE ESSENTIAL PCWER 3" FIXE 0 NUM8ERS IN TH!S DI AGRAM INDICATE THE iWGR FUNTION AL NUMOERS.
THE LOGIC OF REFRIGERATOR 15 OMITTED BECAUSE IT lEPENOS ON H ARDWARE STRUCTURE.
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- RENCE DOCUMENT p, A { 3 t.. A .,jg j; lVAC EMERGENCY COOLING WATER SYS P&O P25-1010 A9 7ture Car"d 1E ACTCR BUILD:NC COOUNG WATER SYS P&D P 21 -10:0
?EACTOR BUILO:NG COOLING WATER SYS/ P21/P41- 1030 VEAC TOR SERVICE WATER SYS 180 I P25-1030 FIGURE 7 3-9 HVAC EMERGENCY COOLING WATER SYSTEM 180 (Sneet I of II) Amenament 31 eBwRIS$hR 2]A6100 Rev 9 21-259 9609090230- C -- L
7 6 S TPU E 3 ~ R10-1030 J TPU t=77 s RESTART TIMER OUTPU T + tal7 s IPO ( ) --*- t=15 s 3 MCRP HECW REFRIGER ATOR SELECT 0 = 0: CON rD, A AU TO r D COS M ANU AL :
'O __
A CON T'O. 0. AU TO r
~
7 k __ (
/
~
~
J ( MCRP _ HECW REFR'CER ATOR ( A) - r START ' CSI AUTO STOP PULL LOCK j ) ) ) = T C f :I-,y7 ' 5 % __ __.J s > LCP ) ) HECW PUMP (A) START CS2 AUTO STOP ) ) )
#88 WCC R/8 ESSENTIAL ELEC q B
EOUIP ROOM (A) VENTILATOR ( A) RUNNING d )
#88 MCC R/8 ESSENTIAL ELEC EOutP ROOM (A)
VENTILATOR (8) RUNNING uCC FECW PUMP (A) \ RUNNING SIGN AL
/
SH 4 A HECW PUMP (A AND D) 137C9445\8075002
i
%-mesaO y 4 3 2 CONTROL SWITCH SELECT SWITCH [0: CONT ] ] , . 7 r , ,c, c ., , ,
M b.2 i _!! > iN POSITION L A: AUTO J j 93 p "9 SH 3 20 5 k
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g92 3
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/
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r, g - 40 2 to m < m u < w , lN1/ PULL NI/ NI/ o h _[ HECW PUVP (A) A'J TOM ATIC ] [ ( RUNNING SIGN AL y {U LOCK 0 Also Aval!abia on N + ~ AptrtuTe Card en s 3 COSx1 CSIX2 CS2X2 [ MCR OPER ATICN AL SWITCH } ( IN (STOP) POSITION j OBSERVATION AL ME ASURES SH 4 USE i SPECIFIC A TION _gCA TION [ 'OPER A TION AL SWI TCH IN [ PULL LOCK] POSITION SH 3 SH 5 ANNUN CI A TOR N O. INDICATOR LOC A TION LOCAL OPER A TION AL SWITCH ) IN (STOPj POStTICN j SH 3 OTHERS INSTAN T INTERRUP. NEED I TION COUNTER MEAS NOT NEED IN CHING NEED OPER A TION NOT NEED VALVE SE AT NG OPEN SIDE FORM CLOSED SiOE THERM AL ExlST O YP ASS NOT EXIST SWCR PWR SUPPLY EXIST OBSERVA TION NOT EXIST ELECTRICAL SOL AC VI V".V PWR SUPPLY DC Vl P25-1030 FIGURE 7 3-9 HVAC EMERGENCY COOLING WATER SYSTEM 180 (Sheet 2 o' 11) Amenc>nent 37 ABwRIS$AR 2]Afl00 Res 9 21~ MO 9609090230 ,
- -l
k d % (_ A l 7 6 5 l LCP LOCAL STOP P8 l STOP E 1
\ R10-1030 '
2 70% BUS VOLTAGC ) =
=
___J HECW PUVP (A) \ % _
~ ~*
AUTOMATIC RUNNING SIGNAL / SH 2 Q d (WO) - OPERATIONAL SW1TCH \ IN [ PULL LOCK] POSITION / SH 2
#52 P/C HECW REFRICERATOR (A) \
i
~
COMPRESSOR STOP
/ _
LOCAL OPERAIlONAl SWITCH \ __ IN [STOP] POSITION / < SH 2
- TS6058-2 MCRP RETURN CHILLED WATER LOW
\ q -
s' TEMP / V ; l #52 P/C HECW REFRICERATOR (D) \ -__ 7 _ COMPRESSOR RUNNING / J SELECT SW IN A:STD8p \ POSITION . D: AUTO f SH 2 8 r HECW PUMP (A) STOP SIGNA SH 4 RtO-1030
<70% DUS VOLTAGE =
A HECW PUMF 137C9445\8075003
.s %
1
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y 4 3 2 HECW REFRIGERATOR ( A) COMPRESSOR STOP SIGN AL CCCIA HECW PUMP (A)
@I
^
l) O = RUN I S'" 2 CON TROL SW!CH d=--
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r % P. '. l ,,
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/ '30 hVO!!dhlD CG O >;panurO cara PBX 2 OBSERVAItON AL ME ASURES USE SPECIFICATION LOC A TION MCRP me g
@' h' as @2 @2 LCe sa m?
ANNUNCI A TOR NO INDICATOR LOCATION OTHERS INSTANT IN TERRUP- NEED O T.ON COUN TER MEAS NOT NEED INCHING NEED OPERATION NOT NEED VALVE SEATING OPEN S!DE FORM CLOSED SIDE THERMAL EXIST B YP ASS NOT exist SWGR PWR SUPPLY EXIST COSERVAT-ON . NOT EXIST O ELECTRICAL SOL AC V VLV PWR SUPPLY DC V P25-1030 D D) FIGURE 7.3-9 HVAC EMERGENCY COOLING WATER SYSTEM 180 (Sheet j of 11) Ame,enent 31 ABwRISSAR 2]A6l00 Rev 9 21-269 l 9609090230 (i -.
es s 7 6 5 l E LOCAL OPERATIONAL SWITCH \ _ IN (STOP) POSITION
/ O SH 2 _
%)
HECW REFRIGERATOR (A) FAULTY SIGNAL r9 r,
#88 MCC HECW PUMP (A)
RUNNING fpU D FIS003A LOCAL 7 j 7 CHILLED WATER LOW FLOW
\ ~_
t=60 s
/
TS60$ A- 2 MCRP RETURN CHILLED WATER HlCH
\
TEMPERATURE /
#52 P/C J HECW REFRIGERATOR (D)
CCMPRESSOR RUNNING
~_
#30 MCC q)
\
( HECW PUMP (0) OVERLOAD TR!P
/
MCC
< HECW REFRICERAIOR (D) \
~
FAULTY SIGN AL
/
B HECW PUMP ( A AND D) A ( 137C9445\B075004
M.wW 4 3 2 _[ HECW PUMP (A) -( STOP SIGNAL SH 3 i s?E
- O.. *T W .s.;/
pg . ,i "~'= P 3 ;, ({" pa .~ p j ., .
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CcNTROL SMTCH [ IO AVOll.lDIO CU
,c ,nc Card OBSERVATIONAL ME ASURES USE SPECIFICATlON LOCA IlON )
i i _[ HECW PUMP ( A) ) 1 -( RUNNING SICN AL ,/ SH 2 ANNUNCI A TOR NO INDiC A TOR LOC A TION l OTHERS INSTANT INTERRUP- NEED TION COUNTER MEAS NOT NEED INCHING NEED OPER A TION NOT NEED VALVE SEATING OPEN SIDE FORV CLOSED SiOE THERM AL E XIS T BYPASS NOT ExlST SWGR PWR SUPPLY E X.S T OBSERVATION NOT EXIST ELECTRICAL SOL AC V VLV PWR SUPPLY DC V P25-1030 FIGURE 7 3-9 HVAC EMERCENCY COOUNG WATER SYSTEM 180 (Sheet 4 of 11) Amenoment J/ ABwRIS$bR 23A6100 Rev 9 ?!-162 9609090230- l
.s +
7 6 5 ( #88 MCC E HECW PUVP (A) RUNNING
/
#52 M/C RCW PUMP (A) RUNNING *
/ =
O
#52 M/C g =
RCW PUMP (D) RUNN!NG LMT SW'TCH O HECW REFRiGERATCR (A) \ SUCTICN VLV FULLY CLOSED / OPERATION AL SWITCH IN \ (PULL LOCX] POSITION f SH 2 TPU c
#52 P/C (_
t=60 $ HECW REFRIGERATOR (A) \ _ q g COMPRESSOR RUNNING / Q G
~
( e 3
< HECW REFRIGERATOR FAULTY SIGN AL
/ (A) \ /
IS LCP 7 CHILLED WATER
\ d OJTLET TEMP =
H!CH [ T 5
) J J B
b TS LCP (WO) ~i = CHILLED WATER OU TLET
\ f i TEMP LOW / ON H TO 8 HOWE AND A HECW REFRICERATOR ( A) COMPRESSOR 137C9445\B075005 l .s ~~r I
l
=
y 4 3 2 ; l l l \ l ! i l l l CONTROL SWITCH gpy, p zm ae 6 :. .a.e I /( rf m i ' ,n. 5: J 5 p:J.a.;a f & HECW REFR:CERATOR ( A) COMPRESSOR STARTUP SIGNAL; (');h {h 1 f A to Availabic on l Aperturo Card 08SERVAflONAL MEASURES USE SPECIFICATION l LOCATION ANNUNCI A TOR NO INDICATOR LOCAllON V (WO) d l l l OTHERS l INSTANT INTERRUP-. NEED TION COUNTER MEAS NOT NEED INCHING NEED
- w REFRICERATOR (B), (C), (0), (E), (F) SUFFlX A CHANGES OPER A TION NOT NEED
$. O, E, F AND OIHERS ARE SIMILAR TO TH AT OF THIS SHEET. OPEN SLOE @, REFRICERATOR (B), (E) IS RCW PUMP (8 OR E); VALVE SEATING $RIGERATOR (C), (F) IS RCW PUVP (C OR F) FORM CLOSED SiOE I l THERMAL E XIS T l l BYPASS NOT EXIST i ! EXIST SWGR PWR SUPPLY l OBSERVATION NOT EXIST ELECTRICAL SOL AC V VLV PWR SUPPLY DC V h P25-1030 I I FIGURE 7.3-9 HVAC EMERCENCY COOUNG WATER SYSTEM 180 (Sheet 5 of 11) Amenenent 37 ABwRissbR 2]A6100 Rev 9 21-263 9609090230- ] = - l
7 6 5 TPU { 7 _ R10-1030 IPU t=77 s RESTAR T TIMER OUTPUT :
/ J t=1/ s TPU 1
(WC) --+- t=15 s 3 MCRP [' HECW REFRIGEffAATGT-r:
, L_W set.ECT E CONT'0; B: AUTO =
D CCS M ANU AL = ,9 =
- 8. CON T'D; E. AU TO rQ 7 ( _
(
~
MCRP __ HECW REFR'GERATOR (8) r
\
START CSI AU TO STOP j ) ) ) PULL LOC < C -I s N -_
-J /
4 \ LCP ) ) 1 l HECW PUMP (8) STARI CS2 AUTO STOP i ) ) )
#88 MCC R/S ESSENTIAL ELEC -] 3 ECUIP ROOM (B) j 8 VENTIL ATOR ( A) RUNN NG d O _!
#88 VCC d '
R/B ESSENTIAL ELEC
< ECUIP ROOM (8)
VENTILATOR (8) RUNNING ON HECW PUVPS (C; HOWEVER, HECW PUC MCC AND HECW PUMP (Ci HECW PUMP (B) \ RUNNING SIGNAL
/
SH 8 A HECW PUMP (8 AND E) ( ~ l 137C9445\B075006 ww
4 3 2 f i k , e ', >,. $ . n ' p.... < q r:, ; , > es,
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