ML20065B712
| ML20065B712 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 03/31/1994 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20065B710 | List: |
| References | |
| NUDOCS 9404040097 | |
| Download: ML20065B712 (58) | |
Text
ABWR DESIGN CERTIFICATION MEMORANDUM ABWR SSAR/CDM CROSS REFERENCE MATERIAL Revised - March 31,1994 GE Nuclear Energy I
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t TABLE OF CONTENTS Page No.
Introduction 3 >
SSAR/ Tier 1 Cross Reference Material Table 1) Core Cooling Analysis 4 Table 2) Containment Pressure / Temperature Response 9 Table 3) -Transient Analysis 12 Table 4) Radiological Analysis 16 Table 5) Overpresssure Protection 17 Table 6) Flooding Protection 19 l Table 7) Fire Protection 25 Table 8) ATWS Analysis 26 Table 9) Generic Safety issues 30 Table 10) TMI Issues 45 q
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Introduction As part of their review of the ABWR Design Certification application, NRC staff have requested that GE provide a cross reference between the following:
4 a) The plant safety analyses and other safety-related issues described in the Safety Analysis Report (SAR) and b) The inspections, tests, analyses and acceptance criteria (ITAAC) included in the Certified Design Material (CDM) being prepared by GE for the ABWR.
This cross reference is intended to be a review aid which will assist the NRC staff in concluding that plant characteristics of particular importance to safety will be adequately confirmed by the CDM ITAAC process; i.e...the as-built facility has the characteristics which are consistent with the plant safety analyses assumptions and the resolutions of the plant safety-related issues described :n the SAR.
The folowing Tables 1 through 10 provide cross references for ten safety-related areas described in the SAR. GE believes the material in this memorandum is aimed at assisting the staff review process; consequently, it is informal in that it is not intended to be part of the SSAR or Design Control Document. These cross references are consistent with ABWR SSAR Tables 14.3-1 through 14.3-10.
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Table 1 Core Cooling Analysis SSAR Verifying SSAR Entry Parameter Value ITAAC 6.3.3.5 Following a LOCA the RHR System is Automatically Directed to the LPFL Mode ----
2.4.1 6.3.3.7.4 The Safety Related Systems Will Operate as Designed with the Loss of AllOffsite AC Power ----
2.12.'! 3 ,
Table 6.31 Low Pressure Flooder System Vessel Pressure at which Flow i May Commence '
(MPaD -- vessel to drywell) 1.55 2.4.1 1
Min. Rated Flow (m3/hr per pump) 954 2.4.1 -l I
at Vessel Pressure l (MPaD -- vessel to drywell) 0.275 2.4.1
)
Initiating Signals Low Water Level ----
2.4.1 l or High Drywell Pressure ----
2.4.1 Maximum Allowable Time Delay from Low Pressure Permissive Signal to injection Valve Fully Open (sec) 36.0 2.4.1 J
4
Table 1 Core Cooling Analysis (Cont.)
SSAR Verifying-SSAR Entrv Parameter Value ITAAC Table 6.3-1 Reactor Core Isolation Cooling System Vessel Pressure at which Flow May Commence (MPaD -- vessel to the air space 8.12 2.4.4 of the compartment containing the water source for the pump suction)
Min. Rated Flow (m3/hr) 182 2.4.4 at Vessel Pressures (MPaD 8.12
-- vessel to the air space of the to compartment containing the 1.03 2.4.4 water source for the pump suction)
Initiating Signals Low Water Level ----
2.4.4 or High Drywell Pressure ----
2.4.4 Maximum Allowable Time Delay from initiating Signal to Rated Flow Available and Injection Valve Fully Open (sec) 29.0 2.4.4 l (Design Des. Only) - ,
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Table 1 Core Cooling Analysis (Cont.)
SSAR Verifying SSAR Entrv Parameter Value ITAAC Table 6.3-1 High Pressure Core Flooder System Vessel Pressure at which Flow May Commence (MPaD -- vessel to the air space 8.12 2.4.2 of the compartment containing the water source for the pump suction)
Minimum Rated Flows (m3/hr per subsystem) 182 to 2.4.2 727 at Vessel Pressures (kg/cm2d 8.12
-- vessel to the air space of the to 2.4.2 compartment containg the water 0.686 source for the pump suction)
Initiating Signals Low Water Level ----
2.4.2 or High Drywell Pressure ----
2.4.2 Maximum Allowable Time Delay from initiating Gignal to Rated Flow Available and injection Valve Fully Open and Power Available at the Emergency Busses (sec). 16.0 2.4.2 ,
Maximum Emergency Diesel Generator Startup Time (sec) 20.0 2.12.13 6
Table 1 Core Cooling Analysis (Cont.)
SSAR Verifying SSAR Entrv Parameter Valve ITAAC Table 6.3-1 Automatic Depressurization System Total Number of Relief Valves with ADS Function 8 2.1.2 Min. Flow Capacity (kg/hr x 106) 2.903 2.1.2 at Vessel Pressure (MPaG) 7.76 2.1.2 initiating Signals Low Water Level ----
2.1.2 and High Drywell Pressure 2.1.2 or High Drywell Pressure Bypass Timer Timed Out ----
2.1.2 Time Delay (sec) 480 2.1.2 Delay Time from AllInitiating Signals Completed to the Time Valves are Open (sec) 29.0 2.1.2 7
9 -
& .a .;.4 ,4+ + p. .w a &..: J =ms e AJ4 ?_a- _ , La 4 Table 1 Core Cooling Analysis (Cont.)
SSAR Verifying SSAR Entrv Parameter y_ahie ITAAC Table 6.3-3 The RHR Subsystems are Divisionally Separated ----
2.4.1 The HPCF Subsystems are Divisionally Separated ----
2.4.2 -
RCIC Operation Does not Required AC Power ----
2.4.4 A Single Failure Will not Prevent the Operation of More Than One ADS Valve ---- 2.1.2 Table 6.3-4 LOCA Break Sizes 2 985 2.1.1 Steamline (cm )
2 839 2.1.1 Feedwater Line (cm )
RHR Shutdown Cooling Suction >
2 792 2.1.1 Line (cm )
RHR Injection Line (cm2) 205 2.1.1 High Pressure Core Flooder (cm2 ) 92 2.1.1 2 20.3 2.1.1 Bottom head Drain Line (cm )
Table 15.6-4 MSIV Closure initiated by High Steam Flow ----
2.4.3 Scram initiated by MSIV Closure ----
2.2.7 .
Table 15.615 Scram initiated by Low Water Level 3 ----
2.2.7 8
Table 2 Containment Pressure / Temperature Response SSAR Verifying SSAR Entrv Parameter Value ITAAC 6.2.1.1.3.3.1.1 Minimum MSIV Closing Time (sec) 3.0 2.1.2 i High Pressure Core Flooder System Minimum Rated Flows (m3/hr per subsystem) 182 '
to 2.4.2 727 at Vessel Pressures (MPaD 8.12
-- vessel to the air space of the to compartment containg the water 0.686 2.4.2 source for the pump suction)
Low Pressure Flooder System Vessel Pressure at which Flow l May Commence l (MPaD -- vessel to drywell) 1.55 2.4.1 Min. Rated Flow (m3/hr per pump) 954 '2.4.1 at Vessel Pressure (MPaD -- vessel to drywell) 0.275 2.4.1 Reactor Core isolation Cooling System Min. Rated Flow (m3/hr) 182 2.4.4 at Vessel Pressures (MPaD 8.12
-- vessel to the air space of the to 2.4.4 i compartment containg the water 1.03 source for the pump suction) -
9
Table 2 Containment Pressure / Temperature Response (Cont.)
SSAR Verifying SSAR Entry Parametet Value ITAAC 6.2.1.1.3.3.2 Maximum MSIV Closing Time (sec) 5.0 2.1.2 .
Total Surface of Drywell Connecting 2 11.3 2.14.1 Vents (m )
Vacuum Breakers Quantity 8 2.14.1 Total Flow Area (m2) 1.53 2.14.1 Table 6.2-2 Drywell Leak Rate (%/ Day) 0.5 2.14.1-Wetwell Leak Rate (%/ Day) 0.5 2.14.1 Min. Suppression Pool 3 3580 2.14.1 Water Volume (m )
Vent System Number of Vents 30 2.14.1
.l Nominal Vent Diameter (m) 0.7 2.14.1 i
2 11.6 2.14.1 Total Horizontal Vent Area (m )
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Table 2 Containment Pressure / Temperature Response (Cont.)
SSAR Verifying SSAR Entrv Parameter Malus ITAAC Table 6.2.2-a Containment Spray Number of RHR Subsystems (Pump Plus Heat Exchanger) 2 2.4.1 Wetwell Spray Flow Rate per RHR Subsystem (kg/hr x 105 ) 1.14 2.4.1 Containment Cooling System Number of RHR Subsystems (Pump Plus Heat Exchanger) 3 2.4.1 Pump Capacity (m3/hr per pump) 954 2.4.1 Overall Heat Transfer Coefficient (kJ/S 0C) 370.5 2.4.1 !
Table 6.3-4 LOCA Break Sizes 2 985 2.1.1 Steamline (cm )
2 839 2.1.1 Feedwater Line (cm )
RHR Shutdown Cooling Suction 2 792 2.1.1 Line (cm )
RHR Injection Line (cm2) 205 2.1.1 High Pressure Core Flooder (cm2 ) 92 2.1.1 2 20.3- 2.1.1 Bottom head Drain Line (cm )
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Table 3 Transient Analysis SSAR Entrv SSAR Verifying Parameter Value ITAAC Table 15.0-1 Reactor internal Recirculation Pumps Number of Pumps 10 2.1.3 Pump Trip Inertia (kg-m2)
Trip Mitigation (maximum) 26.5 2.1.3 Accident (minimum) 17.5 2.1.3 Relief Valve (Relief Function)
Capacity (% NBR Steam Flow at 80.5 kg/cm2g) 91.3 2.1.2 Number of Valves 18 2.1.2 Opening Time (sec) 0.15 2.1.2 (Valve Stroke Time Only. Does not include .1 sec Delay to Energize Solenoid)
High Flux Trip Scram ----
2.2.5 APRM Simulated Thermal Power Trip Scram - --
2.2.5 Total Steamline Volume (m3) 113.2 2.1.2 Table 15; ii FMCRD Scram Times 10H, 9nd Inser1 ion (sec) 0.46 2.2.2 40% Rod Insertion (sec) 1.208 2.2.2 60% Rod Insertion (sec) 1.727 2.2.2 100% Rod Insertion (sec) 3.719 2.2.2 15,1.1.2.2 High Simulated Thermal Power Trip Scram ----
2.2.5 Table 15.1-5 High Water Level 8 Initiates Feedwater Pump Trip ----
2.2.3 12
_ - _ _ - _ _ _ _ _ _ - _ - - _ - _ _ _ - _ _ - - - - _ - - - - - - - - - - - - - - - - - - ~
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l Table 3 Transient Analysis (Cont.)
SSAR Verifying SSAR Entrv Parameter Value ITAAC ;
Table 15.1-5 Turbine Stop Valve Position Switches initiate Reactor Scram ----
2.2.7 Trip of 4 RIPS ----
2.2.8 Table 15.1-7 Low Water Level 2 Initiates Trip of 6 RIPS ----
2.2.8 RCIC System ----
2.4.4 Maximum Startup Time (sec)-- 30 2.4.4 (Includes 1.0 sec for instrument delay) (Design Des. Only)
MSIV Closure on Low Turbine Inlet Pressure ----
2.4.3 15.1.3.3.1 Maximum MSIV Closure Time (sec -- assumes 0.5 sec for instrument delay) 5.0 2.1.2 Table 15.1-9 SRNM High Neutron Flux Scram -----
2.2.5 15.2.1.3.1 TCV Full Stroke Servo Closure (sec) 2.5 2.10.7 Table 15.2-1a Low Water Level 3 Initiates Trip of 4 RIPS ----
2.2.8 Table 15.2-2 High Dome Pressure initiates Trip of 4 RIPS ----
2.2.8 Table 15.2-3 T/G Load Rejection initiates Turbine Control Valve Fast Closure ---
2.10.7 Turbine Bypass System Operation on High Pressure ----
2.10.13 Fast Control Valve Closure initiates Scram ----
2.2.8 Trip of 4 RIPS ---
2.2.8 13
Table 3 Transient Analysis (Cont.)
SSAR Verifying SSAR Entrv Parameter yalue ITAAC 15.2.2.3.1 TCV Full Stroke Fast Closure (sec -- from '
normal operating position) 0.08 2.10.7' Table 15.2-6 Turbine Trip initiates Turbine Control Valve Fast Closure ---- 2.10.7 Turbine Bypass System Operation on High Pressure -----
2.10.13 15.2.3.3.1 Turbine Stop Valve Full Stroke Closure (sec) 0.10 2.10.7 Table 15.2-9 MSIV Position Switches initiate Scram ----
2.2.7 15.2.4.3.1 Minimum MSIV Closure Time (sec) 3.0 2.1.2 L Table 15.2-14 Low Condenser Vacuum initiates MSIV Closure ----
2.4.3 15.2.6.1.1.2 RIP M/G Set Number of RIPS 6 2.2.8 Length of Time Hold Original Speed (sec) 1.0 2.2.8 RIP Coastdown Rate (% per sec) 10 2.2.8 Length of Time (sec) 2.0 2.2.8 Time of RIP Trip (sec) 3.0 2.2.8 ,
Table 15.2-17 Low Water Level 3 Initiates Reactor Scram ----
2.2.7 ,
15.2.7.2.2 Meets Single-failure Criterion ----
2.2.7 - :
l 15.2.9 RHR System has 3 Independent Divisions ----
2.4.1- i l
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Table 3 Transient Analysis (Cont.)
SSAR Verifying SSAR Entrv Parameter Value ITAAC 15.3.1.1.1 No More Than 3 RIPS on One Electrical Power Bus ----
2.2.8 15.3.1.2.2.2 Rapid Core Flow Coastdown Initiates Reactor Scram - --
2.2.5 Mode Switch in the Refuel Position 15.4.1.1.2.2 Refueling Platform Cannot Be Moved Over the Core if a Control Rod is Withdrawn and Fuelis on the Hoist ----
2.5.5 -
15.4.1.1.2.3 Only One or Two Control Rods Associated with the Same HCU Can Be Withdrawn ----
2.2.1 15.4.1.2.1 On Short Flux Period SRNMs Generate Reactor Scram ----
2.2.5 15.4.1.2.3.2 FMCRD Withdrawal Speed (mm/sec) 30 2.2.2 15.4.2.1 At Power the ATLM of the RCIS Prevents Rod Withdrawal Based on MCPR and APLHGR Limits ----
2.2.7 15.4.4.1.1 Overcurrent Protection Logic on the Electrical Bus Which Supplies the Power to the RIPS ----
2.12.1 15.4.9.1 FMCRD Designed to Prevent Rod Ejection ----
2.2.2 15.4.10.1 FMCRD Designed to Prevent Separation ,
of Control Blade and Drive ----
2.2.2 I
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Table 4 Radiological Analysis SSAR Verifying SSAR Entrv Parameter Malue ITAAC Table 15.6-5 Maximum MSIV Closure Time (sec)
(Assumes 0.5 sec for instrument delay.) 5.0 2.1.2 Table 15.6-8 Primary Containment Leakage Rate
(% per day) 0.5 2.14.1 MSIV Total Leakage Rate for All Lines (Um at Standard Conditions) 66.1 2.1.2 SGTS Filter Efficiency Assumed for LOCA (%) 97 2.14.4 Drawdown Time (min) 20 2.14.4 Control Room Recirculation Rates Min. Charcoal Efficiency (%) 95 2.15.5a Table 15.7-8 SGTS Filter Efficiency Assumed for Fuel Handling Accident (%) 99 2.14.4 i
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Table 5 Overpressure Protection SSAR Verifying SSAR Entrv Earameter Value ITAAC 5.2.2.1.4 Direct Scram Signal Generated By:
Position Switches on MSIVs ----
2.2.7 -
Turbine Stop Valves ----
2.2.7 Pressure Swiches on TGV Hydraulic Actuation System Dump Valve ----
2.2.7 Table 5.2-2 Scram Signal on High Flux ----
2.2.5 Recirculation Pump Trip on High Vessel Pressure ----
2.2.8 17
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Table 5 Overpressure Protection (Cont.)
SSAR Verifying SSAR Entrv Parameter Ma!ue ITAAC Table 5.2-3 Safety / Relief Valve Spring Set Pressure 2 SRVs (MPaG) 7.92 2.1.2 Capacity per valve (kg/hr)
(103% Spring Set Pressure) 395000 2.1.2 4 SRVs (MPaG) 7.99 2.1.2 -
Capacity per valve (kg/hr)
(103% Spring Set Pressure) 399000 2.1.2 4 SRVs (MPaG) 8.06 2.1.2 Capacity per valve (kg/hr)
(103% Spring Set Pressure) 402000 2.1.2 4 SRVs (MPaG)- 8.13 2.1.2 Capacity per valve (kg/hr)
(103% Spring Set Pressure) 406000 2.1.2 4 SRW (MPaG) 8.20 2.1.2 Capacity per valve (kg/hr)
(103% Spring Set Pressure) 409000 2.1.2-No. of Valves 18 2.1.2 Figure 5.2-1 SRV Safety Function Opening Time (sec)- 0.3 2.1.2 18
Table 6 l Flooding Protection l l
SSAR Verifying SSAR Entrv Parameter Value ITAAC Reactor and Control Building Flood Protection (from External Sources) l Table 3.4-1 All Penetrations Below Grade Watertight ----
2.15.10 2.15.12 '
Pipe Penetrations Below Design Flood 1 Level Will Be Sealed Against Hydrostatic Head Inside Tunnel or Connecting Building ----
2.15.10 2.15.12 Watertight Doors Installed on All Access Ways Below Design Flood Level ----
2.15.10 (ECCS) 2.15.12 (RCW) 3.4.1.1.1 Min. Wall Thicknesses Below Design Flood Level (m) .61 2.15.10 2.15.12 Piping Tunnels Below Grade do not Penetrate Exterior Walls ----
2.15.10 2.15.12 A
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Table 6 Flooding Protection (Cont.)
SSAR Verifying SSAR Entrv Parameter Value ITAAC Reactor Building Flood Protection (from Internal Sources) 3.4.1.1.2 All Piping, Vessels and Heat Exchangers with Flooding Potential are Seismically Analyzed Standby Liquid Control System ---- 2.2.4 Residual Heat Removal System ----
2.4.1 High Pressure Core Flooder System ----
2.4.2 Reactor Core Isolation Cooling System ----
2.4.4 Reactor Building Cooling Water System ---- 2.11.3 HVAC Emergency Cooling Water Sys. ----
2.11.6 Reactor Service Water System ----
2.11.9 Fire Protection System ----
2.15.6 (Partial coverage)
Oil Storage and Transfer System ----
2.16.2 Main Steamlines (Inside Reactor Bldg) ----
2.1.2 2.10.1 Feedwater Lines (Inside Reactor Bldig) ----
2.1.2 (Only to Seismic interface restraint 20
Table 6 Flooding Protection (Cont.)
SSAR Verifying SSAR Entry Parameter Value ITAAC Reactor Building Flood Protection (from Internal Sources) -- Cont.
Water Sensitive Safety-Related Equipment Raised Off the Floor (mm) 200 2.15.10 All Rooms with a Potential for Flooding Are Supplied With Floor Drains ----
2.9.1 MSIVs Automatically Close on High Temperature in Main Steamline Tunnel ----
2.4.3 3.4.1.1.2.1.1 Evaluation of Floor 100 (B3F)
Watertight Doors on Compartments Containing ECCS Equipment ----
2.15.10 Watertight Doors have Open/Close Status Indicator Lights and Alarms in MCR ----
2.15.10 3.4.1.1.2.1.2 Evaluation of Floor 200 (B2F)
RHR Pressure Lines inside Pipe Chases ----
2.15.10 Minimum Floor Spread Area (m2 ) 300 2.15.10 3.4.1.1.2.1.3 Evaluation of Floor 300 (B1F)
(No Additional Requirements) 3.4.1.1.2.1.4 Evaluation of Floor 400 (1F)
RHR, HPCF and RCIC Lines in Pipe Chases ----
2.15.10 Foam Sprinkler System in Diesel Generator Areas ----
2.15.6 l
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Table 6 Flooding Protection (Cont.)
SSAR Verifying SSAR Entry Parameter Value ITAAC Reactor Building Flood Protection (from Internal Sources)-- Cont.
3.4.1.1.2.1.5 Evaluation of Floor 500 (2F)
Divisional DG Equipment Areas are Separated and Mechanically isolated from Each Other ----
2.15.10 FPC Pools Seismic Category 1 ----
2.15.10 Steamline Tunnel Area isolated by Sealed Doors and Firewalls ----
2.15.10 3.4.1.1.2.1.6 Evaluation of Floor 600 (3F)
Foam Sprinkler System in (Fuel Storage) Day Tank Areas ---- 2.15.6 3.4.1.1.2.1.7 Evaluation of Floor 700 (M4F)
(No Additional Requirements) 3.4.1.1.2.1.8 Evaluation of Floor 800 (4F)
Each RCW Surge Tank A,B & C and its Associated Piping is in a Separate Compartment ----
2.15.10 22
l Table 6 Flooding Protection (Cont.)
SSAR Verifying SSAR Entrv Parameter Value ITAAC 3.4.1.1.2.2 Control Building Flood Protection (from Internal Sources)
No Openings into the Control Building from the Steam Tunnel ----
2.15.12 The Steam Tunnel Sealed At the Reactor Building End ----
2.15.10 All Rooms with a Potential for Flooding Are Supplied With Floor Drains ----
2.9.1 High Water Level Sensors in RCW/RSW Heat Exchanger Room Powered by Class 1E Power Supply ----
2.15.12 Automatically Close RSW isolation Valves and Stop Pumps ----
2.15.12 j 2.11.9 Water Tight Doors on RCW/RSW Heat Exchanger Rooms ----
2.15.12 l Redundant Mechanical Functions are Physically Separated ----
2.15.12 1 Water Sensitive Safety-Related Equipment in Raised Off the Floor All Floors Except Basement (mm) 200 2.15.12 3.4.1.1.2.3 Radwaste Building Flood Protection (from Internal Sources) -l Seismic Category I Substructure ----
2.15.13 1
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Table 6 Flooding Protection (Cont.)
SSAR Verifying SSAR Entry Parameter Value ITAAC 3.4.1.1.2.4 Service Building Flood Protection
- (from Internal Sources) :
Watertight Doors on Access Corridors ----
2.15.11 2.15.12 3.4.1.1.2.5 Turbine Building Flood Protection (from Internal Sources)
Normally Closed Alarmed Door in Passage From Service Building ----
2.15.11 (Doors Only) >
High Water Level in Conden.ser Pit Automatically Shuts Down Circulating Water System ----
-2.10.23 i
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Table 7 Fire Protection (Reactor and Control Building)
SSAR Verifying SSAR Entrv Parameter Value ITAAC 9A.2.4 Electrical Cable Fire-stops Have Fire Rating Equal to Rating of Barrier They Penetrate ----
2.15.10 2.15.12 (Divsional barriers only)
Control, Power or Instrument Cables of Systems Having Similar Safety Related or Shutdown Functions are Located in Separate Fire-resistive Enclosures. ----
2.12.1 A Minimum of Two Fire Suppresssion Means is Available to Each Fire Area ----
2.15.6 9A.4.1.1.1 Drywell inerted During Plant Operation ----
2.14.6 Drywell Has Primary Containment Supply / Exhaust System ----
2.15.5 9A.4.1.1.2 Wetwell inerted During Plant Operation ----
2.14.6 Wetwell Has Spray System ----
2.4.1 Appendix 9A Systems Having Similar Safety Related or !
Shutdown Functions are Located in Separate Fire-resistive Enclosures. ----
2.15.10 2.15.12 (Divsional separation only) .
A Means of Fire Detection, Alarming and Suppression is Provided and Accessible. ----
2.15.6 Fire Stops Are Provided for Cable Tray and Piping Penetrations Through Rated Fire Barriers' ---- 2.15.10 -
2.15.12 Alternate Means of Access and Egress are Provided by a Separate Stair Tower, Elevator or Corridor ----
2.15.10 .;
2.15.12 j 25 -l
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1 Table 8 ATWS Analysis 1
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SSAR Verifying SSAR Entry Parameter .V.alyg ITAAC Nominal Initial Operating Conditions Table 15E-2 Minimum Suppression Pool Volume (m3) 3580 2.14.1 Equipment Performance Characteristics 15.8.2 Minimum SLCS Capacity (1/ min) 378 2.2.4 Table 15E-3 Minimum Closure T;me of MSIV (sec) 3.0 2.1.2 Relief Valve Capacity (%NBR Steam Flow at 80.5 kg/cm2g) 91.3 2.1.2 Number of Valves 18 2.1.2 ,
Opening Time (sec). 0.15 2.1.2 (Valve stroke time only. Does not include 0.1 sec delay to energize solenoid.)
Table 15E-3 Reactor Core Isolation Cooling System Min. Rated Flow (kg/hr) 50.4 2.4.4 at Vessel Pressures (MPaD 8.12
-- vessel to the air space of the to 2.4.4 compartment containg the water 1.03 source for the pump suction)
Initiates on Low Water Level ---- 2.4.4 Maximum Allowable Time Delay !
from Initiating Signal to Rated Flow l Available and injection Valve Fully !
Open (sec) 29.0 2.4.4 l
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. . . - . .. -. .- _ _ . = .--
Table 8 ATWS Analysis (Cont.)
SSAR Verifying SSAR Entry Parameter Value ITAAC Equipment Performance Characteristics High Pressure Core Flooder System Number of Subsystems 2 2.4.2 Minimum Rated Flows (kg/sec per subsystem) 50.4 to 201.6 2.4.2 at Vessel Pressures (MPaD 8.12
-- vessel to the air space of the to 2.4.2 compartment containg the water 0.69 source for the pump suction)
Initiates on Low Water Level ----
2.4.2 Injection Terminated on High Water Level ----
2.4.2 Maximum Allowable Time Delay from initiating Signal to Rated Flow Available and injection Valve Fully Open (Does not include diesel start time and Loading sequence --sec) 20.0 2.4.2 Nominal Recirculation Pump System Inertia (kg-m2) 21.5. 2.1.3 i Maximum Electro-Hydraulic Control Rod insertion Time (sec) 135 2.2.2 Total Minimum RHR Pool Cooling Capacity i For 3 Subsystems (MJ/sec 0C) 1.11 2.4.1 MSIV Closure initiated on Low Water Level ----
2.4.3 MSIV Closure initiated on Low Main ;
Steamline Inlet Pressure to Turbine ----
2.4.3 !
27
. . -.. . . - ._. - . - - ~ .
Table 8 ATWS Analysis (Cont.)
SSAR Verifying SSAR Entry Parameter Value LTf AG ATWS Logic and Setpoints 15E.4 ARI and FMCRD Run-in initiated on High Dome Pressure ----
2.2.8 or Low Water Level 2 ----
2.2.8 SLCS Initiated on an ATWS Trip Signal ----
2.2.4 ATWS Trip Signals for SLCS Initiation High Dome Pressure ----
3.4 and SRNM ATWS Permissive ----
3.4
~
Analytical Time Delay (miniutes) 3 -3.4 or Low Water Level 2 ----
3.4 and SRNM ATWS Permissive ----
3.4 Analytical Time Delay (miniutes) 3 3.4 or Manual AR1/FMCRD Run-in Signals ----
2.2.8 and SRNM ATWS Permissive - ----
2.2.8 Analytical Time Delay (miniutes) 3 2.2.8 RPT (RIPS not Connected to M/G Set)
Initiated on High Dome Pressure ----
2.2.8 28
Table 8 ATWS Analysis (Cont.)
SSAR Verifying SSAR Entrv Parameter Value ITAAC ATWS Logic and Setpoints 15E.4 .RPT (RIPS Connected to M/G Set)
Initiated on Low Water Level 2 ----
2.2.8 Recirculation Runback initated on Any Scram Signal ----
2.2.8 or Any ARl/FMCRD Run-in Signal ----
2.2.8 Feedwater Runback Initiated on an ATWS Trip Signal ----
2.2.3 ATWS Trip Signals for Feedwater Runback High Dome Pressure ----
3.4 and SRNM ATWS Permissive ----
3.4 I Analytical Time Delay (minutes) 2 3.4 ADS inhibit -- Automatic initiation of ADS is Inhibited Uniess Low Water Level 1.5 ----
2.1.2 and APRM ATWS Permissive ----
2.1.2 l
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Table 9 Generic Safety issues SSAR Verifying l SSAR Entrv Parameter Value ITAAC 19B.2-2 A-1: Water Hammer I
Steam Supply System Designed to l Accommodate Steam Hammer ----
3.3 l MSL Designed for Dynamic Loadings Due l to Fast Closing of the Turbine Stop Valves ----
3.3 l l
RCIC System l 1
MUWC to Keep System Filled ----
2.4.4 HPCF System MUWC to Keep System Filled ----
2.4.2 j RHR System l Jockey Pump to Keep System Filled ---- 2.4.;
19B.2-3 A-7: MARK I Long-Term Program Vacuum Breakers Swing Check Type Valves ----
2.14.1 Open Passively on Negative Differential Pressure ----
2.14.1 Require No External Power to Actuate ----
2.14.1 Installed Horizontally Through Pedestal Wall ----
2.14.1 30
Table 9 Generic Safety issues (Cont.)
SSAR Verifying SSAR Entrv Parameter Value ITAAC 198.2-4 A-8: MARK 11 Containment Pool Dynamic Loads Long-Term Program (Refer to response to 198.2-3) 19B.2-5 A-9: ATWS Alternate Rod Insertion Feature Diverse and Independent From RPS ----
2.2.8 ,
Electric Insertion of FMCRD Feature Diverse and independent From RPS --- -
2.2.8 Recirculation Pump Trip on ATWS Signal ----
2.2.8 Automatic initiation of SLC on ATWS Signal ----- 3.4 19B.2-8 A-24: Qualification of Class 1E Safety Related Equipment All Class 1E Electrical Equipment is Environmentally, Dynamically and Seismically Qualified ---- Refer to 1.2(3) 19B.2.-9 A-25: Non Safety Loads on Class 1E Power Sources Non-Class 1E Loads not Connected to I Class 1E Loads Except FMCRD Loads ----
2.12.1 l Class 1E Load Breakers in Division i Between Class 1E Power and Non-Class 1E FMCRD Loads ----
2.12.1
-l l
)
I 31 1
i
Table .9 Generic Safety issues (Cont.)
SSAR Verifying SSAR Entrv Parameter V.aLUS ITAAC 19B.2-10 A-31: Residual Heat Removal (RHR)
Shutdown Requirements RHR System Composed of 3 Electrically '
And Mechanically independent Divisions ----
2.4.1 Shutdown Cooling Can Be Manually Initiated from the Control Room ----
2.4.1 ,
RHR System Can Be powered from Both Offsite and Standby Emergency Electrical Power ----
2.4.1 2.12.1 ,
19B.2-11 A-35: Adequacy of Offsite Power Systems Equipment Qualified for Operation with Voltage up to 10% Less than Normal ----
2.12.1 1 19B.2.12 A-36: Control of Heavy Loads Near Spent Fuel Equipment Handling Components Meet Single Failure Criteria ----
2.15.3 Redundant Safety interlocks and Limit Switches Prevent Heavy Loads Over Spent Fuel ---- 2.15.3 198.2.13 A-39: Determination of Safety Relief Valve Pool Dynamic Loads and Temperature Limits -
Each S/RV Discharge Pipe Fitted with an X-Quencher ----
2.14.1 2.1.2 I (Design J Des. Only) i l
32 1
Table 9 Generic Safety issues (Cont.) >
SSAR- Verifying SSAR Entry Parameter Valve ITAAC 198.2-16 A-44: Station Blackout .,
Sources of Electrical Power No. of Standby Turbine Generators 1 2.12.11 No. of Emergency Diesel Generators 3 2.12.13 19B.2-17 A-47: Safety implications of Control Systems Feedwater Controller Trip Feedpumps on High Water Level ----
2.2.3 Fault Tolerant Through Redundant Micro-processors and Self Diagnostics ---- 2.2.3 19B.2-18 A-48: Hydrogen Control Measures and Effects -
of Hydrogen Burns on Safety Equipment Containment inerted During Normal Operation ----
2.14.6 Permanently Installed Hydrogen Recombiners ----
2.14.8 19B.2-20 B-17: Criteria for Safety-Related Operator Actions RHR Heat Exchanger in LPCI Injection Loop ----
2.4.1 19B.2-22 B-55: Improved Reliability of Target Rock Safety / Relief Valves ABWR Uses a Direct Acting S/RV Design ----
2.1.2 (Design Des. Only) 198.2-23 B-56: Diesel Reliability Independent Diesel Generators 3 2.12.13 Combustion Turbine Generator 1 2.12.11 33
Table 9 l Generic Safety issues (Cont.) ]
i SSAR Verifying SSAR Entry Parameter Value ITAAC 198.2-24 B-61: Allowable ECCS Equipment Outage Periods ECCS Capable of Being Tested During Plant Operation RCIC ----
2.4.4 HPCF ----
2.4.2 RHR ----
2.4.1 198.2-25 B-63: Isolation of Low Pressure Systems Connected to the Reactor Coolant Pressure Boundary Boundary Valves Designed, Fabricated and Tested According to ASME B&PV Code, Section 111 Design Des. Identifies ASME Code Class for System Components RHR System ----
2.4.1 HPCF System ----
2.4.2 RCIC System ----
2.4.4 CRD System ----
2.2.2 SLC System ----
2.2.4 CUW System ----
2.6.1 Nuclear Boiler System ---
2.1.2 ;
Reactor Recirculation System ----
2.1.3 34
I Table 9 Generic Safety issues (Cont.) ;
i SSAR Verifying SSAR Entrv Parameter Value ITAAC 198.2-26 B-66: Control Room Infiltration Measurements Normal AC Filtration Units Number of Divisions 2 2.15.5a Mechanically and Electrically Separate ---- 2.15.5a Number of Outdoor Air Intakes 2 2.15.5a Automatic Switch-over to Emergency Units on High Radiation in Air intake ----
2.15.Sa Emergency Filtration Units Number of Units 2 2.15.5a Mechanically and Electrically Separate ---- 2.15.5a Provisions to Detection Smoke ----
2.15.5a Airborne Radioactive Material ----
2.3.1 Provisions to Remove Smoke and Airborne Radioactive Material ----
2.15.5a 19B.2-27 C-1: Assurance of Continuous Long Term Capability of Hermetic Seals on Instrumontation and Electrical Equipment Safety-related Electrical Equipment is Environmentally Qualified in Accordance with NRC Guidance including NUREG-0588 ---- Refer to 1.2(3)_
35 I
Table 9-Generic Safety issues (Cont.)
SSAR Verifying SSAR Entrv Parameter Valve ITAAC 19B.2-28 C-10: Effective Operation of Containment Sprays in a LOCA SGTS Redundant ----
2.14.4 Filters Gaseous Effluent from Primary and Secaondary Containmnent ----
2.14.4 No. of RHR Subsystems Which Provide Containment Spray 2 2.4.1 Sprays Manually Initiated by Operator ----
2.4.1 Sprays Automatically Terminated When LPFL Injection Valve Opens ----
2.4.1 High Drywell Pressure interlock On Drywell '
Spray Operation ----
2.4.1 19B.2-30 15: Radiation Effects on Reactor Vessel Supports Vessel Support Skirt Located Below Core Beltline ----
2.1.1 Wide Water Flow Region Between Shroud and Vessel Wall ----
2.1.1 l
19B.2-32 25: Automatic Air Header Dump on BWR Scram System ,
i Scram initiated by Low Pressure in the l Common Header Supplying the Charging R Water to the Scram Accumulators ----
2.2.7 l j
36
~
l l
l Table 9 Generic Safety issues (Cont.)
SSAR Verifying SSAR Entry Parameter Value ITAAC 19B.2-33 40: Safety Concerns Associated with Pipe
. Breaks in the BWR Scram System Ball-check Valve in the FMCRD Flange Housing at Connection of the Insert Line with the Drive Scram Port ----
2.2.2 -
19B.2-35 51: Proposed Requirements for improving the Reliability of Open Cycle Service Water Systems A Closed Cooling Water System Will Be Utilized which Transfers Heat Loads Via Heat Exchanger to Service Water System ----
2.11.3 The Safety-Related Portions of the RCW and RSW Will Operate as Designed Assuming Loss of All Offsite Power ----
2.11.3 2.11.9 2.12.1 (Based on Redundancy)
Assuming Any Single Failure ----
2.11.3 2.11.9
)
i 37 l l
l
I Table 9 'l Generic Safety issues (Cont.) 'l SSAR Verifying SSAR Entrv Parameter Value ITAAC ,
19B.2-36 057: Effects of Fire Protection Systems Actuation on Safety-Related Equipment A Means of Fire Detection is Provided ----
2.15.6 All Rooms in the Reactor and Control Buildings with a Potential for Flooding Are Supplied With Floor Drains ----
2.9.2 Safety-Related Equipment Raised Off the Floor ----
2.15.10 Safety-Related Divisions Number 3 2.15.10 2.15.12 Mechanically and Electrically independent ----
Covered by Individual Sys Entries ,
198.2-37 67.3.3: Improved Accident Monitoring Plant Post Accident Monitoring Variables Neutron Flux ----
2.7.1 Control Rod Position ----
2.7.1 Boron Concentration ----
2.11.20 (Sampling Only)
Reactor Coolant System Pressure ----
2.7.1 Drywell Pressure ----
2.7.1 Drywell Sump Level ----
2.7.1 Coolant Level in Reactor ----
2.7.1 Suppression Pool Water Level ----
2.7.1 Containment Area Radiation ----
2.7.1 Primary Containment Pressute ----
2.7.1 Primary Containment isolation Valve Position ----
2.7.1 Coolant Gama ----
2.11.20 (Sampling Only)
Coolant Radiation ----
2.3.1 RHR Flow ----
2.7 1 38 l
l Table 9 Generic Safety issues (Cont.)
SSAR Verifying SSAR Entrv Parameter Value ITAAQ 19B.2-37 67.3.3: Improved Accident Monitoring (Cont.)
Plant Post Accident Monitoring Variables HPCF Flow ----
2.7.1 RHR Heat Exchanger Outlet Temp ----
2.4.1 RCIC Flow ----
2.7.1 SLC Pressure ----
2.7.1 SLCS Storage Tank Level ----
2.7.1 SRV Position ----
2.7.1 Feedwater Flow ----
2.2.3 Standby Energy Status ----
2.7.1 Suppression Pool Water Temp ----
2.7.1 Drywell Air Temperature ----
2.7.1 Drywell/ Containment Hydrogen Concentration ----
-2.7.1 Drywell/ Containment Oxygen Concentration ----
2.7.1 Primary Containment Air Temp ----
2.7.1 Secondary Containment Airspace (e;'uent) Radiation Noble Gas ----
2.3.1 Containment Effluent Radioactivity
- Noble Gas ----
2.11.20 (Sampling Only)
Condensate Storage Tank Level ----
2.7.1 Cooling Water Temperature to ESF System Components ----
2.11.3 Cooling Water Flow to ESF System Components ----
2.11.3 Emergency Ventilation Damper Position---- 2.15.5 Service Area Radiation Exposure Rate ----
2.3.2 Purge Flows - Noble Gases and Vent Flow Rate ----
2.3.1 Identified Release points - Particulates and Halogens - - - -
2.3.1 Airborn Radio Halogens and Particulars---- 2.3.1 39
1 i
Table 9 Generic Safety issues (Cont.)
SSAR Verifying SSAR Entry Parameter Value ITAAC 198.2-38 75: Generic Implications of ATWS Events at Salem Nuclear Plant Separate Scram Groups 4 2.2.7 Solid State Load Drivers Per Scram Group 8 3.4 Contactors for Manual Scram Per Scram Group 2 3.4 198.2-40 83: Control Room Habitability Control Room HVAC Filtration System ----
Refer to 19B.2 Control Room Designed to Withstand Effects of Natural Phenomena ----
2.15.12 Fire Alarm System Provided ----
2.15.6 Fire Hoses and Portable Fire Extinguishers Available ----
2.15.6 19B.2-42 87: Failure of HPCI Steam Line Without Isolation Opening and/or Closing of installed MOVs Used for Isolation of CUW and RCIC Will be Conducted Under Peroperational Differentail Pressure, Fluid Flow and Temperature Conditions ----
2.4.4 2.6.1 Flow Restrictor in CUW Main Suction Line ----
2.6.1 Bottom Head Drainline Tees into CUW Suction Line at an Elevation Above TAF ----
2.6.1 i
l 40
Table 9 Generic Safety Issues (Cont.)
SSAR Verifying SSAR Entrv Parameter .Value ITAAC 198.2-44 103: Design For Probable Maximum Precipitation Design Maximum Rainfall Rate (cm/hr) 49.3 5.0 (Site ,
Parameters)
Design Maximum Short Term Rate (cm/5 min) 15.7 5.0 (Site Parameters) 19B.2-45 105: Interfacing System LOCA at BWRs Design Pressure of Some Low Pressure Components Upgraded to 2.82 MPaG .
(Design Description Only)
RHR System ----
2.4.1 HPCF System ----
2.4.2 RCIC System ----
2.4.4 CRD System ----
2.2.2 SLC System ----
2.2.4 CUW System ----
2.6.1 19B.2-48 118: Tendon Anchorage Failure Primary Containment Structure is of a Reinforced Concrete Design ----
2.14.1 198.2-49 120: On-Line Testability of Protection Systems Manual and Automatic Testability of RPS, LDIS and ECCS Initiation Logic During Reactor Operation ----
3.4 41
l Table 9 Generic Safety issues (Cont.)
SSAR- Verifying SSAR Entry Parameter V3hte ITAAC 198.2-50 121: Hydrogen Control for Large, Dry PWR Containment .
(Not Applicable to BWRs arid Pressure Suppression Containment)
Containment inerted During Normal Operation ---- 2,14.6 19B.2-52 128: Electrical Power Reliability Four Separate and Independent Class 1E dc Divisions ----
2.12.2 No Power Supplied to Non-Class 1E Loads ---- 2.12.1 19B.2-53 142: Leakage Through Electrical isolators in Instrument Circuits Fiber Optic Isolation Devices Used for Electrical isolation of Logic Level and Analog Signals ----
3.4 19B.2-54 143: Availability of Chilled Water Systems and Room Cooling Safety-Related HECW System Provides Chilled Water to Main Control Room Air Conditioning, DG zone Coolers and Control Building Essential Electrical Equipment ----
2.11.6 Essential Equipment HVAC System Provides Controlled Temperature Environment for Safety-Related Equipment Under Accident Conditions ----
2.15.5 42
zi l
Table 9 Generic Safety issues (Cont.)
SSAR Verifying SSAR Entrv Parameter Value ITAAC 198.2-57 153: Loss of Essential Service Water in Light-Water Reactors RSW Divsions Total Number 3 2.11.9 Physically and Electrically Separate ----
2.11.9 RCW Heat Exchangers per Divsion 3 2.11.3 198.2.59 A-17: Systems interaction in Nuclear Power Plants Redundant Safety-Related Equipment and Systems Divisionally Separated ----
Covered by Multiple Sys.
Entries Redundant Electrical Power Systems Divisionally Separated ----
Covered by Multiple Sys.
Entries Divisions Designed Against intra Divisional Flooding ----
2.15.10 2.15.12 19B.2.60 A-29: Nuclear power Plant Design for the Reduction of Vulnerability to Industrial Sabotage Redundant Safety-Related Equipment and Systems Divisionally Separated ----
Covered by Multiple Sys.
Entries Redundant Electrical Power Systems Divisionally Separated ----
Covered by Multiple Sys.
Entries Controlled Access to Safety-Related Areas ----
2.16.3-(Design Des. Only) 43
]
Table 9 '
Generic Safety issues (Cont.)
SSAR Verifying SSAR Entrv Parameter Value ITAAC 198.2.61.1 C-8: Main Steam Line Leakage Control System Main Steamlines and All Branch Lines are Designed to Withstand SSE ----
2.1.2 2.10.1 Non-Safety Main Steam and Bypass Lines at the Turbine Designed to Maintain Structural Integrity Following SSE ----
2.10.1 Condenser Anchorage Designed to Survive SSE ----
2.10.21 198.2.62 029: Bolting Degradation or Failure in Nuclear Power Plants RCPB Component Fabricated, Tested and Installed in Accordance with ASME Code, Sections til and XI ----
2.1.1 (Design Des. Only) 19B.2.63 82: Beyond Design Basis Accidents in Spent Fuel Pools Spent Fuel Pool 5
Seismic Category 1 ----
2.15.10 Low Water Level Alarm ----
2.6.2 (Level Indication only)
Over-Flow Weirs to Skimmer ----
2.6.2 Check Valve in Discharge Line ----
2.6.2 i
44 l l
Table 10 TMIissues SSAR Verifying SSAR Entry Parameter Value ITAAC 19A.2.17 1.D.3 Safety System Status Monitoring Automatic Indication of Bypassed and Inoperable Status of Safety Systems ----
3.4 19B.2.65 1.D.5(2) Plant Status and Post-Accident Monitoring Post-Accident Information Available to Refer to the Operator is in Compliance with RG 1.97 ----
II.F.1 19B.2.66 1.D.5(3) On-Line Reactor Surveillance System ABWR Design incorporates a Reactor Vessel Loose Parts Monitoring System ----
2.8.4 1 A.2.5 ll.B.1 Reactor Coolant System Vents Steam-Driven RCIC 1 2.4.4 Power-Operated Relief Valves Number 18 2.1.2 Dual Position Indication Position Sensors ----
2.1.2 SRV Discharge Temperature Elements ----
2.1.2 Remotely Operable from the Control Room ----
2.1.2 1
45 ,
l i
I Table 10 TMI issues (Cont.)
1 1
SSAR Verifying SSAR Entrv Parameter _V_alg.g ITAAC 1 A.2.6 11.B.2 Plant Shielding to Provide Access to Vital Areas and Protect Safety Equipment for Post-Accident Operation Vital Areas as per NUREG-0737 Accessible Post-LOCA Continuous Occupancy ----
3.2 Non-Continuous Occupancy ----
3.2 1 A.2.7 II.B.3 Post-Accident Sampling Able to Obtain Samples Under Accident Conditions ----
2.11.20 19A.2.21 II.B.8 Rulemaking Proceeding on Degraded Core Accidents inerted Primary Containment ----
2.14.6 Permanently-installed Recombiners ----
2.14.8 1 A.2.9 II.D.1 Testing Requirements SRVs Qualified for Steam Discharge ----
2.1.2 Redundant Logic to Respond to High Water Level Conditions ----
3.4 RHR Shutdown Cooling Systems Number 3 2.4.1 !
I Separate Vessel Penetration and Suction Lines ----
2.4.1 l
46 l
Table 10 TMI issues (Cont.)
SSAR Verifying SSAR Entry Parameter Value ITAAC 1 A.2.10 ll.D.3 Relief and Safety Valve Position Indication Dual Position Indication Position Sensors ----
2.1.2 SRV Discharge Temperature Elements ----
2.1.2 1 A.2.13 II.E.4.1 Decated Penetrations Recombiners in Secondary Containment Number 2 2.14.8 Permanently Installed ----
2.14.8 1 A.2.14 II.E.4.2 Isolation Dependability Diverse Containment isolation Signals ----
2.4.3 Non-Essential Systems.
Automatically isolated On Containment Isolation Signal ----
2.4.3 Redundant isolation Valves ----
2.14.1 Resetting isolation Signal Does Not Automatically Reopen Isolation Valves ----
2.4.3 Containment Purge and Vent Valves Close on isolation Signals ----
2.4.3 Fail Closed ----
2.14.6 Close on High Radiation ---
2.4.3 l
47 l
l
Table 10 TMI issues (Cont.)
SSAR Verifying SSAR Entrv Parameter Value ITAAC 19A.2.27 II.E.4.4 Purging Drywell Has Primary Containment Supply / Exhaust System ----
2.15.5 1 A.2.15 ll.F.1 Additional Accident Monitoring Instrumentation Plant Post Ace' dent Monitoring Variables Neutron Flux ----
2.7.1 Control Rod Position ----
2.7.1 Boron Concentration ----
2.11.20 (Sampling Only)
Reactor Coolant System Pressure ----
2.7.1 Drywell Pressure ----
2.7.1 Drywell Sump Level ----
2.7.1 Coolant Levelin Reactor ----
2.7.1 Suppression Pool Water Level ----
2.7.1 Containment Area Radiation ----
2.7.1 Primary Containment Pressure ----
2.7.1 Primary Containment isolation Valve Position ----
2.7.1 Coolant Gama ----
2.11.20 (Sampling Only)
Coolant Radiation ----
2.3.1 RHR Flow ----
2.7.1 HPCF Flow ----
2.7.1 RHR Heat Exchanger Outlet Temp ----
2.4.1 RCIC Flow ----
2.7.1 SLC Pressure ----
2.7.1 SLCS Storage Tank Level - ----
2.7.1 SRV Position ----
2.7.1 Feedwater Flow ----
2.2.3 Standby Energy Status ----
2.7.1 Suppression Pool Water Temp ----
2.7.1 Drywell Air Temperature ----
2.7.1 Drywell/ Containment Hydrogen Concentration ----
2.7.1 Drywell/ Containment Oxygen Concentration ----
2.7.1 Primary Containment Air Temp ----
2.7.1 48
Table 10 TMl issues (Cont.)
SSAR Verifying SSAR Entrv Parameter Valve ITAAQ 1 A.2.15 ll.F.1 Additional Accident Monitoring instrumentation (Cont.)
Plant Post Accident Monitoring Variables Secondary Containment Airspace (effluent) Radiation Noble Gas ----
2.3.1 Containment Effluent Radioactivity
- Noble Gas ----
2.11.20 (Sampling Only)
Condensate Storage Tank Level ----
2.7.1 Cooling Water Temperature to ESF System Components ----
2.11.3 Cooling Water Flow to ESF System-Components ----
2.11.3 Service Area Radiation Exposure Rate ----
2.3.2 Purge Flows - Noble Gases and Vent Flow Rate ----
2.3.1 Identified Release points - Particulates and Halogens ----
2.3.1 Airborn Radio Halogens and Particulars ---- 2.3.1 1 A.2.16 ll.F.2 Identification of and Recovery from Conditions Leading to inadequate Core Cooling Reactor Wide Range Water Level Number of Divisions 4 2.1.2 Number of Sensors per Division 2 2.1.2 Number of Sets of Sensing Lines per Division 1 2.1.2 ,
Trip Logic per Set of Sensors 2/4 3.4 !
Number of Sets of Sensors 2 ~ 3.4 I
49
Table 10 TMI issues (Cont.)
SSAR Verifying SSAR Entrv Parameter Value ITAAC 1 A.2.17 II.F.3 Instrumentation for Monitoring Accident Conditions ----
Refer to 1 A.2.15 '
1 A.2.18 li.K.1(5) Safety-Related Valve Position Description ----
(Covered by Individual Safety System ITAACs) 1 A.2.20 Describe Automatic and Manual Actions for Proper Functioning of Auxiliary Heat Removal Systems when FW System not Operable Reactor Scram on Low Water Level ----
2.2.7 RCIC System initiates on Low Water Level ----
2.4.4 Terminates injection on High Water Level ----
2.4.4 Restarts on Low Water Level ----
2.4.4 RPV Pressure Controlled by Main Turbine Bypass Valves ----
2.10.13 Safety Relief Valves ----
2.1.2 Discharge to Suppression Pool ----
2.1.2 RHR Systems has Manual Pool Cooling Mode ----
2.4.1 HPCF Systems initiates on Low Water Level ----
2.4.2
- ADS Initiates on Low Water Level ----
2.1.2 RHR - LPFL Mode Initiates on Low Water Level ----
2.4.1 50
Table 10 TMI issues (Cont.)
SSAR Verifying SSAR Entrv Parameter y_alug ITAAC 1 A.2.21 II.K1(23) Describe Uses and Types of RV LevelIndication for Automatic and Manual Initiation of Safety Systems Shutdown Water-Level Measurement Range Top of RPV ----
2.1.2 Bottom of Dryer Skirt ----
2.1.2 Narrow Water-Level Measurement Range Above Main Steam Outlet Nozzle ----
2.1.2 Bottom of Dryer Skirt ----
2.1.2 Low Water Level 3 Automatic initiation Reactor Scram ----
2.2.7 RHR Shutdown Cooling isolation ----
2.4.3 Containment isolation ----
2.4.3 W'ide Water-Level Measurement Range Above Main Steam Outlet Nozzle ----
2.1.2 Top of Active Fuel ----
2.1.2 Low Water Level 2 Automatic initiation RCIC ----
2.4.4 CUW lsolation ----
2.4.3 Low Water Level 1.5 Automatic initiation HPCF ----
2.4.2 MSIV Closure ----
2.4.3 Drywell Cooling System isolation ----
2.4.3 Low Water Level 1 Automatic initiation ADS ----
2.1.2 RHR-LPFL ----
2.4.1 Fuel-Zone Water-Level Measurement Range Above Main Steam Outlet Nozzle ----
2.1.2 Above RIP Deck ----
2.1.2 51
1 Table 10 TMI issues (Cont.)
SSAR Verifying SSAR Entrv Parameter Value ITAAC 1 A.2.22 II.K.3(13) Separation of HPCS and RCIC System Initialon Levels RCIC System initiates on Low Water Level ----
2.4.4 Terminates injection on High Water Level ----
2.4.4 Restarts on Low Water Level ----
2.4.4 HPCF System Initiates on Low Water Level ----
2.4.2 Terminates injection on High Water Level ----
2.4.2 Restarts on Low Water Level ----
2.4.2 1 A.2.23 li.K.3(15) Modify Break Detection Logic to Prevent Spurious isolation of HPCI and RCIC Systems RCIC has a Bypass Start System ----
2.4.4 1 A.2.24 ll.K.3(16) Reduction of Challenges and Failures of Safety Relief Valves - Feasibility Study and System Modification Elimination of Pilot Operated Relief Valves ----
2.1.2 (Design Des. Only)
Redundant Solid State Logic ----
3.4 Pressure Relief Mode Operation is Direct Opening Against Spring Force ----
2.1.2 (Design Des. Only) i l
l 52 l
I i
Table 10 i TMI issues (Cont.) .
.\
I SSAR Verifying SSAR Entrv Parameter Ma!ue ITAAC 1 A.2.26 ll.K.3(18) Modification of ADS Logic-Feasibility Study and Modification for increased Diversity of Some Event Sequences High Drywell Pressure Bypass ,
Timer (minutes) 8 2.1.2 initiates on Low Water Level ----
2.1.2 1 A.2.28 II.K.3(22) Automatic Switchover of RCIC System Suction - Verify Procedures and Modify Design RCIC Automtically Swtiches Pump Suction Source From CSP toSuppression Pool ----
2.4.4 Switchover Signals Low CSP Water Level ----
2.4.4 or High Suppression Pool Level ----
2.4.4 1 A.2.29 II.K.3(24) Confirm Adequacy of Space Cooling Study for HPCI and RCIC Systems individual Room Safety Grade Cooling Units RCIC ----
2.15.5c HPCF ----
2.15.5c Separate Essential Electrical Power Supples RCIC ----
2.4.4 HPCF ----
2.4.2 1 A.2.30 ll.K.3(25) Effect of Loss of AC Power on Pump Seals RCW and RSW Pumps Automatically 2.11.3 Loaded to D / Gs Following LOPP ----
2.11.9 2.12.13 53
Table 10 TMl issues (Cont.)
SSAR Verifying SSAR Entry Parameter Value ITAAC 198.2.21 II.K.3(27) Provide Common Reference Level for Vessel Instrumentation For ABWR the Common Reference for the Reactor Vessel Water Level is at the Top of the Active Fuel ----
2.1.2 (Design Des. Only) 1 A.2.31 II.K.3(28) Study and Verify Qualification of Accumulators on ADS Valves Accumulator Sized to Provide One ADS Actuation with Drywell at Design Pressure ----
2.1.2 Seismic Category l Pneumatic Piping within Primary Containment ----
2.11.13 1 A.2.33.3 II.K.3(46) Response to List of Concerns from ACRS Consultant High Pressure injection ECCS RCIC 1 2.4.4 HPCF 2 2.4.2 Automatic Depressurization on Low Vessel Water Level ----
2.1.2 ECCS Injection Directly into Vessel HPCF 2 2.4.2 RHR LPFL 2 2.4.1 ECCS Injection into Feedwater Lines -
RCIC 1 2.4.4 RHR LPFL 1 2.4.1 ECCS Injection Lines Maintained Filled with Water .
RCIC ----
2.4.4 I HPCF ----
2.4.2 RHR-LPFL ----
2.4.1 I
54
l
. Table 10 l TMI issues (Cont.) l SSAR Verifying SSAR Entrv Parameter Value ITAAC 1 A.2.33.3 II.K.3(46) Response to List of Concerns from ACRS Consultant (Cont.)
High Pressure ECCS Designed to Take Suction from Suppression Pool RCIC ----
2.4.4 HPCF ----
2.4.2 High Pressure ECCS have a Designed Test Mode which Takes Suction from and Discharges to the Suppression Pool RCIC ----
2.4.4 HPCF ----
2.4.2 High Pressure ECCS have a Designed Low Flow Bypass Mode which Dicharges to the Suppression Pool RCIC ----
2.4.4 HPCF ---- 2.4.2 RCIC and HPCF Do not Share Any Common Suction Piping with RHR RCIC ----
2.4.4 HPCF ----
2.4.2 ECCS Have Minimum Flow Protection for All Operating Modes RCIC ----
2.4.4 HPCF ----
2.4.2 RHR ----
2.4.1 55
Table 10 TMI issues (Cont.)
SSAR Verifying SSAR Entry Parameter Value ITAAC 1 A.2.33.3 II.K.3(40) Response to List of Concerns from ACRS Consultant (Cont.)
Number of RCW Divisions 3 2.11.3 Individual ECCS Pumps Can be Isolated Without Affecting Other ECCS Pumps RCIC ----
2.4.4 HPCF ----
2.4.2 ,
RHR ----
2.4.1 ABWR has Water Level Measurement Directly on the Vescel ----
2.1.2 Containment Sprays are Manually Initiated ----
2.4.1 Essential Equipment inside the Containment is Qualified for Harsh Environment ----
2.14.1 ADS Automatically Depressurizes the Vessel on Low Water Level ----
2.1.2 ABWR has Manual Vessel Depressurization Capability ----
2.1.2 56
Table 10 TMI issues (Cont.)
SSAR Verifying SSAR Entrv Parameter Valve ITAAC 1 A.2.34 Ill.D.1.1(1) Review Information Submitted by Licensee Pertainiing to Reducing Leakage from Operating Systems inboard and Outboard Isolation Valves on All Lines Which Penetrate Primary Containment ----
(Covered by individual System Entries)
ABWR has a Leak Detection and Isolation System ----
2.4.3 I MSIV Closure on:
High Temperature in Steam Tunnel ----
2.4.3 High Temperature in Turbine Building ----
2.4.3 High Radiation in HVAC Air Exhaust Results In:
Closure of HVAC Air Ducts to Reactor Building - --
2.4.3 Closure of Containment Purge and Vent Lines ----
2.4.3 Activation of Standby Gas Treatment System ----
2.4.3 l
I l
I 57
Table 10 TMI issues (Cont.)
SSAR Verifying SSAR Entrv Parameter Value ITAAC '
1 A.2.36 Ill.D.3.4 Control Peom Habitability HVAC System Redundant Safety Grade Systems with Outdoor Air Intakes ----
2.15.5c Able to Maintain 3.2 mm WG Positive Pressure in Habitable Control Room ----
2.15.5c Radiatio . and Smoke Sensors in intake Lines to ,oolate Outdoor Air intake ----
2.15.5c Habitable Control Room Shielding Min. Thickness of Concrete Between Habitable Control Room Area and Steam Lines (meters) 1.6 2.15.12 Control Room Constructed Below Grade Level ----
2.15.12 l
1 1
1 58