ML20056F790
| ML20056F790 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 08/25/1993 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9308310085 | |
| Download: ML20056F790 (56) | |
Text
---.
}
GE Nuclear Energy cews rwc : ccwn
, a ce 7 4.e e sv: m u sms August 25, 1993 Docket No. STN 52-001 Chet Poslusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation Room 11-H-3 11555 Rockville Pike Rockville, Maryland 20852
Subject:
ABUR SSAR:
Proposed Chapter 19 Addressing the Design Certification Material Report (DCMR)
Dear Chet:
Attached is a proposed addition to the SSAR Chapter 19 addressing the issue of DCMR contents. This proposal reflects CE's understanding of the-disposition of the road map issues discussed during the GE/NRC meetings 7/27 through 7/29/93 in San Jose. We are submitting a draft version for your preliminary review to obtain your concurrence that this material is consistent with NRC's understanding of what was agreed to.
Please note that:
- 1. The material is preliminary and has not been verified against the latest SSaR.
- 2. GE does not plan to include this material in a formal SSAR amendment until CE/NRC agree that the approach is appropriate.
Tom Boyce has been handling this issue for NRC and should be requested to review the attached material.
-Sincerely,
-h90 JMk N. Fox Advanced Reactor Programs
. attachment cc: A. J. James F. M. Paradiso R. Louison N. E. Hackford E
lh O[
hU 9308310085 930825 PDR ADOCK 05200001 A
PDR d "
I JNF-8-25:c/mj d
j
r h
i DRAFT 19.14 Desinn Certification Material I
Introduction 2
i Section 14.3 of this Safety Analysis Report, METHODOLOGY FOR DETERMINING THE CONTENC GF THE CERTIFIED DESIGN MATERIAL, identifies the criteria for sr,iecting technical material that is to be extracted from the SSAR and included in the Design Certification Material Report (DCMR) (GE document f
24AS447).
In particular, the criteria require the certified design i
descriptions in the DCMR to address the top-level design criteria and performance standards which pertain to the safety of the plant.
Section 14.3 includes tables which identify the ABWR design features which have been included in the DCMR in the area of:
i i
a) ABWR core cooling performance following design basis loss-of-coolant accidents, J
i b) Calculations of primary containment pressure and temperature conditions l
+
4 following a design basis loss-of-coolant accident, I
c) ABWR performance during transient events such as load rejections, inadvertent pump trips, etc.
f i
19.14.1 Summary of the Desicn Certification Material J
4 The safety-related evaluations discussed in the SSAR have been reviewed, and Tables 19.14-1 through 19.14-10 provide a summary of the ABWR design features which have been included in the DCMR as part of the system design description and the associated inspections, tests, analyses and acceptance criteria. The 1
analyses that have been reviewed include the flooding analysis in Chapter 3 of the SSAR, the overpressure protection analysis in Chapter 5, the J
containment cooling analysis in Chapter 6, the core cooling analyses in Chapters 6 and 15, the fire protection analysis in Chapter 9, the safety analysis of transients and the anticipated transients without scram in Chapter 15, the radiological analyses in Chapters 15 and 19, and the evaluation of TM1 issues in Chapter 19.
Important insig; hts from the i
probabilistic risk assessment in Chapter 19 of the SSAR are separately being addressed in Section 19.8 and are not included in these tables.
i a
8 24 AJJ:C-1
i i
DRAFT i
i Summary i
Criteria for selecting ABkT design features to be included in the DCMR are defined in Section 14.3, together with examples of these features for selected plant safety evaluations. Tables 19.14-1 through 19.14-10 provide i
the full set of DCMR entries associated with all of the safety-related evaluations in the SSAR.
?
9 I
?
i f
i i
l.
l j
l l
l' l
8-24 #,JJ:c+2 I
/:
I d O/l /lt_ L
[/l[nfb Table 19.X-1 Core Cooling Analysis i
SSAR SSAR Entry Parameter Value 6.3.3.5 Following a LOCA the RHR System is Automatically Directed to the LPFL Mode l
6.3.3.7.4 The Safety Related Systems Will Operate as Designed with the Loss of AllOffsite AC Power Table 6.3-1 Low Pressure Flooder System Vessel Pressure at which Flow May Commence 2
(kg/cm d -- vessel to drywell) 15.8 3
Min. Rated Flow (m /hr per pump) 954 at Vessel Pressure 2
(kg/cm d -- vessel to drywell) 2.8 Initiating Signals Low Water Level or High Drywell Pressure Maximum Allowable Time Delay from initiating Signal to Pumps at Rated Speed (sec) 29.0 Maximum Allowable Time Delay from Low Pressure Permissive Signal to injection Valve Fully Open (sec) 36.0 2
Table 19.X-1 Core Cooling Analysis (Cont.)
SSAR SSAR Entrv Parameter Value Table 6.3-1 Reactor Core Isolation Cooling System i
Vessel Pressure at which Flow May Commence 2
(kg/cm d -- vessel to drywell) 82.75 Min. Rated Flow (m3/ r) 182 h
2 at Vessel Pressures (kg/cm d 82.75
-- vessel to pump suction) to 10.55 initiating Signals Low Water Level or High Drywell Pressure
[
High Pressure Core Flooder System Vessel Pressure at which Flow May Commence 2
(kg/cm d -- vessel to drywell) 82.75 i
Minimum Rated Flows i
(m3 r per subsystem) 182 to 727 i
/h 2
at Vessel Pressures (kg/cm d 82.75
-- vessel to pump suction) to 7.0 initiating Signals i
Low Water Level or High Drywell Pressure 3
Table 19.X-1 Core Cooling Analysis (Cont.)
SSAR SSAR Entry Parameter Value Table 6.3-1 High Pressure Core Flooder System (Cont.)
f Maximum Allowable Time Delay from initiating Signal to Rated Flow Available and injection Valve Fully Open (sec) 36.0 Automatic Depressurization System Total Number of Relief Valves with ADS Function 8
6 Min. Flow Capacity (kg/hr x 10 )
2.903 l
2 at Vessel Pressure (kg/cm 9) 79,j Initiating Signals Low Water Level
)
and i
High Drywell Pressure or 1
High Drywell Pressure Bypass Timer Timed Out Delay Time from All Initiating Signals Completed to the Time Valves are Open (sec) 29.0 4
h Table 19.X-1 Core Cooling Analysis (Cont.)
SSAR SSAR Entrv Parameter Value Table 6.3-3 The RHR Subsystems are Divisionally Separated The HPCF Subsystems are Divisionally Separated l
RCIC Operation Does not Required AC Power A Single Failure Will not Prevent the Operation of More Than One ADS Valve ----
Table 6.3-4 LOCA Break Sizes 2
Steamline (cm )
984.8 2
Feedwater Line (cm )
838.9 RHR Shutdown Cooling Suction 2
Line (cm )
791.5 2
RHR Injection Line (cm )
205.3 2
High Pressure Core Flooder (cm )
92.0 2
Bottom head Drain Line (cm )
20.25 Table 15.6-4 MSIV Closure Initiated by High r
Steam Flow Scram initiated by MSIV Closure Table 15.6-15 Scram Initiated by Low Water Level 3 5
L Table 19.X-2 Containment Pressure / Temperature Response i
SSAR SSAR Entry Parameter Value 6.2.1.1.4.1 Total Surface of Drywell Connecting 2
Vents (m )
11.3 Vacuum Breakers Diameter (mm) 500 Quantity 8
Table 6.2-2 Drywell Leak Rate (%/ Day) 0.5 Wetwell 1
i Leak Rate (%/ Day) 0.5 Min. Suppression Pool 3
Water Volume (m )
3580 Vent System l
Number of Vents 30 Nominal Vent Diameter (m) 0.7 2
Total Horizontal Vent Area (m )
11.55 i
I a
6
l Table 19.X-2 Containment Pressure / Temperature Response (Cont.)
SSAR SSAR Entrv Parameter Value Table 6.2.2-a Containment Spray l
Number of RHR Subsystems I
(Pump Plus Heat Exchanger) 2 Wetwell Spray Flow Rate j
5 per RHR Subsystem (kg/hr x 10 )
1.12 Containment Cooling System Numberof RHR Subsystems (Pump Plus Heat Exchanger) 3 3
Pump Capacity (m /hr per pump) 954 Overall Heat Transfer Coefficient (kcal/sec 0C) 88.5 Table 6.3-4 LOCA Break Sizes 2
Steamline (cm )
984.8 2
Feedwater Line (cm )
838.9 RHR Shutdown Cooling Suction 2
Line (cm )
791.5 2
RHR Injection Line (cm )
205.3 2
High Pressure Core Flooder (cm )
92.0 2
Bottom head Drain Line (cm )
20.25 i
i l
7
l t
Table 19.X-3 Transient Analysis SSAR SSAR Entry Parameter Value Table 15.0-1 Reactor Internal Recirculation Pumps Numberof Pumps 10 Pump Trip Inertia (kg-m2)
Trip Mitigation (maximum) 26.5 Accident (minimum) 17.5 Relief Valve (Relief Function)
Capacity (% NBR Steam Flow at 80.5 kg/cm2 )
91.3 g
Numberof Valves 18 Opening Time (sec) 0.15 High Flux Trip Scram APRM Simulated Thermal Power Trip Scram Total Steamline Volume (m3) 113.2 Table 15.0-6 FMCRD Scram Times 10% Rod Insertion (sec) 0.46 40% Rod insertion (sec) 1.208 60% Rod insertion (sec) 1.727
)
100% Rod insertion (sec) 3.719 j
i 8
)
d i
Table 19.X-3 Transient Analysis (Cont.)
SSAR SS AR Entry Parameter Value Table 15.1-5 High Water Level 8 Initiates Feedwater Pump Trip l
Turbine Stop Valve Position Switches initiate Reactor Scram Trip of 4 RIPS Table 15.1-6 Low Water Level 2 Initiates Trip of 6 RIPS RCIC System i
Maximum Startup Time (sec)--
30 (includes 1.0 sec for instrument delay)
MSIV Closure on Low Turbine inlet Pressure 15.1.3.11 Maximum MSIV Closure Time (sec -- assumes 0.5 sec for instrument delay) 5.0 15.2.1.3.1 TCV Full Stroke Servo Closure (sec) 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 J
i J
Table 19.X-3 i
Transient Analysis (Cont.)
i SSAR SSAR Entry Parameter Value 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 Trip of 4 RIPS 15.2.2.3.1 TCV Full Stroke Fast Closure (sec) 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 (sec) 0.10 Table 15.2-9 MSIV Position Switches initiate i
Scram 15.2.4.3.1 Minimum MSIV Closure Time (sec) 3.0 Table 15.2-14 Low Condenser Vacuum Initiates MSIV Closure l
i 10 i
l
)
Table 19.X-3 Transient Analysis (Cont.)
l SSAR-SSAR Entry Parameter Value 15.2.6.1.1.2 RlP M/G Set Numberof RIPS 6
Length of Time Hold Original Speed (sec) 1.0 RIP Coastdown Rate (% per sec) 10 Length of Time (sec) 2.0 Time of RIP Trip (sec) 3.0 Table 15.2-17 Low Water Level 3 initiates Reactor Scram i
15.2.7.2.2 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 i
Power Bus l
t 15.3.1.2.2.2 Rapid Core Flow Coastdown Initiates Reactor Scram l
15.4.1.1.2.2 Mode Switch in the Refuel Position l
Refueling Platform Cannot Be Moved Over the Core If a Control Rod is Withdrawn and Fuelis on the Hoist Control Rods Cannot Be Moved if the.
Refueling Platform is Over the Core and i
the Fuelis on the Hoist l
c Only One or Two Control Rods Associated with the Same HCU Can Be Withdrawn I
11
Table 19.X-3 Transient Analysis (Cont.)
(
SSAR SSAR Entry Parameter Value 15.4.1.2.1 On Short Flux Period SRNMs Generate Reactor Scram 15.4.1.2.3.2 FMORD Withdrawal Speed (mm/sec) 30 15.4.2.1 At Power the ATLM of the RCIS Prevents Rod Withdrawal Based on MCPR and APLHGR Limits 15.4.4.1.1 Overcurrent Protection Logic on the Electrical Bus Which Supplies the Power to the RIPS 15.4.8.1 FMCRD Designed to Prevent Rod Ejection 15.4.9.1 FMCRD Designed to Prevent Separation of Control Blade and Drive f
12
Table 19.X-4 Radiological Analysis SSAR SSAR Entry Parameter Value Table 15.6-5 Maximum MSIV Closure Time (sec)
(Assumes 0.5 sec forinstrument delay.) 5.0 Primary Containment Leakage Rate
(% per day) 0.5 MSIV Total Leakage Rate for All Lines (SCFH )
140 SGTS Filter Efficiency (%)
99 il l
l l
13
Table 19.X-5 Overpressure Protection SSAR
+
SSAR Entrv Parameter Value 5.2.2.1.4 Direct Scram Signal Generated By:
Position Switches on MSIVs Turbine Stop Valves Pressure Swiches on TCV Hydraulic Actuation System Dump Valve Table 5.2-2 Scram Signal on High Flux Recirculation Pump Trip on High Vessel Pressure l
f 14
Table 19.X-5 Overpressure Protection (Cont.)
1 SSAR SSAR Entrv Parameter Value Table 5.2-3 Safety / Relief Valve Spring Set Pressure 2
2 SRVs (kg/cr3 g) 80.8 Capacity per valve (kg/hr)
(103% Spring Set Pressure) 395000 4 SRVs (kg/cm2 )
81.5 g
Capacity per valve (kg/hr)
(103% Spring Set Pressure) 399000 l
I 4 SRVs (kg/cm2 )
82.2 9
Capacity per valve (kg/hr)
(103% Spring Set Pressure) 402000
~
4 SRVs (kg/cm2 )
82.9 g
Capacity per valve (kg/hr)
(103% Spring Set Pressure) 406000 4 SRVs (kg/cm2 )
83.6 g
Capacity per valve (kg/hr)
(103% Spring Set Pressure) 409000 l
No. of Valves 18
)
i
~
Figure 5.2-1 SRV Safety Function Opening Time (sec) 0.3 15
4 Table 19.X-6 i
Flooding Protection j
SSAR SSAR Entry Parameter Value Reactor and Control Building Flood Protection (from External Sources) 3.4.1.1 Pipe Penetrations Below Design Flood Level Will Be Sealed Against Hydrostatic Head inside Tunnel or Connecting Building l
3.4.1.1.1 Min. Wall Thicknesses Below Design Flood Level (cm) 60 Watertight Doors and Equipment Hatches Installed Below Design Flood Level Reactor Building Flood Protection (from Internal Sources) 3.4.1.1.2 All Piping, Vessels and Head Exchangers with irlooding Potential are Seismically Analyzed Standby Liquid Control System Residual Heat Removal System High Pressure Core Flooder System Reactor Core Isolation Cooling System ---
Reactor Building Cooling Water System ---
HVAC Emergency Cooling Water Sys.
Reactor Service Water System Oil Storage and Transfer System Main Steamlines (Inside Reactor Bldg)
Feedwater Lines (Inside Reactor Bidig) ----
16
7 Table 19.X-6
')
Flooding Protection (Cont.)
1 SSAR i
SSAR Entrv Earameter Value Reactor Building Flood Protection (from Internal Sources) -- Cont.
MSIVs Automatically Close on High Radiation or Temperature in Main SteamlineTunnel A!! Rooms with a Potential for Flooding Are Supplied With Floor Drains Safety-Related Equipment Raised Off the Floor (cm) 20 3.4.1.1.2.1.1 Evaluation of Floor 100 (B3F)
Watertight Doors on Compartments Containing ECCS Equipment 3.4.1.1.2.1.2 Evaluation of Floor 200 (B2F)
RHR Pressure Lines inside Pipe Chases t
2 Minimum Floor Spread Area (m )
300 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 Foam Sprinkler System in Diesel Generator Areas 17
t Table 19.X-6 Flooding Protection (Cont.)
SSAR SSAR Entry Parameter Value 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 Steamline Tunnel Area isolated by Sealed Doors and Firewalls 3.4.1.1.2.1.6 Evaluation of Floor 600 (3F)
Foam Sprinkler System in Fuel Storage Tank Areas Low Water Level Alarms on Standby Liquid Control Tanks 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,8 & C and its Associated Piping is in a Separate Compartment
\\
\\
l 18
i L
Table 19.X-6 Flooding Protection (Cont.)
.SSAR SSAR Entrv Parameter Value 3.4.1.1.2.2 Control Building Flood Protection i
(from Internal Sources)
No Openings into the Control Building I
from the Steam Tunnel The Steam Tunnel Sealed At the Reactor Building End All Rooms with a Potential for Flooding Are Supplied With Floor Drains i
l High Water Levelin RCW/RSW Heat Exchanger Room Will Automatically Close RSW lsolation Valves and Stop Pumps Water Tight Doors on RCW/RSW Heat Exchanger Rooms i
Redundant Mechanical Functions are Physically Separated Safety-Related Equipment Raised Off the Floor (cm) 20 Turbine Building Flood Protection (from internal Sources)
Normally Closed Alarmed Door in Passage From Service Building High Water Levelin Condenser Pit l
Automatically Shuts Down Circulating Water System l
19
Table 19.X-7 Fire Protection (Reactor and Control Building)
SSAR SSAR Entry Parameter Value 9A.2.4 Electrical Cable Fire-stops Have Fire Rating Equalto Rating of BarrierThey Penetrate Control, Power or lastrument Cables of Systems Having Similar Safety Related or Shutdown Functions are Located in Separate Fire-resistive Enclosures.
A Minimum of Two Fire Suppresssion Means is Available to Each Fire Area 9A.3.2 No Openings in the Steam Tunnel Walls Within the Control Building 9A.4.1.1.1 Drywell Inerted During Plant Operation Drywell Has Purge and Vent System 9 A.4.1.1.2 Wetwell Inerted During Plant Operation Wetwell Has Spray System Appendix 9A Systems Having Similar Safety Related or Shutdown Functions are Located in Separate Fire-resistive Enclosures.
l Appendix 9A A Means of Fire Detection, Alarming and Suppression is Provided and Accessible.
Fire Stops Are Provided for Cable Tray and Piping Penetrations Through Rated Fire Barriers ---
Non-Safety Related Equipment is Located in i
Rooms Separate from Rooms Which Contain Safety Related Equipment Alternate Means of Access and Egress are Provided by a Separate Stair Tower, Elevator or Corridor 20 i
Table 19.X-8 ATWS Analysis SSAR SSAR Entry Parameter Value Nominal Initial Operating Conditions Table 15E-2 Minimum Suppression Pool Volume (m3) 3580 Equipment Performance Characteristics 3
15.8.2 Minimum SLCS Capacity (m /hr) 22.7 Table 15E-3 Minimum Closure Time of MSIV (sec) 3.0 Relief Valve Capacity (%NBR Steam Flow at 80.5 kg/cm2 )
91.3 g
Number of Valves 18 Opening Time (sec) 0.15 Reactor Core Isolation Cooling System Min. Rated Flow (kg/hr) 50.4 2
at Vessel Pressures (kg/cm d 82.75 1
-- vessel to pump suction) to 10.55 initiates on Low Water Level Maximum Allowable Time Delay from initiating Signal to Rated Flow Available and injection Valve Fully Open (sec) 29.0 1
l 21
Table 19.X-8 l
ATWS Analysis (Cont.)
I SSAR j
SSAR Entrv Parameter Va!ue i
Equipment Performance Characteristics i
Tabte 15E-3 High Pressure Core Flooder System Number of Subsystems 2
Minimum Rated Flows (kg/sec per subsystem) 50.4 to 201.6 2
at Vessel Pressures (kg/cm d 82.75
-- vessel to pump suction) to 7.0 4
Initiates on Low Water Level Injection Terminated on High Water Level 4
Max! mum 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 l
Nominal Recirculation Pump System inertia (kg-m2) 21.5 Maximum Electro-Hydraulic Control Rod Insertion Time (sec) 135 l
Total Minimum RHR Pool Cooling Capacity For 3 Subsystems (Kcal/sec 0C) 265 i
MSIV Closure Initiated on Low Water Level l
MSIV Closure Initiated on Low Steamline Pressure i
e l
22
-}
l i
Table 19.X-8 ATWS Analysis (Cont.)
SSAR SSAR Entrv Parameter Value ATWS Logic and Setpoints 15E.4 ARI and FMCRD Run-in initiated on High Dome Pressure l
or Low Water Level 2
8 SLCS Initiated on an ATWS Trip Signal
^
ATWS Trip Signals for SLCS Initiation i
High Dome Pressure and APRM Not Downscale Analytical Time Delay (minutes) 3 or Low Water Level 2 and APRM Not Downscale Analytical Time Delay (minutes) 3 or Manual ARl/FMCRD Run-in Signals and APRM Not Downscale Analytical Tinie Delay (minutes) 3 RPT (RIPS not Connected to M/G Set) i Initiated on f
High Dome Pressure i
23 3
Table 19.X-8 ATWS Analysis (Cont.)
SSAR SSAR Entrv Parameter Value ATWS Logic and Setpoints 15E.4 RPT (RIPS Connected to M/G Set)
Initiated on Low Water Level 2 Recirculation Runback initated on Any Scram Signal or Any ARl/FMCRD Run-in Signal Feedwater Runback initiated on an ATWS Trip Signal ATWS Trip Signals for Feedwater Runback High Dome Pressure and APRM Not Downscale AnalyticalTime Delay (minutes) 2 l
i 24
i Table 19.X-8 ATWS Analysis (Cont.)
SSAR f
SSAR Entry Parameter Value ATWS Logic and Setpoints e
ADS Inhibit initiated on an ATWS Trip Signal ----
ATWS Trip Signals for ADS Inhibit High Dome Pressure and APRM Not Downscale Analytical Time Delay (minutes) 2 or Low Water Level 2 and APRM Not Downscale Analytical Time Delay (seconds) 25 i
25 P
i t
Table 19.X-9 Generic Safety issues SSAR SSAR Entry Parameter Value 198.2-1 Quality and Reliability Assurance Ouality System Requirements have b9en Identified for Each System 19B.2-2 A-1: Water Hammer Steam Supply System Designed to Accommodate Steam Hammer MSL Designed for Dynamic Loadings Due to Fast Closing of the Turbine Stop Valves RCIC System MUWC to Keep System Filled HPCF System MUWC to Keep System Filled RHR System Jockey Pump to Keep System Filled 19B.2-3 A-7: MARK I Long-Term Program i
Vacuum Breakers Swing Check Type Valves Open Passively on Negative Differential Pressure Require No External Powerto Actuate Installed Horizontally Through Pedestal Wall 26
Table 19.X-9 Generic Safety issues (Cont.)
SSAR SSAR Entry Parameter Value 19B.2-4 A-8: MARK 11 Containment Pool Dynamic Loads Long-Term Prograra (Refer to response to 19B.2-3) 19B.2-5 A-9: ATWS Alternate Rod Insertion Feature Diverse and independent From RPS Electric insertion of FMCRD Feature Diverse and Independent From RPS r
Recirculation Pump Trip on ATWS Signal Automatic initiation of SLC on ATWS Signal ----
19B.2-8 A-24: Qualification of Class 1E Safety Related Equipment All Class 1E Electrical Equipment is Environmentally, Dynamically and Seismically Oualified 19B.2.-9 A-25: Non-Safety Loads on Class 1E Power Sources Non-Class 1E Loads not Connected to l
Class 1E Loads Except FMCRD Loads Class 1E Load Breakers in Division i Between Class 1E Power and Non-Class 1E FMCRD Loads 1
27 i
I
Table 19.X-9 Generic Safety Issues (Cont.)
SSAR I
SSAR Entrv Parameter Value 198.2-10 A-31: Residual Heat Removal (RHR)
I Shutdown Requirements RHR System Composed of 3 Electrically And Mechanically Independent Divisions I
Shutdown Cooling Can Be Manually initiated from the Control Room RHR System Can Be powered from Both Offsite and Standby Emergency Electrical Power i
19B.2-11 A-35: Adequacy of Offsite Power Systems Equipment Qualified for Operation with Voltage up to 10% Less than Normal 19B.2.12 A-36: Control of Heavy Loads Near Spent Fuel Equipment Handling Components Meet t
Single Failure Criteria Redundant Safety Interlocks and Limit Switches Prevent Heavy Loads Over Spent Fuel i
19 B.2.19 A-39: Determination of Safety Relief Valve Pool Dynamic Loads and Temperature Limits
]
Each S/RV Discharge Pipe Fitted with an X-Ouencher 28 i
I
1 Table 19.X-9 Generic Safety Issues (Cont.)
F SSAR SSAR Entry Parameter Value 19B.2-16 A-44: Station Blackout Sources of Electrical Power No. of Standby Turbine Generators 1
No. of Emergency Diesel Generators 3
19B.2-17 A-47: Safety implications of Control Systems Feedwater Controller Trip Feedpumps on High Water Level Fault Tolerant Through Redundant Micro-processors and Self Diagnostics ----
19B.2-18 A-48: Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment Containment inerted During Normal Operation l
Permanently Installed Hydrogen Recombiners 19B.2-20 B-17: Criteria for Safety-Related Operator Actions RHR Heat Exchanger in LPCI Injection Loop 19B.2-22 B-55: Improved Reliability of Target Rock Safety / Relief Valves ABWR Uses a Direct Acting S/RV Design 19B.2-23 B-56: Diesel Reliability Independent Diesel Generators 3
Combustion Turbine Generator 1
29
I Table 19.X-9 Generic Safety issues (Cont.)
SSAR SSAR Entrv Earameter Value 19B.2-24 B-61: Allowable ECCS Equipment Outage l
Periods i
ECCS Capable of Being Tested During Plant Operation RCIC HPCF RHR 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 lli RHR System HPCF System e
RCIC System CRD System SLC System CUW System i
Nuclear Boiler System Reactor Recirculation System 4
19B.2-26 B-66: Control Room infiltration Measuremants Normal AC Filtration Units Number of Divisions 2
Mechanically and Electrically Separate ----
l Number of Outdoor Air intakes 2
i 30
Table 19.X-9 Generic Safety issues (Cont.)
SSAR SSAR Entry Parameter Value 19B.2-26 B-66: Control Room infiltration Measurements (Cont.)
Automatic Switch-over to Emergency Units on High Radiation in Air intake Emergency Filtration Units Number of Units 2
Mechanically and Electrically Separate ----
Provisions to Detection Smoke Airborne Radioactive Material Provisions to Remove Smoke and Airborne Radioactive Material 198.2-27 C-1 : Assurance of Continuous Long Term Capability of Hermetic Seals on Instrumentation and Electrical Equipment Safety-related Electrical Equipment is Environmentally Oualified in Accordance with NRC Guidance including NUREG-0588 ---
31
Table 19.X-9 Generic Safety issues (Cont.)
SSAR SSAR Entry Parameter Value 19B.2-28 C-10: Effective Operation of Containment Sprays in a LOCA SGTS Redundant Filters Gaseous Effluent from Primary and Secaondary Containmnent No. of RHR Subsystems Which Provide Containment Spray 2
Sprays Manually initiated by Operator Sprays Automatically Terminated When LPFL Injection Valve Opens High Drywell Pressure Interlock On Drywell Spray Operation 19B.2-30 15: Radiation Effects on Reactor Vessel Supports Vessel Support Skirt Located Below Core Beltline Wide Water Flow Region Between Shroud and Vessel Wall 19B.2-31 23: Reactor Coolant Pump Seal Failures (Not Applicable to ABWR) 32
Table 19.X-9 Generic Safety issues (Cont.)
SSAR SSAR Entry Parameter Value i
198.2-32 25: Automatic Air Header Dump on BWR Scram System Scram initiated by Low Pressure in the Common Header Supplying the Charging Waterto the Scram Accumulators l
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 t
with the Drive Scram Port i
19B.2-35 51: Proposed Requirements for improving the Reliability of Open Cycle Service Water Systems l
t A Closed Cooling Water System Will Be Utilized which Transfers Heat Loads Via Heat Exchanger to Service Water System The Safety-Related Portions of the RCW and RSW Will Operate as Designed Assuming Loss of All Offsite Power Assuming Any Single Failure I
e 33 i
i
Table 19.X-9 Generic Safety issues (Cont.)
SSAR SSAR Entry Parameter Value 19B.2-36 057: Effects of Fire Protection Systems Actuation on Safety-Related Equlpment A Means of Fire Detection is Provided All Rooms in the Reactor and Control Buildings with a Potential for Flooding Are Supplied With Floor Drains Safety-Related Equipment Raised Off the Floor Safety-Related Divisions i
Number 3
I Mechanically and Electrically independent 19B.2-37 67.3.3: Improved Accident Monitoring Plant Post Accident Monitoring Variables Neutron Flux Control Rod Position Boron Concentration Reactor Coolant System Pressure Drywell Pressure Drywell Sump Level Coolant Levelin Reactor t
Suppression Pool Water Level Containment Area Radiation Primary Containment Pressure Primary Containment Isolation Valve Position Coolant Gama Coolant Radiation RHR Flow d
b 1
34
Table 19.X-9 Generic Safety issues (Cont.)
l SSAR SSAR Entrv Parameter Value 19B.2-37 67.3.3: Improved Accident Monitoring (Cont.)
Plant Post Accident Monitoring Variables HPCF Flow RHR Heat Exchanger Outlet Temp RCIC Flow SLC Pressure SLCS Storage Tank Level SRV Position Feedwater Flow Standby Energy Status Suppression Pool Water Temp Drywell Air Temperature Drywell/ Containment Hydrogen Concentration Drywell/ Containment Oxygen Concentration Primary Containment Air Temp Secondary Containment Airspace (effluent) Radiation Noble Gas Containment Effluent Radioactivity
- Noble Gas Condensate Storage Tank Level Cooling Water Temperature to ESF System Components Cooling Water Flow to ESF System Components Emergency Ventilation Damper Position----
Service Area Radiation Exposure Rate ---
Purge Flows - Noble Gases and Vent Flow Rate Identified Release points - Particulates and Halogens Airborn Radio Halogens and Particulars----
35
Table 19.X-9 Generic Safety Issues (Cont.)
SSAR SSAR Entrv Parameter Value j
19B.2-38 75: Generic Implications of ATWS Events at Salem Nuclear Plant A
Separate Scram Groups 4
Solid State Load Drivers Per Scram Group 8
Contactors for Manual Scram Per Scram Group 2
19B.2-40 83: Control Room Habitability i
Control Room HVAC Filtration System Control Room Designed to Withstand Effects of Natural Phenomena Fire Alarm System Provided Fire Hoses and Portable Fire Extinguishers Available 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 l
Flow Restrictor in CUW Main Suction Line Bottom Head Drainline Tees into CUW Suction Line at an Elevation Above TAF 4
36
)
\\
Table 19.X-9 Generic Safety issues (Cont.)
SSAR SSAR Entry Parameter Value 19B.2-44 103: Design For Probable Maximum Precipitation Design Maximum Rainfall Rate (cm/hr) 49.3 Design Maximum Shon Term Rate (cm/5 min) 15.7 19B.2-45 105: Interfacing System LOCA at BWRs Design Pressure of Some Low Pressure Components Upgraded to 28.8 atg RHR System HPCF System RCIC System CRD System 19B.2-47 (Not Used) 198.2-48 118: Tendon Anchorage Failure Primary Containment Structure is of a Reinforced Concrete Design 198.2-49 120: On-Line Testability of Protection Systems Manual and Automatic Testability of RPS, LDIS and ECCS Initiation Logic During Reactor Operation 37
r Table 19.X-9 Generic Safety issues (Cont.)
SSAR r
SSAR Entry Parameter Vafue i
19B.2-50 121: Hydrogen Control for Large, Dry PWR Containment (Not Applicable to BWRs and Pressure Suppression Containment)
Containment inerted During Normal Operation 19B.2-51 124: Auxiliary Feedwater System Reliability (Not Applicable to BWRs) 19B.2-52 128: Electrical Power Reliability Four Class 1E dc Divisions Two-out-of-Four Logic to Activate Safety Systems Two-out-of Three Logic to Activate Safety Systems if One Division is Out of Service Non-Class 1E Loads (Except FMCRD Motors) are Powered from Non-Class 1E Sources dc Divisions Backed by Class 1E D/Gs Non-Class 1E PIP Loads Backed by Off-site Combustion Turbine Generator 19B.2-53 142: Leakage Through Electrical isolators in Instrument Circuits t
Fiber Optic Isolation Devices Used for Electrical isolation of Logic Level and Analog Signals i
38
l Table 19.X-9 Generic Safety issues (Cont.)
SSAR SSAR Entry Parameter Value 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 Essential Equipment HVAC System Provides Controlled Temperature Environment for Safety-Related Equipment Under Accident Conditions 198.2-57 153: Loss of Essential Service Water in Light-Water Reactors RSW Divsions Total Number 3
Physically and Electrically Separate RCW Heat Exchangers per Divsion 3
19B.2.59 A-17: Systems Interaction in Nuclear Power Plants Redundant Safety-Related Equipment and Systems Divisionally Separated Redundant Electrical Power Systems Divisionally Separated Divisions Designed Against 1
intra-Divisional Flooding 39
)
r Table 19.X-9 Generic Safety issues (Cont.)
l SSAR SSAR Entrv Parameter Value 19B.2.60 A-29: Nuclear pow';r Plant Design for the Reduction of Vuln<;rability to Industrial Sabotage Redundant Safety-Related Equipment and Systems Divisionally Separated Redundant Electrical Power Systems Divisionally Separated l
Controlled Access to Safety-Related Areas ----
19B.2.62 029: Bolting Degradation or Failure in Nuclear Power Plants RCPB Component Fabricated, Tested and Installed in Accordance with ASME Code, r
Sections til and XI 19B.2.63 82: Beyond Design Basis Accidents in Spent Fuel Pools i
Spent Fuel Pool Seismic Category 1 Low Water Level Alarm Over-Flow Weirs to Skimmer Check Valve in Discharge Line I
i 40
Table 19.X-10 TMIissues SSAR SSAR Entry Parameter Value 19A.2.17 1.D.3 Safety System Status Monitoring Automatic Indication of Bypassed and inoperable Status of Safety Systems 19B.2.65
!.D.5(2) Plant Status and Post-Accident Monitoring Post-Accident Information Available to the Operator is in Compliance with RG 1.97 --
19B.2.66 1.D.5(3) On-Line Reactor Surveillance System ABWR Design incorporates a Reactor Vessel Loose Parts Monitoring System 1 A.2.5 ll.B.1 Reactor Coolant System Vents Steam-Driven RCIC 1
i Power-Operated Relief Valves Number 18 Dual Position Indication Position Sensors SRV Discharge Temperature Elements Remotely Operable from the Control Room 41
Table 19.X-10 TMI issues (Cont.)
SSAR SSAR Entry Parameter Value 1 A.2.6 ll.B.2 Plant Shielding to Provide Access to Vital Areas and Protect Safety Equipment for Post-Accident Operation Vital Areas as per NUREG-0737 Accessible Post-LOCA Continuous Occupancy Non-Continuous Occupancy 1 A.2.7 II.B.3 Post-Accident Sampling Able to Obtain Samples Under Accident Conditions 19A.2.21 II.B.8 Rulemaking Proceeding on Degraded Core Accidents inerted Primary Containment Permanently-Installed Recombiners 1 A.2.9 II.D.1 Testing Requirements SRVs Qualifieo for Steam Discharge Redundant Logic to Respond to High Water Level Conditions RHR Shutdown Cooling Systems Number 3
Separate Vessel Penetration and Suction Lines 42
Table 19.X-10 TMI issues (Cont.)
SSAR SSAR Entry Parameter Value 1 A.2.10 ll.D.3 Relief and Safety Valve Position Indication Dual Position Indication Position Sensors SRV Discharge Temperature Elements 1 A.2.13 II.E.4.1 Decated Penetrations Recombiners in Secondary Containment Number 2
Permanently Installed 1 A.2.14 II.E.4.2 Isolation Dependability Diverse Containment Isolation Signals Non-Essential Systems Automatically isolated On Containment Isolation Signal Redundant isolation Valves Resetting Isolation Signal Does Not Automatically Reopen isolation Valves Containment Purge and Vent Valves Close on isolation Signals Fail Closed Close on High Radiation 43
Table 19.X-10 TMI issues (Cont.)
SSAR SSAR Entty Parameter Value 1 A.2.15 ll.F.1 Additional Accident Monitoring instrumentation Plant Post Accident Monitoring Variables Neutron Flux Control Rod Position Boron Concentration ReactorCoolant System Pressure Drywell Pressure Drywell Sump Level Coolant Levelin Reactor Suppression Pool Water Level Containment Area Radiation Primary Containment Pressure Primary Containment Isolation Valve Position Coolant Gama Coolant Radiation RHR Flow HPCF Flow RHR Heat Exchanger Outlet Temp RCIC Flow SLC Pressure SLCS Storage Tank Level SRV Position Feedwater Flow Standby Energy Status 44
Table 19.X-10 TMI issues (Cont.)
SSAR SSAR Entrv Parameter Value i
1 A.2.15 ll.F.1 Additional Accident Monitoring 1
Instrumentation (Cont.)
l Plant Post Accident Monitoring Variables Suppression Pool Water Temp Drywell AirTemperature Drywell/ Containment Hydrogen Concentration i
Drywell/ Containment Oxygen Concentration Primary Containment AirTemp Secondary Containment Airspace (effluent) Radiation Noble Gas Containment Effluent Radioactivity
- Noble Gas L
Condensate Storage Tank Level Cooling Water Temperature to ESF i
System Components Cooling Water Flow to ESF System Components Service Area Radiation Exposure Rate Purge Flows - Noble Gases and Vent l
Flow Rate Identified Release points - Particulates and Halogens Airborn Radio Halogens and Particulars ---
i P
f 1
45 t
Table 19.X-10 TMI issues (Cont.)
i SSAR SSAR Entry Parameter Value 1 A.2.16 ll.F.2 Identification of and Recovery from Conditions Leading to inadequate Core Cooling Reactor Wide Range Water Level Numberof Divisions 4
Number of Sensors per Division 2
Number of Sets of Sensing Lines per Division 1
Trip Logic per Set of Sensors 2/4 Number of Sets of Sensors 2
i 1 A.2.17 II.F.3 Instrumentation for Monitoring Accident Conditions l
46
i J
Table 19.X-10 2
TMI iss'Jes (Cont.)
e SSAR SSAR Entry Parameter Value 1 A.2.20 Describe Automatic and Manual Actions for 1
Proper Functioning of Auxiliary Heat Removal Systems when FW System not Operable Reactor Scram on Low Water Level RCIC System initiates on Low Water Level Terminates injection on High Water Level Restarts on Low Water Level RPV Pressure Controlled by Main Turbine Bypass Valves Safety Relief Valves Discharge to Suppression Pool RHR Systems has Manual Pool Cooling Mode i
HPCF Systems Initiates on Low Water Level l
ADS i
Initiates on Low Water Level RHR - LPFL Mode Initiates on Low Water Level I
~
l 1 A.2.21 II.K1(23) Describe Uses and Types of RV Level Indication for Automatic and Manual l
Initiation of Safety Systems i
Shutdown Water-Level Measurement Range Top of RPV Bottom of Dryer Skirt 47 f
h
Table 19.X-10 TMI issues (Cont.)
SSAR SSAR Entry Parameter Value 1 A.2.21 II.K1(23) Describe Uses and Types of RV LevelIndication for Automatic and Manual Initiation of Safety Systems (Cont.)
Narrow Water-Level Measurement Range Above Main Steam Outlet Nozzle Bottom of Dryer Skirt Low Water Level 3 Automatic initiation Reactor Scram RHR Shutdown Cooling Isolation Containment isolation Wide Water-Level Measurement Range Above Main Steam Outlet Nozzle Top of Active Fuel Low Water Level 2 Automatic initiation RCIC CUW lsolation Low Water Level 1.5 Automatic initiation HPCF MSIV Closure Drywell Cool;ng System isolation Low Water Level 1 Automatic initiation ADS RHR-LPFL Fuel-Zone Water-Level Measurement Range Above Main Steam Outlet Nozzle Above RIP Deck 48
Table 19.X-10 TMIissues (Cont.)
SSAR SSAR Entry Parameter Value 1 A.2.22 II.K.3(13) Separation of HPCS and RCIC System initiaion Levels RCIC System initiates on Low Water Level Terminates injection on High Water Level Restarts on Low Water Level HPCF System initiates on Low Water Level Terminates injection on High Water Level Restarts on Low Water Level l
1 A.2.23 II.K.3(15) Modify Break Detection Logic to Prevent Spurious isolation of HPCI and RCIC Systems RCIC has a Bypass Start System 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 Redundant Solid State Logic Pressure Relief Mode Operation is Direct Opening Against Spring Force 49
Table 19.X-10 TMI issues (Cont.)
SSAR SSAR Entrv Parameter Value 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 initiates on Low Water Level 1 A.2.28 II.K.3(22) Automatic Switchover of RCIC System Suction - Verify Procedures and Mod lfy Design RCIC Automtically Swtiches Pump Suction Source From CSP toSuppression Pool Switchover Signals Low CSP Water Level or High Suppression Pool Level 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 HPCF Separate Essential Electrical Power Suppies RCIC HPCF 50
I I
I Table 19.X-10 f
TMI issues (Cont.)
l SSAR SSAR Entry Parameter Value 1 A.2.30 ll.K.3(25) Effect of Loss of AC Power on i
Pump Seals RCW and RSW Pumps Automatically Loaded to D / Gs Following LOPP 1
19B.2.71 II.K.3(27) Provide Common Reference Level for Vessel Instrumentation For ABWR the Common Reference for the Reactor Vessel Water Levelis at the Top of the Active Fuel 1 A.2.31 II.K.3(28) Study and Verify Oualification of Accumulators on ADS Valves Accumulator Sized to Provide One ADS Actuation with Drywell at Design Pressure Seismic Category l Pneumatic Piping within Primary Containment l
1 A.2.33.3 II.K.3(46) Response to List of Concems from ACRS Consultant High Pressure injection ECCS RCIC 1
HPCF 2
Automatic Depressurization on Low Vessel Water Level ECCS Injection Directly into Vessel HPCF 2
RHR-LPFL 2
ECCS Injection into Feedwater Lines RCIC 1
RHR-LPFL 1
]
51
I Table 19.X-10 TMI issues (Cont.)
SSAR SSAR Entry Parameter Value 1 A.2.33.3 II.K.3(46) Response to List of Concerns from ACRS Consultant (Cont.)
ECCS Injection Unes Maintained Filled with Water RCIC HPCF RHR-LPFL High Pressure ECCS Designed to Take Suction from Suppression Pool i
RCIC HPCF High Pressure ECCS have a Designed Test Mode which Takes Suction from and Discharges to the Suppression Pool RCIC HPCF High Pressure ECCS have a Designed Low Flow Bypass Mode which Dicharges to the Suppressien Pool RCIC HPCF i
RCIC and HPCF Do not Share Any Common Suction Piping with RHR RCIC HPCF LPFL ECCS Have. Minimum Flow Protection for All Operating Modes RCIC HPCF RHR 52
1 Table 19.X-10 TMI issues (Cont.)
SSAR SSAR Entry Parameter Value 1 A.2.33.3 II.K.3(46) Response to List of Concerns from ACRS Consultant (Cont.)
Number of RCW Divisions 3
Individual ECCS Pumps Can be Isolated Without Affecting Other ECCS Pumps RCIC HPCF RHR ABWR has Water Level Measurement Directly on the Vessel Containment Sprays are Manually initiated Essential Equipment inside the Containment is Qualified for Harsh Environment ADS Automatically Depressurizes the Vessel on Low Water Level ABWR has Manual Vessel Depressurization Capability 9
i t
53
Table 19.X-10 1
TMI issues (Cont.)
SSAR SSAR Entry Parameter Value 1 A.2.36 Ill.D.3.4 Control Room Habitability HVAC System Redundant Safety Grade Systems with Outdoor Air intakes Able to Maintain 3.2 mm WG Positive Pressure in Habitable Control Room 1
Radiation and Smoke Sensors in Intake Lines to isolate Outdoor Air intake Habitable Control Room Shielding i
Min. Thickness of Concrete Between Habitable Control Room Area and Steam Lines (meters) 1.6 Control Room Constructed Below Grade Level
~
E i
k e
54