ML20062K708
| ML20062K708 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 12/13/1993 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20062K707 | List: |
| References | |
| NUDOCS 9312270059 | |
| Download: ML20062K708 (183) | |
Text
i 1
i
i Amendment 33 - Supplemental Page Change Instruction ;
REMOVE ADD REMOVE ADD PAGE NO. PAGE NO, PAGE NO. PAGE NO. ,
l CilAPTERI CIIAPTER 7 !
1.9-5, 6 1.9-5, 6 7.5-11, 12 7.5-11, 12 f_llAPTER 2 Cil APTER 9 )
2.0-1, 2 2.0-1, 2 9.2-75, 76 9.2-75,76 ,
2.0-3/4 2.0-3/4 9B-15/16 98-15/16 ,
CIIAPTER 3 CIIAPTER 14 [
3.2-17, 18 3.2-17,18 thru 3.2-61,62 thru 3.2-61,62 14.2-17, 18 14.2-17,18 l 3A-xvii /xviii 3A-xvil/xvill CIIAPTER 15 e
3B-37,38 3B-37, 38 15.7-13, 14 15.7-13, 14 ;
3B-45,46 38-45,46 '
3B-49,50 3B-49, 50 CIIAPTER 19 thru 3B-69,70 thru 3H-69,70 :
19.8-9, 10 19.8-9, 10 -!
CIIAPTER 4 '
19.9-3,4 19.9-3,4 43-3,4 43-3,4 ,
19D.6-73, 74 19D.6-73, 74 ;
CIIAPTER 6 CIIAPTER 20 6.0-i,ii 6.0 -1, 11 ;
thru 6.0-xi, xii thru 6.0-xi, xil 203.10-21,22 203.10-21,22 !
l 6.1-3, 4 6.1-3, 4 l
6.2-53,54 6.2-53, 54 1
6 3-15, 16 63-15,16 ;
63-29,30 63-29,30 i 63-45,46 63-45,46 1 6.3-49,50 63-49,50 6 3-83, 84 6 3-83, 84 63-85,86 63-85,86 i 6 3-87 63-87 !
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6.5-1, 2 6.5- 1, 2 )
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L'!RBK 9.LD l
9312270059 931213 PDR ADOCK 05200001 A PDR j
23A6100 Rev. 3 ABWR standardsaferyAnarvsis Report O
v Table 1.9-1 Summary of ABWR Standard Plant COL License Information (Continued)
Item No. Subject Subsection 3.30 Audits of Design Specifications and Design Reports 3.9.7.4 l
3.31 ASME Class 1,2, and 3 Piping System Clearance 3.9.7.5 l ,
Requirements 3.32 As-Built Reconciliation Analysis for ASME Class 1,2 and 3 3.9.7.6 Piping Systems 3.33 Pipe Support Baseplate and Anchor Bolt Design 3.9.7.7 l
3.34 Pipe-Mounted Equipment Allowable Loads 3.9.7.8 l
3.35 Be..chmark Requirements for Computer Codes Used to 3.9.7.9 l ,
Perform Piping Dynamic Analysis -
3.36 ASME Class 1,2, and 3 Piping System Design 3.9.7.10 l
Requirements for Thermal Stratification of Fluids 3.37 Equipment Qualification 3.10.5.1 l
3.38 Dynamic Qualification Report 3.10.5.2 l
3.39 Qualification by Experience 3.10.5.3 l
k 3.40 Environmental Qualification Document (EQD) 3.11.6.1 l
3.11.6.2 l 3.41 Environmental Qualification Records 3.42 Surveillance, Maintenance, and Experience information 3.11.6.3 l
3.43 Radiation Environment Conditions 31.3.3.1 ,
l 4.1 Thermal Hydraulic Stability 4.3.5.1 4.2 Power / Flow Operating Map 4.4.7.1 4.3 Thermal Limits 4.4.7.2 t 4.4 CRD inspection Program 4.5.3.1 5.1 Conversion of Indicators 5.2.6.1 I 5.2 Plant Specific ISI/ PSI 5.2.6.2 ,
5.3 Reactor Vessel Water Level Instrumentation 5.2.6.3 5.4 Fracture Toughness Data 5.3.4.1 5.5 Materials and Surveillance Capsule 5.3.4.2 5.6 Plant Specific Pressure. Temperature information 5.3.4.3 ;
5.7 Testing of Mainsteam isolation Valves 5.4.15.1 l
5.8 Analyses of 8-hour RCIC Capacity 5.4.15.2 l
5.9 ACIWA Flow Reduction 5.4.15.3 l
Y 6.1 Protection Coatings and Organic Materials 6.1.3.1 6.2 Alternate Hydrogen Control 6.2.7.1 l
COL License Information - Amendment 33 L9 5
r-23A6 MD Rev. 3 ABWR standard sareryAnarysis aeport O
Table 1.9-1 Summary of ABWR Standard Plant COL License information (Continued) item No. Subject Subsection 6.3 Administrative Control Maintaining Containment 6.2.7.2 !
l Isolation 6.4 Suppression Pool Cleanliness 6.2.7.3 l
6.5 Wetwell to-Drywell Vacuum Breaker Protection 6.2.7.4 l
6.6 ECCS Performance Results 6.3.6.1 l
6.7 ECCS Testing Requirements 6.3.6.2 l
6.8 Toxic Gases 6.4.7.1 l
6.9 SGTS Performance 6.5.5.1 l '
6.10 PSI and ISI Program Plans 6.6.9.1 l
6.11 Access Requirement 6.6.9.2 <
l ~
7.1 Cooling Temperature Profiles for Class 1E Digital 7.3.3.1 Equipment 7.2 APRM Oscillation Monitoring Logic 7.6.3.1 ,
7.3 Effects of Station Blackout on HVAC 7.8.1 ,
7.4 Electrostatic Discharge on Exposed Equipment 7.8.2 Components 7.5 Localized High Heat Spots in Semiconductor Material for 7.8.3 Computing Devices 8.1 Diesel Generator Reliability 8.1.4.1 l
8.2 Periodic Testing of Offsite Equipment 8.2.4.1 8.3 Procedures When a Reserve or Unit Auxiliary 8.2.4.2 Transformer is Out of Service 8.4 Offsite Power Systems Design Bases 8.2.4.3 8.5 Offsite Power Systems Scope Split 8.2.4.4 8.6 Capacity of AuxiliaryTransformers 8.2.4.5 ;
8.7 Deleted 8.3.4.1 l
8.8 Diesel Generator Design Details 8.3.4.2 l
8.9 Certified Proof Tests on Cable Samples 8.3.4.3 l
8.10 Protective Devices for Electrical Penetration Assemblies 8.3.4.4 l
8.11 Deleted 8.3.4.5 l :
8.12 DC Voltage Analysis 8.3.4.6 ,
l 8.13 Deleted 8.3.4.7 l
8.14 Deleted 8.3.4.8 l
u.S COL License Information - Amendment 33 b
23A6100 Rzv.1 ABWR StandardSafety Analysis Report O
2.0 Site Characteristics 2.0.1 Summary This section defines the envelope of site-related parameters which the ABWR Standard Plant is designed to accommodate. These parameters envelope most potential sites in the U.S. A summag of the site envelope design parameters is given in Table 2.0-1.
O O
2,0- 1 Site Characteristics - Amendment 31
23A6100 R5v. 3
~
ABWR Standsnf SafetyAnalysis Report O
Table 2.0-1 Envelope of ABWR Standard Plant Site Design Parameters Maximum Ground Water Level: 61.0 cm below grade Extreme W'md: Basic Wind Speed: 177 km/hr*/197 km/hr t Maximum Flood (or Tsunami) Level:8 30.5 cm below grade Tornado: - Maximum Tornado Wind Speed: 483 km/hr
- Maximum Rotational Speed: 386 km/hr
-Translational Velocity: 97 km/hr
- Radius: 45.7m
- Maximum Pressure Drop: 0.141 kg/cm2d
-Rate of Pressure Drop: 0.0846 kg/cm2/sec
- Missile Spectra: Spectrum Il Precipitation (for Roof Design): - Maximum Rainfall Rate: 49.3 cm/hr**
- Maximum Snow Load: 0.024 kg/cm 2 Ambient Design Temperature: 1% Exceedance Values
- Maximum: 37.8'C dry bulb 25*C wet bulb (coincident) 26.6*C wet bulb (non-coincident)
- Minimum:-23.3 C 0% Exceedance Values (Historical limit)
- Maximum 46.1 C dry bulb 26.7 C wet bulb (coincident) 27.2*C wet bulb (non-coincident)
- Minimum:-40 C Soil Properties: - Minimum Static Bearing Capacity: 7.32 kg/cm 2n
- Minimum Shear Wave Velocity: 305 m/sec'*
- Liquification Potential: None at plant site resulting from site .
specific SSE ground motion O
2 0-2 Site Characteristics - Amendment 33
e .
23A6100 Rev. 3 standard satary Anatysis aeport ABWR Table 2.0-1 Envelope of ABWR Standard Plant Site Design Parameters (Continued) :i Seismology: - SSE Peak Ground Acceleration: 0.30gf /
- SSE Response Spectra: per RG 1.60
- SSE Time History: Envelope SSE Response Spectra Hazards in Site Vicinity: - Site Proximity Missiles and Aircraft 510~7 per year
-Toxic Gases None
-Volcanic Activity None Exclusion Area Boundary:(EAB) - An area whose boundary has a Chi /O less than or equal to 3
1.37 x 10'3 sec/m t 1.37x10'3 sec/m 3 i
Meteorological Dispersion (Chi /Q): - Maximum 2-hour 95% EAB
- Maximum 2-hour 95% LPZ 4.11x104 sec/m 3 .
I
- Maximum annual average (8760 1.17x10-8 sec/m3 hour) LPZ
- 50-year recurrence interval: value to be utilized for design of non-safety-related structures only.
t 100-year recurrence interval: value to be utilized for design for safety-related structures only.
O
- Probable maximum flood level (PMF), as defined in ANSl/ANS-2.8, " Determining Design Basis Flooding at Power Reactor Sites."
f Spectrum 1 missiles consist of a massive high kinetic energy missile which derforms on impact, a rigid missile to test penetration resistance, and a small rigid missile of a size sufficient to just pass through any openings in protective barriers. These missiles consists of an 1800 kg automobile, a 125 kg,20 cm diameter armor piercing artillery shell, and a 2.54 cm diameter solid steel sphere, all impacting at 35% of the maximum horizontal windspeed of the design basis tornado.The first two missiles are assumed to impact at normal incidence, the last to impinge upon openings in the most damaging directions.
" Maximum value for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> over 2.6 km 2probable maximum precipitation (PMP) with ratio of 5 minutes to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> PMP of 0.32 as found in National Weather Source Publication HMR No. 52.
Maximum short term rate: 15.7 cm/5 min.
tt At foundation level of the reactor and control buildings.
- This is the minimum shear wave velocity at low strains after the soil property uncertainties have i been applied.
f f Free-field, at plant grade elevation.
l l
2 0 3>4 Site Characteristics - Amendment 33 l
1 I
23A6100 Rev. 3 ABWR sentantseveryAnalysisneport -
Table 3.2-1 Classification Summary (Continued)
Quality Quality Assur-Group ance Classi- Require- Seismic Principal Component
- Safg Class Location
- ficationd rnent' Category' Notes B1 Reactor Pressure Vessel System
- 1. Reactor pressure 1 C A B I vessel (RPV)
- 2. Reactor vessel support 1 C A B l skirt and stabilizer
- 3. RPV appurtenances- 1 C A B l reactor coolant pressure boundary portions (RCPB) 4.- Lateral supports for 1 C A B I.
CRD housing and in-core housing
- 5. Reactor internal 2 C - B 1 structures, spargers, for feedwater, RHR shutdown cooling low pressure flooder, and high pressure core flooder systems (see Subsection 3.9.5)
- 6. Reactor internal 2 C - B I
. structures-safety-related components
. (except spargers) including core support structures (See Subsection 3.9.5)
- 7. Reactor internal N C - E- -
structures-non-safety-related components (See Subsection 3.9.5)
- 8. Deleted
- 9. Deleted
- 10. Deleted Reactor Internal Pump C A 'B l
- 11. 1 O Motor Casing (a part of RPV boundary)
Notes and footnotes are listed on pages 3.2-54 through 3.2-61 Classification of Structures, Co'nponents, and Systems - Amendment 33 3.2 17
23A6100 Rsv. 3 ABWR sesaknf suretyAnalysis neport O
Table 3.2-1 Classification Summary (Continued)
Quality Quality Assur-Group ance Classi- Require- Seismic Safeg Category' Notes Principal Component
- Class Location
- ficationd m ent*
B2 Nuclear Boiler System
- 1. Vessels-level 1 C A B I instrumentation condensing chambers
- 2. Vessel-nitrogen 3/N C C B i accumulators (for ADS and SRVs)
- 3. Piping including 3 C C B i (h) supports-safety / relief valve discharge and quencher
- 4. Piping including 1 C,SC A B l supports main steamline (MSL) and ,
feedwater (FW) line up to and including the outermost isolation valve
- 5. Piping including supports
- a. MSL (including 2 SC B B 1 (r) branch lines to first valve) from outermost isolation valve up to and including seismic interface restraint
- b. FW (including 2 SC B B l (r) branch lines to fi~rst valve) from outermost isolation valve to and including the shutoff valve Notes and footnotes are listed on pages 3.2-54 through 3.2-61 3.2-18 Cisssification of Structures. Components, and Systems - Amendment 33
I 23A6100 Rev. 3 ABWR standardsateerAnstrsis Report C'd ,
Table 3.2-1 Classification Summary (Continued)
Quality Quality Assur-Group ance Classi- Require. Seismic Principal Component
- Safeg Class Location
- fication d
m ent* Category' Notes
- 6. Piping including N SC,T B F -
(r) supports-MSL (including branch lines to first valve) from the seismic interface restraint up to but not including the turbine stop valve and turbine bypass valve
- 7. Piping from FW N SC D E I (ee) shutoff valve to seismic interface restraint .
- 8. Deleted
- 9. Deleted
- 10. Pipe whip restraint- 3 SC,C -
B -
MSUFW
- 11. Piping including ,
supports-other within outermost isolation valves
- a. RPV head vent 1 C A B i (g)
- b. Main steam drains 1 C,SC A B I (g)
- 12. Piping including supports-other beyond outermost isolation or shutoff valves
- a. RPV head vent N C C E -
beyond shutoff j valves i
- b. Main steam drains 2/N SC,T B B 1/- (r) to first valve
- c. Main steam drains N SC,T D E -
(r) j beyond first valve Notes and footnotes are listed on pages 3.2-54 through 3.2-61 l
Classalitation of Structures. Components, and Systems - Amendment 33 3 2-19
23A6100 R1v. 3 ACWR srasantsarery A=rysis a:pers O
Table 3.2-1 Classification Summary (Continued)
Quality Quality Assur-Group ance Classi- Require- Seismic Safetg Principal Component
- Class Location
- fication d ment' Category' Notes
- 13. Piping including 2/N C,SC B/D B/E 1/- (g) supports-instrumentation up to and beyond outermost isolation vaives
- 14. Safety / relief valves 1 C A B i
- 16. Valves-FW, other 2 SC B B l (ee) beyond outermost isolation valves up to and including shutoff valves
- 17. Valves-within outermost isolation valves
- a. RPV head vent 1 C A B i (g)
- b. Main steam drains 1 C,SC A B i (9)
- 18. Valves, other
- a. RPV head vent 3 C C B I
- b. First main steam 2/N SC B B 1/- (r) drain valves
- c. Other main steam N SC D E -
(r) drain valves
- 19. Deleted
- 20. Mechanical 3 C,SC - B I modules-instrumentation with safety-related function
- 21. Electrical modules 3 C,SC,X -
B i (i) with safety-related function Notes and footnotes are listed on pages 3.2-54 through 3.2-61 3.2 20 Classi6 cation of Structures. Components. and Systems - Amendment 33
23A6100 Rsv. 3 .
ABWR saadantseterAaaiyak naput
)
I t
t $
f Table 3.2-1 Classification Summary (Continued)
Quality l Quality Assur' -l Group ance _
Classi- Require- Seismic I Principal Component
- Safg Class Location
- fication d m ent' Category I Notes 6 t
- 22. Cable with safety- 3 C,SC,X -
B 1 ,
related function [
l B3 Reactor Recirculation System j (s)(g)
- 1. Piping, Valves and all 2 C B B i )
their supports-Purge System, heat exchanger and primary side of j recirculation motor -l cooling system (RMCS) e
- 2. Pump motor cover, 1 C A A 1 ;
bolts and nuts 1
- 3. Pump non-pressure N C - E -
l i
retaining parts
- including motor, 3 instruments, electrical cables, and seals l
- 4. ATWS equipment N C - E -
(cc) i i
associated with the pump trip function l C1 Rod Control and Information System ,
- 1. Electrical Modules N RZ,X D E -
f
- 2. Cable N SC,RZ,X D E -
I l
C2 CRD System D E I
- 1. Valves with no safety- N SC -
related function (not f part of HCU)
Piping including 2 C,SC B B 1 (j) i
- 2. i supports-insert line j Notes and footnotes are listed on pages 3.2-54 through 3.2-61 ,
i Cia ssi6 cation of Structures. Components, and Systems - Amendment 33 3221
)
23A6100 Rsv. 3 ABWR studentsareryAntysissupon i
O Table 3.2-1 Classification Summary (Corrtinued)
Quality Quality Assur-Group ance Classi- Require- Seismic Safetg Principal Component
- Class Locat'ron* fication d m ent* Category' Notes
- 3. Piping-other (pump N SC D E - (g) suction, pump discharge, drive header)
- 4. Hydraulic control unit 2 SC - B i (k)
- 5. Fine motion drive N C - E -
motor
- 6. CRD N F3 D E -
water pumps
- 7. Control Rod Drive 1/3 C A/- B I
- 8. Electrical modules 3 C,SC - B I with safety-related function
- 9. Cable with safety- 3 C,SC,X - B l related function
- 10. ATWS Equipment N SC - E - (cc) ,
associated with the Alternate Rod insert (ARI) functions C3 Feedwater Control System N C,T,SC, X - E -
C4 Standby Liquid Control Si; stem
- 1. Standby liquid control 2 SC B B i (u) tank including supports Pump including 2 SC B B i (u) 2.
supports
- 3. Pump motor 2 SC - B I
- 4. Valves-injection 1 SC A B i (u)
Valves within injection C,SC A B i (u)
- 5. 1 valves Valves beyond 2 SC B B i (u) 6.
injection valves Notes and footnotes are listed on pages 3.2-54 through 3.2-61 3.2 22 Clessification of Structures Components, and Systems - Amendment 33
. . _ . . . . - - . . . . . ~ . . ..
l 23A6100 Rsv. 3 ;
sawderdseveryaserr sisnever ABWR ;
1 t
l Table 3.2-1 Classification Summary (Continued) !
i Quality ,
Quality Assur- -;
Group ance l Classi- Require- Seismic Safeg Class 1.ocation' fication d m ent* Category' Notes ;
Principal Component
- i C,SC A B i (g,u)
- 7. Piping including 1 supports within injection valves
.2 SC. B B l (g,u)
- 8. Piping including l supports beyond t l
injection valves ,
- 9. Electrical equipment 3/N SC,X -
B/E 1/~ (cc) !
and devices
- 10. Cable 3/N SC,X -
B/E 1/ - - (cc) 7 4
C5 Neutron Monitoring System t 3 SC,X - B l j
- 1. Electrical modules-SRNM, LPRM and i
APRM Cable-SRNM and 3 C,S C,X, - B l -l 2.'
LPRM RZ l
- 3. Detector and tube 2/3 C B/C B I i
'l i
assembly I
C6 Remote Shutdown System !
Components of this system are included under B2, E1, E4, G3, H4, and P2. i 3 C,SC,RZ, - B l
- 1. Electrical modules !
with safety-related X :
functions 3 RZ B l
- 2. Cable with safety- -
related functions ;
Notes and footnotes are listed on peg-c 3.2-54 through 3.2-61 i
F h
3 2-23 Classification of Structures. Componems, and Systems - Amendment 33 l
i
= ,
23A6100 Rsv. 3 ABWR smadartsanyAnna rsissoport l
O\:
Tab e 3.2-1 Classification Summan (Continued)
Quality Ouality Assur-Group ance Classi- Require- Seismic Safeg Principal Component
- Class Locatina* fication d m ent* Cated Notes C7 Reactor Protection System
- 1. Electrical modules 3 SC,X,T, -
B 1 with safety-related RZ functions
- 2. Cable with safety- 3 SC,X,T, - B I related functions RZ,
- 3. Deleted
- 4. Deleted C8 Recirculation Flow Control N X -
E -
System C9 Automatic Power Regulator N X -
E -
System C10 Steam Bypass and Pressure N X -
E -
Control System C11 Process Computer (includes N X -
E -
PMCS & PGCS)
C12 Refueling Platform Control N SC -
E -
Computer C13 CRD Removal Machine N SC -
E -
Control Computer Notes and footnotes are listed on pages 3.2-54 through 3.2-61 O.
3.2 24 Classincstion of Structures, Components, and Systems - Amendment 33
~ ~ , .. .-
~
23A6100 Rsv. 3 saaerd serety Analysis neport .
ABWR :
Table 3.2-1 Classification Summary (Continued) l Quality . .;
Quality Assur- !
Group .ance_ l Classi- Require- Seismic Principal Component' Safg Class Location' fication d m ent' Category' Notes f i
D1 Process Radiation Monitoring System (includes gaseous and liquid effluent monitoring) l i
- 1. Electrical modules- 3 SC,X,RZ - B I -l with safety-related functions (including i menitors) .
L 3 SC,X,RZ B l j
- 2. Cable with safety- -
j related functions Electrical Modules, N T,SC,RZ, - E -
(u) f 3.
other X,W f Cables, other N T,SC,RZ, - E -
-(u) _i
- 4. -
X,W t
D2 Area Radiation Monitoring N X,T,W, -- E -
l System SC,RZ,H ,
i P
D3 Containment Atmospheric Monitoring System Component with 3 C,SC,X - B i ;j 1.
safety-related j
function f
Notes and footnotes are listed on pages 3.1-54 through 3.2-61 l
?
I a
t r
f
~f i
f i
i i
32 l CIsssification of Structures. Components. and Systems - Amendment 33 l
23A6100 RDv,3 ABWR senasntsureryArtysisn:pon O
Table 3.2-1 Classification Summary (Continued)
Quality Ouality Assur-Group ance Classi- Require- Sels.nic Principal Component
- Safeg Class Location
- fication d ment
- Category' Notes E1 RHR System
- 1. Heat exchangers- 2 SC B B 1 primary side i
- 2. Deleted Piping including C,SC A B i (g)
- 3. 1 supports within outermost isolation valves
- 4. Containment spray 2 C,SC B B 1 piping including supports and spargers, within and including the outer most isolation valves 2 SC B B I (g) da. Piping including 2 supports beyond outermost isolation valves Main Pumps including 2 SC B B I 5.
supports
- 6. Main Pump motors 2 SC B B l C,SC A B 1 (g)
- 7. Valves-isolation, 1 (LPFL line) including shutdown suction line isolation valves Valves-isolation, 2 SC B B I (g) 8.
other (pool suction valves and pool test retum valves)
SC B B i (g)
- 9. Valves beyond 2 isolation valves Jockey pumps and 2 SC B B i 10.
motors including supports Notes and footnotes are listed on pages 3.2-54 through 3.2-61 _
O 3.2 26 Classification of Structures. Components, and Systems - Amendment 33
23A6100 Rsv. 3 ABWR smederdsareryAntraiaopers O
Table 3.2-1 Classification Summary (Continued)
Quality Quality Assur-Group anos Classi- Require- Seismic Principal Component
- Safet{
Class Locationfication
- d- m ent* - Category' Notes 11.- Valves to fire N SC - E~ -
protection, Subsystem C (F100C, F103C and l F104C)
E2 High Pressure Core Flooder System
- 1. Reactor pressure 1/2 C,SC A/B B i (g) vessel injection line and connected piping including supports within outermost isolation valve t
- 2. All other piping 2 SC,0 B B l (g) including supports *
- 3. Main Pump 2 SC B B l O 4 Main Pump Motor 3 SC C,SC A
B B
l i (g)
- 5. Valves-other isolation 1 and within the reactor pressure vessel injection line and -
connected lines
- 6. All other valves 2/3 SC B/C B i (g)
- 7. Electrical modules 3 C,SC,X - B I with safety-related functions
- 8. Cable with safety- 3 C,SC,X - B 1 related functions E3 Leak Detection and Isolation System 1.' Temperature sensors 3/N C,SC,T - B/E 1/- (z)
- 2. Pressure transmitters 3 C,SC - B- t/- (z)
- 3. Differential pressure 3 C,SC - B 1/- (z) transmitters (flow)
- 4. Fission Product N SC - E I Monitor O 5. Isolation Valves ' 2/N SC B/C B/E .1 Notes and footnotes are listed on pages 3.2-54 through 3.2-61 Classi& cation of Structures, Components, and Systems - Amendment 33 3227-
23A6100Rzv 3 ABWR standardsareryAmarrsis Report O
Table 3.2-1 Classification Summary (Continued)
Quality Quality Assur-Group ance Classi- Require- Seismic Safetg d Principal Componer.t* Class Location
- fication m ent* Category' Notes
- 6. Instrument lines 3 C,SC B B I
- 7. Sample lines / 2/N C,SC C/D/- B!E t/-
- 8. Flow transmitters N SC -
E -
- 9. Electrical modules 3/N SC,RZ.X -
B/E I/-
- 10. Cables 3/N SC,RZ,X -
B/E I/-
E4 RCIC System l 1. Piping including 1/2 C SC A/B B 1 supports within outermost isolation valves
- 2. Piping including N SC C E - (g) supports-discharge line from vacuum pump to containment isolation valves, and discharge line from condensate pump to the first globe valve
- 3. Piping including 2/3 C,SC B/C B I (g) supports beyond outermost isolation valves up to the turbine exhaust line to the suppression pool, including turbine inlet and outlet drain lines 4 RCIC Pump and piping 2 SC B B I (g) including support, CST suction line from the first RCIC motorized valve, S/P suction line to the pump, discharge line up to the FW line "B" thermal sleeve
- 5. Pump motors N SC -
E I Notes and footnotes are listed on pages 3.2-54 through 3.2-61 3.2-28 Classification of Structures, Components, and Systems - Amendment 33
~.
4
- 23A6100 Rsv. 3 ABHn sundantsarnyAnalysis nopors O
V :,
Table 3.2-1 Classification Summary (Continued) ;
Quality ;
Quality Assur.' i Group ance !
Classi- Require- Seismic Safetg Category' Notec Principal Component
- Class Location
- fication d mont*
Valves-outer 1/2 C,SC A/B B i (g) 6.
isolation and within Valves-outside the 2 SC B B i (g) ;
7.
PCV (except item 8)
Valves-beyond N SC C E I (g) l 8. '
turbine inlet drain line second shutoff l
- 9. Turbine including 2 SC - B i (m) supports
- 10. Electrical modules 3 C SC,X - B I with safety-related functions I
- 11. Cable with safety- 3 C,SC,X - B I related functions l
- 12. Other mechanical and N SC,X -
E - i electrical modules j i
l Fuel Servicing Equipment N/2 SC -/B E/B -
(x) .l l F1 F2 Miscellaneous Servicing N SC,RZ -- E -
A Equipment :
i SC -/B E/B -/l (gg)
, l F3 RPV Servicing Equipment N/2 F4 RPV Internal Servicing N SC - E Equipment :
r F5 Refueling Equipment Refueling equipment N SC - E I (bb) i
- 1.
machine assembly l l
O Notes and footnotes are listed on pages 3.2-54 through 3.2-61 l
i I
l
^
3.2-29 C!sssificst:an of Structures. Components, and Systems - Amendment 33 l
23A6100 R;v. 3 ABM6R smaderdsareryAmarysis sepan l e
O Table 3.2-1 Classification Summary (Continued) ,
l Quality !
Quality Assur-Group ance i Classi- Require. Seismic Principal Component' Safeg Class Location
- fication d
m ent* Category' Notes F6 Fuel Storage Equipment
- 1. Fuel and equipment N SC -
E I (bb) l storage racks-new and spent
- 2. Defective fuel N SC - E -
(bb) l container l 3. Spent fuel pool liner N SC -
E I F7 Under-Vessel Servicing N SC -
E -
(bb)
Equipment F8 CRD Maintenance Facility N SC -
E -
F9 Internal Pump Maintenance N SC -
E -
Facility F10 Fuel Cask Cleaning Facility N SC - E -
F11 Plant Start-up Test N M -
E -
Equipment F12 Inservice inspection N M -
E -
Equipment G1 Reactor Water Cleanup System
- 1. Vessels including N SC C E -
supports (filter /
demineralizer)
- 2. Regenerative heat N SC C E -
exchangers including supports carrying reactor water Notes and footnotes are listed on pages 3.2-54 through 3.2-61 3.2-30 Classification of Structures, Components and Systems - Amendment 33
23As100 nn. 3 1 AMR standardsafetyAnalysis Ropon m
O l Table 3.2-1 Classification Summary (Continued) l Quality j Quality Assur- !
Group ance Classi- Require- Seismic Principal Component
- Safeg Class Location
- fication d ment' Category' Notes
~
- 3. Cleanup recirculation N SC C E -
pump, m'otors 4 Piping including 1 C,SC A B i (g) supports and valves within and including outermost containment isolation valves
- 5. Pump suction and N SC C E -
discharge piping including supports and valves from containment isolation valves back to and including shut-off valve at feedwater line
\g connection
- 6. Piping including N SC C E -
l supports and valves l
leading to radwas'<-
and main condensei
- 7. Non-regenerative heat N SC C E -
exchanger tube inside and piping including supports and valves carrying process water
- 8. Non-regenerative heat N SC D E -
exchanger shell and piping including supports carrying closed cooling water
- 9. Filter /demineralizer N SC D E -
precoat subsystem .
Filter demin holding N SC C E -
10.
pumps including supports-valves and piping including supports h D -
Sample station N SC E 11.
v' Notes and footnotes are listed on pages 3.2-54 through 3.2-61 i
32'31 Cisssification of Structures. Components, and Systems - Amendment 33
23A6100 Rev. 3 ABWR standardsafety Analysis neport O
Table 3.2-1 Classification Summary (Continued)
Quality Quality Assur-Group ance Classi- Require- Seismic Safeg Principal Component' Class Location
- fication d ment' Category' Notes
- 12. Electrical modules and N GC,X - E -
cable with no safety-related functions
- 13. Electrical modules and 3 SC -
B l cable for isolation valves G2 Fuel Pool Cooling and Cleanup System
- 1. Vessels including N SC D E -
supports-filter /
demineralizers
- 2. Piping and valves N SC D E -
including supports upstream of F/D outlet isolation valve
- 3. Piping and valves N SC D E -
including supports downstream of F/D inlet isol:: tion valve
- 4. Heat exchangers N SC C E 1 including supports 5, Pumps including N SC C E I supports
- 6. Pump motors N SC -
E -
- 7. Piping including N SC C E I supports and valves-cooling portion
- 8. Makeup Water System N SC C E I (MUWC) connection including valves and supports
- 9. KHR piping 3 SC C B l connections and valves including supports for safety-related makeup and supplemental cooling Notes and footnotes are listed on pages 3.2-54 through 3.2-61 I 3.2-32 Class l6 cation of Structures, Components. and Systems - Amendment 33
23A6100 Rev. 3 ABWR standardsarery Analysis neport Table 3.2-1 Classification Summary (Continued)
Quality >
Ouality Assur-Group ance Classi- Require- Seismic Safe'g Category' Notes Principal Component
- Ctrss Location
- fication d ment *
- 10. SPCU piping 3 SC C B 1 connections and valves including supports
- 11. Electrical modules and N SC,X -
E -
cables with no safety-re ated function ,
G3 Suppression Pool Cleanup System
- 1. Isolation valves and 2 SC 8 B I piping including supports within outermost isolation valves Pump including N SC C E I 2.
supports Pump motor N SC - E -
3.
- 4. Piping and N SC C E I components beyond outermost-containment isolation valve including supports
- 5. Deleted ,
- 6. Deleted
- 7. Electrical modules and N SC,X - E -
Cables with no safety-related function Electrical modules and 3 SC,X - B i 8.
cables for isolation valves H1 Main Control Room Panels
- 1. Panels 3/N X - 8/E 1/- (aa) ;
i
- Notes and footnotes are listed on pages 3.2-54 through 3.2-61 l
3 2-33 Cisssification of Structures. Components, and Systems - Amendment 33 I
= .
23A6100 Rev 3 ABWR standardsafety Anasysis aeport O
Table 3.2-1 Classification Summary (Continued)
Quality Quality Assur-Group ance Classi- Require- Seismic Principal Component
- Safetg Class Location
- fication d
ment
- Categoryf Notes
- 2. Electrical Modules 3 X - B 1 with safety-related functions
- 3. Cable with safety- 3 X - B I related functions
- 4. Other mechanical and N X - E -
electrical modules H2 Control Room Back Panels '
- 1. Panels 3/N X -
B/E 1/- (aa)
- 2. Electrical modules 3 X - B I with safety-related function Cable with safety- 3 X - B l 3.
related function
- 4. Other mechanical and N X - E -
electrical modules H3 Radioactive Waste Control N W - E - (p)
Panels H4 Local Control Panels
- 1. Panels and Racks 3/N RZ,SC,X -
B/E I/- (aa)
- 2. Electrical modules 3 RZ,SC,X - B I with safety-related functions Cable with safety- 3 RZ,SC,X - B I 3.
related functions
- 4. Other mechanical and N RZ,SC,X - E -
electrical modules H5 Instrument Racks Notes and footnotes are listed on pages 3.2-54 through 3.2-61 3.2-34 Classification of Strucrures Components, and Systems - Amendment 33
23A6100 Rev. 3 ABWR s=ndard sarery Analysis Report I
-(D
)
Table 3.2-1 Classification Summary (Continued)
Quality Ouality Assur- :
Group ance Classi- Require- Seismic Principal Component
- Safetg Class Location
- fication d
ment
- Category' Notes H5 instrument Racks l 1. Mechanical and 3 SC,RZ, -
B -
I electrical with safety- X,W,M l
related functions
- 2. Other mechanical and N SC,RZ,X, -
E -
selected modules T i
H6 Multiplexing System ,
- 1. Electrical module with 3 SC RZ,X -
B I ~
l safety-relatedfunctions (Essential)
- 2. Cable with safety- 3 SC,RZ,X -
B I O related functions (Essential)
]
l,
- 3. Other electrical N SC.RZ X, -
E -
modules and cables W I 1
(Non-essential)
H7 Local Control Boxes
- 1. Electrical modules 3 S C,RZ,X, -
B l with safety-related H,T,W.M functions
- 2. Other electrical N SC,RZ,X, -
E -
modules H,T,W,M, i
i l
J1 Fuel Assembly
~ 1. Fuel assemblies 3 C,SC -
B l
- 2. Control Rods 3 C,SC -
B 1
- 3. Loose Parts Monitoring N C,SC -
E -
System J2 Fuel Channel 3 C.SC -
B l Notes and footnotes are listed on pages 3.2-54 through 3.2-61 Classification of Structures. Components, and Systems - Amendment 33 3.2-35
1 j
22As100 Rsv. 3 standard safery Antysis auport ABWR Table 3.2-1 Classification Summary (Continued) .
Quality Quality Assur.
Group once Classi- Require- Seismic Safeg Class Location
- fication d m ent' Category' Notes Principal Component
- K1 Radwaste System N C,H,SC,T, D E - (p) f
- 1. Drain piping including supports and valves- W,X j radioactive
- 2. Deleted 2 C,SC B B l
- 3. Piping and valves-containment isolation N C.SC B B 1
- 4. Piping including supports and valves forming part of containment boundary _
W E - (p)
- 5. Pressure vessels N -
including supports i E - (p) i Atmospheric tanks N C,SC.H,T, -
- 6. i including supportc W (p)
- 7. 0-15 PSIG Tanks and N W - E -
supports C.SC,W E - (p)
- 8. Heat exchangers and N -
supports C SC,H, E - (p)
- 9. Piping including N -
supports and valves T,W ,
N ALL D E - (p) l
- 10. Other mechanical and ,
electrical modules s C B i
- 11. ECCS equipment room N SC sump backflow protection check valves N1 Turbine Main Steam System
- 1. Deleted (see B2.5) r
- 2. Deleted (see B2.6)
Notes and footnotes are listed on pages 3.2-54 through 3.2-61 Cisssification of Structures. Components. and Systems - Amenc~r-' :)
12-36 t
23A6100 Rev. 3 ABWR standard satory Anarysis aeport 10 O
Table 3.2-1 Classification Summary (Continued)
Quality Quality Assur-Group ance Classi- Require- Seismic Principal Component
- Safetg Class Location
- fication d
ment' Category' Notes N2 Condensate, Feedwater and Condensate Air Extraction System
- 1. Feedwater system N SC,T D E -
(ee) components beyond seismic interface restraint N3 Heater. Drain and Vent N T - E -
System N4 Condensate Purification N T - E -
System N5 Condensate Filter Facility N T - E -
N6 Condensate Domineralizer N T -- E -
N7 Main Turbine N T - E -
N8 Turbine Control System
- 1. Turbine stop valve, N T D E -
(I)(n) turbine bypass valves, (o)(r) and the main steam leads from the turbine stop valve to the turbine casing N9 Turbine Gland Steam N T D E -
System i
N10 Turbine Lubricating Oil N T - E -
System j l
Notes and footnotes are listed on pages 3.2 54 through 3.2-61 Classi6 cation of Structures, Components and Systems - Amendment 33 3237 i
l
23A6100 Rev. 3 l ABWR standardsafety Analysis Report !
O Table 3.2-1 Classilcation Summary (Continued)
Quality Quality Assur-Group ance Classi- Require- Seismic Safetg Category' Notes Principal Component
- Class Location
- fication d m ent*
N11 Moisture Separator Heater N T -
E -
N12 Extraction System N T -
E -
N13 Turbine Bypass System
- 1. Turbine bypass N T D E -
(r) piping including supports up to the condenser N14 Reactor Feedwater Pump N T - E -
Driver N15 Turbine Auxiliary Steam N T -
E -
System N16 Generator N T - E -
N17 Hydrogen Gas Cooling N T - E -
System N18 Generator Cooling System N T - E -
N19 Generator Sealing Oil N T - E -
System N20 Exciter N T - E -
N21 Main Condenser N T -
E -
N22 Offgas System N T -
E -
Notes and footnotes are listad on pages 3.2-54 through 3.2-61 O
i 3.2 38 Classif; cation of Structures, Components, and Systems - Amendment 33
23AG100 Rev. 3 standard sarery Analysis neport ABWR Table 3.2-1 Classification Summary (Continued)
Quality Quality Assur-Group ance Classi- Require- Seismic Safeg Class Location
- fication d ment
- Category' Notes Principal Component
- N23 Circulating Water System N T D E -
N T - E -
N24 Condenser Cleanup Facility N M E -
P0 Makeup Water System (Preparation)
P1 Makeup Water System (Purified)
C B B l
- 1. Piping including 2 supports and valves forming part of the containment boundary N O D E -
V 2. Demineralizer water storage tank including supports
- 3. Piping including N O D E -
supports and valves
- 4. Other components N O D E -
3 P2 Makeup Water System (Condensate)
N O D E -
(w)
- 1. Condensate storage tank including supports SC B B i
- 2. Condensate header- 2 piping including supports, level 7
instrumentation and valves
- 3. Piping including N O D E i
suppons and valves and other components l1 P3 Reactor Building Cooling Water System Notes and footnotes are listed on pages 3.2-54 through 3.2-61 ,
O i V i 3239 Classification of Structures, Components, and Systems - Amendment 33 i
c
a .
23A6100 Rev. 3 ABWR standard safety Anarysis separt O
Table 3.2-1 Classification Summary (Continued)
Quality Quality Assur-Group ance Classi- Require- Seismic Safetg Principal Component' Class Location
- fication d ment
- Category' Notes Piping and valves 2 SC,C B B i (g) 1.
forming part of primary containment boundary
- 2. Other safety-related 3 SC,C C B i piping including supports, pumps and valves 3 SC.C,X - B i
- 3. Electrical modules with safety-related functions
- 4. Cable with safety- 3 SC,C,X - B I related functions
- 5. Other mechanical and N SC,C,X,M D E -
electrical modules P4 Turbine Building Cooling N T D E -
Water System P5 HVAC Normal Cooling Water System
- 1. Piping including N C,SC B E I l
supports and valves forming part of containment boundary
- 2. Other mechanical and N C,SC,RZ - E -
electrical modules T,X, l
l Notes and footnotes are listed on pages 3.2-54 through 3.2-61 l
1 1
Ol 3.2,40 CisssiScstion of Structures, Components and Systems - Amendment 33 1
23A6100 Rev. 3 ABWR standardsevery Analysis Repois
'(x Table 3.2-1 Classification Summary (Continued)
Quality Quality Assur-Group ance Classi- Require- Seismic Principal Component
- Safetg Class 1.ocation' fication d ment
- Category' Notes P6 HVAC Emergency Cooling Water System
- 1. Chillers, pumps, 3 SC,X C B I valves, and piping, including supports
- 2. Electrical modules and 3 RZ,X -
B I cable with safety-related functions P7 Oxygen injection System N T - E -
3 O C B i P8 Ultimate Heat Sink d P9 Reactor Service Water System
- 1. Safety-related piping 3 U,0,X C B i including supports, piping and valves
- 2. Electrical modules and 3 U,0,X - B l cables with safety-related functions
- 3. Other non-safety- N U,0,X - E -
related mechanical and electrical modules P10 Turbine Service Water System
- 1. Non-safety-related N P, O, T - E -
piping including supports, piping and valves
- 2. Electrical modules and N P,0,T - E -
cables with non-safety-related functions Notes and footnotes are hsted on pages 3.2-54 through 3.2-61 32M Cisssification of Structures. Components. and Systems- Amendment 33
23A6100 Rev. 3 ABWR standardsarety Aa,1ysis neport O
Table 3.2-1 Classification Summary (Continued)
Quality Quality Assur-Group ance 7 Classi- Require- Seismic Principal Component
- Safet{
Class Locationfication
- d ment
- Categoryf Notes P11 Station Service Air System
- 1. Containment isolation 2 C,SC B B i including supports, valves and piping
- 2. Other non-safety- N SC,RZ, -
E -
related mechanical X,T,H, and electrical W,C components P12 Instrument Air Service
- 1. Containment isolation 2 C,SC B B I including supports, valves and piping
- 2. Other non-safety- N SC;X C B -
l related mechanical components
- 3. Other non-safety- N SC,RZ,X, -
E -
related electrical T,H, W,C components P13 High Pressure Nitrogen Gas Supply Systems
- 1. Containment isolation 2 C.SC B B I i
including supports, valves and piping
- 2. Gas bottles, piping 3 SC,C C B 1 and valves including supports with safety-related functions
~
- 3. Electric modules with 3 SC,RZ,X - B I safety-related functions 4 Cable with safety- 3 SC,RZ,X - B i related functions
- 5. Other non-safety- N SC,RZ,X C E -
l related mechanical '
components Notes and footnotes are listed on pages 3.2-54 through 3.2-61 l ll 3.2-42 Classification of Structures, Components, and Systems - Amenoment 33 l i
23A6100 Rsv. 3 ABWR stugudsareryAaarysis nevon i
Table 3.2-1 Classification Summary (Continued)
Quality Quality Assur-Group ance Classi- Require- Seismic Safeg Principal Component
- Class Location
- fication d m ent* Category' Notes
- 6. Other non-safety- N SC,RZ,X -
E -
l related electrical components P14 Heating Steam and N ALL -- E -
Condensate Water Return System P15 House Boiler N T -
E -
P16 Hot Water Heating System N ALL -
E -
P17 Hydrogen Water Chemistry N T - E -
( ,)
System
%d P18 Zinc injection System N T - E -
P19 Breathing Air System N C,SC,T - E -
P20 Sampling System (includes N SC,RZ,T - E -
PASS)
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,C,X, - B l distribution equipment RZ inc!uding inverters
- 2. Safety-related Motors 3 SC,C,X,Z, - B l O
Notes and footnotes are listed on pages 3.2-54 through 3.2-61 Cassification of Structures, Components, and Systems - Amendment 33 3243
23A6100 lbv. 3 ACWR starsd:n1suretyAnalysbaoport O
Table 3.2-1 Classification Summary (Continued)
Quality Quality Assur-Group ance Classi- Require- Seismic Principal Component
- Safeg Class Location' fication d m ent* Category' Notes *
- 3. Safety-related 3 SC,X,RZ -
B i Protective relags and control panels
- 4. Safety-related Valve 3 SC,C, X, -
B l ,
l operators RZ l
R2 Unit Auxiliary Transformers
- 1. Unit Auxiliary N O -
E -
Transformers
- 2. Safety-related 3 SC,X,RZ -
B l l
Transformers R3 Isolated Phase Bus N O,T -
E -
R4 Non. Segregated Phase Bus N O,T - E -
R5 Metalciad Switchgear
- 1. Safety-related 6900 3 SC,RZ -
B I l
Volt switchgear R6 Power Center
- 1. Safety-related 480 Volt 3 SC,RZ - B I power centers R7 Motor Control Center
- 1. Safety-related 480 Volt 3 SC,X,RZ -
B i motor control centers Notes and footnotes are listed on pages 3.2-54 through 3.2-61 0
3.2-44 CIsssincation of Structures, Components, and Systems - Amendment 33
23A6100 nsv. 3 ABWR steedertseiery Anstrsis neport l a l Table 3.2-1 Classification Summary (Continued) l Quality !'
Quality Assur-Group ance ;
Require-Safeg Classi- Seismic Principal Component
- Class Location
- ficationd ment
- Category' Notes i R8 Raceway System
- 1. Safety-related control 3 SC,C,X, -
B I l and power cables RZ (including ;
underground cable ;
systems, cable splices, ;
connectors and j terminal blocks l
- 2. Safety-related conduit 3 SC,C,X -
B l and cable trays and RZ their supports l R9 Grounding Wire N SC,C,X,0 - - -
O' u 1
l R10 Safety-related Electrical 3 SC,C -
B l -!
Wiring Penetrations i R11 Combustion Turbine N T -
E -
Generator i
)
R12 Safety-related Direct Current Power Supply
- 1. 125 Volt batteries, 3 SC C,X, -
B I battery racks, battery RZ chargers, and distribution equipment
- 2. Protective relays and 3 SC,X,RZ -
B i control panels
- 3. Motors 3 SC,C,X, -
B l RZ ,
I Notes and footnotes are listed on pages 3.2-54 through 3.2-61 O
~l i
Classification of Structures, Components, and Systems - Amendment 33 3.245 i
.-,,,r.-- .--.n
- n.,--.-,-, .,n v.m - - - r+-e- ,
l 4
23A6100 thv. 3 ABWR stankntseieryAnar sisr auport 91 Table 3.21 Classification Summary (Continued) )
Quality Quality Assur-Group ance Classi- Require- Seismic .
Safetg '
Principal Component
- Clar,s Location
- fication d m ent* Category' Notes R13 Emergency Diesel Generator Svstem Starting air receiver 3 RZ C B I (y) 1.
tanks piping including supports from and including check valve and downstream piping including supports, valves, and compressors.
Starting air 3 RZ - B 1 l 2.
l compressor motors 3 RZ,0 C B l
- 3. Combustion air intake and exhaust system 4 Safety-related piping 3 RZ C B I l
including supports, valves-fuel oil system, diesel cooling ,
water system, and lube oil system 3 RZ B 1 Pump motors-fuel oil l 5.
system, diesel cooling water system and lube oil system B i (y)
- 6. Diesel generators 3 RZ -
3 RZ,X - B 1 r l 7. Mechanical and electrical modules with safety-related functions
- 8. Cable with safety- 3 RZ,0,X - B I related functions
- 9. Other mechanical and N RZ,0 - E electrical modules 3 X - B 1 R14 Safety-related Vital AC Power Supply Notes and footnotes are listed on pages 3.2-54 through 3.2-61 O
Cisssification of Structures, Components. and Systems - Amendment 33
'3.24
23A6100 n:v. 3 ABWR standantsarery Anarrsis neport
/D V
Table 3.2-1 Classification Summary (Continued)
Quality Quality Assur-Group ance Classi- Require- Seismic Safeg Principal Component
- Class Location' fication d m ent* Category' Notes R15 Safety-related Instrument 3 X -
B l and Control Power Supply R16 Communication System N SC,C,RZ, -
B I l
X
+
R17 Lighting and Servicing Power Supply
- 1. Normal Lighting N ALL - E -
l 2. Standby Lighting 3/N ALL C/- B/E 1/- (hh)
- 3. DC Emergency 3/N SC,X,W C/- B/E 1/- (hh) l Lighting
- 4. Guide Lamp Lighting 3/N SC,X C/- B/E 1/-
O l S1 Reserve Auxilian N O -
E -
Transformer TO Primary Containment System
- 1. Suppression 2 C B B l chamber /drywell vacuum breakers T1 Frimary Containment Vessel
- 1. Primary containment 2 C B B i vessel (PCV)-
reinforced concrete containment vessel (RCCV)
- 2. Vent system (vertical 2 C B B I flow channels and horizontal discharges
- 3. PCV penetrations and 2 C B B I drywell steel head
- 4. Upper and lower 2 C,5C - B i
(
drywell airlocks Notes and footnotes are listed on pages 3.2-54 through 3.2-61 ;
i Cisssification of Structures. Components and Systems - Amendment 33 32M r
I
i
. l 23A6100 R2v. 3 standardsareerAnarrsis aupon i ABWR O!1 Table 3.2-1 Classification Summary (Continued) l i
Quality Quality Assur-Group ance Classi- Require- Seismic Safeg d ment
- Category' Notes Principal Component
- Class Location
- fication 2 C,SC - B i
- 5. Upper and lower drywell equipment hatches 2 C - B l
- 6. Lower drywell access tunnels 2 C,SC - B 1
- 7. Suppression chamber access hatch 3 C,SC - B l
- 8. Safety-related instrumentation T2 ConteinmentInternal Structures
- 1. RPV stabilizer truss (see B1.2) 3 C - B 1
- 2. Support structures and equipment for safety-related piping 3 C - B l
- 3. Diaphragm Floor C B I
- 4. UD equipment and 3 -
personnel tunnels 3 C - B 1
- 5. Miscellaneous Platforms T3 R.PV Pedestal 3 C - B l
- 1. RPV pedestal and shield wall T4 Standby Gas Treatment System 3 SC,RZ C B 1
- 1. All equipment except deluge piping and valves l
N SC - E -
l T5 PCV Pressure and Leak j Testing Facility Notes and footnotes are hsted on pages 3.2-54 through 3.2-61 l
i Dessification of Structures Components, and Systems - Amendment 33 3.24 j
23A61C0 Rav. 3 standardsatory Analysis soport ABWR c O
LJ Table 3.2-1 Classification Summary (Continued)
Quality Quality Assur-Group ance Classi- Require- Seismic Safetg d Category' Notes Principal Component
- Class Location
- fication m ent' T6 Atmospheric Control System f
- 1. Nitrogen Storage Tanks N O - E -
- 2. Vaporizers and controls N O -
E -
- 3. Piping including 2 SC B B l supports and valves forming part of conteinment boundary
- 4. Piping including 3 SC C B I supports and valves beyond the first rupture disk up to and including the second O 5.
rupture disk Electrical modules 3 SC,X -
B i with safety-related functions 6, Cables with safety- 3 SC,X - B i related function
- 7. Other non-safety- N SC,RZ,0, -
E -
related mechanical X and electrical components T7 Drywell Cooling System
- 1. Motors N C -
E -
- 2. Fans N C - E -
i Coils, cooling N C -
E - I 3.
- 4. Other mechanical and N C,X -
E -
electrical modules e i
I f T8 Flammability Control 2 SC B B
. System Notes and footnotes are listed on pages 3.2-54 through 3.2-61 Classofication of Structures. Components, and Syste.ns - Amendment 33 3.249 i
1 23A6100 Rsv. 3 :
ABWR standardsarery Aurysis soport l Table 3.2-1 Classification Summary (Continued)
O Quality Quality Assur-Group ance Classi- Require- Seismic Safetg d Principal Component
- Class Location
- fication ment
- Category' Notes T9 Suppression Pool Temperature Monitoring System
- 1. Electrical modules 3 C,X,SC, -
B I ,
with safety-related RZ functions *
- 2. Cable with safety- 3 C,X,SC, -
B i related functions RZ .
U1 Foundation Work 2/3 C,SC,RZ -
B I U2 Turbine Pedestal N T -
E -
U3 Cranes and Hoists
- 1. Reactor Building crane N SC -
E -
(x)
- 2. Refueling Platform N SC -
E -
(x)
Upper Drywell N C E I 3.
Servicing
- 4. Lower Drywell N C - E I Servicing
- 5. Main Steam Tunnel N M - E -
Servicing S. Special Service Rooms N SC,RZ,T, -
E -
W,X U4 Elevator N SC,RZ - E -
l US Heating, Venti!ating and Air Conditioning **
i
- 1. Safety-related equipment" 3 SC,X B f
- a. Fan-coil cooling -
units Notes and footnotes are listed on pages 3.2-54 through 3.2-61 3250 Classification of Structures Components, and Systems - Amendment 33
23A6100 Rsv. 3 ABWR studardsareryAnalysis Report ,
\
Table 3.2-1 Classification Summary (Continued) ;
Quality Quality Assur- .
Group ance Ciassi- Require- Seismic-Principal Component
- Saiog Class Location
- fication d m ent* Category' Notes !
- b. Heating units- 3 SC,RZ,X -
B I i electrical or water ,
- c. Blowers-Air 3 SC,RZ,X -
B I supply or
- d. Ductwork 3 SC,RZ,X -
B 1
- e. Filters- 3 SC,RZ,X - B 1 Equipment areas ,
- f. HEPA Filters, 3 SC,X - B l Charcoal Adsorbers- ;
Control Rooms and Secondary Containment
- g. Valves and 3 SC,RZ - B l i
Dampers-secondary centainment ,
i isolation I
- h. Other safety- 3 H,Z - B I related valves and I dampers ;
- i. Electrical modules 3 SC,RZ, - B l with safety-related H,X )
functions l Cable with safety- 3 SC,RZ, - B l i J.
related functions H,X l Notes and footnotes are listed on pages 3.2-54 through 32-61 -t l
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32-51 Classification of Structures, Components, and Systems - Amendment 33 I
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23A6100 Rev. 3 ABWR standenf satoryAcufysis neport O
Table 3.2-1 Classification Summary (Continued)
Quality Quality Assur-Group ance Classi- Require- Seismic Safeg Principal Component
- Class Location
- fication d ment
- Category' Notes
- 2. Non-safety-related equipment **
- a. HVAC mechanical N SC,RZ,H -
E -
or electrical X,W,T components with non-safety-related functions
- b. Non-safety-related N SC,RZ,H, -
E -
(t)(u) fire protection X,W,T valves and dampers U5.1 Potable and Sanitary Water System
- 1. Potable and sanitary N All -
E -
water equipment (except C,M)
- 2. Drain piping including N All D E -
supports and valves- (except nonradioactive C, M,X) i U6 Fire Protection Sptem
- 1. Other piping including N SC,C,X, D E -
(t) (u) supports and valves RZ,H,T, W,0
- 2. Water storage tank N F D E -
(t) (u)
- 3. Pumps N F D E -
(t) (u)
- a. Motor Driven N F D E I (ff)
- b. Engine Driven N F D E -
(t)(u)
- 4. Pump motors N F - E I (ff)
- 5. Electrical Modules N C,SC,X - E -
(t) (ui !
RZ,H, T,W
- 6. Deleted Notes and footnotes are listed on pages 3.2-54 through 3.2-61 i Classification of Structures, Components, and Systems - Amant 33 3.2-52
23A6100 Rev. 3 ABWR sundardsurety Analysis neport rh
'Q Table 3.2-1 Classification Summary (Continued) :
Quality Quality Assur-Group ance Classi- Require- Seismic Safeg d Category' Notes Principal Component
- Class Location
- fication m ent*
- 7. Cables N SC.C,X - E -
(t) (u)
- 8. Sprinklers or deluge N H,W,SC, D E -
(t) (u) water X,RZ,T
- 9. Foam, reaction or N RZ,T -
E -
(t) (u) deluge U7 Floor Leakage Detection N SC,RZ -
E -
System :
U8 Vacuum Sweep System N C,SC -
E -
C U9 Decontamination System N C,SC,RZ -
E -
( T,W,S,X U10 Reactor Building 3 SC,RZ -
B I U11 Turbine Building N T -
E -
(v)
U12 Control Building 3 X -
B I U13 Radwaste Building
- 1. Structural walls and N W - E -
slabs above grade level (see Subsection 3H.3.3)
- 2. Radwaste Building 3 W - B I Substructure U14 Service Buiiding N H - E -
Stack N RZ -
E -
Y1 Notes and footnotes are listed on pages 3.2-54 through 3.2-61 Classification of Structures, Components, and Systems - Amendment 33 3 2-53
23A6100 nsv 3 ABWR standardsaaryAnaysis neport i
O 1 Table 3.2-1 Classification Summary (Continued) l Quality Ouality Assur-Group ance Classi- Require- Seismic Principal Component
- Safeg Class Location
- ficationd ment
- Category' Notes Y2 Diesel Generator Fuel Oil 3 O,RZ -
B I Storage and Transfer System Y3 Site Security N ALL -
E -
Notes and footnotes are listed on pages 3.2-54 through 3.2-61 Table 32-1 Notes and Footnotes
- The RHR/ECCS low pressure flooder spargers are part of the reactor pressure vessel system, see item B1.5.
t The ECCS high pressure core flooder spargers are part of the Reactor Pres.eure Vessel System, see item B1.5.
- Pool suction piping, suction piping from condensate storage tank, test line to pool, pump discharge piping and return line to pool.
f These sample lines are totally within containment and the fission product monitor provides no isolation function.
" includes Reactor Building, Control Building, and Service Building thermal and radiological environmental control functions within the ABWR Standard Plant.
tt Controls environment in Main and Local control rooms, diesel-generator rooms, battery rooms.
ECCS-RCIC, pump rooms within the ABWR Standard Plant.
- Controls environment in rooms or areas containing non-safety-related equipment within the ABWR Standard Plant,
- a. A module is an assembly ofinterconnected components which constitute an identifiable device or piece of equipment. For example, electrical modules include sensors, power supplies, signal processors, and mechanical modules include turbines, strainers, and orifices.
- b. 1,2,3, N = Nuclear safety-related function designation denned in Subsections 3.2.3 and 3.2.5.
- c. C = Primary Containment H = Service building M = Reactor building steam tunnel O = Outside onsite RZ = Reactor Building Clean Zone (balance portion of the reactor building outside the Secondary Containment Zone)
SC = Secondary Containment portion of the reactor building T = Turbine Building 3 2-54 CIssstfication of Structures, Components, and Systems - Amendment 33
23A6100 Rev. 3 ABWR standardsareryAnatysis noport .
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W = Radwaste Building l
- X = Control Building F = Firewater Pump House
- U = Ultimate Heat Sink Pump House
- P = Power Cycle Heat Sink Pump House *
- d. A,B,C,D= Quality groups defined in Regulatory Guide 1.26 and l
Subsection 3.2.2. The structures, systems and components are designed and constructed in accordance with the requirements identified in Tables 3.2-2 and 3.2-3.
- = Quality Group Classification not applicable to this equipment.
- e. B = The quality assurance requirements of 10CFR50, Appendix B are applied in accordance with the quality assurance program described in Chapter 17.
t E = Elements of 10CFR50, Appendix B are generally applied, commensurate with the importance of the equipment's function,
- f. I = The design requirements of Seismic Category I structures and equipment are applied as described in Section 3.7, Seismic Design.
- = The seismic design requirements for the safe shutdown earthquake (SSE) are not applicable to the equipment. However, the equipment that is not safety-related but which could damage Seismic Category I equipment if its structural integrity failed is checked analytically and designed to assure its integrity under seismic loading resulting from the SSE.
- g. 1. Lines one inch and smaller which are part of the reactor coolant pressure boundary and are ASME Code Section III, Class 2 and Seismic Category I.
- 2. Allinstrumentlin es which are connected to the reactor coolant pressure
- boandary and are utilized to actuate and monitor safety systems shall be Safety Class 2 from the outer isolation valve or the process shutoffvah e (root valve) to the sensing instrumentation.
- 3. All instrument lines which are connected to the reactor coolant pressure i
boundary and are not utilized to actuate and monitor safety systems shall O
- Pump House structures are out of the ABWRStandard Plant scope.
Classification of Structures Components, and Systems- Amendment 33 3265
23A6100 Riv. 3 ACWR sund:rdcareerAnstrsis Report O;
be Code Group D from the outer isolation vahe or the process shutoff valve (root valve) to the sensing instnimentation. -
- 4. All other instrument lines:
- i. Through the root valve the lines shall be of the same classification as the system to which they are attached. ,
ii. Beyond the root valve, if used to actuate a safety system, the lines shall be of the same classification as the system to which they are attached.
iii. Beyond the root valve, if not used to actuate a safety system, the lines may be Code Group D.
- 5. All sample lines from the outer isolation valve or the process root valve through the remainder of the sampling system may be Code Group D.
- 6. All safety-related instnament sensing lines shall be in conformance w?th the criteria of Regulatory Guides 1.11 and 1.151.
- h. Safety / Relief valve discharge line (SRVDL) piping to the quencher shall be Quality Group C and Seismic Category I. In add.ition, all welds in the SRVDL piping in the wetwell above the surface of the suppression pool shall be non-destructively examined to the requirements of ASME Boiler and Pressure ,
Vessel Code,Section III Class 2.
SRVDL piping from the safety /reliefvalve to the quenchers in the suppression pool consists of two parts: the first part is located in the drywell and is attached at one end to the safety / relief valve and attached at its other end to the diaphagm floor penetration. This first part of the SRVDL is analyzed with the main steam piping as a complete system. The second part of the SR\T)L is in the wetwell and extends from the penetration to the quenchers in the suppression pool. Because of the penetration on this part of the line,it is physically decoupled from the main steam piping and the first part of the SRVDL piping and is, therefore, analyved as a separate piping system.
- i. Electrical devices include components such as switches, controllers, solenoids. )
fuses. junction boxes, and transducers which are discrete components of a larger subassembly / module. Nuclear safety-related devices are Seismic l Categog I. Fail-safe devices are non-Seismic Category I. l
- j. The control rod driver insert lines from the drive flange up to and including .l l
the first valve on the hydraulic control unit are Safety Class 2, and non-safety-related beyond the first valve.
i 3.2-S6 Cisssification of Structures. Components. and Systems - Amendment 33 j
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23A6100 Rxv. 3 ACWR standantsafery Analysis neport q
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- k. The hydraulic control unit (HCU) is a factory-assembled engineered module j of valves, tubing, piping, and stored water which controls two control rod i dnves by the application of pressures and flows to accomplish rapid insertion for reactor scram.
Although the hydraulic control unit, as a unit, is field installed and connected to process piping, many ofits internal parts differ markedly from process piping components because of the more complex functions they must provide. Thus, although the codes and standards im oked by Groups A, B, C, and D pressure integrity qualitylevels clearly apply at all levels to the interfaces between the HCU and the connection to conventional piping components (e.g., pipe nipples, fittings, simple hand valves, etc.), it is considered that they do not apply to the specialty parts (e.g., solenoid valves, pneumatic components, and instruments).
The design and construction specifications for the HCU do invoke such codes and standards as can be reasonably applied to individual parts in developing required quality levels, but of the remaining parts and details. For example:
(1) all welds are LP inspected; (2) all socket welds are inspected for gap p between pipe and socket bottom; (3) all welding is performed by qualified D welders; and (4) all work is done per written procedures. Quality Group D is generally applicable because the codes and standards invoked by that group contain clauses which permit the use of manufacturer standards and proven design techniques which are not explicitly defined within the codes for Quality Groups A, B, or C. This is supplemented by the QC technique described.
- 1. The turbine stop valve is designed to withstand the SSE and maintain its integrity.
- m. The RCIC turbine is not included in the scope of standard codes. To assure that the turbine is fabricated to the standards commensurate with safety and performance requirements, General Electric has established specific design requirements for this component which are as follows:
- 1. All welding shall be qualiSed in accordance with Section IX, ASME Boiler and Pressure Vessel Code.
- 2. All pressure <ontaining castings and fabrications shall be hydrotested at 1.5 times the design pressure.
O Cisssification of Stru2ures, Components, and Systems - Amendment 33 3.2-57
23A6100 Rpv. 3 ABWR studardsafatyAnalysisneport O
- 3. All high-pressure castings shall be radiographed according to:
ASTM E-94 E-141 E-142 Maximum feasible volume E-446,186 or 280 Severity level 3
- 4. As-cast surfaces shall be magnetic-particle or liquid-penetrant tested according to AShiE Code,Section III, Paragraphs NB-2545, NG2545, or NB-2546, and NG2546.
- 5. Wheel and shaft forgings shall be ultrasonically tested according to ASTM A-388.
- 6. Butt welds in forgings shall be radiographed and magnetic panicle or liquid penetrant tested according to the ASME Boiler and Pressure Vessel Code,Section III Paragraph NB-2575, NG2575, NB-2545, NC-2545, NB-2546, NG2546 respective'y. Acceptance standards shall be in accordance with ASME Boiler and Pressure Vessel Code Section III, Paragraph NB-5320, NG5320, NB-5340, NG5340, NB-5350, NG5350, respectively.
- 7. Notification shall be made on major repairs and records maintained thereof.
- 8. Record system and traceability shall be according to ASME Section III, NCA-4000.
- 9. Quality control and identification shall be according to ASME Section III, NCA-4000.
- 10. Authorized inspection procedures shall conform to ASME Section III, N&5100 and NG5100.
- 11. Nondestructive examination personnel shall be qualified and certified according to ASME Section III, NS5500 and NG5500.
- n. All cast pressure-retaining parts of a size and configuration for which volumetric methods are effective are examined by radiographic methods by qualified personnel. Ultrasonic examination to equh21ent standards is used as an alternate to radiographic methods. Examination procedures and 3 2-58 Ctsssincation of Structures, Components, and Systems - Amendment 33
23A6100 n2v. 3
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ABWR stuhntsuretyAnalysis soport 0
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acceptance standards are at least equivalent to those defined in Paragraph 136.4, Nonboiler External Piping, ANSI B31.1.
- o. The following qualifications are met with respect to the certification requirements:
- l. The manufacturer of the turbine stop nives, turbine control valves, turbine bypass ulves, and main steam leads from turbine control ulve to turbine casing utilizes quality control procedures equivalent to those defined in GE Publication GEZ-4982A, General Electric Large Steam Turbine Generator Quality Control Program.
- 2. A certification obtained from the manufacturer of these valves and steam loads demonstrates that the quality control program as defined has been accomplished. >
The following requirements shall be met in addition to the Quality Group D requirements:
- 1. All longitudinal and circumferential butt weldjoints shall be radiographed (or ultrasonically tested to equivalent standards). Where
^
size or configuration does not pennit effective volumetric examination.
magnetic particle or liquid penetrate examination may be substituted.
Examination procedures and acceptance standards shall be at least equivalent to those specified as supplementary types of examinations, Paragraph 136.4 in ANSI B31.1.
i
- 2. All fillet and socket welds shall be examined by either magnetic particle or liquid penetrant methods. All structural attachment welds to pressure retaining materials shall be examined by either magnetic particle or l liquid penetrate methods. Examination procedures and acceptance standards shall be atleast equivalent to those specified as supplementary types of examinations, Paragraph 136.4 in ANSI B31.1 l f
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3.2-S9 Classification of Structures. Components, and Systems - Amendment 33 <
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- 3. All inspection records shall be maintained for the life of the plant. These )
records shallinclude data pertaining to qualification ofinspection personnel, examination procedures, and examination results.
- p. A quality assurance program meeting the guidance ofRegulatory Guide 1.143 will be applied dudng design and construction.
- q. Detailed seismic design criteria for the offgas system are provided in Subsection 11.3.4.8.
- r. See Subsection 3.2.5.3.
- s. The recirculation motor cooling system (RMCS) is classified Quality Group B and hfety Class 2 which is consistent with the requirements of 10CFR50.55a.
The RMCS, which is part of the reactor coolant pressure boundary (RCPB) meets 10CFR50.55a (c)(2). Postulated failure of the RMCS piping cannot cause a loss of reactor coolant in excess of normal makeup (CRD return or RCIC flow), and the RMCS is not an engineered safety feature. Thus, in the event of a postulated failure of the RMCS piping during normal operation, the reactor can be shutdown and cooled down in an orderly manner, and reactor coolant makeup can be provided by a normal make up system (e.g., CRD return or RCIC system). Thus, per 10CFR50.55a(c) (2), the RMCS need not be classified Quality Group A or Safety Class 1, however, for plant availability, the system is designed, fabricated and constructed in accordance with ASME Boiler and Pressure Vessel Code,Section III, Class 1 criteria as specified in Subsection 3.9.3.1.4 and Figure 5.4-4.
- t. A quality assurance program for the Fire Protection System meeting the guidance of Branch Technical Position CMEB 9.5-1 (NUREG.0800), is applied.
- u. Special seismic qualification and quality assurance requirements are applied.
- v. See Regulatory Guide 1.143, Paragraph C.5 for the offgas vault seismic requirements.
- w. The condensate storage tank will be designed, fabricated, and tested to meet the intent ofAPI Standard API 650. In addition, the specification for this tank O
3.2-60 CIsssification of Structures, Components, and Systems - Amendment 33
I 23A6100 R2v. 3 ABWR SrndardSafetyAn: lysis Rep:rt O
will require: (1) 100% surface examination of the side wall to bottomjoint and (2) 100% volumetric examination of the side wall weldjoints.
- x. The cranes and safety class 2 fuel senicing equipment are designed to hold up their loads and to maintain their positions over the units under conditions of SSE.
- y. All off-engine components are constructed to the extent possible to the AShfE Code,Section III, Class 3.
- z. Components associated with safety-related function (e.g., isolation) are safety-related.
aa. Structures which support or house safety-related mechanical or elecuical components are safety-related.
bb. All quality assurance requirements shall be applied to ensure that the design, consuuction and testing requirements are met.
cc. A quality assurance program, which meets or exceeds the guidance of Generic Letter 8546, is applied to all non-safety-related ATWS equipment.
dd. Deleted.
ee. Figure 3.2-2 depicts the classification requirements for the feedwater sptem.
At the interface between Seismic and non-Seismic CategoryI feedwater piping system, the Seismic Category I dynamic analyses will be extended to either the first anchor point in the non-seismic system or to sufficient distance in the non-seismic system so as not to degrade the validity of the Seismic Category I 7 analysis.
If. The equipment is not required to be classified as Seismic Category I. However, it is marked as Seismic Category I per PRA recommendation.
I gg. The Head Holding Pedestalis non-safety related and Seismic Category I. All '
other reactor vessel senicing equipment is non-seismic categorygr.nol hh. Light fixtures and bulbs are not seismically qualified but fixtures which receive Class 1E power are seismically supported (see Subsections 9.5.3.2.2.1 and 9.5.3.2.3.1 ) .
I I
i 3.2-61 Classification of Structures. Components, and Systems - Amendment 33
l 23A6100 Rev. 2 l ABWR StandardSafety Analysis Report O
Table 3.2-2 Minimum Design Requirements for an Assigned Safety Designation Minimum Design Requirements Safety Quality Group
- Seismic Categoryl Electrical Quality Designation' Classification ** Assurance" SC-1 A I -
B SC-2 B i - B SC-3 C i 1E B NNS t
- f
- For structural design requirements that are not covered here and in Table 3.2-3, see Section 3.8.
t Safety designations are defined in Subsections 3.2.3 and 3.2.5.
- Table 3.2 3 shows applicable codes and standards for components and structures in accordance with their quality group identified in Table 3.2-1.
Non-nuclear safety (NNS) related structures, systems and equipment that are not assigned a Quality Group in Table 3.2-1 are designed to requirements of applicable industry codes and standards (see Subsection 3.2.5.2).
Some NNS structures, systems, and components are optionally designed to Quality Group C or D requirements of Table 3.2-3, per Quality Group designation on Table 3.2-1. '
f Seismic Category I structures, systems and components meet design and analysis requirements of Subsection 3.7.
Some NNS structures, systems and components are optionally designed to Seismic Category I design criteria as noted on Table 3.2-1. Some safety-related components (e.g., Pipe whip restraints) have no safety-related function in the event of an SSE, and are not Seismic Category 1.
- Safety-related electrical equipment and instrumentation are designated SC-3 and are designed to meet IEEE Class 1E (as well as Seismic Category I) design requirements.
Some NNS electrical equipment and instrumentation i optionally designed to IEEE Class 1E requirements as noted on Table 3.2-1.
tt Safety-related structures, systems and components meet the quality assurance requirements of 10CFR50, Appendix B, as described in Chapter 17.
Some NNS structures, svst' ens, and components meet the QA requirements as noted on Table 3.2-1.
1 i
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3242 Classification of Structures. Components, and Systems - Amendment 32
23A6100 Rev. 3 ABWR standard saretyAnarysisaeport O tiBr or gioo,eB <cootiooeo, :
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22A6100 Rev. 3 ABWR standard sareryAntysis Report O
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23A6100 Rev. 3 ABWR standardsarety Analysis Report
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U-4.3.2.3 Reactivity Coefficients l J
Reactivity coefficients, the differential charv;es in reactivity produced by differential changes in core conditions, are useful in calculating stability and evaluating the '
response of the core to external disturbances. The base initial condition of the system and the postulated initiating event determine which of the several defined coeflicients are significant in evaluating the response of the reactor. The coeEfir%ts or mterest, relative to ABWR systems, are discussed here individually.
There are two primary reactivity coefficients that characterize the dynamic behavior of boiling water reactors: The Doppler reactivity coefficient and the moderator void reactivity coefficient. Also associated with the ABWRis a power reactivity coefficient and a temperature coeflicient. The power coefficient is a combination of the Doppler and void reacthity coeflicients in the power operating range, and the temperature coefficient is merely a combination of the Doppler and moderator temperature coefficients. Power and temperature coefficients are not specifically calculated for reload cores.
4.3.2.3.1 Doppler Reactivity Coefficient The Doppler coefficient is of prime importance in reactor safety. The Doppler coefficient is a measure of the reactivity change associated with an increase in the absorption of resonance-energy neutrons caused by a change in the temperature of the material in question. The Doppler reactivity coefficient provides instantaneous negative reactivity feedback to any rise in fuel temperature, on either a gross or local basis. The magnitude of the Doppler coefficient is inherent in the fuel design and does not vary ,
significantly among BWR designs. For most structural and moderator materials, resonance absorption is not significant, but in U-238 and Pu-240 an increase in temperature produces a comparativelylarge increase in the effective absorption cross-section. The resulting parasitic absorption of neutrons causes a significant loss in reactivity. In ABWR fuel,in which approximately 97% of the uranium in UO2 is U-238, the Doppler coefficient provides an immediate negative reactivity response that opposes increased fuel fission rate changes.
Although the reactivity change caused by the Doppler effect is small compared to other power-related reactivity changes during normal operation, it becomes very important during postulated rapid power excursions in which large fuel temperature changes occur. The most severe power excursions are those associated with rod drop accidents.
l A local Doppler feedback associated with a 1650 C to 2760 C temperature rise is available for terminating the initial excursion.
The Doppler coefficient is determined using the theory and methods described in Reference 4.3-2.
4.3-3 Nuclear Desipn - Amendment 32
23A6100 Rev. 3 ABWR standantsarety Anstrsis Report O
4.3.2.3.2 Moderator Void Coefficient l
The moderator void coefficient should be large enough to prevent power oscillation due to spatial xenon changes yet small enough that pressurization transients do not unduly limit plant operation. In addition, the void coefficient in the ABWR has the ability to flatten the radial power distribution and to provide ease of reactor control due to the void feedback mechanism The overall void coefficient is always negative over the complete operating range since the ABWR design is undermoderated. l A detailed discussion of the methods used to calculate void reactivity coefficients, their 1 accuracy and their application to plant transient analyses,is presented in Reference 4.3-2.
4.3.2.4 Control Requirements The General Electric ABWR control rod system is designed to provide adequate control of the maximum excess reacti ity anticipated during the plant operation. The shutdown capability is evaluated assuming a cold, xenon-free core.
4.3.2.4.1 Shutdown Reactivity The core must be capable of being made subcritical, with margin, in the most reactive condition throughout the operating cycle with the most reactive control rod fully withdrawn and all other rods fully inserted. The shutdown margin is determined by using the BWR simulator code (see Section 4.3.3) to calculate the core multiplication at selected exposure points with the strongest rod fully withdrawn.The shutdown mr.r;;in is ca!culm d based on the carryover of the minimum expected exposure at the end of the previous cycle. The core is assumed to be in the cold, xenon-free condition in order to ensure that the calculated values are conservative. Further discussion of the uncertainty of these calculations is given in Reference 4.3-3.
As exposure accumulates and burnable poison depletes in the lower exposure fuel bundles, an increase in core reactivity may occur. The nature of the increase depends on specifics of fuel loading and control state.
The cold keff is calculated with the strongest control rod out at various exposures through the cycle. A value R is def'med as the difference between the strongest rod out keg at BOC and the maximum calculated strongest rod out k eg at any exposure point.
The strongest rod out k eg at any exposure point in the cycle is equal to or less than:
where:
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.t.34 Nuclear Design - Amendment 33
23A6100 Rev. 3 ABWR standardsaretyAnatrsis neport Chapter 6 Q,o Table of Contents List of Tables.. . . . . . .. . .. . .. . . 6.0-iii List of Figures.. . . .. . .. . . . . . .... . . . .6.0-v 6 Engineered Safety Features . . .. . . . . .. .. . . . .6.0-1 6.0 General . . . . . . . . . . . . .. .. . . . . 6.0-1 6.1 Engineered Safety Feature Materials . . . . . . . . . . . . . ... . . . . . . .. 6.1-1 6.1.1 Metallic Materials . . . . . .. . ... .. . .. . ... 6.1-1 6.1.2 Organic Materials... . . . .. . . . . . . . . . . . . . . . .. . .. 6.1-3 6.1.3 COL License Information.. . . . . . . . . . . ... .. . 6.1-4 6.2 Containment Systems... . . .. . . . .. . . . . . . . . . .. ... 6.2-1 6.2.1 Containment Functional Design. . . . . . . . . . . . .... 6.2-1 6.2.2 Containment Heat Removal System.. . . . . . . . . .. . ... 6.2-41 6.2.3 Secondary Containment Functional Design.. . . . . . . .. ... . .... 6.2-45 6.2.4 Containment Isolation System.. . .... . . . . . . . . .. . .... 6.2-56 6.2.5 Combustible Gas Control in Containment.. . ... ...... ..... . . 6.2-70 6.2.6 Containment Leakage Testing . . . .. .. . ... . . . . . .. . . . . . . . . . .. . 6.2-91 6.2.7 COL License Information.. . . . . . . . . . . . . . . . . 6.2-99 6.2.8 References.. .. . . . . . . . .. .. . . . . . . . . . .... 6.2-100 k 6.3 Emergency Core Cooling Systems... .. .. . . . . . . . . . . . . .. . .. .. .. ... . 6.S l 6.3.1 Design Bases and Summary Description.. . . . . . . . . . . . . . . . . . . 6.S1 6.3.2 System Design . . .. . .. . . . . . . . . . . . . . .. . . 6.S5 6.3.3 ECCS Performance Evaluation. . . ... . . . . . . . . . . . . . . ... 6.3-13 6.3.4 Tests and Inspections. ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .6.3-21 6.3.5 Instrumentation Requirements. ... . . . . . ...... .. .... . . .. .. .. . . . . . . . .. . . 6. 3-2 4 63.6 COL License Irdormation . .................................6.S24 !
6.3.7 Reference . . .. . . . . . .. .. .. . . . . . . . . . . . . ...... 6.S24 6.4 IIabitability Systems . . . . . . . . . . . . . . . . . . . . ... 6.+1 6.4.1 Design Basis . .. . . . . .. . . . . . . . . . . . . ..... . 6.+2 6.4.2 System Design . .. . . . . . . . . . . . . . . . . . . . . . . . . . 6.4-4 6.4.3 System Operation Procedures . . . . . . . . .. . . . . . . . . . . . 6.4-7 6.4.4 Design Evaluations . .. . . . . . . . . . . . . . . . . .... .... ... ... 6.4-8 6.4.5 Testing and Inspection... . . ... . . . . . .... . . . . . . . . .6.4-10 6.4.6 Instrumentation Requirements. . . . . . . . . . . . . . . . . . ... .. .. . . . . 6.4-10 ,
6.4.7 COL License Information.. . . . . . . ...... .. .. .. . . 6.4-10 6.5 Fission Products Removal and Control Systems.. .. . . . . .. . .. . . .. . .. . .. .. . .._ . 6.5-1 6.5.1 Engineered Safety Features Filter Systems.. . . . . . . . . . . = 6.S1 6.5.2 Containment Spray Systems.. .. .. .. . . . . . . . . . . . . . . . . . .... 6.5-10 6.5.3 Fission Product Control Systems . .. . .. .. . . . . . .
.., .. ... ... . . 6.5-1 1 6.5.4 Deleted ... . .... . .. . . .. . . . . . .. ..... .. .. .. .. .. 6.512 s 6.5.5 COL Ucense Information.. . .. .. . . ... . .. .. .. . 6.512 6.5.6 References... . . . . . .. . . . . . . . . . .. 6.512 )
i Table of Contents - Amendenent 33 6.04 !
l
I 1
l 23A6100 Rev. 3 ABWR standardsaretyAnarr sisnepwi i
Table of Contents (Continued) 6.6 Presenice and Insenice Inspection and Testing of Class 2 and 3 Components and Piping.. . . .. . .. . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . .. . .. . .. . . . .. . 6. 6- 1 6.6.1 Class 2 and 3 Svstem Boundaries.. . ... .. . . . . . . . . . . . . . . ... 6.61 6.6.2 Accessibility.. . . . . . . . .. .. . .. . . . . . . . . . . . . . . . .6.64 6.6.3 Examination Categories and Methods.. . . . . . . . . . . . . . . . . . ...... 6.65 6.6.4 Inspection Intervals.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.6-7 6.6.5 Evaluation of Examination Results.. . . . . . . .... . .. . . . 6.68 6.6.6 System Pressure Tests. ... . . . . . . . . . . .. ... . . . . . . ..... . .. 6.68
- 6.6-9 6.6.7 Augmented Insenice Inspection.. . . . . . . . . . . . . .
6.6.8 Code Exemptions. . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . ... 6.6-9 6.6.9 COL License lnformation.. . .. . . . .. . . . . . . . . . . 6.6 10 6.7 High Pressure Nitrogen Gas Supply System.. . . . . . . . . . . . . . . . ....... ..... 6.7-1 6.7.1 Funcdons.. ... .. . .. ...... .. . . . . . . . . . . .. 6.7-1 6.7.2 System Description. . . . . . . . . . . . . . . . . . . 6.7-1 6.7.3 System Evaluation. . . . . . . . . . ..... .... 6.7-2 6.7.4 Inspection and Tesdng Requirements.... ... . . .. . . . ........ .. ... 6.7-3 6.7.5 Instrumentation Requirements... . . . . . . . . . . . . . . . . . . . . . . . . 6.7-5 6.7.6 Analysis and Testing of ADS Accumulator Capacity. .. . . . . . . . . ... . 6.7-4 Appendicies 6A Regulatory Guide 1.52 Section C, Compliance Assessment. . . . . . . . . . . . . .. . ... . . 6A-1 6B SRP 6.5.1, Table 6.5.1-1 Compliance Assessment.. . . . . . . . .. . . . . . . .. ... . .. 6B-1 6C Containment Debris Protecdon for ECCS Strainers... . . . . . . . . . . . . ... 6C-1 6D HPCF Analysis Outlines ... . . . . . . . .. .. . . . . . . . . . . . . . . . . . . . . . . .. 6D.1 i
1 l
1 6.04i Table of Contents- Amendment 33 l
1
23AS100 Rev. 3 ,
ABWR standard sareryAnalysis aeron q Chapter 6 U List of Tables Table 6.1-1 Engineered Safety Features Component Materials.. ... . . . . . . .. .... 6.1-5 Table 6.2-1 Containment Parameters . . . . . . . . . . . .... .. 6.2-101 Table 6.2 2 Containment Design Parameters.. . . . . . . . . . . . . . . . .6.2-102 Table 6.2-2a Engineered Safety Systems Information for Containment Response Analyses., . . . . ... . . . . . . . .. . . . . . .6.2 103 Table 6.2-2b Net Positive Suction Head (NPSH) Available to RHR Pumps.. . . . . . . . . . .. 6.2-105 Table 6.2-2c Net Posidve Suction Head (NPSH) Available to HPCF Pumps... 6.2 106 Table 6.2-2d Secondary Containment Design and Performance Data.... . .... .. . .6.2107 Table 6.2-3 Subcompartment Nodal Description .. . . . ... ... . . . . . . . . . . . . . .6.2 110 Table 6.2-1 Subcompartment Vent Path Description.. . . . . . . . . . . . . . . . . . . 6.2-112 Table 6.2-ta Flow Loss Factor . .. .. . . ... . . . . . . . . . . . . . . . . 6.2-114 Table 6.2-4b Mass and Energy Release Rate.. . . . .. . . . . . . . . . . ... .. .. ... . 6.2-116 k Table 6.2-5 Reactor Coolant Pressure Boundary (RCPB) Influent Lines Penetrating l Dr}well.. . .. . .. . . . . . . . . . .. . .. . . .. . . . 6.2-121 Table 6.2-6 Reactor Coolant Pressure Boundary (RCPB) Effluent Lines Penetrating Dr>well.. . .. . . . . . . . . . . . . . . . . . . . .. .. ... . .. .. .. . .. 6.2-121 Table 6.2-7 Containment Isolation Valve Information.. . .. . . . . . . . . . . . . .. . . 6.2-122 Table 6.2-8 Primary Containment Penetration List * .. ... . . . . . . . . ... 6.2-170 Table 6.2-9 Secondary Containment Penetration List.. .. . . . . . . . . ... ... . ... . 6.2-180 Table 6.2-10 Potential Bypass Leakage Paths . .. . . .. .. . . . . . . . . . . . . ... .. 6.2 182 Table 6.Sl Significant Input Variables Used in the Lossef-Coolant Accident Analysis. . 6.S25 Table 6.S2 Operational Sequence of Emergency Core Cooling System Maximum Core Flooder Line Break..... .. . . . . . .. .. .... .. .. . . ... . . ..... . ... . 6.S2 7 Table 6.S3 Single Failure Evaluation . . . . . . . . . . .. . . . . . . . . . . ..... 6.S28 Table 6.14 Summary of Results of LOCA Analysis.. . . . . . . . . . . . . . . 6.S29 Table 6.S5 Rey to Figures . . . .... . . . . . .. .. . 6.S30 bd Table 6.% Plant Variables with Nominal and Sensitivity Study Values . . . . . . . . . 6.S31 Table 6.S7 MAPLHGRVersus Exposure. .. . . . . . ... . . . . . . .. . - 6.132 List of Tabits- Amendment 33 6 0-in
23A6100 Rev. 3 ABWR standardsareryAnalysis Report List of Tables (Continued)
Table 6.S8 Design Parameters for HPCF System Components... . . . . . . . . . 6.533 Table 6.59 Design Parameters for RIIR System Componems... . . . .. . . . . . . . 6.S$5 Table 6.110 Single Failure Evaluation With One HPCF Subsystem Out of Senice - = 6.S38 Table 6.111 Single Failure Evaluation With One RHR/LPFL Subsystem Out of Senice. 6.S39 Table 6.S12 Single Failure Evaluation With ECCS Division A Out of Senice.. . .. . ... 6.S40 Table 6.513 Single Failure Evaluation With ECCS Division B Out of Senice . 6.141 Table 6.S14 Single Failure Evaluation With ECCS Division C Out of Senice.. .. .. 6.S42 Table 6.S15 Single Failure Evaluation With Two ADS Valves Out of Senice . . . . . .. 6.S43 Table 6.4-1 Identification of Failure /Effect in the Control Room Habitability Area HVAC Sptem. . . . . . . . .. . . . . . . .. . . . . . . 6.4-12 Table 6.4-2 Control Room Habitability Area HVAC System Failure Analysis . 6.4-13 Table 6.5-1 Summary of Major Standby Gas Treatment System Component = ... 6.5-13 I Table 6.5-2 Source Terms Used for SGTS Charcoal Adsorber Design... . . . . ... . 6.515 Table 6.61 Examination Categories and Methods... . . . . . . . . . . . . . . . . . . . . . . . . . .. . 6.611 Table 6.6-1 Examination Categories and Methods (Continued) . . . . . . . . . . . . . . .. . . 6.&25 Table 6.41 Examination Categories and Methods (Continued) .. . . . .6.6 31 1
l l
Table 6.41 Examination Categories and Methods (Continued) .. .. . . . . ... . 6.G34 Table 6.41 Examination Categories and Methods (Continued) = . . . . . . .. . 6.643 Examination Categories and Methods (Continued) .. .. .. .. 6.G44 l Table 6.Gl Table 6.7-1 Nitrogen Gas Demand . . . . . .. .. .. . ... . . . . . . . . . . . . . , .6.7-5 O
S.0 iv List of Tables - Amendment 33
i 23A6100 Rev. 3 ABWR standardsaretyAnstysis aeput .
Chapter 6 List of Figures
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Figure 6.2-1 A Break in a Feedwater Line.. . ... ... . . . . . . . . . . . . . . . . . . 6.2-190 Figure 6.2-2 Feedwater Line Break-RPV Side Break Area . . . . . . . . ... . . 6.2-191 :
Figure 6.2-3 Feedwater Line Break Flow-Feedwater System Side of Break... ... . . . . . 6.2-192 ,
I Figure 6.2-4 Feedwater Line Break Flow Enthalpy-Feedwater System Side of Break... 6.2-193 figure 6.2-5 Lower Drywell Air Transfer Percentage for Model Assumption Versus l Actual Case.. . .. .. . ... . . . . ... . . . .. . . .. 6.2-194 t t
Figure 6.2-6 Pressure Responst of the Primary Containment for Feedwater Line Break 6.2-195 Figure 6.2-7 Temperature Response of the Primary Containment for Feedwater Line 3 Break. . .. . . . . . . ... . .. . . . . . . . . 6.2-196 Figure 6.2-8 Temperature Time History After a Feedwater Line Break.. .. . .. . 6.2-197 j Figure 6.2-9 ABWR Main Steamlines with a Break.. . . . .. ... . . . . . 6.2-198 Figure 6.2-10 MSLB Area as a Function of Time . . . . . . ... . . .. 6.2-199 [
Figure 6.2-11 Feedwater Specific Enthalpy as a Function ofIntegrated Feedwater Flow j Mass.. . . . . . . . . . . . . . . .. 6.2-200
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Figure 6.2-12 Pressure Time History for MSLB with Two-Phase Blowdown Starting l When the RPV Collapsed Level Reaches the Main Steam Nozzle.. . ... ... 6.2-201 Figure 6.2-13 Temperature Time History for MSLB with Two-Phase Blowdown Stardng When the RPV Collapsed Level Reaches the Main Steam Nozzle........... 6.2-202 l
Figure 6.2-14 Pressure Time History for MSLB with Two-Phase Blowdown Starting at One Second ... ... .. . . . . .. . ... . .... 6.2-203 Figure 6.2-15 Temperature Time History for MSLB with Two-Phase Blowdown Starting at One Second. . . . . .. .. . . . . . . ...... .. . 6.2-204 Figure 6.2-16 General Pressure Trends in the Containment During a Post-LOCA Depressurization Transient... . . . . . .. .... .. . . . . ... ... ... . 6.2-205 l
Figure 6.2-17 Differential Pressures in Wetwell and Drywell Reladve to Reactor 2 . 6.2-205 l Building for Vacuum Breaker Size of.771 m ...... . . . . . ..
Figure 6.2-18 Differential Pressures in Wetwell and Drywell Relative to Reactor Building 2 6.2-207 l with Wete11 Spray for Vacuum Breaker Size of.771 m . . . . . .
Figure 6.2-19 Temperan re and Pressure Time Histories in the Containment During Stuck Open Relief Valve Transient . . .. .. .. . . . . .. ... .. .... .. 6.2-208 Deleted.. . ,6.2-209 Figure 6.2-20 . . . . . . . . . . .. .
List of Figures - Amendment 33 6.0 v i
1 1
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2.146 iM Rev. 3 l standard sarery Analysis Report ABWR \
List of Figures (Continued) h !
Figure 6.2-21 Deleted.. . . .. . . . .. . . . . . . . .. . 6.2-210 ;
l Figure 6.2-22 Break Flow Rate and Specific Enthalpy for the Feedwater Une Break Flow '
Coming from the Feedwater System Side.. ... .. .6.2-211 l
Figure 6.2-23 Break Flow Rate and Specific Enthalpy for the Feedwater Line Break Flow Coming from the RPV Side . . . . . ... .. .. . . . . . . . . . . . . .6.2-212 l
Figure 6.2-24 Break Flow Rate and Specific Enthalpy for the Main Steamline Break with Two-Phase Blowdown Starting When the Collapsed Water Level Reaches the Steam Nozzle. .. .. .. . .. . . . . . . . .. . . . . . . 6.2-213 l
Figure 6.2-25 Break Flow Rate and Specific Enthalpy for the Main Steamline Break with Two-Phase Blowdown Starting at One Second... . .. . . . . . .. 6.2-214 Figure 6.2-26 ABWR Containment Boundary Nomenclature. . .. . . . .6.2-215 Three Basic Types of Leakage Paths. . . .6.2-21,6 Figure 6.2-27 . . . . . . ..
Figure 6.2-28 Containment Boundaries in the Reactor Building-Plan Section A-A (0 -180 ) .. .. .. .. . . . . . . . . . . . . . . . 6.2-217 l
Figure 6.2-29 Containment Boundaries in the Reactor Building-Plan Section B-B (90 -270 ) . . .. . . . . . . . . . . . . . . . . . . . . . . . .6.2-217 l
Figure 6.2-30 Containment Boundaries in the Reactor Building-Plan at Elevation -8200 mm . . . . . . . . . .. . . . . . . . . . . . . .6.2-217 l
Figure 6.2-31 Containment Boundaries in the Reactor Building-Plan at Elevation -1700 mm . . . . . . . . . . . .. . . . . . .. 6.2-217 l
Figure 6.2-32 Containment Boundaries in the Reactor Building-Plan at Elevation 4800/8500 mm . . . . .. . .. . . .. .. 6.2-217 l
Figure 6.2-33 Containment Boundaries in the Reactor Building-Plan at Elevation 12300 mm.. . . ... .. . . . . . . . . .. 6.2 217
[
Figure 6.2-34 Containment Boundaries in the Reactor Building-Plan j at Elevation 18100 mm.. . . . . . . . .. . . . . . . . . . . . . . . . . ... . 6.2-217 Figure 6.2-35 Containment Boundaries in the Reactor Building-Plan at Elevation 23500 mm.. . ... ...
= 6.2-217 l
Figure 6.2-36 Containment Boundaries in the Reactor Building-Plan at Elevation 31700 mm.. .. . . . . . . . . . . . . . . . . . . . . . . . . - . 6.2-217 Figure 6.2-37a Secondag Containment Schematic Flow Diagram (ECCS/RCIC) . . ... 6.2-218 Figure 6.2-37b Seconday Containment Schematic Flow Diagram
.6.2-219 l (Main Steam /Feedwater) . . .. . .. . . . . . . . . . . . .
6 0-vi List of figures- Amendment D
23A6100 Rev. 3 standardsarery Analysis aeport ABWR
(
\
List of Figures (Continued)
Figure 6.2-37c Secondary Containment Schematic Flow Diagram (CUW) .. . . . . .6.2-220 Figure 6.2-37d Secondary Containment Schematic Flow Diagram (CUW) . . . .. .6.2-221 Figure 6.2-37e Secondary Containment Schematic Flow Diagram (CUW) . . . .6.2-222 Figure 6.2-38 Plant Requirements. Group Classificadon and Containment Isolation Diagram (Sheets 1 -2). . . . .. . .. .6.2-223 Figure 6.2-39 Atmospheric Control System P&ID (Sheets 1 - 3)... . . . . . . . . . .. . . 6.2-223 Figure 6.2-40 Flammability Control System P&ID (Sheets 1 - 2) . .. .. . 6.2-223 Figure 6.2-41 Hydrogen and Oxygen Concentrations in Containment After Design Basis LOCA. . . . .. . . . . .. . 6.2-224 Figure 6.2-42 Allowable Steam Bypass Leakage Capacity... . . . . . . . . .. . 6.2-22,5 Figure 6.2-43 Typical Pressure Fluctuation Due to CO (See Figure 3B-22)... .. .. . 6.2-226 Figure 6.2-44 Allowable Steam Bypass Leakage Capacity. . . .. . .. . . . . .. . 6.2-227 Figure 6.2-45 Typical Pressure Fluctuation to CH (See Figure 3B-24) Quencher Bubble Pressure Time History.. . ... .. . . . . . . . . 6.2-228 Figure 6.Sl High Pressure Core Flooder System PFD (Sheets 1-2) . . . . . . . . . . . . . . . . 6.3-44 Figure 6.S2 Deleted (See Figure 5.4-9). . . . . .. . . . . . . . . . .... 6.144 Figure 6.S3 Deleted (See Figure 5.4-11). .. . . .. . . . . . . . . . . . . . . . . .. ... .. ... 6.S44 Figure 6.S4 Pressure Versus High Pressure Core Flooder Flow (Per System) Used in LOCA Analysis. . .. .. .. . . . . . . . . . . . . . . . .. . . . . . .... . 6.S45 Figure 6.15 Pressure Versus Reactor Core Isolation Cooling How Used in LOCA Analysis.. . . . . ... . .. . . . . . . . . . . . ... . . -6.S46 Figure 6.16 Pressure Versus Low Pressure Flooder Flow (Per System) Used in LOCA Analysis... .. . . .. . .. ... .. .. . . . . . . . . . . . . . 6.S47 Figure 6.S7 High Pressure Core Flooder System P&lD (Sheets 1-2) . .. . . . . .. 6.S48 Figure 6.18 Deleted (Sec Figure 5.4-8). . . . . . . . . ... .. . . . . . . . . . .. 6.S4 8 Figure 6.19 Deleted (See Figure 5.4-10). . . . . . .. . . . .. . .... . .... 6.S48
( Figure 6.S10 Minimum Water Level Outside Shroud Versus Break Area. . .. . 6.3-49 Figure 6.Sl1 Normalized Core Power Versus Time for Lossof-Coolant Accident Analysis... . . . . . . . . . . . . . . . . . . ... . . . 6.150
^
6.0 vii List of Figures- Amendment 33
23A6100 Rev. 3 ABWR standardsarery Analysis Report O
List of Figures (Continued)
Figure 6.S12 Normalized Core Flow Following a hiain Steamline Break Inside Containment, HPCF Diesel Generator Failure . .. . . . . . .. . . . .. 6.S51 Figure 6.S13 hiinimum Critical Power Ratio Following a hiain Steamline Break Inside Containment, HPCF Diesel Generator Failure.. ... .. . . . . .... 6.551 Figure 6.S14 Water Irvel in Fuel Channels Following a hiain Steamline Break inside Containment, HPCF Diesel Generator Failure . . . . . ..... . ... ... 6.S52 Figure 6.S15 Water level Inside Shroud Following a hiain Steamline Break Inside Containment, HPCF Diesel Generator Failure . . . . . . . . . . . . . ... 6.552 Figure 6.516 Water 12 vel Outside Shroud Following a hiain Steamline Break Inside Containrnent, HPCF Diesel Generator Failure.. . . . . . . . .. . ..-.. . 6.S53 Figure 6.S17 Vessel Pressure Following a hiain Steamline Break Inside Containment, HPCF Diesel Generator Failure.... . .. . . . . . . . . . . . . . . . . . . ... 6.S53 Figurc 6.518 Flow Out ofVessel Following a hiain Steamline Break Inside Containment, .
HPCF Diesel Generator Failure = . . . .. . . . .. .. ...... . 6.S54 Figure 6.319 Flow into Vessel Following a hiain Steamline Break Inside Containment.
HPCF Diesel Generator Failure.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 6.S54 Figure 6.320 Peak Cladding Temperature Following a hiain Steamline Break Inside Containment, HPCF Diesel Generator Failure .. . . . ... 6.S55 Figure 6.121 Normalized Core Flow Following a Feedwater Line Break, HPCF Diesel Generator Failure. . .. . . . . . . . . . . . . . . . . . . . . . . 6.'S55 Figure 6.322 hiinimum Critical Power Ratio Following a Feedwater Line Break, HPCF Diesel Generator Failure.. .. . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . 6.556 Figure 6.S23 Water Izvel in Fuel Channels Following a Feedwater une Break, HPCF Diesel Generator Failure . . . . . . . . . . . . . .. . . . . . . . . . . 6.3-56 Figure 6.S24 Water 12 vel Inside Shroud Following a Feedwater Line Break, HPCF Diesel Generator Failure.... . . . . . . . . . . . . . . . . . . . . . . . .. . . .. 6.S57 Figure 6.S25 Water Izvel Outside Shroud Following a Feedwater Line Break, HPCF l Diesel Generator Failure... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.557 j Figure 6.526 Vessel Pressure Following a Feedwater Line Break, HPCF Diesel Generator Failure. . . . . . . . . . . . . . . . . . . . .. . . . . . . ... .. . ... . ... 6.S58 Figure 6.S27 How Out of Vessel Following a Feedwater Line Break, HPCF Diesel '
Generator Failure.. . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . .. 6.3-58 Figure 6.528 Flow Into Vessel Following a Feedwater Line Break, HPCF Diesel Generator Failure... . . . . . . . . . . . . . . .. . ... 6.S59 6.0-viii List of Figures - Amendment 33
23A6100 Rev. 3 i ABWR standard saretyAnarysis Report ;
k '
List of Figures (Continued) ,
Figure 6.129 Peak Cladding Temperature Following a Feedwater Line Break, HPCF Diesel Generator Failure.. . .. . . . . . . . . . .... . 6.S59 Figure 6.S30 Water Level in Fuel Channels Following an RHR Shutdown Suction Line
. 6.%0 Break, HPCF Diesel Generator Failure.. . . . . . . . . . . . . .
Figure 6.S$1 Water Ixvel Inside Shroud following an RHR Shutdown Sucdon Une ;
Break, HPCF Diesel Generator Failure.. . . . . . . . . . . . . . . . . . . . . . . . .... . . 6.360 Figure 6.132 Water Ixvel Outside Shroud Following an RHR Shutdown Suction Line Break, HPCF Diesel Generator Failure.. .. .... .. . . .. . ... ... 6.%1 Figure 6.S33 Vessel Pressure Following an RHR Shutdown Suction Line Break, HPCF ;
Diesel Generator Failure.. . . . ... ... .. ... 6. % 1 ,
Figure 6.S34 Mow Out of Vessel Following an RHR Shutdown Suction Line Break, HPCF Diesel Gencrator Failure.. . .... .. . . .. . . .
.... 6.%_2 ,
Figure 6.135 Flow Into Vessel Following an RHR Shutdown Sucdon Line Break, HPCF ,
Diesel Generator Failure..... . .. .. . . . . . . . . . . . . . . . . . . ... .6.562 I(
Figure 6.536 Peak CladdingTemperature Following an RHR Shutdown Sucdon Une, HPCF Diesel Generator Failure.. . . . . . . . . . . . . . . . . .. 6.%3 Figure 6.S37 Water Level in Fuel Channels Following an RHR/LPFL Injection Line ,
Break, HPCF Diesel Generator Failure.. ... .. ... . . . . . ...... ... .. ... 6.%3 Figure 6.538 Water level Inside Shroud Following an RHR/LPFL Injecdon Line Break, HPCF Diesel Generator Failure. . . ..... ..... ... . . . . . .. . 6.S 64 Figure 6.S39 Water Level Outside Shroud Following an RHR/LPFL Injection Line Break, HPCF Diesel Generator Failure.. .... . . . . . . . . . ... . . 6.%4 Figure 6.S40 Vessel Pressure Following an RHR/LPFL Injection Line Break, HPCF Diesel Generator Failure .. ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.%5 Figure 6.S41 Flow Out of Vessel Following an RHR/LPFL Injection Une Break, HPCF Diesel Generator Failure.. .... . ....... . . . . ............... . 6.%5 Figure 6.S42 Mow into Vessel Following an RHR/LPFL Injection Line Break, HPCF l Diesel Generator Failure . ..... .. . ... . . . .. . .. . .... . .... 6.%6 l Figure 6.S43 Peak Cladding Temperature Following an RHR/LPFL Injection Line .
6.S66 !
Break, HPCF Diesel Generator Failure.. . . . . . . . . . . . . ..
Figure 6.S44 Normalized Core Flow Following a Core Flooder Line Break, HPCF Diesel Generator Failure... . . . . . . . . . . 6.%7 ]
l Figure 6.545 Minimum Critical Power Ratio Following a Core Flooder Line Break, HPCF Diesel Generator Failure......... .... .... . .. . .... . . 6.%7 6.0-ix List cf Figures - Amendment 33 l l
23A6100 Rev. 3 ABWR standardsaretyAnalysis Report O
List of Figures (Continued)
Figure 6.S46 Water Izvel in Fuel Channels Following a Core Flooder Line Break, HPCF Diesel Generator Failure.. . . . . . . . . . . . . . ... 6.%8 Figure 6.S47 Water Ievel Inside Shroud Following a Core Flooder Line Break, HPCF Diesel Generator Failure.. . . . . . . . . . . . . . . . . . . . . .. ... .. ..... 6.%8 Figure 6.S48 Water Ixvel Outside Shroud Following a Core Flooder Line Break, IIPCF Diesel Generator Failure.. . .. . . .
- 6.%9 Figure 6.S49 Vessel Pressure Following a Core Flooder Une Break, HPCF Diesel Generator Failure.. . . . . . . . . . . . .. 6.%9 Figure 6.%0 Flow Out of Vessel Following a Core Flooder Line Break, HPCF Diesel Generator Failure.. . . . . . . . . . . . . . . . ... . . ... 6.3-70 Figure 6.S51 Flow into Vessel Following a Core Flooder Line Break, HPCF Diesel Generator Failure.. .... . . . . .. .. . . . . 6.3-70 Figure 6.152 Peak Cladding Temperature Following a Core Flooder une Break, .
HPCF Diesel Generator Failure.. ... . - . . . . . . . . . .. . 6.3-71 Figure 6.S53 Water Level in Fuel Channels Following a Bottom Drain Line Break, HPCF Diesel Generator Failure.. .. . . . . .. . . . .
- 6.5 71 Figure 6.S54 Water Level Inside Shroud Following a Bottom Drain Line Break, IIPCF Diesel Generator Failure... . . . . . . . . . . . .... ... .6.3-72 Figure 6.M5 Water Level Outside Shroud Following a Bottom Drain Line Break, HPCF Diesel Generator Failure.... . . . . . .. .6.3-72 Figure 6.M6 Vessel Pressure Following A Bottom Drain Line Break, HPCF Diesel Generator Failure.. . . . . . . . . . . . . . . . . . . . . . ... ......... 6.S73 Figure 6.S57 Flow Out of Vessel Following A Bottom Drain Line Break, HPCF Diesel Generator Failure... . .. . . . . . .. . . . . . . . . . . . . . . . . . . . . . . 6.S73 Figure 6.S58 Flow Into Vessel Following A Bottom Drain Line Break, HPCF Diesel Generator Failure.. ...... .. . . . 6.S 74 Figure 6.S59 Peak Cladding Temperature Following a Bottom Drain Line Break, HPCF Diesel Generator Failure.... . . . . . . .. 6.S74 Figure 6.%0 Water Izvel in Fuel Channels Following a Main Steamline Break Outside Containment HPCF Diesel Generator Failure . .. . .. 6.3-75 Figure 6.%1 Water Ixvel Inside Shroud Following a Main Steamline Break Outside Containment, HPCF Diesel Generator Failure - . . .... .. 6.S75 Figure 6.%2 Water Outside Shroud Following a Main Steamline Break Outside Containment, HPCF Diesel Generator Failure. . . . . . . . . . . . . . . . . . . . . . 6.576 6.0-x List of figures- Amendment 33
I i
i 23A6100 Rev,3 i ABWR standardsaferyAnalysisRepon O
List of Figures (Continued) !
Figure 6.%3 Vessel Pressure Following a Main Steamline Break Outside Containment, HPCF Diesel Generator Failure. . . . . . . . . . . ... ... 6.576 .
Figure 6.%4 Flow Out of Vessel Following a Main Steamline Break Outside i Containment, HPCF Diesel Generator Failure . . . . 6.S77 i Figure 6.%5 How into Vessel Following a Main Steamline Break Outside Containment, HPCF Diesel Generator Failure .... ... .. .. ...... .... ... ..... ..... .. 6.S77 ;
Figure 6.%6 Peak Cladding Temperature Following a Main Steamline Break Outside Containment. HPCF Diesel Generator Failure.. ... ... .. ....... ......... 6.578 Figure 6.S67 Normalized Core Flow Following a Main Steamline Break Outside Containment, HPCF Diesel Generator Failure (Based on Bounding Values) . . .. . . . . . . . .. . .. . . . . . . .. . 6.S78 i Figure 6.%8 Minimum Critical Power Ratio Following a Main Steamline Break -
Outside Containment, HPCF Diesel Generator Failure (Based on Bounding Values).. . . . .. .. .. .. . ...... . .. .. . ..... 6.579 Figure 6.%9 Water I.evel in Fuel Channels Following a Main Steamline Break i Outside Containment, HPCF Diesel Generator Failure (Based on ,
Bounding Values).. ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. ... . 6.S79 ;
Figure 6.S70 Water Level Inside Shroud Following a Main Steamline Break Outside Containment, HPCF Diesel Generator Failure (Based on 95%
Probability Value)... ...... .. . .. . . . . . . . . . . . . . . . . . . . . . . . . . . 6.S80 Figure 6.571 Water Level Outside Shroud Following a Main Steamline Break Outside Containment. HPCF Diesel Generator Failure (Based on Bounding Values)... .. ... . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.S80 Figure 6.S72 Vessel Pressure Following a Main Steamline Break Outside Containment, ,
HPCF Diesel Generator Failure (Based on Bounding Values) . ... ......... . .. 6.381 Figure 6.S73 Flow Out of Vessel Following a Main Steamline Break Outside Containment.
HPCF Diesel Generator Failure (Based on Bounding Values) .. ...... ..... .. 6.341 Figure 6.S74 Flow into Vessel Following a Main Steamline Break Outside Containment, HPCF Diesel Generator Failure (Based on Bounding Values) = .... ... 6.M2 Figure 6.S75 Peak Cladding Temperature Following a Main Steamline Break Outside Containment. HPCF Diesel Generator Failure (Based on BoundingValues)... . . . . . . . . . . . . . . . . . . . . - . . . . . . . . 6.S82 O
Figure 6.S76 Peak Cladding Temperature Versus Break Area with 1 RHR/LPFL + 5 ADS Available.... ... . . . .............. 6.S83 Figure 6.S77 Vessel Water Level Inside Shroud Versus Time Bottom Head Maximum Drainline Break 1 RIIR/LPFL + 5 ADS Available..... . 6.S84 List of Figures- Amendment 33 6.0 xi -
I
. I
\
23A6100 Rev. 3 ABWR standardsareryAnalysis neport List of Figures (Continued)
]
Figure 6.S78 Vessel Pressure Versus Time Bottom Head Maximum Drainline Break 1 RHR/LPFL + 5 ADS Available.. . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . 6.185 Figure 6.579 Peak Cladding Temperature Versus Time Bottom Head Maximum Drainline Break 1 RHR/LPFL + 5 ADS Available. . . . . . . . .. . 6.S86 Figure 6.41 Plant Layout.. . .. . . . . . . . .. . . . . . . . . . . . . . . . .. .. .. 6.4-14 Figure 6.51 Standby Gas Treatment System P&ID (Sheets 1-3). .. ... . .... ... 6.5-16 Figure 6.5-2 Secondary Containment Pressure Transient After Design Basis LOCA. . . 6.5-17 Figure 6.7-1 High Pressure Nitrogen Gas Supply System P&lD.. .. . . . . . . .... 6.7-6 Figure 6D-1 Injection Flow. . . . . . . . . . ... . . . . . . . . .. . .. . 6D.3 O
O1 l
6.0-xii List of Figures - Amendment .D
33A6100 Rev. 3 standard saretyAnalysisneport ABWR n
Because of the methods described above (coolant storage provisions, insulation materials requirements, and the like), as well as the fact that the containment has no significant stored quantities of acidic or basic materials, the post-LOCA aqueous phase pH in all areas of containment will have a flat time history. In other words, the liquid coolant will remain at its design basis pH throughout the event. $
6.1.2 Organic Materials 6.1.2.1 Protective Coatings The use of organic protective coatings within the containment has been kept to a minimum. The major use of such coatings is on the carbon steel containment liner, internal steel structures, and equipment inside the drywell and wetwell. ,
The epoxy coatings are specified to meet the requirements of Regulatory Guide 1.54 l
and are qualified using the standard ANSI tests, including ANSI N101.2. However, because of the impracticability of using these special coatings on all equipment, certain exemptions (e.g., electronic /c!cctrical trim, covers, face plates and valve handles) are allowed. The exemptions are restricted to small-size equipment where, in case of a LOCA, the paint debris is not a safety hazard. Other than these minor exemptions, all coatings within the containment are qualified to Regulatory Guide 1.54. See Subsection 6.1.3.1 for COL license information.
6.1.2.2 Other Organic Materials Materials used in or on the ESF equipment have been reviewed and evaluated in respect to radiolytic and pyrolytic decomposition and attendant effects on safe operation of the system. For example, fluorocarbon plastic (Teflon) is not permitted in emironments 4
that attain temperatures greater than 148.8 C, or radiation exposures above 10 rads.
The 10 reactor internal pump motors each contain less than 10 pounds of polyacryiic and polyethylene motor winding insulation. This material has a design life of 20 years 7
in the emironment ofless than 6 x 10 rads at 60 C maximum.
Other organic materials in the containment are qualified to emironmental condidons in the containment. See Subsection 6.1.3.1 for COL license information.
6.1.2.3 Safety Analysis For each application the materials have been specified to withstand an appropriate radiation dose for their design life, without suffering any significant radiation-induced damage. The specified integrated radiation doses are consistent with those listed in Section 3.11. The various suppliers have indicated their compliance with these requirements.
In addition, since the containment post-accident emironment consists of hot water, air and steam, no significant chemical degradation of these materials is expected because
- 6. r-3 Engineered Safety Feature Materials - Amendment 33
- . l 1
23k61C0 Rev. 3 l standard satctyAnalysis acport I ABWR O l 1
of strict applications ofinspection and testing. No significant amount of solid debris is expected to be generated from these materials.
6.1.3 COL License information 6.1.3.1 Protective Coatings and Organic Materials (1) Indicate the total amount of protective coadngs and organic materials used inside the containment that do not meet the requirements of ANSI N101.2 and Regulatory Guide 1.54.
(2) Evaluate the generation rate as a function of time of combustible gases that can be formed from these unqualified organic materials under DBA conditions that can reach the containment sump.
(3) Provide the technical basis and assumptions used for this evaluation (Subsection 6.1.2.1 and 6.1.2.2). ,
O O
Engineered Safety Feature Materials - Amendment 33 6.14
l l
1 23A6MO Rev. 3 ABWR st=d:rd S:letyAn: lysis R: port l
l 6.2.3.3.1.2.1 Reactor Core isolation Cooling (RCIC) Compartment The RCIC compartment is located in the secondary containment at El-8200 mm,in the 0-90 quadrant of the R/B. The design basis break for the RCIC compartment is detennined to be the single-ended break of the 150A nominal pipe size steam supply line to the RCIC turbine. This line is a high-energy line out to the normally closed isolation valve inside the RCIC compartment. It supplies high<nergy steam to the RCIC turbine in the event of reactor vessel isolation. In the event of a postulated design basis j high-energy line break (HELB), the steam / air mixture from that compartment is directed into adjoining compartments and is eventually purged into the steam tunnel.
6.2.3.3.1.2.2 Reactor Water Cleanup (CUW) Equipment Rooms and Pipe Spaces l The RWCU equipment (pump, heat exchanger, filter /demineralizer, valves) and pipe spaces are located in the 0 - 270 degree quadrant of the reactor building, with floor elevations ranging from elevation -8000 mm to elevation 12300 mm.The design basis pipe break for the CUW System compartment network is determined to be a 200 mm double-ended bren of the cleam p water suction line from the RPV. This high energy piping, which connects the CUW equipment, originates at the reactor pressure vessel. ,
After being routed through the CUW System, this line is directed back to the RPV through special pipe spaces and the steam tunnel. In the event of a postulated design l basis high energy line break in a compartment, the steam / air mixture from that ,
compartment is directed into adjoining compartments and eventually purged into the l turbine building through the steam tunnel.
6.2.3.3.1.2.3 Main Steam Tunnel The Reactor Building main steam tunnel is located between the primary containment vessel and the Turbine Building at elevation 12300 mm and 0* azimuthal position. The ,
DBA for the steam tunnelis the double-ended break of one of the 700 A main steamlines. These lines originate at the RPV and are routed through the main steam l tunnel to the Turbine Building. In the event of a postulated design basis HELB, the ,
l steam / air mixture from the main steam tunnel is purged into the Turbine Building. !
62.3.3.1.3 Design Evaluation ,
The compartment response to the postulated high energy line break was calculated ;
using the engineering computer program SCAM. A detail discussion of methodology l and assumptions used in this program can be found in Reference 6.2-4. f The initial conditions for the analysis include the assumption of 102% rated reactor power and the compartment pressures, temperatures and relative humidity as tabulated in Table 6.2-3. Blowout panels are used in place of open vent pathways when the g l emironmental conditions of an compartment must be isolated from the emironment ,
Containment Systems - Amendment 33 6.2 53 i
23A3100 Rev. 3 ABWR standardsarety Analysis neport O
in another compartment. The blowout panels are assumed to open fully against a 2
differential pressure of 0.0352 kg/cm g, and are assumed to remain open.
For the postulated high energyline break, the blowdown mass and energy release rates from the break were determined using moody's homogeneous equilibrium model for critical flow described in Reference 6.2-2. The blowdown mass and energy release rate for the postulated High Energy Line Break (IIELB) in a given compartment l compromised of initial inventory depletion followed 1.y steady critical flow from the ruptured pipe. After the inventory depletion period, tr cak flow, limited by critical flow consideration, continues until the isolation valve ir .' ally closed.
The following paragraphs describe the key assumpdons and calculadon of mass and ,
l energy release rates for the postulated liELB in the RCIC, CUW and Main Steam Tunnel compartments.
6.2.3.3.1.3.1 RCIC Compartment For RCIC a single-ended pipe break, as noted earlier, was postulated. The mass and .
energy blowdown release rate comprised only of flow from the RPV side. The flow from the other side of the break was assumed to be negligible. The blowdown flow comprised ofinitial inventory depletion followed by steady critical flow from the RPV. In computing the critical flow rate, flow loss factors between RPV and break location were ignored for conservatism. Tabulated values of mass energy release rate for the postulated break is shown in Table 6.2-4b. The total blowdown duration of 41 seconds, l as otnious from tabulated values, is based on assumption that the isolation valve stans closing at 11 seconds (1 second instrument response time and 10 seconds buP ..Jogic time delay) after the break and is fully closed in 30 seconds. Considering tnat the isolation valve is a gate valve, non-linear flow area changes with respect to time were used during the valve closure period.
Figure 6.2-37a shows the compartment nodalization scheme used for the pressurization analysis mode! for different break cases. Table 6.2-3 shows the free volume, initial emironmental conditions and DBA characteristics for the compartments which were analyzed. Table 6.2-4 tabulates subcompanment vent path characteristics. The calculated peak differential pressure for the RCIC compartments are tabulated in Table 6.2-3.
l 6.2.3.3.1.3.2 CUW Compartment For CUW a double-ended pipe break, as noted earlier, was postulated. The mass and energy blowdown release rate comprised of flow from both the RPV and BOP sides of ,
the break location. The flow from the RPV side comprised ofinitial inventory depletion followed by steady critical flow. The flow from the BOP side is the depletion ofinventory 4 between the break locadon and the closest check valve. Flow loss factors due to pipe friction, and other mechanical devices such as valves, elbows, tees, etc. were accounted 6.2 S4 Containment Sptems - Amendmsat 33
e E 23A6100 Rev.1 ABWR standardsateer Analysis neport a
Criterion 4: Coolable Geometry .
" Calculated changes in core geometry shall be such that the core remains amenable to cooling
- As described in Reference 6.2-1,Section III.A, conformance to Criterion 4 is demonstrated by cor3 ormance to Criteria I and 2. ,
Criterion 5: Long-Term Cooling ;
"After any calculated successfulinitial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be l removed for the extended period of time required by the long-lived radioactivity remaining in the core." Conformance to Criterion 5 is demonstra., d generically for GE ,
BWRs in Reference 6.2-1,Section III.A. Briefly summarized, for any LOCA, the water level can be restored to a level above the top of the core and maintained there indefinitely.
6.3.3.3 Single-Failure Considerations The functional consequences of potential operator errors and single failures (including those which might cause any manually controlled electrically operated valve in the t ECCS to move to a position which could adversely affect the ECCS) and the potential
/ for submergence of valve motors in the ECCS are discussed in Subsection 6.3.2. There it was shown that all potential single failures are no more severe than one of the single failures identified in Table 6SS.
It is therefore only necessary to consider each of these single failures in the ECCS performance analyses. The worst failure for any LOCA event is the failure of one of the diesel generators which provide electrical power to one HPCF and one RHR/IPFL.
This failure results in the elimination of the greatest amount of flooding capability at both high and low reactor pressures.
6.3.3.4 System Performance During the Accident In general, the system response to an accident can be described as:
(1) Recching an initiation signal (2) A small lag time (to open all valves and have the pumps up to rated speed) ,
(3) The ECCS flow entering the vessel ,
Key ECCS actuation setpoints and time delays for all the ECCS systems are provided in Table GSI. The minimization of the delay from the receipt of signal until the ECCS I pumps have reached rated speed is limited by the physical constraints on accelerating the diesel-generators and pumps. The delay time due to valve motion in the case of the ,
high pressure system provides a suitably conservative allowance for valves available for this application. In the case of the low pressure system, the time delay for valve motion )
Emergency Core Cooling Systems - Amendment 31 6.3-15
23A6100 Rev. 3 ABWR standardsareryAnar sisnepors 9
is such that tne pumps are at rated speed pdor to the time the vessel pressure reaches the pump shutoff pressure.
The ADS actuation logic includes a 29-second delay timer to confirm the presence of Low Water Level 1 (LWL 1) initiation signal. This timer is initiated upon receipt of a high drywell pressure signal (which is sealed-in) and a LWL 1 signal. The timer setting is consistent with the startup time of the ECCS which also must be running before ADS operation can occur. Once the ADS timer is initiated, it is automatically reset if the reactor water level is restored above the LWL 1 setpoint before ADS operation occurs.
For defense-in-depth protection against inventory decreasing events where a high drywell pressure is not present, the ADS actuation logic also includes an 8-minute high drywell bypass timer. This timer is initiated upon receipt of a LWL 1 signal and is automatically reset if the reactor water level is restored above the LWL 1. After this timer runs out, the need for a high drywell pressure signal to initiate the ADS 29-second delay timer is bypassed (i.e., the 29-second delay timer would require only a LWL 1 signal to initiate). The ADS control system also provides the operator with an ADS inhibit switch which can be used to prevent automatic ADS operation as covered by the engincedng operating procedures (refer to Subsection 7.3).
The flow delivery rates analyzed in Subsection 6.3.3 can be determined from the vessel '
press re versus system flow curves in Figures 6.3-4,6.M and 6.% and the pressure versus time plots discussed in Subsection 6.3.3.7. Simplified piping and instrumen tation and process diagrams for the ECCS are referenced in Subsection 6.3.2. The operational sequence of ECCS for the limiting case is shown in Table 6.3-2.
Operator action is not required, except as a monitoring function, dudng the short-term cooling period following the LOCA. Dunng the long-term cooling period, the operator may need to take action as specified in Subsection 6.2.2.2 to place the containment cooling system into operation for some LOCA events.
6.3.3.5 Use of Dual Function Components for ECCS With the exception of the LPFL systems, the systems of the ECCS are designed to rcomplish only one function: to cool the reactor core following a loss of reactor d it. To this extent, components or portions of these systems (except for pressure rdc4 are not required for operation of other systems which have emergency core cooling functions, er vice versa. Because either the ADS initiating signal or the overpressure signal opens the safety relief valve, no conflict exists.
The LPFL Subsystem is configured from the RHR pumps and some of the RHR valves and piping. When the reactor water level is low- the LPFL Subsystem (line up) has priority through the valve controllogic over the other RHR Subsystems for containment cooling. Immediately following a LOCA, the RHR System is directed to the LPFL mode.
6.3 16 Emergency Core Cooling Systems - Amendment 33
23A6100 Rev. 3 ABWR Standard SafetyAnalysis Report p
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Table 6.3-4 Summary of Results of LOC.A Analysis Break Size PCT Maximum Local Break Location (cm2) Systems Available ( Cl Oxidation Based on Appendix K evaluation models:
Steamline Inside 985 1HPCF + RCIC 552 0.03 %
Containment +2 RHR/LPFL +
8 ADS Feedwater 839 1 HPCF + 542 0.03 %
Line 2 RHR/LPFL +
8 ADS RHR Shutdown 792 1 HPCF + RCIC 542 0.03 %
Cooling Suction + 2 RHR/LPFL+
Line 8 ADS RHR/LPFL Injection 205 1 HPCF + RCIC 542 0.03%
- Line + 1RHR/LPFL +
8 ADS High Pressure Core 92 RCIC+2RHR/ 542 0.03 %
Flooder LPFL + 8 ADS Bottom Head Drain 20.3 1HPCF + RCIC 542 0.03 %
Line + 2 RHR/LPFL + 8 ADS Steamline Outside 3939 1 HPCF + RCIC 0.03 %
Containment + 2 RHR/LPFL + 621 8 ADS Based on bounding values:
Steamline Outside 3939 1 HPCF + RCIC 0.03 %
Containment + 2 RHR/LPFL + 619 8 ADS Note: The core-wide rnetal-water reaction for this analysis has been calculated using method 1 described in Reference 6.3-1. This results in a core-wide metal-water reaction of 0.03%
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I 6.3-29 Emergency Core Cooling Systems - Amendment 33 l
l 23A6100 Rev. 2 ABWR standardSafety Analysis Report O
Table 6.3-5 Key to Figures Appendix K Evaluation Models Bounding Values Main Main Main Steamline Steamline Steamline inside RHR LPFL Core Bottom Outside Outside Contain- Feedwater Suction injection Flood Drt Contain- Contain-ment Line Line Line Line Lne ment ment Core Flow 63-12 6.3-21 6.3-21 6.3-21 6.3-44 6.3-21 6.3-21 6 3-67 Minimum 6.3-13 6.3-22 6 3-22 6.3-22 6.3-45 6.3-22 6 3-22 63-68 Critical Power Ratio Water Level 6.3-14 6.3-23 6.3-30 6.3-37 6.3-46 63-53 6 3-60 6.3-69 in Fuel Channel Water Level 6.3-15 63-24 6.3-31 6.3-38 6.3-47 6.3-54 6 3-61 63-70 inside Shroud 63-48 6.3-55 Water Level 6.3-16 63-25 6.3-32 6.3-39 6.3-62 6.3-71 ,
Outside Shroud Vessel 6.3-17 6 3-26 6.3-33 6.3-40 63-49 6.3-56 6 3-63 63-72 Pressure Flow out of 6.3-18 6.3-27 6.3-34 6.3-41 6.3-50 6.3-57 6 3-64 6.3-73 Vessel Flow into 6.3-19 6 3-28 63-35 6 3-42 6.3-51 6 3-58 6.3-65 6.3-74 Vessel ,
Peak 63-20 6.3-29 6.3-36 63-43 6.3-52 63-59 6.3-66 63-75 Cladding Temperature O
6.3-30 Emergency Core Cooling Systems - Amendment 32
t 23A6100 Rev. 3 ABWR standardsarety Analysis Report 7~
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23A6100 Rett. 2 ABWR standardsafetyAnalysis neport O
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23AS100 Rev.1 ABWR standardsaferyAnalysis Report .
O 6.5 Fission Products Removal and Control Systems 6.5.1 Engineered Safety Features Filter Systems The filter systems required to perform safety-related functions following a design basis accident are:
(1) Standby Gas Treatment System (SGTS)
(2) Control room portion of the HVAC System (HVAC)
The control room portion of the HVAC System is discussed in Section 6.4 and Subsection 9.4.1. The SGTS is discussed in this subsection (6.5.1).
G.5.1.1 Design Basis 6.5.1.1.1 Power Generation Design Basis The SGTS has the capability to filter the gaseous efBuent from the primary containment or from the secondary containment when required to limit the discharge of ,
g radioactivity to the emironment to meet 10CFR100 requirements.
6.5.1.1.2 Safety Design Basis :
The SGTS is designed to accomplish the following:
(1) Maintain a negative pressure in the secondary containment, relative to the outdoor atmosphere, to control the release of fission products to the emironment.
(2) Filter airborne radioactivity (halogen and air particulates) in the effluent to reduce offsite doses to within the limits specified in 10CFR100.
t (3) Ensure that failure of any active component, assuming loss of offsite power, carmot impair the ability of the system to perform its safety function.
(4) Remain intact and functional in the event of a safe shutdown earthquake (SSE).
(5) Meet emironmental qualification requirements established for system operation.
(6) Filter airborne radioactivity (halogens and particulates) in the efDuent to reduce offsite doses during normal and upset operations to within the limits O of10CFR20.
Fission Products Removal and Control Systems - Amendment 31 ,
6.S-1
23A6100 Rev. 3 ABWR standard saretyAnalysis Report O
6.5.1.2 System Design 6.5.1.2.1 General The SGTS P&lD is prosided as Figure 6.5-1.
6.5.1.2.2 Component Description Table 6.5-1 provides a summary of the major SGTS components. The SGTS consists of two parallel and redundant filter trains. fhe two SGTS trains are located in two adjacent rooms. Each train is protected for fire, flood, pipe break and missiles. The electrical separation is provided by connecting the two trains to Divisions 2 and 3 electdc power.
The two trains are mechanically separated also. Suction is taken from the secondary containment, including above the refueling area, or from the primay containment via the Atmospheric Control System (ACS). The treated discharge goes to the main plant stack.
The SGTS consists of the following prirscipal components:
(1) Two filter trains, each consisting of a of a moisture separator, an electric process heater, a prefilter, a high efIiciency particulate air (HEPA) filter, a charcoal adsorber, a second HEPA filter, space heaters, and a cooling fan for the removal of decay heat from the charcoal.
(2) Two independent process fans located downstream of each filter train.
6.5.1.2.3 SGTS Operation 6.5.1.2.3.1 Automatic Upon receipt of a high drywell pressure signal or a low reactor water level signal, or when high radioactivity is detected in the secondary containment or refueling floor ventiladon exhaust, both SGTS trains are automatically actuated and one train is manually placed in the Standby mode. When the operation of both the trains is assured, one train is placed in the Standby mode. In the event that a malfunction disables an operating train, the standby train is automatically initiated.
6.5.1.2.3.2 Manual l
The SGTS is on standby during normal plant operadon.It may be manually inidated for l primay containment de-inerting in accordance with the Technical Specifications when required to limit the discharge of contaminants to the environmentwithin 10CFR20 limits. Normal operation of the SGTS while the plant is in the startup, power, hot f l
standby, and hot shutdown modes of operation is much less than 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year for both trains combined. However, if 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> of operation per year for either tram (excluding tests) is to be exceeded, the COL applicant is required to demonstrate that 6.5-2 Fission Products Removal and Control Systems - Amendment 33
23A6100 Rev. 3 ABWR standardsafery Analysis Report
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5 L
SurTeillance monitoring of the temperatures in the dqwell, is provided by multiple temperature sensors distributed throughout the dowell to detect local area " hot-spots" and to monitor the operability of the drywell ;
cooling system. With this drywell air temperature monitoring system supplied by multiple temperature sensors throughout the drywell, the Regulatory Guide 1.97 requirements for monitoring of dowell air ,
temperature are met and provides the ability to determine drywell bulk average temperature, (k) Dgwell/Wetwell Hydrogen / Oxygen Concentration l The Containment Atmospheric Monitoring System (CAMS) consists of two independent and redundant drywell/ containment oxygen and hydrogen concentration monitoring channels. Emergency response ;
actions regarding these variables are consistently directed toward minimizing the magnitude of these parameters (i.e., there are no safety :
actions which must be taken to increase the hydrogen / oxygen levels if they are low). Consequently, the two channel CAMS design prosides adequate PAM indication, since, in the event that the two channels of j'~'% information disagree, the operator can determine a correct and safe V action based upon the higher of the two (in-range) indications.
(1) Wetwell Atmosphere AirTemperature Surveillance monitoring of temperatures in the wetwell is pro ided by multiple temperature sensors dispersed throughout the wetwell, the required indication of bulk average wetwell atmosphere temperature.
(m) Standby Liquid Control Flow No flow indication is provided for the ABWR design. The positive displacement SLC pumps are designed for constant flow. Any flow blockage or line break would be indicated by abnormal system pressure (high or low as compared to RCS pressure) following SLC initiation. ;
Changing neutron flux, SLC and SLC tank level are substituted for SLC .
flow and are considered adequate to verify proper system function. One !
channel of SLC discharge pressure is provided in addition to the i
monitoring of neutron flux.
(n) Suppression Pool /Wetwell Water Level ,
Regulatory Guide 1.97 suggests two ranges for suppression pool water level (i.e., botom of ECCS suction to 1.5m above normal water level and ,
Information Systems important to Safety- Amendment 33 7.6-11
23A6100 Rev. 3 ABWR studardsareryAnalysis neport j O
top of vent to top of weir wall [B%R 6, Mark III Containment]). The ;
ABWR provides:
(i) Four (4) divisions of narrow range suppression pool water (e.g.,
approximately 0.5 meters above and below normal water level) for control of normal water level and automatic transfer of RCIC and HPCF suctions.
(ii) Two (2) wide range suppression pool /wetwell water level instruments fro approximately the centerline of the ECCS suction ,
piping to the wetwell spray spargers. This range allows for control of suppression pool /wetwell water level in the vicinity of the spray spargers at the high end and the ECCS pumps (vortex limits) at the low end.
Two (2) wide range wetwell levelinstruments are sufficient to control water level at the high level and at the low level by using the highest reading and the lowest reading instruments, respectively, should the instruments disagree. In addition, The low end measurement to the centerline of the ECCS suction piping is considered sufficient since this level measurement is low enough to allow control of the pump vortex limits.
(Note: See dowell water level for instrument range overlap).
(o) Drywell Water Level The lower drywell water level measurement below the RPV (other than sump level) is not warranted because ofits ability to senive a severe accident (core melt) and because of the following: When the suppression pool level is increased to accommodate severe accident drywell flooding (per the ABWR EPGs), suppression pool level will stop increasing while the water spills into the lower drywell through the vents.
Once drywell and werwell water levels equalize, the increase in dr)well level will be monitored by the wetwell water level monitors up to the bottom of the RPV. (See also upper drywell water level monitoring for instrument overlap.)
In aodition to the above discussion oflower drywell water level monitoring, the ABWR design provides for two (2) upper drywell water level monitors. The range of these instruments in from approximately 0.5 meters below the RPV (lower drywell and above wetwell to lower drywell vents) to the maximum primary containment water levellimit
'i (MPC%H.) (upper dr)well and approximately five (5) meters above TAF.). This lower range provides an approximately 0.5 meter Information Systems important to Safety- Amendment 33 7.5 12
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23A6100 Rev. 3 ABWR standardsareryAnalysis neport ,
14.2.12.1.3 Recirculation Flow Control System Preoperational Test (1) Purpose To verify that the operadon of the Recirculadon Flow Control (RFC) System, ,
including that of the adjustable speed drives, RIP trip and runback iogic, and the core flow measurement subsystem, is as specified.
(2) Prerequisites The constmcdon tests have been successfully completed, and the SCG has reviewed the test procedure and approved the initiation of testing. The '
following systems shall be available, as needed, to support the specified tesdng and the corresponding system configurations: Reactor Recirculation System, Feedwater Control System Steam Bypass and Pressure Control System, electric power distribution system /instmmentation and control power supply, Process Computer System, Reactor Water Cleanup System, CRD System, RCIS, ,
Neutron Monitoring System, automatic power regulation system, condensate ;
and feedwater system and Reactor Protecdon System. .
E (3) General Test Methods and Acceptance Criteria Some portions of the RFC System testing described below may be performed in conjunction with that of the recirculation system, as described in '
Subsection 14.2.12.1.2. In any case, close coordination of the testing specified '
for the two systems is required in order to demonstrate the proper integrated sptem response and operation.
Performance shall be observed and recorded during a series ofindisidual component and integrated system tests. These tests shall demonstrate that the RFC System operates properly as specified in Subsecdon 7.7.1.3 and applicable RFC System design specification through the following tesdng:
(a) Proper operation ofinstrumentation and system controls in all combinations oflogic and instrument channel trip, including stability control and protection (SCP), alternate rod insertion (ARI), ,
recirculation flow block, recirculation pump trip (RPT) and runback circuity (RPT testing will specifically include its related A'IWS function) -
(b) Proper functioning ofinstrumentation, including calibration of process ,
l sensors, operator displays and alarm annunciation, confirmation of signal continuity, scaling and validation logic; and operator / technician interfaces and services (c) Proper functioning of the core flow measurement subsystem Spectfic Information to be Included in Final Safety Analysis Reports - Amendment 33 14.2 17
23A6100 n2v.1 ABWR senadantsaferyAnarrsisReport O
(d) Proper operadon of the RFC System control algorithm in all design operating modes and alllevels of controls (c) Proper operation of the adjustable speed drives, recirculadon pump and pump motor component (f) Fault-tolerant capability of the redundant RFC digital controller upon a simulated single processor channel failure (g) Capability of the self-test and online diagnostic features of the FTDC in identifying the presence of a fault and determining the location of a failure (h) Proper operation ofinterlocks and trip logic and all control functions (i) Proper operation of the technical interface unit (TIU) in the various provided operational modes as defined by the RFC design specification
- t (j) Proper steady-state and coastdown performance of M-G sets (k) Capabilities of the FTDC cold and warm start features (i.e., self-starting .
following a power interniption to the full system and bringing a processing channel online with the other channels in operation without the need for operator or technician action)
(1) Proper operation of the RIPS trip function by verifying that RIPS trip in response to simulated high dome pressure, low water level, and both signals as specified by the appropriate RFC System design specification 14.2.12.1.4 Feedwater Control System Preoperational Test (1) Purpose ,
To verify proper operation of the Feedwater Control System (FWCS),
including individual components such as controllers, indicators, and controller software settings such as gains and function generator curves.
(2) Prerequisites The construction tests have been successfully completed, and the SCG has reviewed the test procedures and approved the initiation of testing.
l Preoperational tests must be completed on lower level controllers that do not strictly belong to the FWCS but that may affect system response. All FWCS l components shall have an initial calibration in accordance with vendor l instructions. Appropriate instrumentation and control power supply, Turbme ,
Control System, Reactor Recirculation Flow Control System Condensate and Feedwater System, Proce=s Computer System, Reauor Water Cleanup System. l 14.2 18 Specific Information to be included in Final Safety Analysis Reports - Amendment 31
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23A6100 Rev. 3 ABWR standardsafetyAnalysisneport O 3 l
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l Table 15.7-2 Isotopic Source and Release to the Environment in Curies i Isotope Flow Rates (Design Integrated Releases (at Max !
Basis) Tech Spec) !
T=0 T=2 min T .30 min
. T=30 min Charcoal Total i Delay {
isotope (Ci/sec) (Ci/sec) (Ci/sec) (Ci) (Ci) (Ci) i Kr-83m 3.42 E-3 3.38E-3 2.84 E-3 20.47 11.01 31.47 Kr-85m 6.02E-3 5.99E-3 5.57E-3 40.13 31.02 71.14 ;
Kr-85 1.8BE-5 1.88E-5 1.88E-5 0.14 0.14 0.27 Kr-87 2.07E-2 2.03E-2 1.57E-2 113.35 45.61 158.96 l Kr-88 2.07E 2 2.05E-2 1.83E-2 131.68 87.23 218.90 Kr-89 1.28E-1 8.25E-2 1.77 E-4 1.28 0.00 1.28 Kr-90 2.82E-1 2.15E-2 4.73E-18 0.00 0.00 .0.00 Kr-91 3.42 E-1 2.41 E-5 0 0.00 0.00 0.00 <
Kr-92 3.42E-1 BE-21 0 0.00 0.00 0.00 .>
Kr-93 9.03E-2 3.2E-30 0 0.00 0.00 -0.00 .
Kr-94 2.22E-2 0 0 0.00 0.00 0.00 ,
Kr-95 2.07E-3 0 0 0.00 0.00 0.00 Kr-97 1.35E 5 0 0 0.00 0.00 0.00 Total 1.26 0.15 4.26E-2 307.03 175.00 482.03 Xe-131m 1.47E-5 1.47E-5 1.47E-5 0.11 0.10 0.20 Xe-133m 2.82 E-4 2.82E-4 2.80E-4 2.02 1.24 3.26 ;
Xe-133 7.90E-3 7.90E-3 7.89E 3 56.78 48.57 105.35 l Xe-135m 2.63E-2 2.41 E-2 6.77E-3 48.72 0.00 48.72 l l 2.25E-2 2.17E-2 156.54 9.17 165.72 j Xe-135 2.26E-2 Xe-137 1.47E 1 1.02E-1 6.53E-4 4.70 0.00 4.70 Xe-138 8.66E-2 7.85E-2 2.00E 2 144.09 0.00 144.09 - ,j Xe-139 2.82E-1 3.60E-2 1.09E-14 0.00 0.00 0.00- +
Xe-140 3.01 E-1 6.65E-4 0 0.00 0.00 0.00 Xe-141 2.45E-1 2.44E-22 0 0.00 0.00 0.00 -i Xe-132 7.15E-2 1.76E-31 0' O.00 0.00 0.00. !
Xe-143 1.20E-2 0 0 0.00 0.00 0.00 i Xe-144 5.65E-4 0 0 0.00 0.00 0.00 l Total 1.20 0.27 5.47E-2 412.97 59.08 472.05 .
Kr+ XE 2.46 0.43 0.1 720.00 234.08 954.08 l
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Table 15.7-3 Offgas System Failure Meteorology and Dose Results Meteorology Whole Body Dose .
l 1.37E-3 sec/m3 0.275 Rem l
f i
O 1 i Radioactive Release from Subsystems and Components- Amendment 33 15.7 13 f
23A6100 Rev 3 ABWR sexcudsaretyAnalysis a: port O
Table 15.7-4 (Deleted)
/
l O\ l i
15 7-14 Radioactive Release from Subsystems and Components - Amendment 33
5 23A6100 Rev. 3 standard sareryAnalysis Report ABWR O
Seismic failure of DC power also is assumed to lead directly to core damage. Without ,
l !
DC power, all instrument and equipment control power is lost and the reactor cannot be controlled or depressurized. In the seismic margins analysis it is assumed that this results in a high pressure core melt. The limiting components for DC power are the batteries (HCLPF = 1.13g) and the cable trays (HCLPF = 0.74g).
l It is possible that a large seismic event could impair the ability to scram due to deformation of the channels that enclose each fuel bundle in the event that the scram ;
function is impaired, the only means of reactivity control would be the Standby Liquid Control (SLC) System. Seismic failure of the SLC system to insert borated solution into the reactor is controlled by the seismic capacity of the SLC pump (HCLPF = 0.62g) and the SLC system boron solution tank (HCLPF = 0.62g).
Emergency AC power and plant senice water were both treated as having the same ,
effects in the seismic margins analysis. Failure of either system would require only one additional failure to resultin core damage. The limiting components for seismic failure l
of emergency AC power are the diesel generators (HCLPF = 0.62g), transformers (IICLPF = 0.62g), motor control centers (HCLPF = 0.62g), and circuit breakers
-s (HCLPF = 0.63g). The limiting components for seismic failure of plant senice water are the senice water pumps (HCLPF = 0.63g), room air conditioners (HCLPF = 0.63g), and l
the senice water pump house (HCLPF = 0.60g). ,
Most Sensitive Cornponents The HCLPFs of the accident sequences with the lowest HCLPFs could be increased by l
increasing the individual HCLPFs of the ACIWA pumps, the fuel channels, or the RHR heat exchangers. The HCLPFs of the appropriate accident sequences would be increased by an amount equal to the increase in the HCLPF of any of these components.
The only single item that could, by itself, decrease the HCLPF of any accident sequence below 0.60g is a Categoty I structure having a HCLPF below 0.60g. This would also decrease the HCLPF of accident class IE; ATWS with high pressure melt due to loss of inventory. The lowest HCLPFs for Category I structures are 1.11g and 1.12g.
The only system that could, by itself, result in lowering an accident sequence HCLPF below 0.60g is DC power. DC power has two components that could fail the sequence-the batteries (HCLPF = 1.13g) and the cable trays (HCLPF = 0.74g).
AC Independent Water Addition (ACIWA)
The ACIWA provides a diverse capability to provide water to the reactor in the event that AC power is not available and is important in preventing and mitigating severe j accidents. The system has a diesel driven pump with an independent water supply and q O~ all needed valves can be accessed and operated manually. In addition, support systems ;
normally required for ECCS operation are not required to function for ACIWA operation. The ACIWA can provide either vessel injection or drywell spray in the event Important features identified by the ABWR PRA - Amendment 33 19.8 9 l I
I 23A61CD Rev. 3 ABWR standantsarery Anstrsis Repon 9
all AC power is unavailable. Although the system pumps are housed in an external building (shed), the collapse of the buildir.g would not prevent the diesel driven pump from starting and running.
Seismic Walkdown In addition to the above identified features,it wasjadged important that the seismic walkdown noted in Subsection 19.9.5 be conducted to seek seismic minerabilities.
19.8.3 Important Features from Fire Analyses 19.8.3.1 Summary of Analysis Results An ABWR fire risk screening analysis based on the EPRI Fire Induced Vulnerability Evaluation (FIVE) mehodology was performed to assess vulnerability to fires within the plant. Each scena eu ited was calculated to have a core damage frequency less than IE-6.
19.8.3.2 Logical Process Used to Select important Design Features The screening criterion for EPRI's FIVE methodology provided the primary basis for systematically evaluating important design features. The FIVE methodology provides procedures for ident fying fire coupartments for evaluation purposes, defining fire i ,
ignition frequencies, and performing quantitative screening analyses. The criterion for screening acceptability and dismissal from any more detailed consideration is that the ;
frequency of core damage from any postulated fire be less than IE-6 per year.
Five bounding Sre scenarios and corresponding ignition frequencies were develooed on the basis of the FIVE methodology. Each scenario was calculated to have a core damage frequencyless than 1E-6 and hence screened from furuta consideration.
Validity of these outcomes is contingent upon specific assumptions regarding the design features and performance capabilities of structures and equipment.
Consequently, the study was systematically reviewed to identify those procedures, assumptions, and features which are necessary in the Sre risk assessment analysis to achieve core damage frequencies less than IE-6 and thus pass the FIVE methodology screen.
19.8.3.3 Features Selected Table 19.8-3 lists the features selected and the basis for each feature being considered important. These features are those necessary to maintain fire initiated core damage frequencies below the IE-6 screening criterion. The proper functioning of these features assures the capability to mitigate the postulated fires. Features identified as a result of the review of the Level 1 internal eventt, analysis are also important in die fire analysis but they are not included here unless they have some Src unique significance.
19 B-10 important features identified by the ABWR PRA - Amendment 33
23A6100 Rev. 3 StandardSafety Analysis Report ABWR susceptibility of external floods, plant and site specific procedures will be developed by the COL applicant for severe external flooding using the following guidelines:
(1) Check that the door between the turbine and senice buildings is closed.
(2) Sandbag the external doors to the (a) Reactor building, t (b) Control building, i (c) Service building, (d) Pump house at the ultimate heat sink, (e) Diesel generator fuel oil transfer pits, and 1
(f) Radwaste building. l l
(3) Close and dog all external water tight doors in the reactor and control buildings.
(4) Shut the plant down.
(5) Use power from the diesel generators or CTG if offsi:e power is lost.
Underground passages between buildings would not be affected because they are required to be watertight.
19.9.4 Confirmation of Seismic Capacities Beyond the Plant Design Baser, The seismic analysis w umed seismic capacities for some equipment for which information was not available. It is expected that these capacities can be achieved, but confirmation must be deferred to the COL applicant when sufficient design detail is anilable. The actions specified in Subsection 19H.5 will be taken by the COL applicant.
19.9.5 Plant Walkdowns A plant walkdown to seek seismic vulnerabilities will be conducted by the COL applicant as noted in Subsection loH.5.
Similar walkdowns will be conducted by the COL applicant for intemal fire and flooding events.
19.9.6 Confirmation of Lost of AC Power Event The COL applicant will confirm the frequency estimate for the loss of AC power event (Subsection 19D.3.1.2.4). This review will addr:ss site-specific parameters (as indicated 19 9-3 COL License information - Amendment 33
I 23A6100 !?ev. 3 ABWR standardsaferr Anstrsis Report O
in the staffs licensing review basis document) such as specific causes.(e.g., a severe storm) of the loss of power, and their impact on a timely recovery of AC power.
19.9.7 Procedures and Training for Use of AC-independent Water Addition System Specific, detailed procedures will be developed by the COL applicant for use of the ACindependent Water Addition System (including use of the fire truck) to provide vesselinjection and dowell spray. Training will be included in the COL applicant's crew training program.
The procedures to be developed by the applicant will address operation of the ACRVA for sessel injection or dqwell spray operation. Opention of the ACIWA System in the vessel injection mode requires valves F005, F101, and F102 to be opened and valve F592 to be closed. Reactor depressurization to below ACRVA System operating pressure is required prior to ACRVA operation in the vessel injection mode. Operation of the ACRVA in the drywell spray mode requires valves F017, F018, F101, and F102 to be opened and valve F592 to be closed. These vahrs are shown on Figure 5.4-10. The diesel fire pump will start automatically when the ACRVA is properly aligned for vessel injection or drywell spray. If the normal firewater system water supply is unavailable, the alternate water supply can be made available by opening the manual valve between the diesel driven fire pump and the alternate water supply. This valve is shown in Figure 9.5-
- 4. Ifit is necessary to use a fire truck for vessel injection or drywell spray, valve F103 must l be opened in addition to operation of the valves discussed above for ACRVA operation.
The valve for operation of the ACIWA using the fire truck is also shown on Figure 5.4-
- 10. All of the valves required for ACIWA operation are manually operable.
If it is necessary to operate the ACBVA, radiation levels may be elevated in the rooms where the valves required for ACBVA operation are located. The applicant will make dose rate calculations for the specific configuration being constructed. These calculations willinclude the specific piping layout, shielding considerations, the potential for sptems within the room to have recently been operated and thus contain radioactive coolant, and any other factors that significantly affect the dose rates. These dose rate calculations will be considered in the development of the specific plant procedures for ACIWA operation.
19.9.8 Actions to Avoid Common-Cause Failures in the Essential Multiplexing System (EMUX) and Other Common-Cause Failures ,
To reduce the potential for significant EMUX common cause failures, (Subsection 19N.4.12), the COL applicant will take the following actions:
(1) To eliminate remote multiplexirig unit (IBfU) miscalibration as a credible source of EMUX common cause failure, administrative procedures will be established to perform cross-channel checking of RMU outputs at the main 1994 COL License Information - Amendment 33 i
O O O. .
p Table 19D.6-12 RPS Failure Data ih
= _,
a
{
. Failure Beta Test Interval Mission or Restore g
g l Acronym Event Probebility Rate (/h) Factor (h) (h) Reference
- Note S22BF LOAD DRIVER 1.87E-03 5.06E-06 730 4 8 S23CF LOAD DRIVER 1.87E-03 5.06E-06 730 4 8 h S24DF LOAD DRIVER 1.87E-03 5.06E-06 730 4 8 S31AF LOAD DRIVER 1.87E-03 5.06E-06 730 4 8 S32BF LOAD DRIVER 1.87E-03 5.06E-06 730 4 8 S33CF LOAD DRIVER 1.87E-03 5.06E-06 730 4 8 S34DF LOAD DRIVER 1.87E-03 5.06E-06 730 4 8 S41AF LOAD DRIVER 1.87E-03 5.06E-06 730 4 8 S42BF LOAD DRIVER 1.87E-03 5.06E-06 730 4 8 g S43CF LOAD DRIVER 1.87E-03 5.06E-06 730 4 8 h
S44DF LOAD DRIVER 1.87E-03 5.06E-06 730 4 8 y
~
TFUSEA BACKUP SCRAM LOGIC FUSE 3.29E-03 5.00E-07 13140 4 15 TFUSEB BACKUP SCRAM LOGIC FUSE 3.29E-03 5.00E-07 13140 4 15 TLU1F TRIP LCGIC UNIT 2.95E-04 5.00E-06 0.5 4 18 14 TLU2F TRIP LOGIC UNIT 2.95E-04 5.00E-06 0.5 4 18 14 TLU3F TRIP LOGIC UNIT 2.95E-04 5.00E-06 0.5 4 18 14 %
E TLU4F TRIP LOGIC UNIT 2.95E-04 5.00E-06 0.5 4 18 14 g a.
- See Table 19D.6-13 for References and Notes.
R f
h E
l
- iii.
a D
- i
=
23A61cD Rev. 3 ABWR Standard Safety Analysis Report O'
Table 19D.6-13 PRA Failure Rate Reference Documents and Notes
References:
- 1. Failure Rate Data Manual, NEDE 22056. Rev. 2, GE 1986.
- 3. IEEE Standard 500,1984.
- 4. BWR Owner's Group TechnicalSpecification Improvement Methodology, Part 1, NEDC-30936-P, November 1985.
- 5. Notes from F.E. Cooke regarding the Nuclear Boiler System Logic Diagram (Interlock Block Diagram),10 June 1987. GE DRF A00-05225 Volume 1.
- 7. Common-Cause Fault Rates forInstrumentation and Control Assemblies, NUREGICR-2771 EG&G Idaho, Inc., February 1983.
- 8. Clinton NSPS Self-Test Technical Specification, DRF A00-2373, Page 155,158,159 and 633.
- 9. Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications, NUREG/CR-1278, August 1983.
- 10. Deleted.
- 11. Deleted.
- 12. HPCF Technical Specification, GE Document 22A6278, Rev. 2.
- 13. Alto Lazio Station Reliability Analysis, NEDE 30090, General Electric Company, December 1984.
- 14. Data Summaries of Licensee Event Reports of Selected lastrumentation and Control Compo-nents at U.S. Commercial Nuclear Power Plants, January 1,1976 to December 31,1981, NUREG/CR-1740, Rev.1, July 1984.
- 15. EPRI ALWR Assumptions and Groundrules for Evolutionary Plants.
- 16. NEDC 30851P-A, page C-5, Note 7 (PB-3 for lamba, C-4 for CCF).
- 17. L.G. Frederick, Appendix 19N.
- 18. Study done by Barry Simon, Reference NUMAC Field Data.
- 19. Joint Study Report of SSLC Reliability Analysis, . No. lF-R-389, Page. 7.77 and MIL-HDBK-217C.
- 20. Deleted. ,
- 21. DOE /lD-10206, July 1988.
- 22. EPRI NP/2433.
- 23. Letter March 7,1989, Assumed ASWR Features for PRA, from J.D. Duncan to R.P. Raftery and O. Gokcek.
- 24. C.D. Gentil lon, Component Failure Data Handbook, EGG-EAST-8563, Idaho National Engineering Laboratory, April 1991.
- 25. BWR Scram System Reliability Analysis - Part 1 Class 111, NEDE-21514-1 & 2.
December 1976.
190.634 fault Trees - Amendment 33
23A6100 Rev. 3 ABWR standardsarery Anatysis neport Pr U
a scaling steam evaporator, the ABWR reference design generates this steam from the high pressure heater drain tanks using tank connections such that the incoming dmins ,
are routed via a liquid drain loop seal. Thus, only the minimal amount of cycle gases that may be dissolved in the condensed drains is allowed to enter the drain tanks.
Sealing steam is taken from the drain tanks, through the tank vent, as the degassed drains are allowed to reboil at such a slow rate that no low volatility product can escape the liquid phase and contaminate the vented steam.
Through this process, relatively high purity sealing steam is generated for use during plant normal operation above approximately 50% load. During plant startup, sealing steam is provided directly by main steam but the long term average amount of radioactivity that may be released even with abnormally high levels of fuel failure is still [
quite small as plant startup radioactivity levels are relatively low and duration is relatively short. Finally, to permit continued plant operation even in the extre.nely unlikely presence of multiple fuel rod failures, the gland seal system includes a connection for supplying sealing steam from the plant auxiliary (startup) boiler. -
Question 430.84 .
For turbine bypass system:(10.4.4)
(1) Provide figures which delineate the system and its components.
(2) Clarifywhedier the systen includes pressure-reducer assemblies for the bypass valves to reduce steam pressure prior to steam discharge into the condenser.
i Response 430.84 (1) Figures 10.4-9 and 10.4-10 have been added to delineate the system and its l
components.
(2) The detailed design will follow standard industry practice and reduce the pressure sequentially through orifices prior to entering the condenser. In addition, please note that the valves will be 22.9 cm dia globe type as shown in Figure 10.4-9 which also indicates the actuation mechanism and associated motive power.
Upon a turbine trip or generator load rejection, the start of the bypass valve flow is delayed no more than 0.1 seconds after the start of the main turbine -
stop or control valve fast closure. A minimum of 80% of the bypass s) stem capacity is established within 0.3 seconds after the start of the stop or control valve closure.
[ .
k/ The bypass system quality design codes are defined in Section 3.2.
20.3.10 21 Response to Tenth RAI- Reference 10- Amendment 33
2.1A6100 Rn.1 ABWR staasardsataryAnarrsisnaport O
Question 430.85 For the circulating water system: (10.4.5)
(1) Describe the function cf the waterbox fill and dram subsystem mentioned in ABWR Subect2on 10.4.5.2.1. /dso, describe the " makeup water" shown in SSAR Figure 10.4-3.
(2) Provide the worst possible flood levels that can occur in the applicable plant buildings as a result of circulating water system failure and indicate how safety-related equipment located in the building is protected against such flooding.
Response 430.85 (1) The waterbox fill and drain subsystem performs the following two functions:
(a) Following circulating water astem maintenance and/or inspection from the inside, the subsystem uses turbine senice water outflow to ~
completely refill any previously drained section of the circulating water system. Thus, the circulating water pump can be started without any difficult valve throttling being required and without risk of water -
hammer.
(b) The fill and drain subsystem is also used to perruit rapid draining of the series connected condenser water boxes by gravity flow into the circulating water sump. The sump is provided with a vertical wet pit centrifugal pump which can discharge the collected drams, via the turbine senice water system discharge header, to the power cycle heat sink (cooling tower basin, where applicable).
Overall, the subsystem function is to permit expeditious draining and refill of the condenser tube side and, thus, contribute to the plant ability to respond to potential circulating water leaks with minimal loss of availability.
"Make-upwater" to the circulating water system is provided from the site water supply, as required to compensate for cooling tower evaporation and drift water losses. Makeup water flow rate is normally controlled automatically to maintain a constant level in the cooling tower basin.
(2) As noted in Response to Question 430.73, the worst possible flood that can affect the turbine building would result in a flood level slightly higher than grade. Such a flood, however, would not impact any sa'ety related equipment as no such equipment is located inside the turbine building and all plant safety related facilities are protected against external flooding.
s .-
20.3.10-22 Response to Tenth RAI-Reference 10- Amendment 31
ATTACHMENL3 NRC_C_ommentLoAABWILSSAARAmeto dment2 Land.Dispo sitions item
- Comment Disposition Comment No. Type (See Legend)
Chapter 2, Geosciences:
1 Table 2.1 SRP Section 2.5.3 Surface Faulting incorporated 4 The table indicates that there are no limits on surface faulting at a site. The SSAR table should state that a site is not acceptable of there is fault at or near the ground surface.
2 Typographical error page 2.3-1 Section 2.3.1.2, . Incorporated 1 (1) SSE Ground Motion, third line . 3.7-2 "of" should be ..
3.7-2 "for."
3 Typographical error page 2.3-4 Section 2.3.2.19 incorporated 1 third line, .. " Water / resources * ... should be . " water resources" Chapter 3, Civil Engineering:
4 (Editorial comment): Conventionally, in SI units, kg and kgf To be provided to NRC prior to March 4,1993 (NRC 1 represent the mass and force, respectively. Use of kgf/cm 2 is letter dated November 9,1993).
more appropriate than kg/cm2 for force in all SSAR sections.
5 Subsection 3.3.2.3: .
Incorporated 1 The title should include " systems," and read "Effect of Failure of Structures, Systems or Components Not Design for Tomado Loads."
6 Subsection 3.7.1.2: Incorporated 4 The term (3.5/f)0.2 in Equation 3.7-1 should be (f/3.5)0.2, 7 Subsection 3.7.2 and 3H: Analysis method and procedures provided at 2/22- 3 GE did not provide the analysis method and procedures for 25/93 Structural Audit. Added to Subsection 3.8.5.5.
seismic sliding evaluation in the SSAR as committed in the
- resolution of DFSER Open item 3.7.2-1.
8 Subsection 3.7.5.4 should be revised to reflect a commitment incorporated. 3 for the COL applicant to describe the process for completion of the design of balance-of-plant and non-safety related systems to minimize ll/l interactions and propose procedures for an inspection of the as-built plant for 11/1 '
interactions.
9 (Editorial Comment): Subsection 3.6.2.1.1.4: Incorporated 1 The word "torospherical" should be "torispherical."
1
. . . _ - __ __ - . - . _ _ ~ _
Incorporated 4
~ 10 Subsection 3.8.4.1.5:
The phrase " .. combination of both. Various type of frames form a support system with transverse and longitudinal bracing to the nearest wall or ceiling to take the seismic loads." Should be added to the end of the second paragraph.
Incorporated 4 11 Appendix 3A:
The title of Figures 3A-9 and 3A-10 should be switched with each other.
12 Subsection 3A.3.2: Coetficient of Equation (3A-1) converted to 70. 1 in Equation 3A-1, it appears that the value for the coefficient (1000) and the term (om) need to be corrected to account for the conversion from British units to metric units.
13 Subsection 3H.1.4.5: Incorporated 1 The loading conditions "H" and *L should be "H'" and *Lo",
respectively.
14 Subsection 3H.2.4.5: Incorporated 1 The first seven lines on Page 3H.2-11 should be deleted.
15 Table 3H.2-5, page 3H.2-20: trcorporated 1 "E" should be "E" E' and E are used differently in Section i
3H.3.4.3.3.2. (Usually, e'is used for SSE and E for OBE).
! Similarly, Subsection 3H.3.4.5 and 3H.3.5.3.2, the loading Ccadition "E" on Pages 3H.3-9 and 3H.2-11 should read
. E'."
(Editorial Comment): Table 3H.2.4.5 is duplicated.
Chapter 3, Mechanical Engineering:
incorporated 4 I 16 Subsection 1 A.2.9 - Coolant System Values Testing Requirements (ll.D.1)
Subsection 1 A.3.7, " Testing of SRV and Discharge Piping,"
was added in Amendment 30, and then was deleted in amendment 31. It contained a requirement for the COL applicant to confirm that any SRVs of discharge piping not similar to those that were tested in the generic program will be tested in accordance with NUREG-0737 guidelines. As discussed in FSER Section 14.1.3.3.5.11, this information provided the basis for the resolution of COL Action item 3.9.3.3-1 and 14.1.3.3.5.11-1. Therefore,it should be ,
included in the SSAR. ,
2
w Subsection 3.9.1.5 Inelastic Analysis Methods Subsection 3.9.1 rr<ised to be consistent with 4 17 The CRD outer tube was deleted from the list of Subsection 4.5.1.2.2.9.
components that prevent ejection of the CRD in the unlikely event of a failure of the ASME Class 1 weld that attaches the CRD housing to the stub tube in the bottom head of the reactor pressure vessel. This is now not consistent with the discussion in SSAR Subsection 4.5.1.2.2.9," Integral intemal Blowout Support,* which states that the CRD outer tube and middle flange is one of the safety related components in the load path that provides the anti-ejection function during this postulated event. The staff's evaluation of this issue in the FSER Section 3.9.1 included the outer tube as part of the load path. In addition, based on information in previous SSAR amendments, the staff's discussion in FSER Section 3.9.1 stated that the cylindrical bodies of the CRD guide tube, housing, and outer tube were the only parts of these components that were analyzed by inelastic analysis. SSAR Subsection 3.9.1.5 has now been revised to state that only the cylindrical body of the guide tube was analyzed inelastic atly. The SSAR should be revised to eliminate the discrepancy between Subsections 3.9.1 and 4.5.1.2.2.9.
Addressed in October 22,1993 GE letter to NRC 4 18 Table 3.9 Plant Events The number of cycles / events for most of the plant operating justifying all Table 3.9-1 entries with the exception of events and some of the dynamic loading events listed in this Events 6 and 14 which were increased by a factor of table have been reduced by a f actor of approximately 1.5. 1.5.
This reduces the number of cycles / events back to those reported in the SSAR Amendment 1 numbers for a 60-year plant life. This was Open item 3.9.1-1. In response to this request, GE submitted Amendments 21 and 23 which generally increased these numbers by a factor of 1.5. The staff reported this in its FSER, Section 3.9.1 and found it -
acceptable.' The number of cycles / events reported in Amendments 21 and 23 should be retained unless GE can justify the reduced numbers for a 60 year life.
Table 3.9-8, inservice Testing Safety-Related Pumps and Valves
- a. B21 Nuclear Boiler System Valves, P 3.9-101 Incorporated 4 19 The figure for Valve F039 should be 5.1-3 sh.4 3
d 20 b. C41 Standby Liquid Control System Valves, P 3.9-104 incorporated 4 The test parameter for Valve F003 should be R. Valves F026 and F700 are missing.
21 c. C51 Neutron Monitoring (AlIP) System Valves, P 3.9- Valve data reorganized for better categorizCion. 3 105 The Code category for Valve J004 should be A.C.-
22 d. D23 Containment Atmospheric Monitoring System incorporated 4 Valves, P 3.9-105 The testing of Valves FO01 should be L (test parameter) at RO (test frequency) and S (test parameter) at 3 month (test frequency).-
The testing of Valves F004 through F008 should be L, P (test parameter) at RO (test frequency) and S (test parameter) at 3 month (test frequency).
23 e. E1 Residual Heat Removal System Valves, P 3.9-111 Added valves E11-F718 and F720. 4 Valves F718 and F720 are missing.
l
! 24 f. P54 Instrument Air System Valves, P 3.9132 Incorporated 4 A reference to note (h3) should be added to the description column and S should be added to the test parameter column for Valves F276 and F277.
I- 25 g. P54 High Pressure Nitrogen Gas Supply System Valves, incorporated 4 l P 3.9-132 -
A reference to note (h1) should be added to the -
i . description column for Valve F008. ,
i l
l
.4
.--._v-.-., _,,a av.. . o- -~ e -~ vw v<~ -
e- s <~ ~ne<, e - ,- + - -a+ ,-.- ---- -------=a-----*
4 26 h. T31 Atmospheric Control System Valves, P 3.9-135 and Incorporated 4 3.9-138 A reference to note (h2) should be added to the description column for Valves F001 throu9h F004 and F006.
l The Code category for Valve D001 should be D and its l valve function should be 1, P. I The Code category for Valve D002, the wetwell rupture disk, should be D.
27 1. U41 Heating, Ventilation and Air Conditioning System HVAC does not penetrate containment. "1" is for 3 Valves, P 3.9-139 primary containment isolation only; thus, the valve function is "A*. "L" is leak test for "l* only; thus, testing The valve function for Valves F001 and F002 should be is "P' only.
A, I and their testing should be P. L (test parameter) at 3 month (test frequency).
Valves F003 and F004 are missing. Valves F003 and F004 added.
28 Subsection 3.10.2.1.3.3 - Seismic Qualification by Testing incorporated 3 The next to last sentence should be revised to read:
" Operability of equipment is verified as described in Subsection 3.7.3.2," and the last sentence should be deleted. These changes are necessary in order to become more consistent with the stafl's position in SECY-93-087, Which was approved in the SRM dated July 21,1993.
29 Subsection 3.10.2.2.2 - Seismic Qualification by Testing Subsection 3.10.2.2.2 revised to be more consistent 3 with the criteria in Subsection 3.7.3.2..
For the reasons stated in 7 above, this Subsection should be completely revised to be more consistent with the criteria in Subsection 3.7.3.2.
y .
.=
6 9
5
Editorial Comments:
30 a. Subsection 3.9.3.3.1 - MS Safety / Relief Valves Edited accordingly. 1 The revision which was added to this Subst,ction requires some editorial changes (e.g., missing spaces between words, misspelled words, incomplete sentences).
31 b. Subsection 3.9.6.2.1(1) incorporated 1 Part of one sentence in 'N oecond paragraph is missing.
The sentence should read: "The testing of each size, type, and model shall include test data from the manufacturer, field test data for dedication by the COL applicant, empirical data supported by test, of test (such as prototype) of similar valves that support qualification of the required valve where similarity must be justified by technical data."
32 c. Subsection 3.9.7.9 - Benchmark Problems incorporated 1 The references throughout this subsection were changed from 3.9-11 to 3.9-5. They should remain as 3.9-11.
t 33 d. Subsection 20.3.5 - Response to RAI 210.8 Incorporated 1 The last sentence should state: " . need not be classified Quality Group A or Safety Class I, "
34 Subsection 6.1.2.1 Incorporated 1 The SSAR erroneously refers to ANSI 101.4. The correct reference should be ANSI 101.2. '
,Section 3.11
? The staff concludes the tables in Appendix 31is acceptable.
However, the following are discrepancies discovered in Appendix 31 should be corrected:
3 >35 1. The equipment and zones are not clearly identified in the Subsection 31.2.2 clarified. 3 reference figures discussed in sections 31.2.1 and 31.2.2, the zones cannot be determined from given information.
a 6
36 2. The is a typo in Section 31.3.31, the word
- designated
- Incorporated 1 should be " designed".
Incorporated 4 37 3. In Table 31-8, the Gamma dose rate for the heat exchanger is listed as 2, it should be 20.
- 4. In table 31-10, the integrated Gamma dose for the RCW. Incorporated 4 38 pump and heat exchanger should be 2700 or more.
- 5. It is not clear how the Integrated dose for Gamma and Clanfied in Subsection 3.11.5.2 how the integrated 3 39 beta is determined in Tables 31-16,31-17,31-18 and 31- dose for Gamma and Beta is determined.
19.
Same as item 16. 2 40 ECGB identified an unresolved COL Action item in FSER Section 14.1.3.3.5.11 which apparently has not yet been transmitted to GE. During the staff's review of TMl item li.D 1, SSAR Subsection la.3.7," Testing of SRV and Discharge Piping" was added in Amendment 30 at the staff's request. It contained a commitment for COL applicant to confirm that any SRVs or discharge piping not similar to those that were tested in the EPRI generic program will be tested in accordance with NUREG-0737 guidelines. As discussed on FSER Section 14.1.3.3.5.11, this information provided the basis for the resolution of COL Action items 3.9.3.3-1 and 14.1.3.3.5.11-1. In Amendment 3132, Subsection 1 A.3.7 was deleted. Please inform GE that Subsection 1 A.3.7, as written in Amendment 30 should be included in the SSAR.
Incorporated 3 41 in trying to resolve one of Mr. Michelson's concems the Piping DAC, the staff discussed with GE (T. James and M. Herzog) the need to include the following new statement in the Tier 1 Piping Design Description (Chapter 3.3):
" Structures, systems, and components that shall be required to be functional during and following an SSE shall be protected against the effects of spraying, flooding, pressure, and te' perature due to postulated pipe breaks and cracks in stasmic Category I and NNS piping systems."
7.
Chapter 4 42 4.2 Fuel System design incorporated 3 On page 4.2-1, third paragraph GE should revise to state that "each COL app licant referencing the ABWR design may have different fuel and core designs which will be provided by the COL applicant to USNRC for review and approval instead of information.
Chapter 6 43 Table 6.2-7 did not identify which CIV's are locked closed. The P&lD for each system is shown in Table 6.2-7. 2 (FSER Section 6.2.4) The P&lDs identify which CIVs are locked closed.
The staff concludes that the control room habitability systems meet the acceptance criteria of SRP Section 6.4 and are, therefore, acceptable pending satisfactory resolution of the following dyscrepancies:
44 1. SSAR Section 6.4.2.1 and 6.4.2.4 should be revised to state it was agreed that these subsections can remain 2 that the positive pressure is maintained with respect to the " relative to the outdoor atmosphere". (See item 47).
surrounding spaces.
45 2. SSAR Section 6.4.2.3 has dropped the reference to NAA- Since a pressurization test is being performed as part 2 SR-10100 for performing the control room in-leakage of ITAAC, it was concluded that the in-leakage analysis analysis which was in previous SSAR Amendments. in unnecessary.
46 3. SSAR Section 6.4.2.3 has dropped the list of the leak paths Same as item 45. 2 to and from the MCAE and its evaluated effects, as supported by the performed in-leakage analysis, on MCAE to conform with the requirements of GDC 19.
The staff concludes that the SGTS has a removat efficiency of 99% for all forms of radiciodine. The staff further concludes that the system meets the acceptance criteria of SRP Section 6.5.1 and is, therefore, acceptable pending satisfactory resolution of the following discrepancies:
8
47 1. Revise SSAR Section 6.5.1.1.2 and 6.5.1.3.1 to state that it was agreed that these subsections can remain 3 the negative pressurization is maintained relative to the " relative to the outdoor atmosphere" since the surrounding spaces. instrumentation is located outside tne building. There was no change made to Sut,section 6.5.1.3.2, and Subsection 6.5.1.3.1 was modified.
Subsection 6.5.1.3.3 has been revised as requested. 3 48 2. Revise SSAR Section 6.5.1.3.3 to address IE Bulletin 80-03 to state that the charcoal tray and screen will be all welded construction to preclude the potential loss of charcoal from absorber cells per IE Bulletin 80-03.
- 3. Revise SSAR Appendix 6A, Design Criteria (4), Incorporated 3 49 Maintenance, to state that the design is in compliance with this position since the Surveillance Requirements in SSAR Chapter 14 meets the it tent of the Standard Technical Specifications requirements for SGTS and that it is also stated in SSAR Page 6B-1.
- 4. Revise SSAR Appendix 6A, Design Criteria (5), in-Place incorporated 3 50 Testing Criteria, to add reference to ASME N510 in addition to the " Industrial Ventilation" reference for any testing performed.
" Footnote 2" on page 68-65.
52 6. Revise SSAR Appendix 68, Page 6B-9/10 to state "SRP incorporated 1 Table 6.5.1-1" not "STP Table 6.5.1-1" 53 7. Revise SSAR Appendix 6B, Page 6B-2 to state " Operation incorporated 1 of SGTS to mitigate offsite releases will not be affected by the absence of high flow alarm at the MCR.*
- 8. Revise SSAR Section 6.5.1, Table 6.5-1, and Appsndices 6A Two filter trains already reflected. 2 54 and 6B to reflect two filter trains.
Chapter 7 55 Typographical Error in Table 7.5-2 on page 7.5-21. The correct Corrected to be 0-30 Volume. 1 range required for Drywell/Wetwell Hydrogen Concentration should be 0-10 Volume % instead of 0-0 Volume %
Chapter 8 .
9
incorporated. Statement was not intended to be a 3 SS Section 9.5.3.2.3.1 of SSAR Amendment 32 indicates that the Class 1E Associated Emergency Lighting subsystem is - definition. However, to avoid possible confusion, this classified " Associated" because the subsystem's bulbs are not sentence (and a similar are in Subsection 9.5.3.2.2.1) seismically qualified. This definition for associated is not were deleted. Also, the word "However," has been consistent with the definition for associated that is defined in added at the beginning of the sentences following Section 8.3.3.5.1 of SSAR Amendment 32. these two deletions.
incorporated. Subsection 8.3.1.1.6.4 has been 4 57 Section 8.3.1.1.6.4 of SSAR Amendment 32 indicates that the design for protective relays meets positions 7 of RG 1.9. This is' corrected from " position 7" to " position 8". Similarly, true for Rev. 2 of the RG but not Rev. 3. There is no position 7 Paragraph 14,15 and 16 of Subsection 8.3.1.1.8.2 in RG 1.9 Rev. 3. GE in SSAR Amendment 32 revised their have been corrected from "see C.4....* to see Position SSAR to indicate compliance with RG 1.9 Rev.3 from 1.4.. ". Also, titles the were corrected in Tab!e 1.8-20, compliance with Rev.2 of RG 1.9. and Subsections 8.1.3.1.2.2(2) and 8.3.1.2(2)(b).
Incorporated. This sentence is deleted in Subsection 3 58 The last sentence of Section 8.3.4.14 of SSAR Amendment 32 states "Furthermore, annunciation shall be provided to alarm in 8.3.4.14.
the control room whenever the breakers are in for service" is within GE's scope of supply as indicated in Sections 8.3.1.1.1 and 0.3.2.1.0.1 of SSAR Amendment 32.The design for alarming is not within a COL applicant's scope of responsibility as indicated in Section 8.3.4.14.
The first sentence of the 10th paragraph of Section 8.3.3.1 of incorporated. This first portion of the sentence was 3 59 SSAR Amendment 31 and the March 31,1993 draft SSAR inadvertently deleted because it was thought to be states " Associated Class 1E circuits remain with or are redundant to the remaining portion. However, the physically separated in the same manner as those Class 1E complete sentence has been restored to the original circuits with which they are associated;.. " was deleted from sentence intact.
Amendment 32.
Amendment 31 was consistent with the guidelines of Section 5.5.2 of IEEE 384 and position 4 of RG 1.75. With the deletion of this sentence in Amendment 32, associated circuits which do not have isolation device such as lighting circuits are no longer explicitly addressed in the SSAR, the design description in the SSAR is now inconsistent with the commitment to IEEE 384 buidelines, the deletion may be inconsistent with the staff's safety evaluation report conclusions.
10
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60 Section 8.3.4.21 of SSAR amendment 32 should be revised to The technical specifications already require periodic 3 explicitly state, consistent with other SSAR sections, that the testing of the diesel genc.ator loading capabilities, as COL applicant shall be required to provide appropriate plant stated in Subsection 8.3.4.21. Therefore, no additional procedures for periodic testing of the diesel generator loading plant operating procedures are necessary for this part capabilities and the interlocks which restore the DGs to standby of the comment. However, the second sentence in the event of a LOCA or LOPP. regarding testing of the interlocks has been modified as follows: ' Appropriate plant procedures shall be provided for periodic testing of the interlocks which restore the units to emergency standby on event of a LOCA or LOPP.*
61 Section 8.3.4.17 of SSAR amendment 32 which addresses Subsection 8.3.4.17 does not address regulatory 2 inclusion of regulatory codes and standards in purchase codes and standards. GE agrees that regulatory codes specifications may not be appropriate as a COL action item as and standards are GE responsibility, but these are specified in the SSAR. Specifying which regulatory codes and already addressed in the SSAR text in accordance with standards should be used to meet ABWR plant design the SRP. Rather, Subsection 8.3.4.17 provides a listing requirements is a GE responsibility. Assuring their inclusion in of common industrial standards to be included in the purchase specifications also appears to be within GEs scope of purchase specifications, which would be in addition to responsibility. The SSAR should be revised to indicate that it is the regulatory codes and standards. The content of a GE responsibility to specify which regulatory codes and this section was added in response to a previous NRC standards should be used for the purchase of equipment. This request; not as a licensing issue, but for quality inconsistency affects SER findings addressed in Section assurance purposes.
8.3.6.1 of the SER.
62 The use of the word ' redundant
- In Section 8.3.3.6.2.2 of Same as item 63 2 SSAR amendment 32 and in the 2nd paragraph of Section 8.3.3.6.2.2.3 of SSAR amendment 32 incorrectly implies that safety related equipment need not be protected from design basis events if the event or missile only affects one of two redundant systems. The term
- redundant
- as used in these sections should be deleted.
63 The use of the word
- redundant"in Section 8.3.3.6.2.2 of SSAR incorporated. The three references have been 3 amendment 32 and in the 2nd paragraph of Section resolved as follows:
8.3.3.6.2.2.3 of SSAR amendment 32 incorrectly implies that safety related equipment need not be protected from design Subsections 8.3.3.6.2.2 and 8.3.3.6.1.1: The word basis events if the event or missile affects one or two redundant
- redundant" was deleted from the two referenced systems. The term
- redundant" as used in these section should sentences.
be deleted. Similarly, the term
- redundant" has been incorrectly used in item 2 of Section 8.3.3.6.1.1 of SSAR amendment 32. Subsection 8.3.3.6.2.2.3: The word
- redundant" was replaced with
- Class 1E*, so the sentence reads
..could jeopardize Class 1E cabinets and raceways.'
11
.~ .~ __ ..
incorporated. 3 64 The references 13.6.3 in Section 8.3.4.19 of SSAR amendment 32 should be 8.3.3.6.1.1 (5).
Inserts at the end of the fifth paragraph of Subsection 3 65 GE made a SSAR change due to ITAAC involving de power supplies for the offsite circuits. Amendment 32 did not provide 8.2.3, and at second-to-last sentence in Paragraph (5) the bases to justify this change. This impacted a number of our of Subsection 8.2.3:"The instrumentation and control safety findings regarding independence requirements of GDC circuits for the normal and alternate preferred power
- 17. We have revised three FSER conclusions related this area shall not rely on a single common DC power source to state that this aspect is now open. [See Subsection 8.2.3 items (13) and (15)].
Incorporated. The last sentence of the 7th paragraph 3 66 By SSAR Amendment 32, GE changed their design to specify that the l&C circuits at their de power sources are routed in has been deleted, since the ' preferred
- version of this separate raceways separated to the extent practical versus are information is already contained in the 5th paragraph.
separated by floor, wall, or 50 feet at their power supplies. The remaining first sentence of the 7th paragraph has been moved just ahead of the 5th paragraph, because By draft SSAR 10/12/93, GE further revised Amendment 32 to the commitment for the existence of the interlocks indicate that the instrumentation and control circuits for the should proceed the statement that the interlocks are normal and alternate preferred power shall not rely on a single separated.
common de power source.
Based on these changes to the SSAR, the ABWR design will now permit sharing of de power sources between offsite circuits.
If the two offsite circuits share l'uo or more common de sources, by implication the I&C circuits for the independent offsite sources are interconnected.
The 7th paragraph of SSAR Amendment 32 states that the feeder circuit breakers from the unit auxiliary and resente auxiliary transformers to the medium voltage switchgear are interlocked to prevent paralleling the normal and attemate power sources. With the exception of these interlocks, there are no electrical interconnections between the instrument and control circuits associated with the normal preferred circuits.
This statement in the SSAR is not consistent with the de source interconnection defined above 12
Chapter 9 9.5.1.1 General Evaluation Fire Protection System in Amendment XX GE indicates that they meet the design commitments as specified in the Branch Technical Position CMEB 9.5-1 except in four cases. GE identified the following deviations to the Branch Technical Position:
67 1. Deviation from BTP CMEB 9.5-1 Section 7.j, 3 Diesel Fuel Storage Areas.
The staff finds GE's justification for having the diesel fuel oil Justification provided under Subsection 9.5.1, new day tanks inside the reactor building is acceptable pending item (1).
satisfactory resolution of the following discrepancy:
- Provide capacity in the fuel oil tank rooms to contain total contents of diesel fuel oil day tank and discharge from two fire hoses operating for 1/2 hour.
- 2. Deviation from BTP CMEB 9.51, Section 7.1, Diesel Generator Area.
The staff finds GE's justification acceptable pending satisfactory resolution of the following discrepancies:
68 a. GE is to provide information to demonstrate the See item 69. 3 adequacy of the foam system utilizing closed heads.
69 b. .Should GE change it's design to an open head system, Justification provided under Subsection 19B.2.36. Item 3 then the resolution of GI-57 will need to be revisited. 68 is related to this item and justification for item 69 covers this.
70 c. As discussed with the applicant in a meeting held on it was agreed that it was sufficient for the diked area to 3 September 21,1993, the diked area in the DG room be capable of containing 100% capacity of the tank is to be designed to the appropriate section of NFPA and 1/2 hour of water application from the automatic
- 15. The diked area is to be capable of containing foam sprinkler (2 manual hose stations not required).
100% capacity of the tank and 1/2 hour of water application from the automatic foam sprinkler system and 2 manual hose stations.
13-
- 3. Deviation from BTP CMEB 9.5-1, Section 13, Control Justitication provided under Subsection 9.5.1, new 3 71 Room Complex. item (2).
The staff finds GE's justification acceptable pending l satisfactory resolution of the following discrepancy:
- The applicant is also to provide the rationale for lack of suppression and drainage in the control room subfloor.
- 4. Deviation from BTP CMEB 9.5-1, Section 13, Outdoor Subsection 9A.4.3.2.1 and 9A.4.6 specify that the wall 3 72 transformers. separating the turbine from the transformers will be masonary and rated for h least one hour. NFPA 15 has The staff finds GE's justification is acceptable pending been added to the list of codes and standards, satisfactory resolution of the following discrepancy:
- Specify that the wall separating the turbine from the transformers will be masonry and rated for at least one hour.
Diking will be provided as described in NFPA 15.
9.5.1.3.4 Automatic Foam Fire Suppression Systems GE committed to meet the design aspects of GDC 3, Branch Technical Position 9.5-1 and Generic issue 57, therefore, the staff concluded that the automatic foam fire suppression systems are acceptable pending satisfactory resolution of the following discrepancies:
- 1. GE proposes to utilize closed head sprinklers for the foam Repeat of item 68 2 73 system and has not adequately demonstrated its acceptability to control and extinguish a fire. GE is to provide the technical justification to demonstrate the adequacy of the foam system.
- 2. Should GE change it's design to an open head system, Repeat of item 69 2 74 then the resolution of GI-57 will need to be revisited.
- 3. As discussed with the applicant in a meeting held on Reat of item 70 2 75 September 21,1993, the diked area is to be designed to the appropriate section of NFPA 15. The diked area is to be capable of cc,ntaining 100% capacity of the tank and 1/2 hour of water application from the automatic foam sprinkler system and 2 manual hose stations.
14
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9.4.1.1 Control Room Habitability Area Heating, Venting, and Air Conditioning system (CRHA HVACS)
GE has addressed IE Bulletin 80-03 compliance by providing future SSAR Amendment , which revises SSAR Section 9.4.1.1.4 to state that the charcoal tray and screen will be all welded construction to preclude the potentialloss of charcoal from adsorber cell per IE Bulletin 80-03. Therefore, the emergency air filtration system of the CHRA HVAC system precludes the potentialloss of charcoal from adsorber cells.
By SSAR amendments up to including Amendment ,GE provided the SSAR Appendices 9C and 9D and has provided acceptable justifications for the deviations.
By SSAR Amendment , GE revised SSAR Sections 9.4.1.1.4 and 9.1.1.1.5, stating that the unfiltered inleakage is controlled by the use of welded ducts, except galvanized steel is used for outdoor air intake and exhaust, and unfiltered in-leakage testing will be performed periodically on all system ductwork outside MCAE in accordance with ASME N510, respectively.
By amendments up to and including Amendment ,GE revised the SSAR Section 9.4.1.1 and Table 9.4-4d to include electric heaters in the ESF filter trains.
The staff concludes that the system is nccertWe pending satisfactory resolutions of the followin; ' spancies:
76 1. Revise SSAR Sections 9.4.1.1.3,9.4.1.1.4, and 9.4.1.1.6 It was agreed that these subsections can remain 2 to state that the positive pressurization is maintained " relative to outdoor atmosphere". (See item 47) relative to the surrounding spaces.
77 2. Revise SSAR Section 9.4.1.1.4 to state that the charcoal incorporated 3 tray and screen will be all welded construction to preclude the potential loss of charcoal from adsorber cells per IE Bulletin 80-03.
78 3. SSAR Table 9.4-4 shows heating coil data in kcal/hr for incorporated 3 each CRHA HVAC Division. Revise SSAR Table 9.4-4 to provide the electric heater ratings in kW.
15
. __ __ _ _ , - - _ . -_ ~ _ _ . __ -._ _. - . _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
.m 79 4. The deleted SSAR Table 9.4-4d listed each division's Table 9.4-4 updated to include alt heaters. 3 electric heater capacity as 165 kW for MCR HVAC Divisions A and B emergency filtration units. Revise SSAR Section 9.4.1.1 to restore the above component data either in the tabulated form or in the SSAR text.
80 5. Revise SSAR Section 9.4.1.2.6 to state that tests will be incorporated 3 performed at a test facility to verify that the CRHA HVAC system fire dampers with fusible links close under anticipated air flow conditions.
81 6. Revise SSAR Appendix 9C, Section 9C.1.(4).(d) to state incorporated 1 that the design is in compliance with this position since SR 3.4.3.1 SSAR Chapter 14 meets the intent of the Standard Technical Specifications requirements. Also, revise SSAR Section 9.4.1.1.7 to delete "except as noted in Appendix 9C."
82 7. Revise SSAR Appendix 9D to provide summation of Incorporated 3 pressure drop across the entire system as stated in the SSAR Section 9.4.1.1.6. Also, ASME N509 " Footnote 2*
should be " Footnote 3" on Page 9D-5. ,
83 8. Revise SSAR Section 9.4.1.1.4, stating that the unfiltered Inspection added to verify integrity of system. 3 infeakage is controlled by the use of "All welded black steel ducts except galvanized steel used for outdoor air intake and exhaust".
84 9. Revise SSAR Section 9.1.1.1.5, stating that "The unfiltered See item 83. 3 infeakage testing will be performed periodically on all system ductwork outside MCAE in accordance with ASME N510", as agreed upon with GE for the resolution of USl B-66, Control Room Infiltration Measurements.
85 10. Revise SSAR Figure 9.4-1, Sheet 2 of 5 to reflect an incorporated 4 independent and separate discharge to MCAE and retum i from MCAE to each emergency filtration unit, as shown in
( Sheet 1 of 5. Also, revise SSAR text to include this.
l l 16 l '
86 11. Revise SSAR section 9.4.1.1.5 to state that "The charcoal No change, already in Amendmerd 32 2 filters will be tested with an acceptable gas for bypasses."
9.4.1.2 Control Building Safety-Related Equipment AREA (CBSREA) HVAC System The staff concludes that the CBSREA HVAC system is acceptable pending satisfactory resolution of the following discrepancies:
87 1. SSAR Section 9.4.1.2.3 states that there is an electric Heater deleted from Subsection 9.4.1.2.3. 4 heater for each of the CBSREA HVAC subsystems.
However, SSAR Table 9.4-4d showing the electric heater capacity has been deleted.
88 2. Provide rationale for maintaining a minimum temperature Rationale provided to staff. 2 of 10cc in the winter.
89 3. Reconcile the differences between ITAAC Figures it was agreed to provide this information with 4 2.15.5b,2.15.5c and 2.15.5d and SSAR Section metrification (see item 4).
9.4.1.2.3 and SSAR Figures 9.4-1 Sheets 3,4, and 5 conceming the descriptions of the areas served.
90 4. Revise SSAR Section 9.4.1.2.6 to state that the test will Repeat of item 80. 3 be performed at test facility to verify that the CHRA HVAC system fire dampers with fusible links in HVAC ductwork are capable of closing under anticipated air flow conditions.
9.4.4 Turbine Island HVAC System The staff concludes that the turbine island HVAC system meets the applicable acceptance criteria of SRP Section 9.4.4 and is, therefore, acceptable pending resolution of the following discrepancies:
91 1. Revise titled captions of SSAR Tables 9.4-5 and 9.4-Sa incorporated 4 through 9.4-5c to confirm with SSAR Section 9.4.4.
Reconcile SSAR section 9.4.4.2.1.5 areas with the areas shown in above tables.
92 2. Provide Capacity for the cooling coils serving SJAE A 99,800 kcal/hr capacity provided. 3 area recirculation unit air handler in SSAR Table 9.4-5b.
17
93 3. Verify the capacity of cooling coils serving demineralizer Verified 2 pump and valves area recirculation unit air handler in
.SSAR Table 9.4-Sa.
94 4. Revise Design Description in ITAAC Section 2.15.5 the incorporated. See ITAAC submittal. 4 turbine building (T/B) HVAC system to state "T/B lube oil area exhaust system with two fans." ,
9.4.5.1 R/B Secondary Containment HVAC System The staff concludes that the system complies with applicable SRP Section 9.4.5 acceptance criteria, and, therefore, is acceptable pending resolution of the following discrepancies:
95 1. Section 9.4.5.1.1.2 should replace the words "outside it was agreed that this subsection can remain " relative 2 atmosphere" by the words " surrounding spaces" in to outdoor atmosphere". (See item 47).
relation to the negative pressure of the secondary containment. ITAAC Table 2.15.5 should also be corrected to use the words " surrounding spaces".
96 2. Table 9.4-4g should show that the exhaust f an flow rate is incorporated 3 higher than the supply f an flow rate to ensure that the secondary containment is at a negative pressure with respect to surrounding spaces.
97 3. Table 9.4-4 filter capacity data for secondary containment incorporated 4 exhaust should match with the exhaust fan capacity.
98 4. SSAR Section 9.4.5.1 should state that fire dampers with incorporated 3 fusible links in the HVAC duct work are capable of closing under anticipated air flow conditions (ITAAC items).
l 9.4.5.2 R/B Safety-Related Equipment HVAC System The staff concludes that the R/B safety-related equipment HVAC system is acceptable pending the resolution of the i
following discrepancies:
99 1. All the FCUs are automatkally initiated upon secondary Text modified to match P&lD. 4
, containment exhaust fan felure also since such a failure L will result in the R/B secondary containment HVAC l
system isolation.
l .
18
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- w v. -, ,,,r, 4= - c _ _. ,
100 2. Section 9.4.5, item 2 should delete the words " secondary incorporated 1 containment" from the title of the HVAC system.
101 3. Section 9.4.5.2.2.1 should state that the FCUs will be Already included in Amendment 32 2 sized to maintain the operational temperature of the subject rooms within 40*C, 9.4.5.3 R/B Non-Safety-Related Equipment HVAC System, R/B Mainsteam Tunnel HVAC System, and R/B RIP Power Supply Panel Room HVAC System '
The staff concludes that the R/B non-safety-related equipment HVAC system, R/B main steam tunnel HVAC system, and R/B RIP power supply panel room HVAC system meet the applicablo acceptance criteria of SRP Section 9.4.5 and are, therefore, acceptable pending the resolution of the following discrepancies:
102 1. Section 9.4.5, item 3 should read as "R/B Non-safety Incorporated 1 Related Equipment HVAC System'.
103 2. Section 9.4.5, item 3 and Section 9.4.5.8 should read as Clarification made in text. 3 +
' Reactor Intemal Pump Power Supply Panel HVAC System". SSAR Figure 9.4-5 shows that a closed cooling loop HVAC system cools the RIP power supply panels and not RIP ASD control panel rooms .
Furthermore, RIP ASD control panel rooms are served by the safety related R/B electrical equipment HVAC system.
- 3. Section 9.4.5.3.2 and Figure 9.4-3, do not match with See item 89. 4 104 respect to rooms for which FCUs are provided. GE should correct as appropriate so that the.same names j are used to identify the rooms both in the figure and the !
section.
105 4. The SSAR tables do not list the quantity and capacity of incorporated 3 all equipment for all the 10 rooms serviced by the R/B non-safety-related HVAC system (for example, fans and cooling coils for the 10 rooms are not listed).
S 19
_ _ __ - _ . . . _ . . . _ . - .._ _ . . _ _ . _ . . .. _ _ ., _ . _ _ _ _ _ _ . . _ , ~ . _ . . _ _ _ _ . . _ _ . _ . . . _ . _ . . . _ . _ _ _
106 5. Table 9.4 4h refers to filters for RIP ASD control panels. It is not an error. 2 This should be deleted if it is an error. Equipment listing should be given for the R/B RIP power supply panel i i
HVAC system.
9.4.5.4 R/B Safety-Related Electrical Equipment HVAC System The staff concludes that the R/B safety-related electrical equipment HVAC system complies with applicable GDC referenced in of SRP Section 9.4.5 and, therefore, is acceptable pending the resolution as the following discrepancies:
107 1. Section 9.4.5.4.2 sentence: "The divisional Clarification provided. 3 rooms.... control panel rooms
- is confusing and should be deleted.
Clarification provided. 4 108 2. Item 8 listed in the above section should be deleted. This is because as per Figure 9.4-5, the non-safety-related R/B RIP ASD power supply panel HVAC system takes care of the cooling needs of the power supply panel rooms.
109 3. Item 3 should be re-captioned as RIP ASD control panel Clarification provided. 3 rooms, Divisions B and C. GE should check whether HVAC Divisions A and B serve these control panel rooms since RCW Divisions A and B serve the RIP room coolers.
110 4. The system capability to maintain the rooms other than Provided in Amendment 32. 2 the DG engine rooms below 40*C identified in the ITAAC should be included in the SSAR section.
111 5. GE should explain why electric heaters needed to assure Justification provided to staff. 2 that the temperature in the subject rooms do not dip below 10 C are deleted in Amendment 32.
112 6. SSAR should state that the system has fire dampers with Repeat of item 80. 3 fusible links in the HVAC ductwork which are capable of closing under anticipated air flow conditions (ITAAC Information).
20
__________,_m.._ _ _ ___..-r.r .._,_s ,-c.--- ., ,<~+",%.- ..e r-,.... . ...- .-.w, v4-.. .. . , , , , . - a m,=- % , m i. s.... , 6_.m ..- . , - - . _ , - , , , . . . , y, .
113 7. SSAR should state that Division B of the HVAC system Sheet 2 revised. 4 serves electrical equipment rooms, Division 11 and IV and Figure 9.4-4 Sheet 2 should be revised to reflect the above.
9.4.8 Service Building Ventilation System The staff concludes that the service building ventilation system meets the applicable acceptance criteria of SRP Section 9.4.3 and is, therefore, acceptable pending the resolution of the following discrepancies:
114 1. Like the ITAAC, SSAR should identify 2 HVAC systems: SSAR modified to match wording of Sect %n 11.5, 3 TSC HVAC system and r~ntrolled area HVAC system. ITAAC modified to be consistant.
Staff prefers Section 11.E. '.4 language,i.e., "
controlled area HVAC system'. (ITAAC which says that one of the SB HVAC system is SB HVAC system should be corrected.) Staff has used the above wording in the write-up above.
115 2. Section 9.4.8.1.1 should be corrected since the TSC incorporated 3 HVAC system operates during a high radiation mode in addition to operating during normal operation.
116 a. High radiation mode of operation for the TSC HVAC _ incorporated 3 system.
117 b. Location of both the HVAC systems (ITAAC should incorporated 3 identify the location of the controlled area HVAC system).
118 c. Supply fan and ACU for the controlled area HVAC - ACU does not exist. 2 system.
119 d. Toxic gas protection for applicable COL applicants incorporated 3 (GE should provide COL license information).
21
Incorporated 3 120 e. Provision of 2 recirculation fans for the TSC HVAC rgteri,.
l 4. SSAR Section 9.4.8 should be revised to include the l followine-TSC and OSC added. 3
!. 121 a. Which areas are the ' clean areas".
incorporated 3 122 b. Provision of a radiation monitor in the outside air intake for the TSC HVAC system.
it was agreed that the COL applicant will provide. 2 123 c. The components of ACU (i.e. heating coil and cooling coil) for the controlled area HVAC system)>
2 124 d. Cooling and Heating sources for the ACUS in both the it was agreed that the COL applicant will provide.
HVAC systems.
- e. Common air intake for both the HVAC systems. Incorporated 3 125 it was agreed that this subsection can remain " relative 2 126 5. Both the ITAAC and Section 9.4.8.1.2 should state that the TSC and clean areas are maintained at a positive to outdoor atmosphere". (See item 47).
pressure with respect to surrounding spaces.
CHAPTER 11 The value of "1" reldtred to in the comment is shown in 2 127 ABWR SSAR Table 11.1-6 gives the fraction of steam activity treated by the condensate demineralizer as 1. This is column 2 of Table 11.1-6 and is for the ANS 18-1 inconsistent with the design flow rate of 1022 Cu. Meter / hour " Reference Plant" as defined in ANS 18-1. Table 1, per condensate polisher vessel given in SSAR Table 10.4-4. column 4 which is a non-pumped forward plant and not There are 6 such v.' 4els one of which is standby. The design the ABWR. As noted in the comment, the ABWR flow rate through all five vessels corresponds to 0.67 of the total values which are given in the final row of Table 11.1-7 steam activity being treated by the condensate demineralizer. are correctly indicative of a pumped forward plant .
This is not un-comt,on, since most reactors have forward Therefore, Table 11.1-7 correctly indicates the values pumping. Also, the value of 0.67 agrees with the fractions 0.18 used as indicated by the asterisk in Table 11.6-6.
and 0.01 of steam activity of lodines and others treated by the condensate demineralizer given in SSAR Table 11.1-7.
For the above reasons, the staff requires GE to correct the i
subject entry from 1 to 0.67 in the SSAR Table 11.1-6.
c .
22 -
-'e-a- ,,c< =e- m w.< wee em ener. rw -. '= --mr- ---,r'm=+%== v no- - * .-=- ,-- -g-c-e - - - - + ---*t-+--4-e a- -- ---= ei----- -e -
+i---ere
CHAPTER 12 128 1. Page 12.3-10: delete the first line on the page,11 is repeated incorporated. However, the page change (even though 1 from the last line on page 12.3-9. Also change the Amend. it was an oversight), requires the page to go from no. back to 31 except for the first line, Amend. 32 does not Amendment 32 to Amendment 33.
appear to haw changed this page.
129 2. Page 12.3-Urie 11 from bottom: the last word should be incorporated 1 "RWPs" not " raps".
130 3. Page 12.3-19,line 7 from bottom: line should start 'the TIP incorporated 1 spoolers" not "the TIP spoilers".
131 4. Page 12.3-22: sub-section 12.3.3.1(2) should reference the incorporated 3 DAC Tabte 3.2(b). Suggest revising the penultimate sentence in this sub-section to state, *DAC Table 3.2(b) requires the COL Applicant to perform calculations for the -
expected airborne radionuclide concentrations to verify the adequacy of the ventilation system during the ITAAC stage of plant construction."
132 5. Figure 12.3-43: figure is missing the radiation zone Radiation zone designations added. 4 designations.
CHAPTER 14 Preoperational Test Program 133 in SSAR Section 14.2.3, Test Procedures, the last sentence incorpo,ated 3 should change the word power ascension tests to startup tests to make the sentence more correct with respect to the requirements of RG 1.68 which states that test procedures will be provided to the NRC 60 days before their intended use *or preoperational tests and 60 days before fuel loading for stattop tests (i.e., not power ascension tests).
23
y incorporated 3 134 in Section SSAR 14.2.10.2,2nd sentence, GE states that 'the procedure controlling this movement will specify that shutdown margin and subcritical checks be made at predetermined intervals throughout the loading, thus ensuring safe loading increments." To clarify this sentence, GE should revise this sentence to state "the procedure controlling this movement will specify that partial core shutdown margin demonstration and sub critical checks be made at predetermined intervals throughout the loading, thus ensuring safe loading increments as described in startup test abstraci 14.2.12.2.3, Fuel Loading.
incorporated 3 135 in Section SSAR 14.2.10.3.,1st sentence, GE should delete the first sentence from this section and insert this sentence at the beginning of the paragraph in section 14.2.10.4. The sentence currently states
- Prior to initial criticality, the shutdown margin shall be verified for the fully loaded core. The sentence should be revised to state, "During initial criticality, the full core shutdown margin shall be verified for the fully loaoed core as described in startup test abstract 14.2.12.2.4, Fuli Core Shutdown Margin Demonstration.
Iest /tSttact 14.2.12.1.0. RHR System Prone ational Test 136 The requirements of RG 1.139, Guidance for Residual Heat Removal, Position C.3, state that "to protect the RHR system against accidental over pressurization when it is in operation (not isolated from RCS), pressure relief in the RHR system should be provided with relieving capacity in accordance with the ASME boiler and pressure vessel code."
Test Abstract 14.2.12.1.8, Acceptance Criteria (3)(c), should incorporated 3 state " proper operation of system relief valves including timing, '
position indication, controlling function (if any for air operated valves), and verification of requirements" to meet the requirements of RG 1.139.
24 -
-_.-----o2 -- w e m e- -e m r
-n v - ww>m -m a-u-- e w ,- ,, , y
h PleppEal!QDaLT.es1Atzstract 14.2.12.1.41 '
137 The statf's review identified that preoperational test abstract incorporated 3 14.2.12.1.41, " Pressure Suppression Containment Bypass Leakage Tests" removed reference to Subsection 6.2.6.2 for the applicable test procedures. Reference to subsection 6.2.1.1.5 was added for a description of and criteria for the test method.
The acceptance criteria for the test method. The test abstract should be revised testing method and the acceptance criteria.
138 GE revised Section 6.2.1.1.5 in a markup dated September 30, incorporated 3 1993 to state that "the acceptance criteria for both the high and low pressure leakage tests shall be a measured bypass leakage area which is less than 10% of the suppression pool steam bypass capability specified in subsection 6.2.1.1.5.4 (i.e.,50 cm squared). GE also revised this section to discuss specific types of .tigh and low pressure drywell to wetwell leakage tests that will be performed. GE plans to and cross reference 6.2.6.
into test methods that would similarly be performed on the drywell for the high and low pressure suppression pol bypass leakage tests. The staff finds the incorporated into a future Chapter 14 SSAR amendment and the above changes to chapter 6 are incorporated in a future Chapter 6 SSAR amendment.
PreoptrationaLIRSIAbstract 14.2.12143 GE removed much prerequisite and acceptance criteria information from Test Abstract 14.2.12.1.43, Wetwell to Drywell Vacuum Breaker System (WDVBS) Preoperational Test. GE removed prerequisite (2)(d), which requires approximate power '
sources are available for use to supply electrical power to all .
instrumentation. GE also removed the following acceptance criteria: (1) parts of acceptance criteria (3)(a), for testing of the system logic and timing features for proper operation of vacuum breaker valves, (2) parts of acceptance criteria (3) (b),
verification on the operability conditions of instrumentation and alarms used to monitor WDVBS during loss of preferred power conditions, and (3) acceptance criteria (3)(d), " proper functioning of vacuum breakers test features" I
25
139 After Further review of Section 6.2.1.1.4, it was noted that the Clarification for the deletions have been provided to 3 vacuum relief breaker vaives are swing check valves which the staff. In addition, prerequisite (2)(d) pertaining to
, open passively due to negative pressure across the valve disk electrical power has been reinstated, Further requiring no power source. Acceptance criteria for testing the clarification is contained in Subsection 6.2.11.5.8.1.
system logic and timing feature are not needed for swing check valves (i.e., only required for MOVs). Acceptance criteria for vacuum breaker test features in Section 6.2.1.1.4. The prerequisite on the required more discussion is needed with GE to clarify the exact reason for all of the above deletions to this test abstract.
The staff finds the above changes to the test abstract acceptable assuming the staffs interpretation of the information provided in Chapter 6 is correct as noted below.
(1) Prerequisite (a) includes instrumentation used to monitor system and component parameters needed in this test is energized for the conduct of the test. (2) proper operation of any/allinstrumentation under loss of offsite power is tested per preoperational test 14.2.12.1.45, " Electrical Systems Prooperational Test".
ErespntelinDal test At1stmet 14.2J2.1.52 140 The staff's review identified the following typographical error on incorporated 1 page 14.2-94. Subsection (3) (b), fist paragraph, last sentence uses "my" which should be "may".
Startup Test Program 141 in SSAR section 14.2.12.2, General Discussion of Startup Incorporated 3 Tests, GE should add a sentence which states that "startop test procedures will be provided to the NRC 60 days before fuel loading" to be consistent with similar statements in Sections 14.2.3, Test Procedures, and Section 14.2.12.1, Preoperational Test Procedures.
Starluo Test 14.2.12.1A1 142 The staff's review identified that startup test 14.2.12.2.41 This startup test was deleted in Amendment 32 2 indicates this test is deleted, however, review of prior amendments indicates that startup test 14.2.12.2.41 has never existed. This test should be removed or explanation provided for the test that is being deleted.
26-
_,_ __ ..w.. - , _ . . , - - - w -< - . .. - _,s , _ 4 _ _____ _.____.i_______. _ . _ _ _ _ _ _
n 1
ESAR Sccilon_112,13_C.0L LiCEDse Information 143 in SSAR Section 14.2.13.1, first sentence, GE should delete the incorporated 3 words " site specific" and replace them with the words " COL applicant supplied". The words site specific is heavily used in the definition of interfacing systems. This will avoid any confusion with the 4 systems listed in this section as being interfacing systems since all of the listed systems are not interfacing systems.
144 in SSAR Section 14.2.13.2. Test Procedures /Startup incorporated 3 Administrative Manual, states that the COL applicant will provide the following to the NRC, item (4) "the approved preoperational and startup test procedures approximately 60 days before their intended use (Subsection 14.2.3)." Item 4 should be revised to state that " the approved preoperational test procedures approximately 60 days before their intended use and the startup test procedures approximately 60 days before fuelloading."
Table 14.2-1, Power Ascension Test Matrix 145 Table 14.2-1 should be renamed the Startup Test Matrix instead incorporated 3 of the Power Ascension Test Matrix. l 146 The requirement of RG 1.68, Appendix A. Position 5.c.c state
- demonstrate that gaseous and liquid radwaste processing, .
storage and release systems operate in accordance with -
design." Based on the staff's review of SSAR Amendment 23, the staff requested GE to revise Test Abstract 14.2.12.2.38 and Table 14.2-1 to include the Gaseous Radwaste System as part-
of the Gaseous and Liquid Radwaste Systems Performance Test to meet the intent of RG 1.68. GE revised the test abstract and the table in SSAR Amendment 30.
27
I I After further review of Table 14.2-1, Test Abstracts 14.2.12.2.1, Chemical and Radiochemical Measurements, 14.2.12.2.35, !
Offgas System Test, and 14.2.12.2.3 8, Gaseous and Liquid i Radwaste System Performance Test, and GE Proprietary Sections 11.2, Liquid Radwaste Management and 11.3, Gaseous Radwaste Management System, the staff concluded that the Gaseous Radwaste System is the Offgas System; therefore this system is adequately discussed in test abstract 14.2.12.2.1, Ch Neal and Radiochemical Measurements. The test abstract discusses measurement testing of the release effluents but not radwaste processing and storage testing per the requirements of RG 1.68; therefore, the Liquid Radwaste '
system portion of the test is not an optional test as currently described in Table 14.2-1.
Based on the above, the staff requests GE to make the Incorporated 3 ,
following changes to SSAR Section 14.2 in Amendment 33.
Test Abstract 14.2.12.2.38 should be renamed the Liquid Radwaste Management System Performance and Test Abstract 14.2.12.2.35 should be renamed the Gaseous Radwaste Management /Offgas System Performance per the titles used in Section 11.2 and 11.3. The description and acceptance criteria sections of Test Abstract 14.2.12.2.38 should delete all references to the gaseous radioactive waste system and Section 11.3. Additionally, GE should delete the 3rd through the 6th sentence in the description section. In table 14.2-1, Page 198, GE should substitute the " Gaseous Radwaste Management /Offgas System Performance" test for the "Offgas I
System Perfonnance" test and substitute Liquid Radwaste management Performance test for the " Gaseous and Liquid Radwaste Systems Performance" test. GE should also delete the word optional for the Liquid Radwaste System Performance test in the table.
147 GE needs to verify that the page numbers are correct for all Page numbers verified. 1 pages in Amendment 32. Page numbers were not properly changed from Amendment 31 to Amendment 32 when Section 14.2.13.3 was deleted.-
28
- -n-m .r-...,---ww wi., , ...---..~o: . . ~ . - - , -we = e +,n.--vs-- . w3w. .*>< . . - - . . . ~- ~% -vw :- . ., .,--.iw- - . * . . - - - --_m .- -._
a 9
148 in Table 1.9, item No.14.3, Tests Exempt From License item No.14.3 deleted. 1 Conditions, Subsection 14.2.13.3. Page 1.9-10, should be deleted since Subsection 14.2.13.3 no longer exists in Chapter The following generic comments are provided on problems with GE's use of SI units.
149 Page 15-1, Units for vessel pressure are expressed in See item 4. 1 kg/cm"2*d. Since these are SI units, pressure is expressed in pascals or force per unit area (N/m"2) or kg*m/s"2/m"2 or kg/m's"2. The above expressed units don 1 seem to make sense. d id defined as differential pressure? Even with d defined in this manner, the units don't match. GE seems to be mixing up SI units with english units. English units would be expressed as 1bf/ft"2.
150 Page 164, Same as above. Units for reactor pressure are See item 4. 1 expressed as kg/cm"2 *g.Where g is defined as gauge pressure. Reactor pressure is usually expressed in pascats or force per unit area which breaks down like above to kg/m's"2.
With g defined in this manner, the units don't match.
151 Page 165,2nd paragraph, Same problem. Incorrect units for See item 4. I reactor pressure given as 10.5kg/cm**2 g, where g is defined as gauge pressure. Pressure is force per unit area which is N/m"2.
152 Page 180,1st paragraph. Again same problem with units. See item 4. 1 Vessel dome pressure expressed as 1.76kg/cm"2 d, where d is defined as differential pressure. Correct units are N/m"2.
153 Page 182.1st paragraph, Same problem. Vessel dome See item 4. 1 pressure expressed as 1.76kg/cm"2 d, where d is defined as differential pressure. Correct units are N/m"2.
EQUIPMENT SURVIVABILITY OUESTIONS (SCSB) 154 1. Table 7.5-2 Suppression pool water level only measures 1.5 Instrument ranges changed to permit the EOPs to 4 m above normal water !evel. Bottom of reactor vessel is 6.1 terminate containment flooding.
. m above normallevel and COPS is even higher. How does this effect EOPs to terminate containment flooding?
29-
155 2. Table 7.5-2 Drywell atmosphere temperature only measures Correct. Modified Section 7.5 to reflect required ranges 4 up to 110 C as opposed to Reg. Guide 1.97 of 227 C. DBA of R.G.1.97 temperature reaches over 120 C see figures 6.2-7,8,15.
156 3. Table 7.5-2 Hydrogen concentration measures 0% as Same as item 55. 2 opposed to Reg. Guide 1.97 which indicates 30%.
157 4. Table 7.5-2 Oxygen concentration cross-reference does not Reference incorrect. Changed to Subsection 4 exist. 7.5.2.1(2)(k).
158 5. Table 7.5-2 and test Suppression pool water temperature Subsection 7.5.2.1(2)(1) revised accordingly. 4 indicates 4 divisions with deviations and this is similar to drywell atmosphere temperature with 2 divisions (Reference supp. pool temp. for acceptability).
159 6. Fig. 6.2-13 Graph indicates temp. of 1767.7C. Typographical error. Changed to 176.7"C. 1 160 7. Figs. 6.2-17 and 18 are identical. No. Look at curves past 70 seconds. Figure 6.2-17 2 shows difference betwee drywell and reactor building.
Figure 6.2-18 shows diffe. ace between wetweil and reactor Luilding.
- 8. 7.5.2.1(2)(b) Rupture disks actuate at pressure of 70 psig. They actuate at 90 psig (6.3 Kg/cm2g). The instrument 3 I
161 l range provides a margin of greater than 10%.
162 9. Table 7.5-3 indicates Type A variable for Drywell water level. Deleted Drywell water level from Table 7.5-3 (not a 4 No discussion is provided on ranges, purpose etc. No type A variable). Added discussion on drywell water mention in Table 7.5-2. levelin Table 7.5-2 and Subsection 7.5.2.1(2)(o).
163 10.18A.5 (PC-1) Entry condition for hydrogen is not specified, Hydrogen level COL applicant dependent. 2% is not 2 only blank entry of Hi Alarm Level. EPGs say 2%. specified.
- 11. Deleted 164 12. Table 7.5-2 Suppression pool water temperature range up Changed upper range to 140C to accommodate all 3 to 110 C but HCTL curve have suppression pool possible suppression pool temperatures, temperature up to 150 C.
165 13. Tabic 7.5-2 Drywell atmosphere temperature range up to 110'C changed to 226.7"C. 319.5'C defines the slope 3 110 C but DWSil curves go up to 319.5 C. of the curve, it is not the maximum value.
9 30 w- w,- v,. m , -% . , - - .we c .e-+,w.-v .co-.- -.ws.- =,-.--,-wre 4-+ + w ,. -,- m +~w-,-rew- --w-.w--fr- --w-~s--=- =v- - , . - , w,- --wrmwa' w +- .-w=* er . - - - . , - -w e+-.m-.'-- ,. - . -
g CHAPTER 19 (SPSB) 166 1. A November 3,1993 letter from GE (J. Fox) regarding The interconnecting path between the wetwell and the 3
, " Primary Containment Pressure Control EPG-Low drywell has been added to the bypass study. The Pressure Venting" indicating that there is a potentially results of the study indicate that the bypass risk is significant suppression pool bypass path that was not below the threshold which would require further assessed in SSAR Section 19E.2.3.3, the containment consideration of this in the CETs or the MAAP event trees, or GE's MAAP analyses. This path involves a analysis. Therefore, no further consideration of this common nitrogen makeup line with separate branches to path is necessary. The SSAR has been modified to the drywell and wetwell. These lines (originating at reflect this additional pathway in the bypass study.
penetrations X-60 and X-240) are said to provide a interconnecting path between the wetwell and drywell which equalizes the pressure between the two primary containment volumes. Furthermore, the valves in each of the branches (F040 and F041) are said to be open during normal operation. No instrumentation to detect flow through this path during an accident is apparent.
The bypass analysis should be updated to reflect these lines as potential bypass paths. The validity of the CETs and supporting MAAP analyses should also be justified given this bypass path.
CHAPTER 19 167 Tab!e 19.2-1 Table incorrectly identifies concrete used in The statement in the SSAR is correct. As discussed 2 containment as limestone rather than basaltic. further in Subsection 19E.2.1.2.1(3), the containment is assumed to be made of limestone-sand concrete.
Only sacrificial concrete in the lower drywell is specified to be made of basaltic concrete. However, as this assumption will have little, if any, bearing on the performance of the coritainment, this item is deleted fro,7 the Key Assumptions, Table 19.2-1. Additionally, in reviewing the table, other features were identified which have been incorporated in the standard design.
Thus, they are no longer " assumptions" and have been deleted from the table.
t 31
- _. ._ ~- . _.. . .. _. __ _ - _ . . ~ . _ __ _. . ._ ._ - _ ._
A "OR" has been added to the table as indicated by the 4
- 168 Table 19.3-2 The success criteria for 'all Transients" should have an "OR* before " ADS 8*. Note (9) in the staff.
table under states the requirements for RWCU Note (9) is consistent with the COL Action item in to be usable as a high pressure system to Subsection 19.9.2. No modification to Note (9) is remove decay heat. required.
The new value of 2.7E-10 is correct. No change is 2 169 Table 19.3-5 ATWS Frequency was 1.7 E-10.
required.
As indicated below in the response to item 175, the 2 170 Table 19.3-4 Accident Class 11 frequency should be E-10 not E-
- 12. Table does r'ot address LOCA s Outside of impact of the October 15th submittal on the overall Containment. results of the PRA are not significant. Therefore, the results of study for Class ll events have not been propagated in the SSAR. No change is necessary.
The contribution of LOCAs outside containment to the core damage frequency was found to be negligible, as was the risk associated with these events. The core damage frequency associated with all LOCAs outside containment was found to be approximately 10% of the total core damage. This estimate was very conservative in its treatment of operator recovery actions, therefore, it is not appropriate to propagate the very small numbers which results from the study into the baseline PRA and the SSAR. No change is necessary.
ATWS and transient f requencies have changed f rom 2 171 ATWS and transient frequencies have changed from earlier values. earlier values. The values in the table are correct.
Release frequencies via rupture disc for 16 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 2 172 Table 19.3-6. Release frequencies via rupture disc for 16 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> has changed from 1.1E-10. has changed from 1.1E-10.
Sections 3.2 and 19.4 31 have been revised removing :4 173 Section 19.4.3.1.1 Discussion of Section 3.2 imt"es that ACIWA is Seismic Category I wwn GE has the implication that all systems that must remain functionalin the event of a safe shutdown earthquake indicated that it is not..
are seismic category 1.
CCFP-Pl is identified as 0.004 in SSAR but was 4 174 Section 19.5.3 CCFP-PI is identified as 0.004 is SSAR by was identified as 0.005.in ACRS view graphs. The identified as 0.005 in ACRS view graphs. The 0.004 value fails to include late releases from drywell (>24 0.004 value fails to include late releases from drywell (.24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />). With drywell releases hours). With drywell releases included CCFP=0.0066.
included CCFP= 0.0066.
32 ,
-)
175 Section 19.6.5 October 15th GE submittal on Class 11 indicates The October 15th submittal did not contain sufficient 3 that frequency of containment structural failure detail to calculate the frequency of containment
, from loss of heat removal for intemal events is structural failure. As requested by the staff, the branch 1.1E-8 per year with 1% resulting in core associated with the operation of the rupture disk was damage, not 1.1E-9 with 0.1%. deleted in that submittal. This branch has beer, reinstated in the figure to allow calculation of the structural failure frequency. The figures and text have been added to the SSAR as Subsection 19E.5.14.
Subsection 19.6.5 has also been updated to include the results of the alternative analysis. Note that there is a negligible effect of Class 11 sequences on the overall core damage frequency and risk. Only those reporting parameters relating directly to Class 11 are affected by these changes.
176 The system operation (frequent starting and The RCW and Service Water System are not included 3 aligning of standby pumps) assumed for the in Appendix 19K for inclusion in RAP because no RCW and service water system is not component of these systems appears in the 20 SSCs documented in Appendix 19K for inclusion to of greatest importance for level 1 failure of the systems RAP nor do there appear to be any COL action do not affect level 2 primarily due to COPS. Section items calling for this type of system operation. It 19.9.20 has been added to the COL License is unclear why this assumption would be carried Information requiring the standby pump and heat out by a COL applicant. exchangers to be started and the previously running service and sea water equipment be placed in a standby mode at least once each month.
177 Section 19.6.8 The CCFP was previously 0.005. The current value of 0.002 is correct. The text has 2 been modified slightly to indicate that the value of interest is below the lowest ordinate in the figure.
i 178 Section 19.7.3 ACIWA and CTG do not virtual ly eliminate The statement was intended to indicate that the core 3 station blackout as a contributor to core damage frequency due to station blackout is now very damage frequency. Iow. The text has been modified.
SBO is the largest contributor to core damage frequency in the ABWR PRA.
179 Lower drywell flooder - fusible plug temperature The incorrect value has been corrected. 4 conveds to S00F or 533K.
180 Seismic capacity of added features - Most of Reference to the ACIWA as seismic I and includes the 4 ACIWA is not Seismic Category 1. The ac- AC driven pump has been removed from Section
' driven pump is not part of ACIWA. 19.7.3.
33
= . - .- . _ .
The COL applicant is required to develop procedures 2 181 EPG Improvements - The manual operation of valves to cool the core and provide inventory for the manual operation of MOVs in Subsection makeup in the event of a large seismic event 19.9.15. These procedures will be applicable to a seismic event. No SSAR change is required.
should be added to the COL action items in Section 19.9.
4 The footnote to Table 191-1 has been revised to reflect 3 182 Section 19.8.4 ACIWA System - GE has indicated that if the ACIWA pump and water supply cannot be this commitment. See item 209.
designed and built to a HCLPF of 0.5g in a cost-effective manner, then the COL applicant will provide a fire truck capable of withstanding such an earthquake and if necessary will provide a building capable of withstanding such an
)
eurthquake to house the fire truck. GE needs to provide documentation of this commitment.
The divisional separation of high pressure lines was 2 183 Three Divisions of ESF _- High pressure lines are NOT to penetrate walls or floors separating identified as important to the flooding analysis. This different safety divisions. feature is included in the text of Subsection 19.8.5 and as the last item of Table 19.8.5. No SSAR modification is needed.
2 184 Four Divisions of SSLC - Administrative actions An extensive discussion of causes and defenses of i
to prevent CCF need to be applied to more than CCFs for EMUXs is provided in Appendix N. No EMUX calibration. GE should provide an revision to the SSAR is required.
l expanded explanation of what is expected to l
j-prevent CCF from occurring in the ABWR.
l Section 19.8.2.3 has been modified deleting reference 4 185 Section 19.8.2.3 Shortest path to core damage - It is the NRC's understanding that the limiting components for to the battery racks and chargers as limiting de power (seismic) are the batteries and the de components and adding the cable trays.
cable trays, not the racks or the chargers.
The sentence stating that no single component could 4
-186 Most sensitive components - Batteries or dc cable trays are single components whose cause HCLPF of the plant to drop below 0.6g has been failure could cause the HCLPF of the plant to deleted, drop below 0.6g.
=
34
1 187 ACIWA-The building housing the ACIWA The footnote on Table 191-1 has been revised 3 system (an extemal shed) must not fail consistent with the comment. See item 209.
seismically in such a manner that it would prevent the ACIWA from performing its function.
The ACIWA is not Seismic Category I. The ACIWA discussion should include the fire truck.
188 Section 19.8.4.1 The CDF reported in 19E.2.3.3.4 is 1.3E-8, The values reported in Amendment 32 are correct. The 3 with most (1.3E-8 ) non-bypass and 1.7E-10 table in 19E.2.3.3.4(3) has been clarified to indicate bypass. that the table references to Ex-containment LOCAs only. ,
189 Section 19.8.5.1 The f act that tunnels will connect some A statement was added to Section 19.8.5 to describe 3 building is not reflected in these statements, nor the reasons why flooding in the radwaste tunnels was are the assumptions regarding the adequacy of not treated in the event trees. Basically, the probability the flooding protection provided by the seals at was determined to be significantly lower than the the ends of these tunnels. The adequacy of flooding scenarios resulting from the pipe breaks in the these seals should be included as an Inter face buildings (i.e., a pipe break had to occur plus failure of item. multiple seal failures to cause interbuilding flooding)
The COL applicant must ensure the adequacy of the radwaste tunnel seat designs.
190 Section 19.8.5.3 Need to add (1) NEMA Type 4 enclosures for As requested, requirements for drip proof motors, 3 MCCs and motors that are drip-proof, (2) UHS NEMA 4 enclosures for MCCs, no gravity draining of cannot gravity drain to control building, and (3) the UHS to the control building, and a maximum of max of 4000 m of pipe between isolation valves 4000 meters of RSW piping between the control at the UHS and the RCW/RSW rooms. building and the RSW valves in the pump house were added to 19.8.5.3.
191 Section 19.8.7 Lower drywell design. The interconnection The SSAR has been modified to reflect this number. 3 between the lower drywell and the wetwell is at this change has no effect on the analysis.
an elevation 8.6 m above the floor of the suppression pool.
192 Need to add capability of HPCF p' umps to pump Reference to HPCF capability to pump 340'F has been 3 340 F water from suppression pool. added to Section 19.8.1.
193 Table 19.8-2 ACIWA is not Seismic Category 1. The reference to the ACIWA System as Seismic 4 Category I has been deleted.
194 Section 19.9.1 On May 17,1993, in a fax from Cal Tang to Section 19.9.1 has been revised as requested. 3 Chet Posiusny and George Thomas, GE committed to revise Section 19.9.1 in its .
entirety. It has yet to be modified.
35
item (9) was added to Section 19.9.10 describing a 3 195 Section 19.9 On October 22,1992,in a fax from Glenn Kelly to Jack Duncan, the NRC asked GE to provide COL Action item to ensure that the RSW purnphouse guidance to a COL Applicant on how to assure is designed to prevent interdivisional ficoding and that the assumptions in the ABWR PRA draining of water to the control building.
regarding the UHS and the Reactor Service Water System come true in the as-built plant.
GE needs to provide this guidance.
The SSAR section has been modified. 4 196 Section 19.9.17 Ultimate pressure capability will be shown to be at least 134 psig.
Depressurization with ADS during a station blackout 3 197 Section 19.9 Add a procedure to depressurize with ADS during a station blackout after loss of RCIC. Add a after loss of RCIC is currently in the EPGs. A procedure to backup DC power to ADS valves procedure to backup DC power to ADS has been to keep open for up to 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> following loss of added to Section 19.9.9.
RCIC when its battery fails.
The October 15 submittal was included as a sensitivity 2 198 Section 19D.4 The event trees for each initiation event contribution to Class 11 (TT, TIS,TEO) should be study in new section 19D.5.14. As indicated in Section replaced or supplemented by the modified trees 19D.5.14, the results of the study do not warrant described in the October 15,1993 Class il propagation into the balance of the SSAR. No submittal. modification is needed.
The October 15 submittal was included as sensitivity 3 199 Section 19D.5 Update Sections 19D.5.11.4,19D.5.12.4 and Figure 19d.5-10 to reflect the October 15,1993 study in new section 19D.5.14. A reference to this submittal. section was added to 19D.5.14. As indicated in Section 19D.5.14, the results of the study do not warrant propagation into the balance of the SSAR.
Nonetheless, minor modifications to Subsections 19D.S.12.2 and 19D.S.12.4 were made to prevent confusion.
A modified discussion has been provided in Section 3 200 Section 19D.7.2 "OP f alls to depressurize RCS* said to be
' considered in level 1 analysis and is referred to 19D 7.2 to provide a description of " bounding -
in Section 19D.7.2. Modify SSAR (including analysis". As discussed in the modified text, the value table) to better add-ce Depressurization of the human error portion of the OP node is actions including k a Zwmssurization. approximately 0.002, However, the table has not been -
modified since one would not be able to track this value to a specific node in the CETs. The table does have a footnote directing the reader to the text for this node.
e 36 m_ .. . _ ._____.m _ . _m-_m.__mm._m__.. _ _ __ _ _ _ _ _ _ _ . _.__m_ _ _
l
+
4 The references to Section 190.7.3 in items (g) and (h) 4 201 Section 19D.7.3 Wnte up has been pruned of all discussion specific to associated critical tasks (g) and (h), in the critical tasks list and item (J) in the COL system and COL system operating procedure (j). e.g., operating procedures list in Section 19D.7.6. were in
- valve numbers (including F005. -17 and -18) error. These references have been corrected to refer to
- "must have access under accident conditions" Section 19D.7.2.
GE needs to significantly improve its write up in this area.
Section 19D.7.11. Table 19D.7-11, and Table 19D.7- 3 202 Table 19D.7.12 Control room capability incorrectly indicated for several critical tasks. Table incompletely 12 have been deleted. Section 19D.7.6 has been discusses unambiguous indication and control revised to provide more complete information as room capability for same COL system operating requested.
procedures.
3 203 Section 19D.10.6.1 The list of top 10 contributors to uncertainly Section 19D.10.2 has been revised to address the in CDF needs to be modified to reflect the issue.
Updating of the importance measures for the CTG and the EDGs.
204 Section 19E.2.1.2.2.1 GE needs to clarify wording in SSAR on This subsection considers only the long term station 3 statement regarding core cooling function being blackout scenario. There is no discussion of the use or lost if AC is unavailable af ter eight hours to non-use of the ACIWA during the first few hours of an .
Indicate that this is true only for the intemal accident. A statement was added to indicate that the events analysis and not for seismic events ACIWA cou!d be used to prevent core damage during where ACIWA is credited in the levc! 1 trees. the long-term station blackout scenario.
205 Section 19E.2.1.2.2.2 The discussion on RCIC roam A character was inadvertently dropped f rom the 1 temperature state that the ABWR will be document. The correct temperature is 151*F. The designed to prevent the temperature in the SSAR has been corrected. r room from rising to 15 F. This number is probably incorrect, particularly since the normal room temperature is 104 F.
206 Section 19E.2.1.5 GE states that the " propagation of The word "not" was inadvertently deleted from the text. 1 uncertainty distributions was carried out as The SSAR has been amended.
' done in NUREG-1150'. This is incorrect and significantly overstates the analysis performed by GE.
0 37
207 Section 19E.2.3.3.4 The reported core damage frequency for The core damage frequency for ex-containment 2 the bypass events is approximately one order of LOCAs with bypass decreased due to 1) tne removal
. magnitude less than in the previous submittal. of the sampling lines from consideration since these i GE needs to provide an explanation for the lines do not constitute a LOCA event, and 2) credit for i reduction in CDF. the isciation of RCIC and CUW lines which has been ,
conservatively omitted in the initial analysis. The analysis is appropriately described in the SSAR. No discussion of these modifications to the analysis is required in the SSAR.
208 Table 19H-1 The table needs to be expended to address dc The values given in Table 19H-1 include DC cable 2 power cable trays. trays. No changes to Table 19H-1 are needed.
209 Table 191-1 The table needs to be updated to reflect the The footnote to Table 191-1 has been revised to reflect 3 changes to the assumptions regarding the this commitment.
ACIWA system HCLPFs including the fire truck .
and any building that is required to house it.
Tables 191-2 and 1M-4 have been updated to reflect 4
, 210 Tables 191-2 and -4 The tables need to be updated to reflect the changes in the ACIWA assumptions. changes in ACIWA assumptions.
211 Figure 191-1 It is the NRC's understanding that the fire water Figure 191-1 has been revised to reflect the 0.5g 3 (ACIWA) system's HCLPF is 0.5g. not 0.62g. HCPLF for ACIWA.
212 Figures 191-16 and-19 The titles of these figures do not match The titles to Figures 191-6 to 191-19 have been 1 the contents of the diagrams. corrected to match the contents of the diagrams.
213 Section 19K.5 The statement,"The primary containment and A reference to " Category I" was added to the 4 the reactor building are the structures with the description of the primary containment and the reactor lowest values of HCLPF, ... " is incorrect building, because it does not identify that this is only true A reference to " Cable trays" was added to the list of for structures with safety related equipment that DC power system equipment important to the seismic are in the certified design. The Service Water analysis Pump House has a lower capacity. Further
- Fire truck" was not added to the seismic analysis list down this paragraph " battery cable trays" and of important systems. No credit was taken in the
" fire truck" need to be added to the list of SSCs seismic analysis for the fire truck. This is consistent identified as being important. with the agreed list of important features in Section 19.8.2.
- 214 Section 19K.11.1 The discussion in this section needs to be This section was updated to reflect the importance of 3
- updated to reflect the importance of the CTG the CTG and the EDGs.
and the EDGs.
38 w-w---.e=w--- - - r. --w w . . - e aw we- -
+ w or '-w-w-- er.w- -i=e-=!-%'-< - -e e-ww --w -- --
e v-=e- * - - e* --e**= ++ r % , . . _ - - _ _m--__--_____-__4
l 215 Section 19K.11.7 This section needs to be expanded to DC power cable trays were added to this section, as 3 t address de power cable trays. were the service water system pumps, pump house, and air conditioner. These items were also added to l Table 19K-4.
216 Table 19K-1 This table needs to be updated to reflect the Table 19K-1 was revised to reflect the importance of 3 importance of the CTG and the EDGs. the CTG and the EDGs.
217 Table 19K-4 " COPS AOVs inadvertently left open closed The SSAR bcs been modified as noted. 1 fo!!owing maintenance." Delete "open" The word "open" was deleted as requested.
- 218 Section 19K.11.5 The write up needs to address the opening Section 19.9.7 has been revised to include the 3 valves F-005, F-017, and F-18 for ACIWA operation of valves F005. F017, and F018 for ACIWA operation (core injection or drywell spray). operation. Reference to Section 19.9.7 has been provided in Section 19K.11.5. The testing requirements of RHR system valves F005, F017, and F0tB for RAP has been provided in Section 19K.4.
219 Section 190.4.3 This section on shutdown risk too strongly A paragraph has been added to 190.4.3 describing the 3 downplays the potential effect of loss of potential for high offsite doses if a core melt were to containment integrity during an event in occur when the primary containment was open.
shutdown. At a meeting with GE in San Jose on this issue, the NRC was told that the consequences would be Yaty harsh if there were a core melt during modes 3,4, or 5 with loss of containment integrity such that the pool would be bypassed. GE should more accurately portray the effect of loss of containment integrity in order to not mislead a COL applicant as to the potential seriousness of such an event.
Other Sections 220 Figure 6.2-39, sht 2/3 The atmospheric control system P&lD An enunciation has been added to the drawing. The 4 does not show an annunciation on the vacuum SSAR has been modified.
breakers on position indicated.
39
-- . - - . _- , - .- . -, .- . - ~ . , . _ . . - - .. . .. . - .-
t l
221 Section 5.4.7.1.1.10.1 A GE mark up (Taft tax 8/26/93) is not A discussion of potential dose rates in the areas where 3 incorporated in the SSAR. Certain valves in the the valves necessary to operate the ACIWA System ACIWA needed for drywell spray (and possibly are located has been provided in Section 5.4.1.1.10.
for core injection) are inaccessible after core
- damage if water subsequently is circulated through specific ECCS lines. This fact should be discussed.
222 Section 5.4 The NRC is waiting for a complete description A description of the operation of the ACIWA including 3 of operation of ACIWA including operation of operation of the valves in the yard and how the diesel-valves in the yard and how the diesel-driven driven is operated has been pro'vided in Section
. pump is operated. 19.9.7. A reference to Section 19.9.7 has been provided in Section 5.4.1.1.10.
223 Section 6.2 The NRC is waiting for a complete description Sections 5.4.1.1.10 and 19.9.7 have been revised to 2 of drywell/wetwell spray operation. This include the requested information about ACIWA description should a discussion of the orifices, System operation. No change to Section 6.2 is j alignment of equipment, capabilities of necessary.
local / manual valve actuation, and accessibility / shielding.
$ 'W 40 F . .
j
..L.-- - . . . - -
'l Heat Caoacity Temoerature Limit This issue is being addressed as follows: 3 224 GE's review of certain station blackout (SBO) sequences showed that suppression pool temperature has the potential to exceed the EPG HCTL During a SBO, the only injection A paper will be written by GE and reviewed by Prof.
system available to the RPV is the turbine driven reactor core Sonin of MIT to justify that depressurization of the isolation cooling (RCIC) system. Once the HCTL is exceeds the reactor with the suppression pool at high temperature operator is directed to depressurize the RPV. When RPV will result in stable condensation, and will also address depressurization occurs RCIC, a high pressure injection system, the scrubbing issue. This approach has been would become unavailable for injection and may lead to heat up communicated to the NRC staff.
of the core.
GE then submitted for staff review a revised HCTL with a low-pressure endpoint temperature of 137.7*C instead of 103.9"C.
This upward shifting of the HCTL curve postpones RPV depressurization and would increase the availability of the RCIC. Unfortunately, this upward shift also allows temperatures exceeding saturation to exist within the suppression pool.
There are disadvantages associated with operating the suppression pool at or near saturation. An extended plume of high quality steam was observed during sub-scale experiments performed by Chun and Sonin when the pool reached saturation temperature. The staff is concemed about the existence of large stem bubbles may drift into a relatively cooler area within
- the containment integrity. With the lost of the RHR pumps during a SBO there is a concern of a stratified poolis a possibility.
Another consequence of these extended plumes of steam is the reduction of the scrubbing capability of the suppression pool.
This would result in a direct path from the quencher to the wetwell airspace thus effectively bypassing the suppression pool.
41
The staff acknowledges the value of increasing the availability of the only high pressure injection system RCIC, during a SBO.
The staff does not believe that this increased availability is significant enough to justify operating the disadvantages mentioned above. The staff also believes that the fire water addition system should be available for low pressure injection once RPV depressurization takes place. Therefore, SCSB recommends that the ABWR adopt a HCTL curve that does not exceed saturation temperature for atmospheric conditions such as the one in Amendment 31 of the ABWR EPGs.
225 GE did not include analyses on l&C diversity issue that was included in revised Appendix 7C 3 docketed on June 18,1993. GE plans to include a synopsis of this analyses in Appendix 7C Amendment 33.
i The following action items the staff needs to respond to the ACRS subcommitee or full committee on safeguard information in the future: .
t 226 1. What is the latest safeguards submittal for the ABWR Safeguards submittal as part of Amendment 33. 3
' design? The staff may have to revisit the FSER and to make sure that their safety findings are still valid or not affected?
l 227 2. Should vital area classification of CAS and SAS be NRC action. 2 interface requirement rather than COL action item?
l
. 228 3. Where CAS and SAS will be located in the ABWR design? NRC action. 2 229 4. Will CAS and SDAS be seismically qualified? NRC action. 2 l 2
'- 230 5. Should be an ITAAC/DAC for security? NRC action.
231 6. . How GE perform the sabotage vulnerability analyses when NRC action. 2 they do not a detailed security design? What are the NRC sabotage vulnerability requirements that the staff uses for the ABWR review?
,,*-**...... .. .w,. . , . - . . - , _ . .m....#_ , ,. . .2m.,.. . . , , ,. . . . . . _ _ , .- .. .,_, , ,,,_, m ,,,m _ __ . , ., ,. . . _, m,. , , _
This is a summary of the discrepancies found as a result of the SPLB review of ABWR SSAR Amendment 32. Not included are discrepancies found in SSAR Section 6.4.6.5. 9.4 and 11, which i
have been provided to GE previously.
GliAP_IRBJ 232 1. Modify the 10th paragraph of the SSAR Subsection incorporated 1 3.4.1.1.2 to " Analyses of the worst flooding due to pipe and tank failures and their consequences are performed in this subsection for the Reactor Building, Control Building.
Radwaste Building, Turbine Buildina. and Service Building."
233 2. Modify the third paragraph of SSAR Subsection incorporated 1 3.4.1.1.2.1.2 to correct "SWCU" to "SPCU".
- 3. Make the following modifications to SSAR Table 3.4-1 and related layout drawings, 234 The first column of Table 3.4-1 and Fig 1.2-6 say that the R/B Figure 1.2-6 (4800mm) shows the clean access 3 tunnel between the Reactor Building (RB) and Service path at coordinates RA/R1-7. This path leads to the Building (SB) is at 4800mm but Fig.1.2-6 does not clearly S/B ramp down to 3500mm (Figure 1.2-18). The ramp identify the access way and column 2 of the table, along and the clean access path are identified by with Figs.1.2-14 and 1.2-15, show the access way at clarification's to the figure.
3500mm. Also, Fig.1.2-18 shows the access way at 3500mm (in addition, the access way is not clearly labeled on this figure).-
235 Columns 2 and 3 of Table 3.4-1 state that thrire is an C/B and S/B Figure 1.2-15 (Section B-B) does not 3 access way between the control and service buildings at show the 3500mm access between these building 3500 mm but Fig.1.2-18 does not clearly label the access because the access location is not in the view of way. Section B-B. The access between the C/B and S/B is shown on Figure 1.2-18.
236 Column 2 of Table 3.4-1 states that there is an access way .See the clarification's to Figure 1.2-18 in item 243. 3 between the service and turbine buildings at 3500 mm but Fig.1.2-18 does not clearly label this access way.
t 43
The R/B sumps are at -8200mm and the flow must be 4 237 SSAR Fig. indicates that the radwaste tunnel slopes downward to the -8200 mm elevation at the RB and Control pumped upward to the radwaste bui! ding (Figure 1.2-building (CB) ends of the tunnel. This is in direct 23a).
contradiction to GE's discussion with the staff that the The T/B sumps are at 8800mm and the flow is highest section of the tunnel would be at the RB and CB downward to the radwaste building. The radwaste ends to ensure that any flooding in the tunnel would flow tunnels are sealed at each building waff.
away from safety-related areas.
incorporated 4 238 4. Modify SSAR Subsection 3.5.1.1.3 to refer to Fig. 3.5-2, not 3.5-1 Incorporated 4 239 5. Modify the SSAR Subsection 3.5.4.5 to refer to SSAR Subsection 3.5.1.1.1.3, not 3.5.1.1.3 incorporated as item (12) of subsection 3.6.1.1.3 with a 3 240 6. Add a statement in 3.6.1 that all walls, doors, floors and penetrations which serve as divisional boundaries will be clarification of postulated pipe failures outside primary designed to withstand the worst case pressurizations containment and within secondary containment.
associated with the postulated pipe failures incorporated 3 241 7. GE has agreed to change the reference in SSAR Section 3.11 from Chapter 12 to Chapters 11 and 12.
CHAPTER 6 1
i 242 1. Section 6.2.1.1.1 (page 6.2-1). Item (4): typographic error, Incorporated l
" flow form
- should be read as " flow from".
1 243 2. Section 6.2.1.1.1 (page 6.2-1) Item (%): typographic error, incorporated
" form the reactor core' should be read as *from the reactor".
1 244 3. Section 6.2.1.1.1 (page 6.2-2), item (6), (7), and (8): similar incorporated typographic errors " form" should be read as "from".
245 4. Similar typographic errors on " form" vs. "from" spreading incorporated 1 throughout the rest of Section 6.2.1 and maybe beyond.
This is a generic typographic error. GE should identify all the specific errors and correct them.
44 ,
- - _ _ _ _ _ _ _ _ _ . . _ _. _ __ _ __ _ , _ ]
. . l 1
246 5. Section 6.2.1.1.3.3.1.2 (page 6.2-10): Assumption No. 7 in incorporated 4 the previous Amendments regarding feedwater enthalpy is See item 7 of Subsection 6.2.1.1.3.3.1.2.
missing in Amendment 32. Put it back.
247 6. Section 6.2.1.1.3.3.2.1. (page 6.2-12): Assumption No.1 incorporated 4 regarding critical flow modelin the previous Amendments See item 1 of Subsection 6.2.1.1.3.3.2.1.
was taken away in Amendment 32. Put it back.
248 7. Section 6.2.1.1.3.5.1 (page 0.2-13 ar'd -14); typographic incorporated 1 errors on " Table" and " temperature".
249 6. Section 6.2.1.1.5.6.1 (page 6.2-26): typographic error on incorporated 1 "the".
250 9. Section 6.2.1.2.2 (page 6.2-36): The break sizes of "150A" The "A" carries the dimension of mm. The definition is 2 and *SOA" should be "150mm" and "50mm". provided in Figure 1.7-1. It is not necessary to duplicate.
251 10. Include Tables 6.2-37 a-e in Chapter 6 of the SSAR Tables 6.2-37a - e do not exist. 2 CHAPTER 9 252 1. Add information to the SSAR regarding the COL applicant to Already required by Subsection 9.1.6.1 which 2 provide a criticality analysis showing that the design of the references Subsection 9.1.1.1.1 which in turn requires new storage racks will be such that Keff will not exceed 0.98 the COL applicant to respond to Question 430.180 (all with a fuel load of the highest reactivity, assuming optimum information requested).
moderator conditions (foam, small droplets, spray, or fogging), as described in SRP Section 9.1.1.
253 2. Add information to the SSAR discussing the storage of incorporated in Amendment 31. See Subsection 2 defective fuel assemblies and provide design requirements 9.1.4.2.8 and Table 3.2-1.
in Table 3.2-1 of the SSAR.
254 3. Prov3e dehn requirements for the spent fuel pool liner in incorporated 3 Table 6.F 255 4. Incorporate information regarding protection of the filter- Incorporated 3 demineralizer resins in 9.1.3.
45
Defective fuel storage is stored in the equipment 4 256 5. Modify SSAR Subsection 9.1.4.2.8 to clarify that defective fuel assemblies are placed in special storage containers storage rack. See Subsection 9.1.4.2.8. COL license and stored in the spent fuel storage rack, not the equipment information requirements are incorporated in storage rack, and correct SSAR Subsection 9.1.4.3 to state Subsection 9.1.6.4.
that the COL license information requirements are in SSAR Subsection 9.1.6.4, not 9.1.4.3.
incorporated 4 257 6. Correction SSAR Subsection 9.1.5.8 to refer to SSAR Subsection 9.1.6.6, not 9.1.6.7.
See ITAAC submittal 3 258 7. Modify the second paragraph of ITAAC 2.11.23 to include the Control Building in the list of buildings in which the PSW system is part of the Certified Design.
included in Amendment 32. 2 259 8. Modify SSAR Fig. 9.2-9 to include the discharge from the nonradioactive drain system. This connection is downstream of the hypocontact tank in the figure.
3 260 9. Modify SSAR Subsection 9.2.5.1 (11) to include capability for incorporated full operational inspection and testing incorporated 3 261 10. Modify SSAR Subsection 9.2.5.10 to include inspections and tests during normal operation Reference should be to Table 9.2-3. 3 262 12. Modification SSAR Subsection 9.2.9.1 (5) to refer to Table 9.2-1, not Table 9.2-2. Incorporated.
263 13. Remove references to Fig. 9.2-1a from SSAR Subsection incorporated 1 9.2.11.2 Decision was made between GE and NRC to not make 2 264 14. Modify Tables 9.2-4a-c to refer to safety-related and nonsafety-related instead if essential and nonessential this change.
Decay heat does not appear in Tables 9.2-4 a-c. Each 2 265 15. Clarify the heat capacity. SSAR Subsection 9.2.11.2 states that the 191g1 reactor decay heat 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after shutdown is division of RCW has cooling loads from the RWR 31.8 E6 kcal/hr but Tables 9.2-4 a-c indicate that each system in addition to other cooling loads.
division will need to accommodate approximately 30 E6 t
kcal/hr
. =-
46 Y
]
9 266 16. Modify SSAR Subsection 9.2.14.2.2 to remove the next to Statement is correct. For example. selected sensors 3 the last sentence of the subsection (there is no safety- for the RPS are located in the Turbine Building.
related equipment in the Turbine Building) incorporated 3 267 17. Modify SSAR Subsection 9.2.15.1.4 to clarify that on a LOCA signal. all standby pumps start and all standby valves open.
As indicated in the second paragraph of Subsection 2 268 18. Modify SSAR Subsection 6.7.2 and ITAAC 2.11.13 to clarify that the supply valve to the bottled nitrogen supply also 6.7.2, the values between the non-divisional and opens on a low pressure signal in the nondivisional portion divisional systems close on low pressure. Subsystem of the system. 6.7.2 and ITAAC 2.11.3 are cor.sistent.
269 19. Delete references to Fig. 9.3-9 in SSAR Subsection incorporated 3 9.3.8.1.1 (5) (b)
Incorporated in Subsection 9.3.12.4. 3 270 20. Correct the reference to a COL Action item made in SSAR Subsection 9.3.6.1.1 (5) (c). 9.3.12,1. does not refer to the DTS. It refers to the NRD. Make a separate COL Action item for the DTS.
incorporated in Subsection 9.5.1.3.12. 3 271 21. The staff indicated that GE's design capabilities for fire protection and mitigation in primary containment internal areas during shutdown conditions, supported by operationa!
controls and procedures appear to adequately address the concems. Further evaluation of this information will be completed and follow-up discussions will be conducted to provide feedback to GE and to identify any required SSAR changes if necessary. GE agreed to provide write-up in Section 9.5.1.3.12.
272 22. The staff had requested a change in the SSAR to indicate incorporated in Subsection 9.5.1.1.6. 3 that the smoke control capability would take into account the fact that the fire doors would be maintained open between a fire area and a non fire area. GE provided a revised markup which will be included in SSAR amendment and was found to be acceptable except that the words
maintain open" need to be included.
47
incorporated in Subsections 9.5.1.1.4,9.5.1.1.5, 3~
273 23. The staff identified a statement in the SSAR that cables in trays with bottoms were not considered in the total 9A.2.4 and 9B.2.3.3 combustible loading. This was not in compliance with Generic Letter 86-10 which states that all cables in trays need to be part of the total loading. GE agreed to delete the statements in the SSAR which indicate the exclusion. The staff found this to be acceptable. GE will provide additional changes if other exclusions are found in the SSAR.
incorporated in Subsection 9.A.3.6. 3 274 24. The staff identified that in the SSAR GE had referenced the ICBO 1495 Code for design of the type 1 walls. The staff stated that ASTM E-119 code needed to be referenced. GE committed to revising the SSAR and providing markups of the SSAR.
- 25. GE provided a discussion of deviations from the BTP. GE provided a handout GE which justifies each deviation. The following is a summary of the discussions for each item.
The DG room has an automatic foam system as a fire 2 275 Additionally, GE is to provide a markup regarding DG toom fire and manual FF. suppression with closed head water sprinkfers with fusible I:nks. GE/NRC agree that current detection / suppression systems will prevent inadvertent actuation of the sprinkler system. The DG room has sufficient space to hold the suppressant and it will not cause any overflow should the door be opened for manual fire fighting.
Incorporated, see Subsection 9.5.13.12. 3 276 a. High impedance Faults - A deviation from the specification of the commitment to perform a high impedance f ault analysis to ensure that such f aults could not affect the operation of safety related equipment.. GE provided an acceptable markup.
1
277 typographical error in the SSAR BTP reference. GE provided an acceptable markup which corrected the error.
g i f
(
48-
^ ^ ~ ^
278 c. Diesel Fuel Storage Area - GE has located in the Justification provided under Subsection 9.5.1, new 3 reactor building, outside secondary containment,3 item (1).
diesel fuel tanks which are greater than 1100 gallons in capacity. The staff requested that GE show that the sunken floor below each tank will accommodate fire 4
suppression water and foam for 30 minutes without forcing spilled fuel to migrate to other areas of the plant.
GE agreed to consider the staff's concern.
i
, 279 d. Control Room Complex - GE committed to changing the incorporated in Subsection 9.5.1, item (2) 3 design to add fire detection capability to the sub-floor i
area which was acceptable to the staff.
280 e. Plant Computer Room - GE indicated that this was not a No change. 2 deviation from the SRP and would not need to be further discussed.
281 f. Outdoor Transformers - For this item GE indicated that Repeat of item 72 2 a commitment to NFPA 15 will be added to the SSAR I
and to indicate that the barrier walls to be used will be equivalent to a one-hour fire barrier.
282 26. Clarify that the diesel engine is capable of operating for Clarified 3
. minutes without secondary cooling to ensure that the engine can operate at futiload in excess of the time required to restore cooling water (RCW and RSW) which are sequenced onto the emergency power supply within 1 minute following a Loss of Preferred Power (LOPP) 283 27. Modify SSAR Subsection 9.5.5.2 to state that the CCL incorporated 1 License information is in SSAR Subsection 9.5.13.6, not 9.5.13.5
- 284 28. Modify SSAR Subsection 9.5.5.2 to state that the system is Incorporated. 3 filled with high quality treated water from the Makeup Water (Purified) system, not the Demineralized Water System.
l 285 29. Reinstate note 4 on Fig. 9.5-8 clarifying that the air dryer incorporated. 3 includes both pre- and after-filters i
i e
49
- . - -- - - . . . . - - . - . . . - . - . - - - - . - . - .- --.- . . - . - ~ . _ . . . - . - - - . - _ - -
.--~. . -. . .-
incorporated. 3 286 30. Correction Fig. 9.5-9 to change the flow sense. shown on the tube el sump tank to a level sensor, as had been agreed
- to and modified in an earlier version of the figure.
incorporated. 3 287 31. Mod:fy Fig. 9.5-6 to show the pressure sensors used to detect high pressure conditions in the crankcase (as discussed in SSAR Subsection 9.5.8.2) and to show the 3 differential pressure gauge used to monitor plugging on inlet filters (as discussed in SSAR Subsection 9.5.8.3)
CHAP _IER10 .i incorporated, see ITAAC 2.10.7 . 3 288 1. The Design Description of ITAAC 2.10.7 should add "lVs" on page 2.10.7-2 for the " Actions for Protective Action."
incorporated 3 289 2. Revise SSAR Chapter 15 for the turbine CV trip closum time to *0.08 seconds or greater."
incorporated 1 ,
290 3. Define "NBR" in SSAR Section 10.2.1.3.3 Subsection 10.4.3.5.1.3 is correct. 2 291 4. Revise SSAR Section 10.4.10 to refer to 10.4.3.5.1.2, not 10.4.3.5.1.3.
incorporated 4 292 5. Revise SSAR page 10.0-iii/iv, Tables 10.4-4 through 10.4-6, Figure 10.4-4b, and the text of Section 10.4.6 to reflect system's designation as " Condensate Pu reation System (CPS)."
incorporated 3 293 6. Revise tae last paragraph of SSAR Subsection 10.4.7.2 to state "Th9 system extends... outlet to (but not including) the seismic interf ace restraint outside containment.* and the last paragraph of SSAR Subsection 10.4.7.3 to state "The portion which connects to the seismic interface restraint ,
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outside the containment... Reactor Building." '
Figure 10.4-7 and ITAAC Figure 2.10.2a are not the 2 294 7. SSAR Figure 10.4-7 should reflect instrumentation and its corresponding locations as shown in ITAAC Figure 2.10.2a. same system.
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e 295 8. Delete Subsections 10.4.5.9 and 10.4.5.10 on page 10.4-19 incorporated 4 of the SSAR. These are already printed on page 10.4-18 as Subsections 10.4.5.7 and 10.4.5.8.
9118ElElL11 296 1. Reinstate the PalD s for the ' qviously found in SSAR Chapter 11 _ _
297 Section 18.5 In forth paragraph replace "acticrs item" with
- license information incorporated 1 requirement" 298 Section 18.8.1 in last sentence replace " action" with more suitable phrase Replaced " action items are" with
- license information 1 is" Section 18.8.13 299 Put period at end of second sentence incorporated 1 300 Insert "to" between the words actions and isolate in second Incorpe ated 1 sentence.
301 Insert ")" following " Table 18E-1"in last sentence. Incorporated 1 302 What date for IEC964 on page 18E-20 and IEEE-1023 on Page Dates are provided in Table 1.8-21. 2 18E-21.
303 What are dates ict ANS! HSF-100 and IEG'964 on page 18E- Dates are provided in Table 1.8-21. 2 247 Section 13.5 304 Where is the rest of sentence pertaining to " Loss of Feedwater sentence complete by replacing " " with")" at end of 1 System Failure? on page 13 57 sentence.
305 What happened to autoblowdown in upper portid : of page 13.5- Autoblowdown applies only to PWRs. 2 87 306 What are dates for MIL-H-468558 and MIL-STD-1472D on page Dates are provided in Table 1.8-21. 2 j 13.5-8 i
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Section 15.6 307 Delete Subsaction 15.6.7.2, analysis done using boundary incorporated 3 i Chi /O.
Section 11.2.5 308 Add Sections 1001-2402 after 10CFR20 under item (2). Incorporated 3 309 Insert "10 times" following the word within under item (5). Incorporated 3 Sections 8.2.4 and 8.2.
310 Proposed staff interface requirements / conceptual design is Final version in Amendment 33. 3 provided in Attachment 1. GE to develop, in conjunction with the staff, final version suitable for inclusion with Amendment 33.
COMMENTS RECEIVED 11/18 - 19/93 3.4.1-Flood crotection Pending clarification in the SSAR of the following discrepancies:
311 Modification of the 10th paragraph of SSAR Subsection Same as item 232 2 3.4.1.1.2 to " Analysis of the worst flooding due to pipe and tank failures and their consequences are performed in this subsection for the Reactor Building, Control Building, Radwaste Building,lurbino Building and Service Building.
312 Modification c :ne thiro paragraph of SSAR Subsection Same as item 233 2
- 3.4.1.1.2.1.2 to correct "SWCU" to "SPCU". _
Make the following modifications to SSAR Table 3.4-1 and the related layout drawings, 313 The first column of Table 3.4-1 and Fig.1.2-6 say that the Same as item 234 2 tunnel between the Reactor Building (RB) and Service Building (SB) is at 4800mm but Fig.1.2-6 does not clearly identify the access way and column 2 of the table, along with Figs.1.2-14 and 1.2-15, show the access way at 3500mm.
Also, Fig.1.2-18 shows the access way at 3500mm (in addition, the access way is not clearly labeled on this figure.)
314 Columns 2 and 3 of Table 3.4-1 state that there is an access Same as item 235 2 way between the control and service buildings at 3500mm.
However, Fig.1.2-15 does no t show this access way and Fig.1.2-18 does not clearly label the access way.
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315 Column 2 of Table 3.4-1 states that there is an access way Same as item 236 2 between the service and turbine buildings at 3500mm but Fig.1.2-18 does no t clearly label this access way.
316 SSAR Fig. indicates that the radwaste tunnel slopes Same as item 237 2 downward to the -8200mm elevation at the RB and Control Building (CB) ends of the tunnel. This is in ditect contradiction to GE's discussion with the staff that the highest section of the tunnel would be at the RB and CB ends to ensure that a.ty flooding in the tunnel would flow away f rom safety-related areas 3.5.1.1-lHIEBNALLY-QENERATED MISSILES QUTSIDE C.ONTAINMEMI Pendha resolution of the following discrepancies:
317 Modify SSAR Subsection 3.51.1.1.3 to refer to Fig. 3.5-2, Same as item 238 2 not 3.5-1 318 Modify SSAR Subsection 3.5.4.5 to refer to SSAR Same as item 239 2 Subsection 3.5.1.1.1.3, not 35.1.1.3 3.6.1 _ElEEJAlt,tBES Pending correction of the following discrepancies:
319 Addition of a statement in 3.6.1 that all walls, doors, floors, Same as item 240 2 and penetrations which serve as divisional boundaries will be designed to withstand the worst case presurrizations associated with the postulated pipe failures 320 Inclusion of Tables 6.2-37 a-e in Chapter 6 of SSAR Same as item 251 2 M1-JQUIEMIEMLQUAllHCAHQH 321 SSAR Section 3.11.5.2 states that normal opert*ional exposure Same as item 241 2 is based on the radiation sources provided in chapter 12. The staff has determined that this reference is incorrect. GE indicated that this reference wi!! change to state that the normal operational exposure is based on a source term provided in chapter 11, and inventory provided in chapter 12.
CHAPTER 14 322 Section 14.2.12.1.45.4(3)(j) of SSAR amendment 32 uses the The word is either. It has been corrected. 1 phrase "... powered from wither preferred or standby sources.. "
Clarify the use of the word " wither"in this phrase.
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Subsection 8.3.1.5.2 changed to Subsection 8.3.1.1.5 4 323 Section 14.2.12.1.45.4(3)(i) of SSAR amendment 32 references sutsectinn 8.3.1.1.5.2. There is no subsection in the SSAR. in Subsection 14.2.12.1.45.4(3)(i). Design vol tag 3s are Clarify where the design voltages are specified in the SSAR. not specified in the SSAR. The use of design voltages in terms of i 10% fluctuations are documented in GE letter " Response to NRC comments on SSAR Section 14.2" dated May 13,1993. The phrase "as specified in" has been replaced with the phrase "in accordance with" for clarification of Subsection 14.2.12.1.45.4(3)(i) 324 Section 14.2.12.1.45.4(3)(j) of SSAR amendment 32 indicates Available bus voltages are not specified in the SSAR 3 that available bus voltage are specified in Subsection The phrase "as specified in " has been replaced with 8.3.1.1.8.3. Available bus voltages do not appear to be specified the phrase "in accordance with" for clarification of in this subsection. Provide clarification. Subsection 14.4.12.1.45.(3)(i).
325 Section 14.2.12.1.45.4(3)(b) of SSAR amendment 32 indicates Bus voltage and frequency variations between no load 3 that acceptable bus voltage frequency variations between no conditions are not specified in the SSAR. The phrase load and fu!! load conditions are specified by subsection 8.2.3. "as specified in" has been replaced with the phrase "in Acceptable bus voltage and frequency variations between no accordance with" a clarification of Subsection load and f ull load conditions do not appear to be specified in this 14.2.12.1.45.4.(3)(h).
subsection. Provide clarification.
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