ML20245A179
| ML20245A179 | |
| Person / Time | |
|---|---|
| Site: | 05000605 |
| Issue date: | 03/31/1989 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19297H572 | List: |
| References | |
| NUDOCS 8904250093 | |
| Download: ML20245A179 (600) | |
Text
__
MM 23A6100AC REV.A Remndard Plant
,o; CHAPTER 1 TABLE OF CONTENTS Sectisa Illie P_ase i
s INTRODUCTION AND GENERAL 1
DESCRIPTION OF PLANT
1.1 INTRODUCTION
1.11 1.1.1 Type of License Required 1.11 1.1.2 Number of Plant Units 1.11 1.13 Description oflocation 1.11 1.1.4 Type of Nuclear Steam Supply 1.11 1.1.5 Type of Cor siament 1.1 1 1.1.6 Core Thermal Power Levels 1.11 1.2 GENERAL PLANT DESCRIPTION 1.21 1.2.1 Principal Desip Criteria 1.2-1 1.2.2 ?lant Description 1.2-5 13 COMPARISON TABLES 13-1
- 13.1 Nuclear Steam Supply System Desip Characteristics 13-1 13.2 Engineered Safety Features Desip Characteristics 13-1 133 Containment Desip Characteristics 13-1 4 l
13.4 Structural Desip Characteristics 13-1
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13.5 Instrumentation and Electrical l Systems Design Characteristics 13-1 {
I' 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4-1 l
(\ 8904250093 890331
Amendment I j
MM 21A6100AC Remndard Plant Rev. n l
l CHAPTER 1 !
TABLE OF CONTENTS (Continued) h Section M Eggt 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1.5-1 1.6 MATERIAL INCORPORATED BY REFERENCE 1.6 1 1.7 DRAWINGS AND OTHER DETAILED INFORMATION 1.7-1 1.7.1 Piping and Instrumentation Diagrams 1.7-1 1.7.2 Electric, Instrumentation and Control Drawings 1.7 1 1 1.8 CONFORM ANCE WITH STANDARD REVIEW PLAN AND APPLICABILITYOF CODES AND STANDARDS 1.81 j 1.8.1 Conformance With Standard Review Plan 1.8-1 1.8.2 Applicability of Codes and Standards 1.8-1 1.8.3 Interfaces 1.8-1 .
1.9 INTERFACES 1.9-1 9 ::1 APPENDIX 1A RESPONSES TO TMI 1A-1 RELATED MATTERS 1A.1 Introduction 1Aul.1 1A.2 NRC Position / Responses 1A.2 1 1A.3 Interfaces 1A.3-1
. 1 1A.4 References 1A.4-1 l
i 1
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23A6100AC pe=A=rd Plant REV.A
.fo; SECTION 1.2 1' t'~j.
CONTENTS (Continued)
Sectlori M P_ase 1.2.2.8.2 FuelPoolCooling and Cleanup System (FPCS) 1.2-13 1.2.2.83 Makeup Water (Condensate) System 1.2-14 1.2.2.8.4 Plant Chilled-Water System 1.2 14 1.2.2.8.5 Process Sampling Systems 1.2-14 1.2.2.8.6 Plant Equipment and Floor Drainage 1.2-14 1.2.2.8.7 Service and Instrument Air Systems '1.2-14 1.2.2.8.8 High Pressure Nitrogen (HPIN) Gas Supply 1.2-14 Diesel Generator Fuel-Oil Storage and Transfer l 1.2.2.8.9 System 1.2-14 1.2.2.8.10 Heating, Ventilating, and Air Conditioning (Environmental) System 1.2-15 L2.2.6.11 Lighting Systems 1.2-15 1.2.2A12 Fire Protection System 1.2 15 1.2.2.9 Radiation Monitoring and Control 1.2-15 1.2.2.9.1 Process Radiation Monitoring 1.2-15 1.2.2.9.2 Area Radiation Monitors L2-16 1.2.2.93 Dust Radiation Monitors 1.2 16 1.2.2.9.4 Containment Atmospheric Monitors 1.2-16 1.2.2.10 Shielding 1.2-16 i
i 1.2-vi Amendment 1 l
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1.2-3b 14wegDrywell, Arrangement Elevation, Section 0/180 1.2-19a j 1.2-3c Wetwell, Arrangement Elevation Sections A-A & B-B 1.2-19b 1.2-4 Reactor Building, Arrangement Plan at El (-) 8200 mm 1.2-20 1.2-5 Reactor Building, Arrangement Plan at El (-) 1700 mm 1.2-21 1.2 45 Reactor Building, Arrangement Plan at El 4800 mm 1.2 22 ,
1.2-7 Reactor Building, Arrangement Plan at El 8500 mm 1.2 23 1.2-8 Reactor Building, Arrangement Plan at El 12300 mm 1.2-24 1.2-9 Reactor Building, Arrangement Plan at El 18100 mm 1.2 25 1.2-10 Reactor Building, Arrangement Plan at El 23500 mm 1.2-26 1.2-11 Reactor Building, Arrangement Plan at El 27200 mm 1.2-27 1.2-12 Reactor Building, Arrangement Plan at El 31700 mm 1.2-28 1.2-13 Reactor Building, Arrangement Plan at El 34200 mm 1.2-28 1.2-13a Drywell, Arrangement Plan at El 12300 mm 1.2-29 1.2-13b Drywell, Arrangement Plan at El 15600 mm 1.2-29a 1.2-13c Drywell, Anangement Plan at El18100 mm 1.2-29b 1.2 vii Amendment 6 L_____. . _ _ _ _ _ . _ _ . . _
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MM 23A6100AC REV.C Standard Plant SECTION 12 ILLUSTRATIONS h Figure 11tle P. age 1.2-14 Control and Servicef uilding, Arrangement Plan,Section 0/180 1.2-30 1.2 15 Control and Service guilding, Arrangement Plan, Section 270/90 1.2-31 1.2-16 Control and Service Building, Arrangement Plan, at Elevation (-) 13150mm 1.2-32 1.2-17 Control and Service Building, Arrangement Plan at Elevation (-) 7100mm 1.2-33 1.2 18 Control and Service Building, Arrangement Plan at Elevation (-) 1450mm 1.2-34 1.2-19 Control and Service Building, Arrangement Plan, at Elevation 2950mm 1.2-35 1.2-20 Control r.nd Service Building, Arrangement Plan at Elevation 7350mm 1.2-36 1.2-21 Control and Service Building, Arrangement Plan at Elevation 13295mm 1.2-37 1.2 22 Control and Service Building, Arrangement Plan at Elevation 19700mm 1.2 38 1.2-23 Radwaste Building 1.2-39 l
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COMPARISON OF ENGINEERED SAFETY FEATURES DESIGN CHARACTERISTICS This Plant GESSAR NMP2 Grand Gulf ABWR BWR/6 BWR/5 BWR/6 System /Comnonent 278471 238-748 251 764 251 4 00 Emergency Core Cooling Systems (stred on design power - section 6.3)
Im Pressure Core Snray Systems
- Number ofloops N/A 1- 1 1 flow rate N/A 6000 at 6350 at 7000 at (gpm) N/A 122 psid 128 psid 122 psid Hinh Pressure Core Sorny System **
Number ofloops 2 1 1 1 Flow rate 800 at 1550 at 1550 at 1650 at
[ (gpn) 1147 psid 1130 psid 1147 psid
( 1177 psid 3200 at 6110 at 6350 at 7000 at 100 psid 200 psid 200 pr.id 200 psid Pametor Core Isolation Coolina System (Subsection SA.6)
Flow rate 800 at 700 at 600 at 800 at
- (gpm) 165-1192 165-1192 1173 165-1192 psia psia psia psia reactor reactor reactor reactor Pressure Pressure Pressure Pressure Automatle Depressuration System Number of relief 8 8 7 8 valves
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1 MM 23A6100AC Standard Plant REV.C TABLE 1.3 2 COMPARISON OF ENGINEERED SAFETY FEATURES DESIGN CHARACTERISTIC (Continued)
' Ibis Plant GESSAR NMP.2 Grand Gulf ABWR BWR/6 BWRF / BWR/6 System /Comnonent 278-872 238-748 3115d 251 800 Imw Pressure Coolant Infection
- Number ofloops 3 3 3 3 Number es pumps 3 3 3 3 Flow rate 4200 at 7100 at 7450 at 7450 at (gpm/ pump) 40 psid 20 psid 26 psid 20 psid Auxiliary Systems Residual Heat Removal System (Subsection 5.4.7)
Reactor shutdown cooline mods Number ofloops 3 2 2 2 Number of pumps" 3 2 2 2 Flow rate 4200 7100 7450 7450 (gpm/purnp)
Duty (MBTU/hr*" 29.0 46.9 41.6 50.0 heat exchanger)
Number of heat 3 2 2 2 exchangers Primary contain- 4200 7100 7450 7450 ment cooling mode Flow rate (gpm)
ABIiR design referred to as Low Pressure Flooder The des!gn of the pumps is in part based on the required capacity during the reactor flooding mode.
Heat exchanger duty at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after reactor shutdown.
O Amendment 6 1.3-13
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MM 23A6100AC
' Standanl Plant ~ nev. c r
1(7' 1;6 MATERIALINCORPORATED L BY REFERENCE Table 1.6-1 is a _ list of all GE topical reports and any other report or document which is incorporated in whole or in part by reference in the ABWR SSAR.
1 -
Amendment 6 1.61 R_______________________,_______.._______________ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ . _ . _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ .
MM 23A6100Ac Standard Plant- am c SECTION 1.7 l
O CONTENTS 1- Section Ele Eage 1.7J PininundInstrumentation Diagrams 1.7-1 1.7.2 Electric. Instrumentation. and Control Drawinns 1.7-1 l SECTION 1.7 TABLES Table 31tle East 1.71 Piping and Instrumentation Diagrams 1.7-2 1.7-2 Interconnecting Block Diagrams, Instrumentation Elementary Diagrams and One Line Diagrams 1,7-5 I) ILLUSTRATIONS Figure 3111e East 1.71 Standard Advanced Boiling Water Reactor Symbols 1.7-6 1.7-2 GraphicalSymbols for Use inIBDs 1.7-8 1.7.i1 Amendment 6
1 23A6100AC '
' tandard Plant -
a w c. )
1.7. DRAWINGS
- 1.7.1 Piping and Instrumentation
- ' Diagrams Table 1.7-1 contains a list of system Piping and I
. Instrumentation diagrams provided in the ABWR SSAR. Figure 1.71 defines sysmbols used on piping )
and instrumentation disgrams (P&ID).
1.7.2 Electric, Instrumentation, and Control Drawings Interlocking block diagrams (IBD) and instrument elementary diagrams (IED) are listed in :
Table 1.7-2. Figure 1.7-2 defines the graphic symbols ;
used in the IBDs.
O O 1.71 Amendment 6
ABR 23xaooxc Standard Plant ne c Table 1.7-1
.s
\
PIPING AND INSTRUMENTATION DIAGRAMS
.(
GJ SSAR Fig. No . Page No. 'ntle Type 4.6-8 4.6-24 CRD System P&ID 4.6-9 4.6-26 ChL System PD 5.1-3 5.15 Nuclear Boiler System P&ID 5.2-5 5.2 41 Nuclear Boiler System P&ID 5.44 5.4-47 Reactor Recircula0m System P&ID 5.4-5 5A-48 Reactoc Recirculation System PD 5.4-8 5.4-f,1 Reactor Core Isolation Cooling System P&ID 5.4-9 5.4-53 Reactor Core Isolation Cooling System PD 5.4-10 5.4-55 Residual Heat RemovalSystem P&ID 5.4-11 5.4-59 Residual Heat RemovalSystem PD i 5.4-12 5.4-61 Reactor Water Clean-Up System P&ID
() 5.4-13 5.4-63 Reactor Water Clean-Up System PD 6.3-1 6S25 High Pressure Core Flooder PD 6}2 6126 Residual Heat Removal System PD 6.3-3 6127 Residual Heat Removal System PD 6.3-7 6.3-31 High Pressure Core Flooder P&ID 6.3-8 6.3-32 Reactor Core Isolation Cooling System P&ID 6*9 6.3 33 Residual Heat RemovalSystem P&ID 6.7-1 6.7-4 Nitrogen Gas Supply P&ID 9.11 9.1-23 Fuel Pool Cooling and Cleanup System P&lD 9.1-2 9.1-25 Fuel Pool Cooling and Cleanup System PD 9.21 9.2-26 Reactor Building Cooling Water System P&ID O Amendment 6 1.7-2
M 23A6100AC l Standard Plant no. c i Table 1.71 PIPING AND INSTRUMENTATION DIAGRAMS (Continued)
SSAR Fig. No. Page No. 'Iltle Type 9.2-2 9.2-35 HVAC Normal Coolitg Water System P&ID ,
i 9.2-3 9.2-37 HVAC Emergency Cooling Water System P&ID 9.2-4 9.2-39 Makeup Water System (Condensate) P&ID 9.2-4 9.2-40 Makeup Water System (Purified) P&ID 93 1 9 3-16 Standby Liquid ControlSystem P&ID 93 6 9 3-21 Instrument Air System P&ID 93-7 9 3-22 Servive Air System P&ID 9.4-1 9.4-7a Contrc,1 Building HVAC System P&ID 9.4-2a 9.4-7b Turbine Building Ventilation System P&ID 9.4-2b 9.4-7c Turbine Building Ventilation System P&ID 9.4-3 9.4-7d Secondr.ry Containment HVAC System P&ID 9.4-4 9.4-7e Essential Electrical Equipment HVAC System P&ID 9.4-5 9.4 7f Reactor Internal Pump Control Panel Room HVAC System P&ID 9.4-6 9.4-7g Essential Diesel Generator HVAC System P&ID 9.4-7 9.4-7h Serv:ce Building HVAC System P&ID 9.4-8 9.4-8 Drywell Cooling System P&ID 9.4-9 9.4-9 Drywell Heat Load Area Drawing P&ID 9.5-1 9.5-11 Suppression PoolCleanup System P&ID 10.2-4 10.2-14 Gescrator Hydrogen and CO System P&ID 2
103-1 103-5 Main Steam Supply System P&ID 103-2 103-6 Main Turbine System P&ID 10.4-1 10.4-26 Condenser Evacuation System P&ID O
Amendment 6 1,73
M$ 23A6100AC Standard Plant w.c Table 1.7-1.
PIPING AND INSTRUMENTATION DIAGRAMS (Continued)
SSAR Fig.No. Page No. Title Type 10.4-2 10.4-27 Turbine Gland Scaling System P&ID 10.4 3 10.4-28 Circulating Water System P&ID 10.4-4 10.4-29 Condensate Cleanup System P&ID 10.4-5 10.4-30 Condensate Storage & Transfer System P&ID 10.46 10.4-31 Condensate System P&ID 10.4-7 10.4-32 Feedwater System Pd, :D 10.4-8 10.4-33 LP Extraction Steam Drains and Vents System P&ID 10.4-9 10.4-34 HP Extraction Steam Drains and Vents System P&ID l O Amendment 6 1.7-4
M 23A6100AC Standard Plant m.c Table 1.7-2 INTERCONNECTING BLOCK DIAGRAMS, INSTRUMENTATION ELEMENTARY DIAGRAMS AND ONE LINE DIAGRAMS SSAR Fig. No. Page No. 'Iltle Type 7.3-1 7.3-52 High Pressure Core Flooder System IBD 7.3-2 7.3-59 Nuclear Boiler System IBD 7.3-3 7.3-74 Reactor Core Isolation Cooling System IBD 7.3-4 7.3-90 ResidualHeat RemovalSystem IBD 7.3-5 7.3-104 Leak Detection and Isolation System IBD 7.3-6 7.3-105 Standby Gas Treatment System IBD 7.3-7 7.3-116 Reactor Building Cooling Water System IBD 7.3-8 7.3 132 Essential HVAC System IBD 7.3-9 7.3-133 HVAC Emergency Cooling Water System IBD 7.3 10 7.3 141 High Pressure Nitrogen Gas System IBD 7.4-1 7.4-19 Standby Liquid ControlSystem IBD 7.4-2 7.4-24 Remote Shutdown System IED 7.4-3 7.4-26 Remote Shutdown System IBD 7.6-1 7.6-24 Neutron Monitoring System IED l
l 7.6-2 7.6-28 Neutron Monitoring System IBD 7.6-5 7.6-46 Process Radiation Monitoring System IED 7.6-6 7.6 59 Fuel Pool Cooling and Cleanup Sy>.cm IBD 7.6-7 7.6-64 Containment Atmosphere Monitoring System IED 7.6-8 7.6-69 Containment Atmosphere Monitoring System IBD 7.7-2 7.7-42 Rod Control and Information System IED 7.7-3 7.7-45 Rod Control and Information System IBD I
l l
Amendment 6 1.7-5 l l
MM 23A6100AC Standard Plant am c Table 1.7-2 INTERCONNECTING BLOCK DIAGRAMS, O'
L INSTRUMENTATION ELEMENTARY DIAGRAMS AND ONE LINE DIAGRAMS (Continued)
SSAR Ms.No. Page No. Title Type 7.7-4 7.7-54 Control Rod Drive System IBD 7.7-5 7.7-61 Recirculation Flow ControlSystem IED 7.7-7 7.7-64 Recirculation Flow ControlSystem IBD .
7.7-8 7.7-73 Feedwater ControlSystem IED 7.79 7.7-75 Feedwater ControlSystem IBD 83 8 S 30 Electrical Power Distribution System Single Line 83-2 8 3-31 6.9KV System Single Line 83-3 83 32 480 V System Single Line 83-4 83-33 Instrument Power Supply Single Line 83 5 8 S 34 Process Computer Constant Voltage Constant Frequency Power Supply Single Line
%J 8S6 8 3-35 Safety System Logic and Control Power Supply Single Line 83-7 8136 125 VDC Power System Single Line 818 83-37 250 VDC Power System Single Line i
l l
% Ama 6 1.7-5a i
23A6100AC Simndard Plant n .. c 3
.(a GRAPHIC EXPLANATION NO. FUNCTION OF FWCTON SYMBOL ,
y x SYMBOL INDICATES SIGNAL CONDITION OR ACTION IE.G. VALVE CLOSE SIGNAL).
1 CONDITION ( g ACTION SHIFTS TO RIGHT WHEN SYMBOL OR Y OR X CONDITION IS MET.
StGNAL ( Z l 5 >-+-
Y - 15 lNSTRUMENT NUMBER 2 - REPRESENTS THE NAME OF CONDITION SIGNAL 5 - SHOWS ABOVE OR BELOW SETPOINT" FOR TRANSFER OF SIGNAL CONDITION AND OUTPUT EXISTS IF AND & ONLY IF ALL SPECIFIED NPUTS ExtST A B C O O O 2A A --* O 1 O C
B--* 1 O O 1 1 1 f.
2-INPUT AND 2-@UT AND TRUTH TABLE A B C D 0 0 0 0 AM O O 1 0 2B B ~~* -* D ,
C --+- 1 0 0 0 3-NPUT AND 0
1 1 1 1 A --+ 3-INPUT AND B-> TRUTH TABLE C->
D-> '
4-INNT AND TRUTH TABLE 4-INPUT AND NoT SHoum Figure 1.7 2a GRAPHICAL SYMBOLS FOR USE IN IBDs Anw e' ment 6 1.7-8
MN 23A6100AC Standard Plant a,.. c O
GRAPHIC E XPL ANATION W. FUNCTION SYMBOL OF FUNCTION OUTPUT EXISTS IF AND & ONLY C,V& IF ALL SPECIFIED INPUTS EXIST 2D A*
(2 OF 3. 2 OF 4. OR 3 OF 4) co.ic:,6 ognT B -* 2/3 --* D A B C D
- 8N C-* O O O O
^ O O 1 0 2/3 AND 0 1 0 0 A --* 1 0 0 0 o 1 2E 2/4 -- > E D --> 1 1 0 1 2/4 AND TRUTH TABLE A --> 2/3 AND 0
2F C*
3/4 ->E 2/4 & 3/4 AND'S D --> TRUTH TABLES (NOT SHOWN) 3/4 AND OUTPUT EXISTS ONLY WHEN AT LE AST ONE INPUT EXIST A B C A-* O O O
^
B-+ 0 1 1 1 2-INPUT OR TRUTH TABLE 2- INPUT OR A-* A B C D l
3B C -* O 1 0 1 1 0 0 1 3-INPUT OR 0 1 1 1 1 0 1 1 1 1 0 1 1 1 1 1 TRUTH TABLE 3-INPUT OR Figure 1.7 2b GRAPHICAL SYMBOLS FOR USE IN IBDs Amendment 6 1.7-9
._ o
21A6100AC Standard Plant am e y.
O q I
NO. FUNCTION GRAPHIC ' EXPLANATION SYMBOL OF FUNCTION -'
1 THESE LOGIC SYSBOLS REPRESENT AN EXCLUSfvE "0R" GATE WHOSE OUTPUT ASSUMES 1 ST ATE IF ONE ANO X OR ONLY ONE OF THE LOGIC NPUT ASSUMES (EXCLUStV'6 "OR" ) THE 1STA1E
. A -+
3C C A B C B--* O O O O i 1 1 0 1 1 1 0
)( OR TRUTH TABLE 2 INFUT p_ p EXCLUSfvE "OR"
' N.Y A B U~ 0 0 0 0 0 0 0 1 1 0 1 0 1 1 0 0 1 3 imeuT xon 1 O i 0 0 1 1 0 1 1 0 0 l 1 1 1 0 TRUTH TABLE 3-INPUT E XC..LUSIVE OR
/~'
( Figure 1.7-2c GRAPHICAL SYMBOLS FOR USE IN IBDs Amendment 6 1.7-10
23A6100AC Standard Plant me.. c ,
O NO. M TON GRAPHIC EXPLANATION SYMBOL OF FUNCTION A* ABC0 E 3E EXCLUS,1vE O l "OR' 0%
0010 1 0100 1 1000 1 !
001 1 0 4 i n: pot g 1001 0 01 10 0 1010 0 1 100 0 'l 01 1 1 0 101 1 0 1 101 0 1 1 10 0 1 1 1 1 0 TRUTH T ABLE 4-lNPUT EXC..L USIVE OR" THIS SYLIBOL SHOWS THE 4 NOT A B "NOT" CONDITION. OUTPUT B 15 OPPOSITE TO INPUT A AB 01 10 Figure 1.7 2d GRAPillCAL SYMBOLS FOR USE IN IBDs Amendment 6 1.7 11
23A6100AC l Shndard Plant nev. c 1
I l
' (sl NO' FUNCT M GRAPHIC .
EXPLANATION SYMBOL OF FUNCTION TPU TPU - SIGNAL B 15 ENERGl2ED SA TNER A
] rB WITHIN SPECIFIED TME LNIT (t)
AFTER SIGNAL A IS ENERGl2ED.
ELEMENTS jt src B TERMINATES WHEN A
'"^ '
DELAYED NTIATION A
- - i d B
100 TDO - NTIALLY B IS ENERGl2ED 5B A
] rB WHEN A 15 ENERGl2ED.
SIGNAL B IS DE-ENERGl2ED jt EE WITHIN SPECIFIED TME LNIT (t) l AFTER SIGNAL A 15 DE-ENERGl2ED.
l DELAYED TERMINATION l i
(RESE T) A e
\j B
- -- t d C
WIPE-OUT WHEN SIGNAL C IS NOT PRESENT, 1'
6A (SIGNAL SIGNAL A IS TRANSMITTED TO B.
BLOCK) A - tWO) :- B WHEN SIGNAL C 15 PRESENT, SIGNAL A 15 STOPPED AND DOES (
NOT FLOW TO B. (WO. WIPE-OUT). !
1 ACB j 000 010 101 1 1O ,
l 1
Figure 1.7-2e GRAPHICAL SYMBOIE FOR USE IN IBDs Amendment 6 1.7 12 l
1
M 23A61C0AC Standard Plant new c
'CRAPHic EXPLANATION O
NO. FUNCTION SYMBOL OF FUNCTION DELAYED THE OUTPUT SIGNAL TO B IS 6B WIPE-OUT A STOPPED ArTER TIME (ONE-SHOT) TPU (WOla *8 INTERVAL "t".
7 t SEC.
A I I B
l*- t -*l SELF-HOLD A
- O s
0 WHEN CONDITION C DOES NOT 7/s OR EXIST, CONDITION A HOLDS RESET two)- ITSELF AND THERE IS OUTPUT TO B. THE SELF HOLDING IS 4 RELEASED WHEN CONDITION C C
15 ESTABLISHED AND THERE IS AN OUTPUT TO B ONLY WHEN THERE IS AN A CONDITION IA TAKES PRIORITY).
A--*
'IWOl e B WHEN CONDITION C DOES NOT 78 { EXIST, CONDITION A HOLDS j j
ITSELF AND THERE IS OUTPUT TO B. THE SELF HOLDING IS l
C RELE ASED WHEN CONDITION C 15 ESTABLISHED AND THERE IS i NO OUTPUT TO B.
IC TAKES PRIORITY) l l
Figure 1.7 2f GRAPHICAL SYMBOLS FOR USE IN IBDs Amendment 6 1.7 13 .
l
\ . __- _ _ - .
i M 23A6100AC i
hadard Plant m.v. c f~).
ij NO. FUNCTION GRAPHIC EXPLANATION SYMBOL OF FWCTION 5 - P s. A c E OF stJSTAu.ATIOM X - SWITCH OPERATION NAME 5 Y - SWITCH TYPE, E.G.
X l' OPER CS - CONTROL SWITCH 8 w H 2 --+ SPRING RETURN Y Z
- COS - CONTROL 2 --+ OPERATING SWITCH POSITION HOLD PBS - PUSHBUTTON
?umootun1swmmMno kkt. KEY SWITCH Tn (SPRING RETURN)
KOS - KEY OPERATING SWITCH (POSITION HOcD)
CRT- CRT Touc.H-EcAceAl 2 - SWITCH POSITION 04, DFF, P0t1 HOLO, ,,ETC X
(' V k.-) % 2 SHOWS A COMPONENT OR
-. 2 DEVICE TO BE CONTROLLED X - PART I OF CONTROLLED X DEVICE CONTROL COMPONENT Y Y - CONTROLLED DEVICE 9 NAME OR % 7 DEVICE - E.G. PUMP, VALVE, ETC.
2 2 - CONTROLLED CONDITION E.G. ST ART STOP, ON, OFF,ETC.
OPCN, CLO>E M ELECTRO _* ' E FULLY OPEN THIS SYMBOL REPRESENTS MAGNETIC AN ELECTROMAGNETIC
--* DE FULLY CLOSE VALVE VALVE E - ENERGl2ED DE - DE-ENERGIZED l
i p.
Figure 1.7 2g GRAPHICAL SYMBOIE FOR USE IN IBDs (v) .
l Amendament 6 1.7-14 l
23A6100AC Standard Plant m.y. e
\ NO. FUNCTION GRAPHC SYMBOL EXPLANATION OF FUNCTION
' ' TH:S SYMBOL REPRES!NTS ELECTRO MAGNETIC p AN ELECTROMAGNETIC U PILOT VALVE PILOT -
- E .,
A h DE **
E-- ENERGl2ED DE-- DE-ENERGl2ED
^H'S M :
A ---+ 5 --*C S REPRESENTS " SET MEMORY" MEMU:.y R REPRESENTS RESET MEMORY,,
l (FLIP-FLOP) LOGIC OUTPUT C EXISTS
[
B ---+ R --*D*
WHEN LOGIC INPUT A EXISTS.
'1 C CONTINUES TO EXISTS REGARDLESS OF SUBSEOUENT
- OUTPUT D SHALL STATE OF A AND UNTIL RESET NOT BE SHOWN SY NPUT AT B.
IF NOT USED.
C REMAINS TERMINATED REGARDLESS OF SUBSEOUENT STATE OF B, UNTIL A CAUSES MEMORY TO RESET.
LOGC OUTPUT D, IF USED, EXIST WHEN C DOES NOT /
EXIST, AND D DOES NOT EXIST WhEN C EXISTS.
STATC TH.S DEVICE CONVERTS "E" 13 TRANSDUCER 'E/P H (ELECTRCAL SIGNAL) TO CONVERTER)
P (PNEUMATC SIGNAll ELECTRO- g lE/Pl -
SHOWS A PILOT ELECTRO-g MAGNETIC e MAGNETC VALVE FOR A PILOT A -I : E *, CONTROL VALVE. WHEN THE VALVE FOR PILOT ELECTROMAGNETIC DE
- CONTROL VALVE IS ENERGl2ED BY A SIGNAL FROM A, OPENING OF THE CONTROL VALVE IS ADJUSTED BY A SIGNAL FROM 5.
7 INTENTIONALLY LEFT BL ANK 23 FOR FUTURE ADDITIONS Figure 1.7 2h GRAPHICAL SYMBOLS FOR USE IN IBDs Amendment 6 1,7 13
MM 2SA6100AC Standard Plant nev. c f r~S GRAPHIC . EXPL ANATION NO. FETON SYMBOL OF FUNCTION
- " ON 5gg : NDICATES ELECTRICAL SIGNAL 15 og AND FLOW DIRECTION c c c : INDIC ATES PNEUMATIC LNE AND FLOW DIRECTION
', NDICATES HYDRAULIC PRESSURE LNE AND FLOW DIRECTION
- : : : NDICATES MECHANICAL LNKAGE SIGNAL IS CONNECTED ELECTRICALLY ELEC1RICAL i& SIGNAL CONNECTON
^
a% SIGNAL IS NOT ELECTRICALLY CONNECTED O
17 SIGNAL NPUT M THIS SYMBOL REPRESENTS AN NPUT SIGNAL TO A COMPUTER, N DISPL AY. TEST PANEL, ETC AS DESIGNATED BY THE LETTER X NSIDE THE TRIANGLE. THE LETTER N INDICATES THE ASSIGNED SIGNAL NUMBER.
.D is OPE C Qo NAL W fyES
^"
TT IS USED N SEQUENTIAL !
A" VALVE \ _ O CONTROL j "ULLY
\F OPEN/ ~W l
Figure 1.7 21 GRAPHICAT FMBOIE FOR USE IN IBDs Amendment 6 1.7 16
Mb 23A6100AC Standard Plant m.m e NO. FUNCTON GRAPHIC EXPLANATCN SYMBOL OF FUNCTION l
VIRTUAL y, USED ONLY FOR SIGNALS WHICH DO NOT 19 CONDITION < > ACTUALLY EXGT FOR THE SYSTEM BUT
" ~ ~' "
SIGNAL ARE CONVENIENT TO SHOW ON 180.
PANeuNtr RED NDICATION LIGHT:
20 IN OiC.M TOR.
u t,HG h SHOWS ACTUATON, NPUT AND VALVE OPENING GREEN INDICATION LIGHT:
h SHOWS STOP. INTERRUPTION AND VALVE CLOSURE WHITE IMILK-WHITE) NDICATION LIGHT; b SHOWS CONDITON NDICATION , AUTOMATIC MDDE. 09ER.RTIOPJ... ETC.
ORANGE NDICATION LCHT:
@ SHOWS CAUTION AND FAILURE h COLD 4LEn OR.ffAl6PNtfMT lWicRTiur uvwr c.RT t.tGhr moscAron ro 6d' setem ora e,s.r.
lucicMT*O(r- X AtmeSCMTS T+4 c.o@R. oP Tite 4.;irtif i tretts
@'"T Td sE iootcerreo.
INDICATES AN ANNUNCIATED ALARM qwg A OR WARNING. THE LETTER N j N INDICATES THE ALARM NUMSER X
THIS SYMBOL REPRESENTS THAT THE INPUT SIGNAL SHALL BE DIVSCNALLY ISOLATED at SOLATOR FROM THE OUTPUT SIGNAL y
X - NPUT DIVSCN NUMSER ,
Y - GOLATED OUTPUT DIV!SCN NUMBER p THESE SYMBOLS NDICATE SIGNAL TRANSFER
]
SIGNAL W TO OTHER LOCATONIS).THE UPPER HAL.F OF THE SYMBOL $ USED TO ENTER THE TRANSFER CODE. THE LOwCR LEFT PORTION OF SYMBOL I
33 l TRANSFER S USED TO REFERENCE THE SHEET NUMSER TO GO TO. AND THE LOWER RIGHT HAND PORTCN WILL NDICATE THE LOCATCN WHERE THE SIGNAL CAN BE FOUND. THE TRANSFER CODE SHALL UTILIZE EITHER AN ENGL!SH LETTER OR A NUMBER F SIGNAL TRANSFER S WITHN THE SAME SHEET OR TO OTHER SHEETS OF THE i IBD. FOR SIGNAL TRANSFER FR014/OR TO OTHER ;
MPL SYSTEMS. THE TRANSFER CODE SHALL BE i EXPRESSED WITH 2 ENGLSH LETTERS STARTNG i WITH "AA". ALSO. NDICATE THE SYSTEu MPL l REFERENCE WHERE THE SIGNAL GOES TO OR ORIGNATES EXT TO THE SYM80L.
Figure 1.7 2j GRAPHICAL SYMBOLS FOR USE IN IBDs Amendment 6 1.7 17
r-M 23A6100AC Standard Plant me.. c l
l y-n U
l' NO' FUNCTON GRAPHIC EXPLANATON SYMBOL OF FUNCTION
/ L --- LMT OFF ILSI LIT) -- LMT OFF WITH TOROUE L BACK-UP l MOV'S oPtg TL -- BOTH LMT AND TOROUE OFF 34 POSITION __ ,g,g3, g MOV ( T --- TOROUE OFF
^'
DES NAT 5 TLJ TLIC) - WITH CHATTERNG
_ gggi PREVENTION AT TL ITS)
\
TIC) - WITH CHATTERNG (SEE PREVENTION AT T APPENDIX THE ABOVE LETTER DESIGNATORS
'D FOR LS: LirntT 3g ARE USED TO SHOW THE CONTROL APPLICATION EXAMPLES)
TS1 Tot.OUC W METHODS OF MOTOR-DRIVEN VALVES.
THE CONTROL METHOD SHOULD BE NDICATED ABOVE THE LEFT SIDE OF THE COMPONENT BLOCK FOR VALVE "OPENNG" AND BELOW THE LEFT SDE FOR VALVE " CLOSURE" A --+
35A COMPARATOR ^*B
- 7)-
i
(./
B --*
THESE SYMBOLS REPRESENT A COMPARATOR THAT PROVIDE AN OUTPUT WHEN THE CONDITION
- ~ A B B OR A s B 15 ME T.
358
- B --+ AsR AC OR DC AC 4 OR 36A E D ER HIS SYMBOL REPRESENTS A DR VER STANDARD ST ANDARD LOAD DRIVER.
& OR LOAD d
- E I 1/
368 b DE /3 DRIVI ?j THIS SYMBOL REPRESENTS A i ISOLATED LOAD DRIVER WHOSE OUTPUT !
LOAD DRIVER POWER SIGNAL Div 315 ISOLATED NTERNALLY FROM THE NPUT LOGIC SIGNAL DIVISION 1
/~N F[sure 1.7 2k GRAPHICAL SYMBOIE FOR USE IN IBDs
'\-)
Amendment 6 1.7 18
F- .
l' 23A6100AC Singlard Plant am
- cx
\._-) - SECTION 1.8 CONTENTS Section 11112 fagt 1.8.1 Conformance With Standard Review Plan 151 1.8.2 Applicability of Codes and Standards 1.8-1 1.83 Interfaces 1.8-1 TABLES Table Illlt Eagt 1.8 1 Summary of Differences From SRP Section 1 1.8-2 1.8-2 Summary of Differences From SRP Section 2 1.8-3 1.8-3 Summary of Differences From SRP Section 3 1.8-4 1
O 1.8-4 Summary of Differences From SRP Section 4 1.8-5 1.8-5 Summary of Differences From SRP Section 5 1.8-6 1.8-6 Summary of Differences From SRP Section 6 1.8-7 1.8-7 Summary of Differences From SRP Section 7 1.8-8 1.8-8 Summary of Differences Fr om SRP Section 8 1.8-9 1.8-9 Summary of Differences From SRP Section 9 1.8 10 1.8 10 Summary of Differences From SRP Section 10 1.8-11 1.8-11 Summary of Differences From SRP Section 11 1.8-12 1.8-12 Summary of Differences From SRP Section 12 1.8-13 1.8-13 Summary of Differences From SRP Section 13 1.8-14 1.8-14 Summary of Differences From SRP Section 14 1.8-15 1.8-15 Summary of Differences From SRP Section 15 1.8-16 l
/.
1.8-il i
Amendneemt 6 h
i
ABM 23A6100AC Standard Plant ._ REV.H SECTION 1.8 TABLES (Continued)
Table Iltle Eage 1.8 16 Summary of Differences From SRP Section 16 1.8-17 1.8 17 Summary of Differences From SRP Section 17 1.8-18 1.8-18 Summary of Differences From SRP Section 18 1.8-19 1.8-19 Standard Review Plans and Branch Technical Positions Applicable to ABWR 1.8-20 1.8-20 NRC Regulatory Guides Applicable to ABWR 1.8-38 1.8-21 Industrial Codes and Standards Applicable to ABWR 1.8-51 O
i I
O 1.8-iii Amendment 2 ,
I
.i
)
MD 23A6100AC se==derd Plant REV.C l
- c. . j 1.8 CONFORMANCEWITH STANDARD j
.?' V) ' REVIEW PLAN AND APPLICABILITY OF i CODES AND STANDARDS IJ.1 ConformanceWith Standard Review Plan The subsection provides the information required by 10 CFR 50.34(g) showing conformance with the Standard Review Plan (SRP). This information is presented by SRP section in Tables 1.81 through I.818. See Section 1.8.3 for interface requirements.
IJ.2 ApplicabilityofCodes and Standards Standard Review Plans, Branch Technical Positions, Regulatory Guides, and Industrial Codes and Standards which are applicable to the ABWR design are provided in Tables 1.819,1.820 and 1.8-21. Applicable revisions are also shown.
1.8.3 Interfaces A The SRP sections to be addressed for the
(,) remainder of the plant are those of Table 1.8-19 identified in the comment column, as " site specific" or
- utility".
l
?
'\
Amendment G l 8-1
ABM 23A61MAC Standard Plant m. c
! ; TABLE 1.8-1
SUMMARY
OF DIFFERENCES FROM SRP SECTION 1 Subsection l SRP Specific SRP Summary Description Where i Section Acceptance Criteria of Diffennee Discussed
)
i NONE l
t i
O Amendment 6 12-2
L Mb 23A6100AC Rev. c Standard Plant TABLE 1.8-2
SUMMARY
OF DIFFERENCES FROM SRP SECTION 2 O
Subsection l FRP Specific SRP Summary Description Where Section Acceptance Criteria of Difference Discussed 2.2.1- See Table 2.1 1. Limits imposed on selected SRP 2.1 2.2.2 Section II acceptance criteria by (1) the envelope of the ABWR 2.23 Standard Plant site parameters and (2) evaluations assumptions.
23.1 23.4 2.4.1 2.4.4 2.4.5 2.4.6 2.4.8 2.4.11.6 2.4.12 2.5.2.7 P
2.5.2.7 Minimum value of OBE OBE equal to one-third SSE.
equal to one-half SSE l
Ol\
Akundrnent 6 1.83 i
ABM ux62004c Er==dard Plant Rrv. c f)
W TABLE 1.8 3
SUMMARY
OF DIFFERENCES FROM SRP SECTION 3 Subsection SRP Specific SRP Summary Description Where Section Acceptance Criteria of DiNerence Discussed 3.6.1 II - Postulated pipe Large bore piping will utilize 3.6 and 3.63 and rupture, leak before break option as pro-3.6.2 vided in GDC-4 October 27,1987
' Modification of General Design Criterion 4".
3.73 II.2.b - Determination For equipment and components 3.73.2.2 of number of OBE other than piping,10 rather than cycles. 50 peak OBE stress cycles are used.
r k
j i
l l
1 l
t f~
(
Amiendment 6 1.84
MM 23A6100AC Standard Plant REV.C TABLE 1.8-4 g
SUMMARY
OF DIFFERENCES FROM SRP SECTION 4 i
Subsection l SRP Specific SRP Summary Description Where l Section Acceptance Criteria of Difference Discussed 4.4 II 7 - Requires a loose Loose parts monitor not used. 4.4.6 parts monitor per R.G.
1.133.
O l
9 Amendment 6 1.85 l
23A6100AC m- '--d Plant arv.c
[]
t v
TABLE 1.8-5
SUMMARY
OF DIFFERENCES FROM SRP SECTION 5 Subsection l SRP SpeclRc SRP Summary Description Where Section Acceptance Criteria of Difference Discussed 5.23 II.3.b.(3) - Regulatory Alternate position employed. 5.23.4.23
- Guide 1.71, Welding Qualification for Areas of Limited Accessibil-ity.
5.2.4 11.1 - Inspection of Some welds inaccessible for 5.2.4.2.2 Class 1 pressure con- volumetric examination, taining components.
5.4.6 11.3 - Unavailability Designed for less than two hours 5.4.6.1 of a c power. of operation.
5.4.7 Branch Technical Posi- No diversity ofinterlocks. 5.4.7.1.1.7 tion RSB 51, B.1.(b) and (c) - Diverse inter-p locks for RHR suction d isolation valves.
/'
\
Anwedment 6 1.84
23A6100AC Standard Plant Rev. c TABLE 1.8-6 g
SUMMARY
OF DIFFERENCES FROM SRP SECTION 6 Subsection l SRP Specific SRP Summary Description Where Section Acceptance Criteria of Difference Discussed 6.5.1 11 - Redundancy of fii. Single filter train. 6.5.13 3 ter train.
O l
Amendment 6 1.8-7
(
L 1
ABM DAM 00AC me==dard Plant arv. s
,3 V TABLE 1.8-7
SUMMARY
OF DIFFERENCES FROM SRP SECTION 7 Subsection l-SRP Specific SRP Summary Description Where Section Acceptance Criteria of Difference Discussed 7,1 Table 71: la RHR Annunciation at loop level. 73.23.2(1)
IEEE 279,4.19 73.2.4.2(1) 7.4.23.2(1) 7.1 Table 7-1: 2i Some modes of RHR are not auto- 73.23.2 (2)(b)
GDC 20 matic. 73.2.4.2 (2)(b) 7.4.23.2(1) 7.1 Table 71: 3a Clarification of Requirements. 73.2.1.2. (3)(a)
Reg Guide 1.22 7.1 Table 7-1: 3a HP/LP interlocks cannot be tested 7.6.23.2(3)
Reg Guide 1.22 during power operation.
7.1 Table 71: 3c Continuity testing of certain so- 73.2.1.2. (3)(c) 1m Reg Guide 1.53 lenoids.
% 7.1 Tabic 71: 3c Some components are not redun. 73.2.5.2(3)
Reg Guide l.53 dant. 7.4.2.2.2 (3) 7.1 Table 7-1: 3e Limited redundancy of Remote Shut- 7.4.2.4.2 (1)
Reg Guide 1.53 down. 7.4.2.4.2 (3) 7.1 Table 71: 3e Alternate positions employed. 7.1.2.10.5 Reg Guide 1.75 7.1 ' Table 7-1: 3e Exception of LOCA trip of FMCRD 7.4.2.1.2 (3)(a)
Reg Guide 1.75 feeder breakers.
7.1 Table 7-1: 3h Clarification of Requirements. 7.1.2.10.10 Reg Guide 1.118 7.1 Table 71: 3h Some sensors cannot be tested at 7.2.2.2.1 (7)
Reg Guide 1.118 power operation. 7.2.2.23.1.(10)
~7.2.2.23.1 (21) 7.1 Table 71: 41 Some actuators cannot be exer- 73.2.1.2 (4)(d)
BTP ICSB 22 cised during power optration. 7.4.23.2 (4)(c)
O
's '
l Amendeant 6 1.S-8
1 MM 23Af,100AC Rev. c Standard Plant TABLE 1.8-8 )
SUMMARY
OF DIFFERENCES FROM SRP SECTION 8 Subsection !
SRP Specific SRP Summary Description Wherr Section Acceptance Criteria of Difference Discussed 8.1 Table 81: 2f Exception to LOCA trip for cer- 8.13.1.2.2(6) j Reg Guide 1.75 tain non 1E loads. 83.1.4.2.2.4 l Appendix 9A 8.1 Table 8-1: 2f 15-ft cabic marking intervals. 83.13.1 Reg Guide 1175 8.1 Table 8-1: 2f LDS divisional separation in 83.1.4.2.2.2 IEEE 384 steam tunnel.
O O
Amendment 6 1.8-9 1 - - -
23A6100AC Standard Plant nrv. c
/
. q,<
TABLE 1.8.-9
SUMMARY
OF DIFFERENCES FROM SRP SECTION 9 Subsection l SRP Specific SRP Summary Description Where Section Acceptance Criteria of DifTerence Discussed 93.1- 11.1 - Particles shall Instrument air is filtered to 5 93.7.2 not exceed 3 microns. microns.
93.2 II.k.5 - Capable of sam- Cable of sampling liquids of 1 93.23.1 pling liquid of 10 ci/cm 3.
ci/cm3 9.4.1 GDC 19 Site specific. 6.4.73 1
Amundsncat 6 1.8-10
MM 23A6MAC Standard Plant REv.c TABLE 1.810
SUMMARY
OF DIFFERENCES FROM SRP SECTION 10 Subsection l SRP Specific SRP Summa.y Description Where Section Acceptance Criteria of Difference Discussed NONE O
l I
Oi Amendment 6 1.g.11 l J
23A6100AC
!te==dard Plant arv. c f%,
i) s, TABLE 1.811
SUMMARY
OF DIFFERENCES FROM SRP SECTION 11.
Subsection !
S_RP Specine SRP .
Summary Description Where Section Acceptance Criteria of Difference Discussed 11.1 11.9. BWR GALE Code Alternate computer code. 20.3 (Response to Question 460.1)
O l
..O Amendment 6 1.812
l 23A6100AC Standard Plant any. c TABLE 1.8-12
SUMMARY
OF DIFFERENCES FROM SRP SECTION 12 Subsection l SRP Specific SRP Summary Description W1:ere Section Acceptance Criteria of Difference Discussed l
I NONE i
l l
l O.
Amendment 6 g,g,33
r 23A6100AC R*==dard Plant nev. c TABLE 1.8-13
SUMMARY
OF DIFFERENCES FROM SRP SECTION 13 Subsection l SRP Specific SRP Sumn6ary Description Where Section Acceptance Criteria of Difference Discussed NONE O
O Amendment 6 1.8 14
i l
ABM 23A6100AC REv. c Standard Plant 1
TABLE 1.8-14
SUMMARY
OF DIFFERENCES FROM SRP SECTION 14 l
Subsection SRP Specific SRP Summary Description Where Section Acceptance Criteria of Difference Discussed l
l NONE I
i i
l l
l O
Amendment ' 1.8-15
1 i
ABM 23ssiooxc Standard Plav nev. c 1 TABLE 1.8-15 l
SUMMARY
OF DIFFERENCES FROM SRP SECTION 15
)
i Subsection l SRP Specific SRP Summary Description Where Section Acceptance Criteria of Difference Disen.erl
.15.1.1- Acceptable analytical 3d simulator instead of REDY code 15.1.1 3.2 15.1.4 model. is used. 3D simulator was ap-proved for use in GESTAR review by NRC.
15.2.6 All recirculation pumps Only half of RIPS are tripped. 15.2.6.1.1
- are tripped simulta. This is based on ABWR design.
neously by the initiat-ing event. l 15 3.1- Complete recirculation Trip of all RIPS is classified as 15.3.1.1.1 and 15 3.2 pumps trip is consid- a limiting fault. This is based 203.6 (Response to cred as a moderate- - on ABWR design. Question 490.113) frequency transient.
1533- II.10 - coincident tur- Not analyzed with the assump- 1533.2.2.
/,)
k-.
15 3.4 bine trip, loss of off- tion. If the assumption is made, site power and coast- the consequence would be similar down of undamaged to event shown in 15.2.6.
pumps.
15.4.2 Analysis of uncon- No quantitative analysis is pro- 15.4.2.2 troiIed eontroI rod vided because ABWRs ARBM design withdrawal at power. prevents this transient from oc-curring.
15.4.4- II.2.(b) - Fuel clad. MCPR not calculated since tran- 15.4.4 3 15.4.5 ding integrity sients are very mild. 15.4.5.3.2.1 a n d 15.4.5.3.2.2 15A.8 - No; applicable SRP for Discussion is provide to show 15.4.8 BWR. this event can not occur with ABWR FMCRD design.
154.9 Analysis of rod drop No quantitative analysis is pro- 15.4.9.1 & l5.4.9.2 accidenis vided because ABWR's FMCRD design prevent this accident from occur-ring.
i Amendment 6 1.8 16 l
l w__-___-__-.
ABM 2%6100AC Standard Plant REV.C l
TABLE 1.8-15 g '
SUMMARY
OF DIFFERENCES FROM SRP SECTION 15 (Continued)
Subsection l SRP Specific SRP Summary Description Where f Section Acceptance Criteria of Difference Discussed 15.6.5 II.2 - Use of assump- Suppression pool DF includes mix- 15.6.5.5 i tions outIined in ing in secondary containment and j Regulatory Guide 1.3 condensate without safety grade ,
recirculation system. Fission !
product plateout and removalin i steamlines and condenser. Reduc- l tion in primary containment leak- i age at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. !
15.8 10CFR50.62 requires au- The ABWR design does not need an 15.8.2 & 20.3.6 (Re-(Superseded tomatic SLCS SLCS to response to an ATWS sponse to Question by 10CFR event. 440.104) 60.62)
O O
Amendment 6 1.8-16a
iy .
Mandard Plant *[^[
TABLE 1.8-16
SUMMARY
OF DIFFERENCES FROM SRP SECI' ION 16 Subsection !
SRP Specific SRP Summary Description Where l Section Acceptance Criteria of Difference Discussed i
I i
l NONE i
O O
l Amendment 6 1.8-17
I l
- MM 23A6100AC Ata=dard Plant Rev. c ;
p
( -
TABLE 1.8-17 1 '. . 4
SUMMARY
OF DIFFERENCES FROM SRP SECTION 17 Subsection Specific SRP Summary Description Where l
SRP 1 Section . Acceptance Criteria of Difference Discussed l 17,1 11.1 - A p pli c a n t is GE & major technical associates 17.0 responsible for overall are responsible for their own OA 17.1.1 OA program programs. 17.1.2 17.1 II.2,3,4,7,13,17,18 Reg. Guide 1.28, Rev 3 and Table 17.0-1 "
Meet identified quality alternative positions employed. 17.1.2,17.1.3
' related Reg. Guides 17.1.4,17.1.7 17.1.13, 17.1.17 17.1.18 17.1 11.2 - Meet identified Differences between domestic and 17.1.3 regulations and codes international designs are identified in a controlled list.
/-
O Amendment 6 1.8 18
23A6100AC Standard Plant arv.c TABLE 1.8-18 f
SUMMARY
OF DIFFERENCES FROM SRP SECTION 18 Subsection Summary Description When l
SRP Specific SRP l Section Acceptance Criteria of Difference Discussed NONE O
O Amendment 6 1.8 19
23A6100AC -
handard Plant nev.c
. ,m .
(f TABLE 1.8-19 I Standard Review Plans and Branch Technical Positions Applicable To ABWR ABWR Appl. Issued Appli.
- k SRP No. Enya Dale cable? Comments Chanter 1 Introduction and General Description of Plant z.
1.8 Interfaces for Standard Design . 1 7/81 Yes .j i
Chanter 2 Site Characteristics j 2.1.1 Site location and Description 2 7/81 No Site specific l 2.1.2 ~ Exclusion Area Authority and Control 2 7/81 No Site specific '
2.13 Population Distribution 2 7/81 No Site specific 2.2.1- Identification of Potential Hazards in Site 2.2.2 Vicinity - 2 7/81 No Site specific R
- . 2.23 Evaluation of Potential Accidents 2 7/81 No Site specific 23.1 Regional Climatology 2 7/81 No Site specific 23.2 Local Meteorology 2 7/81 No Site specific 233 Onsite Meteorological Measurements Programs 2 7/81 No Site specific Appendix A 2 7/81 No Site specific 23.4 Short-Term Diffusion Estimates for Accidental Atmospheric Re: eases 1 7/81 No Site specific 23.5 long-Term Diffusion Estimates 2 7/81 No Site specific 2.4.1 Hydrologic Description 2 7/61 No Site specific Appendix A 2 7/81 No Site specific 2.4.2 ' Floods 2 7/81 No Site specific 2.43 Probable Maximum Flood (PMF) on Streams and Rivers 2 7/81 No Site specific p 2.4.4 Potential Dam Failures 2 7/81 No Site specific 2.4.5 Probable Maximum Surge and Seiche Flooding 2 7/81 No Site specific Amendmeni 6 1.8 20
- - -- i
7
[b 1'
L L ABM 23A6100AC l . Standard Plant arv. c n
TJ I' TABLE 1.819-SRPs and BTPs Applicable To ABMR (Continued)
ABWR Appl. Issued Appil.
SRP No. Em Dalt cable? Comments 12.2 Radiation Sources 2 7/81 Yes 12 3 -- Radiation Protection Design Features 2 7/81 Yes 12.4 12.5 Operational Radiation Protection Program 2 7/81 No Utility Chapter 13 Conduct of Operations 13.1.1 Management and Technical Support Organization 2 7/81 No Utility 13.1.2 - Operating Organization 2 7/81 No Utility 13.1 3 13 2 Training (Replaced by SRP Sections 13.2.1 and 13.2.2)
. [~T 13.2.1 Reactor Operator Training 0 7/81 No Utility 13.2.2- Training For Non-Licensed Plant Staff 0 7/81 No Utility 13 3 Emergency Planning 2 7/81 No Utility 13.4 Operational Review 2 7/81 No Utility 13.5 Plant Procedures (Replaced by SRP Sections 13.5.1 and 13.5.2) 13.5.1 Administration Procedures 0 7/81 No Utility 13.5.2 Operating and Maintenance Procedures 1 7/85 No Utility Appendix A 0 7/85 No Utility 1 3.6 Physical Security 2 7/81 Yes ABWR and Utility
.C,hanter 14 Initial Test Procrum 14.1 Initial Plant Test Programs PSAR (Deleted) 14.2 Initial Plant Test Programs FSAR 7/81 Yes
\q/
2 k
i 143 Standard Plant Design, Initial Test Program -
Final Design Approval (FDA) (Deleted)
Amendment 6 1.8-33 l
MM 23A6100AC Standard Plant Rev c 1< TABLE 1.819 SRPs and BTPs Applicable To AB%R (Continued)
ABWR Appl. Issued Appil-SRP No. h Dalg ghirl Comments Chapter 18 Human Factors Encineerine 18.0 Human Factora Engineering / Standard Review Plan Development 1 9/84 Yes
, 18.1 Control Room 0 9/84 Yes Appendix A 0 9/84 Yes 18.2 _ Safety Parameter Display System 0 11/84 Yes Appendix A 0 11/84 Yes
'D t
%)
Amendment 6 1.8-37
)
______-__- - - a
23A6100AC
- lEfe=And Plant . nev. c ;
TABLE 1.8 21-4 b . INDUSTRIAL CODES AND STANDARDS APPLICABLE TO ABWR Code or Standard Number Year 'I1tle ACI
+ 211.1 1981 Practice for Selecting Proportions for Normal, Heavy 3 Weight, and Mass Concrete.
212 1981 Guide for Admixturesin Concrete 214 :1977 Recommended Practice for Evaluation of Strength Test Results of Concrete 301 1984 Specifications for Structural Concrete for Buildings
.304 1973 Practice for Measuring, Mixing, Transporting, and Placing of Concrete
-305 1977- Recommended Practice for Hot Weather Concreting 306 1978 Recommended Practice for Cold Weather Concreting 307 1979 Specification for the Design and Construction of Reinforced Concrete Chimneys 308 1981 Practice for Curing Concrete 309 1972 Practice for Consolidation of Concrete 311.1R 1981 ACI Manual of Concrete Inspection 311.4R 1981 Guide for Concrete Inspection 315- 1980 Details and Detailing of Cocerete Reinforcement 318 1983 Building Code Requirements for Reinforced Concrete Code Requirements for Nuclear Safety-Related Concrete 349 1980 Structures 359 (See ASME BPVC Section III)
Asseadment 6
. _ . _ _ _ . - _ - _ ---__ - - - _ 0
- M 23A6100AC Standard Plant guy. c TABLE 1.8-21 (Con!!nued) g INDUSTRIAL CODES AND STANDARDS APPLICABLE TO ABWR Code or Standard Number Year Title AISC N690 1984 Specifications for the Design, Fabrication, and Erection of Steel Safety Related Structures for Nuclear Facilities SG-673 1986 Specification for the Design of Cold Formed Steel l Structural Members i e
i l
1 I
)
I t
O' Amendment 6 1.8-51a l
[
ABM 2mioo4C Standard Plant REV.C-X TABLE L8-21 (Continued) 0 -
INDUSTRIAL CODES AND STANDARDS APPLICABLE TO ABWR -
Code or Standard Number Year Title ANS 23 1983 Standard ft r Estimating Tornado and Other Extreme Wind Characteristics at Nuclear Power Sites 2.8 1981 Determining Design Basis Flooding at Power Reactor Sites I
5.1 1979 Decay Heat Powerin LWRs 18.1(N237) 1984 Radioactive Source Term for Normal Operation of LWRs 52.1 1983 Nuclear Safety Design Criteria for the Design of Stationary Boiling Water Reactor Plants 55.4 - 1979 Gaseous Radioactive Waste Processing Systems for Light Water Reactors
' 57.1 1980 Design Requirements for LWR Fuel Handling Systems 57.2(N270) 1976 Design Requirements for LWR Spent Fuel Storage Facilities at NPP 58.2 1988 Design Basis for Protection of Light Water NPP Against Effects of Postulated Pipe Rupture 59.51 (N195) 1976 Fuel Oil Systems for Standby Diesel-Generators I
O Amcodment 6 1.8 51b
_ _-_-_ - - - - _ _ _ - - - _ - - - _ - -- - A
l
..ABM 23A61MAC Standard Plant REV C j i
TABLE 1.8-21 (Continued) ]
INDUSTRIAL CODES AND STANDARDS AFPLICABLE TO ABWR i
Code or .
Standard I Number Year '!1tle j i
ANSI l A58.1 1982 Design Loads for Buildings and other Structures, Minimum
)
HEI 1970 Standards for Steam Surface Condenser,6th E., Heat l Exchangers Institute I l
MC11.1 1976 Quality Standard for Instrument Air N5.12 1972 Protective Coatings (Paint) for Nuclear Industry l
N13.1 1%9 Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities N14.6 1986 Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 Kg) or More for Nuclear Materials N510 1980 Testing of Nuclear Air-Cleaning Systems ;
i N101.2 1972 Protective Coatings (Paints) for Light Water Nuclear Containment Facilities 1
N101.4 1972 OA for Protective Coatings Applied to Nuclear l Facilities .
NB5 (See ANS 59.51) ;
N237 (See ANS 18.1 N270 (See ANS 52.2) l l
l O
Amendment 6 1.8-51c
~AB M 23461ooxc Standard Plant REN. C TABLE 1A-21 (Continned) 2[
INDUSTRIAL CODES AND STANDARDS
. APPLICABLE TO ABWR lc Code or l Standard Number Year Title API 620 1986 Rules for Design and Construction of Large, Welded, Low Pressure Storage Tanks 650 1980 . Welded SteelTanks for OilStorage O
O Am**dment 6
ABM ur61mic Standard Plant REV.C TABLE 1.8-21 (Continued) e INPi.STRIAL CODES AND STANDARDS APPLICABLETO ABWR Code or Standard Number Year Title ASHRAE 30 1978 Methods of Testing Liquid Chilling Packages 33 1978 Methods of Testing Forced Circulation Air Cooling and 4 Air Heating Ceils 1
O
)
O Amendment 6
I! & y3yg
!/ Remndant Plant Rev. c _ q bP TABLE 1.8 21(Continued)
+
INDUSTRIAL CODES AND STANDARDS APPLICABLE TO ABWR -
Code er Standard Number Year 11tle ASME o
B30.2 19E3 Overhead and Gantry Cranes B30.9 1984 Slings B30.10 1982 Hooks B30.11 1980 - Monorails and Underhung Cranes l
~ B30.16 1981 Overhead Hoists i B31.1: 1986 ' Power Piping B%.1 1986 Specification for Welded Aluminum-Alloy Storage Tanks N45.2 1977 OA Program Requirements for Nuclear Power Plants N45.2.1 1980 Cleaning of Fluid Systems and Associated Components During construction of Nuclear Power Plants
)
1
)
I A=*adment 6 1.s.31r m _ _ - _ _ _ _ _ - _ _ _ _ _ _ . - - - .
M 23A6100AC Standard Plant REV.C TABLE 1.8-21 (Continued)
INDUSTRIAL CODES AND STANDARDS APPLICABLE TO ABWR Code or Standard Number Year Title ASME N45.2.2 1972 Packaging, Shipping, receiving, Storage, and Handling of Items for Nuclear Power Plants (During Construction Phase)
N45.2.6 1978 Qualifications of Inspection, Examination and Testing Personnel for the Construction Phase of Nuclear Power Plants N45.2.9 1979 Requirements for the Collection, Storage, and Maintenance of OA Records for Nuclear Power Plants N45.4 1972 Leakage Rate Testing of Containment Structures for Nuclear Reactors NOA-1 1983 Quality Assurance Program Requirements for Nuclear Pacilities NOA-1A 1983 Addenda to ANSI /ASME NOA-11986 NOA-2 1983 Quality Assurance Requirements for Nuclear Power Plants I
I l
1 l
\
1 O
Amendment 6 1.8-51g l
-n
-AB M u46i=4c se- da d Plant nev. c TABLE 1.8-21(Continued) fM V' '
INDUSTRIAL CODES AND STANDARDS APPLICABLE TO ABWR Code or Standard Number Year litle ASME Sec11 1986 EPVC Section II, Material Specifications SecIII 1986 BPVC Section III, Rules for Construction of Nuclear Power Plant Components Sec VIII 1986 BPVC Section VIII, Rules for Construction of Pressure Vessel SecIV 1986 BPVC Section IX, Qualification Standard for Welding and Brazing Procedures Welder, Brazers and Welding and Brazing Operators Sec XI 1986 BPVC Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components ASTM (See ASME BPVC Section III)
I
,y l O Amendment 6 1.8-51h j l
-ABM 2346ioore Standard Plant REV.C TABLE 1.8-21(Continued) g INDUSTRIAL CODES AND STANDARDS APPLICABLE TO ABWR Code or Standard Number Year Title AWS A4.2 1986 Procedures for Calibrating Magnetic Instruments to Measure the Delta Ferrite content of Anstenitic Stainless Steel Weld Metal !
l D1.1 1986 SteelStructuralWelding Code i i
O 1
/
)
O' Amendment 6
)
1.8-5 11
23A6100AC mem-d-ed Plant arv.c
/G TABLE 1.8-21(Continued)
, O INDUSTRIAL CODES AND STANDARDS APPLICABLE TO ABWR Code or Standard Number Year 11tle AWWA D100 1984 Welded SteelTanks for Water Storage CMAA70 1983 Specification for Electric Overhead Traveling Cranes O
O Amendment 6 1.8-51J
23A6100AC r Standard Plant any.c TABLE 1.8 21 (" Continued)
INDUSTRIAL CODES AND STANDARDS APPLICABLE TO ABWR Code or Standard Number Year Title IEEE 279 1971 Criteria for Protection Systems for NGPS 308 1980 Criteria for Class 1E Power Systems for NPGS 317 1983 Electrical Penetration Assemblies in Containment Structures for NPGS 323 1983 Qualifying class 1E Equipment for NPGS 334 1974 Motors for NPGS, Type Tests of Continuous Duty class 1E 338 1977 Criteria for the Periodic Testing of NPGS Safety Systems 344 1987 Recommended Practices for Seismic Qualifications of _
Class 1E Equipment for NPGS 379 1977 Standard Application of the Single Failure Criterion to NPGS Safety Systems 382 1985 Qualification of Actuators for Power Operated Valve Assemblies with Safety Related Functions for NPP i 383 1974 Type Test of Class 1E Cables; Field Splices and Connections for NPGS i 384 1981 Criteria for Independence of Class 1E Equipment and Circuits i
387 1984 Criteria for Diesel-Generator Units AppEed as Staudby 1 Power Supplies for NPG3 450 1987 Practice for Maintenance, TetAing, and Replacement of i trage lead Stenge Batteries for Generating; Stations ard Substations 4M 1987 Recommended Practice for the Installation Design and Installation of Large Lead Storage Batteries for NPGS O
Amendment 6 1&51k
. 3r M 23A6100AC Ren=dard Plant REV.Q g TABLE 1.8-21 (Continued) b INDUSTRIAL CODES AND STANDARDS APPLICABLE TO ABWR Code or Standard Number Year 'Ilt'e IEEE (Con't) 485 1983 Recommended Practice for Sizing Large Lead Storage Batteries for NPGS 944 1986 Recommended Practice for the Application and Testing of Uninterruptable Power Supplies for Power Generating Station IPCEA P-46-426 - 1982 Ampacities Including Effect of Shield Losses for Single Conductor Solid-Dielectric Power Cable 15kV through 69kV P-54-440 1987 Ampacities of Cables in Open-Top Cable Trays 8-66-524 1982 Cross Linked Thermosetting Polyethylene Insulated Wire and Cable for Transmission and Distributor of Electrical Energy
(
l I
i l
l- l l.
]
\
1 Amendment 6 1.8-5 11
M 23A6100AC Standard Plant REV.C TABLE 1.8-21(Continued)
INDUSTRIAL CODES AND STANDARDS APPLICABLE TO ABWR Code or Standard Number Year Title NFPA 10 1981 Portable Fire Extinguishers - Installation 10A 1973 Portable Fire Feinguishers - Maintenance and Use 12 1985 Carbon Dioxide Extingniching Systems 13 1985 Installation of Sprinklers Systems 14 1986 Installation of Standpipe and Hose Systems 24 1984 Private Service Mains and their Appurtenances 37 1984 Stationary Combustion Engines and Gas Turbines 70 1987 Handbook 1987 72D 1986 Proprietary Protective Signaling Systems 78 1986 Lightning Protection Code 80 1986 Fire Doors and Windows Installation of Air Conditioning and Ventilating 90A 1985 Systems l
91 1983 Blower and Exhaust Systems
.' ;, 101 1985 Life Safety Code 251 1985 Fire Test, Building Construction and Materials 252 1984 Fire Tests, Door Assetrblies 255 1984 Building Materir.ls, Test of Surface Burnirg Characteristics 321 1987 Classification of Flammable Liquids 801 1986 Facilities Handling Radioactive Ma'.crials O
Amendment 6 1&51m ,
i
pe
,f M 23A6100AC mandard Plant - _ _? c;v y N, TABLE 1.8-21(Continved)
(y INDUSTRIAL CODES AND STANDARDS APPLICABLE TO ABWR Code or Standard
, Number Year litle NFPA (Con't) 802 1988 Nuclear Research Reactors 1%1 1979 Fire Hose 1963 1985 Screw Threads and Gaskets for Fire Hose Connections O
l I
O 1.8-51n
MM 23A6100AC 0" Standard Plant REV.C TABLE 1.8-21 (Continned)
INDUSTRIAL CODES AND STANDARDS APPLICABLE TO ABWR Code or Standard Number Year 'Iltle SSPC PA-1 1972 Shop, Field and Maintenance Painting PA 2 1973 Measurements of Paint Film Thickness with Magnetic Gages SP 1 1982 Solvent Cleaning SP-5 1985 White Metal Blast Cleaning SP-6 1986 commercid Blast Clearimg SP-10 1985 Near-WW P!.ast Cleaning TEMA C 1978 Standards of Tubular Exchanger Manufactures Association UIA4 1983 Rubber-Insulated Wires and Cables O
O Amendment 6
~
23A6100AC ReamAard Plant REV.C
~
4 SECTION 1.9
( .:
CONTENTS Section M Eage 1.9 INTERFACES 1.9-1 TABLES Table Illit Eage 1.9-1 Summary of ABWR Standard Plant Interfaces with Remainder of Plant 1.9-2 O.
4 i
l l
I 1.9 ..u Amendment 6
1 l
M 23A6100AC Standard Plant .
- nev. c 7'^X 1.9 INTERFACES -
' A_/
j.
The ABWR SSAR presents the ABWR Stan-dard Pbet dr.sitin ir.anporating the Nuclear Island Turbine Island and Radwaste Facility. Although this 3 scope is essentially a total plant, there are a modest h
number ofinterfaces between the ABWR Standard L Plant design and the remainder that must be ad-dressed by the parties that reference the ABWR design. The purpose of this sec cn is to identify the SSAR sections where der.criptions'of these interfaces are presented. These interfaces can be classified into one of the following types ofinterfaces:
(1) Design (2) Confirmatory (3) Procedual A design type interface is one in which the ABWR design requires specific design compatibility at the interface (e.g. ultimate heat sink capability). A confirmatory type interface is one in which the party referencing the ABWR design must confirm that the remainder of the plant comforms with a particular design basis (e.g. envelope of site characteristics).
Finally, a procedual type interface is one in which the
}p}
^^
party referencing the ABWR design must agree to provide specific procedures or programs to satisfy TMI'related requirements.
Each of the interfaces are summarized in Table 1.91 in the order they are presented in the SSAR.
An item number has been assigned to each interface
- to faciltate future identification. The subject of the interfa e is provided along with its interface type and corresponding subsection.
%d Amendment 6 1.91
M 23A6100AC Standard Plant nev c f)
%_ )
Table 1.9-1
SUMMARY
OF ABWR STANDARD PLANT INTERFACES WITH REMAINDER OF PLANT ITEM INTERFACE NO. SUBJECT TYPE SUBSECTION 1.1 Standard review plan sections for remainder of Confirmatory 1.83 i plant 1.2 Emergency procedures and emergency procedures Procedural 1A3.1 training program 13 Procedures for removing safety-related systems Procedural 1A3.2 from service 1.4 Inplant radiatiom monitoring Procedural 1A33 2.1 Envelope of ABWR Standard Plant Site Design Design & 2.2.1 Parameters Confirmatory 2.2 Standard Review Plan Site Characteristics Confirmatory 2.2.2
,A 3.1 Effect of remainder of plant structures, Confirmatory 3333 Q) systems and components not designed to tornado loads 3.2 - Flood Elevation Design 3.43.1 33 Ground Water Elevation Design 3.43.2 3.4 Protection of ultimat: heat sink Confirmatory 3.5.4.1 3.5 Missels generated by natural phenomena from Confirmatory 3.5.4.2 remainder of plant 3.6 Site proximity missels and aircraft hazards Confirmatory 3.5.43 3.7 Protection against secondary missels inside Confirmatory 3.5.4.4 containment 3.8 Summary of pipe break analysis results Confirmatory 3.6.4.1 3.9 leak-before-break analysis results Confirmatory 3.6.4.2 3.10 Equipment qualification records Confirmatory 3.10.5.1 l 1
l 3.11 Dynamic qualification report Confirmatory 3.10.5.2 !
,rs
\
8.1 Stability of offsite power system Confirmatory 8.1.4.1
('uY Amendmen 6 1.9-2 j l
1
23A6100AC Standard Plant nem c Table 1.91 g
SUMMARY
OF ABWR STANDARD PLANT INTERFACES WITH REMAINDER OF PLANT ITEM INTERFACE NO. SUBJECT TYPE SUBSECTION 8.2 Adequacy of Station Electric Distribution Confirmatory 8.1.4.2 Systein Voltages 83 Class IE Feeder Circuits Design 8.23.1 ,
8.4 Non-class IE Feeders Design 8.23.2 8.5 Specific ABWR Standard Plant / remainder of plant Design 8.233 power sysytem interfaces 9.1 Ultimate heat sink capability Design 9.2.14.1 9.2 Makeup water system capability Desige 9.2.14.2 93 Contamination of DG combustion air intakes Confirmatory 9.5.10.1 12.1 Regulatory Guide 8.10 Confirmatory 12.1.4.1 i 12.2 Regulatory Guide 1.8 Confirmatory 12.1.4.2 12 3 Occupational radiation exposure Procedural 12.1.4 3 12.4 Long-term training upgrade Procedural 19A3.1 19.1 Long-term program of upgrading of procedures Procedural 19A3.2 19.2 Purge system reliability Procedural 19A33 193 Licensing emergency support facility Procedural 19A3.4 19.4 In-plant radiation monitoring Procedural 19A3.5 l
19.5 Feedback of operating, design and construction Procedural 19A3.6 experience 19.6 Organization and staffing to oversee design and Procedural 19A3.7 Construction O
Amendment 6 1.93 l
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b 23A6100AC Standard Plant nev. c
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APPENDIX 1A TABLE OF CONTENTS Section Title EA22 1A RESPONSE TO TMI RELATED MATTERS 1A.1 INTRODUCTION 1A.1 1 1A.2 NRC POSITIONS / RESPONSES 1A.2 1 1A.2.1 Short Term Accident Analysis Procedure Revision [I.C.1(3)] 1A.2-1 1A.2.2 Control Room Design Revisions / Guidelines and Requirements [I.D.1(1)] 1A.2-2 1A.23 Control Room Design-Plant Safety Parameter Display Console [I.D.2] 1A.2-2 1A.2.4 Scope of Test Program-Preoperational jj s.-
and Iower Power Testing [I.G.1] 1A.2-2 1A.2.5 Reactor Coolant System Vents [II.B.1] 1A.2-3 1A.2.6 Plant Shielding to Provide Access to Vital Areas and Protect Safety Equipment for Post-Accident Operation [II.B.2] 1A.2-6 1A.2.7 Post Accident Sampling [II.B3] 1A.2-7 1A.2.8 Rule Making Proceeding or Degraded Core Accidents [II.B.8] 1A.2-8 1A.2.9 Coolet System Valves-Testing Requirements [II.D.1] 1A.2-8 1A.2.10 Relief and Safety Valve Position Indication [II.D3] 1A.2-8 1A.2.11 Systems Reliability [II.E3.2] 1A.2-8 1A.2.12 Coordinated Study of Shutdown Heat Remos .! Requirements [II.E33] 1A.2-8 1A.2.13 Containment Design-Dedicated
/^g Penetration [II.E.4.1] 1A.2-8 .
I ly l l l l 1A-ii Amendment 6
M 23A6100AC Standard Plant nev. c APPENDIX 1A O
TABLE OF CONTENTS (Continued)
Section M Eagt 1A.2.14 Containment Design-Isolation Dependability [II.E.4.2] 1A.2-8 1A.2.15 Additional Accident-Monitoring Instrumentation [II.F.1(1)] 1A.2-10 1A.2.16 Identification of and Recovery From Conditions Leading to Inadequate Core Cooling [II.F.2] 1A.2-10 1A.2.17 Instruments for Monitoring Accident Conditions [II.F3] 1A.2-11 1 1A.2.18 Safety Related Valve Position Indication [lI.K.1(5)] 1A.2-11 1A.2.19 Review and Modify Procedures for Removing Safety-Related Systems Front Service [II.K.1(10)] 1A.2-11 1A.2.20 Describe Automatic and Manual Actions for Proper Functioning of Auxillary Heat Removal Systems When FW System nc,t Operable [II.K.1(22)] 1A.2-11 1A.2.21 Describe Uses and Types of RV Level Indication for Automatic and Manual Initiation of Safety Systems
[II.K.1(23)] 1A.2-12 1A.2.22 Separation of HPCI and RCIC System 3 Initiation levels [lI.K3(13)] 1A.2-13 1A.2.23 Modify Break-Detection logic to Prevent Spurious Isolation of HPCI and RCIC Systems [II.K3(15)] 1A.2-13 1A.2.24 Reduction of Challenges and Failures of Relief Valves - Feasibility Study and System Modification [II.K3(16)] 1A.2-14 l
O(;
Amendment 4
i L~ ABWR zu6umc Standard Plant aw c 1
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SECTION 1A.2 j CONTENTS Section M East 1A.2.1 Short Term Accident Analysis Procedure Revision [I.C.1(3)] 1A.2-1 1A.2.2 Control Room Design Revisions / Guidelines and Requirements [I.D.1(1)] 1A.2-2 1A.23 Control Room Design-Plant Safety Parameter Display Console [I.D.2] 1A.2-2 1A.2.4 Scope of Test Prograns Preoperational and lower Power Testing [I.G.1] 1A.2-2 1A.2.5 Reactor Coolant System Vents [II.B.1] 1A.2-3 g 1A.2.6 Plant Shielding to Provide Access to Vital Areas and Protect Safety Equipment
\-)
(
for Post-Accident Operation [II.B.2] 1A.2-6 1A.2.7 Post-Accident Sampling [II.B3] 1A.2-7 1A.2.8 Rule Making Proceeding or Degraded Core Accidents [II.B.8] 1A.2-8 1A.2.9 Coolant System Valves-Testing RequirementsIII.D.1] 1A.2-8 1A.2.10 Relief and Safety Valve Position Indication [II.D3] 1A.2-8 1A.2.11 Systems Reliability [II.E3.2] 1A.2-8 1A.2.12 Coordinated Study of Shutdown Heat Removal Requirements [II.E33] 1A.2-8 1A.2.13 Containment Design-Dedicated Penetration [II.E.4.1] 1A.2-8 1A.2.14 Containment Design-Isolation Dependability [II.E.4.2] 1A.2-8 i
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1A.2-ii Assedment 6
MM 23A6100AC
( Standard Plant nev. c 9
SECTION 1A.2 CONTENTS (Continued)
Section Title East 1A.2.15 Additional Accideat-Monitoring Instrumentation [II.F.1(1)] 1A.2-10 1A.2.16 Identification of and Recovery From Conditions Leading to Inadequate Core Cooling [II.F.2] 1A.2-10 1A.2.17 Instruments for Monitoring Accident Conditions [II.F3] 1A.2-11 1A.2.18 Safety-Related Valve Position Indication [II.K.1(5)] 1A.2-11 1A.2.19 Review and Modify Procedures for Removing Safety Related Systems From Service [II.K.1(10)] 1A.2-11 1A.2.20 Describe Automatic and Manual Actions for Proper Functioning of Auxillary Heat Removal Systems When FW System not Operable [II.K.1(22)] 1A.2-11 1A.2.21 Describe Uses and Types of RV Level Indication for Automatic and Manual Initiation of Safety Systems
[II.K.1(23)] 1A.2 12 1A.2.22 Separation of HPCI and RCIC System Initiation Levels [II.K3(13)] 1A.2-13 1A.2.23 Modify Break-Detection Logic to Prevent Spurious Isolation of HPCI and RCIC Systems [II.K.3(15)] 1A.2-13 1 A.2.24 Reduction of Challenges and Failures of i Relief Valves - Feasibility Study and System Modification [II.K.3(16)] 1A.2-14
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1 1A.2-iii 9
Amendment 4
l M 23A6100AC Standard Plant nevc U 1A.2 NRC POSITIONS / RESPONSES GE has participated, and continues to par-ticipate,in the BWR Owners' Group program to de-1A.2.1 Short Term Accident Analysis velop emergency procedure guidelines for General Procedure Revision [I.C.1(3)] Electric BWRs. The resulting emergency procedure guidelines are generally applicable to the ABWR as NRC Position - are the transient and accident analyses. Following is a brief description of the submittals to date, and a in letters of September 13 and 27, October 10 justification of their adequtcy to support guideline and 30, and November 9,1979 (References 4 through development.
8), the Office of Nuclear Reactor Regulation re-quired licensees of operating plants, applicants for (1) Description of Submittals operating licenses and licensees of plants under con-struction to perform analyses of transients and acci- (a) NEDO-24708, AdditionalInformation Re-dents, prepare emergency procedure guidelines, up- quiredfor NRC Staff Generic Report on Boil-grade emergency procedures, including procedures ing WaterReactors, August,1979.
for operating with natural circulation conditions, and to conduct operator retraining (see also Item (b) NEDO-24708A, Revision 1, Additionallnfor-1.A.2.1). Emergency procedures are required to be motion Requiredfor NRC Staff Generic Report consistent with the actions necessary to cope with the on Boiling Water Reactors, December,1980.
transients and accidents analyzed. Analysis of tran- This report was issued via the letter from D.
sients and accidents were to be completed in early B. Waters (BWR Owners' Group) to D. G.
1980 and implementation of procedures and retrain- Eisenhut (NRC) dated March 20,1981.
ing were to ha completed 3 months after emergency procedure guidelines were established; however, (c) BWR Emergency Procedure Guidelines (Re.
D- some difficulty in completing these requirements has vision 0) - submitted in prepublication form
(. . been experienced. Clarification of the scope of the task and appropriate schedule revisions are being June 30,1980.
developed. In the course of review of these matters (d) BWR Emergency Procedure Guidelines (Re-on Babcock and Wilcox (B&W) - designed plants, vision 1) -Issued via the letter frem D. B.
the staff will follow up on the bulletin and order mat- Waters (BWR Owners' Group) to D. G.
ters relating to analysis methods and results, as listed Eisenhut (NRC) dated January 31,1981.
in NUREG-0660, Appendix C (see Table C.1, Items 3,4,16,18,24,25,26,27; Table C.2, items 4,12,17, (e) BWR Emergency Procedure Guidelines (Re-18, 19, 20; and Table C.3, items 6,35,37,38,39,41, vision 2) submitted in prepublication form 47,55,57). June 1,1982, Letter BWROG-8219 from T. J.
Dente (BWR Owners' Group) to D. G.
Response Eisenbut (NRC).
l In the clarification of the NUREG-0737 require- (f) BWR Emergency Procedure Guidelines (Re-ment for reanalysis of transients and accidents and vision 3), submitted in prepublication form inadequate core cooling and preparation of guide- December 22,1982, Letter BWROG-8262 lines for development of emergency procedures, from T. J. Dente (BWR Owners' Group) to NUREG-0737 states: D. G. Eisenhut (NRC).
Owners' group or vendor submittals may be refer- (g) NEDO-31331, BWR Emergency Procedure enced as appropriate to support this reanalysis. If Guidelines (Revision 4), submitted April 23, l
.' owners' group or vendor submittals have already 1987, Letter BWROG 8717, irom T. A.
been forwarded to the stafffor review, a brief de- Pickens (BWR Owners' Group) to T. Murley scription of the submittals andjustification of (NRC). ;
p their adequacy to support guideline development 1 is allthat is required. (2) Adequacy of Submittals Amendment 6 1A.2-1 L______.___-___ __ _ _
Mb 23A6100AC Standard Plant nev c The submittals described in (1) above have been 0700. A DCRDR specified in NUREG-0737 is not dscussed and reviewed extensively among the BWR required by SRP Section 18.1.
Owners' Group, the General Electric Company, and the NRC Staff. 1A.2.3 Control Room Design - Plant Safety Parameter Display Console The NRC has extensively reviewed the latest re. [I.D.2]
vision (Revision 4) of the emergency Procedures Guidelines and issued a SER, Safety Evaluation of NRC Position BWR Owners' Group Ernergency Procedure Guide-lines, Revision 4, NEDO-31331, March 1987, letter In accordance with Task Action Plan I.D.2, each from A. C. Thadani, NRC Office of Nuclear Reactor applicant and licensee shall install a safety parameter Regulation, to D. Grace, Chairman of BWR Owners' display system (SPDS) that will display to operating Group, dated September 12,1988. The SER con- personcel a minimum set of parameters which define cludes that this document is acceptable for imple- the safety status of the plant. This can be attained mentation. It further states that the SER closes all through continuous indication of direct and derived the open items carried from the previous revisions of variables as necessary to assess plant safety status.
the EPG.
. Response GE believes that in view of these findings, no further detailed justification of the analyses or guide- The functions of the SPDS will be integrated into lines is necessary at this time. Interface require- the overall control room design, as permitted by SRP ments pertaining to emergency procedures are dis. Section 18.2.
cussed in Subsection 1A.3.1.
1A.2.4 Scope of Test Program - Preoper.
1A.2.2 Control Room Design Reviews - ational and Lower Power Testing [I.G.1]
Guidelines and Requirements [I.D.1(1)]
NRC Position NRC Position Supplement operator training by completing the In accordance with task Action Plan I.D.1.(1), speciallow-power test program. Tests may be ob-all licensees and applicants for operating licenses will served by other shifts or repeated on other shifts to be required to conduct a detailed control room de- provide training to the operators.
sign review to identify and correct design deficien-cies. This detailed control-room design review is Response expected to take about a year. Therefore, the Office of Nuclear Reactor Regulation (NRR) requires that The initial test program presents an excellent op-those applicants for operating licenses who are portunity for licensed operators and other plant staff unable to complete this review prior to issuance of a members to gain valuable experience and training ,
license make preliminary assessments of their con. and in fact theses benefits are objectives of the pro-trol rooms to identify significant human factors and gram (see Subsection 14.2.1). The degree to which instrumentation problems and establish a schedule the potential benefit is realized will depend on such approved by NRC for correcting deficiencies. These plant specific factors as the organizational makeup of applicants will be required to complete the more the startup group and overall plant staff (see Subsec-detailed control room reviews on the same schedule tions 14.2.2 and 13.1), as well as how the test pro-as licensees with operating plants. gram is conducted (see Subsection 14.2.4).
Response The test program described in Chapter 14 is con-sistent with the BWR Owners' Group response to The design of the main control room will utilize Item I.G.1 of NUREG as documented in a letter of i accepted human factors engineering principles,in- February 4,1981 from D. B. Waters to D. G. l corporating the results of a full systems analysis Eisenhut.
similar to that described in Appendix B of NUREG-Amendment 4 tA.2 2
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. 23A6100AC Standard Plant - nev c p
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's ) contain highly radioactive materials. The design re-view should identify the location of vital areas and equip ment, such as the control room, radwaste con-trol stations, emergency power supplies, motor con-trol centers, and instrument areas,in which person-
- nel occupancy may be unduly limited or safety equip-ment may be unduly degraded by the radiation fields during post accident operation of these systems.
Each licensee shall provide for adequate access to vital areas and protection of safety equipment by, design changes, increased permanent or temporary shielding, or post accident procedural controls. The design review shal.1 determine which types of correc-tive actions are needed for vital areas throughout the facility.
t
Response
Response will be provided by April 28,1989.
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,c Amendment 6 1A.2-6
33A6100AC i Standard Plant asv c l
matically or manually depressurized via safety / relief nuclear boiler.
valves (or via the break, for larger breaks). Low pressure flooder system would then make up inven- (5) The SRV's vent to the containment suppression tory. With the core covered neither the rapid gen- pool, where discharged steam is condensed with- ;
eration of noncondensibles nor their accumulation out causing a rapid containment pressure / tem- )
would be possible. perature transient.
The performance of RCIC under Case 3 is of (6) The SRV's are not smaller than the NRC defined concern only if there has been a very substantial pro- small LOCA. Inadvertent actuation is a design-duction of hydrogen due to core uncovery and there basis event and a demonstrated controllable is a need to start the RCIC. This is extremely transient.
unlikely because it could arise only if the core were a!! owed to remain uncovered for a long period with (7) Inadvertent actuation is of course undesirable, the reactor at high pressure. Automatic depressur- but since the SRV's serve an important protective ization system operation and explicit operating function, no steps such as removal of power instructions and the emergency procedures guide- during normal operation, should be taken to pre-lines are intended to preclude this. If the level has vent inadvertent actuation.
fallen with the reactor at high pressure, the vessel would be depressurized via the relief valves auto- (8) A dual indication of SRV position (LVDT posi-matically or manually to permit low pressure injec- tion transmitter and temperature) is provided in tion independent of RCIC performance, the control room.
Because the relief valves and RCIC will vent the (9) Each SRV is remotely operable from the control reactor continuously, and because containment hy- room.
drogen calculations in normal safety analysis calcula-tions assume continuous venting, no special analyses (10) Block valves are not required, so block valve are required to demonstrate "that the direct venting qualifications are not applicable.
of noncondensible gases with perhaps high hydrogen concentrations does not result in violation of (11) No new 10CFR50.46 conformance calcula-combustible gas concentration limits in contain- tions are required, because the vent provi-ment." sions are part of the plant's original design and are covered by the original design bases.
Conclusion and Comparison with Requirements (12) Plant specific procedures govern the opera-The conclusions from this vent evaluation are as tor's use of the rel:cf mode for venting reac-follows: tor pressure.
(1) Reactor vessel head vent valves exist to relieve 1A.2.6 Plant Shielding to Provide head pressure (at shutdown) to the drywell via Access to Vital Areas and Protect Safety remote operator action. Equipnient for Post-Accident Operation
[II.B.2]
(2) The reactor vessel head is continuously swept to the main condenser and can be vented during NRC Position operating conditions via the SRV's to the sup-pression pool. With the assumption of a post-accident release of radioactivity equivalent to that described in Regula-(3) The RCIC system provides an sdditional vent tory Guide 1.3 and 1.4 (i.e., the equivalent of 50% of pathway to the suppression pool. the core radioiodine,100% of the core noble gas inventory, and 1% of the core solids are contained in (4) The size of the vents is not a critical issue the primary coolant), each licensee shall perform a because BWR SRV's have substantial capacity, radiation and shielding-design review of the spaces exceeding the full power steaming rate of the around systems that may, as a result of an accident, l
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l Amendment 4 tA.2-5 )
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rww-n - r--=--r - ' - - - - - - - - - - - = - - - - - -
MM 23A6100AC Standard Plant arv c h9\,.) .
1A.2.7 Post-Accident Sampling [II.B.3]
position with the following exception. The upper limit of activity in the samples at the time they are NRC Position - taken is as follows:
A design and operational review of the reactor liquid sample 1 C./ml L coolant and containment _ atmosphere sampling line L systems shall be performed to determine the capabil- gas sample 10 pC;/ml ity of personnel to promptly obtain (less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) a sample under accident conditions without incurring a radiation exposure to any individual in excess of 3 and 18-3/4 rem to the whole body or extremities, re-spectively. Accide'nt conditions should assume a Regulatory Guide 1.3 or 1.4 release of fission prod-ucts. If the review indicates that personnel could not promptly and safely obtain the samples, additional
. design features or shielding should be provided to meet the criteria.
A design and operational review of the radio-logical spectrum analysis facilities shall be performed to determine the capability to promptly quantify (in less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) certain radionuclides that are indi-cators of the degree of core damage. Such radionu-clides are noble gases (which indicate cladding failure), iodines and cesiums (which indicate high
.h fuel temperatures), and nonvolatile isotopes (which indicate fuel melting). The initial reactor coolant spectrum should correspond to a Regulatory Guide 1.3 or 1.4 release. The review should also consider the effe~ cts of direct radiation from piping and components in the auxiliary building and possible contamination and direct radiation from airborne effluents. If the review indicates that the analyses required cannot be performed in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to meet
. the criteria..
' In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reac-tor conditions. Procedures shall be provided to perform boron and chloride chemical analyses as-suming a highly radioactive initial sample (Regula-tory Guide 1.3 or 1.4 source term). Both analyses shall be capable of being completed pro. aptly (i.e.,
the boron sample analysis within an hour and the chloride sample analysis within a shift).
Response
The post-accident sampling ssystem described in
[ml' Subsection 9.3.2 meets the requirements of this LJ i
Amenement 6 1A.2-7
i MM 23A6100AC Standard Plant .
Rev c 1A.2.8 Rule Making Proceeding or Response Degraded Core Accidents [II.B.8)
The ABWR Standard Plant safety relief valves Response to this TMI action plan item is ad- are equipped with linear variable differential trans-dressed in Appendix 19A. formers (LVDT's) which are qualified as Class 1E components. These LVDT's are mounted on the 1A.2.9 Coolant System Valves-Testing valve operators and are highly reliable sensors for Requirements [II.D.1] monitoring valve position.
NRC Position in addition, the downstream pipe from each valve line is equipped with thermocouple which signal the Pressurized water reactor and boiling-water annunciator and the plant process computer when reactor licensees and applicants shall conduct testing the temperature in the tailpipe exceeds the predeter-to qualify the reactor coolant system relief and safety mined setpoint.
valves under expected operating conditions for design-basis transients and accidents. These sensors are shown on Figure 5.13 (Nuclear Boiler System P&lD).
Response
1A.2.11 Systems Reliability [II.E.3.2]
The ABWR safety / relief valve (SRV) is postu-lated to discharge steam only, not liguid or two phase This TMi aetion plan item superseded by flow. USIA-45. USI A-45 is addressed in Appendix 19B.
The ABWR Systern logic for response to high 1A.2.12 Coordinated Study of Shutdown water level conditions is described in Subsection Heat Removal Requirements [II.E.3.3]
7.3.1.1.1.1(3) and is considered to be sufficiently redundant that the probability of steam line ficoding This TMI action plan item superseded by USI by ECCS is extremely low. There is no high drywell A-45, USI A-45 is addressed in Appendix 19B.
pressure signal that would inhibit this logic system.
1A.2.13 Containment Design Dedicated in the ABWR design, each of three RHR shut- Penetration [II.E.4.1]
down cooling lines has its own separate containment penetration and its own separate source of suction GE has been evaluating the need for additional from the reactor vessel. Alternate shutdown using hydrogen control during finalization of the ABWR the SRV is therefore not required for ABWR in design. A recombiner, based on the source terms order to meet single failure rules. specified in Table 1 of Regulatory Guide 1.7, will be added to Subsection 6.2.5. Any hydrogen Finally. it is noted that the SRV models which recombiner system provided with primary will be used with ABWR plants are expected to be containment vessel penetrations will be operable in very similar to existing models which have undergone the event of a single active failure.
testing for alternate shutdown cooling mode flow conditions. IA.2.14 Containment Design Isolation Dependability [II.E.4.2]
1A.2.10 Relief and Safety Valve Position Indication [II.D.3] NRC Position NRC Position (1) Containment isolation system designs shall comply with t'3e recommendations of Stan.
Reactor coolant system relief and safety valves dard Review Plan Subsection 6.2.4 (i.e., that shall be provided with a positive indication in the there be diversity in the parameters sensed control room derived from a reliable valve position for the initiation of containment isolation).
detection device or a reliable indication of flow in the
, discharge pipe.
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Amendment 6 tA.2 8 L___
23A6100AC
.(N Efandard Plant : nev c U '(2)- All plant personnel shall give careful consid-eration to the definition of essential and non-essential systems, identify each system deter.
- mined to be essential, identify each system determined to be nonessential, describe the
, - basis for selection of each essential system, 1 modify their containment isolation designs accordingly, and repc-' the results of the j- reevaluation to the NRC.
I (3)- All nonessential systems shall be auto-matically isolated by the containment isolation signal.
(4) The design of control systems for automatic con-tainment isolation valves shall be such that reset-ting the isolation signal will not result in the Q
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Amsedment 6 1A.2-8a
I 23A6100AC Standard Plant nev c
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automatic reopening of containment isolation (5)The ABWR Standard Plant design is consistant valves. Reopening of containment isolation with this position.
valves shall require deliberate operator action. .
(6) All ABWR containment purge valves meet the (5) The containment setpoint pressure that initiates criteria provided in BTP CSB 6-4. The main 22" containment isolation for nonessential penetra. purge valves are fail-closed and are maintained tions must be reduced to the minimum compat- closed through power operation as defined in the ible with normal operating conditions. plant technical specifications. All purge valves are remote operated, fail closed and receive (6) Containment purge valves that do not satisfy the containment isolation signals. Certain valves can operability criteria set forth in Branch Technical be opened in the presents of an isolation signal, Position CSB 6-4 or the Staff Interim Position of but these valves are keylocked and under October 23,1979 must be scaled closed as de- administrative control.
fined in SRP 6.2.4, Item II.3.f during operational conditions 1,2,3, and 4. Furthermore, these valves must be verified to be closed at least every 31 days.
(7) Containment purge and vent isobtion valves must close on a high radiation signal.
Response
(1) Theisolation provisions described in the Stan-dard Review Plan, Subsection 6.2.4 (i.e., that
! there be dh ersny in the parameters sensed for
\
the initiation of containment isolation) were re-viewed in conjunction with the ABWR Standard Plant design. It was determined that the ABWR Standard Plan is designed in accordance with '
these recommendations of the SRP.
(2) This request appears to be directed primarily toward operating plants. However, the classifi-cation of structures, systems and components for the ABWR Standard Plant design is addressed in Section 3.2 of this SSAR. The basis for classi-fication is also presented in Section 3.2. The ABWR Standard Plant fully conforms with the NRC position so far as it relates to the new equipment supplier.
(3) All non-essential systems comply with the NRC poshion to automatically isolate by the contain-ment isolation signals, and by redundant safety grade isolation valves.
(4) Control systems for automatic containment iso-lation valves are designed in accordance with this position for the ABWR Standard Plant tJ Amendesse 6 1A.2 9
M 33A6100AC Standard Plant nev c 1A.2.15 Additional Accident. Monitoring the top of the containment sump. A wide range in-Instrumentation [II.F.1(1)] strument shall also be provided for BWRs and shall cover the range from the bottom of the containment NRC Position to the elevation equivalent to a 600,000 gallon capacity. For BWRs, a wide range instrument shall Noble gas effluent monitors shall be installed be provided and cover the range from the bottom to with an extended range designed to function during 5 feet above the normal water level of the suppres-accident conditiors as well as during normal operat- sion pool.
ing conditions. Multiple rnonitors ar considered necessary to cover the ranges of interest. A continuous indication of hydrogen concentra-tion in the containment atmosphere shall be pro-(1) Noble gas efflynt monitors with an upper range vided in the control room. Measurement capability capacity of 10 Ci/cc (Xe-133) are considered to shall be provided over the range of 0 to 10% hydro-be practical and should be installed in all gen concentration under both positive and negative operating plants. ambient pressure.
(2) Noble gas effluent monitoring shall be provided Response for the total range of concentration extending from normal condition (as low as reasonably GE believes the requirements of Regulatory achievable (pLARA)) concentrations to a maxi- Guide 1.97, Revision 3, incorporate the above re-mum of 10 Ci/cc (Xe-133). Multiple monitors quirements. Section 7.5 compares the ABWR design are considered to be necessary to cover the against this Regulatory Guide.
ranges of interest. The range capacity of individual monitors should overlap by a factor of 1A.2.16 Identification of and Recovery ten. From Conditions Leading to Inadequate Core Cooling [II.F.2]
Because iodine gaseous effluent monitors for the accident condition are not considered to be prac- NRC Position tical at this time, capability for effluent monitoring of radioiodines for the accident condition shall be Licensees shall provide a description of any addi-provided with sampling conducted by absorption on tional instrumentation controls (primary or backup) charcoal or other media, followed by onsite labora- proposed for the plant to supplement existing instru-tory analysis. mentation (including primary coolant saturation j nonitors) in order to provide an unambiguous, easy- i In-containment rgdiation level monitors with a to interpret indication of inadequate core cooling l maximum range of 10 rad /hr shall be installed. A (ICC). A description of the functional design re- l minimum of two such monitors that are physically quirements for the system shall also be included. A l separated shall be provided. Monitors shall be description of the procedures to be used with the developed and qualified to function in an accident proposed equipment, the analysis used in developing environment. these procedures, and a schedule for installing the ;
equipment shall be provided. {
A continuous indication of containment pressure i shall be provided in the control room of each operat- Response ing reactor. Measurement and indication capability shallinclude three times the design pressure of the The direct water level instrumentation provided containment for concrete, four times the design pres- in the ABWR design is capable of detecting sure for steel, and 5 psig for all containments. conditions indicative of inadequate core cooling.
A continuous indication of containment water The ABWR has two sets of four wide range level shall be provided in the control room for all reactor water level sensing units (eight total) which plants. A narrow range instrument shall be provided are used in two separate two out of four logics which for BWRs and cover the range from the bottom to initiate ECCS and other safety functions. Each set of Amendment 6 1 A.210
.M . .-
23A6100AC 7% Standard Plant nev c
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I . four sensors are used in two separate two out of four
- logics which initiate ECCS operation. Four separate-sets of sensing lines, one from each quadrant of the reactor pressure vessel, supply the pressure to the eight sensors for reliability. This ABWR arrangement of reactor water level sensing exceeds or is at least equal to the redundancy and reliability j of the BWR reator water level measurement systems reviewed in BWR Owners Group Report SLI-821, July 1982. The conclusions reached in SLI-821 and companion report SLI-8218, December 1982, also apply to the ABWR. These conclusions meet the NRC staff expectation given in paragraph 4.4.7 of GESSAR 11 SER (Safety Evaluation Report NUREG-0979, April 1983) regarding NUREG-0737, Item 11 F.2.
Based on the above information, the existing highly redundant direct water levelinstrumentation already provides an unambiguous, easy to interpret indication of inadequate core cooling and there are no plans to include core-exit thermocouple in the ABWR design.
C'\
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r Amendment 6 1A.2-10a
23A6100AC Standard Plant nev c T and engineered safeguards systems. the NRC staff and changes should be implemented if g justified by the analysis.
(4) Fuel-zone, water level range: This range used for its RPV taps the elevation above the main Response steam outlet nozzle and the taps just above the internal recirculation pump (RIP) deck. The The ABWR Standard Plant design is consistent zero of the instrument is the bottom of the ac- with this position. The high pressure core flooder tive fuel and the instruments are calibrated to be (HPCF) system is initiated at Level 11/2, and the accurate at 0 psig and saturated condition. The RCIC system is initiated at level 2. At Level 8, the water level measurement design is the conden- injection valves for the HPCF and the RCIC steam sate reference type,is not density compensated, supply and injection valves will automatically close in and uses differential pressure devices as its pri- order to prevent water from entering the main steam mary elements. These instruments provide lines.
input to water-level indication only.
In the unlikely event that a subsequent low level There are common condensate reference cham- recurs, the RCIC steam supply and injection valves -
bers for the narrow range; wide-range; and fuel- will automatically reopen to allow continued flooding zone, water-level ranges. of the vessel. The HPCF injection valves will also automatically reopen unless the operator previously The elevation drop from RPV penetration to the closed them manually. Refer to Subsections drywell penetration is uniform for the narrow range 7.3.1.1.1.1 (HPCF) and 7.3.1.1.1.3 (RCIC) for and wide range water-level instrument lines in order additional details regarding system initiation and to minimize the change in water-level with changes isolation logic.
in drywell temperature.
1A.2.23 Modify Break-Detection IAgic Reactor water levelinstrumentation that ini- to Prevent Spurions Isolation of HPCI
/9 tintes safety systems and engineered safeguards is And RCIC Systems [II.K.3(15)]
(/ shown in Figure 5.1-3.
NRC Position 1A.2.22 Separation of HPCI AND RCIC Systen Initiation I2vels [II.K.3(13)] The high-pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) systems use NRC Position differential pressure sensors on elbow taps in the steam lines to their turbine drives to detect and Currently, the reactor core isolation cooling isolate pipe breaks in the systems. The pipe-break-(RCIC) system and the high-pressure coolant injec- detection circuitry has resulted in spurious isolation tion (H?CI) systems both initiate on the same low- of the HPCI and RCIC systems due to the pressure ]'
water level signal and both isolate on the same spike which accompanies startup of the systems. The high-water-level signal. The HPCI system will pipe-break- detection circuitry should be modified to restart on low water level but the RCIC system will that pressure spikes resulting from HPCI and RCIC not. The RCIC system is a low-flow system when system initiation will not cause inadvertent system compared to the HPCI system. The initiation levels isolation.
of the HPCI and RCIC system should be separated so that the RCIC system initiates at a higher water Response level than the HPCI system. Further, the initiation logic of the RCIC system should be modified so that The ABWR design utilizes the motor driven the RCIC system will restart on low water level. HPCF system rather than the turbine-driven HPCI These changes have the potential to reduce the system for high pressure inventory maintenance.
number of challenges to the HPCI system and could Therefore, this position is only applicable to the result in less stress on the vessel from cold water turbine-driven RCIC system.
injection. Analyses should be performed in evaluate these changes. The analysis should be submitted to In the ABWR Standard Plant design, the high !
t
(
A- he4 1A.2-13
b 23A6100AC Standard Plant nov c differential pressure signals which isolate the RCIC (5) Earlier initiation of ECC systems, turbine are processed through the leak detection and isolation system (LDS). Spurious trips are avoided (6) Heat removal through emergency condensers, because the RCIC has a bypass start system con-trolled by va!ves F045 and F092 (see Figure 5.4-8, (7) Offset valve setpoints to open fewer valves per RCIC P&lD). challenge, On receipt of RCIC start signals, bypass valve (8) Installation of additional relief valves with a block F092 opens to pressurize the line downstream and or isolation-valve feature to eliminate opening of accelerate the turbine. The bypass line via F092 is the safety / relief valves (SRV's), consistent with small (1-inch) and naturally limits the initial flow the ASME Code, surge such that a differential pressure spike in the upstream pipe will not occur. (9) Increasing the high steam line flow setpoint for main steam line isolation valve (MSIV) closure, After a predetermined delay (approximately 510 seconds), steam supply valve F045 opens to ad- (10) Lowering the pressure setpoint for MSIV mit full steam flow to the turbine. At this stage, the Closure, line downstream is already pressurized. Thus, it is highly unlikely that a differential pressure spike (11) Reducing the testing frequency of the could occur during any phase of the normal start up MSIV's, process.
(12) More stringent valve leakage criteria, and 1A.2.24 Reduction of Challenges and Failures of Relief Valves - Feasibility (13) Early removal of leaking valves.
Study and System Modification
[lI.K.3(16)] An investigation of the feasibility and constraints of reducing challenges to the relief valves by use of NRC Position the aforementioned methods should be conducted.
Other methods should also be included in the feasi-The record of relief-valve failures to close for all bility study. Those changes which are shown to boiling water reactors (BWRs)in the past 3 years of reduce relief valve challenges without compromising plant operation is approximately 30 in 73 reactor- the performance of the relief valves or other systems years (0.41 failures per reactor-year). This has dem- should be implemented. Challenges to the relief onstrated that the failure of a relief valve to close valves should be reduced substantially (by an order would be the most likely cause of a small-break loss- of magnitude).
of-coolant accident (LOCA). The high failure rate is the result of a high relief valve challenge rate and a Response relatively high failure rate per challenge (0.16 fail-ures per challenge). Typically, five valves are chal- General Electric and the BWR Owners' Group ,
lenged in each event. This results in an equivalent responded to this requirement in Reference 6. This failure rate per challenge of 0.03. The challenge and response, whichwas based on a review of existing failure rates can be reduced in the folfowing ways: operating information on the challenge rate to relief valves, concluded that the BWR/6 product line had (1) Additional anticipatory scram on loss of feedwa- already achieved the " order of magnitude" level of ter, reduction in SRV challenge rate. The ABWR relief valve system also has similar design features which (2) Revised relief-valve actuation setpoints, also reduce the SRV challenge rate. With regard to inadvertently opened relief vaivt 39RV), the (3) Increased emergency core cooling (ECC) flow, BWR/6 plant design evaluated for the Owners' Group report reflected a reduced level of IORV (4) Lower operating pressures, compared with previous design because of ,
O Amendment 6 1A.2-14
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E ul iABWR 2M61MAC Standard Plant - ne v. c
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' elimination of the pilot operated relief valve type of { '
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' design. The ABWR design has also eliminated the pilot operated relief valve type of design.
For the ABWR.which has a solid state logic desigh with redundancy, the likelihood of an IORV is the same or less than the BWR/6 design evaluated in connection with the Owners' Group report. The j
redundant solid state design has been selected in order that the frequency of IORV with solid state logic becomes low enough so as to achieve the order of magnitude reduction in total SRV challenge rate
! required by NUREG-0737.
.The redundant solid state design for SRV operation in the pressure relief mode consists of two duplicated microprocessor channels. Each microprocessor' channel activates a separate load driver and both load drivers must be activated to L' cause operation of the SRV's in the relief mode.
Operation of the SRV's in the ADS mode also requires activation of two microprocessor channels with separate load drivers to prevent unwanted SRV operation; however, two separate dual channel
- n. systems are used to assure reliable operation in the Q -
. ADS mode. Reliable operation in the pressure re!ief mode is; assured by direct opening of the SRV against spring force.
i.
Amendment 6 1A.214a
I 23A6100AC Standard Plant nev c i (3 1A.2.25 Report on Outages of Emergency feasibility of automating the vessel depressurization d Core-Cooling Systems Licensee Report and Proposed Technical Specification for isolation events with and without a stuck open relief valve, and assessed the changes in overall plant Changes [II.K.3(17)] risk resulting from such automation. 1 NRC Position The intent of this NUREG 0737 item is to j provide more assurance of adequate core cooling in j Several components of the emergency core the event of transients and accidents not producing a cooling (ECC) systems are permitted by techni,:al high drywell pressure signal under conditions such specifications to have substantial outage times (e.g., that high pressure makeup systems are unable to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for one diesel-generator; 14 days for the maintain ; actor inventory. The intent may be HPCI system). In addition, there are no cumulative satisfieu n two ways: the ADS logic may be ]
outage time limitations for ECC systems. Licensees modified to automate the depressurization for these j should submit a report detailing outage dates and additional events, or the operator may be given 1
lengths of outages for all ECC systems for the last 5 specific guidance and training for performing manual l years of operation. The report should also include actions under degraded conditions, the causes of the outages (i.e., controller failure, i spurious isolation). This second course of action has been !
undertaken for ABWR with the implementation of i Response the sympton oriented emergency procedure l guidelines. The transients considered are slow and This requirement does not apply to future well understood, and the operator has sufficient time plants. to assess the plant conditions and initiate the depressurization. The NRC has issued an SER 1A.2.26 Modification of Automatic De- ISafety Evaluation of BWR Owners'
,- Pressurization System IAgic - Feasi- tiroup-Emergency Procedure Guidelines, Rev. 4,
().
bility for Increased Diversity for Some Event Sequences [II.K.3(18)]
. 'EDO 31331, March 1987) which accepts manual con?rol of reactor depressurization in the manner ;
defir ed in the guideline.
NRC Position The automatic depressurization system (ADS) actuation logic should be modified to eliminate the need for manual actuation to assure adequate core cooling. A feasibility and risk assessment study is required to determine the optimum approach. One possible scheme that should be considered is ADS actuation on low reactor vessel water level provided no high-pressure coolant injection (HPCI) or high-pressure coolant system (HPCS) flow exists and a low pressure emergency core cooling (ECC) system is running. This logic would complement, not replace, the existing ADS actuation logic.
Response
General Electric and the BRW Owners' Group participated in a program to evaluate potential modifications of the ADS initiation logic.
The results of the initial study were submitted to the NRC via a letter from D. B. Waters, Chairman of the O
(/
BWR Owr.ers' Group, to D. G. Eisenhut, NRC, dated March 31,1981. This study evaluated the Amendment 6 tA.215
l MM 23A6100AC Standard Plant nev c 1A.2.27 Restart of Core Spray and LPCI reviewed this issue on a generic basis in 1980, and Systems on Low level Design and concluded the NRC suggestions were not required Modification [II.K.3(21)] for plant safety considerations. Justification is provided in the December 29,1980, BWR Owners' NRC Position Group submittal to the NRC (Reference 5). Plant i variations between the BWR and the ABWR designs The core spray and low pressure coolant injec- are not significant to the overall technical conclu-tion (LPCI) system flow may be stopped by the sions of the study.
operator. These systems will not restart automati-cally on loss of water level if an initiation signal is This conclusion is based on the adequacy of the v.ill present. The core spray and LPCI system logic current ECCS logic design coupled with the poten- l 6hould be modified so that these systems will restart, tially negative impact on overall safety of the pro- i i
if required, to assure adequate core cooling. posed changes. For the low pressure ECCS, these.
Bect.use this design modification affects several negative impacts include a significant escalation of j core. cooling modes under accident conditions, a control system complexity and restricted operator j preliminary design should be submitted for staff flexibility when dealing with anticipated events. l review and approval prior to making the actual i modification. A full understanding of the significance of these logic changes must be based on a recognition that Response these systems must consider the possible interactive effects among the other systems making up the over-The ABWR standard plant emergency core all ECCS network. This must also include the cooling system (ECCS) is made up of the high pres- potentialimpact on supporting systems such as the sure core flooder (HPCF) system, the reactor core standby power supplies and the shutdown service isolation cooling (RCIC) system, the automatic de- water systems. Furthermore, the LPFL is one mode pressurization system (ADS) and the low pressure of the RHR system which has other safety related flooder (LPFL) mode of the residual heat removal functions such as suppression pool cooling and (RHR) system, wetwell/drywell spray cooling. Clearly, these other safety functions must not be compromised by any A high water level (Level 8) signal will auto- changes in the LPFL mode of operation, matically close the HPCF injection valves and the RCIC steam supply and injection valves. These The referenced systems analysis took into consid-systems may also be shut down manually. Manual cration these potentialinteractive effects, impacts on action is required to shut down the RHR once it is supporting systems, plant instrumentation and emer-initiated. gency procedure guidelines. The study concluded that auto-restart of these systems is not necessary or in the unlikely event that a subsequent low level appropriate. Therefore, GE concludes that no modi-reoccurs, the RCIC steam supply and injection valves fications should be made with respect to automatic will automatically reopen to allow continued flooding restart of the low pressure ECCS.
of the vessel. The HPCF injection valves will also automatically reopen unless the operator previously 1A.2.28 Automatic Switchover of Reactor closed them manually. Core Isolation Cooling System Suction -
Verify Procedures and Modify Design The NRC has suggested certain modifications to [II.K.3(22)]
the LPCI (LPFL for the ABWR) and core flooder systems provided as part of the ECCS network. NRC Position These NRC suggestions center on control system logic modifications that would provide automatic re- The reactor core isolation cooling (RCIC) system start capability following manual termination of sys- takes suction from the condensate storage tank with tem operation. manual switchover to the suppression pool when the condensate storage tank level is low. This switchover General Electric and the BWR Owners' Group should be made automatically. Until the automatic O
Amendment 4 1A.2 16 l
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MM 23A6100AC Standard Plant nev c
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'/ switchover is implemented, licensees should verify b ) that clear and cogent procedures exist for the manual switchover of th'e RCIC system suction from the con-densate storage tank to the suppression pool.
Response
The RCIC system provided in the ABWR Stan-dard Plant includes an automatic switchover feature which will change the pump suction source from the RCIC storage pool to the suppression pool. The safety-grade switchover will automatically occur upon receipt of a low-level signal from,the conden-sate storage pool or a high-level signal from the sup-pression pool.
See Subsection 7.3.1.1.1.3 for additionalinfor-mation.
1A.2.29 Confirm Adequacy of Space Cooling for High Pressure Coolant Injection and Reactor Core Isolation Cooling Systems [II.K.3(24)]
NRC Position
- &j
( Long term operation of the reactor core isola-tion cooling (RCIC) and high-pressure coolant injec-tion (HFCI) systems may require space cooling to maintair,the pump-room temperatures within allow- i able limits. Licensees should verify the acceptability of the consequences of a complete loss of alternat-ing-current power, The RCIC and HPCI systems should be designed to withstand a complete loss of offsite akernating-current power to their support sys-terts, including coolers, for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Response
The high pressure coolant injection and the re-actor core isolation coolant injection systems are provided space cooling via individual room safety grade air-conditioning systems.
l
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l I
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Amendmesa6 - IA 217 i
MNM 33A6100AC Standard Plant nev c 1A.2.30 Effect of Loss of Alternating- piping within primary containment. Two redundant Current Power on Pump Seals [II.K.3(25)] 7-day supplies of bottled air are available to compensate for leakage during long-term usage, with ,
NRC Position replacement capability being provided for the !
remainder of the postulated accident to assure The licer. sees should determine, on a plant system functional operability. A maximum of three specific basis, by analysis or experiment, the of eight ADS valves need function to meet consequences of a loss of cooling water to the short term demands (see Subsection 19.3.1.3.1) and reactor recirculation pump seal coolers. The pump the functional operability of only one ADS valve will seals should be designed to withstand a complete loss fulfilllonger term needs. Loss of pneumatic supply of alternating-current (ac) power for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. pressure to the ADS SRV accumulator is alarmed to Adequacy of the seal design should be demonstrated. provide the reactor operator with indication of the failure of any of the redundant systems under hostle Response environmental condition.
The ABWR design features internal recircula. The BWR Owners' Group sponsored an tion pumps (RIP) which do not require shaft seals, evaluation of the adequacy of the ADS During a LOPA, the RIPS shutdown automatically configurations. Evaluation results are summarized in but there are no shaft seals which require cooling the following paragraph.
water restoration.
The accumu'lators are designed to provide two I A.2.31 Study and Verify Qualification of ADS actuations at 70% of drywell design pressure, Accumulators on Automatic Depressuri- which is equivalent to 4 to 5 actuations at zation System Valves [II.K.3(28)] atmospheric pressure. The ADS valves are designed to operate at 70% of drywell design pressure because NRC Position that is the maximum pressure for which reactor depressurization through the ADS valves is required.
Safety analysis reports claim that air or nitrogen The greater drywell design pressures are associated accumulators for the automatic depressurization only with the short duration primary system system (ADS) valves are provided with sufficient blowdown in the drywell immediately following a capacity to cycle the valves open five times at design large pipe rupture for which ADS operation is not pressures. GE has also stated that the emergency required. For large breaks which result in higher core cooling (ECC) systems are designed to with- drywell pressure, sufficient reactor depressurization stand a hostile environment and still perform their occurs due to the break to preclude the need for function for 100 days following an accident. Licensee ADS. One ADS actuation at 70% of drywell design should verify that the accumulators on the ADS pressure is !.ufficient to depressurize the reactor and valves meet these requirements, even considering allow inventory makeup by the low pressure ECC normalleakage. If this cannot be demonstrated, the systems. However, for conservatism, the Licensee must show that the accumulator design is accumulators are sized to allow 2 actuators at 70% of still acceptable. drywell pressure. See Subsection 6.8.1 for a description of the ADS N Pneumatic supply.
2 The accumulators for the ADS valves are sized to provide two operating cycles at 70% of drywell design pressure. This cyclic capability is validated during preoperational testing at the station. The accumulators are safety grade components.
The 100-day, post-accident functional operability requirement is met through conservative design and redundancy; eight ADS valves are provided with code-qualified accumulators and seismic Category I 4
Amendment 6 1A.218 1 1 L__ _ _ \
L ABWR 2a6mc Standard Plant nov c 1A.2.32 Revised Small-Break 1A.2.33 Plant-Specific Calculations to IAss-of Coolant Accident Methods to Show Show Compliance with 10 CFR Part 50.46 Compliance with 10 CFR PART 50, [II.K.3(31)]
Appendix K [II.K.3(30)]
NRC Position NRC Position Plant-specific calculations using NRC-approved The analysis methods used by nuclear steam models for small break loss-of collant accidents supply system (NSSS) vendors and/or fuel suppliers (LOC As) as described in Item II.K330 to show for small break loss-of coolant accident (LOCA) compliance with 10 CFR 50.46 should be submitted analysis for compliance with Appendix K to 10 CFR for NRC approval by alllicensees.
Part 50 should be revised, documented, and submit-ted for NRC approval. The revisions should account Response for comparisons with experimental data, including data from the LOFT Test and Semiscale Test The ABWR standard safety small break LOCA facilities. calculations are discussed in Subsection 633.7.
Response The rcferences listed in Subsection 63.6 describe the currently approved Appendix K methodology GE has evaluated the NRC request requiring used to perform these calculations. Co.npliance with that the BWR small break LOCA analysis methods 10CFR50.46 has previously been established for that are to be demonstrated to be in compliance with Ap- methodology.
l pendix K to 10 CFR 50 or that they be brought into compliance by analysis methods changes. The spe- Since, as noted in the previous Item (1A.232), no cific NRC concerns are contained in NUREG-0626, model changes are needed to satisfy NUREG-0737 Appendix F. The specific NRC concerns identified item II.K3(30), there is no need to revise the calcu-O in Subsection 4.2.10 of NUREG-0626 (Appendix F) relate to the following: counter current flow limiting lations presented in Subsection 633.7.
l (CCFL) effects, core bypass modeling, pressure 1A.2.34 Primary Coolant Sources Outside variation in the reactor pressure vessel, integral ex- Containment Structure [III.D.1.1(1)]
perimental verification, quantification of uncertain-ties in predictions, the recirculation line inventory NRC Position modeling, and the homogeneous / equi- librium model. Applicants shallimplement a program to reduce leakage from systems outside containment that The response to the NRC small break model would or could contain highly radioactive fluids dur-concerns was provided at a meeting between the ing a serious transient or accident to as low-as prac-NRC and GE on June 18,1981. Information pro- ticallevels. This program shallinclude the following:
vided at this meeting showed that, based on the TLTA small break test results and sensitivity studies, (1)Immediate leak reduction the existing GE small break LOCA model already satisfies the concerns of NUREG 0626 and is in (a) Implement all practicalleak reduction men-compliance with 10 CFR 50, Appendix K. There- sures for all systems that could carry radioac-fore, the GE model is acceptable relative to the tive fluid outside of containment.
concerns of item II.K3(30), and no model changes need be made to satisfy this item. (b) Measure actualleakage rates with systems in operation and report them to the NRC.
Documentation of the information provided at the June 18,1981 meeting was provided via the letter (2) Continuing Leak Reduction - establish and im-front R. H. Buchholz (CE) to D. G. Eisenhut piement a program of preventive maintenance to I (NRC), dated June 26,1981. reduce leakage to as-low-as-practicallevels. This l
O .
Amendment 6 1 A.219
23A6100AC Standard Plant nevc program shall include periodic integrated leak pass outside of the secondary containment contain tests at intervals not to exceed each refueling leakage control systems or loop seals. These systems cycle. allow the SGTS to maintain a negative pressure rela-tive to the enyhonment and thus limit the amount of Response leakage through the secondary containment. These systems are discussed in Subsections 6.53. Finally, Leak reduction measures of the ABWR Stan- expected liquid leakoff from equipment outside the dard Plant include a number of barricrs to contain- containment is directed to equipment drain sumps ment leakage in the closed systems outside the con- and processed by the radwaste system. These multi-tainment. These closed systems include: ple design features of the ABWR Standard Plant provide substantial capability to limit any potential (1) Residual Heat Removal, release to the environment from systetus likely to contain radioactive material.
Q) High Pressure Core Flooder, Additionally, pressure boundary components of (3) low Pressure Core Flooder, radioactive waste systems are purchased as aug-mented Class D syster: to assure their capability to (4) Reactor Core Isolation Cooling implemented provide integrity.
P Suppression Pool Cleanup (suction and return), 1A.235 In-Plant Radiation Monitoring and [III.D.3.3(3)]
(6) Shutdown Service Water (supply and return). NRC Position Plant procedures will prescribe the method of (1) Each licensee shall provide equipment and asso-leak testing these systems. The testing will be per- ciated training and procedures for accurately de-formed on a schedule appropriate to 10CFR50 Ap- termining the airborne iodine concentration in pendix J type B and C penetrations, that is at each areas within the facility where plant personnel refueling outage. When leakage paths are discov- may be present during an accident.
cred including during these tests, they will be investi-gated and necessary maintenance will be performed (2) Each applicant for a fuel loading license to be to reduce leakage to its lowest practical level. issued prior to January 1,1981 shall provide the equipment, training, and procedures necessary to In addition, lines which penetrate the primary accurately determine the presence of airborne containment contain primary containment isolation radioiodine in areas within the plant where plant valves which are designed in accordance with Gen- personnel may be present during an accident.
eral Design Criteria 55,56 or 57 to provide reliable isolation in the event of line breaks. These isolation Response provisions are discussed in Subsection 6.2.4 and contain both automatic and remote manual-closing (1) This is an interface requirement described in valves. Subsection 1A3.4.
Should a small line break develop within a space (2) Not applicable.
inside the secondary containment concurrent with a significant radioactive source term in the reactor 1A.2.36 Control Room Habitability water,it would be detected by the leakage control [III.D.3.4(1)]
systems described in Subsection 5.2.5 and the line may be isolated. Any release of radioactive material NRC Position from such leaks would also be detected by process radiation monitors which would permit operation of In accordance with Task Action Plan item the standby gas treatment system (Subsection 6.5.1) III.DSA and control room habitability, licensees prior to release to the emironment. All lines which shall assure that control room operators will be Amendment 4 1A.2 20 L _ _ _ _ _ _ _ . _
23A6100AC Standard Plant nev.c
.X.
-() - 1A.3 INTERFACES
' 1A.3.1 Emergency Procedures and Emergency Procedures 'Iraining Program Emergency procedures, developed from the emergency procedures guidelines, shall be provided and implemented prior to fuelloading. (See Subsec-tion 1A.2.1).
1A.3.2 Review and Modify Procedures for Removing Safety-Related Systems From Service Procedures shall br reviewed and modified (as
- required) for removing safety-related systems from service (and restoring to service) to assure operabil-ity status is known. (See Subsection IA.2.19) 1A.3.3 In Plant Radiation Monitoring Equipment and training and procedures shall br provided for accurately determinir3 the airborne io-dine concentration in areas within the facility where
,_. plant personnel may be present during the accident.
/ (See Subsection 1A.2.18)
%))
v -
Amendment 6 1A.3-1
o I
1 23A6100AC Standard Plant ._
nov c
,.g 1A.4 REFERENCES-1
- 1. U. S. Nnclear Regulatory Commission, NRC l
' Action Plan Dewloped as a Result of the TMI-2 Accident, USNRC report NUREG-0660, Vols.
1 and 2, May,1980.
, 2. U. S. Nuclear Regulatory Commission, Clanf- !
cation of TMJ Action Plan Requirements, USNRC Report NUREG-0737, November, 1980.
- 3. U. S. Nuclear Regulatory Commission, Licens-ing Requirements for Pending Applications for Construction Permits and Manufacturing Li-cense, NUREG-0710, Revision 1, June,1981.
- 4. Letter from R. H. Buchholz, GE, to D. F. Ross, NRC,
Subject:
Additional Information Re.
quired for NRC Staff Generic Report on Boiling Water Reactors, November 30,1979, MFN-290-79.
- 5. Letter from D. B. Walters, Chairman, BWR
-Q Owners' Group, to NRC, dated December 29, i 1980, BWR Owners' Group Evaluation of NUREG-0737 Requirements.
- 6. Letter from D. B. Walters, Chairman, BWR Owners' Group, to D. G. Eisenhut, NRC, dated M arch 31,1981, B WR Owners' Group Evaluation of NUREG-0737 Requirements II.K3.16 andII.K3.18.
Amendment 6 1A.4-1
ABM 23A6100AD Standard Plant REV. B f CHAPTER 2 TABLE OF CONTENTS F#_ tion 31tle P.aat l 2 SITE CHARACTERISTICS 2.0
SUMMARY
2.0-1 1
2.1 LIMITS IMPOSED ON SRP SECTION II l '
ACCEITANCE CRITERIA BY AB%R DESIGN 2.11 2.2 INTERFACES 2.2-1 O
O 2-il Amendment 6
MN - 23A6100AD Standard Plant nev. s SECTION 2.0 CONTENTS TABLES Table Iitig P.ast 2.0-1 Envelope of ABWR Standard Plant Site Design Parameters 2.0-2 O
i l
O 2.0-ii Amendment 6
' Mb 23A6100AD Eleandard Plant ' REV.B j2.0'
SUMMARY
{,,i) -
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 summary of the site envelope design parameters is given in Table 2.01. See
- Subsection 2.21.for interface requirements, e,
1 1
10 Amendment 6 2.01
a ABM 23461ooro m.-A-rd Plant REv. n D TABLE 2.0-1 ENVELOPE OF ABWR STANDARD PLANT SITE DESIGN PARAMETERS Maximum Ground Water level: Extreme Wind: Basie Wind Speed:
2 feet below grade 110 mph (1)/130 mph (2)
Maximum Flood (or Tsunaml) 14 vel:(3) Tornado:(4) 1 foot below grade Maximum tornado wind speed: 260 mph
- Transnational velocity: 57 mph
- Radius: 453 ft Precipitation (for Roof Design): Maximum atm AP: 1.46 psid
- Maximum rainfall rate: 10 in/hr - Missile Spectra: Per ANSI /ANS-23
- Maximum snow load: 50 lb/sq. ft.
~ Design Temperatures: Soil Properties:
- Ambient - Minimum Bearing Capacity (demand): 15ksf-
~ 1% Furredance Values - Minimum Shear Wave Velocity: 1000 fps Maximum: 1000F dry bulb /770F coincident wet Liquification Potential; bulb None at plant site resulting . 2 Minimum:-100F from OBE and SSE(7) E-0% Frecedance Values (Historicallimit)
- Maximum: 1150F dry bulb /820F coincident wet Seismology:
iO bulb
- Minimum: 400F
- OBE Peak Ground Acceleration (PGA):
0.10g(5) (6)
- Emergency Cooling Water Inlet: 950F SSE PGA : 030g(5)
- SSE Response Spectra: per Reg. Guide 1.60 SSE Time History: Envelope SSE Response -
Spectra (I) 50-year recurrence interval; value to be utilized for design of non safety-related structures only.
II) 100-year recurrence interval; value to be utilized for design for safety related structures only.
(3) ' Probable maximum flood level (PMF), as defined in ANSl/ANS 2.8, Desemtining Design Basis Flooding at Power Reactor Sites."
' (4) 1,000,000 year tomado recurrence interval, with associatedparameters based on ANSI /ANS 2.3.
(3) Free-field, at plant grade elevation.
(6) For conservatism, a value of 0.15g is employed to evaluate structural and component responses in Chapter 3. .
ill See item 3 in Section 3A.]for additionalinfomsation.
Amendment 6 2.0-2
23A6100AD Standard Plant Rev.n
] SECTION 2.1 CONTENTS TABLES Table D1]r East 2.11 Limits Imposed On SRP Section II Acceptance Criteria by ABWR Design 2.12 i
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O 2 . 1 11 Amendment 6 1
7, r 13A6100AD m ..a.a pi.., mm Li t. 2.1 LIMITS IMPOSED ON SRP -
l-
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SECTION II ACCEFI'ANCE CRITERIA BY ABWR STANDARD PLANT This section defines limits imposed on SRP l Section envelope of11 ABWR acceptance Standard Plantcriteria by (1) the plant site design parameters given in Table 2.0-1 and (2) the assumptions, both implicit and explicit, related to site characteristics employed in the l cvaluation of the 'ABWR design. These limits are presented in Table 2.11 for the following five SRP Section 11 categories;of site characteristics:
(1)' Geography and Demography; (2) Nearby Industrial, Transportation and
Military Facilities; (3) Meteorology; (4) Hydrology Engineering;and (5)' Geology, Seismology and Geotechnical Engineering.
See Subsection 2;2. for interf ace requirements.
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23A6100AD t Stanalard Plant nev. n TABLE 2.1-1
[V0,-
LIMITS IMPOSED ON SRP SECTION II ACCEPTANCE CRITERIA BY ABWR DESIGN SRP SECTION SUBJECT LIMITS GEOGRAPHY AND DEMOGRAPHY 2.1.1 Site Location and Description None.
2.1.2 Exclusion Area Authority and None.
Control
-2.13 Population Distribution None.
NEARBY INDUSTRIAL, TRANSPORTATION AND MILITARY FACILITIES 2.2.1 Identifica! ion of Potential Identify potential hazards it the site sicinity 2.2.2 Hazards in Site Vicinity that have a probability of occurrence >10-7 p per year which produce: (1) missiles more gj energetic than the AISI/ANS-23 missile spectra, or (2) pressure affects in excess of E the design basis tornado.
2.23 Evaluation of Potential Accidents Evaluate only those potential hazards identified above.
METEOROLOGY 23.1 Regional Climatology Per Table 2.0-1.
23.2 Local Meteorology None.
233 Offsite Meterological Measure- None.
ment Programs 23.4 Short Term Dispersion Estimates Show that the site meteorological dispersion for Accidental Atmospheric values as calculated in accordance with Releases Regulatory Guide 1.145, and compared to dose values given in Chapter 15, result in doses i less than stipulated in 10CFR100 and the j applicable portions of SRP Sections 11 and 15.
b
.v 23.5 Long-Term Diffusion Estimates None. I Amendment 6 2.1-2 i
ABM 23A6100AD ,
Standard Plant REV.B f I
TABLE 2.1-1 LIMITS IMPOSED ON SRP SECTION II ACCEPTANCE CRITERIA BY ABWR DESIGN (Continued)
SRP SECTION SUBJECT LIMITS HYDROLOGY ENGINEERING 2.4.1 Hydraulic Desenption Per Table 2.0-1.
2.4.2 Floods Per Table 2.0-1.
2.4.3 Probable Maximum Flood on None.
Streams and Rivers 2.4.4 Potential Dam Failures Demonstrate that failure of existing r,nd Seismically Induced potential upstream or downstream ww.er control structures will not etceed ficoding i foot below grade.
2.4.5 Probable Maximum Serge and Probable maximum serge and sicche flooding 11 Sicche Flooding foot below grade.
2.4.6 ' Probable Maximum Tsunami Probable maximum tsunami flooding 11 foot below grade.
2.4.7 Ice Effects None.
2.4.8 Cooling Water Channels Demonstrate that cooling water channels and and Reservoirs reservoirs supplying emergency cooling water has a failure probability of110-7 per year.
2.4.9 Channel Diversion None.
2.4.10 Flooding Protection Requirements None.
2.4.11- Cooling Water Supply Failure rate of emergency cooling water supply 1 10-7 per year.
2.4.12 Groundwater Per Table 2.0-1.
2.4.13 Accidental Releases of Liquid None.
Effluents in Ground and Surface Waters I 2.4.14 Technical Specifications and None.
Emergency Operation Requirement Amendment 6 2.13
MN 23A6100AD Standard Plant RFV H TABLE 2.1-1
.\ji LIMITS IMPOSED ON SRP SECTION II ACCEPTANCE CRITERIA BY ABWR DESIGN (Continued)
SRP SECTION SUBJECT LIMITS I
GEOLOGY, SEISMOLOGY AND GEOTECHNICAL ENGINEERING 2.5.1 Basic Geology and Seismic None.
Information 2.5.2.7 Vibratory Ground Motion Per Table 2.0-1.* j 2.53 Surface Faulting None.
2.5.4.8 Stability of Subsurface Per Table 2.0-1.
l Materials and Foundations 2.5.5 Stability of Slopes None.
(]
- Acceptance Criteria 11 2.5.2.7 of the SRP Section 2.5.2 specifies the minimum value of the OBE to one-half of the SSE. The ABWR Standard Plant has adopted an OBE equal to one-third of the SSE.
u ,/ j i
The relationship between the magnitude of the OBE and the SSE established in paragraph V of 10 i '
CFR 100 Appendix A is inconsistent with their definitions. The OBE is defined in 10 CFR 100 as the earthquake which could reasonable be expected to affect the plant site during the operating life of the plant, for which those features necessary for continued safe operation of the plant are designed to remain functional. The SSE is based upor 4 maximum earthquake potential which ,
produces the maximum vibratory ground motion for nich certain structures, systems, and I components are designed to remain functional. In coupling the events, as implied by the current regulatory requirement, the intent of the OBE as a reasonably likely event is lost. The use of a 100 year recurrence level of the OBE is appropriate compared to the plant life and is also appropriately conservative relative to the Uniform Building Code requirements for non-safety related structures.
Decoupling the OBE from the SSE has been an issue in the technical community for quite some time. Both industry and regulatory have recognized the inconsistency in the defm' itions and some of the undesirable results such as greatly stiffened structures and systems to meet the more l restrictive OBE stress levels.
Generic Issue 119.3, Decoupling of the OBE from the SSE, was introduced into the regulatory l process by recommendation A-3 of the Piping Review Committee. In the historical background of l this generic issue, it is noted that in developing the current regulations,it was assumed that l the OBE would serve as a separate check of those systems where continued operation was desired at a lower level of ground motion. However, in practice, the assumed load factors, damping, stress n levels, and service limits have caused the OBE, rather than the SSE, to control the design for ,
many systems including concrete and steel structures and nuclear piping. In addition, seismic
(]
design for OBE accounts for certain safety-related factors such as fatigue and seismic anchor Amendment 6 2.14
1 ABM 23A6100AD Standard Plant REV.B TABLE 2.1-1 LIMITS IMPOSED ON SRP SECTION II ACCEPTANCE CRITERIA O
BY ABWR DESIGN (Continued)
SRP SECTION SUBJECT LIMITS GEOLOGY, SEISMOLOGY AND GEOTECHNICAL ENGINEERING movement that ate not considered in the design for the SSE. As a further consequence, structures and systems have been greatly stiffened to meet restrictive OBE allowable stress levels. This stiffening is detrimental to actual plant conditions.
Since the ABWR is a plant of the future and there is sufficient evidence that more flexible designs cau exhibit reliability and safety levels equal to or greater than original stiffer designs (such as piping designs studied by (NUREG/CR-4263), this lower magnitude OBE is warranted.
O 1
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O Amendment 6 2.1-5 1
ABWR. ux6=40 i, ~ Standard Plant REV.B SECTION 2.2 CONTENTS i i
Table Tit]s Page 2.2-1 Envelone of ABWR Standard Plant Site Desien Parameters 2.2-1 2.2.2 Standard Review Plan Site Characteristics 2.2 '
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1 Amendment 6
_7___
ABM 23A6100AD ,
Se==dard Plant aw. n
- 2.2 INTERFACES
{f 2.2.1 Envelope of Standard Plant Design j
I Parameters Compliance with the envelope of ABWR Standard
[4.
. Plant site design parameters of Table 2.01 shall be demonstrated. (See Section 2.0) 2.2.2 Standard Review Plan Site Characteristics Identification and description of all differences from SRP.Section II Acceptance Criteria for site characteristics (as augmented ->
by Table 2.11) shall be provided. Where such differences exist, the evaluation shall discuss how the alternate site characteristic is acceptable.
s%-
O Amcadenens 6 2.2-1
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ __ ._ _ ._3
23A6100AB Standard Plant . REV A CHAPTER 3 l
,3 V TABLE OF CONTENTS Section Title Eage 1
3 DESIGN OF STRUCTURES COMPONENTS, EQUIPMENT, AND SYSTEMS 3.1 CONFORMANCE WITH NRC GENERAL DESIGN CRITERIA 3.1-1 3.1.1 Summary Description 3.1-1 3.1.2 Evaluation Against Criteria 3.1-1 3.2 QASSIFICATION OF STRUCTURES. COMPONENTS.
AND SYSTEMS 3.2-1 3.2.1 Seismic Classification 3.21 3.2.2 System Ouality Group Classifications 3.2-1 3.23 Safety Classifications 3.2-1 3.2.4 Correlation of Safety Classes with
( Industry Codes 3.2-3 3.2.5 Non-Safety Related Structures, Systems, and Components 3.2-3 3.2.6 QualityAssurance 3.2-5 33 WIND AND TORNADO LOADINGS 33.1 1
33.1 Wind Loadings 33-1 33.2 Tornadoloadings 33-1 333 Interfaces 33-2 33.4 References 33-2 3.4 WATER LEVEL (FLOOD) DESIGN 3.4-1 3.4.1 Flood Protection 3.4-1 3.4.2 Analytical and Test Procedures 3.4-7
}
3.0-il Amendment 1 l
1 l
23A6100AE Standard Plant Rev.A )
CHAPTER 3 TABLE OF CONTENTS (Continued) h Section Title Page l
J i
3.43 Interfaces 3.4-7 i i
3.4.4 References 3.4-7 3.5 MISSILE PROTEGILQM 33-1 33.1 Missile Selection and Description 33-1 33.2 Structures, Systems, and Components to be Protected from Externally Generated Missiles 33-7 333 Barrier Design Procedures 3.5-7 33.4 Interfaces 3.5-7 33.5 References 33-8 3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING 3.6-1 3.6.1 Postulated Pipping Failures in Fluid Systems Inside and Outside of Containment 3.6-1 3.6.2 Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping 3.6-6 3.63 Leak-Before-Break Evaluation Procedures 3.6-22 3.6.4 Interfaces 3.6-27 3.6.5 References 3.6-27 3.7 SEISMIC DESIGN 3.7-1 1
! 3.7.1 SeismicInput 3.7-1 3.7.2 Seismic System Analysis 3.7-4 3 . 0 -11i Amendment 1
23A6100AE Standartl Plant Rev n i l
s (1) One locked closed isolation valve inside and significant release of radioactivity.
( ") one locked closed isolation valve outside containment; or ;
(
(2) One automatic isolation valve inside and one !
locked-closed isolation valve outside containment; or The design of the isolation systems detailed (3) One locked-closed isolation valve inside and in the following sections meets the requirements one automatic isolation valve outside of Criterion 55.
containment. A simple check valve may not be used as the automatic isolation valve outside For further discussion, see the following containment; or sections:
(4) One automatic isolation valve inside and one Chapter /
automatic isolation valve outside Section Tills containment. A simple check valve may not be used as the automatic isolation valve outside (1) 5.2 Integrity of Reactor Coolant containment. Pressure Boundary Isolation valves outside containment shall be (2) 6.2.4 Containment Isolation Systems located as close to the containment as practical and upon loss of actuating power, automatic (3)7 Instrumentation and Controls isolation valves shall be designed to take the position that provides greater safety. (4) 15 Accident Analyses Other appropriate requirements to minimize the (5) 16 Technical Specifications (l
C probability or consequences of an accidental rupture of these lines or of lines connected to 3.1.2.5.7 Criterion 56 - Primary Containment them shall be provided as necessary to assure Isolation adequate safety. Determination of the appropri-ateness of these requirements, such as higher 3.1.2.5.7.1 Criterion 56 Statement quality in design, fabrication, and testing, additional provisions for inservice inspection, Each line that connects directly to the con-protection against more severe natural phenomena, tainment atmosphere and penetrates primary reac-and additional isolation valves and containment, tor containment shall be provided with contain-shall include consideration of the population ment isolation valves as follows, unless it can density, use characteristics, and physical be demonstrated that the containment isolation characteristics of the site environs. provisions for a specific class of lines, such as instrument lines, are acceptable on some 3.1.2.5.6.2 Evaluation Against Criterion 55 other defined basis:
The reactor coolant pressore boundary as (1) One locked-closed isolation valve inside and defined in 10CFR50, Section 50.2, consists of the one locked-closed isolation valve outside reactor pressure vessel, pressure retaining containment; or appurtenances attached to the vessel, valves, and pipes which extend from the reactor pressure (2) One automatic isolation valve inside and one vessel up to and including the outermost isola- locked-closed isolation valve outside tion valves. The lines of the reactor coolant containment; or pressure boundary which penetrate the contain-ment have suitable isolation valves capable of (3) One locked closed isolation valve inside and isolating the containment, thereby precluding any
/]
N.)
Amendment 6 11 37 l
ABM 234stooxa Standard Plant REV.A one automatic isolation valve outside 3.1.2.5.8.2 Evaluatloa Against Criterion 57 containment. A simple check valve may not be used as the automatic isolation valve Each line that penetrates containment and is outside containment; or not connected to the containment atmosphere and is not part of the reactor coolant pressure (4) One automatic isolation valve inside and one boundary has at least one isolation valve automatic isolation valve outside contain- located outside containment.
ment. A simple check valve may not be used as the automatic isolation valve outside Details demonstrating conformance with Crite-containment. rion 57 are provided in the following section:
Isolation valves outside containment shall be Chapter /
located as close to the containment as practical Section Illh and upon loss of actuating power, automatic isolation valves shall be designed to take the (1) 6.2.4 Containment Isolation Systems position that provides greater safety.
3.1.2.6 Group VI Fuel and Reactivity Control 3.1.2.5.7.2 Evaluation Against Criterion 56 3.1.2.6.1 Criterion 60 - Control of Releases of The manner in which the containment isclation Radioactive Materials to the Environment system meets this requirement is detailed in the following sections: 3.1.2.6.1.1 Criterion 60 Statement Chapter / The nuclear power unit design shall include Section Illh means to control suitably the release of radio-active materials in gaseous and liquid effluents (1) 6.2.4 Containment Isolation Systems and to handle radioactive solid wastes produced during normal reactor operation including (2) 7 Instrumentation and Controls anticipated operational occurrences. Sufficient holdup capacity shall be provided for retention (3) 15 Accident Analyses of gaseous and liquid effluents containing radioactive materials particularly where (4) 16 Technical Specifications unfavorable site environmental conditions can be expected to impose unusual operational 3.1.2.5.8 Critrrion 57 Closed System limitations upon the release of such effluents Isolation Valves to the environment.
3.1.2.5.8.1 Criterion 57 Statement 3.1.2.6.1.2 Evaluation Against Criterion 60 Each line that penetrates primary reactor 3.1.2.6.1.2.1 Ettluent Monitoring System l containment and is neither part of the reactor l coolant pressure boundary nor connected directly The Nuclear Island effluent monitoring )
to the containment atmosphere shall have at least system is designed with provisions to prevent l
one containment isolation valve which shall be radioactivity releases during accidents from l either automatie, or locked-closed, or capable of exceeding the limits of 10CFR100. I remote manual operation. This valve shall be outside the containment and located as close to Ventilation air from the various Nuclear Is-the containment as practical. A simple check land areas is continuously monitored, and con-valve may not be used as the automatic isolation trolled ventilation air for containment areas is valve. exhausted through HEPA and charcoal filters if el Amendment 1 3.138 1
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ABM 23A6100AE Standard Plant REV B (v ) 3.1.2.6A.2 Evaluation Against Criterion 63 3.1.2.6.5.2 Evaluation Against Criterion 64 3.1.2.6A.2.1 Fuel Storage System Means have been provided for monitoring l radioactivity releases resulting from normal and Appropriate systems have been provided to meet anticipated operational occurrences and from the requirements of this criterion. A postulated accidents. The following releases malfunction of the fuel pool cooling and cleanup are monitored:
system which could result in loss of residual heat removal capability and excessive radiation (1) gaseous releases, and levels is alarmed in the control room. Alarm conditions include low fuel pool cooling water (2) liquid discharge.
pump discharge pressure and high/ low level in the fuel storage pool and skimmer surge tanks. In addition, the containment atmosphere is System temperature is also continuously monitored monitored.
and alarmed in the control room. The area radiation monitors indicate radioactivity in this For further discussion of the same means and area and initiate an alarm in the control room on equipment used for monitoring reactivity re.
abnormal radiation, leases, see the following sections:
The fuel storage system satisfies the Chapter /
requirements of Criterion 63. Section Iiik For further discussion, see the following (1) 5.2.5 Detectio i of Leakage through sections: Reactor Coolant Pressure Boundary (3 Chapter /
() Section Iilk (1) 9.1 Fuel Storage and Handling 3.1.2.6A.2.2 Radioactive Waste System The design of the radioactive waste system (RWS) is out of the Nuclear Island scope. The Nuclear Island / remainder of plant interfaces for the RWS are addressed in Subsection 11.1.8.
l 3.1.2.6.5 Criterion 64 - Monitoring {
Radioactivity Releases 3.1.2.6.5.1 Criterion 64 Statement f
l Means shall be provided for monitoring the j reactor containment atmosphere, spaces containing '
components for recirculation of loss.of-coolant accident fluids, effluent discharge parts, and the plant environs for radioactivity that may be released from normal operations including anticipated operational occurrences and from postulated accidents.
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Amendment 6 3.1-41
23A6100AE
' 00 - '---d ]>launt REV.B TABLE 3.21
,. 1
- l. CLASSIFICATION
SUMMARY
(Continued) l l
Quality Group Quality 14ca. Classi- Assurance Seismic Safet4 Princinal Component" Gail Man
- Beation Requirement' Catan00 b'alta B1 Reactor thssure Vessel System /
Fuel Assemblies
- 1. Reactor vessel 1 C A B I g e
- 2. Reactor vessel support skirt 1 C A B I and stabilizer
- 3. Reactor vessel appurtenances 1 C A B 1 (g)
-pressure retaining portions -
- 4. Supports for CRD housing, . 1 C A B I in-core housing and recircu-lation internal pump
- 5. Reactor internal structures - 3 C --- B I safety related components
'O including core support structures (See Subsection 3.9.5)
- 6. Reactor internal structures - N C --- -- ---
non. safety related components (See Subsection 3.9.5)
- 7. Control rods 3 C -- B 1
- 8. Power range detector hardware 3 C --
B I including startup range detector
- 9. Fuel assemblies 3 C --- B I 52 Noelear Boiler System
- 1. Vessels -levelinstrumenta- 2 C B B I .
tion condensing chambers
- 2. Vessel- air accumulators 3 C C B 1 (for ADS and SRVs)
- 3. Piping including supports - 2/3 C B/3 B I (h) R safety / relief valve discharge $
Amendment 3 3.28
MM 23A6100AE Standard Plant REV B TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued)
Quality Group Quality Safety Loca- Classt. Assurance Seismic Prineinal Comnonenta flau b ge ficationd Reautrement* Catenorvf h B2 Nuclear Boiler System (Continued)
- 4. Piping including supports - 1 C,SC A B 1 main steamline (MSL) and feed.
water (FW)line within outermost isolation valve g 5. Piping incuding supports - 2 SC B B I g MSL and FW from outermost isolation valve to and including seismic interface restraint and FW from outermost isolation to and including shutoff valve
- 6. Piping including supports - MSL N T B - -- g; from the seismic interface AA restraint to the turbine stop valve
- 7. Piping FW from shutoff valve N T D ---
I to seismic interface restraint
- 8. Piping FW beyond seismic N T D -- ---
interface restraint
- 9. Seismic Interface Restraint - 2 SC B B I MSL/FW
- 10. Pipe whip restraints - MSL/FW 3 SC,C --- B --
g 11. Piping including supports other o within outermost isolation valves n
- a. RPV head vent 1 C A B ! (g)
- b. RPV head spray 1 C A B I (g)
- c. Main steam drains 1 C,SC A B I (g) g' 3 12. Pipingincuding supports other o beyond outermost isolation valves
~
- a. RPV head vent N C D -- --
- b. RPV head spray N SC D --- --
- c. Main steam drains N SC D --- ---
O Amendmem 6 3.2-9
r ,
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ABM 234aooxe i Etandard Plant arv g
/~3 TABLE 3.21 ,
j (J CLASSIFICATION
SUMMARY
(Continued) !
l I
Quality l ._- Group Quality l Safety Loca. Classi- Assurance Seismic Princloal Comnonenta Gau b he ficationd Requirement' Catenonf h E3 leak Detection and Isolation System (Continued)
- 6. Differential pressure 3/N X -- B 1/ - (z) switches
- 7. Square root converters 3/N X ---
B 1/ - (z)
- 8. Differential flow summers 3/N X -- B 1/--- (z)
- 9. Differential flow switches 3 X --- B 1
- 10. Timer switches 3 X --- B I
- 11. Power supplies 3 C,SC ---
B I
/
O 12. Radiation monitor N SC --- -- 1
- 13. Instrument lines 3 C,SC B B I
- 14. Sample lines
- 2/N C,SC C/D/--- B/ -- 1/--
- 15. Flow transmitters N SC --- -- -
- 16. Cables 3/N SC,RZ,X - B/- - I/---
E4 RCIC System
- 1. Piping including supports with- 1/2/3 C,SC A/B/C B 1 (g)
R in outermost isolation valves
- 2. Piping including supports - N O SC D - - (g)
Suction line from condensate storage tank beyond second shut-off valve and vacuum pump discharge line from vacuum pump to containment isolation valves p
- 7hese sansple lines are totally withirt containntent and radiation monitoringprovides no irole:lonfunction.
Amendment 6 3.2 16
ABM 2miooxe Standard Plant nev. 8 TABLE 3.2-1 CLASSIFICATION
SUMMARY
(Continued)
Quality Group Quality loca- Classi. Assursace Seismic Safet4 Etavir3mtal* CatenoryI Principal Component
- Clan llan* fication Egigs E4 RCIC System (Continued)
, 3. Pipingincluding supports be- 2/3 C,SC B/C B I (g)
E yond outermost isolation valves B
- 4. RCIC Pump including supports 2 SC B B I i
- 5. Pump motors N C,SC --- --- I i
- 6. Valves - outer isola. 1/2 C,SC A/B B I (g) tion and within
- 7. Valves - shut-off N O,SC D --- -- (g) line from condensate storage beyond second shut off valve. ,
- 8. Valves other* 2/3 SC B/C B I (g) u !
g 9. Turbine including supports 2 SC --- B I (m) j l
- 10. Electrical modules with 3 C,SC,X - - -
B I I safety-related function
- 11. Cable with safety 3 C,SC,X --- B I function
- 12. Other mechanical and N C,SC,X --- -- --- :
electrical modules {
I 1
FI Fuel Servicing Equipment
- 1. Fuel preparations machine N SC - - - ---- ---
- 2. General purpose grapple N SC --- --- ---
i
- RCIC turbine steam admission valve, pump suction valvefrom condensate storage tank, and turbine inlet and exhaust drain valves.
Amendment 3 3.2 17 O
r 23A6100AE
.a pi.., m,,
TABLE 3.21
/~i - CLASSIFICATION
SUMMARY
(Continued) g )
Quality Group Quality l'; Loca- Classi- Assurance Seismic Safet4 Princinal Component
- Clan llen" Deation Requirement' Ca18820 h G1 Reactor Water Cleanup System (Continued)
- 12. Sample station N SC D --- -
, 13. Electrical modules and cable N SC,X D --- - - - 1 g with no safety-related function v
- 14. Electrical modules and cable 3 SC --- B 1 for isolation valves
/
O 3.2-19a Amusement 3
MM 23A6100.E Standard Plant REV.S TABLE 3.21 CLASSIFICATION
SUMMARY
(Continued)
Quality Group Quality Safety 1Aca- Classi. Assurance Seismic Princinal Comnonenta Classb ge ficationd Erguiremente Catenonf h G2 Fuel Pool Cooling and Cleanup System
- 1. Vessels including supports - N SC C --- ---
filter /demineralizers
- 2. Vessels including supports - N SC C -- ---
drain tanks ,
1
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- 3. Heat exchangers including N SC C -- ---
supports
- 4. Pumps including support and N SC C --- ---
pump motors
- 5. Piping including supports, N SC C --- ---
valves
- 6. Normal makeup system N SC,0,T C --- ---
O4 components ,
e A
- 7. RHR connections for 3 SC C B 1 safety-related makeup
$ 8. Electrical modules and N SC,X -- --- --
a cables with no safety-related function G3 Suppression Pool Cleanup System
- 1. Isolation valves and piping 2 C B B I g including supports within e outermost isolation valves n
- 2. Pumps N SC D --- --
- 3. Pump motors N SC --- - - - ---
2 o_ 4 Other piping including supports N SC D --- ---
- 5. Electrical modules and Cables N SC,X --- --- ---
O Amendment 6 3.2-20 l
MM mm
- StandardPlant nev,A HQIES
> .s
- a. A module is an assembly of interconnected components which constitute an identifiable device or piece of equipment. For example, electrical modules include sensors, power supplies, and' -i signal processors and mechanical modules include turbines, strainers, and orifices.
- b. 1,2,3, N = Nuclear safety related function designation defined in Subsections 3.2.3 and -
3.2.5. q
- c. C ' = Primary Containment H = Service Building l
M = any other location O = Outdoors onsite
- RZ = Reactor Building Clean Zone (balance portion of the reactor building outside the
~
kca+ y Containment Zone)
SC = Secondary Containment portion of the reactor building .
T = Turbine Building .
W = Radwaste Building {
X = Control Building
'd. A,B,C,D= Quality groups defined in Regulatory Guide 1.26 and Subsection 3.2.2. The structures, s) stems 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.
G
- (,,) e. B = the quality assuranu: requirements of 10CFR50, Appendix B are applied in accordance f with the quality assurance program described in Chapter 17.
- = Requirements of 10CFR50, Appendix B are not applicable,
- f. I =The design requirements of Seismic Category I structures and equipment are applied i 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, as noted in the ' Comments" column of this table,
]
- g. 1. Lines one inch and smaller which are part of the reactor coolant pressure boundary shall be ASME Code Section III, Class 2 and Seismic Category I.
- 2. Allinstrument lines which are connected to the reactor coolant pressure boundary and are utilized to actuate and monitor safety systems shall be Safety Class 2 from the outer isolation valve or the process shutoff valve (root valve) to the sensing instrumentation. j
- 3. All instrument lines which are connected to the reactor coolant pressure boundary and are not utilized to actuate and monitor safety systems shall be Code Group D from the outer isolation valve or the process shutoff valve (root valve) to the sensing instrumentation. ;
O Amendment 2 3.2 31
)
ABM 2asioore Rev. s
$1andard Plant hEIIS (Continued)
- 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.
- t 6. All safety-relaed instrument sensing lines shall be in conformance with the criteria of
- h. Relief valve discharge piping shall be Quality Group B and Seismic Category '..
Safety / relief valve discharge line (SRVDL) piping from the safety / relief valve to the quen-chers in the suppression pool consists of two parts: the first part is attached at one end to the safety / relief valve and attached at its other end to the diaphragm floor penetration.
This first portion of the safety / relief valve discharge piping is analyzed with the main steam piping as a complete system. The second part of the safety / relief valve discharge piping 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, analyzed 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 Category I. Fail safe devices are non Seismic Category I.
- j. The control rod drive insert lines from the drive flange up to and including the first valve on the hydraulic control unit are Safety Class 2, and non safety related beyond the first valve.
- k. The hydraulic control unit (HCU) is a factory-assembled engineered module of valves, tubing, piping, and stored wat:r which controls two control rod drives 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 of its internal parts differ markedly from process piping components because of the more complex functions they must provide. Thus, although the codes and standards invoked by Groups A, B, C, and D pressure integrity quality levels clearly apply at alllevels 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).
O Amendment 6 3.2-32 1
D 23A6100AE -
_m_ andard Plant REV.B ex
/ i NOTES (Cos inued)
Ci V The oesign and construction specifications for the IICU do invoke such codes and standards as can be reasonably applied to individual parts in developing required quality levels, but these codes and standards are supplemental with additional requirements for these parts and for the remaining parts and details. For example: (1) all welds are LP inspected; (2) all socket welds are inspected for gap between pipe and socket bottom; (3) all welding is performed by qualified welders; and (4) all work is donc per written procedures. Quality Group D is generally applicable because the codes and standards invoked by that group contain c!)uses 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 OC technique described.
I. Not used.
- m. The RCIC turbine is not included in the scope of standard codes. The assure that the turbine is fabricated to the standards commensurate with safety and performance requirements, General Electric has established specific design req Jrements for this component which are as follows:
- 1. All welding shall be qualified in accordance with Section IX, ASME Boiler and Pressure Vessel Code.
- 2. All pressure containing castings and fabrications shall be hydrotested at 1.5 times the design pressure.
- 3. . di high-pressure castings shall be radiograri ed according to:
ASTM E-94 E 141 E 142 maximum feasible volume E-71,186 or 280 Severity level 3
- 4. As cast surfaces shall be magnes particle or liquid-penetrant tested according to ASME Code,Section III, Paragraphs NB-2575, NC-2576, or NB 2576, and NC 2576.
- 5. Wheel and shaft forgings shall be ultrasonically tested according to ASTM A-388.
- 6. Butt welds shall be radiographer and magnetic particle or liquid penetrant tested according to the ASME Boiler and Pressure Vessel Code. Acceptance standards shall be in accordance with ASME Boiler and Pressure Vessel Code Section III, Paragraph NB-5340, NC 5340, NB-5350, or NC-5350, respectively.
- 7. Notification shall be made on major repairs and recorA maintained thereof.
- 8. Record system and traceability shall be accordii( to A ME Section Ill, NCA-4000.
- 9. Control and identification shall be according to ASME Section III, NCA-4000.
r
- 10. Procedures shall conform to ASME Section Ill, NB-5100 and NC-5100.
- 11. Inspection personnel shall be qualified according to ash E Section III, NB 5500 and V{ 4 NC 5500.
Amendment 6 3.2 33
)'
ABWR 2u6me f Standard Plant nev. s j
(
NOTES (Continued) 1
- n. Not used.
l
- o. Not used.
l
- p. Not used. )
I
- q. Section Vill 1 and ANSI B31.1 applies downstream of the outermost isolation vales.
- r. Not used,
- s. The gaseous radwaste system piping, pumps, and valve containing gaseous radwaste shall be constructed in accordance with the applicable codes of Code Group D.
I
- t. There is a limited quality assurance program for the Fire Protection System.
- u. Special seismic goalification and quality assurance requirements are applied.
- v. Not used.
- w. The condensate storage tank will be designed, fabricated, and tested to meet the intent of API Standard API 630. In addition, the specification for this tank will require: (1) 100 %
surface examination of the side wall to bottom joint and (2) 100% volumetric examination of the side wall weld joints.
- x. The cranes are designed to hold up their loads under conditions of OBE and to maintain their positions over the units under conditions of SSE.
- y. M1 off engine component < are constructed to the extent possible to the ASME Code,Section III, Class 3.
- z. Components asmciated with a safety-related function (e.g., isolation) are safer)-related.
aa. Structures which support or house safety related mechanical or electrical components are safety-related.
O Amendment 6 3.2 34
___.______..___________________j
I M 23A6100AE ftta-A-M Plant nev. s A SECTION 3.3 A) '
CONTENTS Section Dilt EAgg 33.1 Wind Loadings 33-1 33.1.1 Design Wind Velocity 33-1 33.1.2 Determination of Applied Forces 33-1 3.3.2 Tornado Loadines 33-1 33.2.1 Applicable Design Parameters 33-1 33.2.2 Determination of Forces on Structures 33-1 33.23 Effea of Failure of Structures or Components Not .>esigned for Tornado leads 33 2 3.3.3 Interfaces . 33-2
'333.1 Sit: Specific Design Basis Wind 33 2 33.1.2 Site-Specific Design Basis Tornado 33-2
(
3333 Effect of Remainder of Plant Structures, Systems, and Components not Designed for Tornado Loads 33-2 33.4 References 33-2 O 33-il Amendment 6
Standard Plant SECTION 33 TABLES Table Edt h 33-1 Importance Factor (I) for Wind Loads 33-3 33-2 Velocity Pressure Distribution and Gust Factors at Various Heights 334 1
0 32.m 9'
Amendment 1
ABM 2:462oort i Standard Plant REV B \
l 3.3 WIND AND TORNADO LOADINGS Reference 1. Reference 2 is used to obtain the effective wind pressures for cases which Refer-h (x' V) l .
ABWR Standard Plant structures which are ence 1 does not cover. Since the Seismic Cat- 1 Seismic Category I are designed for tornado and egory I structures are not slender or flexible, extreme wind phenomena. vortex-shedding analysis is not required and the above wind loading is applied as a static load. j 3.3.1 Wind IAndings j 3.3.2 TornadoImadings 33.1.1 Design Wind Velocity 3.3.2.1 Applicable Design Parameters Seismic Category I structures are designed to ,
withstand a design wind velocity of 130 mph at an The design basis tornado is described by the elevation of 33 feet above grade with a recur- following parameters:
rence interval of 100 years. See Subsection 3.3.3.1 for interface requirement. (1) A maximum tornado wind speed of 260 mph at a radius of 453 feet from the center of the 33.1.2 Determination of Applied Forces tornado; The design wind velocity is converted to (2) A maximum transnational velocity of 57 mph; velocity pressure in accordance with Reference 1 using the formula: (3) A maximum tangential velocity of 203 mph, calculated as defined in Section 3.3 of qz = 0.00256 K z (IV)2 Reference 3, based on the transnational velocity of 57 mph; where K2 = the velocity pressure exposure coefficient which depends upon the (4) A maximum atmospheric pressure drop of 1.46
(] type of exposure and height (z) psi with a rate of the pressure change of
() above ground per Table 6 of Reference 1.
0.27 psi per second in accordance with Reference 3; and I = the importance factor which depends (5) The spectrum of tornado-generated missiles on the type of exposure; appropriate and their pertinent characteristics as given values of I are listed in Table in Subsection 3.5.1.4.
3.31, See Subsection 3.3.3.2 for interface V = design wind velocity of 130 mph, and requirement.
qz = velocity pressure in psf 3.3.2.2 Determint. tion of Forces on Structures The velocity pressure (qz) distribution with The procedures of transforming the tornado height for exposure types C and D of Reference 1 loading intc effective loads and the distribu-are given in Table 3.3-2. tion across the structures are in accordance with Reference 4. The proce ure for transform-The design wind pressures and forces for ing the tornado-generated missile impact into an buildings, components and cladding, and other effective or equivalent static load on struc-structures 8 various heights above the ground tures is given in Subsection 3.5.3.1. The load-arc obtained, in accordance with Table 4 of ing combinations of the individual tornado load.
Reference 1 by multiplying the velocity pressure ing components and the load factors are in accor- i by the appropriate pressure coefficients and gust dance with Reference 4. l factors. Gust factors are in accordance with Table 8 of Reference 1. Appropriate pressure The reactor building and control building are o coefficients are in accordance with Figures 2, not vented structures. The exposed exterior 3a,3b,4, and Tables 9 and 11 through 16 of roofs and walls of these structures are designed for the 1.46 psi pressure drop. Tornado dampers j Amendment 6 3 %1
E M 23A6100AE Standard Plant REV.B are provided on all air intake and exhaust 3. ANSI /ANS 23, American National Standard, openings. These dampers are designed to Estimating Tornado and Extreme Wind withstand a negative 1.46 psi pressure. Characteristics at Nuclear Power Sites, Standards Committee Working Group 33.23 Effect of Failure of Structures or ANS--23, American Nuclear Society.
Components Not Designed for Tornado Loads
- 4. Bechtel Topical Reprt BC-TOP-3-A, Revision All safety related system and components are 3, Tornado and Extreme Wind Design Criteria protected within tornado resistant structures. for Nuclear Power Plants.
See Subsection 3333 for interface requirement.
3.3.3 Interfaces 333.1 Site-Specific Design Basis Wind The site specific design basis wind shall not exceed the design basis wind given in Table 2.0-1 (See Subsection 2.2.1).
333.2 Site-Specific Design Basis Tornado The site specific design basis tornado shall not exceed the design basis tornado given in Table 2.01 (See Subsection 2.2.1).
3.3.3.3 Effect of Remainder of Plant Struc-tures, Systems, and Components not Designed for Tornado Loads All remainder of plant structures, systems, aad components not designed for tornado loads shall be analyzed for the site-specific loadings to ensure that their mode of failure will not effect the abi'hy of the Seismic Category I ABWR Standard Plant structures, systems, and compo-nents to perform their intended safety functions.
(See Subsection 33.23) 3.3.4 References
- 1. ANSI Standard A58.1, Minimum Design Loads for Buildings and Other Structures, Committee A. 58.1, American National Standards Institute.
- 2. ASCE Paper No. 3269, Wind Forces on Structures, Transactions of the American Society of Civil Engineers, Vol.126, Part II.
O Amendment 6 3.3-2
~ _ _ _ _ _ _ -
ABWR . 2w-c Standard Elant an n SECTION 3.4
/
.Y V CONTENTS Section Htle Eagg 3A.1 Mood Protection 3.4-1 3A.1.1 Mood Protection Measures for Seismic CategoryI Structures 3.4-1 3.4.1.1.1 Mood Protection from External Sources 3.4-1 3A.1.1.2 Compartment Flooding from Postulated Component Failures 3.4-2 3A.1.1.2.1 Evaluation of Reactor Building Flood Events 3.4-2 3.4.1.1.2.1.1 Evaluation of Floor 100 (B3F) 3.4-2 3A.1.1.2.1.2 Evaluation of Moor 200 (B2F) 3.4-3 3A.1.1.2.13 Evaluation of Floor 300 (B1F) 3.4-3 3A.1.1.2.1.4 Evaluation of Floor 400(1F) 3.4-4 3.4.1.1.2.1.5 Evaluation of Floor 500 (2F) 3.4-4 3.4.1.1.2.1.6 Evaluatic,n of Moor 600 (3F) 3.4-5 3A.1.1.2.1.7 Evaluation of Floor 700 (M4F) 3.4-5 3A.1.1.2.1.8 Evaluation of Floor 800 (4F) 3A-5 3A.1.1.2.1.9 Floodmg Summary Evaluation 3.4-5 3A.1.1.2.2 Evaluation of Control Building Floodmg Events 3.4-6 3.4.1.2 Permanent Dewatering System 3.4-6 3A.2 Analytical and Test Procedures 3.4-6 3A3 Interfaces 3.4-7 3.43.1 Flood Elevation 3.4-7 3A3.2 Ground Water Elevation 3.4-7 3AA References 3.47 3 . 4 -11 Amendment 6 L . _ _ _ __ _ _ . _ . . _ _ _ _ _ _ - . - - _ _ _ _ _ ________
MM 23A6100AE SLADjard Plant Rev. n i
SECTION 3.4 TABLES 9 Table Illic East 3.4-1 Structures, Penetrations, and Access Openings Designed for Flood Protection 3.4-8. ,
l 4
3.4-2 Reactor Building Access Openings and Penetrations l Below Design Flood Level 3.4-9 i
a i
j i
l l
9I4 l 4 l
l l
l '
l 3.4-iii Amendment 6 j
l4 23A6100AE lihannla d Ph=* amr n 3A WATER LEVEL (FIDOD) DESIGN structures, systems, and components from postulated flooding. Seismic Category I O. . Criteria for the design basis for protection structures required for safe shutdown remain 1 against external flooding shall conform to the accessible during all flood conditions.
- requirements of RG 1.59. The design criteria for protection against the effects of compartment Safety related systems and components are flooding shall conform to the requirements of flood-protected either because of their location ANSI /ANS 56.11. The design basis flood levels above the design flood level or because they are are specified in Table 3.41. enclosed in reinforced concrete Seismic Category N
I structures which have the following 3A.1 Floed Ptotection requirements:
This section discusses the flood protection (1) wall thicknesses below flood level of not measures provided for Seismic Category I struc- less than two feet tures, systems, and components in the ABWR Stan-dard Nuclear Island for both external postulated (2) water stops provided in all construction Gooding and postulated flooding from component joints below flood level;
. failures.
(3) watertight doors and equipment hatches 3.4.1.1 Flood Protection Measures for Seismic installed below design flood level; and CnessayI structures (4) waterproof coating.
The safety-related systems and components in the ABWR Standard Nuclear Island are identified Waterproofing of foundations and walls of in Table 3.2.1. They are either protected Seismic Category I structures below grade is against flood damage or are not subject to damage accomplished principally by the use of water by flooding. Flood protection of safety-related stops at expansion and construction joints. In systems and components is provided for all addition to water stops, waterproofing is O. postulated design flood levels and conditions installed to protect exterior surfaces in areas !
described in Section 2.4. Fostulated flooding of prolonged exposure to water. l from component failures in the building compart-ments'does not adversely affect plant safety nor Additional speciile provisions for flood does it represent any hazard to the public. protection include administrative procedures to assure that all watertight doors and hatch Structures which house the safety related covers are locked in the event of a flood equipment and offer flood protection ere warning. If local seepage occurs through the identified in Table 3.4.1. Descriptions of these walls, it is controlled by sumps and sump pum, structures are provided in Subsection 3.8.4 and 3.8.5. Exterior or access openings and In the event of a flood, flood levels take a penetrations that are below the design flood relatively long time to develop. This allows level are identified in Table 3.4-2. ample lead time to perform necessary emergency actions for all accesses which need to be 3.4.1.1.1 Flood Protection them External protected.
Searces The safety-related components located below Seismic Category I structures that may be the design flood level inside a Seismic Category affected by design basis floods are designed to I structure are shown in Figure 1.2-2. All withstand the floods postulated in Section 2.4 safety related components located blow the using the hardened protection approach, design flood level are protected using the structural provisions with incorporated in the hardened protection approach described above.
plant design to protect safety-related O
A===d= at 6 3.41 1
ABM 23461oore Standard Plant REV B 3.4.1.1.2 Compartment Flooding from Postulated There are no interface requirements made upon Component Fallures the remainder of the plant from possible flooding in the ABWR Standard Plant buildings.
All piping, vessels, and heat exchangers with Other lines, such as storm drains and normal flooding potential in the reactor building are waste lines, interface with plant yard piping.
seismically qualified with one exception, and However, provisions are made in these lines complete failure of a non-seismic tank or piping that, should the yard piping become plugged, system is not applicable. The one exception is crushed, or otherwise inoperable, they will vent the radwaste building which contains no safe onto the ground relieving any flooded condition.
shutdown equipment.
Considering the above criteria and assump-In accordance with Reference 2, leakage cracks tions, analyses of piping failures and their are postulated in any point of moderate-energy consequences are performed to demonstrate the piping larger than nominal one-inch diameter. adequacy of the ABWR desian. These analyses The leakage flow area is assumed to be a circular are provided separately for the reactor and orifice with flow area equal to one-half of the control buildings.
pipe outside diameter multiplied by one half of the pipe nominal wall thickness. Resulting 3.4.1.1.2.1 Evaluation of Reactor Building leakage flow rates are approximated using Hood Events Equation 3 2 from Reference 1 with a flow coefficient of 0.59 and a normal operating Analysis of potential flooding within the pressure in the pipe. reactor building is considered on a floor-by-floor basis.
The only identified worst case of compartment j flooding involving a high energy line is a 3.4.1.1.2.1.1 Evaluation of Hoor100 (E3F) '
feedwater line break in the steam tunnel. All I data a:ecessary for evaluation of this case are Worst case flooding on this floor level would ;
taken from Section 15.1. result from leakage of the RHR 18" suction line )
between the containment wall and the system iso- (
No credit is taken for operation of the drain lation valve (this applies also to the HPCF, i sump pumps although they are expected to operate RCIC, and SPCU suction lines, although in during some of the postulated flooding events. smaller line sizes). Leakage from this source may cause flooding of the affected RHR heat After receiving a flood detection alarm, the exchanger (HX) room at a rate of 1.04 cubic operator has a ten-minute grace period to act in meter / minute (275 gpm) and may continue until cases when flooding can be identified and the line is repaired or equalization of water terminated by a remote action from the control level occurs between this room with the suppres- 1 room. In cases involving visual inspection to sion pool level. Flooding in the room may cause l identify the specific flooding source in the loss of functions for that particular divi-I affected area (except ECCS areas) followed by a sional system loop. This will not impair the remote or local operator action, a minimum of 30 safe shutdown capability of the reactor system. I minutes is provided for the operator. Flooding of other areas is prevented by water tight doors. Suction lines to other services In all instances of compartment flooding, a always remain submerged. Other flooding inci-single failure of an active component is dents may result from failures of other piping considered for systems required to mitigate systems penetrating the RHR HX rooms for each consequences o! .$ eitticular flooding condition. division; these events, however, upon detection The emergency cor uoling system (ECCS) rooms by sump alarms, are controllable by terminating are also evaluated on the basis of a loss of- flow with closure of valves and shutdown of coolant accident (LOCA) and a single active pressure sources.
failure or a LOCA combined with a single passive l failure 10 minutes er more after the LOCA.
O Amendment 6 3.4-2 l
m_ ___ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _._
ABM 2 moan Standard Plant an s HPCF and RCIC systems, while having the to system isolation this would result in a susceptibility to flood their respective flooding rate of 2.8 cubic meters / minute (745 compartments from the suppression pool, do so at gpm). Twenty eight cubic meters in Divisions B lower rates than the RHR system. Failure, and C are spread over a minimum of 360 square i;
however, is guarded against by watertight doors meters, and therefore can be restrained by so that flooding in one compartment does not raised sills at divisional entries; the area in propagate to other areas. Division A is limited and watertight scaled doors will prevent further flooding.
A failure in a nondivisional area, the failure of the 8" suppression pool cleanup system suction in the fourth quadrant, SWCU and CUW F/D line before the isolation valve will perait valve rooms, holding pumps and HNH pump and HX flooding of the fourth quadrant uncontrollably by rooms, may experience flooding from various system elements. The flooding rate, driven by system line failures. The maximum flooding will suppression pool head, would not exceed .27 cubic be from an 8' line at pump pressure resulting in meter / minute (or 70 spm) and, depending upon the discharge of .92 cubic meter / minute (240 gpm).
rate of manual repair, may permit 25 30 cubic meters of water to escape into the fourth The leakage may propagate over the fourth quadrant. Certain functions of the SPCU, CUW, quadrant but an area of 300 square meters is and backwash systems may be lost, but because of available, so that depth adequate to intrude watertight doors on divisional areas no essential into divisional areas, over raised sills will functions would be lost and plant integrity would not occur. Alarming and prompt operator not be in question. isolation of these systems is then performed.
Firefighting in the divisional rooms on this Firefighting on this floor will be accom-floor will be done by sprinklers in the plished manually with a flow of .57 cubic meter /
watertight rooms. Manual firefighting in the minute (150 gpm). Areas of activity are rater fourth quadrant and in the HCU room will bring in large so that this quantity of water (less than O .57 cubic meters / minute (150 gpm). Such for other events above) presents no problem.
activities would not create a water depth which would cause concern. Firefighting in the RIP & FMCRD maintenance rooms will also have little impact; 20 cm (8 3.4.1.1.2.1.2 Evaluation of Floor 200 (B2F) inch) high sills and curbs prevent movement of water from area of application and thus maintain Flooding events on this floor may result from divisional integrity, piping failures in the RHR pipeways, or from piping systems of the HPCF and RCIC. Maximum 3.4.1.1.2.1.3 Evaluation of Floor 300 (B1F)
Gooding would occur from a 10' RHR pressure line at a flow rate of 1.34 cub!c meter / minute (354 Primary flooding events on this floor are spm). Division B and C areas have ampic area to associated with pipeways and pipechases utilized i
spread this water (13.4 cubic meters by the end by RHR,HPCF, and RCIC systems. Maximum icakage of the ten minute response period) but Division A is that postulated for a 10' RHR pressure line has limited flood spread area and will have failure in rooms that are connected to rooms watertight doors to prevent a depth of 0.3 0.4 below, raised sills in these rooms prevent meters (fifteen inches) from flooding Division C intrusion of water from a failed division into area. Division C has an area adequate to another division while clean access control minimize water depth on floors but will have access walls prevent propagation into emergency watertight hatches between Division C areas on electric rooms. Water may spread after a few Floor 200 and Division A areas on Floor 100. minutes over the sill and into the corridors, leakage alarms are provided in these areas for but the area is sufficiently large to keep the detection and operator response. level to 4-5 cm (1-2 inches) during the response period ending in system isolation. Equipment In associated divisional compartments failures hatches in the controlled access area are sealed of 16" RCW pressure lines may also occur. Prior watertight to prevent propagation of a failure Amendment 6 3.4-3
23A6100AE Standard Plant REV.B into a different division below. Minor water Flooding may penetrate the fourth quadrant at a flooding down stairwells may also occur. minimum depth, and leakage may also occur l
through hatches to nondivisional areas below.
Flooding in the emergency electric rooms A, B and C and the remote shutdown rooms may occur Emergency diesel generator A, B and C rooms from lea': age or failures in the heating and contain cooling water piping to components of ventilating chi' ed water suppl) or Emergency this system. Flooding may occur from failures HVAC cooling water system. These failures are of 8" RCW piping serving these cooling needs at limited in potential water release by line a maximum rate of .9 cubic meter / minute (240 inventory and surge tank capacity and will not gpm), which will fill the floor area and escape exceed 8 cubic meters (2000 gallons), causing a into the corridor, with potential cascading down i total water depth of 4-5 cm (1-2 inches), the stairwell. The water will spread over the i side are.as on the lower floor while action to Equipment mounted on pedestals of 20 cm (8") isolrie the failed system takes place. Intru-and door sills at area entrances will prevent sion of we r into other divisional areas is propagation and provide containment for this prevented by .aised sills on the entry passages.
water quantity until response to the failures is made. Leakage of lubricating oilis also possible in the diesel generator rooms, but level indica-Firefighting activity in all areas of this tion provides a continuing control on this level are carried out by manual means at a source. Even major leakage will be contained in maximum rate of .57 cubic meter / minute (150 gpm) the subject rooms due to the small inventory of and no greater effects than those already fluid available.
considered will occur. Raised door sills prevent intrusion of firefighting water into unaffected Firefighting in the diesel generator area division rated rooms. will be provided by CO2 releases. Other firefighting will be by hoses but will be of Failures in the CUW and SPCU systems filter / smaller volumes than those considered, and will demineralizers and associated piping may occur as be of limited duration.
on Floor 200 but will spread over a comparable l area or drain down to Floor 200 or 100 so diat 3A.1.1.2.1.5 Evaluation of Floor 500 (2F) adequate time is available for detection and subsequent system isolation. This floor contain emergency diesel generator l equipment such as fuel lines, control panels, 1 3A.1.1.2.1A Evaluation of Floor 400 (1F) and air cooling systems. The diesel generator rooms are protected from fire by CO 2, and ,
Flooding from the RHR, HPCF, and RCIC systems flooding can occur only from cooling water or ;
may occur in valve rooms A, B and C. Maximum chilled water lines at a rate of .3 cubic flooding is a failure of the 10" RHR pressure meter / minute (84 gpm). With an available flow line with leakage flow of 1.34 cubic meter / area of 144 square meters, flow over the sill minutes (354 gpm). With a room area of about 34 may occur in by 10 minutes. The water will not square meters flow over 20 cm (8') sills occur in intrude upon other division areas but may begin 3-4 minutes but a much larger area is availaR to descend the stairwell.
in corridors and rooms outside these division rooms, so that by the end of the ten minute In the fourth quadrant, leakage may occur l response period, a water level of 2 3 cm (one from the fuel pool cooling system 8 lines. The I
inch) maximum is seen g:nerally outside the maximum leakage rate would be 0.9 cubic meter / l affected room. Raised sills on entry doors tc, minute (240 gpm) but an area of several hundred other divisions prevent flooding propagation into square meters ir available to spread the water other divisions; separation walls between while detection, alarm, and system isolation is controlled access and clean access areas prevents accomplished.
flood propagation to diesel generator areas. I l
Amendment 6 3.4-4
ABM 234sioase me==da d Plant ' nev. s Firefighting in controlled access areas will Flooding on this level may also occur from be done manually under controlled conditions at a room cooling systems or from firefightir ;
O lower flow rate than the above. efforts. Cooling system failures in air supply, exhaust or filter rooms may allow flooding at The steam tunnel area is separated by a the rate of .3 cubic meter / minute (80 gpm) which firewall and scaled doors so that flooding events will flow out into adbcent ;orridor areas If in that area will not propagate into mechanical undetected for 10 minutes, the approximate 3 division rooms on Floor 500. cubic meter (800 gallons) released may create a depth of a few millimeters over the available 3A.1.1.2.1.6 Evaluation of Moor 600 (3F) floor area; a very limited amount of water will cascade down the stairwells. Divisional areas Flooding events at this floor level may encompassing the three emergency electric supply involve fuel oil as well as water. Those fans and the RIP A exhaust willinclude raised divisional rooms associated with the emergency sills to preclude water intrusion although water diesel generator fuel tank and cooling system, depth will be slight. Eqaipment pedestals will have the potential of leakage from the fuel also minimize flooding impact on all equipment.
storage tanks. These rooms must accommodate leakage of 11.4 cubic meter (3000 gallons) for Firefighting activities in this area would each division. Twenty em (8 inches) sills on cause water inflow of .57 cubic meter / minute entry to these areas successfully contain all the (150 gpm) under controlled conditions and volume in the tanks. Leakage from these tanks expected water intrusion is no naore than that will also be monitored through safety grade level above.
indication and alarm equipment so that protracted leakage as well as gross leakage can be 3A.1.1.2.1.8 Evaluatios ef Floor 800 (4F) identified. The rooms are protected by CO2 firefighting system. Water flooding may occur Flooding on this floor can be caused by from the cooling system at about'.15 cubic rupture of the RCW surge tanks A, B & C piping.
meter / minutes (41 gpm). If undetected for However, each tank and its associated piping is O)
\ several hours water may begin cascading down the located in a separate compartment which cm be nearest stairwell but is prevented from entering scaled off in the event of accidental floodeig.
other division areas by raised sills. The use of raised sills on entry ways will contain the seepage to the flooded area. Also, la the SGTS areas, the room cooling equipment the use of pedestals for equipment installation may cause flooding at a rate .15 cubic meter (41 of the RIP supply and exhaused fans and for the gpm). Raised sills prevent intrusion of water DG-C exhaust fans will guard against flooding into rooms of another division. Flooding may this equipment, also occur from manual firefighting in equipment maintenance areas'or from leakage from the Flooding in the main reactor hall may occur standby liquid control tanks. Maximum tank leak from reactor service operations, but will be rate will be .1 cubic meter / minute (25 gpm) so drained into service pools. Firefighting water that a response to tank level alarms within 10 expended into this area would occur at a maximum minutes willlimit loss to one cubic meter (or rate of .57 cubic meter second (150 gpm) but 250 gallons). Large floor areas permit spread of vill spread over the large service area water at limited depth, available. Minor amounts of water may find the way to stairwells, but would not impede 3A.1.1.2.1.7 Evaluatice of Floor 700 (M4F) operations.
Flooding in the FMCRD panel rooms may occur 3A.1.1.2.1.9 Flooding Summary Evaluation from firefighting activities at an input rate of
.57 cubic meters / minute (150 gpm). Since these Floor-by-floor analysis of potential pipe activities are manually controlled, any excessive failure generated flooding events in the reactor l depth of water will be noted and action taken to building shows the following:
mitigate water intrusion to other areas.
A==d=*nt 6 3.4-5
ABM 22461oore Standard Plant RW B (1) Where extensive flooding may occur in a Maximum flooding may occur from leakage in a division rated compartment, propagation to 6-inch chilled water line at the maximum rate of other divisions is prevented by watertight .6 cubic meter / minute (160 gpm). Early doors or scaled hatches. Flooding in one detection by control room personnel will limit division is limited to that division and the extent of flooding which will also be flood water cannot propagate to other mitigated by drainage to exterior of the divisions. building. Total release from the chilled water system will be limited to line inventory and (2) Leakage of water from large circulating surge tank volume; spillage of more than 6 cubic water lines, such as reactor building meter (1500 gallons) is unlikely, cooling water lines may flood rooms and corridors, but through sump alarms and The failure of a cooling water. line in the leakage detection systems the control room mechanical room of the building may result in a is alerted and can control flooding by comparable amount of water flooding that room; system isolation. Divisional areas are elevation differences and separation of the protected by watertight doors, or where only mechanical functions from the remainder of the limited water depth can occur, by raised control building prevent propagation of the sills with pedestal mounted equipment within water to the control area.
the protected rooms.
The evaluation of the control building (3) Limited flooding that may occur from manual flooding events is summarized as follows:
firefighting or from lines and tanks having limited inventory is restrained from Flooding events that may result from failure entering division areas by raind sills and of the service water systems or firefighting elevation differences, systems within the control room do not inhibit plant safety. The proximity of the control area Therefore, within the reactor building, to computer and control racks ensures rapid internal flooding events as postulated will not response to fire alarms or firchead actuation.
prevent the safe shutdown of the reactor. Those water lines carrying water through control areas are shielded to divert water to non 3A.1.1.2.2 Evaluation orControl Building critical areas. Service equipment rooms may Flooding Events build up limited water levels from cooling water or chilled water line failures, but elevation The control building is a four story building differences and raised sills prevent intrusion housing, generally in separate areas, the control of water into control areas; adequate drainage room proper, control and instrument cabinets, and area is available outside the building. Cwrol mechanical equipment (HVAC and chillers) neces- room response to those various levels of sary for building occupation and environmental flooding may ext nd from system isolation and j
control for computer and control equipment. correction to reduction of plant load or ;
Water services to the control building comprise shutdown, but control capability is not compro- j 6 inch fire protection line,6-inch cooling water mised by any of the postulated flooding events. I line to the chiller condenser and a 6-inch chilled water heater. Smaller lines supply 3A.1.2 Permanent Dewatering System l drinking and sanitary water, and makeup for the !
chilled water system. Areas with water pipe There is no permanent dewatering system )
routed through are supplied with floor drains and provided for in the flood design. !
protection supplied to route leakage to the floor so that control or computer equipment is not 3.4.2 Analytical and Test Procedures ;
subjected to water. In these areas where water '
infusion cannot be tolerated, the access door Since the design flood elevation is one foot sills are raised.
e' below the finished plant grade, there is no ,
Amendment 6 3.44 l
r r !
I c ABM 23462aoxe 1 seenAard Plant REV.B dynamic force due to flood. The lateral
.ph hydrostatic pressure on the structures due to the design flood water level, as well as ground water and soil pressures, are calculated.
Structures, systems, and components in the ABWR Standard Nuclear Island designed and analyzed for the maximum hydrostatic and hydrodynamic forces in accordance with the loads and load combinations indicated in Subsection 3.8.4.3 and 3.8.5.3 using well established nicthods based on the general principles of engineering mechanics. All Seismic Category I structures are in stable condition due to either moment or uplift forces which result from the
- proper load combinations including the design basis flood.
3A.3 Interfaces 3A.3.1 Flood Elevation The design basis flood elevation for the ABWR Standard Plant structures is one foot below grade.
, 3A.3.2 GroundWaterElevation The design basis ground water elevation for the ABWR Standard Plant stru:tures is two feet
' below grade.
3AA References
- 1. . Crane Co., Flow of Fluids Through Valves, Fittings, and Pipe, Technical Paper No.
410, 1973.
- 2. ANSI /ANS 56.11, Sta'ndard, Design Criteria
. for Protection Against the Effects of
' Compartment Flooding in Light Water Reactor Plants.
- 3. Regulatory Guide 1.59, Rev. 2 Design Basis Floods for Nuclear Power Plants.
i O I I
Annadment 6 34-7 j
ABM 23462oo4e Standard Plant arv. s Table 3.41 STRUCTURES, PENETRATIONS, AND ACCESS OPENINGS DESIGNED FOR FLOOD PROTECTION Structure Reactor Buildine Control Building Design Flood Level (mm) 11,700 11,700 Reference Plant Grade (mm) 12,000 12,000 Base Slab (mm) -8,200 2,700 ActualPlant Grade (mm) 12,000 12,000-Access Opening and Table 3.4.2 None Penetrations Below Design i Flood level
/
/
%d Amendment 6 144
ABV/R 2346 onis Standard Plant REY.B Table 3.4 2 REACTOR BUILDING ACCESS OPENINGS AND PENE1 RATIONS BELOW DESIGN FLOOD LEVEL Interface Number Llat.Eunist Elevation (mm)
Clean Area Access + 4800 P-21A RCW Supply + 1800 P 21B RCW Supply + 1800 P 21C RCW Supply + 1800 P-21A RCW Return -5200 P 21B RCW Return 5200 P-21C RCW Return -5200 E-22B HPCF Condensate Supply -5200 E-22C HPCF Condensate Supply -5200 NOTE:
0 The above openings are between reactor building and turbine building and are thus not exposed to ground water or outside ground flooding.
I r
O l
Amendment 6 3.4-9
ABWR 22^6ioo^s Standard Plant arv s generated from other natural phenomena, The 3.5.3.1 local Damage Prediction e(U).
missiles considered in the design are per Reference 8; ' The prediction of local damage in the impact -
area depends on the basic material of construc-
. 3.5.1.5 Site Proximity Missiles Except tion of the structure or barrier (i.e., concrete .
Aircran- or steel). The corresponding procedures are presented separately. Composite barriers are External missiles other than those generated not utilized in the ABWR Standard Plant for . l by tornados are not considered as a design basis' missile protection.
" (i.e.1 10 4 er p year).
3.53.1.1 Concrete Structures and Barriers 3.5.1.6 Aircran Hazards .
The mo6fied Petry formula (Reference 3) is Aircraft hazards are not a desji n basis event applied analytically for missile penetration in for the Nuclear Island (i.e.110*' per year). concrete. To prevent perforation, a minimum concrete thickness of 2.2 times the penetration 3.5.2 Structures, Systems, and thickness determined for an infinitely thick Components to be Protected from ' concrete slab is employed. In the event that
- i - Externally Generated Missiles spalling or scabbing is unacceptable, a minimum concrete thickness of 3 times the penetration The sources of external missiles which could thickness de
- ermined for an infinitely thick
. affect the safety of the plant are identified in concrete slab is provided. These design Subs'ection 3.5.1. Certain items in the plant are procedures have been substantiated by full-scale required to safely shut down the reactor and impact tests in which reinforced concrete panels maintain it in a safe condition assuming an (12 to 24 inches thick, 3000 psi design additional single failure. These items, whether strength) were impacted by poles, pipes, and L they be structures, systems, or components, must. rods simulating tornado-borne debris (Reference therefore all be protected from externally 4).
generated missiles.
3.53.1.2 Steel Structures and Barriers These items are the safety related items listed in Table 3.2-1. Appropriate safety The Stanford equation (Reference 5) is classes and equipment locations are given in this applied for steel structures and barriers, table. All of the safety related systems listed are located in buildings which are designed as 3.53.2 Overall Damuse Prediction tornado resistant. Since the tornado missiles are the design _ basis missiles, the systems, The overall response of a structure or structures, and components listed are considered barrier to missile impact depends largely upon to be adequately protected. Provisions are made the location of impact (e.g., near mid-span or to protect the charcoal delay ranks against near a support), dynamic properties of the tornado missiles. structure /barrhr and missile, and on the ki-netic energy of the missile. In general, it has See Subsection 3.5.4.1 for interf ace been assumed that the impact is plastic with all requirement. of the initial momentum of the missile trans-ferred to the structure or barrier and only a 3.5.3 Barrier Design Procedures portion of the kinetic energy absorbed as strain energy within the structure or barrier.
. .The procedures by which structures and barriers are designed to resist the missiles After demonstrating that the missile does not described in Subsection 3.5.1 are presented in perforate the structure or barrier, an equi- ,
1 this section. The following procedures are in i s accordance with Section 3.5.3 of NUREG 0800 l
(Standard Review Plan).
Amendment 6 3 s-7 ,
ABWR -me Standard Plant REV,B valent static load concentrated at the impact 3. A. Amirikan, Design of Protective Strue.
area is determined. The structural response to tures, Bureau of Yards and Docks, Publica-this load, in conjunction with other appropriate tion No. NAVDOCKS P-51, Department of the design loads, is evaluated using an analysis Navy, Washington, D.C., August 1960.
procedure similar to that in Reference 6 for rigid missiles, and the procedure in Reference 7 4. A. E. Stephenson, Full-Scale Tornado Mis-for deformable missiles. site Irnpact Tests, EPRI NP-440, July 1977, prepared for Electric Power Research 3.5.4 Interfaces Institute by Sandia Laboratories.
3.5.4.1 Protection of Ultimate Heat Sink 5. W. B. Cottrell and A. W. Savolainen, U. S.
S Reactor Containment Technology, ORNL.
4 Compliance with Regulatory Guide 1.27 as NSIC-5, Vol.1, chapter 6, Oak Ridge Na-f related to the ultimate heat sink and connecting tional Laboratory.
g conduits being capable of withstanding the effects of externally generated missiles shall be 6. R. A. Williamson and R. R. Alvy, Impact demonstrated (See Subsection 3.5.2). Effect of Fragments Striking Structural Elements Holmes and Narver, Inc., Revised 3.5.4.2 Missiles Generated by Natural Phenomena November 1973.
from Remainder of Plant Structures, Systems and Components 7. J. D. Riera, On the Stress Analysis of Structures Subjected to Aircraft impact The remainder of plant structures, systems, Forces, Nuclear Engineering and Design, and components shall be analytically checked to North Holland Publishing Co., Vol. 8,1968.
ensure that during a site-specific tornado they will not generate missiles exceeding the missiles 8. American Natoinal Standard For Estimating considered under Subsection 3.5.1.4. Tornado and Other Extreme Wind Characteris-tics at Nuclear Power Sites, ANSI /ANS - 2.3. L 3.5.4.3 Site Proximity Missiles and Aircraft l Hazards. 9. River Bend Station Updated Safety Analysis i Report, Docket No. 50 458, Volume 6, pgs. 5 Analyses shall be provided that demonstrate 3.5-4 a n d 3.5-5, A u g u s t 1987.
that the probability of site proximity missiles including aircraft) impacting the ABWR Standard l (Plant and causing consequences greater than 10CFR Part 100 exposure guidelines is .5,10-7 per year (See Subsection 3.5.1.6) 3.5.4.4 Secondary Missiles inside Containment
[
Protection against the secondary missiles inside containment described in Subsection 3.5.1.2.3 shall be demonstrated.
3.5.5 References
- 1. C. V. Moore, The Design of Barricades for Ha:ardous Pressure Systems, Nuclear Engineering and Design, Vol. 5,1967.
- 2. F. J. Moody, Prediction of Blowdown Thrust and let Forces, ASME Publication 69-HT 31, August 1969.
Amendment (, 3.5-8
_ __. l
' MM 23A6100AE hadard Plant nrv s rm 34.2.2 Analytic Methods to Define Blowdown i Forting Functions and Response Models.
34.2.2.1 Analytic Methods to Define Blowdown Forcing Fonctions.
The rupture of a pressurized pipe causes the x flow characteristics of the system to change creating reaction forces which can dynamically excite the piping system. The reaction forces are a function of time and space and depend upon fluid state within the pipe prior to rupture, break flow area, frictional losses, plant system characteristics, piping system, and other factors. The methods used to calculate the reaction forces for various piping systems are presented in the following subsections.
The criteria that are used for calculation of fluid blowdowa forcing functions include:
(1) Circumferential breaks are assumed to result in pipe severance and separation amounting to at least a one diameter lateral displacement of the ruptured piping sections 7
unless physically limited by piping si fb restraints, structural members, or piping stiffness as may be demonstrated by inciastic limit analysis (e.g., a plastic hinge in the piping is not developed under loading).
(2) The dynamic force of the jet discharge at the break location is based.on the cross-sectional flow area of the pipe and on a calculated fluid pressure as modified by analytically. or experimentally-determined thrust coefficient. Line restriction's, flow limiters, positive pump-controlled flow, and the absence of energy reservoirs are taken into accounts, as applicable, in the
- reduction of jet discharge.
(3) all breaks are assumed to attain full size within one millisecond af ter break initiation.
, Blowdown forcing functions are determined by the method specified in Appendix B of
{2M A N S I / A N S.5 8.2 s
o Amendment 6 3613 )
l l l 1 1 l l C_ a
23A6t00AE Standard Plant REv s (3) The analytical model adequately represents the mass / inertia and stiffness properties of the system.
(4) Pipe whipping is assumed to occur in the plane defined by the piping geometry and configuration and to cause pipe movement in the direction of the jet reaction.
(5) Piping within the broken loop is no longer considered part of the RCPB. Plastic
, deformation in the pipe is considered as a E
potential energy absorber. Limits of strain are imposed which are similar to strain levels allowed in restraint plastic members. Piping systems are designed so that plastic instability does not occur in the pipe at the design dynamic and static loads unless damage studies are performed which show the consequences do not result in direct damage to any essential system or component.
3.6.2.2.2 Pipe Whip Dynamic Response (6) Components such as vessel safe ends and val-Analyses ves which are attached to the broken piping system, do not serve a safety-related func.
The prediction of time-dependent and steady- tion, or failure of which would not further thrust reaction loads caused by blowdown of sub- escalate the consequences of the accident cooled, saturated, and two-phase fluid from rup- are not designed to meet ASME Code-imposed tured pipe is used in design and evaluation of limits for essential components under fault-dynamic effects of pipe breaks. A discussion of ed loading. However, if these components the analytical methods employed to compute these are required for safe shutdown or serve a to blowdown loads is given in Subsection 3.6.2.2.1. protect the structural integrity of an es-Following is a discussion of analytical methods sential component, limits to meet the Code used to account for this loading- requirements for faulted conditions and li-mits to ensure required operability will be The criteria used for performing the pipe whip met.
dynamic response analyses include:
The pipe whip analysis is performed using the (1) A pipe whip analysis is performed for each PDA computer program, a program as described in postulated pipe break. However, a given Appendix 3D, which predicts the response of a
. analysis can be used for more than one post- pipe subjected to the thrust force occurring ulated break location if the blowdown forc- after a pipe break. The program treats the ing function, piping and restraint system situation in terms of generic pipe break con-geometry, and piping and restraint system figuration which involves a straight, uniform properties are conservative for other break pipe fixed at one end and subjected to a time-locations. dependent thrust force at the other end. A typical restraint used to reduce the resulting (2) The analysis includes the dynamic response deformation is also included at a location of the pipe in question and the pipe whip be t we e n ilh e t wo e nds. Nonlinear and restraints which transmit loading to the time independent stress strain relationships support structures. are used to model the pipe and the restraint.
Using a plastic hinge concept, bending of the pipe is assumed to occur only at Amendmem 6 3.6-14
r MM 23A6100AE )
mandard Plant - REV.B l 1' conservative,1/2 of minimum percent 2. Recommended stress limits per ASME V elongation as specified in the Code Section III, Subsection NF for applicable ASME Code Section III or faulted conditions if applicable ASTM Specifications.
(3) Type III restraints (fasteners)
- 2. Design steady state load -' The maximum restraint load is limited to (a) Materials 80% of the minimum calculated static Fastener material conforms to ASTM, ultimate restraint strength at the ' ASME, or MIL requirements.
drywell' design temperature. This strain is less than 50% of the (b) Inspection ultimate uniform strain for all All fasteners are inspected or certified materials used for Type I components. per applicable ASTM, ASME, or MIL specifications.
(2) Type .Il restraints (e.g., clevio es, brackets, pins) (c) Design Limits Design limits are the same as for Type (a) Materials 11 restraints.
Materials selection conforms to: (4) Type III restraints (welds)
- 1. ASTM Specifications including (a) Materials consideration for brittle fracture control, or Weld materials for attachments to carbon
~
steel-structures are limited to low
/ 's 2. ASME Code Section III, Subsection NF, bydrogen type.
V if applicable.
(b) Inspection (b) Inspection Liquid penetrant surface inspection are Inspection conforms to: performed per ASTM Specification E165 or AWS Structural Welding Codes, AWS D1.1.
- 1. ASME/ ASTM requirements or process qualification and finished part (c) Design Limits surface inspection per ASTM methods, or The maximum primary weld stress intensity (two times maximum shear
- 2. ASME Code Section III, Subsection NF, stress) is limited to three times AWS or ;
'if applicable. AISC building allowable weld shear l stress.
(c) Design Limits (d) Procedures Design limits are based on the following stress limits: Procedures and welders will be qualified in accordance with the AWS Code for
- 1. Primary stresses (in accordance with welding in building structures.
definitions in ASME Code Section III) are limited to the higher of 70% of (5) Type IV restraints (structural and ciCl Su where Su = minimum ultimate components strength by tests or ASTM e specification or Sy + 1/3 (Su Sy) Material, inspection, and design where Sy = minimum yield strength by requirements for the structural and civil test or ASTM specification, or components are provided by industry Amendmen 6 3.6-21
_ L _ -_--__
ABM 23A6100AE Standard Plant REV.B standards such as AISC, ACl, and ASME Code 3.6.2.4 Guard Pipe Assembly Design
! Section III, Division II, along with appropriate requirements imposed for similar The ABWR single primary containment does not loading events. These components are also require guard pipes to avoid steam bypass of the designed for other operational and accident suppression pool in case of a pipe break or loadings, seismic loadings, wind loadings, crack at the containment penetration. All fluid and tornado loadings. system piping that may require postulation of a pipe break or crack (See introduction to Section The design basis approach of categorizing 3.6 and subsection 3.6.2.1.6) in the containment components is consistent in allowing less penetration areas meet the break exclusion stringent inspection requirements for those requirements of Subsection 3.6.2.1.4.2 or components subject to lower stresses. 3.6.2.1.5.3.1, as applicable.
Considerable strength margins exist in Type 11 through IV components up to the limit of load 3.6.2.5 Material to be Supplied for the capacity (fracture) of a Type I component. Operating License Review Impact properties in all components are considered since brittle type failures could See subsection 3.6.4.1 reduce the restraint system effectiveness.
3.6.3 Leak Before-Break In addition to the design considerations, Evaluation Procedures strain rate effects and other material property variations have been considered in the design of Per Regulatory Guide 1.70, Revision 3, the the pipe whip restraints. The material safety analysis Section 3.6 has traditionally properties utilized in the design have included addressed the protection measures against one or more of the following methods: dyar.mic effects associated with the non-mechanistic or postulated ruptures of piping.
(1) Code minimum or specification yield and The dynamic effects are defined in introduction ultimate strength values for the affected to Section 3.6. Three forms of piping failure components and structures are used for both (full flow area circumferential and longitudinal the dynamic and steady-state events; breaks, and throughwall leakage crack) are postulated in accordance with Subsection 3.6.2 (2) Not more than a 10% increase in minimum code and Branch Technical Position MEB 3-1 of NUREG -
or specification strength values is used 0800 (Standard Review Plan).
when designing components or structures for the dynamic event, and code minimum or However, in accordance with the revised specifica'. ion yield and ultimate strength General Design Criterion 4 (GDC 4) the values are used for the steady-state loads: mechanistic leak-before-break approach (LBB),
justified by appropriate fracture mechanics
-(3) Representative or actual test data values techniques, is now (Reference 1) an acceptable are used in the design of components and procedure to exclude design against the dynamic structures including justifiably elevated effects from the postulation of breaks in high-strain rate-affected stress limits in excess energy piping. Described in this subsection are l of 10%; or the criteria and procedures for the LBB approach which are utilized to qualify piping for (4) Representative or actual test data are used exclusion from postulation of breaks. This for any affected component (s) and the Subsection is based on proposed (Reference 4) minimum code or specification values are Section 3.6.3 of NUREG - 08M.
used for the structures for the dynamic and the steady state events The LBB approach is not used to exclude postulation of cracks and associated effects in ;
O\\
Amendment 6 3.6 22
23A6100AE Standard Plant REV.B
.A U accordance with Subsections 3.6.2.1.5 and portion thereof) is evaluated with the following 3.6.2.1.6.2.' considerations in addition to the deterministic LBB evaluation procedure of Subsection 3_.6.3.2 The LBB approach is not applicable to piping systems where operating experience has indicated (1) Degradation by erosion, crosion/ corrosion particular susceptibility to failure from the and erosion / cavitation due to unfavorable effects of intergranular stress corrosion flow conditions and water chemistry is cracking (IGSCC), water hammer, thermal fatigues, examined. The evaluation is based on the or erosion.' industry expnience and guidelines. Addi-tionally, fabrication wall thinning of el-The LBB approach is not a replacement for bows arnt other fittings is considered in the existing regulations or criteria pertaining to purchase specification to assure that the the design bases of emergency core cooling system code minimum wall requirements are met.
(Subsection 6.3), containment system (Subsection These evaluations demonstrate that these me-6.2) or equipment qualification (Subsection chanisms are not potential sources of pipe 3.11). However, benefits of the LEB procedures rupture to these areas will be taken and the subsections will be revised as the regulations will be (2) The ABWR plant design involves operatbn relaxed by the o 1C. For clarity, it is noted below 7000F in ferritic steel piping and that the LBB app.aach is not used to relax the below 8000 F in austenitic steel piping.
design requirements of the primary containment This assures that creep and creep-fatigue system that includes the primary containment are not potential sources of pipe rupture, vessel (PCV), vent systems (vertical flow channels and horizontal vent discharges), drywell (3) The design also assures that the piping q zones, suppression chamber (wetwell), vacuum material is not susceptible to brittle
't V
r breakers, PCV penetrations, and drywell head, cleavage-type failure over the full range of However, in designing for loads per Table 3.9-2, system operating temperatures (that is, the which does not apply to these PCV subsys- tems, material is on the upper shelf).
the seven types of design loads identified with LOCA induced dynamics of suppression pool or (4) The ABWR plant design specifies use of shield wall annulus pressurization are excluded austenitic stainless steel piping made of if they are a result of LOCA postulated in those material (e.g., nuclear grade or low carbon piping that meet the LBB criteria.' type) that is recognized as resistant to IGSCC. The material of piping in reactor Appendix 3E characterizes fracture mechanics coolant pressure boundary is ferritic steel.
properties of piping rnaterials and analysis me-thods including leakage calculation methods, as (5) A systems evaluation of potential water l required by the criteria of this subsection, hammer is made to assure that pipe rupture Following NRC's review and approval, this appen- due to this mechanism is unlikely. Water hammer is a generic term including various l dix willtobecome lication approved ABWR Standard LBB methodology Plant piping. Appendix for app- high frequency hydrodynamic unanticipated 3F applies these properties and methods to events such as steam hammer and water specific piping to demonstrate their eligibi. slugging. To demonstrate that water hammer lity for exclusion under the LBB approach. See is not a significant contributor to pipe Subsection 3.6.4.2 for interface requirements. rupture, reliance on historical frequency of water hammer events in specific piping systems coupled with a review of operating procedures and conditions is used for this evaluation. The ABWR design includes features such as vacuum breakers and jockey
/]_
(/
3.6.3.1 General Evaluation pumps coupled with improved operational procedures to reduce or climinate the j
i The high energy piping system (or analyzable potential or water hammer identified by past l e Amendment 6 3.6-23 L_--- . _ . .
23A6100AE Standard Plant REV.A experience. Certain anticipated water (1) Use the fracture mechanics and the leak hammer events, such as a closure of a valve, rate computational methods that are accept-are accounted for in the Code design and ed by the NRC staff, or are demonstrated analysis of the piping, accurate with respect to other acceptable computational procedures or with (6) The systems evaluation also addresses a po- experimental data, tential for fatigue cracking or failure from thermal and mechanical induced fatigue. (2) Identify the types of materials and ma-Based on past experience, the piping design terials specifications used for base metal, avoids potential for significant mixing of weldments and safe ends, and provide the high and low- temperature fluids or materials properties including toughness mechanical vibration. The startup and and tensile data, long-term effects such as preoperational monitoring assures avoidance thermal aging, and other limitations.
of detrimental mechanical vibration.
(3) Specify the type and magnitude of the loads (7) Based on experience and studies by Lawrence applied (forces, bending and torsional Livermore . Laboratory, potential indirect moments), their source (s) and method of sources ofindirect pipe rupture are remote combination. For each pipe size in the causes of pipe rupture. Compliance with the functional system, identify the location (s) snubber surveillance requirements of the which have the least favorable combination technical specifications assures that of stress and material properties for base snubber failure rates are acceptably low. metal, weldments and safe ends.
(8) Initial LBB evaluation is based oc the (4) Postulate a throughwall flaw at the design configuration and stress levels that location (s) specified in (3) above. The are acceptably higher than those identified size of the flaw should be large enough so by the initial analysis. This evaluation is that the leakage is assured detection with reconciled when the as-built configuration sufficient margin using the installed leak is documented and the Code stress evaluation detection capability when the pipes are is reconciled. It is assured that the subjected to normal operating loads. If as built configuration does not deviate auxiliary leak detection systems are relied significantly from the design : configuration on, they should be described. For the to invalidate the initial LBB evaluation, or estimation of leakage, the normal operating a new evaluation coupled with necessary loads (i.e., deadweight, thermal expansion, configuration modifications is made to and pressure) are to be combined based on j assure applicability of the LBB procedure, the algebraic sum of individual values, j i
(9) Sufficiently reliable, redundant, dherse Using fracture mechanics stability analysis and sensitive leak detection sprems are or limit load analysis based on (11) below, provided for monitoring of !cak. The system and normal plus SSE loads, determine the that is relied upon to pedict the through- critical crack size for the postulated wall flaw used in the deterministic fracture throughwall crack. Determine crack size mechanics evaluation is sufficiently margin by comparing the selected leakage reliable and sensitive to justify a margin size crack to the critical crack size, of 2 oe. the leakage prediction. Demonstrate that there is a margin of 2 between the leakage and critical crack 3.6.32 Deterministic Evaluation Procedure sizes. The same load combination method selected in (5) below is used to determine The following deterministic analysis and the critical crack size.
evaluation are performed as an NRC-approved I I
method for the ABWR Standard Nuclear Island to (5) Determine margin in terms of applied loads justify applicability of the LBB concept. by a crack stability ant. lysis. Demonstrate i O
Amendment 1 3.6 24
I l
ABWR m6ms Standard Plant REV B q.
including deadweight and SSE (inertial) 3.6A.2 leak-Before Break Analysis Report components.
As required by Reference 1, an LBB analysis Shielded Metal Arc (SMAW) and Submerced Arc report shall be prepared for the piping systems (SAW) Welds proposed for exclusion from the analyses for the dynamic effects due to their failure. The The flow stress used to construct the master report shall include only the piping stress curve is 51 ksi analysis results for the piping systems analyzed and reported for LBB in Appendix 3F in order to The value of SI used to enter the master show that the piping stresses are within the curve for SMAW and SAW is stress levels assumed in Appendix 3F (See Subsec-tion 3.6.3).
Si = M (Pm+Pb+P)Z e (8) 3.6.5 References where L Modification of General Design Criterion 4 P6 = the combined primary bending stress, Requirements for Protection Against Dynamic including deadweight and seismic Effects of Postulated Pipe Rupture, Federal components. Renister. Volume 52, No. 207, Rules and Regulations, Pages 41288 to 41295, October Pc = combined expansion stress at normal 27,1987 operation.
- 2. RELAP 3, A Computer Program for Reactor Z = 1.15 l1.0 + 0.013 (OD-4)] for SMAW, (9) Blowdown Analysis, IN 1321, issued June 1970, Bractor Technoloey TID-4500.
O Z = 1.30 [1.0 + 0.010 (OD-4)] for SAW, (10)
\.) 3. Moody, F. J., Fluid Reactor and impingement and Loads, Vol.1, ASCE Specialty Conference on Structural Design of Nuclear Plant OD = pipe outer diameter in inches. Facilities pp. 219-262, December 1973.
When the allowable flaw length is determined 4. Standard Review Plan; Public Comments from the master curve at the appropriate SI Solicited, Ecderal Recister. Volume 52, No.
value, it can be used to determine if the 167, Notices, Pages 32626 to 32633, August required margins on load and flaw size are 28,1987.
met using the following procedure.
For the method of load combination described in item (5), let M = 1.4, and if the allowable flaw length from the master curve is at least equal to the leakage size flaw, then the margin on load is met.
3.6.4 Interfaces 3.6A.1 Summary of Pipe Break Analysis Results A summary of the dynamic analyses applicable to high.cnergy piping systems shall be prepared in accordance with Subsection 3.6.2.5 of p Regulatory Guide 1.70 (See Subsection 3.6.4.1).
Amendment 6 3.6-27
23A6100AE mandard Plant arv n C 3.7.3.8.2.2 Effect of Differential Building Movements adequately accounted for in the analysis.
la case of buried systems sufficiently flex.
ible relative to the surrounding or under.
The relative displacement between anchors is lying soil, it is assumed that the systems determined from the dynamic analysis of the will follow essentially the displacements structures. The results of the relative anchor- and deformations that the soil would have if point displacement are used in a static analysis the systems were absent. When applicable, to determine the additional stresses due to procedures, which take into account the phe-relative anchor-point displacements. Further nomena of wave travel and wave reflection in details are given in Subsection 3.7.3.8.1.8. cornpacting soil displacements from the ground displacements, are employed.
3.7.3.9 Multiple Supported Equipment Components With Distinct Inputs (2) The effects of static resistance of the sur-rounding soil on piping deformations or The procedure and criteria for analysis are de- displacements, differential movements of scribed in Subsections 3.7.2.13 and 3.733.13. piping anchors, bent geometry and curvature changes, etc., are considered. When appli-3.7.3.10 Use or Constant Vertical Static cable, procedures utilizing the principles Factors of the theory of structures on clastic foun-dations are used.
All Seismic Category I subsystems and compo-nents are subjected to a vertical dynamic anal- (3) When applicable, the effects due to local ysis with the vertical floor spectra or time soil settlements, soil arching, etc., are histories defining the input. A static analysis also considered in the analysis.
n is performed in lieu of dynamic analysis if the
('} peak value of the applicable floor spectra times a factor of 1.5 is used in the analysis. A 3.7.3.13 Interaction of Other Piping with Seismic Category l Piping factor lower than 1.5 can be used if it is justi-fiable. If the fundamental frequency of a compo- In certain instances, non-Seismic Category I nent in the vertical direction is greater than or piping may be connected to Seismic Category 1 M equal to 33 Hz, it is treated as seismically piping at locations other than a piece of equip-
$ rigid and analyzed statically using the zero-pe- ment which, for purposes of analysis, could be riod acceleration of the applicable floor re- represented as an anchor. The transition points sponse spectrum, typically occur at Seismic Category I valves which may or may not be physically anchored.
3.73.11 Torsional Effects of Eccentric Masses Since a dynamic analysis must be modeled from pipe anchor point to anchor point, two options j Torsional effects of eccentric masses are in- exist: j cluded for Seismic Category I subsystems similar j to that for the piping systems discussed in Sub- (1) specify and design a structural anchor at ]
section 3.7.3.3.1.2. the Seismic Category I valve and analyze the Seismic Category I subsystem; or, if imprac-3.73.12 Buried Seismic Category i Piping and tical to design an anchor, Tunnels (2) analyze the subsystem from the anchor point For buried Category I piping systems and in the Seismic Category I subsystem through
' tunnels the following items are considered in the the valve to either the first anchor point analysis: in the non. Seismic Category I subsystern; or to sufficient distance in the non Seismic (1) The inertial effects due to an earthquake Category I Subsystem so as not to sig-upon buried systems and tunnels will be nificantly degrade the accuracy of analysis d of the Seismic Category I piping.
Amendmen 6 3 7-23 l
ABM 23xa mre REV,A Standard Plant _
Where small, non-Seismic category piping is (6) recording and playback equipment; and directly attached to Seismic Category I piping, its effect on the Seismic Category I piping is (7) annunciators.
accounted for by lumping a portion of its mass with the Seismic Category I piping at the point Tht, location of seismic instrumentation is of attachment, outlined in Table 3.7-7.
Furthermore, non Seismic Category I piping 3.7.4.2.1 'Ilme-History Accelerographs (particularly high energy piping as defined in Section 3.6) is designed to withstand the SSE to Time-history accelerographs produce a record avoid jeopardizing adjacent Seismic Category I of the time varying acceleration at the sensor piping if it is not feasible or practical to location. This data is used directly for analy-isolate these two piping systems, sis and comparison with reference information and may be, by calculational methods, converted to 3.7.3.14 Seismk Analysis for Reactor response spectra form for spectra comparisons Internals with design parameters.
The modeling of ,RPV internals is discussed in Each triarial acceleration sensor unit con-Subsection 3.7.2.3.2. The damping values are tains three accelerometers mounted in an ortho-given in Table 3.71. The seismic model of the gonal array (two horizontal and one vertical).
RPV and internal is shown in Figure 3.7 32. All acceleration units have their principal axes oricated identically. The mounted units are 3.7.3.15 Analysis Procedures for Damping oriented so that their axes are aligned with the building major axes used in development of the Analysis procedures for damping are discussed mathematical models for seismic analysis.
in Subsection 3.7.2.15.
One THA is located on the reactor building 3.7.4 Seismic Instrumentation (RB) foundation mat, El (-) 13.2 M, at the base of an RB clean zone for the purpose of measuring 3,7.4.1 Comparison with NRC Regulatory Guide the input vibratory motion of the foundation 1.12 mat. A second THA is located in an RB clean zone at El (+) 26.7 M on the same azimuth as the The seismic instrumentation program is foundation mat "IHA. They provide date on the consistent with Regulatory Guide 1.12. frequency, amplitude, and phase relationship of the seismic response of the reactor building 3.7.4.2 Location and Description of structure. A third THA is located in the frec
- Instrumentation field at the finished grade approximately 160 M l from any station structures with axes oriented in I The following instrumentation and associated the same dire
- tior as the reactor building equipment are used to measure plant response to accelerometers.
carthquake motion:
Two seismic triggers, connected to form redun-(1) three triaxial time history accelerographs dant triggering, are provided to start the THA (THA); recording system. They are located in the free field at the finished grade 160 M from the reac-(2) three peak-recording accelerographs (PRA); tor building. The trigger unit consists of or-thogonally mounted acceleration sensors that act-(3) two triaxial seismic triggers; uate relays wheneven a threshold acceleration is exceeded for any of the three axes. The triggen (4) one seismic switch (SS); is engineered to discriminate against false starts from other operating inputs such as traf-(5) four response spectrum recorders; fic, elevators, people, and rotating equipment.
l Amendment 1 3.7-24
L ABWR 2mimAs Re- ' ed Plant Rev. A N
d SECTION 3.8 CONTENTS (Continued)
Section lllle East !
3.833.1 Load Definitions 3&l8 3.833.2 ' Load Combinations 3.8-18 3.83.4 Design and Analysis Procedures 3.8-18 3.83.4.1 Diaphragm Floor 3.8-18 3.83.4.2- Reactor Pedestal 3&I8 3.83.43 Reactor Shield Wall 3&l9 3.83.4.4 Drywell Equipment and Pipe Support Structure 3.8-19 3.83.4.5 Other Internal Structures 3.8-19 3.83.5 Structural Ac:eptance Criteria 3.8-19 3.83.5.1 Diaphragm Floor 3.8-19 3.83.5.2 Reactor Pedestal 3.8-19 3.83.53 Other Internal Structures 3.8-19 3.83.6 Materials, Quality Control, and Special Construction Techniques 3.8-19 3.83.6.1 Diaphragm Floor 3.8-19 3.83.6.2 Reactor Pedestal 3.8-19 3.83.63 Reactor Shield Wall 3.8-19 3.83.6.4 Drywell Equipment and Pipe Support Structure 3&l9 3.83.6.5 Other Internal Structures 3.8-20 3.83.7 Testing and Inservice Inspection Requirements 3.8-20 4 3.8.4 Other Seismic Calenorv I Structures 3.8-20
]
3.8-vi j Amendment 1
1 23A6100AE Standard Plant nev. n SECTION 3.8 CONTENTS (Continued)
Section Htle P. age 3.8.4.1 Description of the Structures 3.8-20 3.8.4.1.1 Reactor Building Structure 3.8-20 3.8.4.L2 Control Building 3.8-21 3.8.4.13 Radwaste Building Substructure 3.8-21 3.8.4.1.4 Seismic Category I Cable Tray and Conduit Supports 3.8-21 3.8.4.2 Applicable Codes, Standards, and Specifications 3.8-21a 3.8.4.2.1 Reactor Building 33-21a 3.8.4.2.2 Control Building 3.8 22 L'
3.8.4.23 Radwaste Building Substructure 3.8-23 3.8.4.2.4 Seismic Category I Cable Tray and Conduit Supports 3.8-23 3.8.43 Loads and load Combinations 3.8-23 3.8.43.1 Reactor Building 3.8-23 3.8.43.1.1 Loads and Notations 3.8-23 3.8.43.1.2 Load Combinations for Concrete Members 3.8-24 3.8.43.13 I.oad Combinations for Steel Members 3.8-25 3.8.43.2 Ccutrol Building and Radwaste Building Substructure 3.5-25 3.8.433 Seismic Category I Cable Tray and Conduit Supports 3.8-26 3.8.4.4 Design and Analysis Procedures 3.8-26 3.8.4.4.1 Reactor Building 3.8 26 I
a 3.8-vii Amendmen 6 l i
ABWR mawin Re- 'W Pknt nrv n ij s._,
SECTION 3.8 CONTENTS (Continued)
Section Iltle East 3.8.4.4.4 Seismic Category I Cable Tray and Conduit Supports 3.8-26 3.8.4.4.4.1 Cable Tray Supports 3.8-26 3.8.4.4.4.2 Conduit Supports 3.8-26 3.8.4.5 Structural Acceptance Criteria 3.8-26 3.8.4.5.1' Reactor Building 3.8 26 32.4.5.1.1 General Critcria 3.8-26 3.8.43.1.2 Materials Criteria 3.8-27 3.8.4.5.2 Control Building' 3.8-27 3.8.4.53 Radwaste Building Substructure 3.8 27 iQ Q 3.8.5 Foundations 3 3-27 3.8.5.1 Description of the Foundations 3.8-27 3.8.5.2 Applicable Codes, Standards and Specifications 3.8-27 3.8.53 Loads and Load Combinations 3.8-27 3.8.5.4 Design and Analysis Procedures 3.8 27a 3.8.5.5 Structural Acceptance Criteria 3.8-28 32.5.6 Materials, Quality Control, and Special Construction Techniques 3.8 28 33.5.7 Testing and Inservice Inspection Requirements 3.8-28 1'
3.8-viii Amendment 6 l L - - - - - - - - - --------------------U
ABM 23A61ooAs Standard Plant REV.A SECTION 3.8 TABLES ma rua em 3.8-1 Load Combinations, Load Factors, and Acceptance Criteria for the Reinforced Concrete Containment 3.8-29 3.8-2 Major Allowable Stresses in Concrete and Reinforcing Steel 3.8-30 3.63 Stress Intensity Limits 3.8-31 3.8-4 Codes, Standards, Specifications, and Regulations Used in the Design and Construction of Seismic Category I Internal Structures of the Containment 3.8-32 3.8-5 Load Cotabinations, Load Factors, and Acceptance Criteria for the Reinforced Concrete Structures inside the Containment 3.8-34 3.8-6 Load Combinations, load Factors, and Acceptance Criteria for Steel Structures Inside the Containment 3.8-35 3.8-7 Load Combinations for Foundation Design 3.8-36 ILLUSTRATIONS Figure TLtle eagt 3.8-1 RB - Floor Plan El. (-) 13200 3.8-37 3.8-2 RB Floor Plan El. (-) 6700 3.8-38 3.8-3 RB Floor Plan El. (-) 0200 3.8-39 3.8-4 RB - Ground Floor Plan El. 7300 3.8-40 3.8-5 RB - Floor Plan El.13100 3.8-41 l 3.8-6 RB - Floor Plan El.18500 3.8-42 3.8-7 RB - Fioor Plan El. 26700 3.8-43 3.8-8 RB - Roof Plan El. 33200 3.8-44 3.8-ix 0
Amendment 1
_ _. ._. _ _ - - . ______-_-___--___-_D
ABM 23463004e 9mndard Plant nrv. n
( .
t ' 32.1.6.5 Quality Control 3.8.2 STEEL COMPONENTS OFTHE REINFORCED CONCRETE Ouality control procedures are established in CONTAINMENT the Construction Specification and implemented during construction and inspection. The 3.8.2.1 Description of the Containment Construction Specification covers the
' fabrication, furnishing, and installation of each The ABWR has a reinforced concrete structural item and specifies the inspection and containment vessel (RCCV) as described in documentation requirements to ensure that the Subsection 3.8.1. This section will describe l requirements of the ASME Code, Section 111, the following steel components of the concrete Division 2, and the applicable Regulatory Guides containment vessel:
are met.
(1) Personnel Air Locks 3J.1.7 Testing and Inservice Inspection Requirements (2) Equipment Hatches, 3.8.1.7.1 Structural Integrity Pressure Test (3) Penetrations A structural integrity test of the (4) Drywell Head containment structure is performed in accordance with Article CC-6000 of the ASME Code, Section 3.8.2.1.1. Description of Penetrations III, Division 2 and Regulatory Guide 1.136, after completion of the containment construction. The The penetrations through the RCCV include the test is conducted at 115% of the design pressure following.
condition of 45 psig in both the drywell and hw/
suppression chamber, simultaneously. A pressure test for the design differential pressure 3.8.2.1.1.1 Personnel Air Locks condition of 25 psig between the drywell and the Two personnel air locks with an inside suppression chamber is also performed where the diameter sufficient to provide 6 ft., 8 in.,
drywell pressure is greater than the suppression high by 3 ft.,6 in., wide minimum clearance chamber pressure, above the floor at the door way are provided.
One of these air locks provides access to the During these tests the suppression chamber ind upper dry well ar.d the other provides access to spent fuel pool are filled with water to the the lower drywell via the access tunnel.
normal operational water level. Deflection and concrete crack measurements are made to determine Lock and swing of the doors is by manual and that the actual structural response is within the automatic means. The locks extend radically limits predicted by the design analysis, outward from the RCCV into the reactor building and are supported by the RCCV only. The minimum In addition to the deflection and crack clear horizontal distance not impaired by the measurements, the first prototype containment door swing is 6 ft.
structure is instrumented for the measurement of strains in accordance with the provisions of Each personnel air lock has two pressure-Subsubarticle CC-6230 of the ASME Code, Section seated doors interlocked to prevent simultaneous ill, Division 2. opening of both doors and to ensure that one door is completely closed before the opposite 32.1.7.2 Preoperational and Inservice door can be opened. The design is such that the lategrated teak Rate Test interlocking is not defeated by postulated malfunctions of the electrical system. Signals l Preoperational and inservice integrated leak and controls that indicate the operational l rate testing is discussed in Subsection 6.2.1. status of the doors are provided. Provision is t
Amendment 6 3.811
23A6100AE Standard Plant Rev.n made to permit temporary bypassing of the door 3.8.2.1.1.3 Other Penetrations interlock system during plant cold shutdown. The door operation is designed and constructed so The RCCi! penetrations are categorized into either door may be operated from inside the two basic types. These types differ with containment vessel, inside the lock, or from respect to whether the penetration is subjected outside the containment vessel, to a hot or cold operational environment.
The lock is equipped with a digital readout The cold penetrations pass through the RCCV pressure transducer system to read inside and wall and are embedded directly in it. The hot outside pressures. Quick-acting valves are penetrations do not come in direct contact with provided to equalize the pressure in the air lock the RCCV wall but are provided with a thermal when personnel enter or leave the containment sleeve which is attached to the RCCV wall. The vessel. Each door has two separate inflatable thermal sleeve is attached to the process pipe seals with provision for independent pressure and at distance from the RCCV wall to minimize leakage testing. The air supply to the seals is conductive heat transfer to the RCCV wall.
designed so that each redundant seal on a given door is connected to an independent air supply. Besides piping penetrations, several electri-Loss of air supply does not jeopardize the cal penetrations also exist. A description of integrity of the lock. the various penetrations is given Chapter 8.
3.8.2.1.1.2 Equipment if atch 3.8.2.1.1.4 Drywell liead Three equipment hatches are provided. The two A 33 feet 9 inches in diameter opening in the drywell equipment hatches are 9 feet 10 inches in RCCV or upper drywell top slab over the RPV is diameter. One of these serves the upper drywell covered with a removable steel ellipsoidal and the other serves the lower drywell via the drywell head which is part of the pressure access tunnel. The third equipment hatch is 6 boundary. The drywell head is designed for feet 6 inches in diameter and provides personnel removal during reactor refueling and for and equipment access to the suppression chamber replacement prior to reactor operation using the airspace. reactor building crane. One pair of mating flanges is anchored in the drywell top slab and Tht, equipment hatch covers have a double O- the other is welded integrally with the drywell ring-scaled flange with provisions to pressure head. Provisions are made for testing the test the space between the seals of the flange. flange seals v.ithout pressurizing the drywell.
A means for removing and handling the equipment Figure 3.8-16 shows the drywell head.
hatch cover is provided. The hoisting equipment and hoisting guides are arranged to minimize con- The drywell head is attached to the RCCV with tact between the doors and seals during opening a " finger pin
- type closure as shown in Figure and closing. The equipment hatch includes the 3.8-16.
electric motorized hoist with pushbutton control stations, lifting slings, hoist supports, hoist-ing guides, access platforms, and ladders for ac-cess to the dogged position of the door and l
hoist, latches, seats, dogging devices, and tools l required for operation and maintenance of the hatch.
The equipment hatches and covers are entirely supported by the RCCV. Figure 3.815 show gen-eral details of the equipment hatch and cover.
O 3.8-12 Amendment 4
l I
ABM 23461ooxc mem=Aard Plant REV. 6 O'
,d (7) bar and machine steel (A576, carbon content not less than 0.3%); and All radiographic inspection is made in accordance with the requirements of Article NE 5000 of ASME Code Section III except as )
(8) clad (SA-240 typc 304L). specified herein. All shops welds requiring l l
radiography are radiographer in the shop after )
! The structural steel materials located beyond intermediate or final post weld heat treatment (PWHT). Radiographic examination of field l the containment vessel boundaries are as follows: j weldments is performed after any required final I (1) carbon steel (A36 or SA-36) and PWHT. Welds of ASME Code, categories C and D, which cannot be examined by radiography are (2) stainless steel extruded shapes (SA-479). examined in accordance with paragraphs NE-5232 and NE 5242 of ASME Code Section III.
The materials meet requirements as specified !
in Subarticle NE-2000 of ASME Code Section III. 3J.2.7.1.3 Ultrasonic Examination of Plates !'
The lowest service metal temperature is 300F, The ultrasonic examination of materials in the 3A.2.7 Testing and Inservice Inspection steel components is performed in accordance with Repletaments Subarticle NB-2530 of ASME Code Section III.
Leakage of the containment ressel, including 32.2.7.1.4 Magnetic Particle Examination the steel components is described in Subsection 3.8.1.7. Magnetic particle examination is performed on carbon steel welds in accordance with Articles 33.2.7.1 Examination Requirements NE-5000 and CC 5000 of ASME Code Section III.
The magnetic particle acceptance standards are in m The examination and testing of all materials accordance with NE-5340 and CC-5540 0f ASME Code I ) including welding materials are performed in SectionIII.
accordance with the requirements of Articles NE-2000 and NE-5000 of ASME Code Section III. 3.8.2.7.1.5 Liquid Penetrant Examination Nondestructive examinations are performed only by technicians qualified in accordance with Article Liquid penetrant examination is performed on NE 5000 of the ASME Code. all stainless steel welds in accordance with NE-5000 and CC 5000 of the ASME Code. All 32.2.7.1.1 Impact Testing discontinuities exceeding those allowed are removed and the weld repaired, as necessary, in Charpy V notch impact tests are made on ma- accordance with applicable procedures, terial, weld deposit, and the base metal weld heat affected zone employing a test temperature 32.2.7.2 Shop Testing Requirements (
not higher than 00F. The requirements of Sub- {
article NE-2320 of ASME Code Section III are met The shop tests of the personnel air locks in- l for all materials under jurisdiction of the clude operational testing and an overpressure I code. Impact tests of weld deposit and base me- test. After completion of the personnel air tal weld heat-affected zone are made for each locks tests (including all latching mechanisms welding procedure requiring ASME Code Section IX and interlocks) each lock is given an operational qualifications. test consisting of repeated operuing of each door and mechanism to determine whether all parts i 3A.2.7.1.2 Radiography are operating smoothly without binding or ether defects. All defects encountered are corrected All radiography is performed in accordance and retested. The process of testing, correcting with the requirements of Article NE-5000 of ASME defects, and retesting are continued until no Code Section III. defects are detectable.
Amendment 1 3.8-t5
ABM 23A6100AE Standard Plant arv s For the operational test, the personnel air locks are pressurized with air to the maximum Figures 3.8-17 and 3.818 and Figures 1.2-2 '
permissible code test pressure. All welds and through 1,2-13 show an overview of the seals are observed for visual signs of distress containment including the internal structures.
or noticeable leakage. The lock pressure is then reduced to design pressure and a thick bubble 3.83.1.1 Diaphragm Floor solution is applied to all welds and seals and observed for bubbles or dry flaking as indica- The diaphragm floor serves as a barrier tions of leaks. All leaks and questionable areas between the drywell and the suppression are clearly marked for identification and chamber,. It is a reinforced concrete circular subsequent repair. slab, with an outside diameter of 14.5 m (47 ft, 7 in), and a thickness of 1.2 m (3 ft,11 in).
During the overpressure testing, the inner dou are blocked with holddown devices to prevent The diaphragm floor is supported by the reac-unseating of the seals. The internal pressure of tor pedestal and the containment wall. The con-the lock is reduced to atmospheric pressure and nection of the diaphragm floor to the contain-all leaks are repaired. Afterward, the lock is ment wall is a fixed support. The diaphragm again pressurized to the design pressure with air floor connection to the reactor pedestal is a and all areas suspected or known to have leaked hinged npport. The diaphragm floor is pene-during the previous test are retested by the trated by 18,508 mm (20 in) diameter sleeves for bubble technique. This procedure is repeated the SRV lines. Reinforcing steel and end until no leaks are discernible. support details are provided in Figure 3H3-6 andl Table 3H3-1.
3.8.3 Concrete and SteelInternal Structures of the Concrete A 1/4 inch thick, carbon steel liner plate is Containment provided on the bottom of the diaphragm floor, and is anchored to it. The liner plate serves as 3.83.1 Description of the Internal Structures a form during construction and prevents the bypass flow of steam from the upper drywell to The functions of the containment internal the suppression chamber air space during a LOCA.
structures include: support of the reactor vessel radiation shielding, support of piping and 3.83.1.2 Reactor Pedestal equipment, and formation of the pressure suppression boundary. The containment internal A composite steel and concrete pedestal pro-structures are constructed of reinforced concrete vides support for the reactor pressure vessel, and structural steel. The containment internal the reactor shield wall, the diaphragm floor, structures include the following: access tunnels, horizontal vents, and the lower drywell access platforms. The pedestal consists (a) Diaphragm floor of two concentric steel shells tied together by vertical steel diaphragms. The regions formed by (b) Reactor pedestal the steel shells and the vertical diaphragms, except the vents and the vent channels, are (c) Reactor shield wall filled with concrete. There are ten drywell connecting vent (DCV) channels connecting the (d) Drywell and equipment pipe support structure upper drywell to the lower drywell and the horizontal vents.
(c) Miscellaneous platforms The wetted portion of the exterior surface of (f) Lower drywell equipment tunnel the reactor pedestal stcel shell in the suppression chamber is clad with stainless steel (g) Lower drywell personnel tunnel to provide corrosion protection. The extent of the cladding and the reactor pedrstal (h) Reactor shield wall stabilizer configuration is provided in Figure 1.2-2.
Amendment 6 3&l6
__ _ _ _ . . _ . _ s
23A6100AE Standard Plant Rev n p
(a! 3.8.3.13 Reactor Shield Wall 3.8.3.13.2 IAwer Drywell Equipment Tunnel The reactor shield wall is supported by the A steel tunnel is provided at azimuth 180 reactor pedestal and surrounds the reactor degrees for equipment access to the lower dry-pressure vessel. Its ' unction is to attenuate well from the reactor building. The tunnel has radiation emanating from the reactor vessel. In an inside diameter of 4.3 m (14 ft,1 in), is 20 addition, the reactor shield wall provides mm (3/4 in) in thickness, and has a flanged structural support for the reactor vessel closure at the reactor building end. The wetted stabilizer, the reactor vessel insulation and the portion of the tunnel is stainless steel cr drywell equipment and pipe support structure. carbon steel with stainless steel cladding. The Openings are provided in the shield wall to tunnel is attached rigidly to the containment permit the routing of necessary piping to the RPV wall at one end and the reactor pedestal at the and to permit in-service inspection of the RPV other end and is partially submerged in the and piping. suppression pool at normal water level. The tunnel has one or two flexible rings to The shield wall is shaped as a right accommodate differential displacement of the cylinder. The shield wall consists of two containment wall and reactor pedestal. The concentric steel cylindrical shells joined configuration of the tunnel and the connection together by horizontal and vertical steel plate details at the containment wall and reactor diaphragms. Full depth stiffeners are provided pedestal are shown in Figure 1.2-2. Fine motion in the reactor shield wall at the attachment control rod drive piping is routed through the locations of major pipe supports, pipewhip tunnel. The tunnel permits entry from the restraints and beam supports. The annular region reactor building into the lower drywell without between the outer and inner shells is filled with exposure to the suppression chamber atmosphere.
concrete. The arrangement of the reactor shield
[]
LJ wall is provided in Figure 1.2-3. 3.83.1.53 Lower Drywell Personnel Tunnel 3.83.1.4 Drywell Equipment and Pipe Support The lower drywell personnel tunnel is located Structure at azimuth 0 and is similar to the lower drywell equipment tunnel described in Subsection The drywell equipment and pipe support 3.8.3.1.5.2. However, it has a personnel lock structure (DEPSS) consists of various structural at the reactor building end. The arrangement components such as beams and columns. Built up and details of the tunnel are shown in Figure box shapes are used for beams and columns that 1.2-2.
must resist torsion and biaxial bending. The beams span between the reactor shield wall and the vertical support columns which are anchored to the diaphragm floor. The DEPSS provides support for piping, pipe whip restraints, mechanical equipment, electrical equipment and general access platforms and stairs.
3.83.1.5 Other Internal Structures 3.83.1.5.1 Miscellaneous Platforms 3.83.2 Applicable Codes, Standards, and Miscellaneous platforms are designed to allow Specifications access and to provide support for equipment and l q piping. The platforms consists of steel beams The design of the concrete and steel Q and grating.
Amendroess 6 3&l7 I
l
AWM 23461oore Standard Plant arv.A l
f internal structures of the containment conform to (2) Construction Loads--Construction loads are j the applicable codes, standards, and specifica- loads which are applied to the containment
., tions and regulations listed in Table 3.8-4 internal structures from start to completion l except where specifically stated otherwise. of construction. The definitions for D, L l and T are applicable, but are based on j Structure or Specific Reference actuai construction methods and /or comoonent Number conditions.
- q. Diaphragm Floor 14 (3) RV2-leads from component response or direct fluid forces, on components located in the Reactor Pedestal 1 13, 15-22 suppression pool, caused by safety relief valve air cleaning loads. ,
l Reactor Shield Wall 1 13,15-22 (4) RBV--Loads due to reactor building vibra.
DEPSS 15-22 tions caused by an SRV and LOCA event. l Miscellaneous platforms 15 22 (5) AP--Loads and pressures directly on the l reactor shield wall and loads from component l L/D Equipment Tunnel 15-22 response or direct steam flow forces on !
components located in the reactor vessel I L/D PersonnelTunnel 15-22 shield wall annulus region, caused by a
{
rupture of a pipe within the reactor vessel j Reactor Shield Wall 15-22 shield wall annulus region. q Stabilizer -
(6) SL--loads from component response or direct 3333 lands and Load Combinations fluid forces, on components located in the sloshing zone of a pool or component, caused 3333.1 Load Definitions by the sloshing phenomenon from any dynamic event.
The loads and applicable load combinations for which the structure is designed depend on the 3.833.2 Load Combination conditions to which the paticular structure is subjected. The load combinations and associated accep-tance criteria for concrete and steel internal The containment internal structures are de- structures of the containment are listed in Ta-signed in accordance with the loads described in ble 3.8.3-5 and Table 3.8.3-6, respectively; for )
Appendix 3B. These loads and the effects of the reactor shield wall refer to appendix 3B. a these loads are considered in the design of all I internal structures as applicable. The loads 333.4 Design and Analysis Procedures within the loading combinations are combined using the absolute sum technique. (Those loads 3.83.4.1 Diaphragm Floor which are defined as reversible in algebraic sign are combined in such a way as to produce the The design and analysis procedures used for maximum resultant stresses in the structure. All the diaphragm floor are similar to those used other loads are combined in accordance with their for the containment structure. The diaphragm direction of application to the structure.) The slab is included in the finite element model ;
loads are defined in Subsection 3.8.1.3 except as described in Subsection 3.8.1.4.1.1.
follows:
3.8J.4.2 Reactor Pedestal (1) P --Pressure loads resulting from the n8rmal operating pressure difference between The reactor pedestal is included in the the drywell (upper and lower) and the finite element model described in Subsection suppression chamber of the containment. 3.8.1.4.1.1.
Amendment 1 3.8-18
- I
23A6100AE Re '=d Plant nev.s The Reinforced Concrete Containment Vessel 3.8A.I.3 Radwaste Building Substructure
-(RCCV) in the' center of the RB encloses the Reactor Pressure Vessel (RPV). The RCCV supports The radwaste building substructure (RWB) is the upper pool and is integrated with the RB shown in Section 1.2.
structure from the basemat up through the elevation of the RCCV top slab. The interior The radwaste building is a reinforced floors of the RB are also integrated with the concrete structure 53m by 40m by 29.5m high.
RCCV wall. The RB has slabs and beams which join 'lbe building consists of a below grade the exterior wall. Columns support the floor substructure consisting of walls an3 slabs of slabs and beams. The fuel pool girders are reinforced concrete. 0.6m to 1.0m thich, forming integrated with the RCCV top slab and with RB a rigid box structure which se,ves as a well columns. The RB is a shear wall structure container to hold radioactive waste in case of designed to accommodate all seismic loads with an accident. This substructure is located below its walls. Therefore, frame members such as grade to increase shielding capability and to beams or columns are designed to accommodate maximize safety.- It is supported on a separate deformations of the walls in case of earthquake foundation mat whose top is 6.5m below grade.
conditions. In addition, a reinforced concrete superstructure 23m high extends above grade and. j 3.8A.1.2 Control Building houses the balance of the radwaste equipment. l The control building (CB) is located between The radwaste building substructure houses the the reactor building and the turbine building. high and low conductivity tanks, clean up phase It is shown in Section 1.2. separators, spend resin storage tanks, a concentrated waste storage tank, distillate tank The CB houses the essential electrical, and associated filters, and pumps for the control and instrumentation equipment, the radioactive liquid and solid waste 'reatment i control room for the reactor and turbine systems.
buildings, the CB HVAC equipment, RB cooling water pumps and heat-exchangers, the essential Although the radwaste superstructure is not a
'switchgear, essential battery rooms, and the Seismic Category I structure, its major steam tunnel. structural concrete walls and slabs are designed to resist Seismic Category I loads.
The CD is a Seismic Category I structure that houses control equipment and operation personnel 3.8A.1 A Seismic Category I Cable Troy and and is designed to provide missile and tornado Conduit Supports protection. The CB is constructed of reinforced concrete with a steel roof. The CB has three Electrical cables are carried on continuous stories above the ground level and three stories horizontal and vertical runs of steel trays below.' Its shape is a rectangle of 56 m (183 supported at intervals by structural steel l feet,8 inches) in the E-W direction,22 m (72 frames. The tray locations and elevations are feet,2 inches) in the N-S direction, and a predetermined based on the requirements of the height of about 38.7 m (127 feet) from the top of electrical cable network. Generally, several the base mat. trays of different sizes are grouped together and connected to a common support.
The CB is a shear wall structure designed to accomodate all seismic loads with its walls. The support frame spacing is determined by
'therefore, frame members such as beams or columns allowable tray spans, which are governed by are designed to accomodate deformations of the rigidity and stress. The frames may be walls in ease of earthquake conditions. ceiling-supported, or wall-supported, or a LO 3 & 21 Amendment 6
ABM 23x61oors Standard Plant REV.B l
. combination of both. Various type of frames form (c) Regulatory Guide 1.28, Quality Assurance {
a support system with transverse and longitudinal Program Requirements (Design and i bracing to the nearest wall or ceiling to take Construction);
the seismic loads. { '
3.8.4.2 Applicable Codes, Standards, and i
Specifications l
l 3J.4.2.1 Reactor Butiding l
1 The major portion of the reactor building, is not subjected to the abnormal and severe acci-der.t conditions associated with a containment. A listing of applicable documents follows:
(1) ACI 349, Code Requirements for Nuclear Safety Related Concrete Structures;
{
(2) AISC, Specification for Design, Fabrication k j
and Erection of Structural Steel for Buildings; 4 i
(3) ASME Boiler and Pressure Vessel Code Section 111, Subsection NE, Division 1,' Class MC (for design of main steam tunnel embedment t piping anchorage in the RB and CB only);
(4) AWS Structural Welding Code, AWS D1.1; 1
. (5) AWS Structural Welding Code, AWS D12.1;
.1 (6) NRC publications TID 7024 and TID 25021, Nuclear Reactors and Earthquakes and Summary {
of Current' Seismic Design Practice for Nuclear Reactor Facilities; .
{
(7) The inservice inspection requirements for the fuel poolliners in the Reactor Building are in conforraance with ASME Code Section III, Division 2.
(8) NRC Regulatory Guides: !
l (a) Regu.latory Guide 1.10, Mechanical (Cadweld) Splices in Reinforcing Bars of Category 1 Concrete Structures; IJ (b) Regulatory Guide 1.15, Testing of Reinforcing Bars for Category I Concrete Structures;
- 1 0'
Amendment 6 3.8 21a
k u
P r, ,
m LO (d) Regulatory Guide 1.29, Seismic Design Personnel for the Construction Phase of 1/. Classification;. Nuclear Power Plants; 3E .(e) Regulatory Guide 1.31, Control of (e) ANSI N45.2.9, Requirements for the Stainless Steel Welding; Collection, Storage, and Maintenance of OA Records for Nuclear Power Plants; 1 (f) Regulatory Guide 1.'44, Control of the
, 4' Use of Sensitized Stainless Steel; (f) ANSI N45.4, Leakage Rate Testing of Containment Structures for Nuclear (g) Regulatory Guide .1.55, Concrete Reactors;
. Placement in Category I Structures; (g) ANSI N101.2, Protective Coatings (h) Regulatory Guide 1.60, Design Response (Paints) for Light Water Nuclear Spectra for Seismic Design of Nuclear Reactor Containment Facilities; and Power Plants;.
(h) ANSI N101.4, Quality Assurance for (i) Regulatory Guide 1.61,~ Ouality Assurance Protective Coatings Applied to Nuclear Requirements for the Design of Nuclear Facilities; Power Plants; (10) Steel Structures Painting Council Standards (k) Regulatory Guide 1.69, Concrete Radiation-Shields for Nuclear Power (a) SSPC-PA 1, Shop, Field and Maintenance Plants.. Painting; (1); Regulatory Guide 1.76, Design Basis (b) SSPC-PA-2, Measurement of Paint Film p Tornado. Thickness with Magnetic Gages; (c) SSPC-SP-1, Solvent Cleaning;
^
. (m) Regulatory Guide 1.142, Safety-Related Concrete Structures for Nuclear Power Plants (Other than Reactor Vessels and (d) SSPC-SP-5, White Metal Blast Cleaning; Containment); and (c) SSPC-SP-6, Commercial Blast Cleaning; ,
(n) Regulatory Guide 1.94, Quality Assurance and l Requirements for Installation, Inspec-tion, and Testing of Structural Concrete (f) SSPC-SP-10, Near White Blast Cleaning; and Structural Steel During the Con-struction Phase of Nuclear Power Plants. (11) ACI ASCECommittec326, Shear and Diagonal Tension, ACI Manual of Concrete Practice, (9) ANSI: Part 2; (a) ANSI A58.1, Building Code Requirements (12) Applicable ASTM Specifications for for Minimusn Design Loads in Building and Materials and Standards; and Other Structures; (13) AASHTO Standard Specifications for Highway (b) ANSI N5.12, Protective Coatings (Paint) Bridges for truck loading area.
for the Nuclear Industry; 3.g.4.2.2 Control Building
. (c)~ ANSI N45.2, Quality Assurance Program Requirements for Nuclear Power Plants; Refer to Subsection 3.8.4.2.1.
p) ANSI N45.2.6, Qualifications of AM NRC Rules and Regulations Title 10, Chap-g Inspection, Examination and Testing ter 1, Coos of Federal Regulations, Part 73.2 and 73.55.
3,8 22 Amh 1
ABWR - l Standard Plant rIY B ]
3AA.23 Radweste Building Substructure Concrete fIoors and sIabs (including roofs) - 200 psf.
The radwnste building substructure shall be Stairs, stair platforms, grating designed using the same codes and standards as floors, and platforms - 100 psf.
the reactor building. Refer to Subsection Concrete roofs, live or snow load 3.8.4.2.1 for a complete list. (not concurrent) - 50 psf.
Construction live load on floor In addition, the non-Seismic Category I framing in addition to dead reinforced concrete portion of the weight of floor - 50 psf *.
superstructure is designed according to the seismic provisions of Section 2314 of the R = pipe reactions during normal uniform building code, operating or shutdown conditions based on the most critical 32A.2A Seismic Category I Cable Tray and transient or steady-state Conduit Supports condition (1) All codes, standards, and specifications ap- R
= pipe reactions under thermal plicable to'the building structures shall conditions generated by the also apply to cable tray and conduit postulated break and including supports. R 9
(2) AISI, Specification for the Design of Y
= equivalent static load on a Cold-formed Steel Structural Members. structure generated by the reaction on the broken (3) NEMA, Fittings and Supports for Conduit and high energy pipe during the Cable Assemblies. postulated break and including a calculated dynamic factor to 3.8AJ Loads and Load Combinations account for the dynamic nature of the load 3.8AJ.1 Reactor Building Y. = jet impingement equivalent static The temperature and pressure loads caused by I load on a structure generated by a LOCA do not occur on the reactor building. The the postu1ated break and reactor building ventilation system is designed including a calculated dynamic to keep the building within operating design factor to ac- count for the conditions. dynamic nature of the load.
3EA3.1.1 Loads and Notations Y = missile impact equivalent static load on a structure generated by Loads and notations are as follows: or during the postulated break, like pipe whipping, and including D = dead load of structure plus any a calculated dynamic factor to-other permanent load account for the dynamic nature of the load.
L = conventional floor or roof life loads, movable equipment loads, and W =
wind force (Subsection 33.1.)
other variable loads such as construction loads. The following iive Ioads are used:
- If the actual construction live load is greater than this value a design check of the structures willbe inade.
O Amendment 6 3.8-23 l
'?
ABM 23461ccas.
S*- 'ard Plant REVJ
't Wg = tornado load (Subsection 3.3.2)
(Tornado generated missiles are described in Subsection 3.5.1.4, and barrier design procedures in Subsection 3.5.3).
.The pressure part of Wt 55 defined .as either, (1) 360 mph velocity pressure with or without i
9 l
l O
Annendment 6 3.8 23a
ABWR m61**e Standard Plant _
REV,B 1.5 p si neg a t ive pressure Ta = thermal effects (including To )
differential, of (2) 3 psi negative which may occur during a design pressure differential. accident at 1650F maximum 30 minutes after LOCA Pa = internal negative pressure of 3.0 psig due to tornado; accident U = for concrete structures, the section pressure at main steam tunnel piping strength required to resist design embedment loads based on the strength design method described in ACI 318.
B = uplift forces created by tne rise of the ground water table H = loads caused by static or seismic F = internal pressures resulting from carth pressures.
flooding of compartments For structural steel, S is the required sec-E' = safe shu!down earthquake (SSE) loads tion strength based on the clastic design me-as defined in Section 3.7 thods and the allowable stresses defined in Part 1 of the AISC Specification for the Design, E = operating basis earthquake (OBE) Fabrication and Erection of Structural Steel for loads as defined in Section 3.7 Buildings.
To = thermal effects - load effects 3.8.43.1.2 Load Combinations for Concrete induced by normal thermal gradients Members existing through the reactor building wall and roof. Both summer For the load combinations in this subsection, and winter operating conditions are where any load reduces the effects of other considered. In all cases the loads, the corresponding coefficient for that conditions are considered of long load shall be taken as 0.9 if it can be enough duration to result in a demonstrated that the load is always present or straight line temperature gradient. occurs simultaneously with the other loads.
The temperatures are as follows: Otherwise, the coefficient for that load shall be taken as zero.
(1) Summer operation:
(1) Normal operating conditions - The strength (a) air temperature inside building - design method is used and the following 1200F load combinations are sat:l;d:
(b) exterior temperature - 1150F U =1.4 D + 1.7 L + 13 To + 1.7 Ro+
(2) Winter operation: 1.7 H + 1.4 B (a) air temperature inside building - U =1.4 D + 1.7 L + IS To + 1.7 Ro+
700 F 1.7 H + 1.9 E (b) exterior temperature - (-) 400F U el.4 D + 1.7 L + 13 To + 1.7 Ro+
(3) Winter shutdown 1.7 H + 1.7 W (a) air temperature inside building - For fluid pressure F, replace 1.7 H by 1.7 500 F F in the last two of the three equations (b) exterior temperature - (-) 400F above.
For all cases as-constructed temperature is (2) Abnormal / extreme environmental conditions -
600 F The strength design method is used and the O:
Amendment 6 3.8-24
r MM 23A6100AE REV B 1
e Standard Plant .l (T following load combinations are satisfied: (2) Abnormal / extreme environmental conditions - '
U The clastic working stress design method is used and the following load combinations are U = D + L + To + Ro+H+B l satisfied:
U =D + L + To+ Ro + H + E' 1.6 S = D + L + To + Ro + E' U =D + L + To + Ro+H 1.6 S = D + L + To+ Ro + Wg.
l,.1- U =D + L + To+Ro+H+Wg l'
,.i 1.6 S = D + L + To+ Ro + Wg + H
' U -D + L + Ta + R a+ 1.5 Pa + H j; 1.6 S = D + L + aT + Ra+P- a
't' U =D + L + Ta + Ra + 1.25 Pa+H+
1.25 E + '(Yr + Yj + Ym) 1.6 S = D + L + Ta + Ra + 1.0 E + Pa+
(Yj + Yr + Ym)- g U = D 4 L + Ta + R a+P+H a
+ E' + (Yr + Yj + Ym) 1.6 S = D + L + Ta + Ra + E' + Pa+
(Yj + Yr+Ym)-
3.8.4.3.1.3 Load Combinations for Steel l, Members In all these load combinations, both cases of 4 L having its full value or being completely (1) Normal operating conditions the elastic absent are checked, working stress design method is used for the l following load combinations: 3.8.4.3.2 Control. Building and Radwaste l Building Substructure g) i V
S=D+L.
Refer to the loads, notations, and combina-tions established in Subsection 3.8.4.3.1, E=D+L+E except that fluid pressure F, accident pressure S=D+L+W Pa , and pipe break loads Yr . Y J, Y m do not exist and the live loads are as follows:
'Since thermal stresses due to To and R o are present and are secondary and self limiting All concrete floors - 400 psf in nature, the following combinations are also satisfied: Stairs, stair platforms, grating floors, and platforms - 100 psf 1.5 S = D + L + T o + Ro.
Roof live or snow load (non concurrent) - 50 1.5 S = D + L + oT + Ro+E psf 1.5 S = D + L + To + W. Construction live load on floor framing in addition to dead weight of floor - 50 psf' In all these load conditions, both cases of L having its full value or being completely To = thermal s ffects. As constructed absent are checked. temperai u re is 60 0 F. The temperatures inside the building are !
as follows:
If the actual construction live load is Q
V greater than this value a design check of the structures will be made.
Arnendment 6 3.8-25
~ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ . _ _ _ _ . . _ _ _ _ _ _
I~ )
L )
23A6100AE Standard Plant REV.B Operating Conditions (1) Rigid Support with Flexible Tray. In this method, trays are modeled as flexible .
Control room, summer 750F elastic systems and analyzed by the response I winter 700F spectrum method. The resulting reactions HVAC room, summer 950F are used for the design of the supports. j winter 600F Other areas, summer 750F (2) Flexible Support with Flexible Tray. In 4 winter 750F this method, the composite system of trays and supports is modeled and analyzed by Shutdown condition computer as a multidegree of freedom elastic system. The support motions can be Controi room, summer 800F prescribed by the appropriate floor response winter 500F spectrum. The resulting responses are used HVAC room, summer 1040F to obtain design loads for the supports.
winter 500F Other areas, summer 900F 3.8.4.4.4.2 Conduit Supports i winter- 500F The design and analysis of conduit supports 3.8.43.3 Seismic Category I Cable Tray are basically the same as for cable tray and Conduit Supports supports. As conduits are more flexible and have comparatively less dead load, a rigid Loads and load combinations for the tray and support approach is used as d: scribed in method conduit supports shall use the same criteria as (1) of cable tr.v support design.
for the building structures where the supports are located. 3.8.4.5 Structural Acceptance Criteria 3.8.4.4 Design and Analysis Procedures 3.8.4.5.1 Reactor Building 3.8.4.4.1 Reactor Building 3.8.4.5.1.1 General Criteria The design prcicedures for the reactor building The first criterion is that the reactor are discussed in Subsection 3.8.3. The analysis building shall provide biological shielding for procedures for the reactor building are discussed plant perscanel and the public outside of the in Subsection 3.8.1 site boundary. This criterion dictates the minimum wall and roof thicknesses.
3.8.4.4.4 Seismic Category I Cable Tray and Conduit Supports The second criterion is that the reactor building shall protect the reinforced concrete 3.8.4.4.4.1 Cable Tray Supports containment from environmental hazards such as tornado and other site proximity-generated Wherever possible, the supporting frames for a missiles. The shielding thicknesses are tray or group of trays are designed to have sufficient for this purpose.
adequate rigidity to avoid causing additional amplification of seismic acceleration transmitted The reactor building provides a means for by the building structures. Where rigidity collection of fission product leakage from the cannot be achieved without an excessive increase reinforced concrete containment following an in support member size, the design of the accident.
supports is then based on the amplified seismic load obtained from the floor response spectra. The reactor building SGTS is designed to keep the compartments surrounding the reinforced Thus, two metLods are used in design and concrete containment at a negative pressure even analysis of cable tray supports. after a LOCA. In order to achieve a maximum in. leakage rate of 50% per day under a pres 6ure Amendment 6 3.8-26
7 MM 21A6100AE h=d- d Plant arv n differential of 6 mm of water, the reinforcing The reactor building foundation is' a steel is designed to remain clastic during the rectangular reinforced concrete mat 56 m (183 SSE load combinations. feet,8 inches) by 59 m (193 feet,5 inches) and 5.5m (18 feet) thick. The foundation mat is 3A.43.1J Materials Criteria constructed of cast in place conventionally reinforced concrete. It supports the reactor Refer to the materials criteria established in building, the containment structure, the reactor e 3.8.5 for the strength and materials requirements pedestal, and other internal structures. The for the reinforced concrete reactor building. top of the foundation' mat is approximately 20.2.
m(66 feet,3 inches) below grade. The reactor 32.4J.2 Control Building building foundation is shown in Appendix 3H. ,
Structural acceptance criteria are defined in The containment structure foundation, defined '
the AISC Specification and ACI 318 Code. In no as within the perimeter or the exterior surface case does the allowable stress exceed 0.9 F of the containment structure, is integral with is the ' minimum specified yield the reactor building foundation. The where stress. ' TF[e design criteria preclude excessive containment foundation mat details are discussed deformation of the building. The' clearances in Subsection 3.8.1.1.1.
between adjacent buildings are sufficient to prevent impact during a i:ismic event. The The control building foundation is tornado load analysis for this building is the rectangular reinforced concrete mat 22 m (72 same as'the analysis for the reactor building. feet,2 inches) by 56 m (183 feet,8 inches) by 5.5 m (18 feet) thick. The top of the 3.8.43.3 Radwaste Building Substructure foundation mat is 13.2 m (43 feet,4 inches) p below grade.
Structural acceptance criteria are defined in The radwaste building foundation is a-AISC Specification and ASCI 318 Code. In no case does the allowable stress exceed 0.9Fy where rectangular reinforced concrete mat 53m by 40m
- j. Fyis the minimum specified yield stress. The and 3m thick. The foundation mat is constructed
' design criteria preclude excessive deformation of of cast-in place conventionally reinforced the building. The clearance between adjacent concrete. It supports the radwaste building buildings are sufficient to prevent impact during structure, a seismic event.
3.83.2 Applicable Codes, Standards and 3J.5 Foundations Specifications This section describes foundations for all The applicable. codes, standards, seismic Category I structures of the ABWR specifications and regulations are discussed in Standard Plant. Subsection 3.8.1.2 for the containment foundation and in Subsection 3.8.4.2 for the 3AJ.1 Description of the Foundations other seismic Category I foundations.
The foundations of the reactor building and 3.833 L.oads and Load Combinations l
control building are reinforced concrete mat
- l. foundations. The configuration and relative The loads and load combinations for the location of these foundations are shown in containment foundation mat are given in Appendix 3H. S ubsection 3.8.1.3. The loads and load combinations for the other seismic category 1 These two foundation mats are separated from structure foundations are given in Subsection each other by a separation gap of 2 meters (6 3.8.4.3.
T feet,6 inches) wide to minimize the structural 1 interaction between the buildings. The loads and load combinations for all seismic Category I foundations examined to check Amendnwnt 6 3.g.27
, I 23A6100AE l
Standard Plant REV.B against sliding and overturning due to carth- ,
quakes, winds and tornados, and against flo-tation due to floods are listed in Table 3.8 7.
The lateral carth pressure loads are shown in )
Figure 3.8-19, 3.8.5.4 Design and Analysis Procedures The foundations of seismic category I structures are analyzed using well established methods where the transfer of loads from the foundation mat to the supporting foundation media is determined by clastic methods.
Bearing walls and columns carry all the vertical loads from the structure to the foundation mat. Lateral loads are transferred to shear walls by the roof and floor diaphragms.
The shear walls then transmit the loads to the foundation mat.
The design of the mat foundations for the structures of the plant involves primarily determining shear and moments is, the reinforced concrete and determining the O
O Amendment 6 3.8-27a
l MM,,_.
23A6100AE m
t M. interaction of the ' substructure with the The calculated and allowable factors of
- underlying foundation medium. For a mat safety of the Nuclear Island structures for G foundation supported on soil or rock, the overturning, sliding, and flotation are shown in .,
pertinent aspects in the design are to maintain Sections 3H.1 and 3H.2. {
the bearing pressures within allowable limits, _ i particularly due to overturning forces, and to 3AJE Materials, Quality Centrol, and ensure that there is adequate frictional and Special construction Techniques passive resistance to prevent sliding of the structure when subjected to lateral loads. The foundations of seismic Category I structures are constructed of reinforced The design loads considered in analysis of the concrete using proven methods common to heavy foundations are the worst resulting forces from industrial construction. For further discussion the superstructures and loads directly applied to see Subsections 3.8.1.6 and 3.8.4.6.
the foundation mat due to static and dynamic load : ,
combinations. 323.7 Testing and Inservice Inspection - ']
. Itequimments ]
The applicant will evaluate the capability of j
.the foundation to transfer shear if a A formal program of testing and inservice l waterproofing system is used on the foundation. inspection is not planned and is not required j for the seismic Category I structures of the .
The standard Nuclear Island design is' Nuclear Island. !
developed using a range of soil conditions as i detailed in Appendix 3A. The applicant will ;
determine variations of physical properties of !
the site specific subgrade materials. The applicant will calculate settlement of the foundations, differential settlement between O)
( foundations for the site-specific foundations medium and will' design safety related systems (i.e., piping, conduit, etc.) for the calculated settlement of the foundations. The applicant will evaluate the effect cf the site-specific subgrade stiffnets and calculated settlement on the design of the seismic Category I structures and foundations.
- A detailed description of the analytical and design methods for the reactor building foundation mat including the containment
' foundation, is included in Section 3.8.1.4.
3333 Structural Acceptance Criteria l
l The main structural criterion for the L containment portion of the foundation is adequate j- strength to resist loads and sufficient stiffness !
to protect the containment liner from excessive
. strain. The acceptance criteria for the containment portion of the foundation mat is presented in Subsection 3.8.1.5. The structural acceptance criteria for the reactor building
. foundations are described in Subsection 3.8.4.5.
5 Ameedesent 4 3.8-28 l
l-. ... . _ _ _ ___ _ ____ _ ___ _ _____ _.__ _ ___
gs.
23A6t00AE h=dard Plant REV B
. gy Table 3.9-4 DEFORMATION LIMIT FOR SAFETY CLASS REACTOR INTERNAL STRUCTURES ONLY Either One of (Not Both) General Limit
- a. Permittible Deformation. DP S .0S__
Analyzed Deformation .
SFmin Causing Loss of Function, DL
- b. Permissible Deformation. DP 5 LQ_. (page1) 4 Experiment Deformation SFmin si .
Causing loss of Function, DE Where:
DP = Permissible deformation under stated conditions of Service levels A, B, C or D (normal, upset, emergency or fault)
DL = Analyzed deformation which could cause a system loss ;f functions (*)
DE = Experimentally determined deformation which could cause a system loss of function SFmin - Minimum safety factor (see Subsection 3.9.53.6)
(1) Equation b will not be used unless supporting data are provided to the NRC by General Electric.
o A
- Loss of Function" can only be defined quite generally until attention is focused on the component (2) cf interest. In cases of interest, where deformation limits can affect the function of equipment and ' components, they will be specifically delineated. From a practical viewpoint, it is convenient to interchange some deformation condition at which function is assured with the loss of f function condition if the required safety margins from the functioning conditions can be achieved. Therefore, it is often unnecessary to determine the actual loss of function condition because this interchange procedure produces conservative and safe designs. Examples where deformation limits apply are: control rod drive alignment and clearances for proper insertion, reactor internal pump wear, or excess leakage of any component.
1 I
1 V \
Amendment 6 3D
. _ _ _ _ _ - ._______________-_____________D
ABM 23461mse Jgandard Plant nrv. s Table 3.9 5 PRIMARY STRESS LIMIT FOR SAFETY CLASS REACTOR INTERNAL STRUCTURES ONLY Amy One Of (No More Than One Reauired) GeneralIJalt
- a. Pf actic evaluated crimary stresses. PE S .22.1, Permissible primary stresses, PN SF ,.,
- b. Permissible load. LP g _L5_
Largest lower bound limit load, CL SFmm.
- c. Elastic evaluated crimary stress. PE $ ,DJj Conventional ultimate strength SF ,.,
at temperature, US
- d. Flastic. plastic evaluated nominal crimary stress. EP S _QS , J Conventional ultimate strength SF ,.
at temperature, US
- e. Permissible load. LP s .03_ (Note 1)
Plasticinstabilityload, PL SF .
mm
- f. Permissible load. LP 1.QS9_ (Note 1) 3 Ultimate load from fracture SF**. 2 analysis, UF
- g. Permissible load. LP s LQ (Note 1)
Ultimate load or loss of function SF**. '
load from test, LE
)
where l PE = Primary stresses evaluated on an elastic basis. The effective membrane stresses are to i be averaged through the load carrying section of interest. The simplest average bending, shear or torsion stress distribution which will support the external loading ;
will be added to the membrane stresses at the section of interest.
PN = Permissible primary stress levels under service level A or B (normal or upset) conditions under ASME Boiler and Pressure Vessel Code,Section III.
LP = Permissible load under stated conditions of service level A, B, C or D (normal, upset, j emergency or faulted).
I l
O' Amendment 3 3.9-55
MN 23A6100AE Etandard Plant REV.A
.,/^N b) SECTION 3.10 CONTENTS ,
1 Section Title East 3.10.1 Seismic Qualification Criteria (Includine Other
_Dmamic Loads) 3.10-1 3.10.1.1 Selection of Qualification Method 3.10-1 3.10.1.2 Input Motion 3.10-1 3.10.13 Dynamic Qualification Program 3.10-1 3.10.2 Methods and Procedures for On=llMnn Electrical Eautoment and Instrumentation 3.10-2 '
3.10.2.1' Qualification by Testing 3.10-2 3.10.2.1.1 Interface Requirements 3.10-2 3.10.2.1.2 Test Methods 3.10-2
/~N Q 3.10.2.1.2.1 Selection of Test Specimen 3.10-2 3.10.2.1.2.2 Mounting of Test Specimen 3.10-3 3.10.2.1 3 Dynamic Testing Sequence 3.10-3 3.10.2.13.1 Vibration Conditioning 3.10-3 3.10.2.1 3.2 Exploratory Tests 3.10-3 3.10.2.133 OBE Testing Including Other RBV Dynamic leads 3.10-3 3.10.2.1 3.4 SSE Testing Including Other RBV Dynamic Loads 3.10-3 J
3.10.2.13.5 Qualification for Operability 3.10-3 3.10.2.1.4 FinalTest Report 3.10-4 3.10.2.2 Qualification by Analysis 3.10-4 3.10.2.2.1 Analysis Methods 3.10-4 ,
3.10.2.2.2 Analyses for OBE and SSEIncluding 1 Other RBV Dynamicleads 3.10-4 )
O 3.10-ii Amendment 1
g-I 23A6100AE !
l Standard Plant REV.B l n
SECTION 3.10 $i CONTENTS (Continued) i Section 11tig Eagg 3.10.2.23 Documentation of Analysis 3.10 5 3.10.2 3 Qualification by Combined Testing and Analysis 3.10-5 3.10.2 3.1 Low Impedance Excitation 3.10-5 3.10.23.2 Extrapolation of Similar Equipment 3.10-5 3.10.233 Extrapolation of Dynamic Loading Conditions 3.10-5 3.103 Methods and Procedures of Analysis or Testine of Suonorts of Electrical Eauloment and Instrumentation 3.10-5 3.10 3.1 NSSS Electrical Equipment Supports (Other than Motors and Valve-Mounted Equipment) 3.10-6 3.10 3.2 Other Seismic Category I Instrumentation and Electrical Equipment Supports 3.10-6 3.10 3.2.1 Supports for Battery Racks, Instrument Racks, Control Consoles, Cabinets, and Panels 3.10-6 3.103.2.2 Cable Trays and Conduit Supports 3.10-7 3.10 3.2.3 LocalInstrument Supports 3.10-8 3.10 3.2.4 Instrument Tubing Support 3.10-8 3.10.4 Ooeratine License Review (Tests and Analyses Results) 3.10-8 3.10.5 Interfaces 3.10-8 3.10.5.1 Equipment Qualification Records 3.10-8 3.10.5.2 Dynamic Qualification Report 3.10-8 0
3.10-lii Amendment 6
[1 3 :o 1 ,
M 2W100AE '
Standard Plant REV.A Q5 b ' each vendor. The vendor. submits test' data, Tray supports are anchored in the following operating experience, and/or calculations to . . manner:
- verify that _the equipment.will not suffer any loss of function before, during, or after the (a) support members may be attached to specified dynamic disturbance. Analysis and/or . existing structural steel; H testing procedures are in accordance with Subsection 3.10.2. (b) support members may be attached to supplementary steel members spanning In essence, these supports are inseparable = ~ between existing floor beams (or between fross their supported items and are qualified with - existing floor beams and walls); or .
the items. During' testing, the supports are fastened to the test table with fastening devices (c) support members may be attached to '
.or methods used in the actual installation, ' concrete wall / columns using:
thereby qualifying the total installation.
(i) embedded steel plates with shear
'3.10.3.2.2 Cable Theys and Conduit Supports conneetors or I
All Seismic Category I cable trays and conduit (ii) steel plates or strut sections
., ' supports are designed by the response spectrum attached to concrete with concrete method. Analysis and dynamic load restraint expansion anchors.
l measures are based on combined limiting values for static load, span length, and response to (3) leads excitation 'at the natural frequency Restraint against excessive lateral and longitudinal move- (a) Deadloads andliveloads mest nses the structural capacity of the tray to I determine the spacing of the fixed support 75 lb/ linear ft load used for 18 inch points. Provisions for differential motion and wider trays between buildings are made by breaks in the trays 50 lb/ linear ft load used for 12 inch and flexible connections in the conduit. and narrower trays The following criteria are used in the design (b) Dynamic loads - OBE or SSE plus other of Seismic Category I cable tray and conduit RBV dynamic loads
' supports. _
(4) Imad Combination (1) Cable Tray Support Spacings (a) Deadload + liveload Tray support spacings for horizontal or:
vertical runs do not exceed ten feet unless (b) Dead load + live load + OBE + other RBV -
noted otherwise on design drawings (design dynamicloads drawings generally locate supports not more than eight feet apart with a longitudinal (c) Dead load + live load + SSE + other RBV tolerance of 1 foot to avoid interferences), dynamic loads '
Tray supports provide resistance to three (5) Dynamic Analysis cacitation directions by means of vertical, transverse,' and longitudinal support and (a) Regardless of cable tray function, all bracing systems. supports are designed to meet Seismic Category I requirements. Seismic and (2) Supportlocations other RBV dynamic loads are determined O
Assedemet 1 3.10-7
ABM 23462 oars Standard Plant REV.B by dynamic analysis using appropriate response exceeded when the tubing is subjected to the
- spectra. loads specified in Subsection 3.9.2 for Class 2
! and 3 piping. -
(b) Floor Response Spectra 3.10.4 Operating License Review (Tests and (i) Floor response spectra used are Analyses Results) those generated for the supporting floor. In case ;upports are See Subsection 3.10.5.2 for interface attached to the walls or to two requirement.
different locations, the upper bound envelope spectra obtained by 3.10.5 Interfaces superimposing are used.
3.10.5.1 Equipment Qualification Records (ii) In many cases, to facilitate the design, several floor response The equipment qualification records including spectra are combined by an upper the reports (see Subsections 3.10.2.1.4 and bound envelope obtained by 3.10.2.2.3) shall be maintained in a permanent superimposing. file and shall be readily available for audit.
3.103.2.3 Local Instrument Supports 3.10.5.2 Dynamic Qualification Report For field. mounted Seismic Category I A dynamic qualification report (DOR) shall be instruments, the following is applicable: prepared identifying all Seismic Category 1 instrumentation and electrical parts and (1) The mounting structures for the instruments equipment therein and their supports. The DOR have a fundamental frequency above the shall contain the following: (1) A table or file excitation frequency of the RRS. for each system that is identified in Table 3.2-1 to be safety-related or having Seismic -
(2) . The stress level in the mounting structure Category I equipment shall be included in the does not exceed the material allowable DOR containing the MPLitem number and name, the stress when the mounting structure is qualification method and the input motion for subjected to the maximum acceleration level all Seismic Category I equipment and the for its location. supporting structure in the systems and the corresponding qualification summary table or 3.103.2.4 Instrument T;,bing Support vendor's qualification report. (2) The mode of safety-related operation (i.e., active, manual The following bases are used in the seismic active or passive) of the instrumentation and and other RBV dynamic Icads design and analysis equipment along with the manufacturer of Seismic Category I instrument tubing supports: identification and model numbers shall also be tabulated in the DOR. The operational mode (1) The supports are qualified by the response identifies the instrumentation or equipment (a) spectrum method; that performs the safety related functions automatically, (b) that is used by the operators (2) Dynamic load restraint measures and analysis to perform the safety-related functions for the supports are based on combined manually, or (c) whose failure can prevent the limiting values for static load, span satisfactory accomplishment of one or more length, and computed dynamic response; and safety-related functions.
(3) The Seismic Category 1 instrument tubing systems are supported so that the allowable ,
stress permitted by Section til of ASME l Boiler and Pressure Vessel Code are not i Amendment 6 3 lM l
O
. APPENDIX 3B CONTAINMENT HYD.RODYNAMIC LOADS O
I l
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ABWR ux6 min Standard Plant nev. n
, APPENDIX 3B
' (_/
TABLE OF CONTENTS Section Itlle Page 3B CONTAINMENT HYDRODYNAMIC LOADS
38.1 INTRODUCTION
3B.11 Purpose 3B.11 3B.1.2 ABWR Containment Design 3B.1 1 3B.2 REVIEW OF PHENOMENA 3B.2.1 Safety / Relief Valve Actuation 3B.2-1 3B.2.2 Loss-of-Coolant Accidents 3B.2-2 3B3 SAFETY / RELIEF VALVE DISCHARGE IDADS 3B3.1 Ouencher Description and Arrangement 3B3-1 3B3.2 Ouencher Discharge Imads 3B3-1 3B33 Quencher Condensation Performance 3B3-2 3B.4 LOSS-OF-COOLANT ACCIDENT LOADS 3B.4.1 Pressure and Temperature Transients 3B.4-1 3B.4.2 Vem Clearing and Pool Swellleads Methodology 3B.4-1 3B.43 LOCA Steam Condensation Imads 3B.4-4 t
3D-ii Amendment 6 l
ABWR mame Standard Plant n, n APPENDIX JB 1 TABLE OF CONTENTS (Continued) 1 l Etclioll M P. age '
(
3B.5. SUBMERGED STRUCTURE LOADS 3B.5.1 Pool Swell Submerged Structure Loads 3B.5-1 3B.5.2 Condensation Oscillation Submerged Structure loads 3B.5-1 3B.5.3 Chugging Submerged Structure Loads 3B.5-1 3B.5.4 S/RV Submerged Structure Loads 3B.5-2 3B.6 LOADS COMBINATION 3B.6-1 3B.7 REFERENCES 3B.7-1 i
3D-iii O
Amendment 6
23A6100AE Standard Plant nev. n SECTION 3B.1
.i CONTENTS l
Section Elle East 3B.1.1 Purpose 3B,1 1 3B.1.2 ABWR Containment Deslan Features 3B.1 1 ILLUSTRATIONS Figure Ellt East 3B.1-1 ABWR Primary Containment Configuration 3B.12 O
O 3D.1-ii Amcodment 6
- MM 23A6100AE Standard Plant nem n
.3B.1 INTRODUCTION ~
GE PROPRIETARY - provided under separate cover (Pages 3B.11 and 3B.1-2) i l
l l
l O
Amendment 6 3D.1-1
-ABWR 23A61MAE Standard Plant nev. n SECTION 3B.2 CONTENTS Section M Eagt 38.2.1 Safety / Relief Valve Actuation 3B.2-1 3B.2.2 Loss-of-Coolant Accidents 3B. 2-2 3B.2.2.1 Large Break Accident 3B.2-2 3B.2.2.2 Intermediate Break Accident (IBA) 3B.2-3 3B.2.23 Small Break Accident (SBA) 3B.2-3 0 3B.2-ii Amendment 6
ABWR li ' Standard Plant *'^ f -
3B.2 REVIEW OF PHENOMENA xth j GE PROPRIETARY - provided under separate cover (Pages 3B,2-1 through 3B.2-4)
O
)
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Amendment 6 3D.2-1
23A6100AE Standard Plant - nev. n
. SECTION 3B.3 p,
U CONTENTS Section Elle Ea9s 3B3.1 Onencher Description and Arrangement 3B3-1 3B.3.2 Ouencher Discharue Loads 3B3-1 3BS.2.1 ABWR Design Quencher Discharge loads 3B3-1 3B3.2.1.1 Single Valve Discharge 3B3-2 3B3.2.1.2 Multiple Valves Discharge 3B3-2 3R33 Ouencher Condensation Performance 3B3-2 ILLUSTRATIONS Figure Dtig Eage 3B3-1 X-Quencher Schematic 3B3-4 O
-Q 3B3-2 X-Quencher Azimuthal Locations in the ABWR Suppression Pool 3D3-5 3B3-3 Dimensions for P(r) Calculation 3B3-6 3B3-4 Circumferential Distribution 3B3-7 3B3-5 Quencher Bubble Pressure Time History 3B3-8 3B3-6 Psol Boundary Pressure Time-History Measured for One Second of SRV Condensation 333-9 v 3B.3-ii Amendment 6 I
i
L:,
ABWR 23A61jAE.
Standard Plant 9 3B.3 SAFETY / RELIEF VALVE DISCHARGE LOADS
. - -4 XJ ,
e i
k i
e GE PROPRIETARY - provided under separate cover
- (N (Pages 3B3-1 through 3B3-9)
(.)
O -
Amendmen 6 3D.3-1 i
)
ABWR 23A61%AE Standard Plant ne . n f ,g SECTION 3B.4
(
CONTENTS Section Illit East 3BA.1 Pressurr and Temperature Transient 3B.4-1 3BA.2 Vent Clearing and Pool Swell Loads Methodology 3B.4-1 3BA.2.1 PoolBoundary Loads 3B.4-1 3B.4.2.2 loads on Access Tunnel 3B.4-2 3B.4.23 Impact and Drag Loads 3B.4-3 3B.4.2.4 Loads on Diaphragm Floor 3B.4-4 3BA3 LOCA Steam Condensation Loads 3B.4-4 3B.43.1 ABWR Horizontal Vent Test Program 3B.4-4 3B.43.2 Condensation Oscillation (CO) Loads 3B.4-5 3BA3.2.1 Description of CO Data Base 3B.4-5 3B.43.2.2 Evaluation of CO Data Base 3B.4-5 3B.43.23 CO load Definition 3B.4-5 j 3B.43.23.1 Source Ioad Approach 3B.4-5 3D.433 Chugging Loa 3s 3B.4-6 3BA33.1 Description of Chugging Data ' 3B.4-6 3B.433.2 Evaluation of Cbugging Data 3B.4-7 l 3B.4333 Chugging Load Definition 3B.4-7 3B.4333.1 PoolBoundaryLoads 3B.4-7 3B.4333.2 Loads on Access Tunnel 3B.4-7 3B.43333 loads on Horizontal Vent 3B.4-7 id 3B.4-ii Amendment 6
23A6100AE Standard Plant ama SECTION 3B.4 g TABLES Table Iltle Eage 3B.4-1 Pool SwellCalculated Values 3B.4-9 3B.4-2 Final ABWR HVT Test Matrix 3B.4-10 3B 4-3 Mean and Standard Deviation of CH POP and PRMS by Test 3B.4-11 3B.4-4 Average Chug Period by 10-Second Segment for Steam Breaks at 700F 3B.4-12 3B.4-5 Awrage Chug Period by 10-Second Segment for ;
Steam Breaks at 1200F 3D.4-13 3B.4-6 Average Chug Period by 10 Second Segment for Steam Breaks at 1550F 3B.4-14 3B.4-7 Average Chug Period by 10-Second Segment for Stcam Breaks at 70,120, and 1550F 3B.4-15 ILLUSTRATIONS Figure I111e Eagt 3B.4-1 Pool Boundary Pressure During Pool Swell -
Normalized to Bubble Pressure 3B.416 3B.4 2 Time History of Air Bubble Pressure 3B.4-17 3B.4-3 Drag Coefficient for CyGuders Following Impact 3B.4-18 3B.4-4 SS Test Facility 3B.4-19 3B.4-5 FS* Test Facility 3B.4-20 3B.4-6 Test Sensors Common to FS* and SS Tests 3D.4-21 3B.4-7 Test Sensors Unique to FS* and SS Tests 3B.4-22 3B.4-iii Amendment 6
23A6100AE Standard Plant nev n q SECTION 3B.4 V CONTENTS (Continued) ,
ILLUSTRATIONS Figure Iltle Eage 3B.4-8 Envelope PSD at 019P for SST 1,2,3,9,11 and 12 3B.4-23 3B.4-9 Six-Test and Key-Segment Envelope PSDs at 019P 3B.4 24 3B.4-10 ABWR CO Source Load Methodology 3B.4-25 3B.4-11 Typical Large Chug (025P) 3B.4-26 l
3B.4-12 PSD of Typical Large Chug (025P) 3B.4-27 3B.4-13 ABWR CH Source load Methodology 3B.4-28 3B.4-14 Comparison of PSDs Between Analysis and Test 3B.4-29 3B.4-15 Spatialload Distribution for CH 3B.4-30 3B.4-16 Typical Wall Pressure Time History Due to CH 3B.4-31 D 3B.4-17 Circumferential Pressure Distribution on h Access Tunnel Due to CH 3B.4-32 3B.4-18 Typical Test Result of Upward Load Due tc CH 3B.4-33 3D.419 Typical Tests Result of Moment Due to CH 3B.4-34 3B.4-20 Horizontal Vent Upward Leading for Vent Pipe and Pedestal 3B.4-35 3B.4-21 Horizontal Vent Upward Loading for Structure Response Analysis 3B.4-35 3B.4-iv Amendment 6
'M 23A6100AE Standard Plant nev.n
- 3B.4 LOSS OF COOLANT ACCIDENT LOADS I
i I
GE PROPRIETARY - provided under separate cover (Pages 3B.4-1 throtgh 3B.4-35)
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AtueMment 6 3D 41
l ABWR meime Standard Plant nem n 4 1
,e g SECTION 3B.5 CONTENTS l Section lille Eage ,
l I
l 38.5.1- Eggi Swm 3B.5-1 1
38.iL2 Condensation Oscillation Submerned Structure 14Huht 3B.5-1 3BJ.3 Chumafna Submermed Structure Loads 3B.5-1 3B 5.4 S/RV Syhgarmed Structure Loads 3B.5-2 l
I i
I 3B.5-ii Amendment 6
23A6100AE
. Standard Plant ,
a, n t .- 138.5. SUBMERGED STRUCTURE LOADS
.p);
L. :
)
i i
l GE PROPRIETARY - provided under separate cover (Pages 3B.5-1 through 3B.5-2)
O Amendment 6 3D3-1
ABWR .3 23A6100AE Standard Plant Rev. H 38.6 LOADS COMBINATION l l
i f
GE PROPRIETARY - provided under separate cover D,
1 O
,.I- Amendment 6 3D.6-1
23A6100AE Standard Plant a,.. a
.. 3B.7 REFERENCES O'
GE PROPRIETARY - provided under separate cover O
l O 1 l
Amendment 6 3D.7-1
23A6100A2 s Standard Plant un y
a i
~V 3G.3 Additional Parametric stabilizer are increased by the scale factors deter.
Analyses mined from this parametric study. The resulting site-envelope forces are applicable to the design with This section presents additional analyses and and without the RSW stabilizer.
l.
l- their results performed to evaluate the impact on the reactor building resp: nse due to 1) the elimination The RSW stabilizer effects on the floor response of the stabilizer between the reactor shield wall spectra are shown in Fig. 3G.3-1 by comparing re-(RSW) and the RCCV and 2) the revised control sponses with and without the RSW stabilizer at key building design. locaticas. As noted, the spectra at building loca6ons (nodes 88,89, and 95) are not affected by the RSW l
stabilizer. ' At node 33 (RPV/'nain steam nozzle),
3G.3.1 Effect of RSW stabilizer the most pronounced effect due to the RSW stabi- l lizer occurs in a relatively narrow frequency band l The base desi6n considered in the SSI analyses ranging approximately from 10 to 12 Hz in which the presented in Section 3G.2 includes a stabilizer response variation can be accounted for by peak between the RSW and RCCV. This stabihzer, here- broadening and smoothing. Furthermore,in estab-after referred to as the RSW stabilizer, is repre- lishing site envelope floor response spectra the sented by a horizontal spring element spanning from SASSI-envelope spectra are scaled up in a conserva- '
node 70 on the RSW to node 90 on the RCCV as tive manner to account for the effects of the alter-shown in Fig. 3G.21. The stabilizer is inactive in the nate CLASSI/ASD analysis. The margins intro-vertical direction. To investigate the design option duced are sufficient to cover response variations due without the RSW stabilizer, a parametric study is to the elimination of the RSW stabilizer.
performed.
In summary, the effects of the RSW stabilizer on O
d Since the structural responses are found to be governed by stiffer sites in the base case analyses, the the reactor building response are negligible, except at some local RPV/internallocations. The site-en-representative rock site with the HRD85 profile is velope seismic loads presented in Section 3G.4 chosen for this study. The 3 D SSI analysis per- include the RSW effects and thus are applicable to formed for case T10X is repeated by removing the the design with and without the RSW stabilizer.
RSW stabilizer spring from the model.
3G.3.2 Effect of Revised Control Building The maximum forces at key structural elements computed for the case without t. t RSW stabilizer The control building included in the R/B SSI are compared to those corresponN to the case analyses described earlier has been revised as a with the RSW stabilizer in Table 30.3-1. As shown, result of further study of the standard plant buildings the RSW stabilizer has very small effects on the base arrangement. The revised C/B has an overall planar forces on the R/B, RCCV, s nd pedestal. Some force dimensions of 22 m x $6 m (72 ft x 184 ft) as reductions for the case without the RSW stabilizer compared to 16 m x 45 m (52 ft x 147 ft) of the are observed for the RSW and RPV skirt An in- previous de; sign. The embedment depth is increased crene in forces ahghtly higl.u than that noted for from 11.2 cn (40 ft) to 25.7 m (85 ft) which is the the shroud support is found at higher elevations same as the R/B embn' ment depth The revisd along the shroud assemb;y, when the RSW stabilizer C/B remains to be an independent building and is is remove d. The component which is affected the founded on a separate basemat from the R/B. Sam:
most by the RSW .stabidzer is the RPV refuelling as the previous plant layout, the C/B is situated ig bcIlow which is represented by a spring element lectween the R/B and T/B alorg the R/B 0 from node 90 on the RCCV to node 31 on the RPV direction. According to the current plant arrange-as shown in Fig. 3G.2-1. The maximum force on the ment, the T/B is embedded to a depth about 9 m (30 i refuelling bellow without the RSW stabilizer is 73% ft) which is about half of the embedment depth higher than that with the RSW stabilizer. For the considered previously. The effect of the revised C/B structural elements that are subjected to appreciably design and current T/B embedment depth on the f reactor building response due to through soil i higher forces when the RSW stabilizer is eliminated, the site-envelope forces for the design with the RSW structure-to- structure interaction is evaluated. 4 Amen knent 4 3GM i
MM 23A6100AE Standard Plant new s The previous analysis results indicate that the effect of the adjacent control and turbine buildings on the reactorbuilding response is secondary and does not change the response of the reactor building significantly. Softer sites are found to be more i sensitive to structure-to- structure interaction but stiffer sites cause higher R/B response. In view of the site enveloping concept, the VPSD150 site among the four sites previously analyzed for adjacent building effect is chosen for this parametric analysis.
Horizontal analysis for case C6-1X is repeated according to the current C/B and T/B conditions.
The analysis results of maximum forces at key locations in the reactor building complex due to structure-to-structure interaction with the control building of different designs are compared in Table 3G.3-2. The responses predicted for the same site condition of the R/B alone case are also included in this table. The result comparisons between the two cases including adjacent buildings eveal that only the 9-/B base experiences noticeably higher forces resulting from the revised C/B. As compared with the results of the R/B alone case, the analysis including the revised C/B predicts smaller force responses at allocations examined, except for the shroud support moment which is increased by a negligible 1.4% and for the R/B base shear which is increased by 10% The increased base shear of the R/B is, however, only about 57% of the site-envelope load predicted by SASSI (see Table 3G.2-6). Therefore, the site- envelope forces for the reactor building complex established previously remain valid when the effect of the revised design of the C/B is considered.
' The horizontal Door response spectra at four key locaions in the R/B complex calculated for the two cases with the C/B included and the case of the R/B alone are compared in Figure 3G3-2. When the ad..
pcent buildings are considercJ, the puk amplitudes of the case with the revised C/B are generally less J than test of tt-e case with the origbaal C/B. The n sponse spectra ohhe former case are very sindlar ;
l to that of the R/B alone case. No new spectrai !
peaks are introduced as the C/B is revised. This is consistent with the previous finding as described in l subsection 3G.2.4. The, effect of the revised C/L due j to structure-to-structure interaction on the R/B 1 response is therefore secondary and insignificant. l In summary, the site-envelope seismic loads on
( the reactor building complex are not affected by the ,l l revised controlbuilding design.
Amendment 6 3G.3-2 l
}
ABM 23x61004s Standard Plant - nev, s
_,g.
Table 3G3-1 Effect of RSW Stabilizer on Maximum Forces Response HRD85 - T10X Beam Elt location _Iypt. .w. lith. without 28 Shroud support Shear 201 206 Moment 1330 1420 69 RPV Skirt Shear 496 467 Moment 3390 2785 78 RSW Base Shear - 804 672 Moment 4080 3035 86 Pedestal Base Shear - 1230 1294 Moment 23600 22576 92 RCCV Base Shear 6000 6096 Moment 221000 222720 102 R/B Base Shear 5680 5760
, Moment 64800 65200 N]/
Units: Shear in tons; Moment in ton-meters 1 ton = 2,204 lbs; I ton-meter = 86.75 7.10' in-lb.
1 l -
U Amendment 4 3G.3-3
q L -
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MM 23A6100AE ]
Standard PJant Rev. s !
l l- Table 3G.3-2 Effect of Revised Control Building On Maximum Forces .
VPSD150 Profile ,
R/B+ C/B +T/B l Response Case C64X R/B Alone Beam Elt Location h Revised Original case C6X 28 Shroud support Shear 171 206 178 Moment 11 % 1410 1180 69 RPV Skirt Shear 393 422 451 Moment 2360 2540 2870 78 RSW Base Shear 523 549 648 Moment 2670 2550 3220 86 Pedestal Base ' Shear 1320 1250 1530 Moment 24200 22500 28100 92 RCCV Base Shear 6400 5680 6970 Moment 194700 171000 214000 102 R/B Base Shear 8170 5820 7430 Moment 363000 303000 404000 Units: Shear in tons; Moment in ton meters 1 ton = 2,204 lbs; 1 ton-meter = 86.75 x 10'in-lb.
Amendment 6 3G.3 4
ABWR .234aooxn Standard Plant me. n O sectionana TABLES l Table Tult East 3H3-1 Rebar Used In Design Evaluation 3H3-2 3H3-2 Forces and Moments in the Containment Structure Wall 3H3-7 3H3-3 Rebar and Concrete Stresses in the Containment Structure Wall 3H3-13 3H3-4 Forces and Moments in the Containment Structure Top Slab 3H3-19 3H3-5 Rebar and Concrete Stresses in the Containment Structure Top Slab 3H3-21 3H3-6 Forces and Moments in the Foundation Mat 3H3-23 3H3-7 Rebar and Concrete Stresses in the Foundation Mat 3H3-24 3H3-8 Forces and Momentsin the D/F Slab 3H3-26 A 3H3-9 Rebar and Concrete Stresses in the D/F Slab 3H3-29 V 3H3-10 Forces and Moments in the Reactor Pedestal 3H3-32 3H3-11 Steel Stresses in the Reactor Pedestal 3H3-33 3H3-12 Forces and Moments in the Fuel Pool Girder 3H3-35 3H3-13 Rebar and Concrete Stresses in the Fuel Pool Girder 3H3-40 3H3-14 Comparison of Forces and Moments at Various Sections For Uncracked and Cracked Conditions Due to; SRSS of 9 Two Ilorizontal SSE Component with the Vertical SSE Component 3H3-44 3H3-15 Comparison of Rebu and Concrete Stresses at Sections 1,2, and 4 for Uncracked and Cracked Conditions Due to Load Combinations *Sa and 15b 3H3-46 3H3-16 Comparkon of Rotational Strains at Sections 1 and 2 forUncracked and Cracked Conditions Due to lead Combinations 15a and 15b 3H3-47 3H3-17 Steel Stresses in the Reactor Pedestal at Sections 7 and 9 (Cracked Section Analysis) 3H3-48 (y
V 3H3-il Amendment 4
ABM 33462ooin '
= Standard Plant nev n SECTION 3H.3 g ILLUSTRATIONS Figure 11 tic Eage ,
.l 3H3-1 Containment Structure Wall Reinforcement 3H3-50 .l l
1 3H3-2 Containment Structure Opening Reinforcement 3H3-51 3H3-3 Containment Structure Opening Reinforcement 3H3-52 3H3-4 Containment Structure Top Slab Reinforcement 3H3-53 )
3H3-5 Reactor Building Foundation Reinforcement 3H3-54 3H34 Not used 3H3-55 3H3-7 Diaphagm Floor Reinforcement - 3H3-56 I O
i 3H3.m &
Amendment 6
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APPENDIX 3I EQUIPMENT QUALIFICATION ENVIRONMENTAL DESIGN CRITERIA I
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23A6100AE Standard Plant nev.n
. , ,q(s APPENDIX 3I )
TABLE OF CONTENTS i
-sraina noe ran 31 EOUIPMENT-QUALIFICATION ENVIRONMENTAL DESIGN CRITERIA
31.1 INTRODUCTION
31.1 1 31.2 PIANT ZONES 31.2 1 31.2.1~ Inside Primary Containment 31.2-1 31.2.2 Outside Primary Containment 31.2-1 313 ENVIRONMENTAL CONDITIONS PARAMETERS 313-1 31 3.1 Plant hormal Operating Conditions 313-1 31 3.2 Plant Abnormal / Accident Conditions 313-1 b
31-ii Amendment 6
23A6100AE
. Standard Plant __
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31.1 INTRODUCTION
v ,
r GE PROPRIETARY - provided under separate cover iO :
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f ABWR ux62 min Standard Plant REV.B
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r SECTION 31.2 CONTENTS Section Titic East .
i 31.2.1 Inside Primary Containment 31.2-1 31.2.2 Outside Primary Containment 31.2-1 SECTION 31,2 TABLES Table Illit Eagt 312 1 Plant Zones 31.2-2 O
SECTION 31.2 ILLUSTRATIONS Figure I!11e fagt 31.2 1 Zones in Primary Containment Vessel 31.2-6
\ .
31.2-il Amendment 6
. i;.
0- + - 23A6100AE
- Standard Plant a,, n 31.21 PLANT ZONES l
GE PROPRIETARY - provided under separate cover
-h- (Includes pages 31.2-1 through 31.2-6) l i
I Amendment 6 31.2 1
' 23A6100AE Ln ' Standard Piant nev a Q
SECTION 3I3 CONTENTS Section Title Eggg 31.3.1 Plant Normal Ooeratine Conditions 31 3-1 31 3.1.1 Pressure, Temperature and Relative Humidity 31 3-1 31 3.1.2 Radiation Environment Conditions 31 3-1 313.2 Plant Abnormal / Accident Conditions 31 3-1 31 3.2.1 Pressure Temperature and Relative Humidity 31 3-1 31 3.2.2 Radiation Environment Conditions 11 3-1 SECTION 3I3 b TABLES Table Tjils Eage 313-1 Thermal Environment Conditions Inside Primary Containment Vessel, Plant Normal Operating Conditions 313-2 313-2 Thermal Environment Conditions inside Reactor Building (Secondary Containment), Plant Normal Operating Conditions 313-3 31 3-3 Thermal Environment Conditions inside Reactor Building (Outside Secondary Containment), Plant Normal Operating Conditions 31 3-4 31 3-4 Thermal Environment Conditions inside Turbine Building, Plant Normal Operating Conditions 31 3-5 31 3-5 Thermal Environment Conditions inside Control Building, Plant Normal Operating Conditions 313-6 31 3-6 Thermal Environment Conditions inside Radwaste Building, Plant Normal Operating Conditions 31 3-7 31 3-7 Thermal Environment Conditions Inside Service Building, 7
'f Plant Normal Operating Conditions 313-8 31 3-11 Amendment 6 a-_______--_______-
23Abl00AE Standard Plant arv. n SECTION 31.3 TABLES (Continued)
Table I 1112 East 31 3-8 Thermal Environment Conditions in Outdoor Area, Plant Normal Operating Conditions 3I3-9 31 3-9 Radiation Environment Conditions Inside Primary Containment Vessel, Plant Normal Operating Conditions 313-10 313-10 Radiation Environment Conditions Inside Reactor Building (Secondary Containment), Plant Normal Operating Conditions 313-11 !
313-11 Radiation Environment Conditions Inside Reactor Building (Outside Secondary Containment), Plant Normal Operating Conditions 313-12 31 3-12 Radiation Emironment Conditions Inside Turbine Buildia.g, Plant Normal Operating Conditions 313-13 31 3-13 Radiation Environment Conditions inside Service Building, Plant Normal Operating Conditions 313-14 313-14 Thermal Environment Conditions Inside Primary Containment Vessel, Plant Abnormal / Accident Conditions 31 3-15 313-15 Thermal Environment Conditions Inside Reactor Building (Secondary Containment), Plant Abnormal / Accident Conditions 313-16 313-16 Thermal Environment Conditions Inside Reactor Building (Outside Secondary Containment), Plant Abnormal / Accident Conditions 31 3-19 313-17 Thermal Environment Conditions Inside Turbine Building, Plant Abnormal / Accident Conditions 313-20 313-18 Thermal Environment Conditions Inside Control Building, Plant Abnormal / Accident Conditions 313-21 313-19 Radiation Environment Conditions Inside Primary Containment Vessel, Plant Abnormal / Accident Conditions 313-22 0
313 iii Amendment 6 l l
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MAMMAE Standard Plant ,
nnv n SECTION 3I3 '
TABLES (Continued)
Table Iltle P_ ass 313-20 Radiation Environment Conditions Inside Reactor Building (Secondary Containment), Plant Abnormal / Accident Conditions - 31 3-23 313-21 Radiation Environment Conditions laside Control Building, Plant Abnormal /Acciden: Conditions 313-24 1
')
LO 313-iv Amendment 6 l
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23A6100AE Standard Plant _ nev n 3I3 ENVIRONMENTAL CONDITIONS PARAMETERS
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GE PROPRIETARY - provided under separate cover O (Includes pages 3I3-1 through 3I.3-24)
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Amendment 6 3III
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! '. i ABM 23A6100AB R#andant Plant MV. A 4A.3.3.3 Regions of the Power Flow Map 4AJJ.5 Flow Control I
[b, . j Region! This region defines the system The normal plant startup procedure requires I operational capability with the the startup of all RIPS first and maintain at I reactor internal pumps ru.:ni ; at their minimum pump speed (30% of rated), at j their minimum speed (30%). Power which point reactor heatup and pressurization {
changes, during normal startup and can commence. When operating pressure has been i shutdown, will be in this region. The established, reactor power can be increase,'
normal operating procedure is to start This power flow increase will follow a line up along curve 1. within Region I of the flow control map shown in i Figure 4.4-1. The system is then brought to the Region II This is the low power area of the desired power-flow level within the normal oper-operating map where the carryover ating area of the map (Region IV) by increasing through steam separators is expected the RIP speeds and by withdrawing control rods.
to exceed the acceptable value.
Operation within this region is Control rod withdrawal with constant pump precluded by system interlocks. speed will result in power / flow changes along lines of constant pump speed (Curves 1 through Region HI This is the high power / low flow area 8). Change of pump speeds with constant control of the operating map which the system rod position will result in power / flow changes j is the least damped. Operation within along, or nearly parallel to, the rated flow !
this region is precluded by SCRRI control line (curves A through F).
(Selected Control Rods Run-In).
4AJA Temperature Power Operating Map (PWR)
RegionIV This represents the normal operating f zone of the map where power changes Not applicable can be made, by either control rod
(' movement or by core flow changes, 4A3.5 Lead.Following Characteristics through the change of the pemp speeds.
See Appendix A, Subsution A.4.4.3.5 of f 4AJ.3A Design Features for Power-Flow Reference 1.
Control 4AJ.6 Thermal and Hydraulle Characteristics The following limits and design features are Summary Table employed to maintain power flow conditions to the required values shown in Figure 4.4-1: The thermal hydraulic characteristics are provided in Table 4.4-1 for the core and tables (1) Minimum Power Limits at Intermediate and of Section 5.4 for other portions of the reactor High Core Flows: To prevent unacceptable coolant system.
separator performance, the recirculation system is provided with an interlock to 4.4.4 Evaluation reduce the RIP speed.
See Appendix A, Subsection A.4.4.4 of
' (2) Pump Minimum Speed Limit: The Reactor Reference 1.
Internal Pumps (RIPS) are equipped with Anti Rotation Devices (ARD) which prevent a 4.4.5 Testing and Verification tripped RIP from rotating backwards. The ARD begins operating at 300 rpm decreasing See Appendix A, Subsection A.4.4.5 of speed. In order to prevent mechanical wear Reference 1.
in the ARD, minimum speed is specified at 300 rpm. However, to provide a stable Chapter 16 includes the requirement that core 7 operation, the minimum pump speed is set at flow will be checked at least once every 24 (d 450 rpm (30% of required). hours to detect flow reduction.
4.4-3 1
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L i i
23A6100AB Standard Plant anv. c
, 4.4.6 Instrumentation Requirements See Appendix A, Subsection A.4.4.6 of 4
Reference 1.- A loose parts monitoring system is not employed in ABWR due to the existance of considerable successful experience without one in operating BWRs and because the normal operational testing will detect the existance of loose parts that might affect CRD performance and because operationalinstrumentation will detect any flow blockage. The impact of a loose parts monitoring system on plant operation and construction is not considered warranted when compared to its benefit.
4.4.7 References
- 1. General Electric Standard Application for j.. Reactor Fuel, (NEDE 24011, latest approved i revision).
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O Amendment 6 4.44
ABWR mum.
senadard Plant an. s d SECTION S.2 CONTENTS (Continued)
Section M East 533.2.23 Sources ofImpurities 5.29 5.23.2.2.4 IGSCC Considerations 5.212 j 5.23.23' Compatibility of Construedon Materials with Reactor Coolant 1 2-12 l 5.23.2.4 Compatibility of Construction
- Materials with External Insuladon 5.2-12 5.233 Fabrication and Processing of Ferride Materials 5.2 12 5.2 12 5.233.1 Fracture Tou6hness
-5.233.2 Control of Welding 5.2 13 l
(
5.233.2.1 Regulatory Guide 1.50: Control of Prebeat Temperature Employed for Welding of 1ow-Alloy Steel 5.2 13 l 5.233.2.2 Regulatory Guide 134: Control Electroslag Weld Properdes 5.2 13 5.233.23 Regulatory Guide 1.71: Welder Qualification for Areas of 1.imited Acceuibility 5.213 5.2333 Regulatory Guide 1.66:
Nondestructive Examination of Tubular Products 5.2 13 5.233.4 Moisture Control for 1.cw Hydrogen, Covered Arc Welding Electrodes 5.2-14 l 5.23.4 Fabrication and Proceuing of Austenitic Stainless Steels 5.2 14 l 5.23.4.1 Avoidance of Stress / Corrosion Cracking 5.2 14 l 5.23.4.1.1 Avoidance of Significant Sensitizati on 5.2 14 1 3.2iv Amendment 2
1 k 23A6100AB ;
Standard Plant REV.C !
SECTION 5.2 )
CONTENTS (Continued)'
Section Iltle P_ase i
i 5.23.4.1.2 Process Controls to Minimize j Exposure to Contaminants 5.2-14 5.23.4.13 Cold Worked Austenitic Stainless ~l Stcels 5.2-15 l
l 5.23.4.2 Control of Welding 5.2-15 j l
5.23.4.2.1 Avoidance of Hot Cracking 5.2-15 )
I
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5.23.4.2.2 Regulatory Guide 134:
Electroslag Welds 5.2-15 l 5.23.4.23 Regulatory Guide 1.71: Welder Qualification for. Areas of 1.imited Accessibility 5.2-15 f4 5.23.43 Regulatory Guide 1.66: Non-destructive Examination of Tubular Products 5.2-15a 5.2.4 In-Service Insnection and 7estine of Reactor Coolant Pressure Boundarv 5.2-15a 5.2.4.1 System Boundary Subject to Inspection 5.2-15a 5.2.4.2 Provisions for Access to the Reactor Coolant Pressure Boundary 5.2-15a 5.2.4.2.1 Design and Arrangement of Reactor Coolant Boundary Components 5.2-15a 5.2,4.2.2 Reactor Pressure Vessel 5.2-15a 5.2.4.23 Pipe, Pumps, and Valves 5.2-16 5.2.43 Examination Techniques and Procedures 5.2-16 5.2.43.1 Equipment for Inservice Inspection 5.2-16 5.2.43.2 Visual Examination 5.2-16a 5.2 v O
Amendment 6 l
_ _ _ _ _ _ _ _ __ l
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. ABM 234siocxs j me=Ard Plant REv. c )
i3 or less in thickness. . impact testing is boundary are fabricated from carbon steel l
.V J' performed in accordance with NB 2332 for. materials.
thicknesses of 21/2 inches or less. The materials comply with Appendix G, Section Preheat temperature employed for welding of G 3100 of ASME Code Section 111. Iow alloy steel meet or exceed the recommenda-i tions of ASME Code Section Ill, Subsection NA.
,h' -
'(2) Materials for bolting with nomin'ai diameters Components are either held for an extended time exceeding one inch are required to meet both at preheat temperature to assure removal of hy.
the 25 mils lateral expansion specified in drogen, or preheat is maintained until post-weld
'!' NB 2333 and the 45 ft lb Charpy V value spe- heat treatment. The minimum preheat and maximum cified in 10CFR$0, Appendix G. The 45 ft lb interpass temperatures are specificd and j
. requirement stems from the ASME Code where monitored.
g it applies to bolts over 4 inches in' diame-g ter, starting Summer 1973 Addenda. Prior to this, the Code referred to only 2 sizes of ,
bolts (11 inch and > 1 inch). GE continued-the two size categories, and added the 45; ft lb as a more conservative requirement. ;
(3) The reactor vessel complies with the requi- ]
rements of NB-2331. The reference tempera-ture (RTNDT) is _ established for-all required pressure retaining materials used
-in the construction of Class 1 vessels.
f~ This includes plates, forgings, weld q
'( .
material, and heat affected zone. The RTNDT differs from the nil-ductility , l temperature (NDT) in that in addition to All welds were nondestructively examined by passing the drop test, three Charpy V-Notch radiographic methods. In addition, a supple-specimens (traverse) must exhibit 50 ft.lb mental ultrasonic examination was performed.
absorbed energy and 35 mil lateral expansion at 600F above the RTNDT. The core 5.23J.2.2 Regulatory Guide 134:. Control of l beltline material must meet 75 ft lb Electroslag Weld Properties absorbed upper shelf energy.
No electroslag welding is performed on BWR (4) Calibration of instrument and equipment components.
3 shall meet the requirements of the ASME A Code, Section Ill, paragraph NB 2360. 5.233.23 Regulatory Guide 1.71: Welder Qualification for Areas of Limited Accessibility 5.233.2 Control of Welding Welder qualification for areas of limited
.5.233.2.1 Regulatory Guide l.50: Control of accessibility is discussed in Subsection Preheat Temperature Employed for Welding of 5.2.3.4.2.3.
Law AlloySteel 5.233.3 Regulatory Guide 1.66: Nondestruc-Regulatory Guide 1.50 delineates preheat tem- tive Examination of Tubular Prmlucts perature control requirements and welding procc-dure qualifications supplementing those in ASME. Regulatory Guide 1.66 describes a method of Sections ill and IX. implementing requirements acceptable to NRC re-garding nondestructive examination requirements
- bO The use of low-alloy steel is restricted to of tubular products used in RCPB. This Regula-the reactor pressure vessel. Other ferritic tory Guide was withdrawn on September 28,1977, components in the reactor coolant pressure by the NRC because the additional requirements Amendment 6 5.2 13
.Z__._ _ _ . _ _ _ _ _ _ _ _ _ . _ _ . _ _
ABM zwious StandardPlant arv s imposed by the guide were satisfied by the AGME with the guide lines of NUREG 0313, to avoid Code. significant sensitization. l Wrought tubular products were supplied in Process control are exercised during all accordance with applicable ASTM /ASME material stages of component manufacturing and construe.
specifications. Additionally, the specification tion to minimize contaminants. Cleanhaess con-for the tubular products used for CRD bousings trols are applied prior to any eleva.ted temper-specified ultrasonic examination to paragraph ature treatment. For applications where stain.
NB 2550 of ASME Code Section 111. less steel surfaces are exposed to water at tem-peratures above 200'F low carbon (<0.03%)
These RCPB components meet 10CFR50 Appendix B grade matcrials are used. For critical applica-requirements and the ASME Code requirements thus tions, nuclear grade saaterials (carbon content assuring adequate control of quality for the .s. 0.02%) are used. All materials are supplied products. in the solution heat treated condition. Special ,
sensitization tests are applied to assure that ;';
5.233.4 Moisture Control for Low Hydrogen, the material is in the annealed condition. A Covertd Arc Welding Electrudes During fabrication, any heating operation All low hydrogen covered welding electrodes (except welding) between 800 1800'F are are stored in controlled storage areas and only avoided, unless followed by solution heat authorized persons are permitted to release and treatment. During welding, heat input is distribute electrodes. Electrodes are received controlled. The interpass temperature is also in hermetically sealed etnisters. After removal controlled. Where practical, shop welds are from the sealed containers, electrodes which are solution heat treated. In general, weld filler not immediately used are placed in storage ovens material used for austenitic stainless steel which are maintained at about 2500F (generally base metals is Type 308L/3161/3091 with an 2000F minimum). average of 8% (of Fn) ferrite content. <
j Electrodes are distributed from scaled con- 5.23.4.1.2 Process Controls to Minimize j tainers or ovens as required. At the end of each Erposure to Contaminants l work shift, unused electrodes are returned to the storage ovens. Electrodes which are damaged, Exposure to contaminants capable of causing wet, or contaminated are discarded. If any stress / corrosion cracking of austenitic stain- <
electrodes are inadvertently left out of the less steel components was avoided by carefully I ovens for more than one shift, they are discard- controlling all cleaning and processing mate-ed or reconditioned in accordance with rials which contact the stainless steel during manufacturer instructions. manufacture, construction, and instrilation.
5.23.4 Fabrication and Processing of Special care was exercised to insure removal ,
Aurtenitic Stainless Steels of surface contaminants prior to any beating operations. Water quality for cleaning, 5.23.4.1 Avoidance of Stress / Corrosion rinsing, flushing, and testing was controlled Cracking and monitored. Suitable protective packaging was provided for components to maintain 533.4.1.1 Avoidance of SigniDeast cleanliness during shipping and storage.
.. Sensitizarlon The degree of surface cleanliness obtained by When austenitic stainless steels are heated in these procedures meets Ibc requirements of l ,
the temperature range 800' 1800 F, they Regulatory Guides 1.37 and 1.44. l
! ; are considered to become " sensitized" or i N susceptible to intergranular corrosion. The ABWR f design complies with Regulatory Guide 1.44 and l
)
Arneodmeru 1 5.2 14
_ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ l
MN ae" -d Plant asAstorAs nev s SJJMJ Cold Werhad Amstenitic Stataless high alloy steels or other materials such as Sessis static and centrifugal castings and binetallic joints should comply with fabrication require-Cold work controls are applied for components meats of Sections III and IX of the ASME Boiler made of austenitic stainless steel. These mate- and Pressure Vessel Code. It also requires rials are used la the cast condition. During additional performance qualifications for fabrication cold work is controlled by applying weldingin arcas oflimited access.
limits is hardness, bend radii and surface finish en ground surfaces. All ASME Section III welds are fabricated in accordance with the requirements of Sections III
$2JM Centrol of Welding and IX of the ASME Boiler and Pressure Vessel Code. There are few restrictive welds involved 5JJ.4.2.1 Aveldamee of Hot Crocidag in the fabrication of BWR components. Welder qualification for welds with the most restric-Regulatory Guide 131 describes the acceptable tive access is accomplished by mockup welding, method of implementing requirements with regard Mock up is examined sectioning and radiography :::
to the control of welding when fabricating and (orUT). y joining austenitic stainless steel components and systems.
The Acceptance Criterion !!.3.b.(3) of SRP Section 5.2.3 is based on Regulatory Guide Written welding procedures which are approved 1.71. The AEWR design meets the intent of this ey GE are required for all primary pressure boun- regulatory guide by utilizing the alternate dary welds. These procedures comply with the approach as follows: '
requirements of Sections 111 and IX of the ASME Boiler Pressure Vessel Code and applicable NRC When access to a non volumetrically examined Regulatory Guides. ASME Section !!! production weld (1) is less than 12 inches in any direction and (2) allows
(- All austenitic stainless steel weld filler welding from one access direction only, such materials were required by specification to have weld and repairs to welds in wrought and cast a minimum delta ferrite content of 5 FN (ferrite low alloy steels, austenitic stainless steels number) determined on undiluted weld pads by and high nickel alloys and in any combination of usagnetic measuring instruments calibrated in these materials shall comply with the fabrica-accordance with AWS specification A4.2 74. tion requirements specified in ASME Boiler and Pressure Vessel Code Section !!! and with the Delta ferrite measurements are not made on requirements of Section IX invoked by Section qualification welds. Both the ASME Boiler and Ill, supplemented by the following requirements:
Pressure Vessel Code and Regulatory Guide 131 specify that ferrite measurements be performed on (1) The welder performance qualification test undiluted weld filler material pads when magnetic assembly required by ASME Section IX shall instruments are used. There are no requirements be welded under simulated access condi- =
for ferrite measurement on qualification welds. tions. An acceptable test assembly will E provide both a Section IX welder 5.23M.2 Regulatoe7 Guide 1J4: Electroslag performance qualification required by this Welds Regulatory guide.
Electroslag welding was not employed for If the test assembly weld is to be judged reactor coolant pressure boundary componento by bend tests, a test specimen shall be removed from the location least favorable 5.13.4.23 Regulatory Guide 1.71: Welder for the welder. If ahis test specimeo QuellAcation or Artas ofIJmited Accessibility caanot be remoyed froa a Ioeatioo prescribed by Section IX, an additional l Regulatory Guide 1.71 requires that weld bend test specimen will be required. If j fabrication and repair for wrought low alloy and the test assembly weld is to be judged by Amenemm 2 12.u
ABM '='gne isaa>anas me nev a .
radiography or UT, the length of weld to be system, or connected to the reactor coolant I I
esamined shall include Ihe location least systems, up to and including: l }
favorable for the melder.
(1) The outermost containment isolation valve {
Records of the results obtained in welder in system piping that penentrates the I accessibility qualification shall be as primary rcactor ea='aia=*=r certklied by the manufacturer or installer,-
shall be maintained and shall be made (2) The second of two valves normally closed 1 accessible to anthoriand personnel. during normal reactor operation in system piping that does not penestrate primary l Socket weld with a 2 la. sominal pipe size reactor e= tan ==ent.
and under are escluded from the above requirements. (3) The reactor coolant system and relief valves,
- (2) (a) For accessibility, when more restricted y access conditions than qualified will 5.2.4J Provisions for Acesss to the Rasctor obscure the welder's line of sight to Coolant Pressure Boundary the catent that production welding will require the use of visual aids such as 5.2A.2.1 Design and Arrangement of Reactor mirrors. The tequalification test as- Coolant Pressure Boundary Components sembly shall be welded under the more restricted access conditions using the Accessibility in accordance with ASME Code '
a visual aid required for production Section XI, fWA 1500 is provided as described in welding. the following paragraphs. E i
(b) GE complies with ASME Section IX. 5.2A.2.2 Reactor Pressure Vessel (3) Surveillance of accessibility qualification Access for examination of the RPV has been l I requirements will be performed along with provided through provisions incorporated into normal surveillance of ASME Section IX the design of the vessel, shield wall, and performance qualification requirements. vesselinsulation as follows:
5J3A.3 Regalatory Golde 1.66: (1) The shield wall and vessel insulation be-Nondestructive h=1==elos of Tubular Products hind the shield wall are spaced away from I the RPV outside surface. Access ports are For discussion of compliance with Regulatory located at each reactor pressure vessel Guide 1.66, see Subsection 5.2333. nozzle. The annular space between the {
reactor vessel outside surface and insula-5.2.4 Inservice Inspection and Testing tion inside surface permits insertion of of Reactor Coolant Pressure Boundary remotely operated ultrasonic devices for examination of vessel longitudinal and This subsection discusses the itservice in- circumferential welds. . Access for i spection and testing program for the NRC Ouality insertion of the automated devices is Group A components;i.e., ASME Boiler and Pres- provided through removable insulalation sure Vessel Code Section III, Class 1, compo- panels at the top of the shield wall.
seats. It will show how the program meets requires, ats of hr'L= Xi of thc ASME Code. (2) Access to the reactor pressure circum-ferential, longitudinal, and nozzle-5.2A.1 System Boundary Subject to laspection to vessel welds above the shield wall is provided through use of removable The system boundary subject to inspection insulation panels. Either manual or j includes all pressure vessels, piping, pumps, and valves which are part of the reactor coolant automated examination methods may be employed. { }
Anwadowat 2 5.2 15s
ABM us6toorn Standard Plant nev c
()
c (3) The vessel flange area and vessel closure 5.2.4.23 Pipe, Pumps, and Valves l
head can be examined during normal refueling using manual ultrasonic methods. With the Physical arrangement of piping, pumps, and closure head removed, access is provided to valves provide personnel access to each weld 4 the upper interior portion of the vessel by location. Working platforms are provided at I removal of the steam dryer and steam areas to facilitate servicing of pumps and separator assemblies and core flooder valves. Platforms and ladders are provided for sparger. Removal of these components also access to piping welds including the enables examination of the remaining pipe-to-reactor vessel nozzle welds. Removable internal reactor pressure vessel components thermal insulation is provided on welds and utilizing remote visual techniques. The components which require frequent access for examination of the flange-to-vessel weld can examination or are located in high radiation be performed manually from the flange seal areas.
surface and/or the RPV outer surface.
Welds are located to permi ultrasonic (4) The closure head is dry stored during examination from at least one sije, but where refueling. Removable insulation will permit component geometries permit, acc:ss from both manual examination of all welds on the sides is provided. Consideration will be given vessel head from the outside surface. The to weld joint configurations and surfaces during nuts and washers are dry stored during fabrication to permit thorough ultrasonic refueling and may be examined at that time. examinations. Consideration will also be given All RPV studs are accessible for required to minimizing the number of welds during examination during refueling either in place fabrication.
or when removed.
5.2.43 Examination Techniques and Procedures 6
) (5) Openings in the RPV support pedestal and V removable insulation panels around the Examination techniques and procedures, cylindrical lower portion of the vessel including any special techniques and procedures, provide access for manual or automated will be written in accordance with the ultrasonic mrJhods for examination of the requirements of Table IWB 2500 of ASME Code RPV welds within the support skirt. Section XI. Wherever possible, automated, computerized examinations will be performed with SRP Seetion 5.2.4, 11.1 states the the data stored in computerized format for easy inspection requirements of 10 CFR Part 50 retnieval and comparison.
para. 50.55a as detailed in Section XI of the code must be met for all class 1 5.2.43.1 Equipment for Inservice Inspection pressure-containing components (and their supports). Certain areas of the vessel Whenever possible, mechanized automated will not be completely accessible for inspection is planned with manual supplement for examination. The nozzcl-to-nozzel welds the preoperational examination of the welds in will be partially inaccessible for reactor pressure vessel top and bottom heads volumetric examination due to their including flange-to vessel weld. Remote configuration. ( N o t e t h a t 5.2.4.2.2 ultrasonic scanning will be used to examine the addresses only examination of the vessel cir c u m f e r en tiaI,1ongit u dinal, a nd from the outside surface. If the vessel nozzle-to vessel welds on the balance of the were examined from the inside surface, the vessel.
nozzel to nozzel welds could be made accessible.) Welds in the bottom head will Remote visual examination techniques, where also be partially inaccessible for utilized, will provide resolution capabilities volumetric examination due to CRD at least equivalent to that obtained by direct p penetrations. visual observations.
j Amendmem 6 5.2-16
r-- - - -
!~
23A6100AB Standard Plant _
REV.C h
5.2AJ.2 Visual Examination Visual examination shall be conducted with
- lighting good enough to resolve a 1/32 inch black line on a neutral gray card.
Where direct manual operations are required without the use of mirrors or other viewing aids, clearance (at least 24 inch clear space) is provided where feasible for the head and shoulders of a man within a working arm's length
. (20 inches) of the surface to be examined.
At locations where leakages are normally expected are collected (for example, valve stems, J
pump seals), the. examination shall verify that the leakage collection system is operative.
Piping runs are clearly identified and laid 3 out so that insulation damage, leaks, and j structural distress will be evident to a trained observer walking the line.
Ol I
I l
i l
i Amendment 6 5.216a
i 23A6100AB L. si ; E!ansmAa d Plant Rev.c
- (] SECTION 5A CONTENTS (Continued)
Section Thlt East 5.4.6.2.5.1 . Standby Mode 5.4-15 5.4.6.2.5.2 Emergency Mode 5.4-15 5.4.6.2.53 Test Mode 5.4-16 5.4.6.2.5.4 Limiting Single Failure 5.4-16 5.4.63 Performance Evaluation 5.4-16 5.4.6.4 PreoperationalTesting 5.4-16 5.4.7 Residual Heat Removal System 5.4-16 5.4.7.1 . Design Bases . 5.4-16 5.4.7.1.1- Functional Design Basis 5.4-16 5.4-17 f~~ 5.4.7.1.1.1 Low Pressure Flooder (LPFL) Mode 5.4.7.1.1.2 Test Mode 5.4 5.4.7.1.13 Minimum Flow Mode 5.4-17 5.4.7.1.1.4 Standby Mode 5.4-17 5.4.7.1.1.5 Suppression Pool Cooling 5.4-17 5.4.7.1.1.6 Wetwell and Drywell Spray Cooling 5.4-18 5.4.7.1.1.7 Shutdown Cooling ~5.4-18 5.4.7.1.1.8 ' Fuel Pool Cooling 5.4-18 5.4.7.1.1.9 Reactor Well and Equipment Pool Drain 5.4-18 5.4.7.1.2 Design Basis for Isolation of RHR System from Reactor Coolant System 5.4-18 j I
5.4.7.13 Design Basis for Pressure Relief Capacity 5.4-18a p
5.4.7.1.4 Design Basis with Respect to GeneralDesign Criterion 5 5.4-19 l l
5.4-iv Amendment 6
l- MM 23A6100AB b Standard Plant REV.A l
SECTION 5.4
! CONTENTS (Continued) hl Section Illit East 5.4.7.1.5 Design Basis for Reliability and Operability 5.4-19 5.4.7.1.6 Design Basis for Protection from Physical Damage 5.4-19 5.4.7.2 Systems Design 5.4-19 5.4.7.2.1 System Diagrams 5.4-19 5.4.7.2.2 Equipment and Component Description 5.4-19 5.4.7.23 Controls and Instrumentation 5.4-20 5.4.7.23.1 Interlocks 5.4-21 5.4.7.2.4 Applicable Codes and Classification 5.4-21 5.4.7.2.5 Reliability Considerations 5.4-21 5.4.7.2.6 Manual Action 5.4-21 5.4.73 Performance Evaluation 5.4-23 5.4.73.1 Shutdown with All Components Available 5.4-23 5.4.73.2 Worst Case Transient 5.4-23 5.4.733 Emergency Shutdown Cooling 5.4-24 5.4.73.4 Normal Shutdown Cooling 5.4-24 5.4.7.4 Pre-operational Testing 5.4-24 5.4.8 Reactor Water Cleanun System 5.4-24 5.4.8.1 Design Bases 5.4-25 5.4.8.2 System Description 5.4-25 5.4.83 System Evaluation 5.4-26 5.4.9 Main Steamlines and Feedwater Ploine 5.4-27 5.4.9.1 Safety Design Bases 5.4-27 9
5.4-v
)
i 23A6100AD REV C Standard Plant i crated within design limits has an extremely low 5.4.3 Reactor Coolant Piping probability of failure caused by any known ,
failure mechanism. GE purchase specifications Since the RIPS are located inside the RPV, require that integrity of the pump motor case be there is no major external reactor coolant maintained through all transients and that the piping connected to the ABWR pressure vessel, pump remain operable through all normal and upset transients. The design of the pump and motor 5.4.4 Main Steamline Flow Restrictors bearings are required to be such that dynamic load capability at rated operating conditions is 5.4.4.1 Safety Design Bases not exceeded during the safe shutdown earthquake.
The main steamline flow restrictors were de-Pump overspeed will not occur during the signed to:
course of a loss of-coolant accident (LOCA) due to a anti rotation device (ARD) which is located (1) Limit the loss of coolant from the reactor at the bottom of the RIP motor and prevents a vessel following a steam line rupture backward rotation of the RIP. The ARD also outside the containment to the extent that prevents backward rotation during normal plant the reactor vessel water level remains high operation when one RIP is stopped and the other enough to provide cooling within the time re-RIPS are operating. quired to close the main steam line isola-tion valves; A failure modes and effects analysis of RIP is presented in Appendix 15B. (2) Withstand the maximum pressure difference ex-pected across the restrictor following com-S.4.1.6 Inspection and Testing plete severance of a main steamline; Quality control methods are used during fabri- (3) Limit the amount of radiological rclease cation and assembly of the reactor recirculation outside of the drywell prior to MSIV system to assure that design specifications are closure; and met. Inspection and testing procedures are de-scribed in Chapter 3. The reactor recirculation (4) Provide trip signals for MSIV closure.
system is thoroughly cleaned and flushed before fuel is loaded initially. 5.4.4.2 Description During the pre-operational test program, the A main steamline flow restrictor (Figure reactor recirculation system is hydrostatically 5.4-6) is provided for each of the four main tested at 125% reactor vessel design pressure. steamlines by giving the inside bore of each RPV Preoperational tests on the reactor recirculation steam outlet nozzle the shape of a flow restrict-system also include checking operation of the ing venturi.
pumps, flow control system, and valves and is dis-cussed in Chapter 14. The restrictor limits the coolant blowdown rate from the reactor vessel in the event a main During the startup test program, horizontal steamline break occurs outside the containment 6 and vertical motions of the RIP motor casing is to the maximum (choke) flow of 83 x 10 lb/hr observed; supports are adjusted as necessary to at 1025 psig upstream pressure. The flow assure that the motor casings are free to rnove as restrictor is designed and fabricated in accor-designed. Nuclear system responses to dance with ASME Code, Fluid Meters.
recirculation pump trips at rated temperatures and pressure are evaluated during the startup The flow restrictor has no moving parts, its tests and plant power response to recirculation mechanical structure can withstand the flow control is determined. velocities and forces associated with a main 5.4.2 Steam Generators (PWR)
Not applicable to this BWR.
54-5 Amendment 6
l f
M 23A6100AD Standard Plant REV.A J
steamline break. The maximum differential restrictor material because it has excellent re-pressure is conservatively assumed to be 1375 sistance to erosion / corrosion in a high velocity psi, the reactor vessel ASME Code limit pressure. steam atmosphere. The excellent performance of stainless steel in high velocity steam appears ,
The ratio of venturi throat diameter to to be due to its resistance to corrosion. A pro- I steamline inside diameter of approximately 0.5 tective surface film forms on the stainless results in a maximum pressure differential steel which prevents any surface attack and this (unrecovered pressure) of about 14 psi at 100% of film is not removed by the steam.
rated flow. This design limits the steam flow in a severed line to less than 200% rated flow, yet Hardness has no significant effect on it results in negligible increase in steam mois- erosion / corrosion. For example hardened carbon ture content during normal operation. The steel or alloy steel will erode rapidly in appli-restrictor is also used to measure steam flow to cations where soft stainless steel is unaf-1 initiate closure of the main steamline isolation fected.
valves when the steam flow exceeds preselected op- l erational limits. The vessel dome pressure and Surface finish has a minor effect on the venturi throat pressure are used as the high erosion / corrosion. If very rough surfaces are and low flow sensing locations. exposed, the protruding ridges or points will erode more rapidly than a smooth surface. Expe-5.4.4.3 Safety Evaluation rience shows that a machined or a ground surface is sufficiently smooth and that no detrimental In the event a main steamline should break erosion will occur.
outside the containment the critical flow phenom-enon would restrict the steam flow rate in the 5.4.4.4 Inspection and Testing venturi throat to 200% of the rated value. Prior to isolation valve closure, the total coolant Because the flow restrictor forms a perma-lo:ses from the vessel are not sufficient to nent part of the RPV steam outlet nozzle and has cause core uncovering and the core is thus ad- no moving components, no testing program beyond equately cooled at all times. the RPV inservice inspection is planned. Very slow erosion which occurs with time, has been ac-Analysis of the steamline rupture accident counted for in the ASME,Section III design (Subsection 15.6.4) shows that the core remains analysis. Stainless steel resistance to erosion covered with water and that the amount of radioac- has been substantiated by turbine inspections at tive materials released to the environs through the Dresden Unit 1 facility. These inspections the main steamline break does not exceed the have revealed no noticeable effects from erosion guideline values of published regulations. on the stainless steel nozzle partitions. The Dresden inlet velocities are about 300 ft/sec Tests on a scale model determined final design and the exit velocities are 600 to 900 ft/sec.
and performance characteristics of the flow However, calculations show that, even if the restrictor. The characteristics include maximum erosion rates are as high as 0.004 in. per year, flow rate of the restrictor corresponding to the after 40 years of operation, the increase in accident conditions, unrecoverable losses under restrictor-choked flow rate would be no more I normal plant operating conditions, and discharge than 5%. A 5% increase in the radiological dose moisture level. The tests showed that flow calculated for the postulated main steamline restriction at critical throat velocities is break accident is insignificant.
stable and predictable.
5A.5 Main SteamlineIsolation System The steam flow restrictor is exposed to steam of about 2/10% moisture flowing at velocities of 5.4.5.1 Safety Design Bases 150 ft/sec (steam piping ID) to 600 ft/sec (steam restrictor throat). ASTM A351 Type 304 cast The main steamline isolation valves, indi-stainless steel was selected for the steam flow vidually or collectively, shall:
9 144
MN.
se==d==vl Plant 23A6100AB' REV.A system, each valve is tested as discussed in may cause the nuclear SRVs to open briefly, but Chapter 14. j the rise in fuel cladding temperature will be in-significant. No fuel damage results. Two isolation valves provide redundancy in each steamline so either can perform the isola- I The ability of this Y pattern globe valve to tion function and either can be tested for l close in a few seconds after a steamline break, leakage after the other is closed. The inside under conditions of high pressure differentials valve, the outside valve, and the respective and fluid flows with fluid mixtures ranging from . control systems are separated physically, mostly steam to mostly water, has been demon-strated in a series of dynamic tests. A The isolation valve is analyzed and tested full size,20 inch valve was tested in a range of for earthquake loading. The loading caused by steam-water blowdown conditions simulating pot.tu- the specified earthquake loading is required to lated accident conditions (Reference 1). be within allowable stress limits and with no malfunctions that would prevent the valve from The following specified hydrostatic, leakage, closing as required.
and stroking tests, as a minimum, are performed by the valve manufacturer in shop usts: Electrical equipment that is associated that is associated with the isolation valves and oper.
(1)- To. verify its capability to close at set- ated in an accident environment is limited to tings between 3 and 10 sec (response time the wiring, solenoid valves, and position for full closure is set prior to plant op- switches on the isolation valves. The expected eration at '3.0.sec minimum, 5.0 see pressure and temperature transients following an maximum), each valve is tested at rated pres- accident are discussed in Chapter 15.
sure (1000 psig) and no flow. The valve is
/^ stroked several times and the closing time 5.4.5.4 Inspection and Testing is recorded.
The main steam isolation valves can be func.
(2) Leakage is measured with the valve seated. tionally tested for operability during plant op-The specified maximum seat leakage, using cration and refueling outages. The test provi-water at design pressure, is 2 sions are listed below. During refueling outage colp cm /hr/in, of nominal valve size. In addi- the main steam isolation valves can be function-tion, an air seat leakage test is conducted ally tested, leak tested, and visually inspect-using 50 psi pressure upstream. Maximum per- ed.
missible leakage is 0.1 scfh/in of nominal valve size. The main steamline isolation valves can also be tested and exercised individually to the (3) Each valve is hydrostatically tested in ac- fully closed position in the slow closing mode cordance with the requirements of the appli- if reactor power is reduced sufficiently to cable edition and addenda of the ASME Code. avoid scram from reactor pressure core flux or During valve fabrication, extensive high flow through the other steamline flow nondestructive tests and examinations are restrictors.
conducted. Tests include radiographic, liquid penetrant, or magnetic particle ex. Leakage past the valve disk backseat and aminations of casting, forgings, welds, piston rings is collected and measured by the hardfacings, and bolts, drywell drain system. During shutdown while the nuclear system is pressurized, the leak rate (4) The large diameter piston inside the valve through the piston rings can be measured by col-body (See Figure 5.4-7) assures correct lecting and timing the leakage.
alignment of the actuating components.
The leak through the pipeline valve seats After the valves are installed in the nuclear can be measured accurately during shutdown by l
I 5.4-9
i l
l ABM 234stoors l Standard Plant REV C the following suggested procedure: enough period to obtain meaningful data. An l alternate means of leak testing the outer (1) With the reactor at approximately 1250F and isolation valve is to utilize the previously normal water level and decay heat being re- noted steamline plug and to determine leak-moved by the RHR system in the shutdown age by pressure decay or by inflow of the cooling mode, all main steam isolation valves test medium to maintain the specific test are closed utilizing both spring force and pressure.
air pressure on the operating cylinder.
During pre-startup tests following an exten-
!?) Nitrogen is introduced into the reactor sive shutdown, the valves will receive the same vessel above normal water level and into the hydro tests that are imposed on the primary connecting main steamlines and pressure is system.
raised to 20 30 psig. An alternate means of pressurizing the upstream side of the inside Such a test and leakage measurement program isolation valve is to utilize a steamline ensures that the valves are operating correctly.
plug capable of accepting the 20 to 30 psig pressure acting in a direction opposite the 5.4.6 Reactor Core Isolation Cooling System hydrostatic pressure of the fully flooded reactor vessel. Evaluations of the reactor core isolation ,
cooling system against the applicable General De- g t3) A pressure gage and flow meter are connected sign Criteria are provided in Subsection 3.1.2.
- to the test tap between each set of main steam isolation valves. Pressure is held 5.4.6.1 Design Basis below 1 psig, and flow out of the space between each set of valves is measured to The reactor core isolation cooling (RCIC) establish the leak rate of the inside isola- system is a safety system which consists of a tion valve. turbine, pump, piping, valves, accessories, and instrumentation designed to assure that suffi.
(4) To leak check the outer isolation valve, the cient reactor water inventory is maintained in reactor and connecting steamlines are flooded the reactor vessel to permit adequate core cool-to a water level that gives a hydrostatic ing to take place. This prevents reactor fuel head at the inlet to the inner isolation overheating during the following conditions:
valves slightly higner than the pneumatic test pressure to be applied between the (1) a loss-of coolant (LOCA) event; valves. This assures essentially zero leakage through the inner valves, if neces- (2) vessel isolated and maintained at hot sary to achieve the desired water pressure at standby; the inlet to the inner isolation valves, gas from a suitable pneumatic supply is intro- (3) vessel isolated and accompanied by loss of duced into the reactor vessel top head. Ni- coolant flow from the reactor feedwater trogen pressure (20 to 30 psig) is then system; applied to the space between the isolation valves. The pistons are checked for leak (4) complete plant shutdown with loss of normal tightness. Once any detectable piston ring feedwater before the reactor is depressur-leakage to the drain system has been ac- ized to a level where the shutdown coolant counted for, the seat leakage test is con- system can be placed in operation; or ducted by shutting off the pressurizing gas and observing any pressure decay. The volume (5) loss of AC power for 30 minutes.
between the closed valves is accurately known. Correction for temperature variation Acceptence criteria 11.3 of SRP Section 5.4.6 during the test period are made,if necessary, states that the RCIC system must perform its to obtain the required accuracy. Pressure function without the availability of any a c and temperature are recorded over a long power. Review Procedure 111.7 further requires Amendment 6 5 4-10
' i , ib cf .
, y.
s ,
' L:);
'* DWR . .
mei-
. StamA=4 Pi==* - nev. c
- 1 '
1 "that there be sufficient battery capability for
- two hours'.~of ' operation. , While RCIC is designed, i ,;.m ~
for.30 minu;es. of operation during loss-of ac
- power, the' batt'ery_ capacity 'should allow over x , four hours of operation, which would meet this -
~
n ,
requirement.
Following a reactor scram, steam generation '.
%g .
will continue at a reduced rate due to the core h .;
i fission product decay heat.. At this time the turbine bypass system will divert the steam to.
4 5
\
\ c._
l'F:
l Anwedment 6 5.410e
h MN 2sA61oors e - %: n.., m l shutdown condition using safety grade . normal plant operation, the pump is started and e -- equipment as well as the non-safety the return line to the suppression pool is L functions associated with refueling and opened. A reverse sequence is used to terminate b servicing operations. this test. Upon receipt of an automatic'initi-ation signal while in the flow testing mode, the
. The RHR system has nine different operational system is returned to automatic control.
configuraticas that are discussed separately to provide clarity. 5A.7.1.1.3 Minimens Flow Mode 5A.7.1.1.1 Imw Pressure Flooder (LPFL) Mode : If the aain discharge flow reaches a predetermined low value, the minimum flow valve Each loop '.n the low pressure flooder- la that loop will automatically open to provide subsystem provides core cooling water supply to some pump flow. During this mode, water is compensate for water loss beyond the normal pumped from the suppression pool and returned to control range from any cause up to and including tw su,apression pool via the low flow bypass the design basis (LOCA). This subsystem is line. Sufficient main discharge flow will cause initiated automatically by a low water level in the minimum flow valve to close automatically, the reactor vessel or high pressure in the drywell. Each loop in the system can also be 5A.7.1.tA Standby Mode placed in operation by means of a manual initiation pushbutton switch. During normal plant operation, the RHR loops are in a. standby condition with the During the low pressure flooder mode, water is motor. operated valves in the normally open or pumped from the suppression pool through the RHR normally closed position. The valves on the heat exchangers, and diverted through the minimum suppression pool suction line are open and the Q flow lines until the injection valve in the minimum flow valves are open; the test valves -
- U discharge line'is signalled to open on low _ and injection valves are closed. The RHR pumps reactor. pressure. The RHR shall be capable of are not running while the standby water leg delivering flow into the reactor vessel within 36 pumps are running to keep the pump discharge seconds. after receipt of the low pressure- lines filled. The relief valves in the pump permissive signal following system initiation. suction and pump discharge lines protect the
- This assumes a one second delay for the lines against overpressure.
Instrumentation to detect the low pressure permissive and generate an initiation signal to SA.7.1.1.5 Suppression Pool Cooling the injection valve. Consequently, the 36-second RHR requirement is consistent with the 37 second The suppression pool cooling subsystem pro-
[
injection time . assumed in' LOCA analyses. vides means to remove heat released into the Additionally, the time for the pumps to reach suppression pool, as necessary, following heat
,1 rated speed, from the receipt of at least one additions to the pool. During this mode of actuation signal, is 27 seconds. operation, water is pumped from the suppression pool through the RHR heat exchangers, and back The system remains in this mode until manually to the suppression pool. This subsystem is stopped by the operator. initiated automatically by a high suppression pool temperature signal or a high wetwell air-5A.7.1.1.2 Test Mode space pressure signal if no LOCA signal is present. If both a LOCA signal and a pool cool-Full flow functional tests of the RHR can be ing initiation signal are present, the RHR will performed during normal plant operation or during function in the low pressure flooder mode. It plant shutdown by manual operation of the RHR can also be manually initiated for testing and from the control room. For plant testing during other purposes.
O
$A47
ABM a34620048 nry c 1
j Standard Plant 1
5.4.7.1.1.6 Wetwell and Drywell Spray Cooling 2-out.of-4 logic, which has four independent pressure sensor and transmitter inputs. The l Two of the RHR loops provide containment independence is provided by each being in a j spray cooling subsystems. Each subsystem separate instrument division. Furthermore, the i provides both wetwell and drywell spray cooling. inboard and outboard valves of a common shutdown This subsystem providec steam condensation and cooling suction line are operated by different primary containment atmospheric cooling following electrical divisions.
a LOCA by pumping water from the suppression )
pool, through the heat exchangers and into the 5.4.7.1.1J Fuel Pool Cooling wetwell and drywell spray spargers in the primary containment. The drywell spray mode is initiated Two of the RHR loops provide supplemental f by operator action post-LOCA in the presence of fuel pool cooling during normal refueling j high drywell pressure. The wetwell spray mode activities and any time the fuel pool heat load j actuates automatically whenever the RHR is exceeds the cooling capacity of the fuel pool J operated in the suppression pool cooling mode. heat errhangert,. For normal refueling activities 1 These modes are terminated automatically by a where the reactor well is flooded and the fuel LOCA signal, or manually by operator action. pool gates are open, water is drawn from the reactor shutdown suction lines, pumped through 5.4.7.1.1.7 Shutdown Cooling the RHR heat exchangers and discharged through the fuel pool distribution spargers. For 100%
The shutdown cooling subsystem is manually core removal, if necessary, water is drawn from activated by the operator following insertion of the fuel pool cooling system (FPCS) skimmer surge i the control rods and normal Howdown to the main tanks, pumped through the RHR heat exchangers and l condenser. In this mode, the RHR removes returned to the fuel via the FPCS coc, ling lines.
residual heat (decay and sensible) from the These operations are initiated and shut down by reactor vessel water at a rate sufficient to cool operator action. The FPCS is automatically it to 1250F within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after the control isolated from the RHR when:
rods are inserted. The subsystem can maintain or reduce this temperature further so that the (1) The safety function of the RHR is required.
reactor can be refueled and serviced.
(2) The shutdown cooling subsystem is in Reactor water is cooled by pumping it directly operation.
from the reactor shutdown cooling nozzles, through the RHR heat exchangers, and back to the 5.4.7.1.1.9 Reactor Well and Equipment Pool vessel (via feedwater on one loop and via the Drain core cooling system on the other two loops).
The RHR pravides routing and connections for This subsystem is initiated and ii ~ n by emptying the reactor well and equipment pool to operator action. the suppression pool after servicing. Water is pumped or drained by gravity through the FPCS The Branch Technical Position RSB 5-1, section return lines to the RHR shutdown suction lines B.I.(b) and (c), of the RHR Standard Review Plan, and then to the suppression pool.
SRP 5.4.7, requires the RHR suction side isolation valves to have independent diverse 5.4.7.1.2 Design Basis for isolation oI RHR interlocks io prevent the valves from being System from Reactor Coolant System opened unless the reactor coolant system (RCS) pressure is below the RHR system design The low pressure portions of the RHR system pressure. While the ABWR RHR design does not are isolated from full reactor pressure whenever explicitly meet this requirement for diversity, the primary system pressure is above the RHR it does meet the intent of the requirement to system design pressure. (See Subsection provide high reliability against inadvertent 5.4.7.1.3 for further details.) In addition, opening of the valves. The pressure signal that automatic isolation may occur for reasons of provides the interlock function is supplied from vessei water inventory retention which are l l
Amendment 6 5 4-1B l
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'i 1
MM 23A6tooAB REV.C j
,. !handard Pinnt '
.(y_
l k unrelated to line pressure rating.' (Subsection 5.2.5 provides an explanation of the' leak detection system and.the isolation signals.)~
F The RHR pumps are protected against damage from' a closed discharge valve by mean_s of automatic minimum flow valves which open on low mainline flow and close on high mainline flow..
- 5.4.7.13 Design Basis for Pressure Relief '
Capacity -
The relief' valves in the RHR system are sized on one of three bases:
(1) thermal relief only, 4 (2) valve bypassleakage only, or
'(3) control. valve failure and the subsequent uncontrolled flow which results.-
v( W i,
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l- Amendment 6 5.418a l
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l ABM 23A6100AD mem-A= 4 Plant m !
,Qr3 SECTION 6.2 1 CONTENTS (Continued Section Dile P_ age 6.2.6.1.2.2 Acceptance Criteria 6.2-41 6.2.6.1.23 Test Frequency 6.2-41 6.2.6.13 Additional Criteria for Integrated Leakage Rate Test 6.2-41 6.2.6.2 Containment Penetration Leakage Rate Test (Type B) 6.2-42 6.2.6.2.1 General 6.2-42 6.2.6.2.2 Acceptance Criteria 6.2-42 6.2.6.23 Retest Frequency 6.2-42 6.2.6.2.4 Oc6fgn Provisions for Periodic Pressurization 6.2-42 6.2.63 Containment Isolation Vaive leakage Rate Test (Type C) 6.2-43 6.2.63.1 General 6.2-43 6.2.63.2 Acceptance Criteria 6.2-43 6.2.6.4 Scheduling and Reporting of Periodic Tests 6.2-43 6.2.6.5 SpecialTesting Requirements 6.2-43 6.2.7 References 6.2-44 l
O 6.2-viii
ABM 234 1oo4.
nrv. c Standard Plant i
TABLES Table 31 tin P.ast $
6.2-1 Containment Parameters 6.2-45 6.22 Containment Design Parameters 6.2-46 l
6.2-2a Engineered Safety Systems Information for Containment Response Analyses 6.2-46a M 6.2-2b Net Positive Suction Head (NPSH) Available I to RHR Pumps 6.2 46c 6.2-2c Net Positive Suction Head (NPSH) Available to HPCF Pumps 6.2-46d 6.2-2d Secondary Containment Design and Performance Data 6.2-46e 6.2-3 Compartment Nodal Description 6.2-47 6.2-4 Compartment Vent Path Description 6.2-48 6.25 Reactor Coolant Pressure Boundary (RCPB)
Influent Lines Penetrating Drywell 6.2-49 6.2-6 Reactor Coolant Pressure Boundary (RCPB)
Effluent Lines Penetrating Drywell 6.2-50 ILLUSTRATIONS Figure Ittle P_ age 6.2-1 A Break in a Feedwater Line 6.2-51 6.22 Feedwater Line Break--RPV Side Break Area 6.2-52 6.23 Feedsater Line Break Flow -Feedwater System Side of Break 6.2 53 6.2-4 Feedwater Line Break Flow Enthalpy--
Feedwater System Side of Break 6.2-54 6.25 Lower Drywell Air Tramfer Percentage for Model Assumption Versus Actual Case 6.2 55 6.2-6 Pressure Response of the Pri nary Containment 6.2-56 e
Amendment 6 L_______-_ _ _ _ .
j
-AB M 2a63oo48
- me.=a.ed piant arv. c
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SECTION 6.2 ILLUSTRATIONS (Continued)
Figure Title Eage 6.2-7 Temperature Response of the Primary Containment 6.2-57
-6.2-8 Temperature Time History After a Feedwater Line Break 6.2-58 6.2-9 ABWR Main Steamlines with a Break 6.2-59 6.2-ixa Amendswn 6
ABM 23xsioorn REV C Standard Plant
) Subsection 5.4.7.4 for further discussion of initial suppression pool temperature and the RHR
'- preoperational testing.) service water temperature are at their maximum values. This assumption maximizes the heat sink i 6.2.2.3 Design Evaluation or the Containment temperature to which the containment beat is Cooling System rejected and thus maximizes the containment temperature. In addition, the RHR heat 6.2.2.3.1 System Operation sad Sequence of exchanger is assumed to be in a fully fouled Events condition at the time the accident occurs. This conservatively minimizes the heat exchanger heat In the event of the postulated LOCA, the removal capacity. Even with the degraded short term energy release from the reactor pri- conditions outlined above, the maximum mary system will be dumped to the suppression temperature is maintained below the design limit pool. Subsequent to the accident, fission pro- specified in Subsection 6.2.2.1.
duct decay heat will result in a continuing en-crgy input to the pool. The RHR LPFL rnode and It should be noted that, when evaluating this suppression pool cooling mode will remove this long-term suppression pool transient, all heat entegy which is released into the primary contain. sources in the containment are considered with ment system, thus resulting in acceptable sup- no credit taken for any heat losses other than pression pool temperatures and containment through the RHR heat exchanger. These heat pressures. sources are discussed to Subsection 6.2.1.3.
In order to evaluate the adequacy of the RHR lt can be concluded that the conservative system, the following is assumed: evaluation procedure d: scribed above clearly demonstrates that the RHR system in the (1) With the reactor initially operating at 102% suppression pool cooling mode limits the
,' ] of rated power, a LOCA occurs. post-LOCA containment temperature transient.
L i (2) A single failure of a RHR beat exchanger is 6.2.2.4 Test and Inspections the most limiting single failure.
The containment cooling system is required (3) The ECCS flows assumed available are 2 HPCF, to have scheduled maintenance. The system 1 RCIC, and 2 LPFL (RHR). testing and inspection will be performed
( periodically during the plant normal operation g (4) Containment cooling is initiated after 10 and after each plant shutdown. Functional 6 minutes. (See Response to Question 430.26) testing will be performed on all active ,
O components and controls. The system reference Analysis of the net positive suction head characteristics will be established during g f. (NPSH) available to the RHR and HPCF pumps in preoperational testing to be used as base points g5* accordance with the recommendations of Regulatory for checking measurements obtained from the Guide 1.1 is provided in Table 6.2 2b. system tests during the plant operation.
General compliance for Regulatory Guide 1.26 The preoperational test program of the may be found in Subsection 3.2.2. containment cooling system is described in Subsection 14.2.12. The following functional tests will be performed. The RHR pump will be tested through the suppression pool cooling loop operation by measuring flow and pressure. Each 6.2.2.3.2 Summary of Containment Cooling pump will be tested individually.
Analysis j Containment spray spargers will be tested !
g When calculating the long term, post LOCA pool during reactor shutdown by air, and by visual
- ; temperature transient, it is assumed that the inspection to verify that all the nozzles are v
Amendmem 6 62 17 l
ABM u4swes.
Standard Plant arv c clear. RHR but exchangers will b; checked for source of radioactive release after an effectiveness by measuring inlet and outlet accident. During normal plant operation, the temperatures at the tube and shell sides. secondary containment areas are kept at a negative pressure with respect to the All motor and air operated valves required environment and ,-lean zone by the HVAC system.
for safety are capable of being exercised med Following an accident, the standby gas treatment ,
their operation demonstrateci. The layout and system (STGS) provides this function. These arrangement of critical equipment outside the systems are described in subsections 9.4.5 and drywell is designed to permit access for 6.5.1, respectively.
appropriate equipment used in testing and inspecting system integrity. Relief valves on Fission products that may leak from the the low pressure lines are removable for testing. primary to secondary are processed by the SGTS before being discharged to the environment. The Periodic inspection and maintenance of the HVAC exhaust systems and SGTS are located within main system pumps, pump motors, and heat the secondary containmeist to assure collection exchangers are conducted in accordance with t!n of any leakage. The se:condary containment manufacturer's instructions. provides detection of the level of radioactivity released to the environment during abnormal and During the normal plant operation, the pumps, accident plant conditions. Personnel or heat exchangers, valves, piping, instrumentation, material entrances to the: secondary containment wiring and other components outside the consist of airlocks with interlocked doors or containment can be inspected visually at any hatches.
time. Testing frequencies are generally correlated with testing frequencies of the There are basically three types of potential associated controls and instrumentation. When a leakage paths for the release of fission product pump or valve control is tested, the operability during and following an accident. These leakage of that pump or valve and its associated paths are shown in Figure 6.2 27, instrumentation is tested by the same action.
When a system is tested, operation of the 6.23.1 Design Bases components is indicated by installed instrumentation. Relief valves are removed as (1) Secondary containment is provided to scheduled at refueling; outages for benet tests collect fission products which may leak and setting adjustment. from the primary containment following a DBA. This collection allows filtration by 6.2.2.5 Instrumentation Requirements the SGTS prior to reIease to the environment. The secondary containment Details of the instrumentation are provided in region completely surrounds the primary Section 7.3. Tbc suppression pool cooling mode containment vessel.
of the RHR system is automatically initiated. j (2) During a DBA, the secondary containment and 6.2.3 Secondary Containtnent Functional supporting systems such as the SGTS, is Design designed to limit the thyroid and whole body doses to less than 10CFR100 guidelines The secondary containment boundary, as shown at the site boundary and low population in Figure 6.2 26, completely surrounds the zone and to less than 10CFR50, Appendix A, )
primary containment vessel (PCV) except for the General Design Criterion 19 doses for the j basemat and together with clean zoce comprises control room operator. J the reactor t uilding. The secondary containment 1 encloses all penetrations through the primary (3) The mechanical, electrical, instrumen- i containment and all those systems external to the tation, and structural components of the i primary containment that may become a potential secondary containment design are protected .
as necessary from internally and O\
Amendment 5 6218 l
_ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ . 6
i MM 23A6100AB nry A Standard Plant y
) externally generated missiles, dynamic zone or the environment are automatically effects associated with pipe whip and jet isolated during an accident. These systems forces, and environmental conditions from an cannot be automatically initiated following accident, and are designed to Seismic Cate- an accident.
gory I requirements. These i! cms and equip-ment of supporting systems required to func- (9) The exhaust unit of the secondary tion after an accident are designed for sin- containment HVAC maintains the secondary gle active failure, loss of offsite power containment air flow pattern from areas of coincident with an accident,30-day accident normally low contamination to those with a duration for radiological analysis and 100- potentially high level.
day duration for operational capability. No credit is assumed for non-seismic category I (10) High energy pipe brea)u within the piping, power supplies and equipment. secondary containment are detected by the leak detection and isolation system.
(4) The design allowable inleakage at the Blowout panels are provided where necessary secondary containment-environs boundary is to relieve the thermal and pressure buildup within the capability of the SGTS to in the various subcompartments.
maintain the pressure inside secondary containment at -1/4-in. water gage relative (11) All effluents from the secondary to the environs, under design exterior wind containment areas are monitored for gamma conditions. This prevents exfiltration such radiation leve! prior to being released to that 10CFR100 guidelines will not be the environment, exceeded following a DBA.
(12) Adequate instrumentation and control room Post-accident pressure transients do not indications are provided to monitor all
[3 cause 10CFR100 guidelines to be exceeded important secondary containment parameters in onder to maintain the plant within the k) because of exfiltration, technical specification limits and provide (5) Automatic shutoff of the normal HVAC air information for operator actions.
supply and ventilation exhaust and of other systems after a LOCA or on detection of high (13) The secondary containment, in conjunction radiation in effluent prevents airborne with supporting systems, will be leakage from escaping the secondary periodically tested to assure that i co r.t ain m e n t. performance requirements can be met.
(6) All openings through the secondary 6.23.2 System Design containment boundary, such as personnel and equipment doors, remain closed after a DBA The secondary containment is a reinforced by interlocks or administrative control. concrete building that forms an envelope These doors are provided with position surrounding the PCV above the basernat. The indicators and alarms having readout and secondary containment has isolation systems on annunciation capability in the control room. piping, doors and other penetrations. This permits maintaining the secondary containment (7) Liquid leakage from the se:ondary volume at a slightly negative pressure so all containment to the clean zone or to the PCV leakage can be collected and treated before environment is controlled by means of water release. Details of structural design and loop seals, automatic shutoff valves in arrangement of compartments for various systems series, or piping upgrade to safety class. are described in Section 3.8.
(8) All operating systems that transport liquid The boundary of the secondary containment is from the secondary containment to the clean shown in the following figures:
(o) a 6.2 19
ABWR zus>=^n aw e Standard Plant 6.2-28 Containment Boundaries in the Reactor fuel storage pools, RWCU, FP/SPCU and other Building Plan Section A-A (0-180 ) potentially radioactive sources in the secondary containment. The HVAC exhaust systems and SGTS 6.2-29 Containment Boundaries in the Reactor are a1so 1oeated within tbe secondary Building-Plan Section B B (90 -270 ) containment to assure collection of any leakage. The RHR and HPCS pump seals and valve 6.2 30 Containment Boundaries in the Reactor packings and RCIC system components are a Building-Plan at El (-) 13200 mm poter.lial source of radioactive release and are located within the secondary containment.
6.2-31 Containment Boundaries in the Reactor Building-Plan at El (-) 6700 mm During refueling operations, the drywell head j is removed and the secondary containment becomes a 6.2-32 Containment Boundaries in the Reactor the containment envelope. Therefore, entry iv.o . I I
Building-Plan at El (-) 200 mni the secondary containment is provided via double doo- vestibules, or, in the case of the main 6.2-33 Containment Boundr. ries in the Reactor equipment batch, a double door entry. This Building-Plan at E17300 mm assures the integrity of the secondary 1 containment envelope with effluent monitoring j 6.2-34 Containment Boundaries in the Reactor and treatment of airborne radioactive material Building-Plan at E113100 mm resulting f om normal plant or refueling i t
operations or from abnormal events such as a 6.2-35 Containment Boundaries in the Reactor fuel drop accident.
Building-Plan at El 18500 mm The airborne fission product is contained by 6.2-36 Containment uoun daries in the Reactor maintaining all portions of Ihe seccndary Building-Plan at E126700 =m containment at a negative 1/4 in. of water gage relative to the lowest pressure boundary outside 8 Secondary containment design and performance the secondary containment. This negative
& data is provided in Table 6.2-2d. pressure is achieved following an accident by '
the SGTS.
During normal operation, the secondary I containment system is operated at a slightly The airborne fission product leakage from the negative pressure relative to the atmosphere. primary containment is processed by the SGTS.
This assures that any leakage from the primary The SGTS achieves a 99.99% removal of halogen containment will be collected and can be treated (stable and radioactive) and a 99.9% of airborne before release if its radioactivity level is particulate prior to discharge to the above prescribed limits. The secondary environment. This removal efficiency will be 1 containment HVAC system operates on a feed and periodically tested in accordance with bleed principle with internal recirculation. Air regulatory requirements. The dose limit flow is from clean to potentially contaminated evaluation takes credit for 99% airborne halogen areas, and particulate for this type of leakage. A ,
99% removal credit is allowed even though the The building effluents are monitored for design will achieve 99.99% removal capability.
radioactivity by the area radiation monitoring system. If the radioactive level rises above set The SGTS will maintain the secondary contain-levels, the secondary containment discharge can ment air flow pattern from potentially low to be routed through the SGTS for treatment before high level contaminated areas. The potentially '
release. The operation of the secondary high level contaminated areas are the following:
containment SGTS and HVAC are discussed in more detail in Subsection 6.5.1 and Chapter 9, (1) RWCU System Rooms respectively.
(2) RCIC System Room i During normal operation, the secondary f containment is the envelope that forces )
collection of airborne radioactive material from l
t Amendment 6 6.2 20 1 \
i l
____________________a
ABM me.=Aad Plant 22461 min arv. c
- a p TABLE 6.2 2b ij
^
NET POSITIVE SUCTION HEAD (NPSH) AVAILABLE TO RHR PU3MPS . '
r A. Suppression pool is at its minimum depth, El. -3740mm ( 12.27 Ft).
, B. Centerline of pump suction is at El. -7200mm (23.62 Ft).
C. Suppressgo poolgater is at its maximum temperatue for the given operating mode,97 C (207 F).
D. Pressure is atmospheric above the suppression pool.
E. Maximum suction strainer losses are 2.0 psi.
NPSH = HATM + "A '"VAP ""F where:
HATM = atmospheric head H = static head M S p; H VAP = vapor pressure head k Hp = Frictional head including strainer Minimurn Expected NPSH 3
RHR Pump Runout is 1100 m /h (4843 gpm).
Maximum suppression pool temperature is 97 C (207 F)
H = 10.73m (35.20 Ft)
ATM H =
3.46m (1135 Ft) 3 H VAP * * *I' I)
Hp =
1.82m (5.97 Ft)
Strainer head loss = 2.0 psi = 1.46m = 4.80 Ft NPSH available = 10.73 + 3.46 - 9.74 - 1.82 = 2.63m (8.63 Ft)
NPSH required - 2.4m (7.87 Ft)
Amendment 3 6.2-46c
- p. - - -
r ABWR ur6 oars Standard Plant nev. c h ; TABLE 6.2 2c NET POSITIVE SUCTION HEAD (NPSH) AVAILABLE TO HPCF PUMPS ,
L A. Suppression poolis at its minimum depth, El. 3740mm ( 12.27 Ft).
B. ~ Centerline of pump suction
- is at El. -7200mm (23.62 Ft). !
C. Suppression pool water is at its maximum temperatue for the given operating
'*~
mode,97 C (207"F). .
1 D. Pressure is atmospheric above the suppression pocl.
I' E. Maximum suction strainer losses are 2.0 psi.
NPSH = HATM + HA -HVAP Hp where:
HATM = atmospheric head H3 = static head g HVAp = vapor pressure head Hp = Frictional head including strainer MmimmnfugtSkt.d.hl?.SB HPCF Pump Runout is 890 m3 /h (3918 gpm).
Maximum suppression pool temperature is 97 C (207 F)
=
HATM 10.73m (35.20 Ft)
=
H3 3.46m (11.35 Ft)
=
HVAP 9.74m (31.95 FL)
Hp =
1.82m (5.97 Ft)
]
Strainer head loss = 2.0 psi = 1.46m = 4.80 Ft NPSH available = 10.73 + 3.46 9.74 - 1.82 = 2.63m (8.63 Ft) i i
1 NPSH required = 2.2m (7.22 Ft)
NPSH Reference Point O
Amendment 6 6.2-46d
l 1J 23A6100AB Standard Plant ne.. c
.q
/T! ).
Table 6.2-2d SECONDARY CONTAINMENT DESIGN AND PERFORMANCE DATA DESCRIPTION UNIT VALUE A. Secondary Containment ps 3 6
- 1. Free Volume ft 3.0 x 10 1
- 2. Pressure, inches of water, gauge inch H O (..)0.25 2
- 3. Leak Rate at Postaccident Pressure %/ day
(*of Secondary Containment Free Volume)
- 4. Exhaust Fans Number - 2 Type - Centrifugal 85
- 5. Filters $
)3 (a) Basicspecification
.t Number of filter train -- 1
+ Type - Dust (b) Component specification f3 (1) Prefilter Number of set - 1 Type- - Dry (2)- HEPA Glters Number of set - 2 Type (Material) Glass fiber (3) Charcoal absorber Number of set - 1 Type - Deep bed
/
k Amendment 6 6.2-46e
23A6100AD Standard Plant nev. c Table 6.2 2d SECONDARY CONTAINMENT DESIGN AND PERFORMANCE DATA O
(Continued) l DESCRIPTION UNIT VALUE B. Tramient Analysis
- 1. Initial conditions (a) Primary Containment (1) Pressure psia 15.45 (2) Temperature KF 135.0 (3) Outside air temperature Summer operation: %F 115.0 g KF e Winter operation: (-)40.0 (b) Se,ondary Containment (1) Pressure inch H O (-)0.25 2
(2) Tempert.ture Max value in summer KF 104.0 Min valae in winter %F 50.0
- 2. Thickness of Secondary Containment Wall.
Wall thickness range from inch (11.81 - 59.06) 1
- 3. Thickness of Primary Containment Wall (a) Concrete Wall inch 79.0 (b) Liner Plate inch 0.25 O
Amendment 6 6.246f
23A6100AB Standard Plant ne.. c ,
-( Table 6.2 2d
% ]
SECONDARY CONTAINMENT DESIGN AND PERFORMANCE DATA 1 i
(Continued)
- DESCRIPTION UNIT VALUE C. The.ma! Qaractenstics
- 1. Primary Containment Wall (a) Coefficient of Linear Expandon Concrete Wall in/in HF 0.55x10"5
-5 Liner Plate in/in-%F 0.73x10 (b) Modulus of Elasticity 0 i
Concrete Wall psi 3.41xig Liner Plate psi 27.2x10 g l, 7
(c) Thermal Conductivity 3 Concrete Walt Dtu/hr-ft KF 0.941 l
Liner Plate Bru/hr-ft ..MF 30.91 (d) Thermal Capacitance Concrete Wall Btu /ft -KF 28.84 Liner Plate Bte/ft -KF 53.56
- 2. Secondary Containment Wall 3
(a) Thermal Conductivity Btu /hr-ft -hF 0.941 (b) Thermal Capacitance Btu /ft -KF 28.84 (D
L)
Amendment 6 6.2-46g f
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23A6100AB Standard Plant , nev. c Table 6.2-2d SECONDARY CONTAINMENT DESIGN AND PERFORMANCE DATA (Continued)
DESCRIPTION UNIT VALUE _
- 3. Heat Tramfer Coefficients (a) Primary Containment Atmosphere to Primary 2
Contaimnent Wall Btu /hr ft %F 0.246 (b) Primary Contaimnent Wall to Secondary Containmen; 2
0.614
!4 Atmosphere Blu/hr-ft -%F (c) Secondary Containment Wall to Secondary Containment Atmosphere Btu /ht-ft hF 0.614 (d) Primary Containment Emissivity - 0.95 (e) Seconc'.ary Containment - 0.95 f Emissivity i
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- MM 23A6100AD nev.n Remndmed Plant rw SECTION 6.3 (m )\
CONTENTS (CONTINUED)
Section Title Easc j 633.7.7 Line Breaks Outside Containment 6 3-12 633.7.8 Bounding Peak Peak Cladding Temperature 6 3-12 633.8 LOCA Analysis Conclusions 6 3-13 63.4 Tests and Inspections 6 3-13 63.4.1 ECCS Performance Tests 6 3-13 63.4.2 Reliability Tests and Inspections 6 3-13 63.4.2.1 HPCF Testing 6 3-14 63.4.2.2 ADS Testing 6 3-14 63.4.23 RHR Testing 6 3-14 63.4.2.4 RCIC Testing 63-14 63.5 Instrumentation Reautrements 6 3-14 63.6 References 63 15 O
63-iv Amendown: 2
'M 23A6100AB Standard Plant REV.C SECTION 63 TABLES !
i Table Iltle Page j 1
63-1 Significant Input Variable Used in the {
Loss-of Coolant Accident Analysis 6 3-16
'63-2 Operational Sequence of Emergency Core f
Cooling Systems, Maximum Core Spray Line Break 6 3-19 63-3 Single Failure Evaluation 6 3-20 63-4 Summary of Results of LOCA Analysis 6 3-21 63-5 Key to Figures 63-22 63-6 Plant Variables with Nominal and i Sensitivity Study Values 6 3-23 63-7 MAPLHGR Versus Exposure 6 3-24 63 8 Design Parameters for HPCF System Components63-24a 63-9 Design Parameters for RHR System Components63-24d O
ILLUSTRATIONS Figure litic Eage 63-1 HPCF Process Diagram 6 3-25 63-2 RHR Process Diagram 6 3-26 63-3 RHR Process Diagram 6 3-27 63-4 Vessel Pressure Versus High Pressure Core Spray Flow (per System) Used in LOCA Analysis 63-28 63-5 Vessel Pressure Versus Reactor Core Isolation Cooling Flow Used in LOCA Analysis 6 3-29 63-6 Vessel Pressure Versus Low Pressure Flooder Flow (Per System) Used in ,
LOCA Analysis 6 3-30 !
I 63-7 HPCF P&ID 6 3-31 63-v Amendment 6 I
l
- _ _ _ _ _ _ _ _ i
ABM 23A6100AD RIN. B 9mndard PInnt ECCS pump rooms, when required, protects against r< c onditions is also briefly described.
mass flooding of redundant ECCS pumps.
63.1.2.1 High Pressure Core Flooder Mechanical separation outside the drywell is achieved as follows: The HPCF pumps water through an overhead flooder sparger mounted above the reactor core.
(1) The ECCS shall be separated into three func- Coolant is supplied over the entire range of tional groups: system operation pressures. The primary purpose of HPCF is to maintain reactor vessel inventory after small breaks which do not depressurize the (a) RCIC + 1RHR reactor vessel.
(b) 1 HPCF + 1 RHR 63.1.2.2 Residual Heat Removal (c) 1 HPCF + 1 RHR The RHR system has three independent loops (2) The equipment in each group shall be sepa- and delivers water to the core at relatively low rated from that in the other two groups. In reactor pressures. The primary purpose of the addition, the HPCF and RCIC shall be sepa- RHR is to provide inventory makeup and core 5 rated. cooling during large breaks and to provide con-tainment cooling. Following ADS initiation, RHR (3) Separation barriers shall be contructed provides inventory makeup following a small between the functional groups, as required, break, to assure that environmental disturbances such as fire, pipe rupture, falling objects, 63.1.23 Reactor Cort Isolation Cooling etc., affecting one functional group will not affect the remaining groups. In addi. The RCIC system injects water into a feedwa-O tion, separation barriers shall be provided ter line, using a pump driven by a steam as required to assure that such disturbances turbine. The RCIC steam supply line branches do not affect both the RCIC and the HPCF. off one of the main steam lines leaving the reactor pressure vessel and goes to the RCIC 63.1.1.4 ECCS Environmental Design Basis turbine. Makeup water is supplied from the condensate storage tank (CST) or the wetwell Each ECCS has a safety-related injection / with the preferred source being the CST.
isolation testable check valve located in piping within the drywell, except RCIC and RHR Division 63.1.2.4 Automatic Depressurization System A, which connect to feedwater lines outside the drywell. In addition, the RCIC System has an The ADS utilizes a number of the reactor isolation valve in the drywell portion of its safety / relief valves (SRVs) to reduce reactor steam supply piping. The portions of ECCS piping pressure during small breaks in the event of and equipment located outside the drywell and HPCF failure. When the vessel pressure is re-within the secondary containment are qualified duced to within the capacity of the low pressure for the environmental conditions defined in system, these systems provide inventory makeup Section 3.11. so that acceptable post accident temperatures are maintained.
63.1.2 Summary Descriptions of ECCS 6.3.2 System Design The ECCS injection network comprised of a RCIC system; a HPCF system; and RHR system. These A more detailed description of the individual systems are briefly described here as an introduc- systems including individual design characteris-tion to the more detailed system design descrip- tics of the systems are covered in detail in tions provided in Subsection 6.3.2. The ADS S u b s e c t io n s 6.3.2.1 t h r o u g h 6.3.2.4.
which assists the injection network under 6.3-3 Amendment 2 I-L__________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ __ _ _ _ _
ABM 23A6100AB Standard Plant REV. c The following discussion provides details of the 6.3.2.2.1 High Pnssure Core Flooder (HPCF) combined systems; in particular, those design System features and characteristics which are common to all systems. The HPCF system is composed of two HPCF loops (B and C) injecting water above the core. Each 6.3.2.1 Schematic Piping and lastrumentation of the two loops belongs to a separate division; Diagrams electrical and mechanical separation between the two divisions is complete. Physical separation The P& ids for the ECCS are identified in is also assured by locating each division in a Subsection 6.3.2.2. The process diagrams which different area of the reactor building. The two identify the various operating modes of each loops are both high pressure pumping syster. s system are identified in Subsection 6.3.2.2. (i.e., they are capable of injecting water into I the reactor vessel over the entire operating !
6.3.2.2 Equipment and Component Descriptions pressure range). Rated flow at both high and low pressure is the same for each loop. The The starting signal for the ECCS comes from piping and instrumentation diagram and process I four independent and redundant sensors of drywell diagram are given in Figures 6.3-7 and 6.31 pressure and low reactor water level. The ECCS respectively, is actuated automatically and requires no operator action during the first 30 min following The reference pressure for the operating the accident. A time sequence for starting of performance of the system at high pressure is the systems is provided in Table 6.3-2. the lowest spring (safety) setpoint of the SRVs.
Electric power for operation of the ECCS is Both HPCF divisions take primary suction from ,
from regular AC power sources. Upon loss of the the CST and secondary suction from the l regular power, operation is from onsite emergency suppression pool. In the event CST water level standby AC power sources. Emergency sources have falls below a predetermined setpoint or sufficient capacity so that all ECCS requirements suppression pool water level rises above a are satisfied. Each of the three ECCS functional predetermined setpoint, the pump suction will groups identified in Subsection 6.3.1.1.3(1) has automatically transfer from the CST to the j its own diesel generator emergency power source. suppression pool. Both HPCF system loops have -
Section 8.3 contains a more detailed description suction lines that are separate from RHR loops.
of the power supplies for the ECCS.
The HPCF pumps are located at an elevation Regulatory Guide 1.1 prohibits design reliance which is below the water level in the '
on pressure and/or temperature transients suppression pool. This assures a flooded pump l expected during a LOCA for assuring adequate suction. The motor-operated valve in the i NPSH. The requirements of this regulatory guide suction line from the suppression pool on each are applicable to the HPCF, RCIC and RHR pumps. division is normally closed since primary suction is taken from the CST. This valve The BWR design conservatively assumes 0 psig automatically opens on receipt of either of the containment pressure and maximum expected suction transfer signals noted above. The i temperatures of the pumped fluids. Thus, no suppression pool suction valves on each loop are l reliance is placed on pressure and/or temperature capable of being closed from the control room if ;
transients to assure adequate NPSH. a leak develops in the system piping downstream of the isolation valves. Overpressure protec-Requirements for NPSH are given in Figures tion of the pump suction line is provided by a 8 6.3-1 (HPCF),6.3-2 (RCIC) and 6.3-3 (RHR). Pump relief valve to the pump minimum flow line. $
characteristics curves are given in Figures 6.3-4 .
l (HPCF),6.3 5 (RCIC) and 6.3-6 (RHR). Each of the two high pressure core flooder j l loops discharges water into the core via a l
The design parameters for the HPCF and RHR separate overhead spray sparger. Internal 9 system components are provided in Tables 6.3-8 l g i
! and 6.3-9 respectively.
Amendment 6 6.3-4 1
L______-
MM 23A6100AD Standard Plant REV C
~
I piping connects each sparger to the vessel sufficient to provide the required pump nozzle. cooling. A flow element in the main discharge line measure < system flow rate during LOCA and Each HPCF discharge line to the reactor is test conditi. :s and automatically controls the provided with two isolation valves in series. motor operated globe valve on the bypass line.
One of these valves is a testable check valve located inside the drywell as close as practical The HPCF is designed to operate from normal to the reactor vessel. HPCF injection flow offsite auxiliary power or from emergency diesel causes this valve to open during LOCA conditions; generators if offsite power is not available.
thus, no power is required for valve actuation if normal auxiliary power is lost, the onsite during LOCA. If an HPCF line should break out- power source (diesel generator) for that side the drywell, the check valve in that line division is started. The onsite power source inside the drywell will prevent loss of reactor for any division is capable of carrying all of water outside the drywell. The other isolation the division emergency loads, including the HPCF valve (which is also referred to as HPCF injec- pump and valve motors. Manually operated remote tion valve) is a motor operated gate valve locat- controls for system components (such as HPCF ed outside the drywell and as close as practical pumps, valves, etc.) and diesel generators are to the HPCF discharge line drywell penetration. provided in the plant control room.
This valve is capable of opening with the maximum pressure differential across the valve expected Full flow functional tests of the HPCF can be for any system operating mode including HPCF pump performed during normal plant operation or shutoff head. This valve is normally closed as a during plant shutdown by manual operation of the backup to the inside testable check valve for HPCF from the control room. For plant testing containment integrity purposes. A drain line is during normal plant operation, the pump suction provided between the two valves. A normally open is transferred to the suppression pool, the pump
( i manual isolation valve inside the drywell is is started, and the test discharge line to the Cl provided for HPCF loop maintenance during a plant suppression pool is opened. A reverse sequence refueling or maintenance outage. is used to terminate this test. Upon receipt of an automatic initiation signal while in the flow For each loop, a full flow line is provided testing mode, the system shall be returned to with discharge to the suppression pool to allow automatic control.
for full flow test of the system during normal operation. The valves in these lines are closed 6.3.2.2.2 Automatic Depressurization System during normal operation. A full flow test return (ADS) line is consistent with established BWR practice. There is no Regulatory Guide requiring if the RCIC and HPCF cannot maintain the this feature, but all BWRs have a 100% capacity reactor water level, the ADS, which is test return line, and the Chapter 16 Technical independent of any other ECCS, reduces the Specifications specify periodic full flow system reactor pressure so that flow from the RHR functional tests. There are no specific system operating in the low pressure flooder requirements for testing at runout flow; however, mode enters the reactor vessel in time to cool the system does have this capability. the core and limit fuel cladding temperature.
For each loop, a full flow bypass line is also The ADS employs nuclear system pressure provided to return water to the suppression pool relief valves to relieve high pressure steam to to prevent pump damage due to overheating when the suppression pool. The design number, the injection valves on the main discharge lines location, des eription, operational are closed. The bypass line connects to the main characteristics and evaluation of the pressure discharge line between the main pump and the relief valves are discussed in detail in discharge check valve. A motor-operated globe Subsection 5.2.2. The instrumentation and valve on the bypass line automatically closes controls for ADS are discussed in Subsection (p) when flow in the main discharge line is 7.3.1.1.1.2.
Amendment 6 6.3-5
ABM 2346i004 8 Standard Plant REV.B 63.2.2.3 Reactor Core Isolation Cooling it is fully pressurized. This combination of System (RCIC) systems will provide adequate core cooling until vessel pressure drops to the point at which the The RCIC system consists of a 100% capacity low pressure flooder (LPFL) subsystems of the steam-driven turbine which drives a 100 capacity RHR can be placed in operation.
pump assembly and the turbine and pump accessories. The system also includes piping, During RCIC operation, the suppression pool valves, and instrumentation necessary to acts as the heat sink for steam generated by implement several flow paths. The RCIC steam reactor decay heat. This will result in a rise supply line branches off one of the main steam in pool water temperature. Heat exchangers in lines (leaving the reactor pressure vessel) and the RHR system are used to maintain pool water goes to the RCIC turbine with drainage provision temperature within acceptable limits by cooling to the main condenser. The turbine exhausts to the pool water directly during normal plant the wetwell pool with vacuum breaking operation.
protection. Makeup water is supplied from the CST and the wetwell with the preferred source A design flow functional test of the RCIC being the CST. RCIC pump discharge lines include system may be performed during normal plant the main discharge line to the feedwater line, a operation by drawing suction from the test return line to the wetwell, a minimum flow suppression pool and discharging through a full bypass line to the wetwell, and a cooling water flow test return line back to the suppression supply line to auxiliary equipment. The piping pool. The discharge valve to the vessel remains configuration and instrumentation is shown in closed during the test, and reactor operation Figure 6.3-8. The process diagram is given in remains undisturbed. All components of the RCIC Figure 6.3 2. system are capable of individual functional testing during normal plant operation. Should Following the reactor scram, steam generation an initiation signal occur during test mode in the reactor core will continue at a reduced operation, the pump suction will automatically rate due to the core fission product decay heat. revert to the CST and the flow will be The turbine bypass system will divert the steam automatically directed to the vessel.
to the main condenser, and the feedwater system will supply the makeup water required to maintain 63.2.2.4 Residual Heat Removal System (RHR) reactor vessel inventory.
The RHR is a closed system consisting of In the event the reactor vessel is isolated, three independent pump loops which inject water and the feedwater supply is unavaili.ble, relief into the vessel and/or remove heat from the valves are provided to automatically (or remote reactor core or containment. Each of the pump manually) maintain vessel pressure within loops contains the necessary piping, pumps desirable limits. The water levelin the reactor valves and heat exchangers. The piping and vessel will drop due to continued steam instrumentation diagram and process diagram are generation by decay heat. Upon reaching a given in Figures 6.3-9 and 6.3 3 respectively.
predetermined low level, the RCIC system is In the core cooling mode, each loop draws water initiated automatically. The turbine-driven pump from the suppression pool and injects the water will supply water from the wetwell or from the into the vessel outside the core shroud (via the CST to the reactor vessel. The turbine will be feedwater line on one loop and via the core driven with a portion of the decay heat steam cooling subsystem discharge return line on two from the reactor vessel, and will exhaust to the loops). In the heat removal mode, pump suction wetwell pool. may be taken either from the suppression pool or the reactor pressure vessel. With the pump In the event there is a LOCA event, the RCIC suction being taken from the suppression pool, system in conjunction with the two HPCF systems, the pump discharge within these loops provides a is designed to pump water into the vessel while flow path to the following points: !
O Amendment 2 6.3-6
]
MM 23A6100AB REV C Standard Plant o) v (1) Suppression pool signals unless the associate pump motor starter contacts are closed.
(2) The reactor pressure vessel (via feedwater on one loop and via the core cooling Each loop contains instruments necessary to subsystem return lines on the other two maintain a ready condition, to evaluate loop loops). performance, and to operate the minimum flow valve.
(3) The wetwell and drywell spruy spargers (on two loops only). Each RHR pump discharge line is maintained in a filled condition to minimize the time lag In the shu:down cooling mode, with the pump between a starting signal and initiation of flow suction being taken from the reactor pressure into the reactor vessel and to minimize momentum vessel (via the shutdown cooling lines), the pump forces associated with accelerating fluid into discharge within these loops provides a flow path an empty pipe.
back to the reactor vessel via the core cooling discharge return lines, and feedwater line, or to Each divisim. is provided with a discharge the upper reactor well via the fuel cooling line fill pump, which takes suction from the system (on two loops only). suppression pool suction line. A check valve is located in the discharge line at an elevation With the pump suction being taken from the lower than the suppression pool minimum water skimmer surge tanks of the fuel pool cooling level line to prevent backflow from emptying the system, the pump discharge is returned to the lines into the suppression pool.
fuel pool on two loops only.
Full flow functional tests of the RHR can be Each loop is in a single quadrant of the performed during normal plant operation or ra reactc. buildir:g and rec:ives its electric power during plant shutdown by manual operation of the from a bus separate from those serving tLe other RHR from the control room. For plant testing
(~) two loops. Each bus is supplied from both during normal plant operation, the pump is on-site and off-site power sources. started and the return line to the suppression pool is opened. A revere sequence is used to For each loop, a full flow line is provided terminate this test. Upon receipt of an with discharge to the suppression pool to allow automatic initiation signal while in the flow for full flow test of the system during normal testing mode, the system is returned to operation. The valves in these lines are closed automatic control.
during normal operation.
6.3.2.2.5 ECCS Discharge Line Fill System For each loop, a full flow bypass line is also provided to return water to the suppression pool A requirement of the core cooling systems is to prevent pump damage due to overheating when that cooling water flow to the reactor vessel be the injection valves on the main discharge lines initiated rapidly when the system is called on are closed. The bypass line connects to the main to perform its function. This quick start discharge lines between the main pump and the system characteristic is provided by discharge check valve. A motor-operated globe quick-opening valves, quick-start pumps, and valve on the bypass line automatically closes standby AC power source The lag between the when flow in the mein discharge line is signal to start the pump and the initiation of sufficient to provide the required pump cooling. flow into the RPV can be minimized by keeping A flow element in the main discharge line the core cooling pump discharge lines full.
measures system flow rate during LOCA and test Additionally, if these lines were empty when the conditions and automatically controls the motor systems were called for, the large momentum operated globe valve oa the bypass lines. The forces associated with accelerating fluid into a motor-operated valve does not receive automatic dry pipe could cause physical damage to the Amendment 6 6.3-7
23A6100AB Standard Plant REV.C piping. Therefore, the ECCS discharge line fill as metallic materials, etc., are selected as a system is designed to maintain the pump discharge result of an engineering review and evaluation lines in a filled condition. for compatibility with other materials in the system and the surroundings with concern for Since the ECCS discharge lines are elevated chemical, radiolytic, mechanical and nuclear above the suppression pool, check or stop check effects. Materials used are reviewed and valves are provided near the pumps to prevent evaluated with regard to radiolytic and back flow from emptying the lines into the pyrolytic decomposition and attendant effects on suppression pool. Past experience has shown that safe operation of the ECCS.
these valves will leak slightly, producing a small back flow that will eventually empty the 63.2.5 System Reliability discharge piping. To ensure that this leakage from the discharge lines is replaced and the A single failure analysis shows that no lines are always kept filled, a water leg pump is single failure prevents the starting of the provided for each of the three RHR loops. The ECCS, when required, or the delivery of coolant power supply to these pumps is classified as to the reactor vessel. No individual system of essential when the main ECCS pumps are the ECCS is single failure proof with the deactivated. Indication is provided in the exception of the ADS; hence, it is expected that control room as to whether these pumps are single failures will disable individual systems operating, and alarms indicate low discharge line of the ECCS. The most severe effects of single level. The RCIC loop and the two HPCF loops are failures with respect to loss of equipment occur maintained full by connection to the makeup water if the LOCA occurs in combination with an ECCS (condensate). pipe break coincident with a loss of off site power. The consequences of the most severe 63.23 Applicable Codes and Classifications single failures are shown in Table 6.3-3.
The applicable codes and classification of the 63.2.6 Protection Provisions ECCS are specified in Section 3.2. All piping systems and components (pumps, valves, etc.) for Protection provisions are included in the the ECCS comply with applicable codes, addenda, design of the ECCS. Protection is afforded code cases and errata in effect at the time the against missiles, pipe whip and flooding. Also equipment is procured. The piping and components a counted for in the design are thermal of each ECCS within the containment and out to stresses, loadings vom a LOCA, and seismic the including the pressure retaining injection effects.
valve are Safety Class 1. The remaining pipng and components are Safety Class 2,3, or noncode, The ECCS piping and components located as indicated in Section 3.2 and on the individual outside the drywell are protected from system P&lD. The equipment and piping of the internally and external generated missiles by ECCS are designed to the requirements of Seismic the reinforced concrete structure of the ECCS 1 Category 1. This seismic designation applies to pump rooms. The watertight construction of all structures and equipment essential to the these ECCS pumps rooms also protects the core cooling function. IEEE codes applicable to equipment against flooding.
the controb and power supply are specified in Section 7.1. The ECCS is protected against the effects of pipe whip, which might result from piping 63.2A Materials Specifications and failures up to and including the design basis 1 Compatibility event LOCA. This protection is provided by 1 separation, pipe whip restraints, and energy )
Materials specifications and compatibility for absorbing materials. One of these three methods l the ECCS are presented in Sections 6.1 and 3.2. is applied to provide protection against damage
' Nonmetallic materials such as lubricants, seals, to piping and components of the ECCS which packings, paints and primers, insulation, as well O
Ameridment 6 6.3-8 L_____.._.____._.
ABM 23A6100AB men =dard Plant REV.B
[V to valve motion in the case of the high pressure 633.73 Break Spectrum Calculations system provides a suitably conservative allowance )
l for valves available for this application. In A complete spectrum of postulated break i the case of the low pressure system, the time sizes and locations is considered in the delay for valve motion is such that the pumps are evaluation of ECCS performance. For case of at rated speed prior to the time the vessel reference, a summary of figures presented in pressure reaches the pump shutoff pressure. Subsection 63.3 is shown in Table 6.3-5. 1 i
l The flow delivery rates analyzed in Subsection A summary of results of the break spectrum j 633 can be determined from the head-flow curves calculations is shown in tabular form in Table !
in Figures 6.3-4, 6.3-5 and 6.3 6 and the 6.3-4 and graphically in Figure 6.3 8. {
pressure versus time plots discussed in Conformance to the acceptance criteria (PCT = {
Subsection 6.3.3.7. Simplified piping and 2200 F, local oxidation = 17% and core-wide j instrumentation and process diagrams for the ECCS metal water reaction = 1%) is demonstrated I are referenced in Subsection 6.3.2. The Details of calculations for specific breaks are operational sequence of ECCS for the limiting included in subsequent paragraphs. l case is shown in Table 63-2.
633.7.4 Large Line Breaks inside Containment Operator action is not required, except as a monitoring function, during the short-term Since the ABWR design has no recirculation cooling long-term period following cooling period, the LOCA.
the operator During may need the lines, maximum the maximum feedwater steamline lines break (.903 ft andbreak (1.Qp),
to take action as specified in Subsection 6.2.2.2 the maxigum RHR shutdown suction line break to place the containment cooling system into (.853 ft ) become the large break cases.
operation for some LOCA events. Important output variables from the analyses of f p these events are shown in Figures 63-12 through
( 63.33 Use of Dual Function Components for 6 3-36.
ECCS There variables are:
See Appendix A, Subsection A.6.3.3.5 of Reference 4. (1) core flow as a function of time; 6.3.3.6 Limits om ECCS System Parameters (2) minimum critical power ratio as a function of time; See Appendix A, Subsection A.6.3.3.6 of Reference 4. (3) water level in the fuel channels as a function of time; 6.3.3.7 ECCS Analyses for LOCA (4) water level inside the shroud as a function 6.3.3.7.1 LOCA Analysis Procedures and Input of time; Variables (5) water level outside the shroud as a See Appendix A, Subsection A.6.3.3.7.1 of function of time; Reference 4. The significant input variables used by the LOCA codes are listed in Table 63-1 (6) vessel pressure as a function of time; and Figure 63-11.
(7) flows out of the vessel as a function of 6.33.7.2 Accident Description time; See Appendix A, Subsection A.6.3.3.7.2 of (8) flows into the vessel as a function of Reference 1. time; O
Amendment 2 6.3-11
ABM 234610a4n Standard Plant _ REV.C (6) vessel pressure as a function of time; elevation it is the lowest break on the vessel except for the drainline break. Furthermore, the (7) flows out of the vessel as a function of worst break / failure combination leaves the time; fewest number of ECC systems remaining and no high pressure core flooder systems. LOCA (8} Oows into the vessel as a function of time; analyses using break areas less than the maximum values were also considered. The cases analyzed (9) peak cladding temperature as a function of are indicated on the break spectrum plot (refer time. to Figure 6.3-10). From these results it is clear that the overall most limiting break in A conservative licensing assumption is that terms of minimum transient water level in the all offsite AC power is lost simultaneously with downcomer, is the maximum core flooder line the initiation of the LOCA. As a further conser- break case, vatism, all reactor internal pumps are tripped at the start of LOCA event even though this in 633.7.7 Line Breaks Outside Containment itself is considered to be an accident (See Subsection 15.3.1). The resulting rapid core This group of breaks is characterized by a flow coastdown produces a calculated departure rapid isolation of the break. Since a maximum from nucleate boiling in the hot bundles within steam line break outside the containment pro-the first few seconds of the transient. duces more vessel inventory loss before isola-tion than other breaks in this category, the LOCA analyses using break areas less than the results of this case are bounding for all breaks maximum values were also considered for the in this group. Important variables from these steamline, feedwater line, and RHR shutdown analyses are shown in Figure 6.3-60 through suction line locations. The cases analyzed are 6.3-66.
indicated on the break spectrum plot (refer to Figure 63-10). In general, the largest break at As discussed in Subsection 6.3.3.7.4, the each location is the worst in terms of minimum trip of all reactor internal pumps at the start transient water level in the downcomer. of the LOCA produces a calculated departure from nucleate boiling for all LOCA events. Further-633.7.5 Intermediate Line Breaks Inside more, the high void content in the bundles Containment following a large steamline break produces the ,
earliest times of loss of nucleate boiling for ]
For these cases the maximum RHR/LPFL injection any LOCA event. Thus, the summary of results in '
ft2 line break (0.221 ) was analyzed. Important Table 6.3-4 show that, though the PCTs for all variables from this analyses are shown in Figures break locations are similar, the steamline 63 37 through 63-43. breaks result in higher calculated PCTs and the outside steamline break is the overall most l 633.7.6 Small Line Breaks inside Containment limiting case in terms of the highest calculated PCT.
For these cases the maximum high pressure core flooder line break (0.099 ft 2) and the maximum 633.7.8 Bounding Peak Cladding Temperature l bottom head drain line break (0.0218 ft2 ) were Calculation !
analyzed. Important variables from these analyses are shown in Figure 6.3 44 through Consistent with the SAFER application 6.3 59. The drainline break analysis is also methodology i- Reference 2, the Appendix K peak I bounding for any credible break within the cladding temperatures calculated in the previous reactor internal pump recirculation system and sections must be compared to a statistically j its associated motor bounding and cover. calculated 95% probability value. Table 6.3-6 l presents the significant plant variables which !
As expected, the core flooder line break is were considered in the determination of the 95% ;
the worst break location in terms of minimum probability PCT. Again, since the ABWR LOCA .
I transient water level in the downcomer, in results have a large margin to the acceptance Amendment 6 6.3-12
. , - _ _ _ - _ - _ _ . _ _ - - - . _ . - - . . - _ _ . - . _ - . . _ . = .
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Esandard Plant . REV.C:
N criteria, a ~ conservative' PCT calculation was -
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.This bounding PCT was calculated by varying all .
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LO Amendment 6 6.312a L
a i
ME 23AH00AB Remndard Plant _
n, . c
[ ('
V ' Table 63-8 ;
1 DESIGN PARAMETERS FOR HPCF SYSTEM COMPONENTS 1 (1) Main Pumps (C001)
Number of Pumps 2 Pump Type Centrifugal Drive Unit . Motor How Rate 800 gpm @ 1192 psia reactor pressure L 3200 gpm @ 115 psia reactor pressure .l l
Developed Head 2926 ft. @ 1192 psia reactor pressure 725 ft. @ 115 psia reactor pressure Maximum Runout Flow 3800 gpm @ 247 psia reactor pressure Minimum Bypass Flow 400 gpm E
,q- Water Temperature Range 50H to 2126F NPSH Required 8 ft.
(2) Strainer (D001) location Suppression Pool Size 50% plugged shall meet pump NPSH requirements (3) Restricting Orifice (D002) location Pump discharge line Size Lir:iit pump flow to values specified (4) Condensate Storage Tank 150,000 gal reserve storage for HPCF and RCIC systems combi.aed I
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6.3 24a Amendment 6 l
23A6100AB Standard Plant me ,. e Table 63-8 DESIGN PARAMETERS FOR HPCF SYSTEM COMPONENTS (Continued)
(5) Flow Eternents (FE001) location Pump discharge downstream of minimum flow bypass line Head Loss 20 ft. maximum @ 3200 gpm Accuracy 125% combined element, transmitter and indicator at maxirrum rated flow (6) Core Flooder Sparger Flow Rate 3200 gpm minimum @ 115 psia reactor pressure E
Pressure Drop 260 ft. mad.num @> 3200 gpm 5 (7) Piping and Valves Design Pressures 70 psig - suction and discharge connected to suppression pool 200 psig - pump suction 1575 psig pump discharge Design Temperatures 100%F - condensate tank suction 212%F- pump suction and dhcharge 575%F - discharge to vessel 1
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_ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ - - - - . . - - - . - - J
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s*=ndard Plant n. . c Table 6.3-8 (h,
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DESIGN PARAMETERS FOR HPCF SYSTEM COMPONENTS (Continued)
L (8) . Valve Operation
- Pump Suction Valve, Normally closed, opens on low water Suppression Pool (F001) levelin condensate tank Pump Suction Check Valve, Prevents backflow into suppression pool Supprest, ion Pool (F002)
Pump Su'etion Valve, Normally open, closes on low water
' Condensate Tank (F009) ~ levelin condensate tank Pump Suction Check Valve, Prevents backflow into condensate tank .
Condensate Tank (F010)
Pump Discharge Valve, Normally closed, opens within Reactor Injection Valve (F004) 36 seconds after inidation sigt.al Testable Check Valve, Prevent loss of coolant outside drywell Reactor Injection Line (F005) for line break Maintenance Valve, Normally open, used to isolate system Reactor Injection Line (F006) from teactor for maintenance purposes Pump Test Line Valves (F015, Normally closed, throttle valves used F016) to test system flow at rated and runout conditions Pump Minimum Flow Line Normally closed, opens on signal when Valves (F012, F013) pump is running and no flow through flew meter. Used to protect pump from overheating.
O Amsadment 6 6.3-24c
L- ,
MM 23A6100AB .
Standard Plant g _. Rev S i Table 63-9 1 DESIGN PARAMETERS FOR RHR SYSTEM COMPONENTS (1) Main Pumps (C001)
. Number of Pumps 3 Pump Type Centrifugal I
i Drive Unit Motor Flow Rate 4200 gpm Developed Head 400 ft @ 115 psia reactor pressure Maximum Runout Flow 5000 gpm Maximum Bypass Flow 650 gpm l Minimum Shutoff Head 640 ft.
^
Maximum Pump Brake Horsepower 550 kw Water Temperature Range 40% to 358KF
( NPSH Required 8 ft.
't]"%
(2) Heat Exchangers (B001)
Number of units 3 Seismie , CategoryI design and analysis Types of exchangers Horizontal U Tube /Shell Maximum primary side pressure 500 psig Design Point Function Cooling Post LOCA Containment Amendment 6 6.3 24d
23A6100AB Standard Plant nev. c Table 6.3-9 O
DESIGN PARAMETERS FOR RHR SYSTEM COMPONENTS (Continued)
Primary side (tube side) performance data (1) Flow 4200 gpm (2) Inlet temperature 358HF maximum (3) Allowable pressure drop 10 psi (Max)
(4) Type water Suppression Pool or Reactor Water (5) Fouling factor 0.0005 Secondary side (shell side) performance data (1) Flow 5280 gpm V
(2) Inlet temperature 105HF maximum g (3) Allowable pressure drop 10 psi (Max)
- 14) Type water Reactor Building Coohng Water (5) Fouling factor 0.0005 (3) Strainer (D008)
Location Suppression Pool i
Size Meet pump NPSH requirements when 50% plugged (4) Restricting Orifices location (D003) Vessel return line Size Limit flow to vessel to 4200 gpm location (D002) Suppression pool return lie l
l O
Amendment 6 6.3-24c
ABM 23A6100AB n Standard Plant n,. c
/7 Table 6.3 9 GI g DESIGN PARAMETERS FOR RHR SYSTEM COMPONENTS (Continued)
Size Limit flow during suppression pool cooling to 4200 gpm 14 cation (D004) Fuel pool return line Size Limit flow during fuel pool cooling to 4200 gpm location (D001) Pump minimum flowline Size Limit pump flow through the bypass line to 650 gpm location (D005) Discharge line to wetwell spray K
Size Limit wetwell spray sparger flow to 500 gpm y location (D006) - Discharge line to drywell sparger Size Limit drywell spray sparger flow to 3700 gpm
-. p ~(5) Flow Elements (F6009) lacation Pump discharge line, downstream of heat exchanger bypass return Rated Flow 4200 gpm Headloss 20 ft maximum @ 4200 gpm Accuracy 3.2.5% combined element, transmitter and indicatt at rated flow (6) Vessel Flooder Sparger Flow Rate 4200 gpm Pressure Drop O
.w. cat c 6.3-24f L. __ _ _ _ - - _ _ _ _ _ _ - _ _ -
ABWR uAMMAB Standard Plant - Pru:
Table 63-9 DESIGN PARAME'i"ERS FOR RHR SYSTEM COMPONENTS O
(Continued) l i
)
(7) Wetwell Spray Sparger )
l Flow Rate 500 rpm i Pressure Drop 94 ft. @ 500 gpm (8) Drywell Spray Sparger Flow Rate 3700 gpm Pressure Drop 94 ft. @ 3700 gpm (9) Piping and Valves !
Design Pressures 70 psig - discharge piping connected to suppression pool g 4
100 psig - suction piping connected to suppression pool 125 psig - wetwell and drywell sparger piping l 200 psig pump suction piping 500 psig - pump discharge piping 1250 psig vessel suction and return piping Design Temperatures 212HF - suppression pool piping and wetwell and drywell sparger piping 358HF - pump suction and discharge piping 575HF - vessel suction and return piping (10) Valve Operation See Table 5.4-3, RHR PUMP / VALVE LOGIC I
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. Standard Plant V Food storage space is provided as a part of operation. All equipment is designed to the kitchen lunchroom adjacent to the control facilitate the above discussed test and equipment room. Water and food storage adequate inspection functions.
for 12 people for 5 days is stored in this area.
The storage cabinets have a net volume of 28 Failure of any system or component to ft3 useable for food storage. In addition, the properly perform its assigned function during refrigerator has a net volume of 10 ft 3 any test or inspection is grounds for repair or available. Potable water is stored in scaled replacement.
sanitary containers in the kitchen-lunchroom.
6.4.6 Instrumentation Requirements All foodstuffs and water intended for emergency use must be so labeled and not be used A complete description of the required for normal conditions, thus ensuring and adequate instrumentation is given in Subsection supply at all times for emergency use. 6.3.1.1.8.
The sanitary facilities are located across the 6.4.7 Interfaces hall from the control room. !
The control room habitability system design 6.4.5 Testing and Inspection was based on the following environmental conditions.
The system is designed to permit periodic inspection of important components (e.g., fans, 6.4.7.1 External Temperature motors, belts, coils, filters, ductwork, piping, dampers, control instrurnentation and valves), to The maximum external air temperature is assure the integrity and efficiency of the 1150F and the minimum external air
[,_\ system. Local display and indicating devices are temperature is -400F.
provided for periodic inspection of vital parameters such as air temperature upstream and 6.4.7.2 Meterology (x/O's) downstream of the heating and cooling coils, cooling water inlet temperatures, filter pressure The x/O's used for evaluation of the drop, duct static pressures, and water pressures control room operator dose to meet General at the inlet and outlet of coils. Design Criterion 19 were derived from Regulatory Guide 1.3 for ground level release.
Test connections are provided in the duct work Specific values and assumptions are presented and piping for periodic checking of air and water in Subsection 15.6.5.
flows for conformance to design requirements.
All features are periodically tested by 6.4.7.3 Toxic Gases initiating all dampers during normal operation.
Tbc operating system is proven operable by its General Design Criterion 19, as related to performance during normal plant operations. The providing adequate protection to permit access HEPA filters are periodically tested with DOP and occupancy of the control room under smoke per ANSI N101.1. The charcoal filters re accident conditions. Acceptance is based upon to be periodically tested with a freon gas for the meeting the guidance of Regulatory Guide adsorption efficiency. Inspection and sampling 1.78 relating to instrumentation to detect and connection are provided for on site filter alarm any hazardous chemict.1 release in the testing. plant vicinity and relating to the systems capability to isolate the control room from Filter pressure drop is to be routinely such releases; and Regulatory Guide 1.95 monitored and a high differential alarm alerts relating to the systems capability to limit the the operator to switch over to standby system. accumulation of chlorine within the control N room. The ABWR is not design for any hazardous (d The systems are to be tested periodically by chemical release. The control room is provided initiating the changeover sequence during normal with an isolation system for radioactivity Amendment 6 6.4-7 l
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l ' release which can be easily modified to handle additional sensors. Chemical accidents -
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( ,) safety feature since it mitigates the the LOCA and on the duration of these l consequences of a postulated accident by conditions. )
controlling and reducing the release of j radioacti"!ty to the environment. The entire (6) The SGTS component design criteria and j SGTS component, except for the deluge, is qualification testing are in the accordance with {
classified as important to safety with the the recommendations of Regulatory Guide 1.52. '
classifications of Quality Group B. Seismic Category I, and Class IE. 6.5.1.3.2 Sizing Basis {
The SGTS has independent, redundant active The SGTS flow capacity is 1200 cfm based on components. Should any active component fail, the following the function can be performed by a redundant active component. The electrical devices of (1) SGTS sizing is based on the secondary independent components are powered from separate containment inleakage rate.
Class 1E electrical buses.
(2) Free air volume of the secondary containment is 3,000,000 ft3, The SGTS passive filter train (filter casing, (3) Secondary containment inleakage rate is 50%
particulate filters and charcoal adsorber) of the free air volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at a present high reliability in practice. Therefore, differential pressure of (-) 1/4 inch water gage one filter train will be used to perform the SGTS with respect to the outdoor atmosphere.
function. Subsection 6.5.1.3.3 provides further evaluation of the single filter tra% oesign. (4) The above basis gives a flow rate of 1050 cfm. With added margin, the SGTS is sized for Active components, such as exhaust fan and 1200 cfm.
[mV) process heater, are redundant. The dryer unit, containing demister and process heater, and the 6.5.133 Single Filter Train exhaust fan are equipped in two sets, indepen-dently. The dryer unit is installed outside the Acceptance Criteria 11 of SRP Section 6.5.1 filter train and is located upstream of the requires the STGS to be redundant. The entire exhaust fan because 1) The electric process STGS is redundant with the exception of its heater is considered active and redundancy is filter train. The SGTS filter train does not required and 2) this placement meets mechanical require redundancy because the filters and and electrical separation requirements. charcoal adsorber are considered passive components and present high reliability in (4) The SGTS is designed to Seismic Category I practice. Several design features in the SGTS j
requirements as specified in Section 3.2. The are provided to reduce the possibility for SGTS is housed in a Category I structure. All long term passive failure of the filters. These surrounding equipment, components, and supports design features are:
are designed to pertinent safety class and Category I requirements. (1) Advance design of the filter housing and flow pattern virtually eliminates any untreated (5) The SGTS equipment is designed and qualified bypass of the filters.
to the process gas condition and to the environmental conditions that it performs. The (2) Review of the operating plants in the U.S.
SGTS design is based on the maximum pressure and shows that failure in the SGTS filters occurred differential pressure, maximum integrated dose primarily in the charcoal due to inadvertent rate, maximum relative humidity, maximum water spray. The wet charcoal event occurred l temperature and other conditions resulting from when an operator mistakenly opened a deluge
) (3 valve and allowed water to spray into the l () charcoal. The design will eliminate this failure by not normally connecting a water l
Amendmem 6 6.5-3 I
ABM ux6toorn Standard Plant REV.A source to the deluge piping. Water hose and percent hardness, percent moisture, particle connection will be placed nearby. size distribution, an ash content. These tests meet the intent of Regulatory Guide 1.52.
(3) Degradation in charcoal adsorber and filter Elemental and methyl iodine removal and effectiveness due to again effects is avoided by retention capabilities are measure (at periodic testing. These filters will be replaced postulated accident conditions) in accordance with
- fresh" filter if they are not up to their with RDT Standard M16-IT. Impregnated content, expected efficiency rating. leachout, and charcoal ignition temperature are also determined. HEPA filter banks are tested 6.5.1.3.4 Source Terms for SGTS Design in lace before operation to verify 99.9 percent retention based on the DOP smoke penetration The SGTS design is based upon the requirements test. The charcoal filter banks are of Regulatory Guide 1.52. A single charcoal preoperationallly leak tested using a gaseous absorber of 6 inch depth with redundant active halogenated hydrocarbon refrigerant, to measure components (pumps, heaters, demisters, etc.) is bypass leakage and element imperfection in provided to insure iodine filter efficiencies of accordance with ANSI N510, Section 12.
greater than 99% based upon Regulatory Guides 1.3,1.52, and 1.25. 6.5.1.4.2 Periodic Surveillance Testing 6.5.1.4 Tests and Inspection Periodic Surveillance testing of the SGTS is conducted in accordance with the surveillance The SGTS and its components are thoroughly r e q uir e m e e,t s give n in t h e t e ch nical tested in a program consisting of the following: specifications.
(1) Manufacturer's qualification. 6.5.1.5 Instrumentation Requirements (2) Preoperational tests. Control and instrumentation for the SGTS are discussed in Section 7.5. The SGTS is designed (3) Periodic surveillance tests. to function automatically upon receipt of an applicable safety actuation signal.
The above tests are performed in accordance with the objective of Regulatory Guide 1.52. The SGTS is instrumented to display and alarm in the control room under the parameters as 6.e.l.4.1 Preoperational Testing shown in Table 6.5 3. Fan opersting status and valve operating status are also displayed.
The SGTS charcoal filter train housings are pressure tested to demonstrate a leakage of less Differential static pressure indicators is than 0.02 percent of ratd air flow. The housing provided throughout the secondary containment to '
leak test is performed in accordance with ANSI ensure negative pressure is maintained. An N510, Section 6. alar.n shall be annunciated in the control room when the differential pressure exceeds the HEPA filters are shop tested prior to required negative pressure.
installation, in accordance with MIL-F51068 and MIL-STD 282, at 100 percent and 20 percent of Temperature sensors for the charcoal adsorber rated flow, have the capability of detecting the charcoal temperature reliably and accurately on a Impregnated activated carbon is tested before continuously basis, with and without air flow, installation in accordance with the methods and is designed to resist damage by a fire in specified in RDT Standard M16-IT. Tests the charcoal. The sensors shell be set to alarm determine apparent density, degree of activation, in the control room prior to the iodine O
6.54
23A6100All Standard Plant nev. n
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1 CHAPTER 7 TABLE OF CONTENTS Section M East 7 INSTRUMENTATION AND CONTROL SYSTEMS 7.1 GENERAL 7.1-1 7.1.1 Identification of Safety-Related Systems 7.1 1 7.1.2 Identification of Safety Criteria 7.1-3 7.2 REACTOR PROTECTION (TRIP) SYSTEM 7.2-1 73 ENGINEERED SAFE'IY FEATURES SYSTEMS.
INSTRUMENTATION AND CONTROL 73-1 73.1 Description 73-1 73.2 Analysts 7 3-34
(,/ - Emergency Core Cooling Systems -
Instrumentation and Controls 7 3-34
- I.cak Detection and Isolation System Instrumentation and Controls 7 3-39
- RHR/Wetwell and DrywellSpray Mode -
Instrumentation and Controls 73-40
- RHR/ Suppression Pool Cooling Mode -
Instrumentation and Controls 7 3-41
- Standby Gas Treatment System -
Instrumentation and Controls 73-43
- Emergency Diesel Generator Support Systeu -Instrumentation and Controls 7 3-44
- Reactor Building Coolits Water System - Instrumentation and Controls 7.5-c
- Essential HVAC Systems -
Instrumentation and Controls 7 3-46 O
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Amendment 6
Mb 23A6100AF Standard Plant REV.A CH/ PTER 7 TABLE OF CONTENTS (Continued) h Section M Eage HVAC Emergency Cooling Water l
System - Instrumentation and Controls 7 3-48
- High Pressure Nitrogen Gas Supply System - Instrumentation and Controls 7 3-49
- Additional Design Considerations q Aralyses 73-51 I
- Periodic Testing of ESF Instrumentation 73-51 7.4 SYSTEMS REOUIRED.FOR SAFE h.tiL'TDOWN 7.4-1 7.4.1 Description 7.4-1 7.4.2 Conformance Analysis 7.4-10
- Alternate Rod Insertion Function 7.4-10
- Standby Liquid ControlSystem 7.4-11 RHR/ Shutdown Cooling Mode 7.4-13
- Remote Shutdown System 7.4-15 7.5 SAFETY RELATED DISPLAY INSTRUMENTATION 7.5-1 7.6 ALL OTHER INSTRUMENTATION SYSTEMS REOUIRED FOR SAFETY 7.6-1 7.6.1 Description 7.6-1 7.6.2 Analysis 7.6-12 7-iii O
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MM 23A6100AF Aimadard Plant REV H p
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7.1.2.1.6 Protection System In Service Testability traces can be made for each rod scrammed. Prior to the test, a physics review is conducted to assure that the rod pattern during scram testing The RPS and ESF systems can be tested during will not create a rod of unacceptable reactivity reactor operation by six separate tests. The worth.
first five tests are primarily manual tests and, although each individually is a partial test, The fourth test involves applying test combined with the sixth test they constitute a signals to each Digital Trip Module (DTM) in complete system test. The sixth test is the turn and observing on a digital display that the self test of the safety system logic and control channel trips change state. One method utilizus which automatically tests the complete system electrical signals generated by the calibration excluding sensors and actuators. and trip test unit (CTTU) and fed to the DTM while bypassing the transmitter. If desired, The first of these is the manual scram test. the transmitter may be used directly in the i The manual scram test verifies the ability to test. In this method, the test signals simulate I de energize the scram pilot valve solenoids with- the actual process signal. The digital, micro-out scram by using the manual scram pushbutton processor controlled, SSLC circuitry does not switches. By depressing the manual scram button require calibration. Automatic self-test func-for one trip logic, half of the scram solenoids tions will simulate programmed trip setpoints are de energized. After the first trip logic is and monitor the response. Analog sensor inputs reset, the second trip logic is tripped manually are calibrated at the analog inputs of the to complete the test for the two manual scram remote multiplexing units. With a division-of-buttons. In addition to control room and sensors bypass in place, calibrated, variable computer printout indications, scram groups ramp signals are injected in place of the sensor indicator lights indicate that the actuator trip signals and monitored at the SSLC control room l[,}
v logics have de energized the scram pilot valve panels for linearity, accuracy, fault response, solenoids. and downscale and upscale trip response. The test signals shall be adjustable manually from On the back panels, a separate, manual the control room and also shall be capable of pushbutton switch in each of the four divisions performing an automatic sequence of events, provides means to manually trip all trip When surveillance testing during plant shutdown, actuators in that division. This scaled in trip coincidence and actuated device operation division manual trip is equivalent to a sealed-in can be verified by simultaneous trip tests of automatic trip from the same division of trip coincident channels. Pressure transmitters and logic. (An alternative manual scram can be level transmitters are located on their respec-accomplished by depressing any two or more of the tive local panels. The transmitters can be in-four divisional manual trip pushbuttons.) dividually valved out of service and subjected to test pressure to verify operability of the The second test includes calibration of the transmitters as well as verification of calibra-neutron monitoring system by means of simulated tion range. To gain access to the field con-inputs from calibration signal units. Calibra- trols on each transmitter, a cover plate or tion and test controls for the neutron monitoring seniing device must be removed. The access to system are located in the control building equip- the field controls is administratively control-ment room. They are under the administrative led. Only qualified personnel are granted ac-control of the control room operator and can be cess for the purpose of testing or calibration done either manually or automatically, adjustments.
Subsection 7.6.1.1, " Neutron Monitoring System,"
describes the calibration procedure. The fifth test is the sensor check. Digital inputs are tested by varying the monitored vari-The third test is the single rod scram test able (e.g., stop valve closure, control valve
,A which verifies the capability of each rod to fast closure, main steamline isolation valve
( scram. It is accomplished by operating switches closure) or by disconnecting the sensor from the for the particular control rod drive. Timing process variable and inputting and varying a i
Ameneartat 6 7.15
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-hhh 23A6100AF REv.B dtandard Plant l test source. In those cases where the sensor is ..
disconnected from the process variable, an out of service alarm will be indicated in the j main control room. Analog input is checked by 1 cross comparison of the instrument channels j measuring the same variable. {
j The sixth test is an automatic self-test performed by the circuit boards within the safety system logic and control (SSLC).
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This self test subsystem (STS) is an overlay (2) A more complete, manually initiated, gd i - testing and surveillance subsystem which provides internal self test is available when a unit the capability to continuously and automatically is off-line for surveillance or maintenance perform end to end testing of all active testing. This test exercises the trip circuitry, within the SSLC panels, essential to outputs. The channel containing the unit is the safe shutdown'of the reactor, bypassed during this phase of testing.
The primary purpose of the STS is to improve A fault is considen.d the inability to open the availability of the SSLC by optimizing the or dose any control circuit.
time to detect and determine the location of a failure in the functional system. It is not in- Self test failures are displayed on a front tended that the STS climinate the need for the panel readout device or other diagnostic other five manual tests. Rather, by continuously unit.
providing on line periodic test, most faults are detected more quickly than by manual testing (3) Passive test monitoring available only. The STS is classified as Safety Associated continuously shall include power supply and its equipment wherever it interfaces with voltage levels, card out of-file interlocks, safety equipment is qualified to Class 1E and battery voltage levels on battery-backed standards. memory cards (if used). Out of tolerance conditions shall result in an inoperative The hierarchy of test capability is provided (out-of service) condition for that to ensure full coverage of all EMS /SSLC func- particular system function, tions, including logic functions and data com-munications links (See Figure 7.1-1). Testing (4) An automatic system self test is provided to shall include: inject test patterns through the EMS communications links to the RMUs, test the (n)
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(1) Internal self-test of each signal proce.tsing module from input to output. Teuing is ability of the RMUs to format and transmit sensor data through and across the EMS /SSLC automatic and performed once every ten interface in the prescribed time, and test minutes during normal operation. These the trip logic from seal-in to actuation of verify the basic integrity of each card or the load drivers and solenoids through the module on the microprocessor bus. All dedicated (non multiplexed) communication operations are part of norrmal data process- links, ing intervals and do not affect system re-sponse to incoming, trip initiation signals. Self test of the load drivers is performed A fault is considered the discrepancy by injection of test patterns representing between an expected output of a permissive short duration trip levels through the circuit and the existing present state. divisional isolators to the output driving circuitry. Voltage and current sensing Actuation of the trip function is not feedback to the test unit verifies that the performed during this test. power switchiog device is capable of de energizing on command, but the solenoid The self test function is capable of detect- does not operate, ing and logging intermittent failures with-out stopping system operation. Normal sur- Automatic system self testing occurs during veillance by plant personnel can identify a portion of every periodic transmission these failures via a diagnostic display for period of the data communication network, preventive maintenance. Since exhaustive tests cannot be performed during any one transmission interval, the Self test failures (except intermittent test software is written so that sufficient failuses) are annunciated to the operator at overlap coverage is provided to prove system
[b the main control room console el logged by the process computer, performance during tests of portions of the circuitry, as allowed in IEEE Std. 338.
Ameadhnent 2 7.16
MM 23A6100AF Standard Plant REV.B
.q V' applicable to the diesel generator and its 7.1.23.8 Essential HVAC Systems-auxiliaries are listed in Table 7.1-2. Instrumentation and Controls (2) Nonsafety-Related Design Bases (1) SafetyDesignBases There is no power generation design basis See Subsections 9.4.1.1.l and 9.4.5.1.1.
for this system.
7.1.23.9 HVAC Emergency Cooling Water System.
7.1.2.3.7 Reactor Building Co'> ling Water lastrumentation and Controls System Instrumentation and Controls (1) Safety Design Bases (1) Safety Design Bases General Functional Requiremer.ts:
General Functional Requirements:
The general functional requirements of the The general functional requirements of the HVAC emergency cooling water system i tstru-instrumentation and controls of this system mentation and controls shall provide control shall be to: for cooling units that ensure a controlled environment for essential equipment and (a) maintain control of cooling water to control room areas following a loss-of.
equipment that requires cooling during coolant accident, loss of preferred power, reactor shutdown modes and following a or isolation of normal heating, venting, and LOCA; air conditioning (HVAC).
(b) provide for the automatic isolation of Specific Regulatory Requirements:
the non essential parts of the reactor U' $.
building cooling water system (except CRD pump oil coolers and instrument air The specific regulatory requirements appli-
. cable to the system instrumentation and coolers) from the essential parts during control are given in Table 7.1-2.
LOCA or upon detection of a major RCW leak in the non-essential system; (2) Nonsafety-Related Design Bases (c) satisfy Seismic Category I design The system shall provide a continuous supply requirements of chilled water to the cooling coils of air conditioning systems which provide a con-Specific Regulatory Requirements: trolled temperature environment and proper humidity to ensure the comfort of the The specific regulatory requirements operating personnel and to provide a applicable to the system instrumentation and suitable atmosphere for the operation of controls are given in Table 7.1-2. control equipment.
(2) Nonsafety Related Design Bases 7.1.2.3.10 High Pressure Nitrogen Gas Supply System - Instrumentation and Control (a) Controls and instrumentation shall be provided to control and monitor the (1) Safety Design Bases ;
distribution of reactor building cooling water to remove heat from plant auxilia- General Functional Requirements:
ries during normal plant operation.
The general functional requirements of the (b) The essential service water system shall instrumentation and controls shall provide be capable of being tested during normal automatic and manual coritrol of the nitrogen
( plant operation. gas supply to assure its operation during Amendment 6 7.111
M 23A6100AF Standard Plant REV.A 1
I all modes of plant operation; and to automa- (e) mitigate the consequences of anticipated tically initiate the emergency nitrogen bot- transient without scram (ATWS) events.
tie supply (on low nitrogen supply pressure) to assure adequate supply of nitrogen to 7.1.2.4.2 Standby Liquid Control System automatic depressurization safety / relief (SLCS) - Instrumentation and Controls valves and to nitrogen using equipment and valves in the reactor building. (1) Safety Design Bases Specific Regulatory Requirements: GeneralFunctional Requirements:
The specific regulatory requirements appli. The 'neral functional requirements of this cable to this system are listed in Table equipment are to provide necessary control 7.1-2. of the SLC equipment for shutting the reac-tor down from full power to cold shutdown (2) Nonsafety-Related Design Bases and maintaining the reactor in a suberitical state at atmospheric temperature and pres-There is no power generation design basis sure conditions by pumping sodium pentabo-for this system. rate, a neutron absorber, into the reactor.
S.I.2.4 Safe Shutdown Systems-Instrumentation Specific Regulatory Requirements:
and Control The specific regulatory requirements 7.1.2.4.1 Alternate Rod Insertion Function applicable to this system are given in Table (ARI) - Instrumentation and Controls 7.1 2.
(1) Safety Design Bases (2) Nonsafety-Re:ated Design Bases None. None.
(2) Nonsafety-Related Design Bases 7.1.2.4.3 RHR Reactor Shutdown Cooling Mode (RHRS) Instrumentation and Controls The general functional requirements of the instrumentation and controls of the ARI (1) Safety Design Bases function are to:
General Functional Requirements:
(a) Provide alternate and diverse method for inserting control rods using fine motion The general functional requirements of the control rod drive (FMCRD) electric shutdown cooling mode of the RHR are to pro-motors; vide monitoring and control as required to:
(b) provide for automatic and manual (a) enable the system to remove the residual operation of the system; heat (decay heat and sensible heat) from the reactor vessel during normal (c) provide assurance that the ARI shall be shutdown; highly reliable and functional in spite I of a single failure; (b) provide manual controls for the shutdown ,
cooling system in the main control room l
[ and at the remote shutdown panel; and l l (d) provide assurance that the ARI shall operate when necessary (fine-motion i
i control rod drive motors shall be (c) indicate performance of the shutdown connected to the emergency diesel cooling system by main control room generators); and instrumentation and instrumentation in the remote shutdown panel.
7.1-12 Amendment 2 l
P MM Ef endmed Plant 23A6100AH REv. n I
O secTioN 7.3 CONTENTS Section Title Eage A
l 73.1 Description 73-1
.y )
73.1.1 Systems Descriptions 73-1 73.1.1.1 Emergency Core Cooling Systems Instrumentation i and Controls 73-1 73.1.1.1.1 High Pressure Core Flooder System Instrumentation and Controls 73-1 73.1.1.1.2 Automatic Depressurization System Instrumentation and Controls 73-4 73.1.1.13 Reactor Core Isolation Cooling (RCIC) System -
Instrumentation and Controls 73-9 73.1.1.1.4 RHR/ Low Pressure Flooder (LPFL) Instrutuentation and Controls 73-1.,
N 73.1.1.2 Leak Detection and Isolation System (LD&IS)
Instrumentation and Controls 7 3-19 ,
73.1.13 RHR/Wetwell and Drywell Spray Cooling Mode Instrumentation and Controls 7 3-20 73.1.1.4 RHR/ Suppression Pool Cooling Mode-Instrumentation and Control 73-22 73.1.1.5 Standby Gas Treatment System Instrumentation and Controls 7 3-24 73.1.1.6 Emergency Diesel Generator Support Systems 7 3-26 73.1.1.7 Reactor Building Cooling Water System -
Instrumentation and Controls 7 3-26
= 73.1.1.8 Essential HVAC Systems - Instrumentation and Controls 7 3-28 73.1.1.9 HVAC Emergency Cooling Water System Instrumentation and Controls 7 3-28 73.1.1.10 High Pressure Nitrogen Gas Supply System
- p. Instrumentation and Controls 7 3-31 ll q l-73-il i
Amundment 6 h I
MM 23A6100AH Standard Plant REV.B SECTION 7.3 CONTENTS (Continued) 9 Section Tttig Eggt 73.1.2 Design Basis Information 73=32 73.13 System Drawings 7 3-34 73.2 Anallli 73 34 73.2.1 Emergency Core Coo!ir.g Systems Instrumentation and Controls 7 3-34 73.2.1.1 General Functional Requirements Conformance 7 3-34 l
< l 73.2.1.2 Specific Regulatory Requirements Conformance 7 3-36 i
73.2.2 Leak Detection and Isolatiou System Instrumentation and Controls 73 39 73.2.2.1 General Functional Requirement Conformance 73-39
]
73.2.2.2 Specific Regulatory Requirements Conformance 73-39a 73.23 RHR/Wetwell and Drywell Spray Mode Instrumentation and Controls 73-40 l
73.2.4 RHR/ Suppression Pool Cooling Mode -
Instrt. mentation and Controls 73-41 73.2.4.1 G eneral Functional Requirements Conformance 73-41 1
73.2.4.2 Specific Regulatory Requirements Conforinance 73-42 73.2.5 Standby Gas Treatment System Instrumentation i and Controls 7 3-43 :
73.2.5.1 Conformance to General Function:.1 Requiremer.s 73-43 73.2.5.2 Specific Regulatory Requirements Conformance 7 3-43 73.2.6 Emergency Diesel Generator Support System Instrumentation and Control 7 3-45 r !
7 3 -111 l
1 Amendment 6
I LABWR 23samar REV.A Remndard Plant l v-C) SECTION 7.3 CONTENTS (Continued)
Section M Eagt 73.2.6.1 Conformance. to General Functional Requirements 7 3-45 73.2.6.2 Specific Regulatory Requirements Conformance 7 3-45 73.2.7 Reactor Building Cooling Water System Instrumentation end Controls 7 3-46 73.2.7.1 Conformance to General Functional Requirements 7 3 73.2.7.2 Specinc Regulatory Requirements Conformance 7 3-46 73.2.8 : Essential HVAC Systems Instrumentation and Control 73 47 73.2.8.1 Conformance to General Functional Requirements 73 47.
73.2.8.2 Specific Regulatory Requirements Conformance 73-47 fq 73.2.9 HVAC Emergency Cooling Water System
-Q Instrumentation and Control 7 3-48 73.2.9.1 Conformance to General Functional Requirements 7 3-48 73.2.9.2 Specific Regulatory Requirements Conformance 7 3-48 73.2.10 High Pressure Nitrogen Gas Supply System 7 3-49 73.2.10.1 Conformance to General Functional Requirements 73 50 73.2.10.2 Specific Regulatory Requirements Conformance 7 3-50 7 3.2.11 Additional Design Consideration 4 Analyses 7 3-51 7 3.2.11.1 General Plant Safety Analysis 73-51 73.2.11.2 less of Plant Instrument Air System 7 3-51 73.2.113 kss of Cooling Water to Vital Equipment '73-51 7 3.2.12 Periodic Testing of ESF Instruinentation 73-51 73-iv Amsadment 2
MM 23AM00All nev.n Standard Plant __ _
.. SECTION 7.3 ILLUSTRATIONS hl Figure Thlt Eagt 73-1 HPCF System IBD 73-52 73-2 Nuclear Boiler System IBD 73-59 73-3 RCIC System IBD 73 74 73-4 RHR System IBD 73-90 73-5 Leak Detection and1 solation System 7 3-104 73-6 Standby Gas Treatment System IBD 73-105 73-7 Reactor Building Cooling Water System IBD 7 3-116 73-8 E:sential HVAC System IBD 73-132 j 73-9 HVAC Emergency Cooling Water System IBD 7 3-133 7 3-10 High Pressure Nitrogen Gas System IBD 7 3-141 O
i i
73-v i
Amendment 6 e __ _ _ - _ _ _ _ _ _ _ _ _ _
a ,e t '
i j
23A6100AF t
Standard Plant ._ REV,j} j m , ,
(') : 7.3 ENGINEERED SAFETY FEATURE SYSTEMS, INSTRUMENTATION AND (2) automatic depressurization system (ADS) (in-cludes safety / relief valve [SRV) electrical
. CONTROL activation logic);
7.3.1 Description (3) reactor core isolation cooling (RCIC) system; and 7.3.1.1 Systems Descriptions (4) low pressure flooder (LPFL) mode of the re-This subsection describes the instrumentation sidual heat removal (RHR) system, and controls for the various engineered safety features (ESF) systems. It provides design basis The purpose of ECCS instrumentation and con-information as called for by IEEE 279 and pro- trols is to sense the need for Er"t action and vides reference to system diagrams which are in- :o initiate appropriate response .. ':. he system cluded in the Safety Analysis Report. in the even: of an accident requiring its action.
Sapporting systems for the control and instru-mentation (C&I) equipment include the instrument, The emergency core cooling syster" instrument logic, control and motive power sources and are channels detect a need for cos e coo:ing spiems addressed under the heading of power supplies for operation, the logic makes approj;. iate deci.
each system. sions, and the trip actuators initiate the ,
appropriate equ'ipment operation.
The EhF systems described in this section include the following: 7.3.1.1.1.1 High Pressure Core Flooder System Instrumentation and Controls (1) emergency core cooling systems (ECCS);
.( (1) System Ider.tification !
L- (2) ' leak detection and isolation system (LDS);
The control and instrumentation components (3) wetwell and drywell spray mode of RHR for the high pressure core flooder (HPCF)-
(WDCS-RHR); system, except as noted in this subsection, are located outside the drywell. Pressure (4) suppression pool cooling mode of RHR and level transducers used for HPCF (SPC-RHR); initiation are part of the nuclear boiler system and are located on racks outside the !
(5) : standby 6as treatment system (SGTS); drywc!1. The system is arranged to allow a design flow functional test during normal (6) ' emergeney diesel generator support systems; reactor power operation. The piping and instrumentation diagram is shown in Section (7) reactor beilding cooling water system; 6.3 and the interlock block diagram is shown on Figure 7.3-1.
(8) essential HVAC systems; (2) Supporting Systems (Power Supplies)
(9) HVAC emergency cooling water syS' m; and Supporting systems for the HPCF C&l consist (10) high pressure nitrogen gas supply system. only of the instrumentation, Iogic and motive power supplies. The controls 7.3.1.1.1 Emergency Core Cooling Systems instrumentation and logic power is obtained Instrumentation and Controls from the SSLC Divisions 2 and 3,120-VAC UPS buses discussed in Section 8.3. The logic The emergency core cooling systems (ECCS) are power is as described in Section 7.2 for "a network of the following systems: the RPS portion of the $5LC.
(1) high pressure core flooder (HPCF) system;
' Amendment 6 7.34
s .
N, ' ~ ABWR m6-Standatti Plant __.,_ __ nnv. A i'
(3) Equipment Des'gu This valve will reopen if reactor .
water level subsequently decreases to The HPCF system is designed ito operate from the low initiation level. The system is preferred offsite power sources or fmm the arranged to allow automatic or manual -l Divisions 2 and 3 diesel generatou if off- operation. The HPCF initiation signal site (preferred) power is not available, from the nuclear boiler system also initiates the standby diesels in the 1 (a) Initiating Circuits respcctive divisions.
f Reactor. vessel low water level is mon- Two AC moto:t operated valves and two itored by four level transmitters (one check valves are provided in ca:h of the 1 in each of the four electrical divis- two putup suction loops. Within each ions) that sense the difference between loop (B and C), one valve lines up pump the pressure due to a constant reference suction from the condensate storage leg of water and the pressure due to the pool, the other from the suppression actual height of water in the vessel, pool The control arrangemen,t is shown .
I Each level transmitter provides an input in Figure 7.3-1 Reactor grade water in to local multiplexer units which perform the condensate storage pool is the pre-signal conditioning and analog to-digi- ferred source. On receipt of an HPCF tal conversion. The formatted, digit- initiation algnal, the condensate stor-ized sensor input is multiplexed with age pool suction valves are automatic-other sensor signals over an optical fi- ally signaled to open (they are normally
' ber data link'to the logic processing in the open position unless the units in the main control room. All suppression pool suction valves are four transmitter signals are fed in to open). If the water level in the :
the two out of four logic for each of condensate storage pool fails below a i the two divisiotis (II & 111). The init- preselected level, first the suppression '
iation logic for HPCF sensors is shown pool suction valves automatically open in Figure 7.31, and then the condensate storage pool j suction valves automatically clase. Two j 4
Drywell pressure is monitored by four level transducers are used to detect low pressure transmhters in the same four- water lesel in the condensate storage divis. ion configuration described above. pool. Either transducer can cause the
) Insttument sensing lines that terminate suppression pool suction valves to open j outside the drywell allow the transmit- and the condensate storage valves to i ter to communicate with the drywellint- close. The suppression pool suction erior. Each drywell high-pressure trip valves also autotnatically open if high channel provides an input into two-out- water level is detected in the sup. j of-four trip logic shown in Figure pression pool. Two level transducers {
7.3 1. monitor this water level and either -
transducer can initiate opening of the -
The HPCF system is initiated on receipt suppression poc>l suction valves and I of a reactor vessel low water level sig- closure of condensate storage pool nul (level 1.5) or drywell high-pres- suction valves.
sure signal from the trip logic. The HPCF system reaches lts design flow rate (b) Logic and Sequencing within 36 seconds of receipt of initia-tion signal. Makeup water is discharpd Either reactor vessel low water level to the reactor vessel until the reactor (Level 1.5) or high drywell pressure high water levelis reached. The HPCF automatically starts the HPCF as then automatically stops flow by closing indicated in Figure 7.3-1. ,
the injection valve if the high water level signa) is available. (c) Bypasses at;d Icterlocis Amendment 2 712
y ,
I l
MM 23A6100AF 1
Standard Plant any A !
.. ,m All of the safety related portions of b
k") the RCW system are started automatical-(d) Redundancy and Diversity ly upon a LOCA and/or LOPP (emergency The RCW system instrumentation and
. diesel starting signal due to loss of power supplies are separated into three offsite power). The containment isola- divisions such that no single occur-tion valves are closed automatically rence results in the loss of function upon receipt of the LOCA signal o.r may of more than one division. Overall be closed manually from the control redundancy is provided by separated, ;
toom, divisional service water loops for ]
Divisions I, II, and III. 1 (b) 1.ogic and Sequencing I
-(e) Actuated Devices l The I.OCA signal used tc> actuate the RCW l water isolation systein is derived from The automatically actuated isolation the two-out of-four logic of reactor low valves in the RCW system are provided level or high drywell pressure trip with electric motor operators. The signals. The signal is generated by: valve limit switches turn off the motor ]
when the valves are fully open and Two-of-four level sensors being permit torque switches to control valve tripped; or motor forces while the valves are seating in the closed direction. Other l Two-of four pressure sensors being valves have torque limits in the open tripped; or direction except at breakaway and torque limits on closing.
Both sets of the above.
['T (f) Separation
\_/ Once an initiation signal is received, the signal is scaled in RCW system trip channels, logic cir-until manually reset. cuits, manual controls, cabling and instruments are mounted so that Divi-The isolation valves stay closed sion I, Division II, and Division III until the LOCA signal is no longer separation is maintained in accordance present or a control switch is with criteria stated in Section operated in the control room. 8.3.1.3.
1 (c) Bypra and Interlocks (g) Testability 1 The LOCA nignal that autanatically ini- The RCW system has the capability of tintes the consafety-related service wa- being tested during normal plant ter isolation system can be overridfen operation.
by a control switch in the cortrol room. if the operator determines that System control and logic circuits can the nonsafety-related auxiliaries are be individually checked by applying operable, flow can be initiated by a test or calibration signals and observ-combination LOCA, override and manual ing the system tesponse. The control valse-opening operation. The remote clicuitry is designed to restore the rhutdown panel has control transfer system to the required operation if a capability to take manual control of LOCA occurs during a test.
Divisions I and II of the RCW. (See subsection 7. 4.1. 4 . 4 ( 5 ) for RSS (h) Eiwiror.mentalCotuiderations j interface.)
- (m) The only control components pertinent v
9 h
MM 23A6100AF Standard Plant REV.H to this system that are located inside ation of the RHR, HECW, FPC, CAM, and the containment are nuclear boiler emergency diesel generator systems.
system sensors that generate signals for ,
the LOCA signal logic. Refer to Section When the plant is in the hot standby or 3.11 for environmental qualifications of cooldown mode, safety related RCW
! this equipment. cooling water is required for the RHR
! heat exchangers. Refer to Subsection (i) SafetyInterfaces 7.3.1.1.4 for a discussion of the manual or automatic operation of the The safety interfaces for the RCW system RHR heat exchanger inlet and outlet Divisions I, II, and III controls are as isolation valves, follows:
Process operating parameters and equip-LOCA signals to Divisions I,II, and ment status information are provided in III RCW pumps; the control room for the operator to accurately assess system performance.
Divisions I, II, and III RCW pump . Alarms are also provided to indicate manual start signals from the main malfunction in the system. Refer to control room (MCR) and the remote IBD Figure 7.3-7 for specific indica.
shutdown system (RSS); tion of equipment status in the control room. See Chapter 16 for setpoints and Divisions I, II, and III RCW pump margin.
running signals to the MCR and RSS; (k) Parts of System Not Required for Safety Divisions I, it, and 111 cooling .
water supply low pressure signals to The nonsafety-related portions of the the MCR and RSS; RCW system include the annunciators and the computer. Other instrumentation Divisions I, II, and III RCW flow considered nonsafety-related are those signals to the MCR and RSS; indicators that are provided for operator information, but are not RCW Hx A or D strainer differential essential to correct operator action.
pressure MCR annunciator; 7.3.1.1.8 EssentialIWAC Systems -
Overload and power failure signals Instrumentation and Controls from all RCW and RSW pumps to the MCR annunciator; See Subsections 9.4.1 and 9.4.5, RCW surge tank low and high level 7.3.1.1,9 HVAC Emergency Cooling Water System signals to the MCR annunciator. Instrumentation and Controls RCW cooling water high temperature (1) System Identification signals to the MCR annunciator. ,
The HVAC emergency cooling water system !
(j) OperationalConsiderations (HECW) is the system which supplies demin-eralized chilled water to the cooling coih The RCW system is capable of operating of the control building safety-related at a variety of cooling load conditions electrical equipment rooms and main control as required for all plant operating room coolers, sad the diesel generator zone modes, including normal and emergency air conditioning systems. The system is conditions. composed of two divisions each containing two 50% refrigerators and chilled water Cooling water is required for the oper- pumps, AmeMment 6 7.3-28 i i
)
l m ; .
.j
yt y a di.# x ABM Standard Plant . _ _ , _
234sioaxe REV H i (2) Variables (6) Range of Energy Supply and Environmental '
Conditicas of Safety Related Systems
, The plant variables that are monitored to provide automatic protective actions are See Section 3.11 for environmental condi-discussed in the initiating circuits ' tions and Chapter 8 for the range of energy
' sections for each system. For additional supply conditions.
information, see Chapaer 15, where safety analysis parameters for each event are ECCS 125 VDC power is provided by the four cited. divisions of station batteries. ECCS 120 VAC power is provided by the SSLC buses.
(3) Number of Sensors and Location ESF systems motor-operated valve power is i
There are no sensors in the LDS or ECCS supplied from motor control centers.
which have a spatial dependence and therefore, location information is not (7) Malfunctions., Accidents, and Other Unusual j relevant. The only sensors used to detect Events Which Could Cause Damage to Safety-essential variables of significant spatial Related Systerrs dependence are the neutron flux detectors (see Subsection 7.2.2.1(6)) and the radia- Chapter 3 covers the description of the tion detectors of the process radiation following single credible accidents and monitoring system. These are in Section events: flood, storm, tornado, carthquake, 7.6. All other systems discussed in Section fire, LOCA, pipe break outside containment, ,
7,3 have sensors which have no spatial' and feedwater line break. Each of these h dependence. events is discussed below for the ESF systems and ECCS.
(4) Operational Units i( (a) Mood g Prudent operational Hmits for each safety related variable trip setting are The buildings containing ESF systems selected to be far enough above or below and ECCS comp <ments have been designed normal operating levels so that a t,purious to meet the probable rnaximum flood
[
l ESF system initiation is avoided. Analysis (PMF) at the site location. This j then verifies that the release of radio- ensures that the buildings will remain active materials,~ following postulate gross watertight under PMF conditions includ-
> failures of the fuel or the nuclear system ing wind generated wave action and wave process barrier, is kept within established runup.
limits.- Operational limits contained in the i '
techrical specifications for the ECCS and (b) Storm (Tornado)
LDS arc based on operating experience and conurained by the safety design basis and The buildings contairdng ESF components the safety analyses. bave been designed to withstand mete-orological evento described in Sub-(5) Margin Between OperationalLimits seetion 3.3.2.
d- The naargin between operational limits and Superficial damage may occur to miscel- )
the' limiting conditions of operation for the laneous station property during a pos-ESF systems instruments are listed in Chap- tulated tornado, but this will not im-1 ter ItL The mdgin includes the conside- pair the protection system capabili-
,o ration of 4 ea<,or and instrament channel ac- ties.
E4 c6 racy, resporde times, and setpoint drift.
Hh. .
(c) Earthquake I- . Indicators are provided to alert the reactor
( M[
L oteraf.or of the onset 'of unside conditions. The structures containing ESF compo-zgy '
7.3-33
.g Meerdre:m 6 ,
.f ; ' f. ,
ABM 23462 mar REV.B Standard Plant nents have been seismically qualified as shall, as a minimum, initiate safety action described in Sections 3.7 ard 3.8, and in a sufficient number of systems and sub-vill remain functional during and fol- systems to accomplish timely initiation of kowing a safe shutdown earthquake any required safety function under condi-(SSE). Seismic qualification of instru- tions of a single design basis event with mentation and electrical equipment is its consequential damages and a single fail-discussed in Section 3.10. ure together with its consequential dama-ges.
To protect ESF systems in the event of a postulated fire, the redundant portions Trip points are within the operating rauge of the systems are separated by fire of instruments with full allowance for barricis. If an internal fire were to instrument error, drift, and setting error.
occur within one of the sections of a main control room panel or in the area 73.13 System Drawings of one of ?he local panels, the ESF sys-tems functions would not be prevented by A list of the drawings is provided in Section the fire. The use of separation and 1.7. P&lDs are provided within Chapters 5,6, fire barriers ensures that, even though and 9, and are referenced where appropriate in some portion of the system may be Chapter 7. All other diagrams, tables, and fig-affected, the ESF system will continue ures are inclnded in Chapter 7 as appropriate.
to provide the required protective Subsection 1.7.3 provides keys for the interpre-action. The remote shutdown system tation of symbols used in these documents.
provides redundancy in the event of sig'aificant exposure fires in the 7.3.2 Analysis control room.
Failure modes and effects analyses for ESF The plant fire protection system is systems are provided in Chapter 15.
discussed in Section 9.5.
73.2.1 Emergency Core Cooling Systems ,
l (d) LOCA Instrumentation and Controls The following ESF system instrument taps 73.2.1.1 General Functional Requirements i and sensing lines are located inside the Conformance drywell and terminate outside the dry-well. They could be subjected to the Chapters 15. " Accident Analysis," and 6, effects of a design basis loss of-cool- " Engineered Safety Feature Systems," evaluate ant accident (LOCA): the individual and combined capabilities of the emergency cooling systems. For the entire range Reactor vessel pressure of nuclear process system break sizes, the cooling systems provide adequate removal of Reactor vessel water level decay heat from the reactor core.
j Drywell pressure Instrumentation for the emergency core cooling systems must respond to the potential These items have been environmentally inadequacy of core cooling regardless of the qualified to remain functional during location of a breach in the reactor coolant and following a LOCA as discussed in pressure boundary. Such a breach inside or Section 3.11. outside the containment is sensed by reactor low water level. The reactor vessel low water level I (8) Minimum Perfownce Requirements signal is the only emergency core cooling system iritiating function that is completely The instrumentation and control for the indep~ndent of breach location. Consequently, i various systems ie:,cribed in this section it can actuate HPCF, RCIC, ADS and LPFL.
l l
Amendment 6 7.3-34
23A6100AF
, Standard Plant REV.B s
i l K '
(c) TMI II.K.3(13) - HPCI and RCIC Initia-in two-out of four logic permissives to tion Levels automatically close the LPFL injection valves should reactor pressure exceed (d) TM1 II.K.3(15) - HPCI and RCIC the low pressure system design pres. Initia- tion Levels sure. Therefore, the . CCS is full com-pliance with this BTP. (d) TMI II.K.3(15) - Isolation of HPCI and RCIC (b) BTP ICSB 20 - Design of instruments-tion and Controls Provided to Accomp. (c) TMIII.K.3(18) ADSActuation lish Changeover from injection to Re.
circulation Mode: The ABWR, as with (f) TMI II.K.3(21) - Restart of LPCS and the BWR, has entirely separate systems LPCI for vessel injection and for vessel recirculation. Therefore, this BTP is (g) TMI II.K.3(22) - RCIC Automatic not applicable to the ABWR, Switch over (c) BTP ICSB 21 - Guidance for Appliation These, and all other TMI action plan re-of Regulatory Guide 1.47: The ABWR quirements, are addressed in ApperAix 1A.
design is a single unit. Therefore, item B 2 of the BTP is not applicable. 7.3.2.2 Leak Detection and Isolation System Otherwise, the ECCS is in full comp- Instrumentation and Controls liance with this BTP.
n (d) BTP IGSB 22 - Guidance for Application 7.3.2.2.1 General functional Requirements Con-
) of Regulatory Guide 1.22: In general, formance actuated equipment within the reactor protection system can be fully tested The leak detection and isolation system during reactor operation. Exceptions (LDS) is analyzed in this subsection. This sys-for the RPS scram function are dis- tem is described in Subsection 7.3.1.1.2, and cussed in 7.2.2.2.3.1 (10). Exceptions that description is used as the basis for this for ECCS include the ADS valve pilot so- analysis. The safety design bases and specific lenoids and the LPFL shutdown valves regulatory requirements of this system are which cannot be opened while the reac- stated in Section 7.1.
tor is pressurized. However, both can be tested during reactor shutdown. In addi. The isolation function of the LDS in con-tion, the ADS valve solenoids are junction with other safety systems, is designed monitored for continuity during the to provide timely protection against the onset logic selftest. and consequences of the gross release of radioac-tive matcrials from fuel and reactor coolant (5) TMI Action Plan Requirements (TMI): pressure boundaries. Chapter 15 identifies and evaluates postulated events that can result in In accordance with the Standard Review Plan gross failure of fuel and reactor coolant pres-for Section 7.3, and with Table 7.1-2, the sure boundaries. The consequences of such gross following TMis are considered applicable for failt.res are described and evaluated. Chapter the ECCS: 15 also evaluates a gross breach in a main steamline outside the containment during op-(a) TMl ll.D.3 - Relief and Safety Valve Po- eration at rated power. The evaluation shows sition Indication that the mtiin steamlines are automatically iso-7, lated in time to prevent . !oss of coolant (3) TM111 EA.2 - Containment Isolation De- from being great enough to allow uncovering of (v )
pendability Positions the core. These results are true even if the longest closing time of the valve is assumed.
Amendment 6 73-39 l
ABM 234sioore Standard Plant anv. n 7.3.2.2.2 Specific Regulatory Requirements Con- dant divisions.
formance All of these signals are multiplexed and Table 7.1-2 identifies the leak detection passed through fiber optic medium before and isolation system and the associated codes and entering the voting logic of the reoundant standards applied in accordance with the Standard divisions involved in the isolation valve Review Plan. The following analysis lists the ap- logic. Separation and isolation are thus plicable criteria in order of the listing on the preserved both mechanically and electri-table, and discusses the degree of conformance cally in accordance with IEEE 279 and for each. Any exceptions or clarifications are Regrlatory Guide 1.75.
so noted.
Other requirements of IEEE 279 such as (1) 10CFR50.55a (IEEE 279): testing, bypasses, manual initiation, logic seal-in, etc., are described in Sub-The leak detection and isolation system is a section 7.3.1.1.2.
four-division system which is redundantly de-signed so that failure of any single element (2) General Design Criteria (GDC):
will not interfere with a required detection of leakage or isolation. In accordann with the Standard Review Plan for Section 7.3, and with Table All components used for the safety isolation 7.12, the following GDCs are addressed functions are qualified for the environments for the LDSt.
in which they are located (Sections 3.10 and 3.11). Most initiation parameters are repre- (a) Criteria: G DCs 2, 4,13,16,19, 20, sented by all four divisions which actuate 21, 22, 23, 24, 29, 34, 35, 38, 41, the isolation functions via two out-of four and 44.
logic permissives. Most of the sensors are provided by the nuclear boiler system. (b) Conformance: The LDS is in full compli-Theso instruments are shared by the ECCS as ance with all GDCs identified in (a) weh as the RPS and other systems which as discussed in Subsection 3.1.2.
require actuation signals from these essen-tial variables. However, each system re- The following clarification should be ceives all four signals as input to its own made with respect to GDC 23: The RPS unique voting logic incorporated in the is designed to fail in a safe state safety system logic and control (SSLC) (i.e., de-energize to actuate). This network. If individual channels are by- is also true for most isolation valves passed fcr service or testing, the voting including the MSIVs. However, the RHR logic reverts to two-out of three. and RCIC isolation valves are desigced to " fail as is" in that these are The containment is divided into four qua- motor-operated valves and require drants, each housing the electrical equip- power to both open and close. In addi-ment which, in general, corresponds to the tion, should the RHR or RCIC be in op-l mechanically separated divisions assigned to eration when valve power is lost, it each section (i.e., mechanical divisions A, essential these valves remain open so B, C, and D correspond with electrical divi- the systems can continue their safety sions I, II, III and IV, respectively). functions.
Some exceptions are necessary where a given mechanical division has more than one elec- (3) Regulatory Guides (RGs):
trical division within the quadrant. For example, the MSIVs have redundant solenoid In accordance with the Standard Review operators which require separate divisional Plan for Section 7.3, and with Table power interfaces. However, electrical 7.1-2, the following RGs are addressed for separation is maintained between the redun- the LDS:
Amendment 4 7.3-39a
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APPENDIX 7A J DESIGN RESPONSE TO APPENDIX B OF ABWR LICENSING REVIEW BASES: INSTRUMENTATION AND CONTROLS
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!. 'ABWR ustor Standard Plant __ _
w.n AJ'PENDIX 7A k TABLE OF CONTENTS s'
Section Ihle Pagg 7A DESIGN RESPONSE TO APPENDIX B OF ABWR LICENSING REVIEW -