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IPS AIINDIGI!T 11 UI'.. H. ZIICER 1:UCLEAR POWER STATION CONTEllT AI;D SUIIIf.ARY OF INSTRUCTIO!! SHEETS FOR A!EITDIEI!T 11 VOLUI4E 1 SECTION - TABLE OF CONTEIITS Pages 7, 17 through 19 Pages 2.0-ix, 2.0-xiv, 5 0-v, 7.0-1, 7.0-xv, 8.0-111, 8.0-vi, 9.o-iv, 10.0-v, 10.0-vi, 10.0-vii, 13 0-iv, 13 0-v and I.0-1 SECTION 1.0 - INTRODUCTION AND SUI 4!!ARY Pages 1.1-1, 1 3-19, 1 3-20, 1 3-21 and 1.4-1 SECTION 2.0 - SITE Pages 2.0-ix and 2.0-xiv through 2 5-84, Pages 2.2-1, 2.2-1425-87,25-8b,23-33,25-812.5-91, 2.5-91.1, 2 2 5-98, 2 5-99 and 2 5-101 Figures 2.2-2, 2.2-3, 2.2-9, 2 3-2, 2 3-3 Page 2.2-4.1, 2.2 5-1 and 2 5 4-1 VOLUIE 2 Pages 7, 17 through 19 SECTION 4.0 - REACTOR COOLANT SYSTD4 Page 4.4-1 SECTION 5 0 - CONTAINMENT Page 5 0-v Page 5.2-6 Figure 5 2-15 -
Pages 5.o-1 and 5 0-2 i
l VOLUIG 3 Pages 7,17 through 19 870924020e 870921 PDR FOIA ~1~
NENZ87-111 PDR 7 C 7 g g 7,f {,[
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?/E'!D C T l' SECTION 7.0 - CONTROL A' D INSTRUMENTATION Pages 7.0-1 cnd 7 0-xv Pages 7.1-1 7.2-3, 7.2-4, 7.2-6,
' 7.2-14,7.1-1.1, 7.1-2, 7 2 through 7.3-3,73-5,73-6 -2, 7 3-8, 7 3-12,
- 7. -13, 7 3-19, 7 3-20, 7.3-22,'7 3-23, 7 3-24, 7.p1-2, 7.4-3, 7.4-10, 7.4-15, 7.4-17, 7.4-19, 7.4-20, 7.4-31, 7 5-17, 7.8-5, 7.8-7 and 7.12-11 Figure 7 3-2 72-2,7.2-3,72-4,72-5,7.23.6-1, Page7.2 7.1-1, 3 9-1,7 2-1,4.3-1, 7 7 5 7 3 3-1, 7 5 8-1, 7.6.3-1, ;
7.8 5-1 and 7.8.5 2-1 l l
VOLUME 4 Pages 7, 17 through 19 SECTION 8.0 - ELECTRICAL POWER SYSTEM Pages 8.0-111 and 8.0-vi Pages 8.10 8.5-1 4.8.5-1.1, 8.6-1, 8.7-1, 8.7-2 8.8-1, 8.10-4, 1
8.10-14 , 8.10-7, 8.10-9, 8.10-12, $.10-12.1 and Figure 8.3-2 SECTION 9.0 - RADIOACTIVE WASTE (RADWASTE) SYSTEMS Page 9 0-iv 1 Pages 9 2-3, 9 2-5, 9 2-7, 9 2-8 Figure 9 2-1 Pages 9.4.6-1 through 9 4.6-4 SECTION 10.0 - AUXILIARY SYSTEMS Pages 10.0-v through 10.0-vii Pages 10.20-1 and 10.20-2 Figure 10.6-1 Pages 10.0-1 and 10.19-2 SECTION 12.0 - STATION S*rRUCTURES AND SHIELDING 12 3-4, 12 3-5, 12.3-21, Pages12.4-2, 12.2-1,12.4-2.1 12.2-13,and 12 3-1,4-3 12.
Pages 12.2.1.1-2, 1237-1,12.4.4-1,1251-1 SECTION 13.0 - CONDUCT OF OPERATIONS
- Pages 13 0-iv and 13 0-v Pese 13.6-2, 13.6-3 through 13 6-15 l Pages 13 0-3, 13.6.4-1 l
l SECTION 14.0 - PLANT SAFETY ANALYSIS
' Pages 14.9 1-1 and 14 9 1-2 ZPS j j AMEND! INT 11 1
VOLUME 5
)
Pages 7, 17 through 19 l APPENDIX D.0 - QUALITY CONTROL SYSTEM Pages D.0-1, D.6-12 and D.6-13
{ APPENDIX I.0 - PROCEDURES FOR THE SEISMIC ANALYSIS OF i CRITICAL HUCLEAR POWER FIANT STRUCTURES, SYSTDIS AND EQUIPMENT ;
4 Page I.0-1 Pages I.2-8 and I.10-4 l
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ZPS i
AMENDMENT 11 1 INSTRUCTIONS FOR UPDATING YOUR PSAR l
VOLUME 1 SECTION - TABLE OF CONTENTS All changes have been indicated by a vertical line and the Amendment Number (11) in the right margin of the page.
- 1. In Volume 1, SECTION - TABLE OF CONTENTS remove and destroy Page 7 and replace with amended Page 7. Remove and destroy Pages 17 through 19 and replace with amended Pages 17 through 19.
- 2. In Volume 1 SECTION - TABLE OF CONTENTS, remove and destroy the fol-lowing pages and replace them with the appropriate amended pages )
listed below: )
l Remove Pages Replace With Amended Pages 2.0-ix 2.0-ix 2.0-xiv 2.0-xiv 5.0-v 5.0-v 7.0-1 7.0-1 7.0-xv 7.0-ry 8.0-111 8.0-111 8.0-vi 8.0-vi 9.0-iv 9.0-iv 10.0-v 10.0-v 10.0-vi 10.0-vi 10.0-vii 13.0-iv 13.0-tv 13.0-v 13.0-v I.0-1 I.0-1
l ZPS l
AMENDMENT 11 VO_L,JIME I.
TABLE OF CO'.TEhTS, (Continued) i j
PACE 19.1
SUMMARY
DESCRIPTION 10.1-1
- 14.2 NIM FUEL STORACE L ~
10.2-1
'. *i l:
I l .s. 3 SPE!G FUEL STORACE T .-
10.3-1 10.4 TOOLS AND SERVICING EQUIPMENT 10.4-1 19.5 FUEL POOL COOLING AND CLEANUP SYSTEM 10.5-1 i
18.6 REACTOR BtJILDING CIDSED COOLING WATER SYSTEM 10.6-1 10.7 TURBINE BUILDING CIASED COOLING WATER SYSTEM 10.7-1 10.8 SERVICE WATER SYSTEM 10.8-1 10.9 FIRE PROTECTION SYSTEM 10.9-1 10.10 HEATING, VE!EILATION, AND AIR CONDITIONING SYSTEMS 10.10-1 10.11 MAKf. Ur WATER TREATMENT SYSTEM 10.11-1 10.12 INSTRUMENT AND SERVICE AIR SYSTEMS 10.12-1 10.13 POTABLE AND SANITARY WATER SYSTEM 4
10.13-1
! 10.14 EQUIIMENT AND FLOOR DRAINAGE SYSTEMS 10.14-1 i s l
l 10.15 PLANT PROCESS SAMPLING SYSTEM 10.15-1 1u.16 COMMUNICATION SYSTEM 10.16-1
. 10.17 LICllTING SYSTEM 10.17-1 10.18 IIEATING BOILERS 10.18-1 10.19 PRIMARY CONTAINMENT MONITORING SYSTEM 10.19-1 .
10.20 PRIMARY CONTAINMENT HYDROGEN, OXYGEN AND FISSION l
PRODUCTS SAMPLING 10.20-1 11 11.0 STEAM AND POWER CONVERSION SYSTEM TABLE OF CONTENTS 11.0-1 '
11.1
SUMMARY
DESCRIPTION 11.1-1 7
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I ZPS l
I AMENDMENT 11 LIST OF ZPS, FEBRUARY 23, 1971
, AEC QUESTIONS l
j AEC QUESTION RENUMBERED VOLUME l
l NIMBER AS QUESTION PAGE OF PSAR 2.12 '2.2.3-2 2,2.3-12 1 I 2.13 2. 3. 2.1 -2 2.3.2.1-2 1 2.14 2.3.2.1-3 ,
- 2. 3. 2.1-3 1
! 2.15- Later Later Later l 2.16 '2.3.8-1 2. 3.8 -1 1 4.9 4.7-2 4.7-2 2 9 4.10 4.7-1 4.7-1 2 i 4.11 4.9-1 4.9-1 2 9 4.12 4.0-1 4.0-1 2 1
5.11 5.2.3.7-1 5.2.3.7-1 2 5.12 10.19-1 10.19-1 2 5.13 Later Later Later 7 5.14 Later Later Later 5.15 . Later Later Later !
5.16 Later Later Later
- 5.17 5.2.3-1 5.2.3-1 2 4
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, 7.1 5.3.3.3.3-1 5.3.3.3.3-1 2 7.2 5.3.3.3.2-1 5. 3. 3. 3. 2-1 2 f
7.3 5.3.3.3.3-2 5. 3. 3. 3. 3 -2 2 1
I 7.4 7.1-1 7.1-1 3 11 7.5 Later Later Later j 7.6 4.4-1 4.4-1 2 i
7.7 Later Later Later )
i 7.8 7. 2. 3. 6 - 1 7. 2. 3. 6 -1 3 )
11 7.9 7.2-1 7.2-1 3 7.10 7.2.3.9-1 f.2.3.9-1 3 7.11 Later Later Later 17 l9 i
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ZPS AMENDMENT 11 1
LIST OF ZPS, FEBRUArt 23, 1971
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AEC QUESTIGiS, (Continued)
AEC QUESTION RENUMBERED VOLUME NUML2R AS QUESTION PAGE OF PSAR 7.12 7.2-2 7.2-2 3 11 ;
7.13 Later Later Later 7.14 '
-' Later Later Later 7.15 7.4.3-1 7.4.3-1 3
'7.16 7.5.7.3.3-1 7.5.7.3.3-1 3
_. 7.17 7.8.5-1 7.8.5-1 3 11 i
7.18 7.5.8-1 7.5.8-1 3 l
7.19 7.6.3-1 7.6.3-1 3 7.20 7.8.5.2-1 7.8.5.2 ~ 3 7.21 Later Later Later L 7.22 Later Later. Later 7.23 Later Later Later 7 7.24 Later Later Later l9 7.25 D.0-1 D.0-1 5 l11 7.26 10.10.3-1 10.10.3-1 4 7.27 7.2-3 7.2-5 3 l11 7.28 Later Later Later 7.29 10.19-2 10.19-2 4 11 i
l- 7.30 Later Later Later 7.31 Later Later Later i
j 8.1 8.3.2.1-1 8.3.2.1-1 4 I
. 8.2 8.3.2-1 8.3.2-1 4 8.3 8.3.3-1 8.3.3-1 4 8.4 8.4.3-1 8.4.3-1 4 8.5 8.5.4-1 8.5 1 4 8.6 8.4.3-2 8.4.1-2 4 8.7 8.5.3.1-1 8.5.3.1-1 4 8.o 8.0-1 8.0-1 4 18 9
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1 ZPS AMENDMENT'11 LIST OF ZPS, FEBRUARY 23, 1971 AEC QUESTIONS, (Continued)
AEC QUESTION REKUMBERED VOLUME NUMBER AS QUESTION PAGE OF PSAR 8.9 8.0-2 8.0-2 4 8.10 '8.9-1 8.9-1 4
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8.11 8.10-1 8.10-1 4 9.1 9.2.4-1 9. 2.4 -1 4 9.2 9.2.4.6-1 9.2.4.6-1 4 9.3 9.4-1 9.4-1 4 9.4 9.4.6-1 9.4.6-1 4 l 11 9.5 9.2.4.7-1 9.2.4.7-1 4 9.6 9.4.3-1 9.4.3-1 4 10.1 Later Later Later 10.2 10.5-1 10.5-1 4 10.3 10.0-1 10.0-1 4 l 11 10.4 10.11.2-1 10.11.2-1 4 7 12.22 12.6.1-1 12.6.1-1 4 12.23 12.5.6-1 12.5.6-1 4 13.1 13.0-1 13.0-1 4 13.2 13.2.1.6-1 13.2.1.6-1 4 13.3 13.2.1.2-1 13.2.1.2-1 4 9 1.1. 4 13.0-2 13.0-2 4 13.5 13.3-1 13.3-1 4 13.6 13.6.4-1 13.6.4-1 4 13.7 13.0-3 13.0-3 4 11 14.12 14.9.1-1 14.9.1-1 4 14.13 Later Later Later 1
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ZP3 AMEND!!EhT 11 TABLE OF COWENTS, (Continued)
PACE 2.5.4.4.3.1 Evaluation of Liquefaction Potential 2.5-88 2.5.4.4.3.2 Soil Liquefaction Analysis 2.5-91 2.5.4.4.3.3 Bearing capacities 2.5-91.1 11 2.5.4.4.3.4 Static Settlement 2.5-91.1 2.5.4.4.3.5 Dynamic Settlement 2.5-94 2.5.4.4.3.6 Rock - Soil - Structure Interaction 2.5-94 2.5.4.5 Subsurface Walls 2.5-94 5 !
2.5.4.6 River Bank Stability 2.5-98 2.5.4.7 Effects of Nearby Quarry Blasting on Plant construction and Operation 2.5-98 2.5.4.8 Ebture Units 2.5-99 2.5.4.9 References 2.5-100 l
i 2.5.4.10 Soil Liquefaction Analysis Report No. 43 2.5-102 i l i
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r ZPS AMENDMENT 11 SECTION 2.0 - SITE LIST OF FIGURES FICURE NUMBER TITLE 2.2-1 Plant Site Area Topography 2.2-2 Local Area Topography 2.2-3 Plant Site Aerial Photograph 2.2-4 Plant Site Topdgraphy -
l7 2.2-5 Present and Future Population Distribution 0-5 Hiles 2.2-6 Present and Future Population Distribution 0-10 Miles 2.2-7 Present and Future Population Distribution 0-50 Miles 2.2-8 Land Use Map 2.2-9 Exclusion Area Boundary 4 II 2.2-10 Exclusion Area and Dominant Station Structures l7 2.3-1 Summary of Climatetlogical Data 2.3-2 Weather Station Locations at The Wm. H. Zimmer Nuclear 7 11 Power Station Site 2.3-3 Weather Station and Smoke Release Locations (Short Term 7 gi Intensive Program) 2.3-4 Comparison of Greater Cincinnati Airport and S.ite Wind Directions
'2.3-5 Comparison of Site Wind Directions l
2.3-6 Annual Wind Rose for Greater Cincinnati Airport !
2.3-7 4 i Inversion Wind Rose for Creater Cincinnati Airport s
1 2.3-8 Maximum Single Sector Wind Persistence for Greater ,
Cincinnati Airport
{
2.3-9 Comparison for Airport PSiS and Airport Wind Speed at CVG 2.3-10 l Typical Flow Pattern Around Building '
2.3-11 Wind Tunnel Wake Pattern 2.3-12 Tower Wake Pattern at Paradise Plant 2.3-13 Tower Wake Pattern at Paradise Plant 2.3-14 Plume Rise in Neutral Stability 2.3-15 Plume Rise in F Stability i
2.0-xiv
ZPS AMENDMENT 11 SECTION 5.0 - CONTAINMENT LIST OF FIGURES FICURE NUMBER TITLE 5.2-1 Primary and Secondary Concrete Containment ,
7 Structures
- 5.2-2 Column and wall Base Detail at Floor Liner Plate
- 5.2-2.1 Drywell Floor Joint At Containment Wall l l7 1 5.2-3 Typical Section at Buttress it 5.2-4 i Tendon Access Gallery 5.2-5 Typical Leak Test Chamber r
5.2-6 Drywell Heat Attachment Detail - Tendon
-l Anchor at Drywell Head 7 j l
5.2-7 Primary Containment System Hot Process Line f Penetration j
5.2-8 Primary Containment System Cold Process Line Penetration 5.2-9 Typical Electrical Penetration Assembly 5.2-10 Personnel Access Lock 5.2-11 Drywell Cooling and Ventilation System 5.2-12 Emergency Lock and Equipment Hatch 5.2-13 Typical Layout of Hoop Tendons 5.2-14 Typical Layout of Vertical Tendons 5.2-15 Reactor Containment Development Elevation 11 5.3-1 Standby Cas Treatment System f5 5.3-2 CSCS Equipment Area Cooling System 5.3-3 Reactor Building Ventilation System h
7 5.3-4 Standby Cas Treatment System Equipment Train
{ 5.3-5 Schematic-Showing Mixing Effect of Supply Outlet 5.0-v l
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ZPS AMENDMENT 11 SECTION 7.0 - CONTROL AND INSTRUMENTATION j
TABLE OF CONTENTS l PEE I
7.0 CONTROL AND INSTRUMENTATION 7.1-1 l 7.1
SUMMARY
DESCRIPTION 7.1-1
~
l 7.1.1 Protection Systems '
7.1-1
,I 7.1.2 Power Generation Systems. 7.1-1 f 7.1.3 Protective Functions 7.1-1 l11
[ 7.1.4 Plant Operational Control 7.1-2
[ 7.1.5 Definitions 7.1-3 7.2 REACTOR PROTECTION SYSTEM 7.2-1 i
7.2.1 Safety Objective 7.2-1 7.2.2 Safety Design Basis 7.2-1 7.2.3 Description 7.2-3
- f. 7.2.3.1 Identification 7.2-3
) 7.2.3.2 Power Supply 7.2-4 7.2.3.3 Physical Arrangement 7.2-4 7.2.3.4 Logic 7.2-4 7.2.3.5 Operation 7.2-5
- 7. 2. 3.6 Scram Functioss and Settings 7.2-7 7.2.3.7 Mode Switch 7.2-11 7.2.3.8 Scram Bypasses 7.2-12 7.2.3.9 Instrumentation 7.2-13 7.2.3.'10 Wiring 7.2-17
, 7.2.4 Safety Evaluation 7.2-19 l
7.2.5 Inspection and Testing 7.2-25 l s7.2.6 Operational Nuclear Safety Requirements 7.2-27 7,2.6.1 Limiting Conditions for Operation 7.2-27 7.2.6.2 Surveillance Requirements 7.2-34 7.0 1 ,
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l ZPS
' AMENDMENT 11 i
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SECTION 7.0 - CONTROL AND INSTRUMENTATION LIST OF FICURES FIGURE NUMBER TITIE 7.1-1 Use of Protection System Control &
, Instrumentation Definitions 6 7.2-1
. Reactor Protection System Schematic Diagram 1 7.2-2 Reactor Protection System Typical Con-trol Room Panel for One Trip System 7.2-3 Reactor Protection System Functional 6
n Control Diagran
, 7.2-4 Reactor Protection System Scram Functions 7.2-5 Reactor Protection System Instrumentation 7.3-1 Nuclear Boiler System Piping & 6 Instrum2ntation Diagram 7.3-2 Main Steam Line Isolation valve 11 Schematic Control Diagram 7.3-3.1 Nuclear Eoiler System Functional
, Control Diagram, Part 1
} 7.3-3.2 Nuclear Boiler System Functional 8
Control Diagram, Part 2 7.4-1 CSCS Network Models f 7.4-2 High-Pressure Core Spray Piping &
Instrumentation Diagram 7.4-3.1 High-Pressure Core Spray System Functional Cantrol Diagram, Part 1 6 7.4-3.2 High-Pressure Core Spray System Functional Control Diagram, Part 2 7.4-3.3 High-Pressure Core Spray System Functional Control Diagram, Part 3 7.4-3.4 High-Pressure Core Spray System Functional Control Diagram, Part 4
- 7. 4-4 Instrumentation Objectives I
7.4-5 Initiation Logic - RHR B and C, HPCS, RCIC 7.4-6 Automatic Depressurization System Functional Control Diagram 7.0-xv
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l ZPS AMENDMENT 11
-TABLE OF CONTENTS, (Continued)
PAGE 8.10 DESIGN CRITERIA FOR CABLE INSTALLATION 8.10-1 8.10.1 Obj ective 8.10-1 ;
- 8.10.2 Design Basis (Ger eral criteria) 8.10-1 8.10.2.1 Definitions and Descriptions 8.10-1 8.10.2.1.1 Cable Pans 8.10-1 8.10.2.1.2 Cable Rack 8.10-2
- 8. 10.2.1.3 Conduit 8.10-2 8.10.2.1.4 Power Cables 8.10-2 8.10.2.1.5 Control cables 8.10-2 8.10.2.1.6 Instrument cable 8.10-3 8.10.2.1.7 Segregation Division (ESS and PCI) 8.10-3 8.10.2.1.8 Channel 8.10-4 8.10.2.1.9 Primary Containment Penetrations 8.10-4 4
8.10.2.2 Cable Ampacity 8. 10-4 8.10.2.3 Cable Pan Loading Criteria 8.10 l. 1 l11 8.10.3 Engineered Safeguards Systems Cable Pan Segregation Criteria 8.10-6 8.10.3.1 Design Basis 8.10-6 l 8.10.3.2 Power Supplies (Control and Power) 8.10-9 8.10.3.3 Separation Details for Pans and Conduits 8.10-10 8.10.3.3.1 Mechanical Damage (Missile) Area 8.10-10 8.10.3.3.2 Fire Protection Criteria 8.10-10 3 8.10.3.4 Cables Entering Panels 8.10-12.1 l11 8.10.3.5 Detection Circuitry for Primary Containment Isolation System 8.10-13 8.10.4 Segregation Criteria for Reactor Protection System Cables 8.10-13 8.10.4.1 General 8.10-13 8.10.4.2 Cable Separation in Reactor Protection Systeta 8.10-13 8.10.4.3 Neutron Monitoring 8.10-14 8.10.4.3.1 LPRM Inputs 8.10-14 8.0-111 1
l ZPS
,g AMENDMENT 11 SECTION 8.0 - ELECTRICAL POWER SYSTEMS l l
LIST OF FIGURES l FIGURE NUMBER TITLE 3
8.1-1 Single Line Diagram 4 8.3-1 345 KV Transmission System 8.3-2 Property Arrangement $45 KV Switchyard 7 11 and Transmission Lines
)
8.6-1 250V De Power Distribution System 125V Dc Power Distribution System 4 8.7-1 8.8-1 24V De and 48V De Power Distribution i System ,
8.9-1 120V Ac Distribution Instrument Buses and Uninterruptable Bus 7
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AMENDMEh7 11 ZPS I
SECTION 9.0 - RADIOACTIVE WASTE (RADWASTE) SYSTEMS LIST OF FIGURES FIGURE NUMBER TITIZ !
' 1 9.2 1 Liquid Radwaste System-4 7 7 Process Diagram !
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ZPS AMENDMENT 11 ,
TABLE OF CONTENTS,(Continued)
PAGE 10.20 Primary Containment Hydrogen, Oxygen and Fission Products Sampling 10.20-1 10.20.1 Safety Objective 2 10.20-1 10.20.2 Safety Design Basis _ . , ~ ~f , '
, 10.20-1 11 10.20.3 Description A: - / >: .,4 10.20-1 10.20.4 Safety Evaluation 10.20-1 L
10.20.5 Inspection and Testing 10.20-2
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ZPS AMENDMENT 11 SECTION 10.0 - AUXILIARY SYSTDtS LIST OF TABLES TABLE NUMBER TITLE PACE 10.4-1 Tools and Servicing Equipment 4
..,. 10.4-2 10.5-1 Puel Fool Cooling & Cleanup System
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10.5-2 4 Design Specs. ,
10.15-1 Typical Locations of Sampling Points 10.15-3 i
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ZPS AMENDMElff 11 SECTION 10.0 - AUIII.IARY SYSTEMS LIST OF FICURES FICURE NUMBER TITE -
10.2-1 Fuel Storage Arrangement 10.2-2 New Fuel Storage Rack
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10.3-1' Spent Fuel Storage Rack 10.5-1 Fuel' Pool Cooling and Cleanup System 10.6-1 Reactor Building Closed Cooling Water System 11 10.7-1 Turbine Building Closed Cooling Water System 10.8-1 Service Water System 10.10-1 Control Room HVAC System 10.10-2 Station Ventilation System 10.10-3 Diesel-Generator Ventilation Systym, Service Water Pump House Ventilation System and Switchgear Heat Removal System 7 l t
10.10-4 Service Water Pump House Yeatilation, Make-Up and Service Water Pump Rooms Heat Removal Systems ,
10.12-1 Station Service Air System l 4
10.12-2 Control and Instrument Air System i
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ZPS AMENDMENT 11 f
)
TABLE & CONTENTS. (Continued) j PAGE l l
13.6.4.1 General 13.6 -2 l 13.6.4.2 General Description of Site and Surrounding Terrain 13.6-3 l 2
13.6.4.3 rmergency organization 13.6-4 13.6.4.4 Liaison : Local and Civil Authorities and Other Agencies 13.6-7 j g
13.6.4.5 Protective Measures: On and off site 13.6-9 13.6.4.6 Emergency Treatment; Decontamination; Transportation 13.6 -14 13.6.4.7 Training 13.6-15 13.6.4.8 Recovery and Re-Entry 13.6-15 13.7 RECCEtDS 13.7-1 13.7.1 Initial Tests and Operations 13.7-I 13.7.2 Normal Operations 13.7-1 13.7.3 Maintenance And Testing 13.7-1 13.7.4 Other Records 13.7-2 13.8 OPERATIONAL REVIEW AND AUDITS 13.8-1 13.8.1 Adalaistrative Control 13.8-1 13.8.2 Routine Reviews 13.8-1 13.8.3 Operations Review CMttee 13.8-1 13.9 REFUELING OPERATIONS 13.9-1 13.9.1 General 13.9-1 13.9.2 Training for Refueling Operations 13.9-1 4 ,
13.9.3 Inspection Procedures 13.9-1 4
13.9.4 Emergency Procedures 13 9-1 13.9.4 Emergency Procedures 13.9-1
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I l AMENDMENT 11 I
. SECTION 13.0 - CONDUCT OF OPER2 10NS !
If LIST OF TABLES l
l TABLE NUMBER TITLE PAGE l
( 13.3-1 Pre-Operational Nuclear Train- 13,3-2 ing Tentative Course Outlime 13.6-1 Letter of Agreement 13.6-10 between CG&E and the 11 !
h Ries }bnufacturing Comptiny ,
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ZPS AMENDMENT 11 APPENDIX I.O - PROCEDURES FOR THE SEISMIC ANALYSIS OF CRITICAL NUCLEAR POWER PLANT STRUCTURES. SYSTEMS AND EQUIPMENT TABLE OF CONTENTS PAGE I.1 INTRODUCTION - r I.1-1 I.2 ANALYSIS OF BUILDINGS AND MAJOR SIRUCTURES ,
I.2-1 I.2.1 Dynamic Seisn.!c Analysis of Shear Structures (DSASS) I.2-2 I.2.2 Matrix Analysis of Seismic Stresses (MASS IV) I.2-5 I.2.3 Dynamic Analysis of Structures (DYNAS) I.2 I.2.4 Applications of Programs I.2-7 I.2.5 Damping I.2-8 ,
I.2.6 Interconnecting Class I and Class II Structures I.2-8 l gg I.3 DEVELOPMENT OF EQUIPMENT DESIGN CRITERIA I.3-1 1.4 ANALYSIS OF COMPONENTS SUPPLIED BY HANUFACIURER I.4-1 I 5 SHEIL SIRUCHTRES I.5-1 I.6 DYNAMIC SOIL PRESSURES , I.6-1 9 I.6.1 References I.6-2 I.7 DESIGN OF STRUCRTRES AND COMPONENTS FOR RELATIVE HOVEMENT EFFECTS DUE TO SEISMIC EXCITATION , I.7-1 I.8 SP.ISMIC DESIGN OF CLASS I CABLE PAN SUPPORTS I.8-1 I.9 SPRI30G-SLAB ANALYSIS ' I.9-1 I.9.1 Introduc tion I 9-1 I.9.2 Program Description I.9-1 I.10 SEISMIC DESIGN CRITERIA FOR CLASS I SYSTEMS AND EQUIPMENT I.10-1 I.10.1 Introduction I.10-1 I.10.1.1 Scope I.10-1 I.10.1.2 Definitions I.10-1 1.0-1
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ZPS AMENDMENT 11 INSTRUCTIONS FOR UPDATING YOIR PSAR WOLUME 1 l
SECTION 1.0 - INHODUCTION AND
SUMMARY
l i
t This section has been revised to incorporate minor editorial changes and -
additional information. 4 All changes have been indicated by a vertical line and the amendment number (11) in the right margin of the page.
All pages (text, tables, figures) with changes have also been marked in the upper right corner of the page with " AMENDMENT 11".
Figures that have been altered in any way are indicated by the amendment e number in the upper right corner of the figure; note that there are no other marks that would indicate changes in figure. On the page marked " LIST OF 1 FIGURES", figures that have changed in any way are designated by a vertical line with the amendment number alongside the title of the figure. See example below:
FIGURE NUMBER TITLE 2.2-1 Station Site Area Topography 11 To update your copy of t' he us. H. Zimmer Nuclear Power Station PSAR, please use the following procedure:
- 1. In Volume 1, SECTION 1.0 - INTRODUCTION AND SIDEARY, remove and destroy the fellowing text paFes and replace with the appropriate pages listed below:
l l REMOVE 'PAGE REPLACE WIN AMENDED PAGES l
1.1-1 1.1-1
'f 1.3-19 1.3-19 1.3-20 1.3-20 1.3-21 1.3-21 1.4-1 1.4-1 w_- - - - -- -- -
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1 ZPS j AMENDMENT 11 l SECTION 1.0 - INTRODUCTION AhD SIMtARY 1.1 PROJECT IDENTIFICATION ;
1.1.1 Introduction and Identification of Site This Preliminary Safety Analysis Report (PSAR) is subr.itted in support i of the application of the Cincinnati Cas & Electric Company, Columbus & Southern 5 Ohio Electric Company, and the Dayton Power and Light Company for a construction
, permit and facility license for a nuclear power station to be located at a site in Washington Township, Clermont County, Ohio. The information developed in this PSAR relates to a reactor unit, rated at a power level of 2436 MWt. The i site has been planned for additions of other nuclear units and certain provisions l1 j will be made to accomodate these future additions, but this PSAR covers only the ' '
l first unit designated as the Wm. H. Zimmer Nuclear Power Station, Unit 1. 4 The Um. H. Zimer Nuclear Power Station is scheduled to operate at a
- j}
j gross el ctrical power output of approximately 840 MWe and a net electricel power l4 jf output of approximately 807 MWe when the condenser is operating at 1.5 inches Hg Abs.
i
. The Ha. H. Zimer Nuclear Power Station will use a s *ngle cycle, forced 4
'] circulation, boiling water reactor similar to the Edwin I. Hatch Nuclear Plant,
[ Unit 1, now under construction by the Georgia Power Company. The Docket Number i of the Hatch Plant is 50-321. Its Permit Number is CPPR-65, issued September 30, 1 1969.
I j 1.1.2 Identification and Qualifications of Applicant and Contractors l 1.1.2.1 Applicant J
} The Cincinnati Cas & Electric Company (CG&E), Columbus & Southern Ohio Electric Company (C&SOE), and the Dayton Power and Light Company (DPL), are the
! applicants for the construction permit and facility license for the Wm. H. Zimmer j
! Nuclear Power Station Unit 1, near Moscow, Clermont County, Ohio, 25 miles south- 4 l
, east of Cincinnati on the Ohio River. The three companies will own the Station and Unit 1 as tenants in common, with the tentative percentages shares of owner-
, ship being as follows:
CG&E 377.
C&SOE 327. l DPL 31%
Six co=monly owned fossil-fueled units are scheduled for operation prior to the startup of the Wm. H. Zimer Nuclear Power Station Unit 1.
Unit 1 is scheduled for commercial operation January 1,1975. To meet 4 this schedule, a construction permit will be needed by April 1,1971.
1.1-1 i
ZPS AMENDMENT 11 TABLE 1.3-7 SAFETY AND SAFETY-REL HD SYSTEMS AND ESSENTIAL COMPONENTS OF SAFETY AND SAFETY-REIATED SYSTEMS Discussion in PSAR Subsection or Paragraph
\
- 1. High-Pressure Core Spray (HPCS) i
- a. HPCS pump. . . . . . ............. 6.4.1
- b. Diesel-generator associated with HPCS .... 8.5.1 - 8.5.3
- c. Associated piping, valves, controls ..... 6.4.2 j instrumentation and electrical equipment
- 2. Low Pressure Core Spray (LPCS)
- a. LPCS pump. . . . . . . . . . . . . . . . . . . 6.5.3.3 l
- b. Associated piping, valves, controls, instrumentation and electrical equipment ... 6.4.3 - 6.4.4
- 3. Residual Heat Removal (RHR) . . . . . . . . . . . 4.8
- a. RHR pumps. . . . . . . . . . . . . . . . . . . 4.8
- b. RHR heat exchangers . . . . . . . ...... 4.8
- c. Associated piping, valves, controls, instrumentation and electrical equipment . .. 4.8 4 Reactor Core Isolation Cooling System ..... . 4.7
- a. RCIC pump. . . . . ........ ...... 4.7
- b. RCIC turbine .... ....... ..... . 4.7
- c. Associated piping, valves, control instrumentation and electrical equipment ... 0.7 l
l S. Reactor Building Closed Cooling Water (RBCCW) .. 10.6 1 Piping, valves, controls instrumentations and electrical equipment associated with the l
RBCCW system providing cooling to:
i
- a. Three (3) RHR pump motor seal coolers . . . . . 10.6
- b. Room coolers for core standby cooling systems (CSCS) equipment .......... . 10.6 -5.3 l11 (1) Northeast corner room (RHR pump snd LPCS pump)
(2) Northwest corner room (2RHR pumps)
(3) Southeast corner room (HPCS pump) l11 1.3-19
ZPS 7
AMENDMENT 11 TABLE 1.3-7, (Continued)
Discussion in PSAR Subsection or Parag raph
- c. RCIC equipment room cooler (southwest j corner) . . . . . . . . . . . . . . . . . . . . 10.6 -
5.3 111
- d. Cont rol roca and auxiliary electrical l equipment room HVAC . . . . . . . . . . . . . . 10.6 - 10.10 11
- 6. Service t'-te r Sys tem (SWS) . . . . . . . . . . . . 10. 8
- a. Pumps ....... . . . . . . . . . . . . . 10.8
- b. Strainers . . . . . . . . . . . . . . . . . . 10.8
- c. Piping, valves, controls, instrumentation . . 10.8 and electrical equipment associated with the SWS providing water to: !
(1) Diesel generators l l (2) ERR heat exchangers (3) KBCCW heat exchangers l l
(4) Service water pump room coolers ~
- 7. Automatic Depressurization System (ADS) . . . . . 4.4.5 - 7.4.3.3
- 8. Reactor Protection System . . . . . . . . . . . . 7.2
{
- 9. Primary Containment and Reactor Vessel l
! Isolation Control System . . . . . . . . . . . . . 7.3 7 1
- 10. S tandby Liquid Control . . . . . . . . . . . . . . 3.8 I
- a. Tank . . . . . . . . . . . . . . . . . . . . . 3.8
- b. Pumps ....... . . . . . . . . . . . . . 3.8
- c. Explosive valves . . . . . . . . . . . . . . . 3.8 '
- d. Associated piping, valves, heat tracing, controls, instrumentation and electrical 1 equipment. ..... . . . . . . . . . . . . . 3.8 I
- 11. Standby Cas Treatment and Reactor Euilding l Recirculation ..... . . . . . . . . . . . . . 5.3.3.3.3-5.3.4.4 111
- a. Fans . ....... . . . . . . . . . . . . . 5.3.3.3.3 I
- b. Filters ....... ............5.3.3.3.3
- c. Associated duct work, pipe, valves, dampers, controls, instrumentation and electrical e ,uipment. . . . . . . . . . . .. ... . 5.3.3.3.3
. )
1.3-20 i
l
7 i
ZPS AMEND {ENT 11 IABLE 1,3-7, (Continued)
Discussion in PSAR l i
Subsection or Pa rag raph 12.125V de Power ' System . ..............8.7
- a. Batteries
- b. Battery chargers ,
- 13. 250V de Power System . . . . . '. . . . . . . . . . .~8.6
- a. Batteries
~'~
- b. Battery chargers -
- 14. Essential ventilation and cooling systems for:
- a. Di e s e l-g e ne ra t or ro oms . . . . . . . . . . . . . 10.10.3.6
- b. Control room and auxiliary electrical equipment room (HEAC) . . . . . . . . . . . . . . 10.10.3.2
- c. Essential switchgear rooms and battery rooms . . . . . . . . . . . . . . . . . . . . . 10.10.3. 3
- d. Service water pump room. . . . . . . . . . . . 10.10. 3. 7,
- e. CSCS equipment rooms: . . . . . . . . . . . . 5.3.3.3.2 (1) Northeast corner room 7 (2) Northwest corner room (3) Southeast corner room til
- f. RCIC equipment room (southwest corner) I
- g. Ductwork, pipe, valves, dampers; controls, instrumentation and electrical equipment associated with ventilation of above areas
- 15. Neutron Monitoring System (portion required to detect overpower condition and provide sdram signal) . . 7. 5. 5. 3.5-7. 5. 7.1 11 1
- 16. Main Steam Radiation Monitoring System . . . . . . 7.12.1
- 17. Standby ac Power System . . . . . . . . . . . . . 8.5
- a. Diesel-generator for HPCS . . . . . . . . . . 8.5
- b. Diesel generators for Bus 1-3 and Bus 1-4 . . 8.5
- c. Diesel-generator water pumps . . . . . . . . . 8.5
- d. Diesel-generator fuel supply system (in- <
cluding transfer pumps and storage tanks) . . 8. 5
- e. Diesel . carting air system (including compr,ssors) . . . . . . . . . . . . . . . . . 8.5 4
1.3-21
ZpS AMENDMEhT 11 1.4 SITE AND ENVIRONMEffT DESCRIPTION 1.4.1 Location and Size of Site The site is located on the Ohio River in Washington Township, Clermont County, Ohio. It is approximately 24 miles southeast of downtown Cincinnati, Ohio and approximately one-half alle north of Moscow, Ohio. Centerline co-ordinates of the reactor area are 38' 51' 55" north latitude and 86* 13' 45" west longitude. Figures 1.4-1, and 1.4-2, show the relation of the site to the surrounding areas and the site boundaries. Figure 1.4-3 shows the site area topography. 4 1.4.1.1 Site ownership The site is Wiolly owned by the Cincinnati Cas & Electric company and '
consists of approximately 635 acres. O gg 1.4.1.2 Access to the Site Several transportation facilities provide access to the site. U. S.
Highway 52 passes through the site, about one-half saile east of the plant location. The Chesapeake and Ohio Railroad traverses the Kentucky side of the Ohio River, within several hundred feet of the shore. The Ohio River is of course navigable at this location. ,
1 1.4.2 Description of the Environs !
- a. General i
, I The western portion of the site, between U. S. 52 and the Ohio River, 1 is primarily farmland and is relatively level at 500 feet elevation above sea
,' level. The central area of this portion is cut by a small stream with a bed elevation of about 470 feet. The remainder of the site area is someidiat hilly and partially wooded, with elevations varying up to 800 feet. The average grade is approximately 495 feet above mean sea level.
- b. Population Within a ten-mile radius at the site the 1960 population was 21,526, estimated to increase to 30,164 by 1985. The largest city within this radius is New Richmond, Ohio, with a 1960 population of 2,834.
1 The largest city within 50 miles of the site is Cincinnati, Ohio with a 1960 population of 502,500.
- c. Land Use Within ten miles of the site are included portions of four counties, three of which are in Kentucky. In 1964 over half of the land area of each of these counties was being used for farming endeavors. Approximately one-third 1.4-1
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[ AMENDMENT 11 I
i P INSTRUCTIONS FOR UPDATING YOUR PSAR VOLINE F /
- SECTION 2.0 - SITE V
F This section has been revised to reflect new information and minor editorial changes. .
All changes have been indicated by a vertical line and the amendment
[ number (11) in the right margin of the page.
I.
- All pages (text, tables, figures) with changes have also been marked in the upper right corner of the page with " AMENDMENT 11".
Figures that have been altered in any way are iudicated by the amendment number in the upper right corner of the figure; note that there are no other marks that would indicate changes in figure. On the page marked " LIST OF FIGURES", figures that have changed in any way are designated by a vertical line with the amendment number alongside the title of the figure. See example below:
FIGURE NINBER TITLE
? 2.2-1 Station Site Area Topography l 11
) To update your copy of the Wm. H. Zinaner Nuclear Power Station PSAR, please use the following procedure:
- 1. In Volume 1, SECTION 2.0 - SITE, remove and destroy Table of Contents 'Pages 2.0-ix and 2.0-xiv and replace with amended Pages 2.0-ix and 2.0-xiv.
- 2. In Volume 1, SECTION 2.0 - SITE, remove and destroy the following text pages and replace with the appropriate pages listed below:
REMOVE PAGE REPLACE WITH AMENDED PAGE f 2.2-1 2.2-1 2.2-14 2.2-14 2.3-33 2.3-33 2.5-81 through 2.5-84 2.5-81 through 2.5-84 2.5-87 2.5-87 2.5-88 2.5-88 2.5-91 2.5-91 t
2.5-91.1 l
___________.______m _ _ . _
ZPS AMENDMENT 11 REMOVE PAGE REPIACE WITH AMENDED PAGE 2.5-92 2.5-92 2.5-94 2.5-94 1
2.5-98 2.5-98 2.5-99 2.5-99
~
2.5-101 2.5-101
- 3. In Volume 1, SECTION 2.0 - SITE, remove and destroy the following figures and replace with the appropriate figures listed below:
REMOVE FIGURE REPIACE WITH AMENDED FIGtRE 2.2-2 2.2-2 2.2-3 2.2-3 2.2-9 2.2-9 2.3-2 2.3-2 2.3-3 2.3-3
- 4. In Volume 1, SECTION 2.0 - SITE, behind the red tabbed divider page titled " Amendments to Section 2.0" remove and destroy Pages 2.2.4-1, 2.2.5-1, and 2.5.4-1 and replace with Amended Pages 2. 2-4.1, 2. 2. 5-1 and 2.5.4-1.
I ZPS AMEND."EhT 11 TAH.E OF CONTENTS, (Continued)
PACE 2.5.4.4.3.1 Evaluation of 1. liquefaction Potential 2.5-88 2.5.4.4.3.2 Soil Liquefaction Analysis 2.5-91 2.5.4.4.3.3 Bearing capacities 2.5-91.1 2.5.4.4.3.4 Static Sectiement 2.5-91.1 2.5.4.4.3.5 Dynamic Settlement 2.5-94 2.5.4.4.3.6 Rock - Soil - Structure Interaction 2.5-94 .
2,5.4.5 Subsurface Walls 2.5-94 5 2.5.4.6 River Bank Stability 2.5-98 2.5.4.7 Effects of Nearby Quarry Blasting on Plant Construction and Operation 2.5-98 2.5.4.8 ruture Units 2.5-99
]
2.5.4.9 References 2.5-200 2.5.4.10 Soil Liquefaction Analysis Report No. 43 2.5-102 l
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ZPS AMENDMENT 11 SECTION 2.0 - SITE LIST OF FICITES ,
FIGURE NUMBER TITLE 2.'2- 1 Plant Site Area Topography i 2.2-2 Local Area Topography 7 2.2-3 Plaar Site Aerial Photograph .
~ ~ '
2.2-4 Plant Site' Topography l7 2.2-5 Present and Future Population Distribution 0-5 Miles i 2.2-6 Present and Future Population Distribution 0-10 Miles 2.2-7 Present and Future Population Distribution 0-50 Miles 2.2-8 Land Use Map 2.2-9 Exclusion Area Boundary 0 l1 2.2'-10 Exclusion Area and Dominant Station Structures l7 2.3-1 Sunnary of Climatological Data 2.3-2 Weather Station Locations at The Wm. H. Zinner Nuclear Power Station Site 7, 11 2.3-3 , Weather Station and Smoke Release Locations (Short Term 7 gy
, . Intensive Program) 2.3-4 Comparison of Creater Cincinnati Airport and Site Wind Directions ,
' 2. 3-bi . Comparison of Site Wind Directions l
2.3-6 U Annuci Wind Rose for Greater Cincinnati Airport 2.3-7 4 :
Inversion Wind Rose for Greater Cincinnati Airport
'/ ' '
2.3-8 Maximum Single Sector Wind Persistence for Creater i
", +
Cincinnati Airport j I 2.3-9 Comparison for Airport HWS and Airport Wind Speed at CVG 2.3-10 Typical Flow Pattern Around Building
>' j 2.3-11 Wind Tunnel Wake Pattern j 2.3-12 Towcq Wake Pattern at Paradise Plant !
2.3-13 Tower Wake Pattern at Paradise Plant '
2.3-14 , Plume Rise in Neutral Stability 2.3-15 Plume Rise in F Stability
[
2.0-xiv
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I ZPS I AMENDMENT 11 I
I 2.2 SITE DESCRIPTION 2.2.1' Location The site, consisting of 635 a:res, is located 24 miles southeast' of i
Cincinnati, Moscow, Ohio.
Ohio, on the Ohio side of the Ohio River and 2,700 feet ' north of 111 l1 l State of Ohio. The site is in Washington Township, Clermont County, in the Figure l
} ares topography out to 5 2 2-1 illustrates the relationship of the plant site to miles.
Figure 2.2-2. An aerial photograph The local area topography is illustrated by I of the general area with the site outlined is shown on Figure 2.2-3. Figure 2.2-4 shows the site topography. j i
2.2.2 Population
)
f The distribution of the present and future population within 5,10, and i
50 miles of the site is shown on Figures 2. 2-5, 2. 2-6, and 2. 2-7. The popula-tion distribution areas are comprised of sixteen, 22 degree, radial sectors with the appropriate mileage increments.
In 1960 4,572 persons lived within a 5-mile radius of the site. 'Itte estimated be 5,775. 1970 population is 4,937 and by 1985 the population is estimated to j The nearest 1960 population of 438. town within the 5-mile radius is Moscow, Ohio, with a ]
I The 1960 population in the 0-to 10-mile radius was 21.526 and the 1970 projection is 24,176. It is estimated that by 1985 30,164 people will live in t this area. Table 2.2-1 shows the population and location cf all cities within ten miles of the site. As indicated in Table 2.2-1, the only city with a 1960 i population over 1,000 in the 10-mile radius is New Richmond, Ohio.
The total 1960 population within 50 miles of the site was 1,705,535.
This 1985. figure is estimated to increase to 2,043,363 by 1970 and to 2,688,025 by A listing of cities with populations over 10,000 within the 50-mile radius is shown in Table 2.2-2 The closest city with a population over 10,000 is Fort Thomas, Kentucky. Table 2.2-3 presents the. population density within the 50-mile radius. <
The list of references used to obtain the population data and projec-tions can be found in Paragraph 2.2.4.
2.2.3 Land Use
! The plant is to be located on the east bank of the Ohio River, about one-half mile north of the city of Moscow, Ohio, in Clermont County. Portions of four different counties are included within ten miles of the site. Besides Clermont, in Kentucky. these counties are Campbell, Pendleton, and Bracken, all of which are In 1964, as can be seen in Table 2.2-4, over half of the total area of each of these four counties was being used for farms and farming 2.2-1
/
I I
ZPS f
AMENDMENT 11 2.2.5 Low Population Zone ,
la l
The Low Population Zone (LPZ) is indicated in Figure 1.4-3 and a description of the activity within the LPZ is covered in Paragraph 13.6.4. The gg evacuation procedures to be followed in the LPZ af ter a reactor incident are described in Paragraph 13.6.4.5.
y
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2.2-14 i
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ZPS AMENDMENT 11 I
t r R V l Case F c c 4
Neutral (AM) 0.331 0.311 14.8 m/s l
Neutral (PM) 0.597 0.378 12.2 {
Stable 0.305 0.303 15.2 1
4 From the above, for wind speeds less than 12 m/s, the effects of cooling tower !
l wake on the plume can be neglected. j Of major concern as to plume effects is the probability of the plume reaching ground causing both fog and/or icing. Two primary terrain features of I concern near the tower are: the proposed switching yard approximately 1.5 km northeast of the tower and the city of Moscow, Ohio, about 1 km southeast of the l7 1 tower. From the above considerations and calculations, and assuming adequate drift eliminators have been installed on the tower, the plume will not reach ground if the wind speed is below about 25 mph for neutral conditions and 30 mph k for stable air. For wind speeds greater than 25 mph, the plume will be caught 4 )
in the wake and be brought down to the ground at a distat.ce of two to four tower i heights away. For unstable air, the plume will be of such short length so as to '
be of no concern. Since category F does not occur with such high winds, this ,
situation can be safely neglected. The plume from the cooling tower even under relatively poor diffusion conditions will rise to between 500 and 600 meters above the base of the tower at a downwind distance of 1 km. This will be well above any nearby terrain feature or plant structures. Additional calculations
, have shown that for wind speeds greater than 8 m/s the plume may be caught in
.l the tower's aerodynamic wake and brought to the ground. However, with such high 7
wind speeds at the site, the plume will diffuse quite rapidly within a short j 1
- - distance of the tower and should be of no concern. In addition, wind speeds of j greater than 25 mph will occur less than 1 percent of the time. Fogging and j icing is a normal phenomenon which occurs at many power plant sites. Over many i
l years of operation no adverse effects have been recorded with the exception of )
ice accumuisting on transmission lines. Details of the design criteria for ice 11
- e loading on transmission lines are discussed in Paragraph 8.3.3. The operation of the cooling tower will add an incrementally small amount of fogging and icing i l over and above that which will occur normally. In summary, it can be concluded 4 that the plume from the cooling tower will have little effect on the surrounding areas and the plant safety will not be affected.
A failure of the Class-II cooling tower will not damage any plant struc-
- tures essential to a safe plant shutdown. For example, assuming an improbable failure, that is, one in which the tower tips over, the circular area in which 7
the cooling tower could fall has a radius of 700 f t (see Figure 1.5-1). No Class-I structures or other essential equipment are located in this 700 f t radius. Any lesser structural failure of the tower would have a smaller fall radius than that shown in Figure 1.5-1.
2.3.9 Conclusions 4
The climatological data from Greater Cincinnati Airport (CVG) has been i
2.3-33
i ZPS A.ENDMENT 11 2.5.4 Foundations
.: . 5. 4.1 Summary The site is considered suitable from a foundations standpoint, for the construction of the nuclear power station facilities. Major plant structures will be supported on mat foundations. Mat loadings and foundation grades are 5:
shown in Table 2.5-16, and ultimate bearing capabilities for mat foundations are ,
summarized in Table 2.5-19.
A comprehensive evaluation of the behavior of the on-site soils under earthquake loading indicates a low margin of safety against potential liquefac-tion for sandy soils above elevation 450. Foundation preparation will consist of densification of soils between foundation level and elevation 450, by de-watering, excavating, and recompacting existing soils.
Assuming densification of soils above elevation 450, total and differen-tial settlements of mat foundations supporting various units will be structurally and operationally permissible. Computed settlements are given in Tables 2.5-20 and 2.5-21, 5
No unusual problems are associated with site preparation. A dewatering system will assure that construction operations are performed under dry condi-tions during normal and flood stages of the Ohio River. Structural fill and backfill materials will consist of compacted on-site granular soils placed under engineering supervision. General fill materials required to raise the site to elevation 520 stage will be obtained from on-site and of f-site sources.
Consideration was given to alternative forms of foundations, however, mat foundations are considered to be the most satisfactory form of foundation support. 1 2.5.4.2 Planned Construction Construction at the plant site will consist of one reactor unit and ap-purtenant facilities. Finished plant grade has been established at elevation 520. (Mean Sea Level Datum); attaining final grade will require the placement of approximately 25 f t of fill.
Pertinent data regarding major plant structures are presented in Table 2.5-16. The location and arrangement of the major structures are shown on Figure 12.5-1.
The reactor unit will be housed within the reactor building; its founda-tion elevations tabulated above are the elevations of the tendon access tunnels.
An approximately 16 foot wide radwaste tunnel will join the reactor unit to the 5
!' radwaste building. This tunnel will have a base elevation ranging from 476 to i 11 l I
2.5-81
)
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ZPS 1
1 AMENDMENT.11 l
TABIE 2.5-16 STRUCTURAL IDADING CONDITIONS I DIMENSIONS MUNDATION NUNDATION IDADING STRUCTURE (FEET) EIEVATION (KIPS /SO.FT.)*
Reactor 163 x 145 468'6" : 9,200 Building 1
Auxiliary 106 x 93 466'5" 5,800 Building 64 x 93 489'0" 5,000 Diesel Generator 56 x 105 496'0" 5,000 Building 11 ,
Turbine 350 x 130 466'5" 5,400 Building 489'0" 4,450 Heater Bay 280 x 65 489'0" 3,700 466'5" 3,700 l f 473'0" 3,700 Radwaste 153 x 78 489'0" 3,640
- Dead + Live feads i
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1 2.5-82
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ZPS ;
AMENDMENT 11 l
481. The central portion of the Turbine Building will have a base elevation of 466'5". .The northern and southern sections will be founded at elevation 489.
The Heater Bay is divided into approximately three equal sections with base L 11 elevations of 489, 466'5" and 473. l Auxiliary structures including tanks, service area, warehouse, circulat-ing water pump house and offices will be supported at shallow depths. l 2.5.4.3 Site conditions 2.5.4.3.1 Surface Conditions The proposed Unit I construction area is presently an undulating field blanketed throughout the western two-thirds of the area by a thf ek mat of long grass, and throughout the eastern third by trees which vary up to six inches in i' diameter. The ground surface level generally ranges between elevation 490 and elevatiun 500; however, a small irrigation channel with a lowest level of eleva-tion 480 encroaches onto the northeastern corner of the' construction area. Top-soil with heavy roots extends to an average of nine inches below existing ground surface within the grassed area and to an average of 15 inches below ground l surface within the wooded area.
The clayey silt and silty cicy stratum is the alluvium described in Paragraph 2.5.2.4. Throughout the construction area, the alluvium is a stiff to very stiff predominantly clayey silt soil with random lenses and seams of clean or silty fine sand. Random thin seams of hard carbonaceous material are present in the stratum.
The sands are glacial outwash material. The upper 10 to 20 feet of sand l is predominantly fine to coarse sand containing a trace of silt and some fine gravel. Approximately 60 percent of this upper sand is in the medium size range.
The observed relative density of this layer is variable. The progressively deeper 10 to 20 foot thickness of sand is predominantly a medium sand containing 5 some coarse sand and some fine cand. Below elevation 450 the sand grading is I generally fine to coarse containing some gravel. I The thin stratum of sand which immediately overlies the bedrock is a fine to coarse sand containing a considerable quantity of fine to medium gravel. ,
The s11t content of this stratum varies up to approximately ten percent. This i fine to coarse sand stratum is very dense and has a relative density in excess l 2
of 80 percent. The relative densities for the various strata are listed in Table 2.5-18.
- i. A description and discussion of the bedrock is presented in Paragraph I l 2.5.2.4. In the specific plant area, the bedrock immediately underlying the soil overburden is a crystalline fossiliferous limestone with irregular shale ,
3 I
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1 l
)
l 2.5-83 i
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1 l
1 l 1
t ZPS f
3 AMENDENT 11 ,
. ,f partings which vary up to three inches in thickness. Throughout the upper 20 t feet of rock the total volume of shale varies from approximately 3 to 17 per-cent. 5!
Generally, the limestone is closely fractured and has bedding between !
layers which is wavey and irregular. I There is a relatively steep partially eroded bank adjacent to the Ohio normal poolthe River and river has a surface water elevation at approximately 455 during stage.
2.5.4.3.2 Subsurface conditions l
This section is intended as a modification and extension of the Paragraph 2.5.2.4.2.
Borings drilled within the area of the proposed unit structures at the locations shown on Figure 2.5-2.2, reveal.ed that the site is underlain by soil i
and rock in the sequence tabulated in Table 2.5-17, 2.5.4.3.3 Ground Water 2.4.2.3. Site and regional ground water conditions are discussed in Paragraph Water level measurements within the alluvium or glacial outwash mate-rials in the immediate plant area are discussed in this section together with ground water effects on subsurface design and construction. 5 3
Ground water level observations were made during and subsequent to drill-ing operations. For this purpose, a total of 21 well point piezometers were installed throughout level. In October 1969 the general site areas to monitor the ground water surface the surface of the ground water was observed at between elevation 456 and 458 throughout the proposed structure area. An isolated ob-f servation of perched water with a surface elevation of 477 ft also was made. In i
May 1970 'in the same area the ground water surface was observed bretween 465 and 466 ft; approximately two weeks later the ground water surface was observed at selevations which ranged from elevation 461 to 465 ft. Similarly, during this r
later at observation elevation 476 fperiod,
- t. another perched water level was observed in one boring ed at approximately elevation 467.In settlement and bearing capacity analyses, gro In bearing capacity analysis, the ground I water condition resulting in the lowest factor of safety for each structure was celected; water Icvel elevations ranging from 455 to 520 were considered in the y1 analysis.
For liquefaction analyses, the ground water level has been assumed to be at elevation 508.6. i 2 5.4.3.4 Frost The maximum depth of frost penetration in the vicinity of the site is on the order of 30 inches.
2.5-84
l ZPS I
AMENDMENT 11 elevation 450 and the planned base elevation of the foundation. For the reactor l5 l5 building, the excavation will extend laterally an additional 30 feet from the
- distance indicated above. ,
l 5
Prior to excavation below the ground water table, the area will be de- i
! watered as outlined in Paragraph 2.5.4.4.2.4. All excavation faces will be cut
! with a slope no steeper than 1-1/2:1 (horizontal: vertical). A minimum factor of safety against slope failure of 1.4 has been calculated for this condition. l-2.5.4.4.2.4 Dewaterina Dewatering will be accomplished in a manner which will allow earthwork operations to be performed under dry conditions. For this purpose, well-points and/or a system of deep wells will be utilized during all excavating, backfill-ing, and construction operations below the ground water table. The water sur-face throughout the entire excavation area will be maintained at Icast three feet below the working surface of the excavation. The dewatering system will be designed and operated with sufficient reserve capacity to maintain water levels at desired elevations during flood stages of the Ohio River.
Observation wells will be installed at selected locations throughout the construction area so that the ground water level can be continuously monitored.
l Staged shut-down of the dewatering system will be allowed as construction at lower elevations is completed. l3 2.5.4.4.2.5 Excavation Base Treatment i
Following dewatering and excavaidt ; to elevation 450, the exposed base vill be inspected to assure that sand anu/or sand with gravel is exposed through-
- out. Sof t, loose or diturbed materials at tne bass of the excavation will be
' removed. The in-situ undisturbed soil will be tested to verify that its in-place relative density is within the range measured by field investigation studies. The exposed surface at the base of the excavation will then be compact- 5 ed to a relative density of at least 85 per cent as determined by ASIM Test Designation - 2049- 64T.
2 5.4.4.2.6 Structural Fill and Backfill a
- Placement of compacted fill will be required above elevation 450 (a) to attain desired foundation grades; and (b) to backfill adjacent to substructure walls. All fill which is placed for foundation support or as structure backiill will be composed of clean sand or sand with gravel. Fill naterial will be 4 placed in thin, near horizontal lif ts and each lift will be compacted to a minimum relative density of 85 per cent as determined by ASTM Test Designation 2049 -64T. The placement and compaction of fill and backfill will be continuously supervised and in-place density tests will be performed in each lift to verify satisfactory compaction prior to the placement of a subsequent lif t. Comple ted fills which are exposed to disturbance during construction operations will be recompacted prior to foundation construction to density loosened surface soils.
i l
2.5-87 )
i
l ZPS f
a AMENDMENT 11 l
I 2.5.4.4.2.7 General Fill j5 0,1 Approximately 25 feet of fill will be placed to raise the existing site grade to elevation 520. All fill which will support structures and all backfill which will influence structures will be placed and compacted in the manner described in Paragraph 2.5.4.4.2.6. However, fill which will have no influence on structures will be designated as general fill. General fill vill consist of clayey silt and silty clay soils excavated from the construction area and off-g
' site borrow materials. These off-site borrow materials may include sand dredged from the Ohio River and placed by hydraulic fill methods. Fill materials and placement methods will be selected to satisfy pertinent design requirements. 11 2 5.4.4.3 Mat Foundations l5 2.5.4.4.3.1 Evaluation of Liquefaction Potential Detailed analysis of the liquefaction potential of on-site granular soils has been performed to verify that sands supporting Class.I structures and equipment will not liquefy during the design basis earthquake. The method of performing these studies, data and assumptions used in analysis, and the results obtained are stanmarized below:
- a. Soils at the site were investigated by standard penetration testing and by undisturbed sampling of in-situ soils. Predominantly granular soils, which are difficult to sample in an undisturbed j state, were successfully recovered by use of a special cryogenic j sampler in which soil was retair.ed in the sampling device by a l frozen plug of soil in the sampler bit.
I
! b. Standard peactration test results were converted to relative densi-l.
ties by methods porposed by Bazarra (1967) and Gibbs and Holtz i
(1957). Results of these conversions are shown on Figure 2.5-9(A) and Figure 2.5-9(B) . 5
- c. The relative densities of undisturbed samples were determined by comparing the in-place density of the recovered sample with the !
maximim and minimum density of identical or comparable soils. The results of these relative density determinations are shown on Figure 2.5-9(C). l5
- d. Based on the comparison of all data shown on Figure 2.5-9, it was concluded that the Gibbs and Holtz method of conversion had been confirmed for soils at this site. The Bazarra method was not considered further in liquefaction analysis. Soil density l
i 2.5-88 i l
l i
_ _ _ _ _ Y
( ZPS i '
AMENDMENT 11 l3 2.5.4.4.3.2 Soil Liquefaction Analysis Soil liquefaction is a phenomenon which occurs when loose, saturated, granular 4. oils are subjected to vibratory loading. Using the soil properties found at the site together with the appropriate seismic excitation, this phenom-enon has been investigated. It has been determined that, in all cases investi-gated, the induced stresses across the soil profile are less than those required for liquefaction.
7 l For a better understanding the full text of the Sargent & Lundy Report l
- No. 43, " at included Soil theLiquefaction Analysis" end of Subsection forParagraph 2.5, Wm. H. Zimmer Nuclear Power Station, isl 5 2.5.4.10. ,
I l Based on the above analysis, granular soils at the site are considered to have an insufficient margin of safety against liquefaction above elevation f 450. Soils below elevation 450 indicate an acceptable margin of safety by the above outlined analysis, and reference to other published data confirms this l conclusion. (Ambraseys and Sarma,1969: Castro,1969; Seed,1969).
Conclusions resulting from the above outlined study have been used in 5 selecting the method of foundation support; namely, mat foundation support on a prepared base of granular soils compacted to a relative density of 85 per cent and extending to elevation 450. Generalized properties of soil strata at the site af ter compaction of foundation materials and placement of fill and backfill materials adjacent to structures, are presented in Table 2 5-18.
Although the compacted soils immediately below mat foundations will be stable under earthquake loading, the soils above elevation 450 and outside the compacted zone were assumed to be susceptible to liquefaction during the design {
l basis earthquake. Analyses were performed to evaluate the possibility that the j j liquefaction of these loose adjacent soils might affect the stability of the i
compacted soils and the foundations supported thereon. The analytical procedure used in these studies was a pseudo-dynamic slope stability analysis with the following principal assumptions:
11
- a. Ground surface elevation at 520; water level elevation at 508.6.
- b. No shearing resistance in the liquefied soils.
- c. Weight of saturated liquefied soils equal to 110 pcf.
- d. Horizontal acceleration of 0.2g throughout the soil profile.
2.5-91 R
2PS AMEhDMEhT 11
- c. Horizontal dynamic force equal to 0.2 s; vertical load at any point in the soil profile, applied as an extra-static force in the direc-tion of slope instability.
- f. Maximum foundation pressures for dynamic loading equal to 150 per cent of dead plus live static loading tabulated in Table 2.5-16.
i l
The results of the stability analyses indicate that all Class I struc-tures, other than the Reactor Building, will be stabic during the postulated de-sign basis earthquake provided that the compacted fill underlying foundations extends a minimum lateral distance beyond foundation lines equal to the vertical 11 distance between elevation 450 and the planned base elevation of the foundation.
For the Reactor Building, the bottom of the compacted fill will extend an addi-tional 30 feet from the distance indicated above. The resulting factor of safety for this condition is 1.0 for the Reactor Building, and in excess of 1.0 for other Class I structures.
1 A factor of safety equal to or greater than 1.0 is considered acceptable '
given the very conservative nature of the assumptions including: (a) the assump-tion of a zone of liquefied soils extending from elevation 450 to the ground surface; (b) the use of a horizontal acceleration of 0.2g throughout the soil profile. (A horizontal acceleration of 0.2g will occur only for one or two peak cycles during the postulated earthquake and only at or near the ground surface.) ,
I 2. 5. 4.4.3. 3 scaring capacities Mejor structures will be s.upported on mat foundation established on structural fill at the elevations tabulated in Table 2.5-16. Ultimate bearing capacities and indicated factors of safety for mat foundations are presented in Table 2.5-19.
The tabulated factors of safety have been determined by assuming g that each structure, or portion of it with a variable mat elevation is isolated ,
from the adjacent structure.
All structures will have foundations proportioned such that the peak i foundation loading during seismic loading condition will not exceed 150 per cent '
of the foundation loeda tabulated in Table 2 5-19. Thus, factors of safety under short duration seicmic loading will not be less than about two-thirds of l the values tabulated in Table 2.5-19.
11 From a bearing capacity standpoint the most unfavorable condition during dynamic loading will occur if soils abo,ve elevation 450 and outside the compact-ed zone are assumed to liquefy. Under this condition, the bearing capacity of the soils will be reduced. Factors of safety during this condition vill not be less than about one-third of the values tabulated in Table 2.5-19. *
- 2. 5. 4. 4. 3. 4 Static Settlement Total and differential settlements which the proposed structures will 5 l undergo due to the static loads have been examined by the following caethods: I
- a. One dimensional conventional settlement analysis using the results 2.5-91.1
ZPS AMENDMENT 11 l
TABLE 2.5-19 k
ULTIMATE BEARING CAPACITIES !
I l STATIC ULTIMATE IISICATED l FOUNDATION BEARDiG FACTOR f FOUIWATION IDADING CAPACITY OF STRtrTURE EIEVATION , KIPS /SO.FT. KIPS /SO.FT. SAFETY Reactor 468'6" 9.2 217 24 Building Auxiliary 466'5" 5.8 196 34 )
Building 489'0" 5.0 119 24 Diesel 496'0" 5.0 95 19 Generator gy Building i
Turbine 466'5" 5.4 194 36 l Building 489'0" 4.5 127 28 Hester Bay 489'0" 3.7 119 32 466'5" 3.7 181 49 473'0" 3.7 163 44 Radwaste 489'0" 3.6 126 35 Prailding i l
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2.5-92 A
(
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2PS AMENDMENT 11 of laboratory consolidation tests.
- b. Emporical methods (Bazarra,1967, Meyerhof,1965) .
- c. Elastic compression using elastic deformation moduli data.
The estimated total and differential settlements for the proposed struc-tures derived from these analyses are presented in Table 2.5-20.
Assuming that structures or portions of structures which have their foundations at differing elevations are structurally disconnected, the marh==
estimated differential settlements between various structural units are tabulst-ed in Table 2,5-21.
The differential settlements tabulated above are based on the very conservative asstanption that all building and fill loads are applied instantane-ously. Actual sectiement will occur with application of loads, and more accu-race assessment cf differential settlements will be made when the construction 5 sequence is determined. Settlements will occur rapidly with the application of loads and no appreciabic residual settlement vill occur after completion of building construction and attaining the final site grade.
- 2. 5. 4. 4. 3. 5 Dynamic Settlement Earthquake motion will cause settlements which are additional to the static settlements presented in Paragraph 2.5.4.4.3.3. These dynamic settle-asents will consist of !!) elastic deformation due to increased dynamic bearing pressures developed during earthquake loading; and, (2) non-elastic deformation due to a possible slight densification of aiands below elevation 450 generated by both the vertical and horizontal components of the earthquake motion.
Elastic settlement wil1 be calculated and provided for in design using deformation moduli presented in Table 2.5-22. Non-elastic settlement is estimated to be less than one-half inch. .
If liquefaction of the natural soils above elevation 450 and adjacent to the compacted fill surrounding Class I structures should occur, ground displace-ments adjacent to building areas will result. The most sevete condition that can be postulated would be a lateral displacement of the soil mass overlying the liquefied layer with an associated depression termed a " graben," as described by Seed (1968), at the back end of the slide mass and adjacent to the structures, 37 This would not affect the stability or cause displacement of Class I structures '
as has been demonstrated in Paragraphs 2.5.4.4.3.2 and 2. 5.4.4.3.3. s 2 5.4.4.3.6 Rock-Soil-Structure Interaction Parameters dervied for analysis of bedrock-soil-structure interaction are summarized in Table 2.5-22. These values have been determined by field and D [
laboratory testing ad by reference to various published data. 1 2.5.4.5 Subsurface Walls )
1 l Subsurface walls will be designed to resist both static and dynamic later-al pressures. Estimated lateral pressures, expressed as equivalent fluid pressurers
]
j for various wall and loading conditions, are presented in Table 2.5-23. l5 ;
- 2. 5-94
)
_ _ _ _ - a
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ZPS AMENDMENT 11 Surcharge pressures from adjacent structures or loads will be added to the tabulated lateral pressures in the design of all walls.
The lateral pressures for rigid walls subjected to static loading have been assumed equal to earth pressure at rest.. For cantilever walls subjected to i static loads, active earth pressure has been assumed. In design, a factor of l safety of 1.5 will be applied in the use cf these values to allow for residual soil pressures which could result from the high degree of compaction which the backfill soils will receive.
The lateral pressure for both rigid walls and cantilever walls subjected to dynamic loading have been determined in accordance with the methods outlined by Seed and Whitman (1970). In design, a factor of safety of 1.1 will be applied in the use of these values.
- 2. 5.4. 6 River Bank Stability i
There is evidence of sicughing along the Ohio River bank on the west edge of the site. Blocks of sile and sand have slipped down to the river shore-line carr#ng trees and brush. It is believed that this instability is a result j of normal erosion due to undercutting of the bank by wave action. '
No deep-seated stability failures have been reported or observed along this reach of the Ohio River durin8 the past 200 years. There was no reported damage of the river banks in the site vicinity during the 1811-1812 New Madrid earthquakes, and it is believed that the intensities felt in the site vicinity from these s hocks were the largest from any known seismic event. On this basis, 4 deep-seated bank stability problems are not expected under normal conditions. I Analysis of the river bank stability has been performed for both static and pseudo-dynamic loading conditions. A minimum factor of safety of 15 against deep-seated failure has been obtained for static loading conditions.
Under pseudo-dynamic loading conditions factors of safety against deep-seated failures of 1.0 and 1.2 have been obtained for horizontal accelerations of 0.2g
, and 0.lg respectively. 11 The effect of liquefaction of the loose sands adjacent to the River embankment was not included in the above analysis. Due to the pre-earthquake stress conditions in earth banks it has been shown by Seed (1968) that liquefac-tion of a sand layer adjacent to a slope is less likely than well behind the lope. The mode of failure of the entire soil mass including the river embank-ent under these conditions would be as described in Paragraph 2.5.4.4.3.5.
- 2. 5.4. 7 Effects of Nearby Quarry Blasting on Plant Construction and Operation The Black River Mining Company operates an underground limestone mine at Carntown, Kentucky, approximately two miles upriver from the proposed site of the Zimmer Nuclear Power Plant, Moscow, Ohio. j l
2.5-98
~
4 L
ZPS AMENDMENT 11 l
l The mine is about 600 feet below the level of the river, and the under-I ground workings are quite extensive. The excavation is accomplished by use of explosives..
Normal operations include blasting once a day (excluding Saturday and Sunday) at 11:30 p.m. The morning work shif t removes what was blasted on the previous evening, and the afternoon work shift continues the removal process in addition to loading all holes that are to be blasted.
i One days supply of explosives, approximately 6,000 pounds, is kept under-l ground, as allowed by State Regulations. A regulation powder magazine is on the surface, and holds approximately 75,000 pounds of explosives at any time.
Electrical blasting caps are used, and both millisecond and regular delays are used in the blasting circuits.
Fram conversations with management persocnel at the mine, it is under-stood that mining operations will continue indefinitely. At present, roughly two million pounds of explosives are used a year, and this 12 expected to in-crease to three million pounds a year in the foreseeable future.
A limited number of measurements have been made of the nature of the ground motion effects of blast, both at the mine and at the plant site two adles distant. In the mine, at a distance of 1,000 to 1,200 from the blasting operations, ground motion velocities on the order of 0.1 to 0.4 inches per i second were measured. At the site, ground notion velocities of 0.0001 to 0.0002 have been measured during blasting at the mine.
2.5.4.8 future Units
! Consideration has been given to the effects that the construction of i i future units miEht have on the design of the unit. No adverse effects are i
l4 anticipated. The effects of renoval or reduction of lateral support from por-l I
tions of the unit substructures during construction of future units will bc considered in design. l4 l
)
2.5-99 ,
5 1
1 2PS l AMENDMENT 11 Division, ASCE, Vol. 93, No. SM3, May, pp.83-108. -
- 14. Seed , H.B. and Idriso, I.M. (1967) " Analysis f Soil Liquefaction:
Niigara Earthquake," Journal of the Soil Mechanics and Foundations Division, ASCE, Vol. 93, No. SM3, May, pp.
l
- 15. Seed, H.B.. and Whitman, Robert V. (1970) " Design of Earthquake Retaining Structures for Dynamic Loads," ASCE Specialty Conference.
Lateral Stresses in the Ground and Design of Earth Retaining I Structures, Cornell University, Ithaca, New York.
- 16. Seed, H.B. (1968) "The Fourth Terzaghi Iscture: Landslides During Earthquakes Due to Liquefaction", Journsi of the Soil gg Nechanics and Foundation Division, ASCE, Vol. 94, No. SMS, ,
September. '
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NT 11 ZPS AMENDMEh7 1:
2.2.4-1 (AEC - October 13. 1970. Ques tion 2.8)
I QUESTION Figure 1.5-1 shows the exclusion area boundary line passing through a secties of land east of the reactor building which is not owned by the applicant.
Describe the activities conducted within this area and indicate how access will be controlled. Figure 1.5-1 does not have a distance scale indicated. ' Provide a revised drawing showing this scale and also the Kentucky side of the Ohio River.
The cantour lines on Figure 1.5-1 are not clearly labeled as to elevations.
Please add these on the revised drawing. Indicate the distances from both the stack and the turbine building to the exclusion boundary.
ANSWER Tie answer can be found in Paragraph 2.2.4 of Amendment 4 and Paragraphs 13.6.4.4 of Amendment 11. The exclusion area boundary is indicated in Figures 2 2.2-3 and 2. 2-10.
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ZPS AMENDMEhT 11 2
2.2.5-1 (AEC - October 13, 1970, Question 2.10) gqESTION Identify and show on a recent map all schools, institutions and hospitals in the low population zone and describe how they will be considered in your evacuation plans. ,
AN3WER , r c-The answer can be found in Paragraph 2.2.5 of Amendment 4 and Paragraphs 13.6.4.4 and 13.6.4.5 of Amendment 11. The Low Population Zone is indicated in 11 Figure 1.4-3.
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.I ZPS AMENDMEF 11 2.5.4-1 (A m - October 13, 1970, Question 2.1) !
QUESTION A description of a number of possible foundation schemes for the facility structures is presented in Paragraph 2.5.4.3.1. The proposed foundation eleva-tions for the major buildings are summarized in Table 2.5-11. When the type of foundations to be employed is determined, submit foundation details to permit evaluation of the adequacy of the plant design to resist seismic effects. The j u information supplied should include a description of the structure supported by l the foundation, the dimensions of the foundation element, including length of l 1
. pipe or caissons and the depth to shich they are driven or drilled, and any other l l pertinent features needed to describe the foundation.
f ANSWER l'
2 Pile foundations will be utilized to provide foundation support for all structures. he pile foundations and structural pile caps will resist downward, ]<
uplift and lateral loads for all conditions of static and dynamic loading. Pile load tests will be performed to confirm design capacities for downward and uplift loads. He piling will be driven to bedrock which is approximately 50 ft below grade. Se type of pile and method of driving will be determined by a test pil-
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ing program. A more detailed description of the structures supported by the foundation, the dimensions of the foundation element and other pertinent infor- j mation can be found in Paragraph 2.5.4 of the Wm. H. Zimmer PSAR Amendment 11. l11 l
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J' ZPS AMENDMENT 11 INSTRUCTIONS FOR UPDATING YOUR PSAR VOLtME 2 All changes have been indicated by a vertical line and the Amendment Number (11) in the right margin of the page.
- 1. At the beginning of Volume 2 remove and destroy Page 7 and replace with amended Page 7. Remove and destroy Pages 17 through 19 and re-place with amended Pages 17 through 19.
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. - a, ZPS AMENDMENT 11 VOLU5tE 4 TABLE OF CONTENTS, (Continued)
PAGE 19.1 St#f!ARY DESCRIPTION 10.1-1 j '). 2 NEM FUEL STORACE 10.2-1 i '.t . 3 SPENT FUEL STORAGE 10.3-1
- 1. 8. 4 TOOLS AND SERVICING EQUIPMENT 10.4-1 1.) . 5 FUEL POOL COOLING AND CLEANUP SYSTEM 10.5-1 lo.6 REACTOR BUILDING CLOSED COOLING WATER SYSTEM 10.6-1 10.7 TURBINE BUILDING CLOSED COOLING WATER SYSTEM 10.7-1 h) . 6 SERVICE WATER SYSTEM 10.8-1 10.9 FIRE PROTECTION SYSTEM 10.9-1 10.10 f(EATING, VENTILATION, AND AIR CONDITIONING SYSTEMS 10.10-1 10.11 MAKE-UP WATER TREATMENT SYSTEM 10.11-1 lu.12 INSTRUMENT AND SERVICE AIR SYSTEMS 10.12-1 10.1J POTABLE AND SANITARY WATER SYSTEM 10.13-1 10.14 EQUIPMENT AND FLOOR DRAINAGE SYSTEMS 10.14-1 10.1 ', PLANT PROCESS SAMPLING SYSTEM 10.15-1 10.16 CO.'NIINICATION SYSTEM 10.16-1 19.17 LIGilTING SYSTEM 10.17-1 10.18 ilEATING BOILERS 10.18-1 10.19 PRIMARY CONTAINMENT MONITORING SYSTEM 10.19-3 10.20 PRIMARY CONTAINMENT HYDROGEN, OXYGEN AND FISSION PRODUCTS SAMPLING 10.20-1 l' 11.0 STEAM AND POWER CONVERSION SYSTEM TABLE OF CONTENTS 11.0-1 11.1
SUMMARY
DESCRIPTION 11.1-1 7
. _ . . . - - - . _ _ _ _ . _ - _ . - - . - _ _ _ _ _ - - _ _ _ . _ . _ . _ - - _ - *m
ZPS AMENDMENT 11 LIST OF ZPS, FEBRUARY 23, 1971 AEC QUESTIONS AEC QUESTION RENUMBERED VOLUME NUMBER AS QUESTION PAGE OF PSAR l 2.12 2.2.3-2 2.2.3-12 1 2.13 2.3.2.1-2 2.3.2.1-2 1 2.14 2.3.2.1-3 2.3.2.1-3 1 2.15 Later Later Later I
2.16 2.3.8-1 2.3.8-1 1 f 4.9 4.10 4.7-2 4.7-2 2 l9 l 4.7-1 4.7-1 2 4.11 4.9-1 4.9-1 2 9 4.12 4.0-1 4.0-1 l 2
5.11 5.2.3.7-1 5.2.3.7-1 2 5.12 10.19-1 10.19-1 2 5.13 Later Later Later 7 5.14 Later Later Later 5.15 Later Later Later 5.16 Later Later Later 5.17 5.2.3-1 5.2.3-1 2 7.1 5.3.3.3.3-1 5.3.3.3.3-1 2 7.2 5.3.3.3.2-1 5.3.3.3.2-1 2 7.3 5.3.3.3.3-2 5.3.3.3.3-2 2 7.4 7.1-1 7.1-1 3 11 7.5 Later Ieter Lt.ter 7.6 4.4-1 4.4-1 2 7.7 Later Later Later 11 7.8 7.2.3.6-1 7.2.3.6-1 3 l 11 7.9 7.2-1 7.2-1 3 7.10 7.2.3.9-1 7.2.3.9-1 3 ;
7.11 Later Later Later 17 l9
ZPS AMENDMENT 11 LIST OF 2PS, FEBRUARY 23, 1971 AEC QUESTIONS, (Continued)
(- AEC QUESTION RENUMBERED VOLUME NUMBER AS QUESTION PACE OF PSAR i
7.12 7.2-2 7.2-2 3 11 7.13 Later Later Later
'. 33 '
7.14 Later. ' '
Later Later j 7.15 7.4.3-1, q 7.4.3-1 3 :
7.16 7.5.7.3.3-1 7.5.7.3.3-1 3 7.17 7.8.5-1 7.8.5-1 3 11 '
9 7.18 7.5.8-1 7.5.8 3 7.19 7.6.3-1 7.6.3-1 3 9
7.20 7.8.5.2-1 7.8.5.2-1 3 7.21 La ter Later Later 7.22 Later Later Later 7.23 Later Later Later 7 7.24 Later Later Later l9 7.25 D.0-1 D.0-1 5 l11 7.26 10.10.3-1 10.10.3-1 4 7.27 7.2-3 7.2-5 3 l11 1 7.28 Later Later Later J 7.29 10.19-2 10.19-2 4 11 j 7.30 s Later Later Later 7.31 Later Later Later 8.1 8.3.2.1-1 8.3.2.1-1 4 1
8.2 8.3.2-1 8.3.2-1 4 8.3 8.3.3-1 8.3.3-1 4 8.4 8.4.3-1 8.4.3-1 4 8.5 8.5.4-1 8.5.4-1 4 1 j 8.6 8.4.3-2 8.4.3-2 4 i
8.7 8.5.3.1-1 8.5.3.1-1 4 l l 8.8 8.0-1 8.0-1 4 i J e i f
- l. 18 9 i
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XPS
/MENDMENT 11 LIST OF ZPS, F EBRUARY 23. 1971 AEC QUESTIONS, (Con t inued )
AEC QUESTION RENUMBERED VOLUME NUHBER AS QUESTION OF PSAR
- _PA_G_E_
8.9 8.0-2 8.0-2 4
., 8.10 .
8.9-1 8.9-1 4 8.11 -
8.10-1 8.10-1 4 9.1 9.2.4-1 9.2.4-1 4 ,
'[ -
9.2 9.2.4.6-1 9.2.4.6-1 4 l 9.3 9.4-1 9.4-1 4
( 9.4 9.4.6-1 9.4.6-1 4 l 9.5 9.2.4.7-1 9.2.4.7-1 4 9.6 9.4.3-1 9.4.3-1 4 10.1 I4ter Later Later 10.2 10.5-1 10.5-1 4 10.3 10.4 10.0-1 10.0-1 4 l1 10.11.2-1 10.11.2-1 4 7 12.22 12.6.1-1 12.6.1-1 4 12.23 12.5.6-1 12.5.6-1 4 13.1 13.0-1 13.0-1 4 13.2 13.2.1.6-1 13.2.1.6-1 4 13.3 13.2.1.2-1 13.2.1.2-1 4 9 13.4 13.0-2 13.0-2 4 13.5 13.3-1 13.3-1 4 13.6 13.6.4-1 13.6.4-1 4 13.7 13.0-3 13.0-3 4 33 14.12 14.9.1-1 14.9.1-1 4 14.13 Later Later Later 19 9
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ZPS i
AMENDMENT 11 l INS 7 RUCTIONS FOR UPDATING YOUR PSAR VOLINE 2 JECTION 4.0 - REACTOR COOIANT SYSTEM i
This section has been updated with the inclusion of the answer to AEC-DRC j question 7.6. .
l l
All changes have been indicated by a vertical line and the annendment l number (11) in the right margin of the page. l All pages (text, tables, figures) with changes have also been marked in the upper right corner of the page with "JMENDMENT 11".
Figures that have been altered is any way are indicated by the amendment number in the upper right corner of the figure; note that there are no other marks that would indicate changes in figure. On the page marked " LIST OF FIGURES", figures that have changed in any way are designated by a vertical line with the amendment number alongside the title of the figure. See example below:
FIGURE NINBER TITIE 2.2-1 Station Site Area Topography 11 To update your copy of the Wm. H. Zimmer Nuclear Power Station PSAR, please use the following procedure:
- 1. In Volume 2, SECTION 4.0 - REACTOR COOIANT SYSTEM, behind the red tabbed divider page titled " Amendment to Section 4.0" after Page 4.2.6-5 insert Page 4.4-1.
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AJ0ENDMENT 11 4.4-1 (ZPS - February 23, 1971, AEC Question 7.6)
} QUESTION l State' your design criteria and describe the design of the pressure controllers used to actuate the safety-relief valves so as to limit reactor l j
over-pressure during expected operational transients. Discuss the consequences 1 i
of single failures in the design of this system.
l ANSWER
.l j The pressure switch used to actuate each safety relief valve shall l
make and break electrical contact with a difference in pressure of 1% or less of set pressure.
f When the pressure increases to the set point, the high pressure switch in the dual control pressure switch is actuated and completes the relay circuit that energizes the solenoid of the air line valve on the safety / relief valve piston actuator.
The low pressure switch provides for relay control below the actuation value of the high pressure switch there by allowing an adjustable blow down range for the safety / relief valve.
As indicated in the Wh, H. Zimmer PSAR Table 4.4-1 each safety relief valve has its own pressure controller pressure switch to initiate the relief function of the valve. Failure of a pressure switch could prevent only one safety /
relief valve from opening at its set pressure. Such a failure would slightly increase the peak pressure of the reactor coolant system for a full load turbine trip transient. This transient as analyzed in the Um. H. Zimmer PSAR Paragraph 14.5.1.2.1 indicates a margin of 40 psi between peak pressure and :
lowest' safety valve setpoint. It is estimated that this margin would decrease l by no more than 10 psi if one valve failed to operate.
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AMENDMENT 11 INSTRUCTIONS FOR UPDATING YOUR PSAR b VOLUME 2
.) -
); SECTION 5.0 - CONTA'INMENT bi An answer to a DRL-AEC question has been revised.
If.
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All changes have been indicated by a vertical line and the amendment (11) 4 , in the right margin of the page.
d ,
Q All pages (text, tables, figures) with changes have also been marked in i the upper right corner of the page with " AMENDMENT 11".
5 f) g
. Figures that have been altered in any way are indicated by the amendment number in the upper right corner of the figure; note that there are no other marks
[ that would indicate changes in figure. On the page marked " LIST OF FIGURES", fig-T. ures that have changed in any way are designated by a vertical line with the amend-f, ment number alongside the title of the figure. See example below:
FIGURE NUMBER TITLE P
l 2.2-1 Station Site Area Topography l 11
^
1
[ To update your copy of the WM. H. Zimmer Nuclear Power Station PSAR, r
please use the following procedure:
,, 1. In Volume 2. SECTION 5.0 - CONTAINMENT, remove and destroy Table t
f of Content Page 5.0-v and replace with amended Page 5.0-v. !
- 2. In Volume 2, SECTION 5.0 - CONTAINMENT remove and destroy text Page 5.2-6 and replace with amended Page 5.2-6.
l 3. In volume 2, SECTION 5.0 - CONTAINMENT after Figure 5.2-14 insert new Figure 5.2-15.
- 4 In Volume 2, SECTION 5.0 - CONTAINMENT, behind the red tabbed titled " Amendment to Section 5.0", remove and des troy Pages 5.0-1 and 5.0-2 and replace with amended Pages 5.0-1 and 5.0-2.
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AMENDMENT 11 SECTION 5.0 - CONTAINMENT
{ LIST OF FIGURES l
! FIGURE NUMBER TITLE l
t 5.2-1 Primary and Secondary Concrete Containment Structures i
5.2-2 '
Column and Wall Base Detail at Floor Liner Plate
[ 5.2-2.1 Drywell Floor Joint At Containment Wall I l 5.2-3 Typical Section at Buttress 5.2-4 Tendon Access Gallery 5.2-5 Typical Leak Test Chamber 5.2-6 Drywell Heat Attachment Detail - Tendon 7-Anchor at Drywell Head 5.2-7 Primary Containment System Hot Process Line Penetration 5.2-8 Primary Containment System Cold Process Line Penetration 5.2-9 Typical Electrical Penetration Assembly 5.2-10 Personnel Access Lock 5.2-11 Drywell cooling and Ventilation System 5.2-12 Emergency Lock and Equipment Hatch 5.2-13 Typical Layout of Hoop Tendons 5.2-14 Typical Layout of Vertical Tendons 1
, 5.2-15 Reactor Containment Development Elevation i iI 11 5.3-1 Standby Cas Treatment System l5
- 5.3-2 CSCS Equipment Area Cooling System I.
5.3-3 Reactor Building Ventilation System
( .I' i 7 i-. 5.3-4 Standby Gas Treatment System Equipment Train a
5.3-5 Schematic-Showing Mixing Effect of Supply Outlet i
-5.0-v
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1 1
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2PS l AMENDMENT 11
- a. They are capable of withstanding the peak transient pressure.
- b. They are capable of withstanding the forces caused by impingement of the fluid from the rupture of the largest local pipe os connection without failure.
7 c. They are capable of accommodating the thermal and mechanical stresses which may be encountered during all modes of operation without fail- )
ure. . j Refer to Table 5.2-2 for approximate number and size of these penetra- l P tions and Figure 5.2-15 for the location of the penetrations.
l13 5.2.3.4. 2 Pipe Penetrations
, Pipe penetrations will be of the type as shown in Figures 5.2-7 and l
5.2-8 for all process lines penetrating the containment. Piping penetration locations will be consistent with the requirements of the safety design basis 7 f ;
- for the piping, control and instrument systems. The pipe will be welded )
> directly to the sleeve which is imbedded into the concrete as it penetrates the containmen t. Insulation and air gaps are' provided around the pipe to reduce
! thermal stress in the containment during normal operations.
In addition to their function as a primary containment barrier, the penetrations serve as anchors to the pipes. Thermal growth and movement will be taken up in the piping system. Guided supports will be used where required to direct pipe expansion. The piping system will be designed such that the result-ant combined stress in the pipe and penetration components under normal and ac-cident conditions do not exceed the code allowable design limits ASME Nuclear Vessel Code,Section III Subsection B.
5.2.3.4.3 Electrical Penetrations Electrical penetrations will be designed to accommodate the electrical requirements of the plant. These are functionally grouped into low voltage power and control cable penetration assemblies, high voltage power cable pene-tration assemblies, and shielded cable penetration assemblies. Each penetration seal will have the same basic configuration. An assembly will be sized to be inserted in the 12 inch schedule 80 penetration nozzles which are furnished as part of the containment structure. Installation of the penetration assembly will be accomplished by inserting it from outside of the containment into the penetration nozzle. Three field welds are required to complete the installation of the assembly in the penetration nozzle. It is intended that the penetration canister assemblies be supplied from coacnercially available designs (Figure 1
5.2-9).
Peaderplates conforming to the inner diameter of the penetration nozzle will be provided at each end of the penetration assembly, forming a double pres-sure barrier. Radiation shielding will be attached to the penetrations on the 5.2-6
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.i REACTOR CONTAINMENT l l
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ZPS t
AMENDMENT 11 I 5.0-1 (AEC - Oc tober 13, 1970, Question 5.6) l
)
I l QUESTION i
In connection with the material presented in Section 5.0 and Section 12.C I of the PSAR, the following information is required:
! l
{ (a) Provide a more detailed description of the manner la which the ,
reactor vessel is supported both vertically and laterally.
l (b) Since the dividing floor between the dry well and the suppression chamber will be a conventionally reinforced concrete floor supported by the center pedestal on a sedes of concrete columns, and from the containment vall at the periphery of the slab, provisions are neces-sary to accommodate the required differential motion for thermal and seismic effects. It is not clear from the description given whether there is a seal between the floor and the containment wall separating the two chambers. Describe the details, if such a seal is provided, or other means to accommodate the required differential motion.
l (c) Since the dry well primary containment structure is prestressed, l provide additional information with respect to the details of the i
penetrations through this structure and particularly the methods of analysis employed in checking the adequacy of the penetrations.
ANSWER i a.
I The reactor pressure vessel is supported vertically by the building foundation through the reactor pressure vessel concrete pedestal and laterally at its base by the reactor pressure vessel concrete pedes-tal and at its three quarter he'ight by the top of the shield wall through the reactor pressure vessel stabilizers. The reactor pressure vessel support is shown in the attached Figure.
- b. The joint between dry well floor and containment will be treated as follows:
l1 The dry well floor will be rigidly connected to the containment wall.
A full moment and shear connection will be provided by dovels and shear lugs welded to the reinforced liner plate (similar to Figure 5.2-2). The thermal expansion will become part of the containment design, the resulting forces and moments will be acccamodated within the allowable stress limits. There is no need for a seal or shear keys. (See Paragraph 5.2.3.1 Amendment 7) l 1:
- c. The concrete containment will be analyzed for the effects of penetra-tions on the stresses in the containment walls for the various load-ing combinations included in Table 12.2-2. The stresses for the con-tainment will first be determined neglecting the openizgs. These stresses will then be used to formulate boundary conditions around 5.0-1 I
/
ZPS AMENDMENT 11 the large penetrations. Utilizing the PLFEM Computer Program, the seem-brane forces, moments and shears will be calculated. The wall thick-l ness around the opening will be thicknened if necessary and reinforced i to carry the stress concentration. The design will insure .that the re-forcing provided will replace the strength remioved by the opening.
f l
4 The prestressing tendons will be deflected around all penetrations l
) rather than anchored to the openings.
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- i. ZPS p
jf AMENDMENI 11
' INSTRUCTIONS FOR UPDATING YOUR PSAR k
VOLUME 3 All changes have bee.. indicated by a vertical line and the Amendment Number (11) in the right margin of the page.
- 1. At the beginning of Volume 3 remove and destroy Page 7 and replace with amended Page 7. Remove and destroy Pages 17 through 19 and ~re-place with amended Pages 17 through 19.
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! ZPS ANENINENT 11
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_VOLtNE 4 TABLE ,0F CONTENTS, (Continued)
PAGE 10.1 SLMIARY DESCRIPTION 2 10.1-1
} l'). 2 NfM FUEL STORACE -
~
10.2-1 i .8. 3 SPEfft FUEL STORACE l_ 10.3-1 1:8.4 TOOLS AND SERVICING EQUIPMENT 10.4-1 t-4 1.1. 5 i FU! L POOL COOLING AND CLEANUP SYSTEM 10.5-1 t
{ lo.6 i
REACTOR BUILDIE: CLOSED COOLING WATER SYSTEM 10.6-1 I -
10.7 TURBINE BUILDINC CLOSED COOLING WATER SYSTEM 10.7-1 I id.8 SERVICE i'ATER SYSTEM 10.8-1
- 10.9 FIRE PROTECTION SYSTD1 10.9-1 f 10.10 HEATING, VEttilIATION, AND AIR CONDITIONING SYSTEMS 4 10.10-1
- 10.11 MAKE-UP WATER TREATMENT SYSTEM 10.11-1
'i , 10.12 INSTRUMENT AND SERVICE AIR SYSTEMS I 10.12-1 10.13 POTABLE AND SANITARY WATER SYSTEM 10.13-1 10.14 EQUIINENT AND FLOOR DRAINAGE SYSTEMS 10.14-1 r
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. PLANT PROCESS SAMPLING SYSTEM 10.15-1 10.16 C0bNUNICAI10N SYSTEM 10.16-1 10.17 LIGitTING SYSTEM 10.17-!
10.18 HEATING BOILERS 10.18-1 10.19 PRIMARY CONTAINMENT MONITORING SYSTEM 10.19-3 .
10.20 PRIMARY CONTAINMENT IIYDROGEN, OXYGEN AND FISSION
, j PRODUCTS SAMPLING 10.20-1
- 11 l 11.0 STEAM AND POWER CONVERSION SYSTEM l
l TABLE OF CONTENTS 11.0-1 l 11.1
SUMMARY
DESCRIPTION l 11.1-1 7
=
1
/
2PS AMENDMENT 11 LIST OF ZPS, FEBRUARY 23, 1971 AEC QUESTIONS AEC QUESTION RENUMBE RED VOLUME NUMBER AS QUESTION PAGE OF PSAR 2.12 2.2.3-2 2.2.3-12 1 2.13 2.3.2.1-2 2.3.2.1-2 1 2.14 2.3.2.1-3 2.3.2.1-3 1 2.15 Later Later Later 2.16 2.3.8-1 2.3.8-1 1 4.9 4.7-2 4.7-2 2 9 4.10 4.7-1 4.7-1 2 4.11 4.9-1 4.9-1 2 9 4.12 4.0-1 4.0-1 2 5.11 5.2.3.7-1 5.2.3.7-1 2 5.12 10.19-1 10.19-1 2 5.13 Later La ter Later 7 5.14 Later Later Later 5.15 Later Later Later 5.16 Later Later Later 5.17 5.2.3-1 5. 2. 3 -1 2 7.1 5.3.3.3.3-1 5.3.3.3.3-1 2 7.2 5.3.3.3.2-1 5.3.3.3.2-1 2 7.3 5.3.3.3.3-2 5.3.3.3.3-2 2 7.4 7.1-1 7.1-1 3 11 7.5 Later Later Later 7.6 4.4-1 4.4-1 2 7.7 Later Later Later 7.8 7.2.3.6-1 7.2.3.6-1 3 1 11 ,'
7.9 7.2-1 7.2-1 3 l 7.10 7.2.3.9-1 7.2.3.9-1 I 3 7.11 Later Later
(
Later 17 l9 m
/
ZPS AMENDMENT 11 LIST OF ZPS, FEBRUARY 23, 1971 AEC QUESTIONS, (Continued)
AEC QUESTION RENUMBERED VOIIME NUMBER AS QUESTION ,PAGE OF PSAR 7.12 7.2-2 7.2-2 3 f I' 7.13 Later Later Later 7.14 Later Later later 7.15 7.4.3-1 7.4.3-1 3 7.16 7.5.7.3.3-1 7.5.7.3.3-1 3 7.17 7.8.5 7.8.5-1 3 11-g :
7.18 7.5.8-1 7.5.8-1 3 7.19 7.6.3-1 7.6.3-1 3 3 7.20 7.8.5.2-1 7.8.5.2-1 3 7.21 La ter Later Later 7.22 Later Later. La ter 7.23 Later Later Later 7 7.24 Later Later Later l9 7.25 D.0-1 D.0-1 5 l11 7.26 10.10.3-1 10.10.3-1 4 7.27 7.2-3 7.2-5 3 l11 7.28 Later Later Later 7.29 10.19-2 10.19-2 4 11 7.30 Later Later Later 7.31 Later Later Later 8.1 8.3.2.1-1 8. 3. 2.1 -1 4 i 8.2 8.3.2-1 8. 3. 2-1 4 8.3 8.3.3-1 8.3.3-1 4 8.4 8.4.3-1 8.4.3-1 4 8.5 8.5.4-1 8.5.4-1 4 8.6 8.4.3-2 8.4.3-2 4 l 8.7 8.5.3.1-1 8. 5. 3.1 -1 4 i ,! 8.8 8.0-1 8.0-1 4 l 18 9
ZPS AMENDMENT 11 E'T OF ZPS, FEBRUARY 23, 1971 AEC QUESTIONS, (Continued)"
AEC QUESTION I ENUMBERED VOLLHE NUMBER Ai QUESTION PAGE OF PSAR
}
, 8.9 8.0-2 8.0-2 4 8.10 8. 9 -1 8.9-1 4 l '
j 8.11 8.10-1 8.10-1 4
' 9.2.4-1 9.1 9.2.4-1 4 9.2 9. 2. 4. 6-1 9.2.4.6-1 4 9.3 9.4-1 9.4-1 4 l
9.4 9.4.6-1 9.4.6-1 4 l1:
9.5 9. 2.4. 7-1 9.2.4.7-1 4 9.6 9.4.3-1 9.4.3-1 4 10.1 Later Later Later 10.2 10.5-1 10.5-1 4 10.3 10.0-1 10.0-1 4 l 11 10.4 10.11.2-1 10.11.2-1 4 7 12.22 12.6.1-1 12.6.1-1 4 12.23 12.5.6-1 12.5.6-1 4 13.1 13.0-1 13.0-1 4 13.2 13.2.1.6-1 13.2.1.6-1 4 13.3 13.2.1.2-1 13.2.1.2-1 4 9 13.4 13.0-2 13.0-2 4 13.5 13.3-1 13.3-1 4 13.6 13.6.4-1 13.6.4-1 4 13.7 13.0-3 13.0-3 4 11 14.12 14.9.1-1 14.9.1-1 4 14.13 Later Later Later 19 9
.1
ZPS AMENDMENT i i l
INSTRUCTIONS FOR UPDATING YOUR PSAR
_ VOLUME 3_
_ SECTIO 7.0 - CONTROL AND INSTRUMENTATION This sectinn minor ch,anges in thehas text.been revised to reflect new information and incorporate;
)
j All changes have been indicated by a. vertical line and the amendment number (11) in the right margin of the page, i
?
i All pages (text, tables, figures) with changes have also been marked I in the upper right corner of the page with " AMENDMENT 11".
g Figures that have been altered in any way are indicated by the amend- ;
ment number in the upper right corner of the figure; note that there are no i other marks that would indicate changes in figure. On the page marked " LIST OF FIGURES", figures that have changed in any way are designated by a vertical l
line with the amendment number alongside the title of the figure. See example below: l
)
FIGURE NUMBER
_ TITLE l ,
2.2-1 Station Site Area Topography l:
please useTo update your copy the following of the Wm. H. Zimmer Nuclear Power Station PSAR, procedure: ,
1.
{ In Voltane 3, SECTION 7.0 - CONTROL AND INSTRUMENTATION, remove and l destroy Table of Contents Pages 7.0-1 and 7.0,xv and replace with ,
amended Pages 7.0-1 and 7.0-xv. l
! I
- 2. \
l In Voltane 3, SECTION 7.0 - CONTROL AND INSTktfENTATION, remove and j l destroy the following pages and replace with the appropriate pages listed below:
) '
REMOVE PAGE
_REPIACE WITH AMENDED PAGES 7.1-1 7.1-1 I 1 I
7.1-2 7.1-1.1 I
} 7.1-2 l 7.2-2 7.2-2
- i. 7.2-3 {
s 7.2-3 j
7.2-4 7.2-4 {
7.2-8 7.2 -8 j
7.2-14 i . 7 2-14 7.3-3 I 7. 3 -3 7.3-5 7.3-5 7.3-6 through 7.3-8 7.3-6 through 7.3-8 l l 7.3-12 l l 7.3-13 7.3-12 7.3-13 A.
_ _ _ _ _ _ _ _ _ _ . . _ _ - _ - - - - - - - - - - - - J
ZPS 11 AMENDMENT 11
_ REMOVE PAGE REPLACE WITH AMENDED PAGES 7.3-19 7.3-19 7.3 20 -
7.3 20
, 7.3 22 7.3-22 7.3 23 7.3-23 g, 7.3 24 7.3-24
~
7.4 2 ,
7.4 2 -
7.4-3 7.4-3 7.4-10 -
7.4-10 7.4-15 7.4-15 7.4-17 7.4-17 7.4 19 7.4-19 7.4 20 7.4-20 7.4-31 7.4-31 7.5-17 7.5-17 7.8-5 7.8-5
.7.8-7 7.8-7 7.12-11 7.12-11
- 3. In Volume 3, SECTION 7.0 - CONTROL AND INSTRUMENTATION, remove and l11 destroy Figure 7.3-2 and replace with Amended Figure 7.3-2
- 4. In Volume 3, SECTION_7.0 - CONTROL AND INSTRUMENTATION, behind the red tabbed divider. page titled " Amendments to Section 7.0", insert the following pa:ges in the order given below:
PAGE NUMBER 7.1-1 7.2-1 7.2-2 7.2-3 7.2-4 7.2-5 7.2.3.6-1 7.2.3.9-1 7.4.3-1
- 7. 5. 7.3. 3- 1 3
7.5.8-1 7.6.3-1 7.8.5-1 r 7. 8.5.2- 1 4
4 .
- } ZPS AMENDMENT 11 SECTION 7.0 - CONTROL AND INSTRUMENTATION TABLE OF CONTENTS PACE 7.0 CONTROL AND INSTRUMENTATION 7.1-1 7.1 SIDEARY DESCRIPTION 7.1-1 j 7.1.1 Protection Systems -
7.1-1 7.1.2 Power Generation Systems 7.1-1 7.1.3 Protective Functions 7.1-1 l11 7.1.4 Plant Operational Control 7.1-2 !
7.1.5 Definitions 7.1-3 7.2 REACTOR PROTECTION SYSTEM 7.2-1 7.2.1 Safety Objective 7.2-1 7.2.2 Safety Design Basis 7.2-1 7.2.3 Description 7.2-3
- 7. 2. 3.1 Identification 7.2-3 7.2.3.2 Power Supply 7.2-4 7.2.3.3 Physical Arrangement 7.2-4 7.2.3.4 Logic 7.2-4 7.2.3.5 Operation 7.2-5 7.2.3.6 Scram Functions and Settings 7.2-7 s
7.2.3.7 Mode Switch 7.2-11 7.2.3.8 Scram Bypasses 7.2-12 7.2.3.9 Instrumentation 7.2-13
, 7.2.3.10 Wiring 7.2-17 7.2.4 Safety Evaluation 7.2-19 7.2.5 Inspection and Testing 7.2-25 7.2.6 Operational Nuclear Safety Requirements 7.2-27 7.2.6.1 Limiting Conditions for Operation 7.2-27 7.2.6.2 Surveillance Requirements 7.2-34 7.0-1
-T; ,
N
,P
! .I ZPS L ,
l AMENDMENT 11 h SECTION 7.0 - CONTROL AND INSTRUMENTATION Q,
LIST OF FIGURES f.-
L FIGURE NUMBER TITE 7.1-1 Use of Protection System Control &
Instrumentation Definitions 7.2-1 . Reactor Protection System Schematic .
Diagram )
7.2-2 Reactor Protection System Typical Oma-11 L
' P trol Room Panel for One Trip System - i 7.2-3 Reactor Protection System Punctional Control Diagram a 7.2-4 Reactor Protectico System Scram Bactions 7.2-5 Reactor Protection System Instrumentation l
! 743 I Nuclear Boiler System Piping &
! Instrumentation Diagram fj
- 7. 3 -2 Main Steam Line Isolation valve Schematic Control Diagram j
! 7.3-3.1 Nuclear Boiler System Functional Control Diagram, Part 1 7.3-3.2 Nuclear Boiler System Functional Control Diagram, Part 2 l 7.4-1 4 CSCS Network Models I
7.4-2 High-Pressure Core Spray Piping &
Instrumentation Diagram I 7.4-3.1 Ifigh-Pressure Core Spray System I Functional Control Diagram, Part 1 7.4-3.2 High-Pressure Core Spray System ,
Functional Control Diagram, Part 2 '
i 7.4-3.3 High-Pressure Core Spray System Functi<xs1 Control Diagram, Part 3 l 7.4-3.4 High-Pressure Core Spray System Punctional Control Diagram, Part 4 7,4-4 Instrumentation Objectives l
Initiation Logic - ER B and C, HPCS, RCIC 7.4-5 7.4-6 Automatic Depressurization System Functional Control Diagram 7.0-xv
I I ZPS
' ~
AMENDMENT 11 l
SECTION 7.0 - CONTROL AND INSTRUMENTATION l I
7.1
SUMMARY
DESCRIPTION i The control and instrumentation section presents the details of the more i compisx control and instrumentation systems in the station. Some of these sys - J tema ar nuclear protect!on systems, while others are power generation systems.
7.1.1 Nuclear Protection Syshms_
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i P
i ZPS AMENDMENT 11 i
side the primary containment and inside the reactor building; !
they are physically separated from each other and tap off the reactor vessel at widely separated points. The reactor protection system pressure switches, as well as instruments for other systems, sense pressure and level from these same pipes. The physical separation and signal arrangement assure that no single physical event can prevent a scram due to reactor vessel low water level.
Temperature equalizing columns are used to increase the accuracy of the level measurements.
- 4. Turbine stop valve closure inputs to the reactor protection sys-tem are from valve stem position switches mounted on the four turbine stop valves. Each of the double-pole, single-throw switches is arranged to open before the valve is more than 10% closed to provide the earliest positive indication of closure. Either of the two trip channels associated with one stop valve can signal valve closure. The logic is arranged so that closure of any two
. valves causes a single trip system trip, and closure of three or more valves initiates a scram.
- 5. Turbine control valve fast closure inputs to reactor protection system are from four (4) pressure switches on the control valve hydraulic fluid discharge header. The pressure switches monitor the loss of hyrauulic fluid pressure which will result in the fast closure of the control valves. The pressure switches provide signals to the four reactor protection system trip channels. The arrangement l' is a one out of two taken twice logic.
- 6. Main steamline isolation valve closure inputs to the reactor pro-tection system are from valve stem position switches mounted on the eight main steamline isolation valv Each o~ the double-pole, single-threw switches is arranged to open before the valve is more than 10% closed to provide the earliest positive indication of closure. Either of the two trip channels associated with one iso-lation valve can signal valve closure. To facilitate the descrip-tion of the logic arrangement, the position sensing channels for each valve are identified as follows:* l6)
Position Sensing Valve Identification Channels ;
i Main steam line A, inboard valve a,b !
Main steam line A, outboard valve c.d Main steam line B, inboard valve e,f I Main steam line B, outboard valve g,h Main steam line C, inboard valve j,k
- Additional information is available in CE Topical Report NED 0-10139 l6 7.2-14 1
e
ZPS AMENDMINT 11 f
l control system to respond correct 1z to essential monitored variables, i
- b. The system shall be designed for a high probability that when any essential monitored variabJe exceeds the isolation set point, the event shall either result in automatic isolation or shall not impair the ability of the system to respond correctly as other monitored variables exceed their trip points.
, l c. Where a plant condition that requires isolation can be brought on by a failure or malfunction of a control or regulating system, and the same failure or malfunction prevents action by one or more isolation contrei system channels designed to provide protection against the un-safe condition, the remaining portions of the isolation control system shall meet the requirements of safety design bases 1, 2, 3, and 7a. i
- d. The power supplies for the primary containment and reactor vessel isolation control system shall be arranged so that loss of one supply cannot prevent automatic isolation when required.
- e. 1he system shall be designed so that, once initiated, automatic isolation action goes to completion. Return
,1 to normal operation after isolation action shall require
,f - deliverate operator action.
- f. 1here shall be sufficient electrical, and physical wiring l11 i
. and piping separation between trip channels monitordng the l same essential variable to prevent environmental factors, electrical faults, and physicci events from impairing the i ability of the system to respond correctly. s 6
- 8. To reduce the probability that the operational reliability and precision of the primary containment and reactor vessel isolation control system will be degraded by operator error, the following safety design bases are specified for Class A and Class B .Suto-matic isolation valves: l
- a. Access to all trip settiags, component calibration controls, test points, and other tenninal points for equipment associated with essential monitored variables shall be under the physical control af supervision or of the control room operator.
.I 7.3-3
ZPS ANINDHENT 11 l
7.3.4.2 Power Supply ,
The pcwer supply for the trip systems and trip logics is fed from the same two electrical buses that supply the reactor protection system trip systems. Each of the two buses has its own motor-generator set. Either bus can receive alternate power from a bus that can be energized by standby power. The buses cannot be simultaneously supplied from the same power source. Isolation valves receive electrical power from buses that are reli-able, in that power will be available from standby power sources. One solenoid on each valve is powered by the RPS "A" bus, and the second by the RPE "B" 6 ll ;
bus. The main steam isolation valves, which are described in detail later, use ac and pneumatic pressure in the control scheme. Table 7.3-1 lists the power supply for each isolation valve.
7.3.4.3 Physical Arrangement Table 7.3-1 lists the pipelines that penetrate the primary containment and indicates the types and locations of the isolation valves installed in each pipeline. Figure 7.3-1 identifies some of these pipelines. Pipelines which penetrate the primary containment and are in direct communication with the reactor vessel generally have two Class A isolation valves, one inside l the primary containment and one outside the primary containment. Pipelines j which penetrate the primary containment and which communicate with the primary I containment free space, but which do not communicate directly with the reactor vessel, generally have two Class B isolation valves located outside the pri-mar, containment. Class A and Class B automatic isolation valves are consider-ed essential for protection against the gross release of radioactive material in the event of a breach in the nuclear system process barrier. Process pipelines that penetrate the primary containment but do not communicate directly with the reactor vessel, the primary containment free space, or the environs, have at least one Class C isolation valve located outside the primary contt.nment which is capable of remote manual operation. Table 7.3-1 yy presents information about piping penetrations in the primary contain-ment. Only the controls for the automatic isolation valves are discussed in this part of the safety analysis report. The valves which are the sub.
ject of this text are specifically identified in the detailed descriptions which follow.
Power cables are run in conduits from appropriate electrical sources i to the motor or solenoid involved in the operation of each isolation valve. l The control arrangement for the main steam line isolation valves includes j 7.3-5
l I
P_ROCESS PIPELINES PENETR, l (Numbers in parenthesis are keyed to notes on ;!
k Valve Approx, Valves Location No. of Pipe Per Ref. to Valve Line Isolated . Lines Size (in.) Line Class Drywell
' ' Type (j 5 l11 Main Steam Line O
"4 24 1' A Inside A0 Glo 1 A .Outside A0 Glo '
Main Steam Line Drain 1 3 1 A Inside M0 Gate 1 A Outside M0 Cats '
'From Rx. Feedwater 2 18 1 A Inside Check 1 A Outside Check Rx. Water Sample 1 A Outside M0 Chec 1 3/4 1 A Inside 50 Valsl 1 A Outside SO Valv-Control Rod liyd. Return 1 3 1 A Inside Chec15 1 A Outside Check Control Rod Drive Outlet 137 1 2 A Outside SO Valvi Control Rod Drive Inlet 137 1 2 A Outside SO Valw RHR (Shutdown Cooling) Ret. 2 10 1 A Inside A0 ChecI 1 A Outside MO Gloix' RHR (Shutdown Cooling) Suct. I 18 1 A Inside M0 Gate (from recire, syst.) 1 A Outaide MO Gate RHR - LPCI to Rx. 3 10 1 A Inside A0 Check f
3 1 A Outside MO Ga te
'1 i RHR,- Rx llcad Spray 1 6 1 A Inside MO Gate j 1 A Outside MC Globe RHR to Supp. Spray Header 2 Outside f
6 1 B M0 Gate RHR Containment Spray 2 10 2 B Outside M0 Cate
! RHR Test Line to Supp. Pool 2 8 2 B Outside M0 Gate RHR Pump Min. Flow Bypass
,- to Supp. Pool 3 8 1 B Outside Check i
RHR Suct. from Supp. Pool 3 20 1 B Outside MO Gate
, Standby Liq. Control Sys. I 1) 1 A Inside Check 1 A Outside Check j
- See Note (9) l @ Maximum closing time
/
--...-l !
l l
l TABIE l MTING PF followi/ ,
5$ ,
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BLANK PAGE e
i Ck , i O T-d 50 g, i
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i I
I 1
l
l Z PS AMEND.T."I . :
7.3-1 PRItRRY COVTAINME.W ring pages; signal codes are listed on following pages.) )4 Min.
Closing Normal Power to Power to Isolation Rate or Status Open (5) (6) Close _(5) (6) Signal Time (7.11.12) (9. 10)_ Remarks Air & AC Air /S pring B ,C ,D ,P 3 sec Open Note (1)
Air & AC Air / Spring B,C,D,P 3 sec Open Note (1) f6 AC AC B ,C,D , P 12"/ min Closed
- Note (9)
DC DC B ,C ,D , P 12"/ min Closed
- Note (9)
Process Rev. Flow - Open
- Process Rev. Flow - Open l1, AC AC RM Standard Open .
AC Spring B,C,D,P Standard Closed
- Note (9)
AC Spring B ,C ,D , P St.anda rd Closed
- Note (9) ,
- Process Rev. Flow - Opens on Rod ) f
- Process Rev. Flow - Movement & )
AC Spring Note (4) -
Closed all ) Note (4)
AC Spring Note (4) - other times )
Note (3) Note (3) Note (3) - Closed Note (3)
AC AC C 37"/ min Closed AC AC RM Standard Closed l11 DC DC RM Standard Closed Note (3) Note (3) Note (3) -
g Closed Note (3)
AC AC C 12 see Closed l11 AC AC A,F,U Standard Closed DC DC A,F,U Standard Closed AC AC G Standard closed Note (2)
AC AC G Standard Closed Note (2)
. Process Rev. Flow -
Closed RM Standard Open 11 AC AC
- Process ele v. Flow - Closed
- Procesa Rev. Flow - Closed I
l l
l
)
l TABII 7.3-1 (Co j Valve Approx, Valves Location No. of Pipe Per Ref, to Valve Line Isolated Lines Size (in.) Line Class Drvwell Tvoe (61 Rx. Water Clean Up from Rx. l 4
1 6 1 A Inside MO Gate J
1 A Outside MD Cate '
Rx. Water Clean Up Ret. 1 m. 4 1 A Outside MO Gate
- .. 1 A Outside Qieck I RCIC-Turbine Steam Supply 1 3 1 A Inside MO Gate 1 A Outside M0 Gate RCIC-Turbine Exhaust 1 8 1 B Outside Check ,
1 B Outside Stop Check i RCIC Pump Suction 1 6 2 B Outside M0 Gate
] (from Supp. Pool)
! LPCS to Rx. 1 10 1 A Inside AO Check
! 1 A Outside m . Gate LPCS Pump Suction f 1 18 1 Outside M Gate
! (from Supp. Pool)
HPCS to Rx. 1 10 1 A Inside AO Check 1 A Outside MO Gate HPCS Pump Suction 1 14 1 B Outside MO Gate HPCS Test Line to Supp. Pool 1 10 1 B Outside M0 Globe HPCS Min. Flow Bypass 1 6 1 B Outside Check to Supp. Pool Drywell Eqp't. Drain Disch. 1 3 2 B Outside AO Gate Drywell Ftr. Drain Disch. 1 3 2 B Outside AO Gate l Traversing In-Core Probe 4 1 1 A Outside 50 Shear 1 A Outside 50 Ball Instrument Sensing Line 50 1 1 A Outside Hand Globe 1 A Outside Flow Check Steam Flow Measurement 8 1 1 A Outside Hand Globe A Outside Flow Check Instrument Sensing Drywell 2 1 B Outside Hand Globe Press. {
1 B Outside Flow Check Service Air to Drywell 1 1 1 B Outside Check 1 B Outside AO Globe Instrument Air to Drywell 3 2 1 B Outside Check ,
1 B Outside AD Globe l l
l
I ntinued) ZPS AMENDENT Min.
Closing Normal Power to Power to Isolation Rate or Status-Ooen (5) (6) Close (5)-(6) Signal Time (7.11.12) (9. 10) Remarks AC AC A,W,Y,RM 12"/ min Open )
DC DC A,W,Y,RM 12"/ min Open ) Pumps stopped as a re-AC AC A,W,Y,RM 12"/ min Open ) sult of valve closure Process Rev. Flow -
Open see signal J AC AC K Standard Open ) Signal B opens, signa:
DC DC K Standard Open ) overides to close Forward Flow Process - Rev. Flow -
Closed ) Closes on rev. flow or k Forward Flow Process Rev. Flow -
Closed ) low exh. press.
Fe DC l DC RM Standard Closed Note (3) Note (3) Note (3) Not Applicable Closed AC AC RM Note (11) Closed AC AC RM Standard Open Note (3) Note (3) Note (3) Not Applicable Closed AC AC RM Note (11)~ Closed l AC AC RM Standard Closed AC AC G Standard Closed
- Note (9)
- Process Rev. Flow - ' Closed Air /AC Spring A,F S tandard Open Air /AC Spring A,F Standard Open. s DC DC RM - Open AC AC F - Closed
- be- Hand Hand - - Open ) Typical of Class A ek Spring - Excess Flow - Open ) inst. lines be Hand Hand - Open ) Typical of Class A inst. lines ek S pring -
Excess Flow - Open )
be Hand Hand - - Open Typical of Class B j ek Spring - Excess Flow - Open inst. lines l
- Process Rev. Flow Standard Open Air Spring RM Standard Open
- Process Rev. Flow - J Air S P ring B Standard Open l
7.3-7 l6
2 . . . . - . . . . . - . . - . - - . .
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j BLANK PAGE 2 ,
I F
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3 J
l Lk ZPS AMENDMENI 11 Min.
Closing Normal Isolation Rate or Status (6), Signal Time (7.11.12) (9. 10)
Remarks A,W,Y,RM 12"/ min Open )
A,W,Y,RM 12"/ min A,W,Y,RM Open ) Pumps stopped as a re-12"/ min Open O Rev. Flow ) sult of valve closure '
Open see signal J l 11
- K Standard i
Open ) Signal B opens, signal K e4 - K j Standard Open ) overides to close l gg j Rev. Flow -
Closed Rev. Flow
) Closes on rev. flow or j Closed ) low exh. press. f RM Standard Closed l 11 Note (3) Not Applicable Closed RM Note (11) e Closed -
RM Standard Open Note (3) Not Applicable Closed -
RM Note (11) Closed RM Standard Closed G Standard Closed
- Note (9)
Rev, Flow i -
' Closed A,F S tandard Open l A,F Standard Open s s RM -
Open i F t .
Closed *
- f. -
Open ) Typical of Class A
{ Excess Flow -
Open ) inst. lines l -
Open ) Typical of Class A I
Excess Flow -
Opeo ) inst. lines j - -
Open Typical of Class B
) Excess Flow -
Open inst. lines i l Rev. Flow Standard Open ) i 11
} RM Standard Open
! Rev. Flow -
l3 B Standard Open i
i 7.3-7 l6
/
TABII 7.3-1 Valve Approx. Valves Location No. of Pipe Per Ref. to Valve Line Isolated Lines Size (in.) Line Class Drywell Troe (6)
- Rx Bldg. Closed Cooling 1 6 1 C Inside m Cate Watur Inlet 1 C Outside NO Cate Rx. Bldg. Closed Cooling 1 6 1 C Inside m Cate Water Outlet 1 C Outside M0 Cate Demineralized Water in 1 4 1 B Inside Check 1 B Outside Check Drywell Purge Inlet 1 18 Outside 1 B m Butterf !
1 B Inside W ButterfI Dryvell Main Exhaust 1 18 '
1 B Inside )DButterfl Drywell Exh. Val. Bypass 1 2 1 B Outside A0 Clobe Supp. Chamber Purge Inlet 1 18 2 B Outside NO Butterf}l 1 B Inside H0 Butterfl!
Supp. Chamber Exh. Val.
Bypass 1 2 Outside 1 B A0 Clobe Supp. Chamber Main Exh. 1 18 Inside 1 B }D Butterf1:
Drywell Air Sample 1 % 2 B Outside SO Supp. Chamber Air Samples 1 1 2 B Outside SO Drywell Air Samples 2 2 B Outside SO Supp Chamber Air Samples 2 b 2 B Outside SO Supp. Chamber Air Samples 2 1 2 B Outside SO Drywell & Supp. Chamber Purge Exh. Pan Suction 18 1 1 B Outside NO Butterfly Standby Cas Treatment System Suction 1 18 1 B Outside FD Butterfly SO - Solenoi.
MO - Motor 0,!
AO - Air Ope l i '
7 ZPS AENDMENT 11 I ntinued) i
-1 (Cont '
Min. .;
Closing No ma t , i wer to Fower to IsolaMon Rate or Status Powe .. n 5 6 Close (5) (6). Signal 0 _n Time (7.17 12) (9, 10) Rema rks AC AC RM Standard 'Open AC AC RM Standard Open AC AC RM Standard Open AC AC RM Standard Open Frocess Rev. Flow -
Closed Frocess Rev. Flow -
Closed AC AC F,A,Z, (8) Standard Closed efly A AC AC F,A,Z (8) Standard Closed ,
cfly t AC AC F,A,Z (8) S tandard Closed Note (13) l cf17 '
ir/AC Spring F,A,Z (8) Standard Closed Note (13)
Air AC AC F,A,Z (8) Standard Closed
Air AC AC F,A,Z (8) Standard Closed Note (13)
AC Spring F,A,Z (8) Standard Open A
AC Spring F,A,Z (8) Standard Open A
AC Spring F,A,Z (8) S tandard Closed AC Spring F,A,Z (8) Standard Closed ,
A i AC Spring F,A,Z (8) Standard Closed A
i 7y AC AC F,A,Z (8) Standard Closed AC AC F,A,Z (8) S tandard Closed i i
Operated- 1 oid op ersted i opera ated pe rate !
j 2 7.3-8 l6 l
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1 ZPS AMENDMENT 11 Min.
Closing Normal ,
I Isolation Rate or Status l Signal Time (7.11,12) (9. 10) Remarks ,
RM S tandard Open RM Standard Open j RM Standard Open i RM Standard Open Rev. Flow -
Closed ;
Rev. Flow -
Closed !
t F,A,2,(8) S tandard Closed '
F,A,Z (8) Standard Closed f
F,A,Z (8) S tandard Closed Note (13) :
F,A,Z (8) Standard Closed I Kate (13) l F,A,Z (8) Standard Closed F,A,Z (8) S tandard Closed F,A,Z (8) Standard Closed Note (13) 11 F,A,Z (8) Standard Closed Note (13)
F,A,Z (8) Standard Open F,A,Z (8) S tandard Open '
j F,A,Z (8) Standard Closed F,A,Z (8) Standa rd Closed '
F,A,Z (8) Standard Closed f s l I
l F,A,Z (8) S tandard Closed I F,A,Z (8) Standard Closed f
I i l 5
9 h'
i 7.3-8 l6
ZPS AMENDMENT 11 TABLE 7.3-1, (Continued)
NOTES FOR TABLE 7.3-1
- 8. Reactor building vent enaust high radiation signal "2" is generated by two trip channels, each channel has two trip units. This requires one unit at high trip or one unit at downscale (instrument failure) trip, on one trip channel and one unit at high trip or one unit at downscale trip on the other trip channel in order to initiate isolation.
- 9. Valves identified by an asterisk in the " Normal Status" column can be opened or closed by remote manual switch for operating convenience during any mods of reactor operation except when automatic signal is present.
10.
Normal status position of valve (open or closed) is the position during normal power operation of the reactor (see " Normal Status" column).
- 11. 'Ihe specified closure rates are as required for containment isolation only.
- 12. Minimum closing rate is based on valve and line size.
- 13. A manual switch overrides all automatic signals on the drywell exhaust valve bypass valve, suppression chamber exhaust valve bypass valve, drywell exhaust valve and suppression charrber exhaust valve, E
7.3-12
l ZPS AMENDMENT pneumatic piping and an accumulator for those valves for which air is considere<
the emergency source of motive power for closing. Pressure and water level I
sensors are mounted on instrument racks in either the reactor building or the turbine building. Valve position switches are mounted on the valve for which f position is to be indicated. Switches are enclosed in cases to protect them l from er.vironmental conditions. Cables from each sen.sor are routed in conduits
( and cable trays to the control room. All signals transmitted to the control
- room are electrical; no pipe from the nuclear system or the primary containment j penetrates the control room. Pipes used to transmit level information from the j reactor vessel to sensing instruments terminate inside the secondary containment l
? (reactor building). The sensor cables and power supply cables are routed to cab 11 ineta in the control room where the logic arrangements of the system are formed. j 1
Electrical panels, junction boxes, and components of the reactor protec-tion system are prominently identified by nameplate. Circuits entering junction boxes or pull boxes are conspicuously marked inside the boxes. Wiring and ca-bling outside cabinets and panels are identified by color, tag, or other con-spicuous means.
7.3.4.4 Logic The basic logic arrangement is one in which an automatic isolation valve is controlled by two trip systems. Where many isolation valves close on the same signal, two trip systems control the entire group. Where just one or two valves must close in response to a special signal, two trip systems may be formed from the instruments provided to sense the special condition. Valves that respond to the signals from common trip systems are identified in the detailed descriptions of isolation functions.
11 Each trip system has two trip logics each of which receives input sig-nals, from at least one trip channel for each monitored variable. Thus, two trip channels are required for each essential monitored variable to provide independ-ent inputs to the trip logics of one trip system. A total of four trip channels for each essential monitored variable is required for the trip logics of both trip systems.
The trip actuators associated with one trip logic provide inputs into each of the trip actuator logics for that trip system. Thus, either of the two automatic trip logics associated with one trip system can produce a trip system trip. The logic is a one-out-of-two arrangement.
To initiate valve closure the trip actuator logics of both trip systems must be tripped. The overall logic of the system could be termed one-out-of-two taken twice.
l l
The basic logic arrangement described above does not apply to Class C isolation valves and testable check valves. Exceptions to the basic logic arrangement are made in several instances for certain Class A and Class B isolation valves. The reasons are explained later.
7.3.4.5 Operation During normal operation of the isolation control system, when isolation 7.3-13 I
I
/
ZPS :
ti AMENDMENT ed function plays in initiating isolation of barrier valves or groups of valves 1 illustrated in the functional control diagram on Figure 7.3-3a and 7.3-3b.
{
1.
Reactor vessel low water level. A low water level in the reacto-l vessel could indicate that reactor coolant is being lost through i a breach in the nuclear system process barrier and that the core !
t is in danger of becoming overheated as the reactor coolant in-
~
ventory diminishes.
t ab- Reactor vessel low water level initiates closure of various Class
- j. - A valves and Class B valves. The closure of Class A valves is L intended to either isolate a breach in any of the pipelines in ' I 0~
which valves are closed or conserve reactor coolant by closing of I 3" process lines. The closure of Class B valves is intended to 11 prevent the escape of radioactive materials frna the primary containment through process lines which are in communication with '
the primary containment free space.
Two reactor vessel low water level isolation trip settings are fe used to complete the isolation of the primary containment and the reactor vessel. The first reactor vessel low water level isola-tion trip setting, which occurs at a higher water level than the second setting, initiates closure of all Class A and Class B ,
valves in major process pipelines except the main steam lines. '
The main steam lines are lef t open to allow the removal of heat '
from the reactor core. The second and lower reactor vessel low water level isolation trip setting completes the isolation of the ,
primary containment and reactor vessel by initiating closure of I
? the main steam isolation valves and any other Class A or. Class B !
_ valves must be shut to isolate minor process lines. i i q The first low water level setting, which is coincidentally the same as the reactor vessel low water level scram setting, was selected to initiate isolation at the earliest indication of a possible breach in the nuclear system process barrier yet far enough below normal operational levels to avoid spurious isola-tion. Isolation of the following pipelines is initiated when reactor vessel low water level falls to this first setting (Table 7.3-1, signal A):
RilR reactor head spray 11:
Reactor water cleanup Drywell equipment drain discharge !
Drywell floor drain discharge f Drywell purge inlet I Drywell main exhaust Suppression chamber purge inlet Suppression chamber exhaust 11 Suppression chamber exhaust valve bypass 7.3-19 l
l
/
, 2PS A i
~
- y. .//) q AMENDMENT 11 Drywell exhaust valve bypass Drywell 02 analyzer sample gg e Suppression thamber 02 analyzer sample Drywell 6 suppression chamber purge exhaust suction.
Standby gas treatment system suction The second and lower of the reactor vessel low water level isolation settings, which is coincidentally the same water level i
setting at' which the RCIC system is pieced into operacion, was selected low enough to allow the removal of heat frois the reactor e
for a predetermined time following the scram and high enough sto complete isolation in time for t! e operation of core standby couling systems in the event of a 1s:rge break in the nuclear system process barrier. Isolation et u.+ following pipelines is
- initiated when the reactor vessel water level falls to this 1 second setting (Table 7,3-1, signal B):
All four main steam lines Main steam line drain Reactor water sample line
- 2. Main steam line high radiation. High radiation in the vicinity
' of the main steam lines could indicate a gross release of fission products from the fuel. High radiation near the main steam lines initiates isolation of the following pipelines (Table 7.3-1, signal C):
All main steam lines Main steem line drain '
Reactor water sample line The high radiation trip setting is selected high enough above background radiation levels to avoid spurious isolation, yet low enough to promptly detect a gross release of fission products l from the fuel. Further information regarding the high radiation j set point is available in the " Process Radiation Monitortrag" i section.
l
- 3. Main steam line space high temperature. High temperature in the space in which the main steam lines are located outside of the primary containment could indicate a breach in a main steam line.
The automatic closure of various Class A valves prevents the excessive loss of reactor coolant and the release of significant amounts of radioactive material from the nuclear system process ,
I barrier. When high temperatures cccur in the main steam 1.ine space, the following pipelines are isolated (Table 7.3-1, signal D):
7.3-20
/ l EPS
, AMEND *iENT 11 All four main steam lines Main steam drain line Reactor water sample line The low steam pressure isolation setting was selected far enough below normal turbine inlet pressures to avoid spurious isolation yet high enough to provide timely detection of a pressure regu-lator malfunction. Although this isolation function is not required to satisfy any of the safety design bases for this sys-tem, this discussion is included here to make the listing of isolation functions complete.
- 6. Primary containment (drywell) hiRh pressure l i
High pressure in the drywell could indicate a breach of the nuclear j system process barrier inside the drywell. The automatic closure '
of various Class B valves prevents the release of significant amounts of radioactive material from the primary containment. Upon detection of a high drywell pressure, the following pipelines are i isolated (Table 7.3-1, signal F): j RHR reactor head spray I 11 i
i Drywell equipment drain discharge i Drywell floor drain discharge !
Traversing in-core probe tubes Drywell purge inlet k Drywell main exhaust Suppression chamber purge inlet Suppression chamber exhaust Drywell exhaust valve bypass ;>
Drywell 02 analyzer sample Suppression chamber 02 analyzer sample gg i Drywell & suppression chamber purge exhaust fan suction f' Standby gas treatment system suction Suppression chamber exhaust valve bypass .'
l The primary containment high pressure isolation setting was {
selected to be as low as possible without inducing spurious iso-1ation trips. -]
,i
- 7. RCIC turbine steam line space high temperature (Table 7.3-1, -
signal K) ,
High temperature in the vicinity of the RCIC turbine steam line '
outside the primary containment could indicate a break in the -
RCIC s team line. The automatic closure of certain Class A valves prevents the excessive loss of reactor coolant and the release of signific.snt amounts of radioactive material from the nuclear system process barrier. When high temperature occurs in the RCIC steam i
7.3-22 -
,t 1
1 1
ZPS l
AMENDMENT 11
{
l line space the RCIC turbine steam line is isolated. The high temperature isolation setting was selected far enough above anticipated nonnal RCIC system operational levels to avoid spurious operation but low enough to provide timely detection of an RCIC turbine steam line break.
- 4. RCIC turbine high steam flow (Table 7.3-1, signal K)
RCIC turbine high steam flow could indicate a break in tt'e RCIC turbine steam line. The automatic closure of certain Class A valves prevents the excessive loss of reactor coolant and the q release of significant amounts of radioactive materials frams the '
nuclear system process barrier. Upon detection of RCIC turbine high steam flow the RCIC turbine steam line is isolated. The high steam flow trip setting was selected high enough to avoid spurious isolation yet low enough to provide timely detection of an RCIC turbine steam line break.
The logic arrangement used for this function is shown on Figure l 4.7-2A and is an exception to the usual logic requirement be- l11 cause high steam flow is the second method of detecting an RCIC turbine steam line break.
- 9. RCIC turbine steam line low pressure (Table 7.3-1, signal K) l RCIC turbine steam line low pressure is used to automatically >
close the two isolation valves in the RCIC turbine 'steam line so l that steam and radioactive gases will not escape from the RCIC 8 i
I turbine shaf t seals into the reactor building af ter steam pres-sure has decreased to such a low value that the turbine cannot be operated. The isolation setpoint is chosen at a pressure be-low that at which the RCIC ' turbine can operate effectively.
I
- 10. Reactor building ventilation exhaust high rediation j i
High radiation in the reactor building ventilation exhaust could i
indicate a breach of the nuclear system process barrier inside the primary containment which would result in increased airborne radioactivity levels in the primary containment exhaust to the secondary containment. The automatic closure of certain Class B valves acts to close off release routes for radioactive material from the primary containment into the secondary containment (reactor building). Reactor building ventilation exhaust high radiation initiates isolation of the following pipelines (Table 7.3-1, signal Z):
l i Drywell purge inlet !
Drywell main exhaust 7.3-23 l
1
pn F , ZPS I AMENDMENT 11 y
Suppression chamber purge inlet Suppression chamber. exhaust l14 Exhaust valve bypass i Drywell 02 analyzer sample Suppression chamber 02 analyzer sample Suppression chamber exhaust valve bypass 11l Drywell & suppression chamber purge exhaust fan suction Standby gas treatment suction i 7
l The high radiation trip setting selected is far enough abeve back- ;
ground radiation levels to avoid spurious isolation, but low ;
enough to provide tianely detection of nuclear system process nar- I rier leaks inside the primary containment. Because the primary I containment high pressure isolation function and the reactor vessel low water level isolation function are adequate in effecting
. appropriate isolation of the above pipelines for gross breaks, the reactor building ventilation exhaust high radiation isolation function is provided as a third redundant method of detecting breaks in the nuclear systes process barrier significant enough to i require automatic isolation.
7.3.4.8 Instrumen tation Sensors providing inputs to the primary containment and reactor vessel ;
isolation control system are not used for the automatic control cf procedo sys- '
tem, thus separating the functional control of protection and process system.
Trip channels are physically and electrically separated to reduce the probability that a single physical event will prevent isolation. Trip channels for one monitored variable that are grouped near to each other provide inputs to dif-ferent isolation trip systems. The sensors, which are described in the following paragraphs, are used functionally in the isolation control system as illustrated in Figures 7.3-3.1 and 7.3-3.2.
Table 7.3-2 lists instrument characteristics. l6
- 1. Reactor vessel low water level signals are initiated from eight con-tacts on four indicating t>pe differential pressure switches which 6 li '
sense the difference between the pressure due to a constant reference column of water and the pressure due to the actual wacer level in the vessel. Four of the contacts are used to indicate that l6 water level has decreased to the first and higher low water level isolation setting; the other four are used to indicate that water icvel has decreased to the second and lower of the two low water level isolation settings.
The four contacts for each level setting are arranged in pairs; l 6 !
each contact in a pair provides a signal to a different trip system. !
IW pipelines, attached to taps above and below the water level on the reactor vessel, are required for the differential pressure measurement for each pair of contacts. The two pairs of pipelines l6 7.3-24 g
2PS
.1 AMENDMENT 11 l11 a. Appropriate responses of the core. standby cooling system shall be initiated automatically by control systems so that no deci-sion or manipulation of controls is required of plant opera-l tions personnel.
l gi ; / b. Intelligence of the responses of the standby cooling systems shall be provided to the operator by control room instruments-tion so that faults ir the actuation of safety equipment can be
, diagnosed.
- c. Facilities for manual actuation of the standby cooling systems shall be provided in the control room so that operator judgment 7
i and action is possible, yet reserved for the remedy of a deficiency in the autamatic actuation of the safety equipment.
- 5. To meet the reliability requirements of safety design bases I and 2, the following safety cesign bases are specified:
- a. No single failure, maintenance, calibration, or test operation shall prevent the integrated operations of the core standby cooling systems from providing adequate core cooling.
- b. Any installed means of manually interrupting the availability l6 of the core standby cooling systems shall be under the physical control of the main control room operator or other supervisory
- personnel.
- c. The power supplies for the controls and instrumentation for the l 6 core standby cooling systems shall be chosen so that core cooling can be accomplished concurrently with a loss of off-j site auxiliary ac power.
$ d. The physical events that accompany a loss-of-coolant accident l6 l shall not interfere with the ability of the core standby I cooling systems' controls and instrumentation to function
} properly.
11 ?
- e. , Earthquake ground motion shall not impair the ability of the l6 control and instrumentation of the essential core standby cooling system network to function properly.
l 11
, 6. To provide the operator with the means to verify the availability
) of the core standby cooling systems, it shall be possible to test j the responses of the controls and instrumentation to conditions i
representative of abnormal or accident situations.
V l 7.4.3 Description l
i 7.4.3.1 Identification j The controls and instrumentation for the core standby cooling systems
, are identified as that equipment required for the initiation and control of the
}
H 7.4-2
/
ZPS AMENDMENT 11
> I following:
High pressure core spray (HPCS) system Automatic depressurization system (ADS) j Low pressure core spray (LPCS) system Low pressure coolant injection (an operating mode of the residual heat removal system RHRS)
The equipment involved in the control of these systems includes automatic injection valves, electric pump controls, relief valve controls, and the switches, contacts, and relays that make up sensory logic channels. Testable ,
{ check valves and certain automatic isolation valves are not included in this !
y description because they are pertinent to the primary containment and reactor i vessel isolation control system, Subsection 7.3, " Primary Containment and i Reactor Vessel Isolation Control System".
I
!. Electrical panels, junction boxes, and components of the reactor protec-l tion system are prominently identified by nameplate. Circuits entering junction 11
.I boxes or pull boxes are conspicuously marked inside the boxes. Wiring and i
cabling outside cabinets and pant..s are identified by color, tag, or other i
conpicuous means.
To assure the functional capabilities of the core standby cooling sys-tems during and after earthquake ground motions, the controls and instruments-
, tion for each of the systems are designed as Class I seismic design equipment as described in Appendix C.O. This meets safety design basis 5.
Successful core cooling for a specified line break accident is depicted in Figore 7.4-1 for small line breaks,
b, the low pressure core cooling phase is accomplished by LPCS or any 6 2 RHR Systems or HPCS.
~'
Similarly, the large break model requires LPCS, the 3 RHR Systems or HPCS for successful core cooling.
7.4.3.2 High Pressure Core Spray (HPCS System Control and Instrumentation)
- 7. 4. 3.2.1 Identification and Physf cal Arrangement When actuated, the HPCS system pumps water from either the condensate storage tank or the suppression chamber to the reactor vessel via a sparger located above the core. The HPCS includes one electrically driven pump, one diesel engine-generator and auxiliaries, appropriate valves and piping, control devices for this equipment, censors, trip channels, and logic circuitry. When off-site auxiliary power is not available, the HPCS system is powered by a diesel generator driving the single motor patep. The HPCS system is designed to operate of f the off-site auxiliary power sources as well as the diesel gener-ator. The piping and instrumentation diagram is shown in Figure 7.4-2.
l6 7.4-3
\ l t
ZPS AMENDMENT 11
}
l l the reactor vessel low water level and primary containment high pressure j initiating signals. By manually resetting these signals the delay times are re- l cycled. The operator can use the reset switch to delay or prevent automatic opening of the relief valves if such delay or prevention is prudent.
Two ADS trip systems are provided, ADS A and ADS B (see Figure 7.4-6).
Division I sensors for low reactor water level and high drywell pressure initiate ADS A, and Division II sensors initiate ADS B. The relays of one trip ,
system are mounted in a different cabinet than the relays of the other trip system.I ,I l The ADS A trip system actuates the "A" solenoid of each ADS valve. Simi- i larly, the ADS B trip system actuates the "B" solenoid of each ADS valve. Actua-tion of either solenoid causes the ADS valve to' open to provide depressurization. i Within each trip system is a timer to delay actuation of the ADS valves t for a pre-determined interval. ' l surization if it is operable. This delay Instrument permits and specifications HPCS settings to are provide listed rea in Table 7.4-2.
i The reactor vessel low water level initiation setting for the automatic s depressurization system is selected to open the relief valves to depressurize the reactor vessel in time to allow adequate cooling of the fuel oy the LPCI sys-tem following a loss-of-coolant accident in which the HPCS fails to perform its function adequately. The primary. containment high pressure setting is selected j to be as low as possible without inducing spurious initiation of the automatic "
depressurization system. This provides timely depressurization of the reactor vessel in the event that' the HPCS fails af ter it successfully starts following a loss-of-coolant accident.
The low pressure pump discharge pressure setting used as a permissive for 6
depressurizations is selected to assure that at least one of the enree LPCI pumps 4 i
for the LPCS pump has received electrical power, started, and is capable of j j delivering water into the vessel. The setting is high enough to assure that the j i pump will deliver at near rated flow without being so low as to provide an l erroneous signal that the pump is actually running. {
7.4.3.3.3 Automatic Depressurization System Initiating Instrumentation J
! The pressure and level switches used to initiate one ADS solenoid are i l separated from those used to initiate the other solenoid on the same ADS valve. 6 t
- Reactor vessel low water level is detected by four switches that measure differ-
! rential pressure. Primary containment high pressure is detected by four pressure
' switches, which are located outside the primary containment and inside the reac-tor building. The level instruments are piped individually so that an instrument i
' pipeline break will not inadvertently initiate auto blowdown. The reactor vessel lou water level signals and the primary containment hfEh pressure signals are arranged to seal into the control circuitry; they must be manually reset to clear.
The LPCS pump and LPCI pumps discharge pressure switches are discussed in Paragraph 7.4.3.4 and 7.4.3.5 respectively. In this instance, pump discharge pressure twitches must be connected into both ADS trip systems to preclude initia-tion of ADS when the low pressure core cooling systems are not available. 6
}
l 7.4-10 e
i 4
ZPS j AMENDMENT 11
- 1. Test bypass valves are closed and interlocked to prevent opening.
- 2. When ac power is available, the LPCS starts.
- 3. When reactor vessel pressure drops to a preselected value, valves open in the pump discharge lines allowing water to be sprayed over the Core.
Two automatic initiation ftmetions are provided for the LNS system; reactor vessel low water level and primary containment (drywell) high pressure.
Reactor vessel level indicates that the fuel is in danger of being overheated because of insufficient coolant inventory. Primary containment high pressure is indicative of a break of the nuclear system process barrier inside the drywell.
-0 h The LPCS (and LPCI A) initiation logic is depicted on Figure 7.4-7 in a one-out-of-two twice network using both level and pressure sensors. The initia-I l tion signal vill be generated whenever:
a
- 1. both level sensors are tripped, or when
'11 2. both pressure sensors are tripped, or when j 3. either of two other combinations of one level sensor and one i pressure sensor are tripped.
i This initiation signal is derived from sensors separated from the other CSCS sensors, and is used to start both LPCS and LPCI A. The instruments used to j
detect reactor vessel low water level and primary containment high pressure are the
- the same ones used to initiate LPCI A. Once an initiation signal is received by
/ the LPCS control circuitry, the signal is sealed in until manually reset. The i seal-in feature is sho.m in Figure 7.4-9 g.
7.4. 3. 4. 3 IECS System Pump Control g' The control arrangements for the LPCS pump is shown in Figure 7.4-8. The l6 LPCS pump can be controlled by a control room remote switch, or the automatic con-trol system. A pressure indicator switch is installed in the pump discharge pipe-line upstream of the pump discharge check valves. Such an arrangement provides indication of proper pump operation following an initiation signal. This pressure signal is used in the automatic depresserization system to indicate pump failure. i The location of the pressure instrument relative to the discharge check valves presents the operating pump discharge pressure from concealing a pump failure.
7.4.3.4.4 IECS System Valve Control The control arrangements for the various automatic valves in the low pres-sure core spray system are indicates in Figure 7.4-9. Motor-operated valves are ,
prorided with limit swiches to turn off hee motors when the full open positions l11!
are resched. Torque-switches are also provided to control valve motor forces when valves are closing. In addition to torque-switches thermal overload devices are used to trip motor operated valves when they are overloaded. Appropriate inter- 7' '
locis prevent the incorrect positioning of the valves by manual action af ter the system has ocen automatically actuated. All motor-operated valves are equipped with limit switches that provide control room indication of valve position. I Each automatic valve can be operated from the control room.
Valves which have vessel and containment isolation requirements are I described in Subsection 7.3. The time required for the valves pertinent to l 7.4-15
5
?PS i i
AMENDXENT 11 Figures 7.4-10.1 and 7.4-10.2 show the entire residual heat rencva] l6 '
system including the equipment used for ?PCI operatim. The following list of equipment itemizes essential components for which control or instrumentation is required:
I Three KHRS main system pumps i
Pump suction valves '
LPCI injection valves 1
The instrumentation for LPCI operation provides inputs to the control circuitry for other valves in the residual heat removal system. This is neces- !
sary to ensure that the water pumped from the suppression chamber by the main i system pumps is routed directly to the reactor. 'Ihese interlocking features are described in this section.
ft LPCI operation uses three pump loops, each loop with its own separate i 6 vessel injection nozzle. Figures 7.4-10.1 and 7.4-10.2 show the location of in-l 6
)
g -
struments, control equipment, and LPCI components relative to the primary con- {
tainment. Except for the LPCI testable check valves the components pertinent J l
t to LPCI operation are located outside the primary containment. j i The power for the main system pumps is supplied from ac buses that can receive standby ac. power. Two of the pumps derive their power from a dif-ferent bus than that used for the other pump. Motive power for the automatic q valves used during LPCI operation comes from the same bus from which the pumps J11 for each side are supplied. Control power for the LPCI components comes free the de buses. Trip channels for LPCI "A" and LPCI "B" and "C" are shown in Figures 7.4-11.1 and 7.4-11.2. 6
}6 LPCI is arranged for automatic and remote-manual operation from the control room. The equipment provided for manual operation of the system allows the operator to take action independent of the automatic controls in the event of a loss-of-coolant accident.
7.4. 3. 5. 2 LPCI Initiating Signals and Logic The overall operating sequence for LPCI following the receipt of an initiation signal is as follows:
- 1. If off-site auxiliary ac power is available, all three main g system pumps start with no delay, taking suction from the sup- j pression chamber. The valves in the suction paths to the sup-pression chamber are maintained open so that no automatic i f,
action is required to line up suction. l
- 2. If off-site auxiliary ac power is not available, two of the main i
system pu=ps start immediately upon restoration of bus voltage l6 11 i and the other one starts after a 5-second delay to prevent over-loading of the source of standby power.
P 7.4-17 I
t
4 AMENDMENT 11 ZPS 1
- 3. Valves in other systems are automatically positioned so that the water pumped from the suppression chamber is routed properly.
4 I 4 When nuclear system pressure has dropped to a value at which the main system pumps are capable of injecting water into the vessel, the LPCI injection valves automatically open.
- 5. The LPCI loops then deliver water to the reactor vessel until reactor vessel water level is adequate to provide core cooling.
In the ' descriptions of LPCI controls and instrumentation that follow, Figures 7.4-11.1 and 7.4-11.2 can be used to determine the physical locations of 6 sensors. Pigures 7.4-11.1 and 7.4-11.2 can be used to determine the functional use of each sensor in the control circuitry for the various LPCI components.
Instrument characteristics and settings are given in Table 7.4-4.
Two automatic initiation functions are provided for the LPCI: reactor i vessel low water level and primary containment (drywell) high pressure.
! Reactor vessel low water level indicates that the fuel is in danger of being overheated because of insufficient coolant inventory. Primary containment high pressure is indicative of a break of the nuclear system process barrier inside the drywell. LPCI A initiation logic is consnon to the LPCS and is separated from the initiation logic for LPCI B and LPCI C. Each initiation logic uses the same one-out-of-2 twice form; however, one trip system uses 6 only uivision I sensors (LPCI A) and the other trip system uses only Division II sensors (LPCI B, LPCI C). Each trip system consists of 2 level switches and 2 drywell high pressure switches connected into a one-out-of-two twice configur-ation.
Once an initiation signal is received by the LPCI control circuitry, the signal is sealed in until manually reset. The seal-in feature is shown in Figure 7'.4 11.1.
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- 7. 4 . 3.5. 3 LPCI Mode Pump Control The functional control arrangement for the main system pumps is shown i in Figure 7.4-11.1. The reaction of the pumps to an initiation signal depends on l11 the availability of power. If of f-site auxiliary ac power is evailable, all three LPCI pumps start with no delay. If off-site auxiliary ac power is not available, two main system pumps start immediately and the other one starts after a 5-second delay that is designed to prevent overloading the source of standby 6l11 power. The time delays are provided by timers which are listed in Table 7.4-4.
The delay times for the pumps to start under conditions where normal ac pcwer is not available include approximately 3 seconds for the start signal to develop f af ter the actual reactor vessel low water level or primary containment high pressure occurs,10 seconds for the standby power to become available, and a l
I 7.4-19
AMENDE NT 11 ZPS 1
i l- { scquencing delay to prevent overloading the source of Atandby power. The )
j following total delay times are from the time of ths. accident to the start of
- the main system pumps
Pump A -
18 seconds delay l
Pump B -
18 seconds delay i Pump C - 13 secorbi delay 1
The operator can manually control the r.:mps from the main control room. j This is so that the operator may use the pumps ,or other purposes (e.g., con- l tainment cooling) after enough time has passed for LPCI operation to flood the l core to approximately 2/3 of the height of _ the core, which is above the re- '
quired flood level for adequate core cooling.
A pressure indicator and switch is installed in each pump discharge pipeline upstream of the pump discharge check valves. Such an arrangement pro- l vides indication of proper pump operation following an initiation signal.
This pressure signal is used in the automatic depressurization system to in-dicate pump failure. The location of the pressure instruments relative to the discharge check valves prevents the operating pump discharge pressure from concealing a pump failure.
The main system pump motors are provided with overload protection. 6 The overload relays are applied so as to maintain power on the motor as long as possibic without harm to the motor or'immediate damage to the emergency power system. i 6
7.4.3.5.4 LPCI Valve Control All automatic valves used in the LPCI function are equipped with re-mote-manual test capability, so that the entire system can be operated from the control room. Motor-operated valves are provided with limit switches to turn off the motors when the full open positions are reached. Torque switches ;
are also provided to control valve motor forces when valves are closing. Ther- '
d mal overload devices are used to trip overloaded motor operated valves. Valves . 6 which have vessel and containment isolation requirements are described in Sub- i section 7.3. The time required for the valves pertinent to LPCI operation to travel full stroke is as follows:
LPCI injection valves -
12 seconds Containment cooling valves -
Standard The main system pump suction valves to the suppression pool are nor-mally open. To position the valves a keylock switch inust be turned in the l control room. Upon receipt of an LPCI initiation signal certain reactor 7.4-20 I
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/ l A>ENDMENT 11 !
ZPS Each detailed requirement in Table 7.4-9 is referenced, there possible, to the most significant condition originating the need for the requirements by identifying a matrix block on one of the six matrices 3 of Appendix G.O.
The matrix block references are given in parenthesis beneath the detailed re-quirements in the " minimum required for action" columns of Table 7.4-9 and are coded as follows:
Example of Matrix Reference F34-84 where F signifies BWR operating state F, 34 signifies a loss of coolant accident (row #34), 1 and 84 signifies the incident detection circuitry (column #84)
In most cases, the basis for an operational nuclear safety requirement is clear from the information provided by the previously noted references.
The incident detection circuitry requirements in states C, D, E, and F re-sult from considerations for the loss of coolant accident or lesser ' cases of this design basis accident. There are no requirements on the IDC in states A and B. Manual start is shown in Table 7.4-9 to indicate the need for the CSCS in these states, but none of the IDC components are required to assure the manual start capability.
5 There is one HPCS trip system and two ADS trip systmas. HPCS and ADS '6 11; systems function as a pair to satisfy the single failure criterion whenever the nuclear system is pressurized above 113 psig. Below 113 psig, the low
- pressure core standby cooling systems, as well as the HPCs, can deliver 100 percent of design flow and no requirements are made upon the ADS trip system.
There is one trip system for 2 LPCI loops and another trip system for s
the LPCS loop and the other LPCI loop. These trip systems must be operable any time the nuclear system is pressurized. They must be operable above 113 psig, because they would be required any time the ADS system was actuated.
The operable LPCI and LPCS pump discharge pressure channels required 11 in the ADS trip system must be in operable low pressure pump cooling paths.
A low pressure pump cooling path includes an RHRS or LPCS pump and the cor-responding piping and equipment required to complete a core cooling path.
Tt.e operational nuclear safety requirements found in Table 7.4-9 are sumarized in Table 7.4-10.
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ZPS A.'fE2 MENT 1:
l and signal readout equipment. Each APRM channel can average the output signals f rom up to 24 LPRM's. The number .of LPRM's associated with an Ah M depends upon the number of LPRM detector assemblies in the reactor core. Assignment of LPRM's to an APRM is made using the pattern illustrated in Figure 7.5-2.
The letters at the detector locations in Figure 7.5-7 refer to the exial posi-tions of the detectors in the LPRM detector assembly. Position A is the Lot-tom position, positions B and C are above position A, and position D is the topmost LPRM detector position. APRM's A. C, and E are powered from the same ac bus used for trip system A of the reactor protection system; APRM's B, D and F are powered from the ac bus used for trip system B. The + 20-volt de bus used for a given APRM channel is the same as that used for the LPRM'e p ro-viding inputs to the APRM. The pattern in Figure 7.5-7 is for the APRM's associated, with trip system A of the reactor protection system. The pattern provides out the core. LPRM signals from all four core axial LPRM detector poritions through-Assignments of LPRM's to APRM's associated with trip system B j of the reactor protection system are made using the same array as for trip sys- !
tem A, but shif ted over by one LPRM detector assembly.
The APRM amplifier gain can be adjusted by a combination of fixed rie- 1 l
sistors and potentiometers to allow calibration to power as determined by a heat balance. The averaging circuit automatically corrects for the number of (
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unbypassed LPRM amplifiers providing inputs to the APRM.
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Each APRM channel receives tvc independent redundant flow signals re- l preventative of total core flow. These two flow signals are used with the three APRM's in one trip system. Each signal is provided by summing the flow signals II((
from the two recirculation loops. These redundant flow signals are sensed I' rom j
two flow elements, one in each recirculation loop. Each flow element has tsao i sets of taps, the differential pressure from these taps is routed separately to eight differential pressure transducers. The transducers and other signal conditioning equipment are separated in a way which provides two independent flow !
signals for use by the three APRM's in each trip system. No single active component failure can cause more than one of these two redundant signals to read in correctly.
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in order to obtain the proper (most conservative) reference signal ocsder '
single failure conditions, each APRM is supplied with the two redundant and isolated flow signals associated with that trip system. These flow signals are routed to a low actuation circuit which selects the lower (more conservative) of the two signals for use as the scram trip reference for that particular APRL Because there are two redundant flow units assigned to each trip system, one flow unit in each trip system can be bypassed for a short period of time. This design meets the requirements of IEEE 279, dated August 30, 1968.
7.5.7.3.4 Trip Function The trip units for the APRM's supply trip signals to the reactor pro-tection system and the reactor manual control system. Table 7.5-4 f cemizes 7.5-17
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2PS AMENDMENT 11 There are numerous indications of reactor vessel water level in the reactor building. Almost all of the level-measuring instruments indicate locally, as shown in Figures 7.8-2. Some of the instruments derive their level measurements from the instrument pipelines in which the temperature-equalizing columns are installed. Thus, temperature-compensated, as well as uncompensated, level indications are available in the reactor building.
There are seven separate reactor vessel water level indications provided in the control room. These are continuously displayed on various bench boards.
Two of the control room level indications are derived from the lines using the temperature-compensated reference columns of water. Of the five remaining level measurements, two come from the level transmitters differential pressure switches provided for the feedwater control system, two (one of which is record- 11 ed) come from the instruments used to measure the water level inside the core shroud for RHR operation, and one uses a separate reference column of water located so that water level indication is possible all the way to the top of the vessel. A level recorder that receives level signals from level transmitters in the feedwater control system provides a continuous record of reactor vessel water level. This same recorder provides high and low level alarms. Table 7.8-1 lists the specifications for level instruments not previously described with other systems.
Figures 7.8-2 gives a chart showing the water levels at which various automatic alarms and safety actions are initiated. Each of the actions listed
.1 is described and evaluated in the subsection of the Safety Analysis Report where the system involved is described. The following list tells where various level measuring components and their set points are discussed.
Level Instrumentation Subsection in Which Discussg Level switches for initiating scram Reactor Protection System (7.2)
Level switches for initiating pri- Primary Containment and Reactor mary containment or reactor vessel Vessel Isolation Control System isolation (7.3) i j Level switches used for HPCS, Core Standby Cooling Systems Control l
LPCI, automatic depressurization and Instrumentation (7.4)
- system, or recirculation pump trip i Level switches and recorder used to Core Standby Cooling Systems Control measure water level inside core and Instrumentation (7.4) shroud level transmitters and recorders Feedwater System Control and used for feedwater control Instrumentation (7.10) j t
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l ZPS AMENDMENT 11 1 I
support assembly differs from the pressure at the core exf t by the height of water external to the fuel channels, the pressure contributions from flow l through the control rod guide tubes, and any leakage around the junctions of the fuel support piece and the core support assembly or fuel assemblies.
Instrument pipelines leading from the reactor vessel to locations outside the drywell are provided with one manual isolation valve and one excess I flow check valve. All of the flow and differential pressure instruments are l located outside the primary containment. i l
This instrumentation permits the determination of total core flow in j 11 two ways. The first method is the readout of the summed flow measurements from all the jet pumps. The second method includes the use of jet pump prototype performance data, the jet pump differential pressures, and the differential pressure between the reactor vessel annulus and the core inlet plenum. A temporary correlation can also be made to define core flow as a function of reactor operating power level and the readout of the pressure difference between the reactor vessel annulus and the core inlet plenum. This correlation is of a temporary nature because it will change with a fixed core arrangement over a period of time as a result of crud buildup on the fuel. The control room flow rate readouts of the specially calibrated jet pumps can be used to cross-check the flow rate readouts of'all the other jet pumps. A discrepancy in the cross-checks is reason enough to check local flow indications.
Flow in each recirculation loop is used to change the APRM scram setting f gg as described in Paragraph 7.5.7.3.3. Flow in each recirculation loop is measured l by a flow element as shown in Figure 7.8-1. Indicated recirculation loop flow rates can be checked by using recirculation pump performance curves and the differential pressure between the reactor vessel annulus and the core inlet plenum. Extreme accuracy of the flow rate operational readouts in the main control room is not necessary because precise measurements can be obtained during reactor operation if they are desired. It is sufficient to periodically demonstrate that the reactor recirculation flow rate is at least the design flow rate during operation at rated power.
7.8.5.4 Reactor Vessel Internal Pressure _
i Reactor vessel internal pressure is detected by pressure switches, indicators, and transmitters from the same instrument pipelines useu for reactor vessel water level measurements. Two pressure indicators that sense pressure from different, separated instrument pipelines provide pressure indications in the reactor building. Two reactor vessel pressure indications are provided in the control room. These cotre from the two pressure trans-mitters used in the feedwater control system. Reactor vessel pressure is i continuously recorded in the control room on two recorders.
Each recorder receives a pressure signal from one of the feedwater control system pressure transmitters.
7.8-7 ,
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ZPS AMENDMENT 11 TABLE 7.12-4 PROCESS RADIATION MONITORING SYSTEM ENVIRONMENTAL AND_
POWER SUPPLY DESIGN CONDITIONS Sensor Location Control Room Design Design Parameter Value Range Value Range Temperature 25'c O*C to +60*C 25'c S'to +50'c ,
77'F 32*F to 140*F 77'F 40*F to 120'F g '.
Relative 5'to 60%
Humidity 50% 20*to 98% 50%
Power, AC 115V/2307 1 10% 115V/230V i 10%
50/60 cps 15% 50/60 cps 15%
Power, DC +24 vde +21.5 to +29.5 vde
-24 vde ~21.5 to -29.5 vde I
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AMENDMENT 11
/ EGEND
/ - 4 WA Y VAL VE (MA/N VALVE /N OPEN PCSITION) 2 - 4 WAY VAL VE 3- AIR STORAGE l n--R" ' !
l TA N K U* i' "l l Y' Ih% f AIR MA/N CYLINDER I64LVE 0/V l
? 4-3 WAY VALVE l 5-3 WAY VALVE l 6- 3 WAY VALVE 7- SMD COWA y ,wg VALVE B- HYDRAULC CYL/NCER
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9- SWING CHECK YA LVE. , i 4
SLOW SPEED T T EXERCISING CIRCu/T f,lRt6T ~
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W l20V- 60~ E F2OW SQL -
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N op 3 AIR SUPPLY WM. H. ZIMMER NUCLEAR POWER STATION PRELIMINARY $AFETY ANALYSIS REICET FIGURE 7.3-2 MAIN STEAM LINE ISOLATION VALVE SCHEMATIC CONTROL DIAGRAM T.
i ZPS AMENDMENT 11 7.1-'1 (ZPS - February 23, 1971, AEC Question 7.4) 3 QUESTION The l'isting of protection systems on Page 7.1-1 appears incomplete since systems such as SGTS, refueling interlocks, reactor manual control system (portions), auxiliary supporting systems, and process radiation moni-toring system (pertions) are not included. You are requested to re-examine this listing to assure completeness and discuss the basis for omitting any of the above-mentioned systems in the present or revised listing.-
ANSWER For the answer refer to Paragraphs 7.1, 7.1.1, 7.1.2 and 7.1.3 of the Wm. H. Zimmer PSAR, also Paragraphs 7.7.4.1 and 7.7.5 Protection systems that are not listed in Section 7.0 ar2 listed in other applicable sections.
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1 ZPS AMENDNINT 11 1 7.2-1 (ZPS - February 23, 1971, AEC Question 7.9)
QUESTION There are several areas of the protection system design described in PSAR which appear to be inconsistent with your safety design basis (SDB). For-example, (1) a single scram reset switch appears to violate SDB 7f and 7g; (2) the scram discharge volume high water level scram channels appear to violate SDS 11; (3) the bypass of the turbine control valve fast closure and turbine stop valve closure scrams by low reactor pressure and turbine first stage pressure signals respectively each appear to violate SDB 7a and 7g.
You are requested to identify all areas of protection system design which are in conflict with the SDB and provide analyses to support the exceptions.
ANSWER Compliance of these RPS designs with IEEE-279 is given in the following
! sections of Topical Report NEDO-10139, June,1970:
(1) RPS scram reset switch Section 2.2.19 pages 2-205 thru 2-210 (2) RPS discharge volume bypass Section 2.2.14 pages 2-169 thru 2-176 (3) RPS turbine trips bypass Section 2.2.16 pages 2-185 thru 2-190 l
l In the first instance, the RPS reset switch is located " upstream" of {
the individual reset relays, and complete failure of the reset switch will not prevent a reactor scram from being initiated or going to completion.
Initiation is accomplished by opening of trip channel contacts to the RPS scram
'contactors. Completion of protective action is assured by the recent addition of time delay relays to inhibit manual reset within the first ten seconds
, following initiation of reactor scram. Seismic requirements are also satisfied for all protection system " essential" components, (See answer to question February 23, 1971, AEC Question 7.24).
In the second instance, oper'ational availability of each trip channel and trip logic is possible during reactor operation as described in paragraphs 2.2.14.9 and 2.2.14.10 on pages 2-170 and 2-171 of Topical Report NEDO-10139 (June, 1970) . It is not possible to test the scram discharge volume bypass in the startup and run modes of reactor operation due to its interconnection with the reactor system mode switch. Nevertheless, it is possible to test that "no" bypass is in effect at any time; this test assures that plant safety is lef t intact. Imposition of the bypass in startup or run modes of operation is not a l permitted condition; therefore, a test of its availability is merely for opera- l ting information rather than for any safety considerations. l Finally, the remaining trip bypasses for turbine trip and generator trip are in compliance with the design criteria as discussed on page 2-186 7.2-1
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2PS I
AMENDMENT 11
- of the Topical Report. During test of each individual bypass circuit, the protective action trip channel will be momentarily bypassed; however, the 1 actual interval is kept as short as possible in accordance with IEEE-279 paragraph 4.11.
In summary, the RPS design is in conformance with the Safety Design i Basis and with IEEE-279. j l
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ZPS AMEND m 7 11 7.2-2 (ZPS - February 23, 1971 AEC Question 7.12)
QUESTION Provide the criteria and their bases used in establishing the ranges ]
of the protection system sensors.
ANSWER Setpoints for the RPS sensor trips are determined from operational I experience and analyses of transients and are set sufficiently far from
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expected operating values of process variables such as to avoid spurious scrams and operating experience yet close enough to satisfy the design bases .
and safety requirements. 4 For the majority of RPS sensors, the sensor range is dictated by the results of the BWR transient analysis. This analysis specifies the protection j action setpoint ior the following sensors: '
Main Steamline Isolation Valve Switch 90 to 100 percent open Closure Turbine Stop Valve Closure Switch 90 to 100 percent open j i
Turbine Control Valve Fast Closure Refer to Page 7.2-14 item 5
)
Reactor Vessel Low Water Level Fixed height above vessel zero Neutron Monitoring System (APRM) Fixed 120% of rated power Neutron Monitoring System (IRM) Fixed 95% of full scale j t i Reactor Vessel High Pressure Fixed 1055 psig setpoint
{ Turbine First Stage Pressure Fixed pressure corresponding to j (BYPASS) 30% power i
5 The remaining sensor setpoints are dictated by the design character-istics of the individual plant systems, such as:
Discharge volume High Water Fixed gallonage in instrument Level volume Main Steamline High Radiation Fixed multiples of normal back-ground Primary Containment High Pressure Fixed 2 psig serpoint .
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l ZPS 1
AMENDMENT 11 Reactor Vessel Low Pressure Fixed 600 psig setpoint (Bypass) )
From these established setpoints, the sensor range is determined taking into account the available industry instruments, the maximum and minimum values for the operating variables through the plant lifetime, and the amount of margin considered prudent by the instrttnentation designer of the plant.
h 7.2-4
ZPS ,
AMENDMENT 11 7.2-3 (ZPS - February 23. 1971, AEC Question 7.27)
QUESTION Describe how reactor protection system and engineered safety equipment will be identified physically as safety related equipment in the plant to assure appropriate treatment, particularly duririg maintenance and testing operations.
ANSWER Equipment associated with the Reactor Protectima Systems (RPS), the Nuclear Steam Shutoff System (NSSS) and the Core Stan&y Cooling System (CSCS) will be identified ,so that it is apparent that:
l 2. the grouping (or division) is part of the enforced segregation with
! which the particular equipment item is associated.
j This identification will consist of marting panels and equipment racks t of the RPS, NSSS, and CSCS with marker plates that are conspicoonsly different in color than those for other panels or racks. These markers will incit.de identification of the proper division of equipment vitkin the system.
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! Physical identification of all components of other reactor protection
! system and engineered safety equipment is described in #etail in Subsection 8.10
. and Table 8.10-3 of the Wm. H. Zininer Nuclear Pcwer Stztion ISAR, Amendment 7.
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2PS AMENDMENT 11 l 7.2.3.6-1 (ZPS - February 23, ,* 971 AEC Question 7.8)
{
QUESTION j In previous BWR de. signs a reactor trip function derived from sensors monitoring low-condenser vacuum was provided. Discuss your reasons for l deletion of this reactor trip function in your design.
ANSWER On GE BWR plants prior to Dresden 2, reactor scram was not directly initiated from either turbine trip or generator trip. However, loss of con-denser vacuum signalled a direct reactor scram for these plants anticipating loss of the reactor heat sink.
Starting with the Dresden 2, turbine trip scram and generator
- load-rejection trip scram were added to the Reactor Protection System along '
with retention of the Condenser Low Vacuum scram.
In late 1967, it was determined that 'since loss of condenser vacuum results in direct closure of the turbine stop valves through the turbine control system and since the turbine trip rer 4 ting from stop valve closure would initiate a direct reactor scram, a change was instituted to remove the condenser low vacuum scram on the basis that: )
i (1) Reactor scram would be initiated from turbine stop valve closure E f (2) Condenser low vacuum scram' was redundant with the turbine trip scram j
(3) Turbine trip scram encompasses other initiating events besides condenser low vacuum p (4) Turbine trip is a more direct variable for protection of the IE reactor m
1 Consequently, protection of the reactor is assured by sensing an im:cincat loss of reactor heat sink through closure of the turbine stop valm fast-closure of the turbine c~. :rol valves or closure of the main stee lic.e isolation valves. Adequi . reactor protection for the loss of condenser vacuum is ass 0 red through the turbine trip scram resulting from turbine step valve closure. This protection will be provided for the Wm. H.
l_ Zirmaer Nuclear Power Station.
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ZPS AMENDMENT 11 7.2.3.9-1 (ZPS - February 23, 1971 AEC Question 7.10)
QUESTION Resolve the inconsistency which exists between Table 7.2-1 and Page 7.2-14 concerning the turbine control valve fast closure design.
ANSWER Refer to Paragraph 7.2.3.9 and Table 7.2-1 of the Wm. H. Zimmer Nuclear i Power Station PSAR Amendment 11. _
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ZPS
, AMENDMENT 11 7.4.3-1 (ZPS - February 23, 1971, AEC Question 7.15)
QUESTION From our review, it appears that the signals required to actuate LPCS and LPCI lack diversity. Injection valve differential pressure signals are needed in coincidence with either low reactor water level or high drywell pressure to permit the injection valves of these redundant systems to open.
You should evaluate the design in this regard and justify the lack of diversity or submit a description of the design changes which will provide the diversity of signals required to actuate these redundant emergency core cooling systems.
ANSWER Reactor low water level or high drywell pressure signals are required to start the CSCS pumps. Differential pressure switches are used to open the injection valves. The use of differential pressure switches to pennit open-ing of low-pressure core cooling systems injection valves is justified because:
- a. of the redundancy between the diverse systems which compose the CSCS network
- b. all active and control components and power sources are tested with sufficient frequency to insure a high individual component availability
- c. the combined CSCS network is designed to be single failure proof although the individual systems within the CSCS network, with the exception of the ADS, are not designed to be single failure proof.
7.4.3-1
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i ZPS f AMENDMENT 1; 7.5.7.3.3-1 _(EPS - February 23. 1971. AEC Ouestion 7.16)
OUESTION l
i j An inconsistency exists on Page 7.5-17 of the PSAR with respect to the number of flow signals to be provided to bias each APRM channel. There is also I
an apparent conflict of design requirements for this system in that it is first stated that the design will function with a " single active component failure" and then it is stated that IEEE-279 will be satisfied. IEEE-279 requires that any single failure should not preclude protective action. Re-evaluate your design in this regard and submit a revised description of your design intent for 1 each of these items.
ANSWER l
i Page 7.5-17 has been revised in Amendment 11of the Wm. H. Zimmer PSAR, to remove an ambiguity that may have caused an interpretation of inconsistency.
The tenn " single active component failure" is used with the Neutron Monitoring System flow reference scram portion to exclude postulat d failures of the recirculation loop flow element from consideration of the system relative to i IEEE-279. Once the flow element is excluded, the flow reference scram meets the f single failure criterion of IEEE-279 Paragraph 4.2. j
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Failure of the recirculation loop flow element is very improbable. Of all possible failures, those failures such that apparent flow is decreased are inherently " fail safe" in that the trip point is decreased, thereby causing a scram at lower neutron flux than if the instrument were correct. Other postulated failure which would cause an apparent flow increase would cause an increase in
All calculations of transient analysis assume a 120% flux scram, therefore, even this type of failure" is accommodated.
Hence, the Neutron Nbnitoring System responses are in full compliance with IEEE-279.
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l ZPS AMENDMENT 11 7.5.8 -1 (ZPS - February 23, 1971, AEC Question 7.18) i
- l QUESTION I Justify the lack of safety design basis and system operational requirements for the RBM system. Also describe the RBM design to show that the l three level trip setting and the bypass circuitry meet IEEE-279.
]
ANSWER Because the RIN system is not considered a safety system, no safety de- 3 sign basis is given. This system provides a signal to permit operator evaluation of the change of local relative power level during control rod movement. The system is designed to meet " appropriate protection system criteria... acceptable to the Regulatory Staff". (1) Details of exceptions to the IEEE-279 criteria are explained in Paragraph 7.5.8.
The RBM system consists of two RBM channels, each of which has three level setpoints (upper, intermediate and low). The three setpoints on each RBM channel are independent of those on the other RBM channel. Each RBM channel is sensed by separate APRM reference channels.
The three bypass functions provided are:
- a. Manual Bypass - A manual bypass switch is provided with a mechanical barrier between Channel A and Channel B;
- b. Peripheral Rod - Both RBM channels are bypassed when a peripheral control rod is selected; and
- c. Power level - the RBM outputs are bypassed when the power level is less than 307..
1 Ebin I Hatch Nuclear Plant Unit 1, Docket #50-321 Amendment 7, RBM Design Changes, June 1969.
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ZPS l
AMENDMENT 11 !
l 7.6.3-1 (ZPS - February 23, 1971, AEC Question 7.19)
QUESTION Submit a discussion of the consequences of single failures in the circuitry identified on Page 7.6-1 which senses that all rods are fully inserted. The discussion should cover the circuitry from the sensors to l the final actuated device.
I ANSWER The fully-inserted condition is sensed by the closure of' a magnetically operated reed switch situated at position 00 in the rod position indicator probe. There is one probe for each control rod. A column of reed switches is mounted on the probe structure, a cylindrical magnet encircles the probe and actuates the nearest reed switch. As the control rod moves, it moves the magnet axially with respect to the probe. This causes individual columnar switches to operate. The placement of the switches in the column is such that I only one switch at a time closes.
l The switches on each probe are connected to a corresponding probe buffer card in the Rod Position Information System (RPIS) cabinet which con-
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tains electronic circuits that process the input data signals. Each probe J buffer card receives position data from its associated position indicator l probe at all times. The RPIS uses diode transistor-logic (DTL) microcircuit modules . Two grouping terminals, used for detecting the fully inserted rod condition, will have some of their inputs at a low voltage level until all rods are fully inserted. Then, as all inputs assume the high voltage level a relay driver transistor will conduct; thus furnishing current to the external relays. Contacts from these relays are then used in the refueling interlock
!. circuit (rod block) of the reactor manual control system.,
f f Since the DTL integrated circuit gate in each prnbe buffer card has I separate output signals (rod fully-inserted) to two separate grouping terminals, and these terminals are connected into two normally energized independent )
logic strings in the reactor manual control system, the. design meets the singic failure criterion, i
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l ZPS AMENDMENT 11 7.3,5-1 (ZPS - February 23, 1971, AEC Question 7.17)
WJESTION '
Discuss the design intent with respect to satisfying the safety design basie listed on Page 7.8-1 of the PSAR for the reactor vessel instrumentation during plant operation. State your basis for not including requirements for these instruments in the technical specifications.
ANSWER The design intent of the reactor core flow and vessel water level instrumentation is stated in the safety and power generation objectives listed in Paragraphs 7.8.1 and 7.8.2.
As indicated in Paragraph 7.8.5.2, there are a total of seven reactor vessel water level indicators provided in the control room. As indicated in that paragraph, several of these systems provide automatic actions as portions of other systems and specific reference is made to the other systems. Also as indicated in this subsection some of this instrumentation is connected to the feedwater control system which provides automatic control of reactor vessel water level and this control is further described in Subsection 7.10.
There are a number of technical specification requirements listed for trip settings of the water level instruments in proposed technical specification B.3 " Limiting Safety System Settings." These include settings for scram, isolation and core standby cooling system operation, all of which are for automatic actuation. As indicated in the bases for these technical specification settings, the settings to be chosen are those used in the analysis of transients in Section 14.0.
Indication of reactor core flow rate is available from instrumentation described in Paragraph,7.8.5.3. Automatic safety actions am initiated in part through the use of recirculation flow signals to the neutron monitoring system which is discussed in Subsection 7.5. Core flow determinations for operator use include the total jet pump flow recorder, the jet pump flow associated with each recirculation driving loop, and the core AP as sensed across the core support place. The.se readouts are mutually corroborating.
Recirculation flow is specified in the technical specifications in that it is one of the parameters referenced on the Fuel Cledding Integrity Safety Limit, Figure B.2-1 and also on the normal Fuel Operation Envelop, Figure B.5-1.
- 7. 8. 5-1 1
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ZPS AMENDMENT 11 !
7.8.5.2-1 _(ZPS - February 23, 1971, AEC Question 7.20)
QUESTION It appears that Figures 7.4-3 and 7.8-2 do not agree with the discussion of the automatic depressurization system level instrumentation of Page 7.8-5 of
- the PSAR. Resolve this apparent inconsistency.
ANSWER l Figures 7.4-3.1 through 7.4-3.4, 7.8-2.1 and 7.8 2.2 were updated in Amendnent 6, and Paragraph 7.8.5.2 has been updated in Amendment 11 to resolve this inconsistency.
I The above mentioned inconsistency appears to be in reference to " level transmitters" for the feedwater system. As now indicated in Paragraph 7.8.5.2; these devices are actually differential pressure switches.
- Also, one of the two signals from the instruments used to measure the I
i water level inside the core shroud is now specified as being recorded.
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ZPS AMDMENT 11 INSTRUCTIONS FOR UPDATING YOUR PSAR VOLIME 4 All changes have been indicated by a vertical line' and the Amendment i Number (11) in the right margin of the page.
- 1. At the beginning of Volume 4 remove and destroy Page 7 and replace
- with amended Page 7. Remove and destroy Pages 17 through 19 and re-tes 17 through 19.
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place with amende<.
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ZPS AMENDMENT 11 VOLthtE '.
TABLE OF CD'.TE'.TS, (Continued) i '
l, PAGE I !".! SLMtARY DESCRIPTION 10.1-1 y!
'4
. N.2 NEW Fl!EL STOP. ACE 10.2-1 8.1 SPEhT FifEL STORAGE '
10.3-1 1 a.4 TOOLS AND SERVICING EQUIht!LKT 10.4-1 18.5 FUEL POOL COOLING AND CLEANUP SYSTEK 10.5-1 I
lo.6 REACTOR BUILDING CLOSED COOLING WATER SYSTEM 10.6-1 11.7 TURBINE StflLDING CLOSED COOLING WATER SYSTEM 10.7-1 ns.b SERVICE WATER SYSTEM 10.8-1
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id.Y FIRE PROTECTION SYSTEM 10.9-1 T 1.s.10 llEATING, VE!TTIIATION, AND AIR CONDITIONING SYSTEMS 10.10-1 118. 1 1 MAKE-UP WATER TREATMEST SYSTEh 10.11-1 14.12 INSTRUMENT AND SERVICE AIR SYSTEMS 10.12-1 10.13 POTABLE AND SANITARY WATER SYSTEM 10.13-1 I".14 EQUIPMENT AND FIDOR DRAINAGE SYSTEMS 10.14-1 ,
l o. I ',
hANT PROCESS SAMPLING SYSTEM 10.15-1 I".16 C05t!UNICAf10N SYSTEM 10.16-1 l'8.17 LICllTING SYSTEM 4
10.17-1 Id.!b IIEATING BOILERS 10.18-1 10.19 PRIMARY CONTAINMENT MONITORING SYSTEM 10.19-3 .
10.20 PRIMARY CONTAINMENT liYDROGEN, OXYGEN AND FISSICN PRODUCTS SAMPLING 10.20-1 11 1
11.0 STEAM AND POWER CONVERSION SYSTEM TABLE OF CONTENTS 11.0-1 11.1
SUMMARY
DESCRIPTION 11.1-1 7
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.s I ZPS AMENDMENT 11 I
l LIST OF ZPS, FEBRUA r 23, 1971 l AEC QUESTICES f
!: AEC QUESTION RENUMBERED VOLUME
! ! NUMBER AS QUESTION PAGE OF PSAR 2.12 2.2.3-2 2.2.3-12 1 b 2.13 2.3.2.1-2 2.3.2.1-2 , 1 2.14 2.3.2.1-3 2.3.2.1-3 1
! 2.15 Later Later Later
[ 2.16 2.3.8-1 2.3.8-1 1 4.9 4.7-2 4.7-2 2 9 I 4.10 4.7-1 4.7-1 2 4.11 4.9-1 4.9-1 2 9 1
. 4.12 4.0-1 4.0-1 2 I
! 5.11 5.2.3.7-1 5.2.3.7-1 2 l 5.12 10.19-1 10.19-1 2 5.13 Later Later Later 7 5.14- Later Later Later 5.15 Later ,
Later Later 5.16 Later Later Later 5.17 5.2.3-1 5.2.3-1 2 4
7.1 5.3.3.3.3-1 s 5.3.3.3.3-1 2 7.2 5.3.3.3.2-1 5.3.3.3.2-1 2 7.3 5.3.3.3.3-2 5.3.3.3.3-2 2 7.4 7.1-1 7.1-1 3 11 7.5 Later Later Later 7.6 4.4-1 4.4-1 2 7.7 Later Later Later 7.8 7.2.3.6-1 7.2.3.6-1 3 11 7.9 7.2-1 7.2-1 3 7.10 7.2.3.9-1 7. 2. 3. 9 -1 3 7.11 Later Later Later l
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i- j ZPS .
I AMENDMENT 11 LIST OF ZPS, FEBRUARY. 23, 1971 AEC QUESTIONS, (Continued)
AEC QUESTION RENUMBERED VOLUME I NUMBER AS QUESTION PAGE OF PSAR 7.12 , 7.2-2 7.2-2 3 11 l 7.13 Later Later Later
- i. 7.14 Later La ter Later j i 7.15 7.4.3-1 7.4.3-1 3 f 7.16 7.5.7.3.3-1 7.5.7.3.3-1 3 I 7.17 7. 8 . 5 -1 7.8.5-1 I
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! 7.18 7.5.8-1 - 7.5.8 -1 i 3
i j 7.19 7.6.3-1 7.6.3-1 6
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,! 7.20 7.8.5.2-1 7.8.5.2-1 3 I I
l- 7.21 La ter Later Later 7.22 Ia te r La te r Later 7.23 I4ter Later Later 7 7.24 Later Later Later l9 7.25 D.0~1 D.0-1 5 l11 7.26 10.10.3-1 10.10.3-1 4 7.27 7.2-3 7.2-5 3 l11 l 7.28 later La ter Later 7.29 10.19-2 10.19-2 4 11 7.30 later Later Later 7.31 Later Later Later l l
8.1 8.3. 2.1 -1 8.3.2.1-1 4 j 8.2 6.3.2-1 8.3.2-1 4 l
8.3 8.3. 3 - 1 8.3.3-1 4 8.4 8.4. 3 - 1 8.4.3-1 4 8.5 S.5.4-1 8.5.4-1 4 8.6 8.4.3-2 8.4.3-2 4 8.7 6. 5. 3.1 - 1 8.5.3.1-1 4 8.8 6.0-1 8.0-1 4 18 9
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i ZPS AMENDMENT 11 l LIST OF 2PS, FEBRUARY 23, 1971 1
AEC QUESTIONS, (Continued)
AEC QUESTION RENUMBERED VOLUME NUMBER AS QUESTION PAGE OF PSAR l 8.9 8.0-2 8.0-2 4
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i 8.10 8.9-1 8.9-1 4 l' :
I ! 8.11 8.10-1 8.10-1 4 1
i 9.1 9.2.4-1 9.2.4-1 4 9.2 9.2.4.6-1 9. 2. 4. 6 -1 4 j 9.3 9.4-1 9.4-1 4 l 9.4 9'. 4. 6 -1 9.4.6 1 4 l 11 9.5 9.2.4.7-1 9. 2. 4. 7-1 4 I I
9.6 9.4.3-1 9. 4. 3 -1 4 l 10.1 Later I
Later Later
, 10.2 10.5-1 10.5-1 4 10.3 10.0-1 10.0-1 4 l 11 10.4 10.11.2-1 10.11.2-1 4 7 12.22 12.6.1-1 12.6.1-1 4 12.23 12.5.6-1 12.5.6-1 4 13.1 13.0-1 13.0-1 4 13.2 13.2.1.6-1 13.2.1.6 -1 4 13.3 13.2.1.2-1 13.2.1.2-1 4 !
9 13.4 13.0-2 13.0-2 4 13.5 13.3-1 13.3-1 4 13.6 13.6.4-1 13.6.4-1 4 13.7 13.0-3 13.0-3 4 11 14.12 14.9.1-1 14.9.1-1 4 14.13 Later Later Later 19 9 s
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ZPS AMENDMENT 11 j
INSTRUCTIONS FOR UPDATING YOIR PSAR VOLINE 4 SECTIO!! 8.0 - ELECTRICAL POWER SYSTEM This section has been revised to incorporate additional information requested by the AEC.
All changes have been indicated by a vertical line and the amendment number (11) in the right margin of the page. )
All pages (text, tables, figures) with changes have also been marked in the upper right corner of the page with " AMENDMENT 11". !
Figures that have been sitered in any way are indicated by the amendment {
manber in the upper right < crner of the figure; note that there are no other marks that would indicate changes in figure. On the page marked " LIST OF FIGURES", figures that have changed ic any way are designated by a vertical line with the amendment number alonbside the title of the figure. See example below:
FIGURE NUMBER TITLE l 2.2-1 Station Site Area Topography 11 )
To update your copy of the Wm. H. Zimmer Nuclear Power Station PSAR, t
please use the following procedure: j
- 1. In Volume 4, SECTION 8.0 - ELECTRICAL POWER SYSTEMS, Table of I
Contents, remove and destroy Pages 8.0-111 and 8.0-vi and replace with amended Pages 8.0-111 and 8.0-vi.
- 2. In Volume 4, SECTION 8.0 - ELECTRICAL POWER' SYSTEMS, remove and destroy the following text pages and replace with the appropriate pag.s listed below:
REMOVE PACE REPLACE WITH AMENDED PAGE 8.5-1 8.5-1 8.5-1-1 8.6-1 8.6-1 8.7-1 8.7-1 8.7-2 8.7-2 1
l 8.8-1 8.8-1 l 8.10-4 * ~'
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ZPS AMENDMENT 11 REMOVE PACE REPLACE WITH AMENDED PAGE 8.10-9 8.10-9 8.10-12 8.10-12 8.10-12.1 8.10-14 8.10-14
.3. In Volume 4, SECTION 8.0 - EIECTRICAL POWER SYS'IEM, remove and destroy Figure 8.3-2 and replace with amended Figure 8.3-2.
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j AMENDMENT 11 TABLE OF CONTENTS, (Continued)
PAGE 8.10 DESIGN CRITERIA FOR CABLE INSTALLATION 8.10-1 8.10.1 Obj ective 8.10-1 8.10.2 Design Basis (General Criteria) 8.10-1 8.10.2.1 Definitions and Descriptions J" 8.10-1 8.10.2.1.1 Cable Pans ~
8.10-1 8.10.2.1.2 Cable Rack 8.10-1 8.10.2.1.3 Conduit 8.10-2 8.10.2.1.4 Power Cables 8.10-2 8.10.2.1.5 Control Cables 8.10-2
- - 8.10.2.1.6 Instrument Cable 8.10-3 8.10.2.1.7 Segregation Division (ESS and PCI) 8.10-3 8.10.2.1.8 Channel 8. 10-4 8.10.2.1.9 Primary Containment Penetrations 8.10-4 8.10.2.2 Cable Ampacity 4 8.10-4 8.10.2.3 Cable Pan Loading Criteria 8. 10-4.1 l11 1 8.10.3 Engineered Safeguards Systems Cable Pan Segregation Criteria 8.10-6 8.10.3.1 Design Basis 8.10-6 8.10.3.2 Power Supplies (Control and Power) 8.10-9 8.10.3.3 Separation Details for Pans and Conduits s
- 8. 10-10 8.10.3.3.1 Mechanical Damage (Missile) Area 8.10-10 8.10.3.3.2 Fire Protection Criteria 8.10-10 8.10.3.4 Cables Entering Panels-8.10-12.1 l11 ,
8.10.3.5 Detection Circuitry for Primary Containment Isolation System 8.10-13
, 8.10.4 Segregation Criteria for Reactor Protection System Cables 8.10-13 8.10.4.1 General 8.10-13 8.10.4.2 Cable Separation in Reactor Protection System 8.10-13 8.10.4.3 Neutron Monitoring 8.10-14 8.10.4.3.1 LPRM Inputs 8.10-14 8.0-111
ZPS AMENDMENT 11 SECTION 8.0 - ELECTRICAL POWER SYSTEMS LIST OF FIGURES FIGURE NUMBER TITLE i
8.1-1 Single Line Diagram 4 8.3-1 345 KV Transmission System 8.3-2 Property Arrangement 345 KV Switchyard and Transmission Lines 7lI11 8.6-1 250V De Power Distribution System 8.7-1 125V De Power Distribution System 4 I 8.8-1 24V Dc and 48V De Power Distribution i System 8.9-1 120V Ac Distribution Instrument Buses and Uninterruptable Bus 7 1
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ZPS AMENDMENT 11 8.5 STANDBY AC POWER SUPPLY AND DISTRIBUTION 8.5.1 Safety Obiective Safety Objective of the standby ac power supply is to provide a source of electrical power which is self-contained within the plant and which is not dependent on auxiliary transformer sources of supply. The standby power supply will be capable of supplying the power for those electrical loads which are re-quired for the safe shut-down of the plant.
l4 Amplu capacity will be provided for the condition in which the unit may be involved in a loss-of-coolant accident as well as all condition of shut-down without loss-of-coolant. The system will be designed to meet this requirement 4 with any one of the diesel sets out of service.
852 Safety Desian Basis The standby ac power supply will produce ac power at a voltage and fre-quency compatible with normal bus requirements. The standby power sets will be of such number, and applied to the various plant buses so that the loss of any one of the power units will not prohibit the safe shut-down of the unit. The l total system will satisfy single failure criteria. g I
i j 8.5.3 Description
> 8 5.3.1 General i
The standby ac power supply will consist of three diesel generator sets, each of which will serve one of the three 4160 volt switch groups. The diesel generator sets will be located within Class I enclosures, and each will be seg-regated from the others so that failure of any component of one will not jeopard-ize proper functioning of the others. The load distribution will be arranged so that redundant auxiliary systems will be supplied from alternate switchgroup power supplies so that safe shut-down can be achieved with any two of the three 4 segregated power supplies.
Each diesel generator set will have ample capacity to supply the maximum 4 load required from its associated switchgroup division for safe shut-down of the l11 unit under normal as well as accident conditions, including the condition in which the unit may be involved in a loss of coolant accident. It will also meet this requirement in the event that one of the two remaining divisions of segre-gated power supply may be inoperative.
The diesel generator sets will be sized so that the loss of one diesel will not restrict the availability of adequate power to carry the CSCS power re-
) quirements of the unit for 8000 hours0.0926 days <br />2.222 hours <br />0.0132 weeks <br />0.00304 months <br /> continuously, or that power necessary to L safely shut-down the unit and maintain it is a safe shut-down condition. In ad-dicion, the system will be of sufficient capacity to start all inertial loads it
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ZPS AMENDMENT 11 is expected to drive.
Each diesel generator set will be designed to start automatically and be 4 ready to accept load within 10 sec. It will also be provided with manual start
, control and provision for manual synchronizing with other auxiliary power sup-plies. The starting system will be pneumatic, with two full capacity air '
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2PS i
l AMENDMENT 11 8.6 '
250 VOLT DC POWER SUPPLY AND DISTRIBUTION b 8.6.1 Safety Objective The objective of the 250 volt de system is to supply a reliable source of de power with the capability of supplying larger de loads, such as pumps and large valves for a time adequate to safeguard the plant until power is restored to the ac power buses.
8.6.2 Safety Design Basis The 250 volt system will be arranged so that more than one failure is required before normal plant needs and reactor safety are not served.
8.6.3 Descr'ption
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The 250walls. volt batteries will be located in ventilated battery rooms having concrete The battery chargers will be sized with a capacity suit- l4 able for restoring the battery to full charge under normal (not emergency) load in a time (approximately 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />) commensurate with the recommendations of the battery manufacturer, 4
Two (2) 250 volt battery systems will be provided for the unit. Each system will be complete with a main distribution panel, battery charger, and accessory equipment, and will be physically segregated so that any failure in-volving one system will not jeopardize the other system. The two battery systems will be associated with two of the three segregated divisions of the essential auxiliary power supply in regard to segregation requirements. The battery sys- 4 tem will be as shown in Figure 8.6-1 A spare battery charger will be provided and arranged to replace a normal charger if required. j All essential loads for reactor safety requiring alternate supplies will be arranged so that the normal supply will be from one battery system and the alternate supply from the other. Loads will be divided between the two battery systems. See Paragraph 8.7.2 for location of the 250 volt batteries.
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The 250 volt battery system will also supply other balance of plant loads supplied by separate circuits fed from circuit breakers on the battery distri- 4 bution panels so that these loads may be disconnected if required.
The ampere hour capacity of each battery will be suitable for supplying all essential battery chargers. loads required until ac power is restored for operation of the The ampere hour capacity will in no case be less than that required for a minimum of one hour operation without a battery charger.
1 Table 8.6-1 is a preifminary estimate of loads to be supplied from the 250 volt battery system. It is expected that these loads will be modified and 4
possibly additional loads added as design details are completed 8.6-1
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ZPS AMENDMENT 11 8.7 125 VOLT DC POWER SUPPLY AND DISTRIBUTION 8.7.1 Safety Objective
, The objective of the 125 volt de system is to provide a highly reliable source of control power for relay and control circuits used with the circuit breakers and associated equipment of the auxiliary power system and for other essential protective relay and control equipment provided for the proper and safe operation of the plant, as well as safe reactor shutdown under postulated ]
emergency or accident conditions.
The system will also provide power for operation of essential alarm and '
indication circuits, emergency lighting, starting and control of standby diesel-generator equipment, and similar equipment requiring a source of de control power of maximum reliability.
To ensure the required physical segregation between battery systems, three separate battery rooms will be provided each of which will be physically 37
> located in areas adjacent to the division of switchgear with which it is associated.
8.7.2 Safety DesiRn Basis The 125 volt de system will be designed to meet the single failure cri-terion in which failure of any single component of the system will not jeopard-ize proper operation of equipment required for the safe reactor operation or shut-down of the plant. This will be accomplished by providing a separate bat-tery, distribution panel, and battery charger for each division of the essential 4
auxiliary power supply system. Therefore, any single failure cannot jeopardize more than one division.
8 7.3 Description l
The l'25 volt battery system will be. as shcwn in' Figure 8.7-1. A sepa-rate, battery, distribution panel, and battery charger will be provided for each 1 of the three segregated divisions of the essential auxiliary power supply system. !
In addition, a spare battery charger will be provided and arranged so that it 4 !
may be used to replace any of the normal charges in the event that they may be- l come inoperative. Each battery and its associated equipment will be physically )
segregated from other batteries and their associated equipment so that any fail- )
ure will not affect more than one system.
The battery serving switchgear division 1-3 will be located in a sepa-rate battery room in the 1-3 switchgear area at floor elevation 456'. This i battery room will also contain a 250 volt battery and a 24/48 volt battery all 11 1 of which will be associated with the 1-3 division of segregation. The battery l serving switchgear division 1-4 will be located in a separate battery room in the 1-4 switchgear area at floor elevation 525'. This battery room will also j
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.ZPS
-LJ AMENDMENT 11 contaic a 250 volt battery and a 24/48 volt battery all of which wf11 be associ-ated with the 1-4 division of segregation. The battery serving switchgear divi-sion 1-5 will be located in a separate battery room in the 1-5 switchgear area at elevation 510'. This battery room will contain only the 125 voIt battery 11 serving division 1-5. i tery room.
A suitable ventilation fan will be provided :for each bat- l The battery systerrs serving the three divisions of essential auxiliary !
power will also serve other balance of plant equipment. Such balance of plant equipment will be supplii o by separate circutts fed from circuit breakers on the battery distribution panels, ,o that these loads may be disconnected if required. .4 The battery chargers will have ample capacity to supply noranal loads supplied from the battery system plus charging cepacity suitable for restoring ,
the battery to a fully charged condition within approx;:estely 14 hoeurs following a complete discharge. l l
The ampere hour capacity of each battery will be suitable for supplying all essential loads required until ac power is restored for operatiam of the .
battery chargers. The ampere hour capacity will in no case be less than that !
required for a minimum of one hour operation without a battery charger.
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Table 8.7-1 is a preliminary estimate of loads to be supplied from the ;
125 volt battery system for one unit. It is expected that these loads will be modified and possibly additional loads added as design details are carnpleted.
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l ZPS l AMENDMENT 11 8.6 '* 24 VOLT DC POWER SUPPLY AND DISTRIBUTION 8.8.1 Power Generation Objective The objective of the 24/48 vole de system is to provide a reliable
, source of power for the startup range neutron monitoring systems, the feedva-ter valve control circuits and the process radiation instrument panel.
8.8.2 Power Generation Design Basis The 24/48 volt de system is to be designed to provide 24/48 vole power to selected instrument panels and is to have 100% backup redundancy for each l j
unit.
8.8.3 Description
. The electrical supply for the source range monitor and intermediate range monitor systems will consist of two duplicate 48/24 volt 3 wire, grounded neutral systems. See Figure 8.8-1. Esch system will consist of two 24 volt, 80 ampere hour (8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> rate) batteries in series and connected to a de distri-bution panel. There will be two silicon rectifier type 24 ampere battery chargers on each system, one of which is connected to each of the 24 volt bat-teries. Power supply for the battery chargers will be from a source having backup supply from the standby diesel-generator system. Each 48/24 volt sys-
- =m will be equipped with modervoltage and overvoltage alarms. The battery chargers will be capable of completely recharging the battery in approximately 8.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> while simultaneously supplying the normal continuous load of 15 mm-peres (estimated). See Paragraph 8.7.2 for location of the 24/48 volt batteries. 11 8.8.4 Inspection and Testing s The batteries and other equipment associated with the 48/24 volt de system are easily accessible for inspection and testing. Service and testing will be accomplished on a routine basis in accordance with recomunendations of the manufacturer. Typical inspections would include visual inspections for leaks and corrosion, checking all batteries for voltage, specific gravity and l
level of electrolyte, and simulated load test.
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ZPS AMENDMENT 11 As a result of the foregoing, cable pan divisions are established and cables will be assigned with a division designation which will determine the l routing or enclosures in which the cables will be directed. '
Primary containment isolation actuation cables which go from relay cabi-nets in the auxiliary electric equipment room to motor control centers and from the motor control centers to their respective motor-operated valves will have a similar designation. The cables for inboard isolation valves will be routed in ,
separate pans from the cables associated with the outboard isolation valves.
8.10.2.1.8 Channel l I
The term channel is used in relation to the logic and sensor circuits for i the neutron monitoring system, primary containment isolation system, reactor pro-tection system, etc. to define separation units.
8.10.2.1.9 Primary Containment Penetrations l The electrical properties of all cables which enter the primary contain-ment will be preserved by assigning cables to penetrations containing groups of cables within a limited voltage level. 4 The only meditas voltage cables entering the containment feed power to the reactor circulating ptamp motors. Cables entering the containment for this pur-pose will be routed through penetrations assigned specifically for that function.
Low voltage cables feeding motor-operated valves, ventilation equipment, interlock circuits, etc. will be routed through "lov voltage" penetrations.
Instrument cables such as thermocouple leadwires and RID cables will be routed through separate instrument penetrations.
Redundant cables such as those powering ventilation equipment or sump ptumps, must enter the containment by way of separate electrical penetrations.
Where possible, redundant cables will utilize penetrations arranged horizontally with respect to one another rather than vertically. l11 Cables, feeding redundant safety equipment inside the containment, will enter through separate cable penetrations located in separate sections. (Sec-tions, refers to the 3 areas formed by the natural separation of the containment into thirds by the tendon anchor buttresses.)
Neutron monitoring cables will enter the containment through four (4) 11 penetrations having at least one penetration of a different type between any two neutron-monitoring penetrations. This configuration will guarantee a minimum center distance of 4 f t between any two neutron-monitoring penetrations.
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AMENDMENT 11 l
I 8.10.2.2 Cable Ampacity All power cables to be used in Wm. H. Zimmer Nuclear Power Station will 'I be assigned in accordance with Table 8.10-1. The tables for power cable loading are based on IPCEA Publication No. P-46-426.
8.10.2.3 Cable Pan Loadinst criteria ,
Dee to thermal buildup within cable pans carrying power cables, it has been established that no more than 42 conductors shall be routed in a single pan except as follows* l l
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- a. In applications where cable loading factors are relatively low, more {
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AMENDMENT 11 ;
- i. Those portions of the auxiliary ac power system and auxiliary de power systems which furnish motor and control power to the foregoing systems.
J. Service water syt tem.
- k. Reactor building closed cooling' water system.
4 i Cables associated with the foregoing systems must be installed in a i manner such that no single credible failure will prevent an adequate number of l components from providing the core cooling, isolation, and ventilation re-quirements for the safe shut-down of the unit in the unlikely event that a de- i sign-basis accident occurs. To satisfy this criterion, the cables for the above ;
systems will be separated into two or three groups, as required, which will be '
known as Divisions 1-3,1-4 and 1-5, respectively, in accordance with Table 8.10- 2. The functions of the standby-gas treatment will be divided between Div.
1-3 and Div.1-4, and the location of redundant components on separate levels will .
of itself provide adequate cable separation and isolation to meet the required ' '
criteria. i It is a basic requirement of the Wm. H. Zinner Nuclear Power Station de-sign that, with any single failure, one of the three folleving combinations shall be available for core-coolings ,
- a. 1-HPCS pump,1-LPCS pump,1-RHR pump.
l-
- b. 1-HPCS pump, 2-RHR pumps.
l7
- c. 1-LPCS pump, 3-RHR pumps, auto depressurization system.
In addition to the above requirement, there are certain circumstances ;
when the HPCS system shall be redundant to the RCIC systess; thus those esbles 11!
serving the RCIC auxiliaries must be in a division separate from that which is associated with the High-Pressure Core Spray System.
As a result of the foregoing, three cable divisions are required for separation purposes. By assigning the balance of plant cables to the pan divisions associated with the switch groups from which they originate, the pan system may be held to three divisions.
Neutron monitoring cables will be handled with a separate duct system ;
having a 4-channel segregation and using a rectangular duct. '
Conduit will be used for specific purposes such as primary containment isolation sensors and the logic circuits, reactor protection system sensors, logic and actuation circuits, and some of the area and process radiation ;
circuits.
. I i
8.10-7 1
i i
ZPS AMENDMEhT 11 I i
8.10.3.2 Power Supplies (Control and Power)
Control power cables for the t .ree ESS Divisions will be routed in the control pans of the ESS Division with 5 .!c.h they are associated.
De three 125 volt de systems originate in three separate battery rooms located on three separate floor levels in the auxiliary building. Control power cables from the individual battery distribution panels will be routed in control pans to the control panels and switchgear of their respective, ESS Divisions.
- 120 volt ac control power originates from the two instrument buses, one in Division 1-3 and one in Division 1-4. 120 volt control power froso these buses 1:
will be routed in their respective division control pans. The 120 volt ac con-trol power source for Division 1-5 is from the secondary of the distribution transfonner which is powered from the diesel generator No. 3 4ky bus. Cables from this distribution panel will be routed in control pans for Division 1-5.
%e power cables for ESS motors, heaters, etc., will be routed in the power pans of the ESS division with which they are associated. The power cables will originate at 4ky and 480 volt switchgear located on the same three floor levels as the battery room of the ESS division to which they are assigned.
Power sources which feed the three divisions of switchgear are described in Paragraph 8.4.3.
P l
)
1 i
8.10-9 I
- _ A
ZPS AMENDMENT 11 i
I
[ adequate horizontal separation will exist between cables and compo- 'l7 I nents to prevent comon damage. Adequate horizontal separation be- l tween cables and components of redundant functions, unless separated l by a barrier, is considered to be ten times the largest horizontal l dimension of any combustible component or cable bundling involved in either function within the panel, however, a minimum separation of 6 in. will exist. Cables of non-redundant systems shall not provide l7 a combustible path 'between two redundant systems. Components and cabling of redundant systems will not exist in a vertical configura- l7 tion within the same panel.
If ad dacent panels contain circuits of redundant systems, barriers between the two panels will be installed. Panel ends closed by 17 steel end plates are considered to be acceptable barriers provided that terminal boards are not mounted on the end place and sufficient 4 air space exists between cabling and the end plate.
Floor-to-panel steel barriers will be provided on elevated panels l7 ,
with closed ends. f Penetration of separation barriers shall not he permitted.
Where redundant devices are located in separate compartments of a single cabinet, routing of external cables to the devices will en-l7 sure adequate cable isolation.
i e. Fire Stops Fire stops will be installed in the cable-pan system; wall penetra- l tions to mitigate the migration of fire from one area or pan system to another also will be installed.
Tray penetrations are generally used for convenience where the main-tenance of a pressure differential is not critical. Such openings, however,' should be sealed with fire stops to prevent flame migration from one area to another. Because the wall opening is usually 11 ,
larger than thac taken up by the installed tray and cable, the open- l ing will be partially closed by a berrier designed in accordance with the wall structure itself. The remaining open space around the cables and tray then can be filled with any effective combination of
[ fire-resistant wools and/or compounds. Time ratings for fire-i resistant walls will be maintained in the selection of materials used in penetration of such walls. In the case of power cables, possible effects on cable ampacity will be considered in establish-l ing the lengthwise depth of the fire-resistant fill in the penetre-
- i. tion.
8.10-12
- )
J ZPS AMENDMENT 11 Cable in trays penetrating floors will be completely enclosed for a distance not less than 6 feet above the flcer. The closed tray will contain a tray fire stop obtained by using fire-resistant wool and/or nonflowing fire-resistant compounds.
No vertical path will be lef t unplugged thereby preventing chimney effects. As in the case of wall openings, the time ratings for fire resistant floors will be maintained in the design of the fire stops.
Sleeve penetrations through floors will be used when better control of air, gas, flame or radiation interchange among areas is desired than could be afforded with tray penetrations alone. The most com- gg mon use is beneath control boards or other panels. Af ter cable in-stallation, all voids will be plugged with a nonflowing, fire-resistant material.
Where sleeves are installed in metal plates in the floor, the plates will be coated with a fire-resistant compound.
- f. Smoke Detectors Heat and/or smoke detectors will be installed in confined areas con-taining heavy concentrations of wire and cable such as cable-spread-ing areas, control room, computer room, relay room, and electrical-equipment rooms. The design and configuration of the area will influence the choice and location of the detection equipment.
8.10.3.4 Cables Entering Panels Engineered safeguards systems cables entering Control Room and Auxiliary Electric Equipment Room panels from the cable spreading areas directly between these rooms must meet the following separation requirements which modify those contained in the previous section.
Control Room and Auxiliary Electric Equipment Room panels generally will be designed so that cables for redundant engineered safeguards systems entering '
4 panel sections will be separated by a minimum distance of three f t horizontally, in which case the criteria of the previous section will apply.
If a minimum separation of three ft horizontally between cables of re-dundant divisions cannot be attained, the cables of one of the redundant divi-sions will be installed in conduit from a point inside the panel where the fire barrier between divisions is effective to that point where a minimum separation of three ft is attained.
f Balance of plant cables routed with cables of one of the redundant safe-ty systems divisions will be treated as part of the redundant system cables where l they enter panels and will meet the above mentioned criteria in these areas. {
8.10-12.1 d
I i ZPS i AME ..' MENT 11 i 1
f of more than one of the above groups will not be installed in the same cable tray or conduit. No minimum distances between conduit / cable pans for reactor protection system cables are required, because this system is designed to be fail safe; that is, loss or malfunction of cables or components will tend ,to initiate, rather than to prevent, a reactor scram.
8.10.4.3 Neutron Monitoring Certain neutron monitoring trip signals are used in the logic circuits of the reactor protection system. As a result, the neutron cables from sensors to relay cabinets in the control room must retain separation in accordance with that used for the reactor protection system.
Neutron monitoring cables must be kept separate in four channels. Each channel will contain cables from 1 - SRM sensor, + 2 - IRM sensors + 4 1; of the LPRM sensors. This means that there will be 4 containment penetrations for neutron monitoring. Cables will be routed in conduit or rectangular duct. 11 8.10.4'.3.1 LPRM Inputs LPRM cables in the neutron monitoring system must be created differ- 4 ently from the RPS description because some, but not all, of them are averaged into six APRM outputs, two of which are subdivided to provide eight inputs to the reactor protection system (two to each subchannel). Therefore, the following special requirements will apply:
- a. LPRM cables beneath the reactor vessel and inside the support pedestal will be neither grouped nor separated, because this location is a distribution area for these cables to their re-
'f' spective detectors and because this location provides an adequate l'
' degree of protection from external elements during plant operating and shut-down periods,
- b. LPRM cables will be grouped at the inner end of the pedestal pene-trations through which they pass and will be routed inside the ;
containment in four separate conduits / cable trays to their re-spective containment electrical penetrations. Each of these groups will consist of an assemblage of LPRM cables such that the loss of all cables in a single group will not prevent a high neutron flux scram.
l
- c. The four groups of LPRM cables established in accordance with b) f above will be maintained through the containment electrical pene- j tration and will be installed in four separate cable trays / conduit which will carry them to the power range monitoring cabinet in the control room. *
- d. LPRM cables which are not averaged may be routed in the same 8.10-14 1.
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. '.. WM. H.ZIMMER NUCLEAR POWER STATION i
-.- 4
, _ . g *1,. *i' PaELIMIN ARY SAFETY ANALYSIS REPORT s
- ' ~ ~ ~ , . .* *
..y*"**.m'.......... j F8GURE 8.3 2 l' "
- t.::=;h,' .
PROPERTY ARRANGEMENT \
,_ ' . 7 ^ - _. w
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- 2 N. - ky .' .j,[. . ,
TRANSMISSION LINES l
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A BLANK PAGE
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i l
ZPS' j
?*
%e AMENDMENT I2 i
INSTRUCTIONS FOR UPDATING YOUR PSAR VOLtME 4 .
SECTION 9.0 - RADIOACTIVE WASTE (RADWASTE) SYSTEMS This section has been revised to incorporate new information and provide as answer to an AEC-DRL question.
All changes have been indicated by a vertical line and the amendment number (11) in the right margin of the page.
All pages (text, tables, figures) with changes have also been marked in the upper right corner of the page with "AME5DMENT 11".
Figures that have been altered in any way are indicated by the amendment number in the upper right corner of the figure; note that there arre no other marks that would indicate changes in figure. On the page marked " LIST OF FICIRES", figures that have changed in any way are designated by a vertical line with the amendment number alongside the title of the figure. See example below-FIGURE NIMBER TITLE 2.2-1 Station Site Area Topography f 1:
To update your copy of the Wm. H. Zimmer Nuclear Power Station PSAR, please use the following procedure:
- 1. In Volume 4, SECTION 9.0 - RADI0 ACTIVE WASTE (RADWASTE) SYSTEMS, remove and destroy Table of Contents Page 9.0-iv and replace with amended Page 9.0-iv.
- 2. In Volume 4, SECTION 9.0 - RADIOACTIVE WASTE (RADWASTE) SYSTEMS, remove and destroy the following text pages and replace with the appropriate pages listed below:
REMOVE PAGE REPLACE WITH AMENDED PA2 9.2-3 9.2-3 9.2-5 9.2-5 9.2-7 9.2-7 9.2-8 9.2-8
- 3. In Volume 4, SECTION 9.0 - RADI0 ACTIVE WASTE (RADWASTE) SYSTEMS, remove and destroy Figure 9.2-1 and replace with the amended Figure 9.2-1.
=
ZPS AMENDMENT 11
.i -
- 4. In Volume 4, _SECTION 9.0 - RADIOACTIVE WASTE (RADWASTE) SYSTEMS, behind the red tabbed divider page titled " Amendments to Section 9.0" af ter Page 9.4.3-1 insert Pages 9.4.6-1 through 9.4.6-4.
'l . ~
i l
_ _--_ __ ~
AMENDMENT 11 ZPS SECTION 9.0 - RADIOACTIVE WASTE (RADWASTE) SYSTEMS LIST OF FIGURES F'ICURE NUMBER TITIE 9.2-1 Liquid Radwaste System-4 7 gy Process Diagram 8
i
(.
9.0-iv l
b i si i ..i is
, ,si..i
l ZPS AMENDMENT 11 i
9.2.4.1 High-Puri ty Was tes l
High-purity (low conductivity) liquid wastes are collected in the waste collector tank from the following sources: ,3 Drywell floor drain sump i i Drywell equipment drain sump Reactor building equipment drain sump Radwaste facility equipment drain sump Turbine building equipment drain sump R' wtor cleanup system .
)
RHR System (Suppression pool cicanup)
Cleanup phase separators ,.
Condensate demineralized backvssh Fuel pool system The high-purity wastes are processed by filtration and ion exchange thorugh the waste filter and waste demineralizers. Af ter processing, the waste is pumped to the waste samp e tank where it is sampled and then, if satisfacrery for reuse, transferred to the condensate storage tank as make -up water.
If the analysis of the sample reveals water of high conductivity or high radioactivity concentration, it is returned to the system for additional proces-sing either by the waste filter - demineralized train or by the radwaste eva-porator.
9.2.4.2 Low Purity Wastes Low-purity (high conductivity) liquid wastes are collected in the floor drain co11cetor tank from the following sources:
i Reactor building floor drain sumps l 11 Turbine building floor drain sump These wastes generally have low concentrations of radioactive impurities. Proc-essing consists of filtration, or filtration plus ion exchange, and subsequent 4 transfer to either the floor drain or the waste sample tank for sampling and analysis . The results of the sample analysis indicate whether the waste can be ,
l transferred directly to condensate storage or whether it must undergo further l 8
processing by either ion exchange or evaporation. Normally, low-purity waste is i reclaimed for reuse in the plant. i I
9.2.4.3 Chemical Wastes !
Chemical wastes which are collected in the chemical waste tank are from 4 the following sources:
Shop decontamination solutions I
- 9. 2- 3 7
ZPS AMENDMENT 11 9.2.4.6 _ System Analysis R1 Figure 9.2-1 " Liquid Radwaste System Process Diagram" supplies the fol-1.owing information for cach of the major flow paths of the system:
Normal batches per day
[ ' Volume per batch Normal daily volume '
4 Maximum activity concentration Flow rate On'ly the most contaminated influents to each subsystem are considered; all other influents will have a lower radioactivity concentration than those shown. The following bases were used in determining estimated values for the tabulated quantities:
- 1. Maximum activity concentrations are for fission products only and exclude tritium. The reactor water fission product concentra-tions are calculated for a diffusion mixture of isotopes and are 7 based on continued operation with sufficient fuel cladding defects t
to cause an off-gas activity of 100,000 p ci/see of noble gases at 30 minutes decay.
- 2. Activated corrosion products are assumed to be effectively removed i by the waste and floor drain filters. 7 g : 3. Activity concentration in the primary steam are based on a reactor i
water-to-st' cam decontamination factor (DF) of 10 3
~
- 4. A decontamination factor (DF) of 102 is assumed for the waste cation
- and anion demineralizers; a DF of 10 is assumed for the waste mixed bed demineralized and the evaporator mixed bed demineralized.
4
- 5. A DF of 106 is assumed for the waste evaporators. .
- 6. No activity decay is assumed although the equipment drain and floor drain tanks and sumps are sized to hold one day's accumulation of waste.
L
- 7. Activity concentrations in the condensate demineralized regenerants are based on the regeneration of each condensate demineralized every l11 115 days af ter having processed approximately 495 million gallons of l reactor condensate. !
P
( Table 9.2-1 provides an estimate of the activity released to the environs
) per year from the liquid radwaste system. The following assumptions were used
- 9. 2- 5
'/ .
ZPS AMENDMEhT 11 in determining these estimates:
All normal influents to the equipment drain subsystem are 1.
processed and returned to the reclaimed condensate storage l tanks for reuse in the plant.
- 2. All normal radioactive influents to the floor drain sub-system are processed and returned to the reclaimed conden-sate storage tanks for. reuse in the plant.
- 3. All the normal influents to the laundry drain subsystem are processed and retained for reuse in the laundry facility.
- 4. The majority e' treated radioactive waste released from the plant or . nate as condensate demineralized regener-ants that have been processed through the chemical waste subsystem. Processing of these regenerants on a batch basis occurs about once in every twenty days with a batch consisting of 18,500 gallons of neutralized chemicals.
9.2.4.7 Release of Processed Waste Processed waste is released to the river from one of two discharge tanks on a batch basis. Before release, the waste batch is sampled and the sample is analyzed for gross beta, gamma activity. The resulting activity 7
concentration is used to determine the rate of waste pump-out and is recorded so that an accurate account may be kept of the activity discharged to the river. Processed waste released to the river will not normally be analyzed
, for specific nuclide concentrations. I l 3 i
i All processed liquid waste to be released from the plant is sampled at -
l two points before the discharge procedure is initiated. The actual release to f
the river is from one of two discharge tanks (see the previous paragraph). In l 11 these tanks the final sample is taken and analyzed. Before being routed to the discharge tanks, processed waste is sampled in one of the several sample tanks l provided as a final check point in each of the liquid radwaste subsystems.
j These tanks are as follows:
l Waste Sample Tanks (3 provided) j Floor Drain Sample Tanks (2 provided)
- Laundry Sample Tanks (2 provided)
Evaporator Monitor Tanks (2 provided) l1 A batch of processed waste present in one of these tanks is sampled and l j , the sample is analyzed for gross beta, gansna activity. The results of this !
analysis dictate whether the waste batch, destined for release, is routed to the I l discharge tanks or back to the radwaste system for additional processing. If '
l the activity concentration is low enough to permit release, the processed waste batch is routed to the discharge tanks. If the activity concentration is too 9.2- 7 t
/
,l
! ZPS j AMENDMENT 11 l
1
) high to permit release, the waste batch is routed back to a processing system l and the sample is analyzed for specific nuclide content. This more detailed analysis will indicate if there has been a malfunction in the specific liquid radwaste subsystem or if there is an unusual or unexpected contaminant present. -
Af ter reprocessing, the waste batch is again analyzed for gross beta, gamma activity and its future vetermined.
I Processed waste to be released from the plant is pumped into the service 7' water discharge canal for dilution prior to entering the river. During periods of waste release, the discharge system is continuously sampled and a composite sample for the release period is analyzed for gross beta, gamma and tritium activity. Detailed records are maintained of all radioactive waste discharged to the environs.
If required during the periodic intervals of release of processed liquid ,
waste, the average concentration of unidentified radionuclides in the service water discharge canal shall be allowed to reach 10% of the limits specified in 10CFR20 (1 x 10-8 pct /cc). This limit shall apply for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> not 11l more of ten than once every two weeks. During the infrequent intervals of re-lease of tritiated water for the purpose of primary water tritium dilution, the average concentration of tritium in the service water discharge canal shall be maintained at less than 1% of the limits for tritium specified in 10CFR20 (3 x 10-5 p Ci/c e) .
9.2.5 Evaluation of Power Generatio,p.
Liquid radwastes discharged from the plant have a radioactivity concen-tration which is as low as practicable and is always below 1% of the limits set by 10CFR20. The effluent from the plant to the service water discharge canal is 4 monitored by taking batch samples; records are kept of the concentration levels.
A process monitoring syrtem is provided to indicate high-radiation levels in the discharge canal. Af ter the annunciation of the high-level alarm, the discharge of the liquid radwastes stops automatically.
The minimum service water flow from the plant to the Ohio River is 12,500 gpm. This is a typical vinter flow rate and is maintained by the opera-tion of a single service water pump.
Processed liquid waste vill be recycled for reuse within the plant as long as there is storage volume available in the reclaimed condensate storage tanks. 7 Processed waste to be returned to the reclaimed condensate storage tanks must meet the following requirements:
Conduc tivity <.0 mho/cm at 25'C Chlorides (as C1) 50.05 ppm Boron (as B03 ) 11 0 ppm pH 16-8 at 25'C Radioactivity Concentration 13 x 10-3 9Ci/cc 9.2-8 i
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Pact. MIN ARY S AFETY AN ALYstS REPORT FIGURE 9.2 1 LIQUID RADWASTE SYSTEM PROCESS DIAGRAM 4
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ZPS AMENDMENT 11
- 9. 4. 6-1 (ZPS - February 23, 1971, AEC Question 9.4)
QUESTION It is stated in Paragraph 9.4.6 that gaseous radwastes are discharged through the off-gas vent and that radiation levels of the releases are contin-uously monitored to ensure operation within 1% of the limits of Part 20. Please describe these monitors in more detail. If the radiation monitoring instrumen-tation is designed to measure iodine, please indicate the response time so that the reactor operator may be able to take corrective action in the event of abnormal releases.
Please indicate if the vent gas monitor can detect iodine in the presence of noble gases. Provide the following information relative to trititsa:
- a. Discuss the uncertainties associated with estimating the amounts of tritium generated.
- b. Provide your analysis of the fraction of primary coolant tritium that will be released to the circulating water discharge canal.
- c. Describe the methods to be employed for monitoring tritium '
concentrations prior to discharge from the plant.
ANSWER i The offgas vent radiation monitor presently planned for use on the ,
Wm. H. Zimmer Nuclear Power Station is chown in Figure 7.12-1 and described in
]
Paragraph 7.12.3 of the PSAR. As descrioed thers, the gas sample is drawn from l the vent stream with an isokinetic probe and di:ected to the off gas vont pipe sample rack. In the sample rack, the gas is drawn through a halogen and a par-ticulate filter and a sample chamber containing a 8" x 2" Sodium Iodide Crystal photomultiplier scintillation detector. Routine periodic analysis of the filters ]
provides adequate information to the operator regarding the release rate of i iodine. Noble gas monitoring is accomplished with the scintillation detector.
Expected normal continuous releases from the Wm. H. Zinner Station off-gas sys+
tem are of the order of a few hundred gerocurigper second. The predominant isotopes in the release are Kr885m, Kr and Xe .
- a. The scintillation detector / log count rate meter sensitivity is such that these levels of radioactivity will be readily detected by the system. Operator action to control release above these levels can be brought about as a result of 9.4.6-1 3
7
\
ZPS i
AMENDMENT 11 l
initiation of two high level alarms in the main control room which will be pre-set at appropriate levels. J methods: In a boiling water reactor (BWR) tritium is produced by three principal 1.
Activation of naturally occurring deuterian in the primary coolant. l
- 2. Nuclear fission of UO2 fuel.
3.
Neutron capture reactions with boron used in control rods and poison
" curtains" used with the initial core.
With regards to tritium which may be released from a BWR in liquid or gaseous vastes, the tritium formed in control rods or poison " curtains" which is released is believed to be negligible. Some fission product tritium may also be released. ,
This discussion is limited to the uncertainties associated with esti - !
mating the amounts of tritium generated in a BWR which are available for release.
1 All of the tritium produced by activation of deuterium in the primary {
coolant of a BWR can be calculated using the following equation:
k- "
Fact --E4Y A 1 3.7x104p where:
Ra c t = tritium formation rate by deuterium activation
, (pCL/sec/MWt) -
E =
macroscopic thermal neutron cross section (cm-1) 4 =
thermal neutron flux (neutrons /(cm2 ) (sec) )
V =
coolant volume in core (cm3) l A =
tritium radioactive decay constant (1.78x10~9 sec-1)
P =
reactor power level (HWt)
MWt.
For present day BWR's Ract is calculated to be 1.310.4x10-4 pCi/sec/
The uncertainty indicated is derived from the estimated errors in selecting values for the coolant volume in the core, coolant density in the core, abundance of deuterium in light water (some additional deuterium will be present because of the H (n,y) D reaction), thermal neutron flux, and microscopic cross section for deuterium.
The fraction of tritium produced by fission which may be released to the coolant is much more difficult to estimate. However, since Zircalloy-clad fuel rods are used in BWR's essentially all fission product tritium will remain in l
- 9. 4. 6-2
/
2PS AMEND E ni 11 the fuel rods (1) unless defects are present in the cladding material.
The study made at Dresden 1 in 1968 by the U.S. Public Health Service (2) suggests that essentially all of the tritium released from the plant could be accounted for by the deuterium activation source. For purposes of estimating the release of tritium from defected fuel, we can make the assumption that it is released in a manner similar to the release of noble gases. We can thus use the emperical relationship described as the " diffusion mixture" used for predicting the release of individual noble gas radioisotopes as a function of total noble gas released from the fuel. The equation which describes this relationship is:
Eggg = Ky(I~
where:
Rdif =
release rate of the radioisotope ( Ci/sec) from the fuel y = fission yield fraction A = radioactive decay constant (sec-1)
K =
a constant related to total release rate If the total noble gas activity is 105 UCi/sec after 30 minutes delay we would calculate that about 0.24 UCi/sec of tritium would be released from the fuel. To place this value in perspective in the USPHS study, the obcerved re-lease rate of Kr-85 (which has a half life similar to that of tritium) was 0.06 to 0.4 times that calculated using the " diffusion mixture" reistionship.(2)
This would suggest that the actual tritium release rate might range from 0.015 to 0.10 pCi/sec,
- b. The liquid radwaste system collects, monitors, treats, and returns processed radwastive liquid wastes to the plant for reuse. The system is de-signed to recycle as much processed liquid waste au can be accommodated within the plant balance. Triated liquid waste, cc11ected in the plant equipment and floor drains, will be processed and returned to the reclaimed condensate stor-age tanks. Tritium will also be added to the primary water in the condensate from the recombiners in the off-gas system. Both these factors will tend to increase the equilibrium concentration of tritium in the primary water as well l i
c.s lengthen the period of time necessary to reach this equilibrium c:ncentra-tion. Once equilibrium is reached, a major fraction of the tritium entering the primary coolant will be released to the environs, either as water vapor and gas to the atmosphere, or as liquid waste to the service water discharge (1) Ray, J.W., " Tritium in Power Reactors," Reactor and Fuel-Processing Technology,12 (1), pp 10-26, Winter 1968-1969.
(2) Kahn, B. , e t. al. " Radiological Surveillance Studies at a Boiler Wa ter Nuc. lear Power Reactor," BRH/ DER 70-1, March,19 70.
9.4.6-3
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ZPS A'GNDMENT 11 canal.- Tritium losses due to radioactive decay will be small due to its 12.26 year half lif e. If the primary water tritium inventory builds up to a point where tritiated water must be discharged, the average concentration in the service water canal will be limited to 17. of the 10CFR20 limits (i.e. 3x10-5 pCi/ce).
- c. Since the routine plant release will not be for the purpose of
. primary water tritium dilution there is no sampling for tritium in the dis-charge tanks. However, at infrequent intervals during the plant life-time tritiated waste will have to be released in order to maintain the tritium concentration in the primary water to acceptable levels. All waste released for the purpose of tritium dilution will be sampled for tritium concentration i in the discharge tanks before release. The released waste is sampled for .
tritium under conditions as described in the second paragraph of Page 9.2-8.
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ZPS AMENDMENT 11 INSTRUCTIONS FOR UPDATING YOUP PSAR !
VOLUME 4 SECTION 10.0 - AUXILIARY SYSTEMS This section has been revised with the inclusion of a new subsection, answers to AEC-DRL questions and a revised figure.
All changes have been indicated by a vertical line and the amendment number (11) in the right margin of the page.
All pages (text, tables, figures) with changes have also been anarked in the upper right corner of the page with " AMENDMENT 11".
Figures that have been' altered in any way are indicated by the amendment number in the upper right corner of the figure; note that there are no other marks that would indicate changes in figure. On the page marked " LIST OF FIGURES", figures that have changed in any way are designated by a vertical line with the amendment number alongside the title of the figure. See example below:
FIGURE NUMBER TITLE 2.2-1 Station Site Area Topography i To update your copy of the Wm. H. Zimmer Nuclear Power Station PSAR, f {
please use the following procedure:
j h 1. In Volume 4, SECTION 10.0. AUXILIARY SYSTDfS, remove and destroy '
- f. Table of Contents Pages 10.0-v and 10.0-vi and replace with the l
amended Pages 10.0-v through 10.0-vii.
- 2. In Volume 4, SECTION 10.0 - AUXILIARY SYSTDfS, af ter text Page 10.19-2 insert amended Pages 10.20-1 through 10.20-2. '
- 3. In Volume 4, SECTION 10.0 - AUXILIARY SYSTEMS, remove and destroy Figure 10.6-1 and replace with amended Figure 10.6-1.
- 4. In Volume 4, SECTION 10.0 - AUXILIARY SYSTEMS, behind the red tabbed divider page titled " Amendments to Section 10.0":
- a. In front of Page 10.5-1 insert Page 10.0-1.
- b. Af ter Page 10.19-1 insert Page 10.19-2.
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AMENDMENT 11 TABLE OF CONTENTS, (Continued)
EACE 10.20 Primary Containment Hydrogen, Oxygen and Fission <
Products Sampling 10.20-1 l l
10.20.1 Safety Objective 10.20-1 10.20.2 Safety Design Basis 10.20-1 10.20.3 Description 11 {
10.20-1 10.20.4 Safety Evaluation 10.20-1 10.20.5 Inspection and Testing 10.20-2
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l ZPS AMENDMENT 11 SECTION 10.0 - AUXILIARY SYSTEMS _
LIST OF TABLES ,
TABLE NUMBER TITLE PAGE 10.4-1 Tools and Servicing Equipment 10.4-2 1- 10.5-1 Puel Fool Cooling & Cleanup System 10.5-2 l4 Design Specs.
10.15-1 Typical Locations of Sampling Points 10.15-3 1
10.0-vi 11
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ZPS AMENDMENT 11 SECTION 10.0 - AUXILIARY SYSTDiS LIST OF FIGURES FIGURE EUMER TITLE 10.2-1 Fuel Storage Arrangement i
10.2 2 New Fuel Storage Rack j 10.3 1 Spent Fuel Storage Rack 10.5-1 Fuel Pool Cooling and Cleanup System 10.'-1 6 Reactor Building Closed Cooling Water System l11 )
10.7-1 Turbine Building Closed Cooling Water System 4
10.8-1 Service Water System 10.10-1 Control Room BVAC System i
10.10-2 Station Ventilation System 10.10-3 Diesel-Generator Ventilation System, Service Water Pump House Ventilation System and Switchgear Heat Removal ;
Systen 7 10.10-4 Service Water Pump House Ventilation, Make-Up and Service Water Pug, Rooms Heat Removal Systems !
10.12- 1 Station Service Air System 4
10.12-2 Control and Instrument Air System I
10.0-vii i
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2PS AMENDriENT 11 10.20 PRIMARY CONTAINMENT HYDROGEN. OXYGEN AND FISSION PRODUCTS SAMPLING 10.20.1 Safety objective The objective of the ' hydrogen, oxygen and fission products sampling system is to provide operations personnel with information concerning levels
, of hydrogen, oxygen and fission products in the primary containment following isolation of the primary containment due to a LOCA.
l 10.20.2 Safetv Desinn Rasis
! The hydrogen, oxygen and fission products sampling system will be j designed to initiate primary containment air sampling upon isolation of the i primary containment. 'Ibe system shall initiate an alarm in the main control l room as soon as the predetermined level of hydrogen, oxygen, or fission products f is detected in the primary containment air. l
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! 10.20.3 Description There will be two independent, physically separated systems. Both systems will be placed in service automatically when the primary containment is isolated. In addition, each system shall be capable of manual operation.
Each system will draw air samples from one open ended sample line in the drywell and one open ended sample line in the suppression chamber and pass 11 l
through analyzers for hydrogen, oxygen, and fission products. The sample then !
will pass to a ptanp and be returned to the suppression chamber by another opeo ended semple line. Ecch one of the six sample lines (three per system) will be equipped with one normally open hand operated valve and two spring to close, power to open solenoid valves. Upon isolation of the primary containment, the sunple line solenoid valves will automatically open. The sample lines solenoid J valves will be capable of remote manual as well as automatic operation. Four j sample inlet lines will le equipped with one grab sample tap, each with two l normally closed hand operated valves. {
The power for this system will be from the 120 Volt ac Instrument Power Supply described in Paragraph 8.9.1. ;
10.20.4 Safety Evaluation ;
f Levels of hydrogen, oxygen, and fission products will be recorded in ,
the main control room. When the levels of the hydrogen, oxygen, and/or fission -
products reaches preset levels, an alarm will be initiated in the main control room. Low sample flow which would be indicative of closed sample valve, sample pump not operating, or similar sampling failure, would be annunciated in the main control room. High sample flow, which would be indicative of a sample jI line break, will initiate closure of the solenoid valves in that sample line ~,l and would initiate an alarm in the main control room.
10.20-1 y
l ZPS i; AMENDMENT 11 f 1
1 Pressure control devices shall be located in the sample lines downstrem=
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' of the solenoid valves prior to the analyzers. This assures a constant sample j pressure to the analyzers. i l
10.20 5 Inspection and Testina The systems normally operate upon isolation of the primary containment
)
and infrequently by manual control during normal plant operation. Therefore, 1 each system can be inspected, calibrated, and tested during normal plant opera- l tion by using tht grab sample taps on the sample lines, or by using calibration l g taps on the analyzers, if provided.
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7 ZPS AMENDMENT 11 10.0-1 (ZPS - Februa ry 23, 1971, AEC Question 10.3)
QUESTION Describe the separation criteria that will be used for redundant compo-nents in an essential auxiliary system.
ANSWER
, Redundant compor.ents of an essential auxiliary system will be separated to the extent necessary to assure that in the event of any one of the following events sufficient equipment would remain operational to pemit a safe shut down of the unit.
- 1. Flooding or steam release from equipment failure such as pipe or tank rupture.
- 2. Pipe whip and jet forces resulting from pipe rupture.
- 3. Missiles which may result from equipment failure.
- 4. Fire.
Components will be separated Lc the excent necessary to allow separation of essential electrical cables and conduits in accordance with IFEE 279.
Separation may be in the form of distance, barriers, and restraints or various combinations of these fcrms.
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ZPS AMENDMENT 11 10.19-2 (ZPS - February 23, 1971, AEC Question 7.29)
QUESTION Provide a description of the instrumentation systems included in your design for remote monitoring of post-eccident conditions within the primary containment. Submit an analysis to show that these systens provide appropriate wide range information for the full spectrum of postulated accidents.
ANSWER The answer to this question can be found in Subsection 10.19 " Primary Coccainment Monitoring" of the Wm. H. Zimmer PSAR Amendment 7 and Subsection 10.20 " Primary Containment Hydrogen, Oxygen and Fission Products Sampling" of Amendment 11.
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2PS AME.'DMENT 11 INSTRUCTIONS FOR UPDATING YOUR PSAR VOLUME 4 SECTION 12.0 - STATION STRUCTURES AND SHIELDING i
This section has been revised and answers to AEC questions have been l provided. 1 All changes have been indicated by a vertical line and the amendment '
number (11) in the right .argin of the page.
All pages (text, tables, figures) with changes have also been marked it '
the upper right corner of the page with " AMENDMENT 11".
Figures that have been altered in any way are indicated by the amendmen ;
number in the upper right corner of the figure; note that there are no other taarks that would indicate changes-in figure. On the page marked " LIST OF FIGURES", figures that have changed in any way are designated by a vertical lin with the amendment number alongside the title of the figure. See example below FIGURE NUMBER TITLE 2.2-1 Station Site Area Topography To update your copy of the Wo. H. Zimer Nuclear Power Station PSAR, please use the following procedure:
- 1. In Volume 4, SECTION 12.0 - STATION STRUCTURES AND SHIELDING, remove and destroy the following text pages and replace with the appropriate pages listed below:
REMOVE PAGE REPIACE WITH AMENDED PAGE 12.2-1 12.2-1 ,
12.2-13 12.2-13 12.3-1 12.3-1 12.3-4 12.3-4 12.3-5 12.3-5 12.3-21 12.3-21 12.4-2 12.4-2 and 12.4-2.1 12.4-3 12.4-3
- 2. In Volume 4, SECTION 12.0 - STATION STRUCTURES AND SHIELDING, behind the red tabbed divider page titled " Amendments to Section 12.0", remove and destroy the following pages and replace with the appropriate pages listed below:
W ZPS AMENDMENT 11 RDf0VE PAGE REPIACE WITH AMENDED PAGE 12.2.1.1-2 12.2.1.1-2 12.3.7-1 12.3.7-1 12.4.4-1 12.4.4-1 12.5.1-1 12.5.1-1
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ZPS AMENDMENT 11 12.2 ~DESIGN CRITERIA l 12.2.1 Ceneral Design Data l'
l The design of the structure will conoider the environmental conditions of the site as evaluated in Section 2.0, and all pressure and temperature loads associated with the design basis accident as defined in Section 5.0 and Section
- - 14.0 plus all dead and live loads. The design will consider the possible effects of tornado wind pressures and tornado borne missiles ^on structures housing equip ,
l ment essential to safe shutdown of the reactor. Table 12.2-1 lists the General Design Data.
12.2.1.1 Interconnected Class I and Class II Structures Class II structures rigidly connected to Class I structures will be in-cluded in the dynamic analysis of the class I structures in order to get a true response of the system. Class II structures will not be designed to class I criteria.
The method that will be utilized for the design of interconnecting Class I and Class II structures is explained in Paragraph I.2.6 entitled "Intercon-necting Class I and Class II Structures" of Appendix I.O.
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ZPS AMENDMEhT 11 The base slab will act prie:arily in bending and will be designed by' the working stress method. Sufficient prestressing will be provided in the cylindri-cal walls to climinate membrane tensile stress across the entire section under 3 working stress design loading combinations in order to :aintain the sheer capa- l bility of the concrete section, i I
All Class I structures required for safe shut-down will be fIood protected I to elevation 545.4.
i l No increase in the allowable stresses delineated in the applicable building
! codes will be allowed in the design of Class I structures for earthquake loading.
J- I i a i Major plant structures will be supported on mat foundations. Mat loading and foundation grades are shown in Table 2.5-16 and ultinate leaning capabilities for mat foundations are summarized in Table 2.5-19.
The service water pipe located between the river intake structure and the
( main building complex will be supported on a pile foundation. An evaluation of the ,
j on-site soils under earthquake loading indicates a low margin of safety against i i potential liquification for sandy soils above elevation 450'. The pile foundation i i will resist all downward, uplif t and lateral loads for all conditions of static and dynamic loading.
The method that will be used to analyze the service water piping and support i
, system is explained in Appendix J.0 titled " Analysis of Underground Service Pipes and Supporting Piles".
j 3 11 l l j i ! 12.2.2.4 Structural Instrumentation Program
' The containment instrumentation will consist of the following:
- a. Strain and stress meters embedded in the concrete at critical sections along one 2eridian and at the equipment hatch as shown on Figure 12.2-1.
These instruments will measure concrete strains and stresses in the meridional and hoop direction during post-tensioning and proof testing.
These values will be compared with the design values to insure behavior of the containment is consistent with design assumptions.
- b. Deformation meters will be installed to measure the overall deformations l l of the structures during the pressure test. The readings can be compared with the design values. The deformation meters will be used again during subsequent pressure tests and the measurements compared with the orig-inal ones.
1 12.2.2.5 Seismic Recording System The seismic recording system will consist of a renote Sensing Accelero-graph located some distance from the structures to establish the true ground motion during a seismic disturbance , uneffected by any structure response. The l Sensing Accelerograph would trigger three triaxial accelerometers located inside l the reactor building and connected remotely to a magnetic tape. One of the j sensors will be placed on the reactor, one on the bottom slab and one on the re- '
fueling floor at the top of the containment.
)
12.2-33 I i i
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ZPS AMENDMENT 11 12.3 ANALYTICAL TECHNIQUES i
12.3.1 Seismic Analysis I The seismic design of Class I structures for the Operatfmg Basis Earth-I quake (OBE) will be a horizontal acceleration of 10 percent gravity at the ground surface and a horizontal acceleration of 6.5 percent gravity at rock sur ?
face. Vertical accelerations will be 7 percent gravity at groused surface.
The maximum vertical ground acceleration will be considered as occurrin-simultaneously with the horizontal ground acceleration. The loading due to thi combination will not exceed the values as outlined in Table 12.2-2. The stress resulting from the horizontal excitation will be combined linearly with those resulting from the vertical excitation.
7 The combined stresses resulting from functional loadings and from a I design basis earthquake having a horizontal ground acceleration of .20 g at )
ground surface and .13g at rock surface as specified in Section 2.0, will be I 3
such that a safe shut-down can be achieved. The vertical acceleration will be
.14 g at ground surface.
11 The design will be reviewed to assure that any resulting deflections or j distortions will not prevent the proper functioning of any Class I structure or <
part thereof and will not endanger adjacent structures or componannts. The stresses resulting from the horizontal excitation will be combined linearly with those resulting from the vertical excitation.
3 I
7 The dynamic seismic analyses of a nuclear power station ecumplex can
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generally be divided into two broad categories.. The first is the analysis of I the major buildings and structures which house and/or support the systems which have been defined from a functional standpoint as being Class I. ne second is
. the analyses of the individual systems themselves. The general snethods and
{ philosophies used to accomplish the analyses in these categorf es are presented
, in the Appendix I.0 titled " Procedures for the Seismic Analysis of Critical
- Nuclear Power Plant Structures, Systems and Equipment" a summary of this appendif gl7 -
is provided in Paragraphs 12.3.2 through 12.3.2.15. .
N 12.3.2 Asymetric Shear-Beam and Shear Wall Structures
, For structutes that have shear walls as the lateral load resisting elements, that are asymetric in plan and have aspect ratios, i.e.,, ratio of total height to length in plan, of less than unity, an appropriate solution would be to develop a mathematical model which would allow the representation I l of the actual structure by the analytical model described below. l In this model the floor slabs are considered to be infinitely rigid in their own planes. The rigid body motion of each slab consists of 3 degrees of jl freedom; horizontal translation in two perpendicular directions and rotation Ui i
12.3-1
j.
2PS AMENDMENT 11 12.3.2.6' Horizontal Floor Response Spectra l
A time-history analysis of the building structure seismic model will i I made and the resulting motions of the floor slabs will be used to generate {
- m. : response spectra for the slabs. The resulting peaked spectra will .be smoothed 1 l
and the smoothed curves presented as part of the seismic design criteria for 7 f equipment elements supported on the slabs.
{
l 12.3.2.7 Forcing Functions DE
.s The design criteria for the station provide recomended response spect '
ses for the seismic analysis. The floor slabs response spectra are obtained by averaging the results from four scaled sets of earthquake records.
12.3.2.8 Response Spectra Curves -
Response spectra curves are generated for the OBE and DBE condition frc j the demped structural model and the four sets of scaled earthquake records.
N-S and E-W floor spectra curves are obtained from the concurrent N-S excitation and E-W excitation of the building structure. Then at each spectra g period for a given spectra damping the average response from the four N-S floor spectra and the average response from the four E-W floor spectra are calculated I
and the maximum of the average are plotted. The plotted spectra curves, with their valleys and peak 4, are smoothed by enveloping the peaks to either side of l the peak's period.
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7 12.3.2.9 Vertical Amplification Effects
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The amplification, if any, of the vertical ground motion by the floor sinbs and supporting systems will be determined by investigating specific cases '
in the building structures using dynamic analysis procedures. In the event amplification effects are found to be of a significant nature, either vertical spectra or static coefficients will be calculated to be used as loading for
> 11 the component items for which this loading is required.
12.3.2.10 Analysis by Owner The seismic analvsis of the component Class I systems that are not done by the manufacturer are done by Sargent & Lundy. Analyses performed by Sargent
& Lundy follow the procedures delineated in Appendix I.O.
12.3.2.11 Responsibility The dynanic analysis of the component Class I systems wi11 be acec,m-p11shed by the manufacturers who are responsible for supplying the systems or components to the owner.
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12.3-4
ZPS 33 AMENDMENT 11
- 12. 3.2.12 neview of Vendors' Seismic Analysis
' be The three Sargent & Lundy design disciplines, the Mechanical, Electric and Structrral Engineering Departments have 'the responsibility for the review ed f the vennices' seismic analyses. The Project Engineer of the engineering dis cipline anaer whose direction Class I equipment will be purchased is responsit ,
for insuring that the Class I equipment, component system or structure is de-signed, fericated and erected to withstand the specified seismic loading. Th Mechanical Project Engineer maintains overall responsibility for insuring that j- these criteria are satisfied and that uniform and consistent procedures are etro "" *
[
4 12.3.2.13 manping h degree of damping present in a structure is a function of several variables. Among these variables are the method of construction, the materials -
" of which the structuras are composed and the stress levels attained in the dynamic response of tne structure. A table of damping values thought to be responsible yet suitably conservative are presented in Appendix I.O.
f, 12.3.2.14 sinamic soil Pressures
' d resultant dyanamic lateral force on the walls is obtained by a meth<
developed by lm)nonobe and Okabe. The pressure distribution is assumed to be a of 7 one-half sine curve. Additional dynamic pressure due to water below the water level will be calculated using a modification of the Westerguard theory. The l total dynamic pressure is then obtained by combining the soil and water pressurc
,1 as determined Ly the above procedures. '
i' j- 12.3.2.15 mesign for Relative Movement Effects t
j Class I Components interconnected to various structures on the plant 1 l site will be esigned and analyzed for the effects of the specified seismic
- excitation. Atried electrical ducts and essential piping are examples of the j type of compenent under consideration.
j The p.aperties of both the structures themselves and the material on which they are supported will be used with the specified seismic excitation to se ;
determine the maximum relative displacements.
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ZPS I
AMENDMENT 11 11 drywell head is removed it will be secured to prevent its sliding into the pool.
The capacity of the fuel pool to withstand the effects of tornado gener--
ric al ated missiles has been investigated by the General Electric Compa::y and is pra-ew sented in their report " Tornado Protection for the Spent Fuel Storage Pools" Jis- -
APED-5696 Class I, November,1968.
sible f- 12.3.8.3 Turbine Generated Missiles Tha ,
- at
- Ir. the unlikely event of a failure of the Digital Analoc Control Apparat i
Governor, which is supplied with the turbine to control speed, a empletely re-I dundant overspeed trip system will prevent the turbine from exceeding approximat 111% of rated speed. ;
Redundant overspeed trip signals are being provided for each steam inlet 3
- valve to both the high and low pressure turbines.
als l The main steam inlet lines to the high pressure turbine will have. sepa-rate stop-throttle and control valves, both of which will be tripped closed should the turbine speed exceed approximately lill of rated speed. These valves 11 will be tripped closed by both (a) a mechanical overspeed trip weight and (b) a back up electrical trip. The maximum speed that the turbine would reach is 1207..
7 Steam lines from the moisture separator reheater tc the low pressure tur-
.thod bine inlets will be provided with separate stop and interceptor valves, each of i a
which will be tripped closed at approximately 111% overspeed in the same manner
- er _
as the high pressure turbine inlet valves, ures a Based on the above, a disc failure at a design overspeed of 1207. has been l assumed. Such a failure would only occur in the unlikely event of serious ma-terial deficiency. Analyses have established that a 90 degree disc segment poses the greatest threat as an external missile and that only one such disc would pene-trate the turbine casing at 120% overspeed.
The disc segments could be eje'eted from either low pressure turbine in any radial direction, plus or minus 5' from the original disc plane. The mis-
. site would therefore not come into contact with the primary contalement, control
__ room, or fuel pool.
E Safe plant shut down will be assured by designing the impact area of che following structures to withstand the impact of the disc segment etissile:
- a. Auxiliary bay roof and walls,
- b. Refueling floor of the reactor building.
- c. Reactor building walls below the refueling floor.
12.3.8.4 Re fe rences
- 1. Amirikan, A., Design of Prctective Structures, Bureau of Yards &
Docks No. NAVDOCKS P-51,1950.
- 2. White, R. W. & Botsford, N. B. , Containment of Fragments from a ,
Runaway Reactor, Stanford Research Institute SRIA-113 September 15,196l 12.3-21
ZPS A?ENDMENT 11 will develop 1007. of the specified minimum guaranteed ultimate strength of the tendons listed below: ,
- Minimum guaranteed tensile strength of nonbutton headed wire (f',) 240,000 psi Minimum guaranteed tensile strength of button headed wire 240,000 psi l
Minimum guaranteed yield strength i measured by the 17. total extension 192,000 psi under load method (f, ) (0.8 f',)
12.4.3.3 Tendon Conduits Tendon conduits will be 26-gauge cold rolled steel strip, spiral wound and interlocked. The conduit wall will be reinforced with ribs to withstand the concrete placement loading.
All splices will be sealed to prevent intrusion of cement paste. The tendon sheath splice will be made using a snug fitting coupling approximately one foot long. The joints between the sheath and the coupling will be taped. 3 The radius of curvature will be a minimum of 20 f t.
l 11 12.4.4 Reinforcing Bars Reinforcing bars for concrete will consist of deforned bars meeting the requirements of ASIM A-615 for Class I structures. Reinforcing cos: forming to ASTM A-616 or ASTM A-615 will be used for Class II structores. Placing and splicing of bars will be in accordance with the requirements of ACI 318. Grade classification will be as follows:
Sizes Gra de
- 3 to #5 40 *
- 6 to '#11 60
- 14 to #18 40 or 50 (See Note 1)
Note 1 - #14 or #18 Rars will conform to ASIM A-615, Grade 40 or will be specially ordered reinforcement modified to obtain a guaranteed minimum yield strength cf 50,000 psi and a minimum ultimate strength of 70,000 but maintaining a ductility of 12 per cent elongation in a 2" oecimen.
II Mill test results in accordance with ASTM A-615 will be obtained from the reinforcing steel supplier for each heat of steel to substantiate the re-quired e mposit2on, strength and ductility.
In addition, bend tests and tension tests shall be made using the full 12.4-2
/
ZPS AMD*WlNT 11 section of the bar as rolled, using the largest bar size in each heat. Copics of each test made shall be submitted for approval. All reports shall be docu-mented and submitted even though they might indicate a heat that is inadequate.
The tests shall document yield, ultimate strength, percer:t elongation and chem- '
ical composition. 11 ;
Four full size bars of sufficient length to permit a bend test will be ;
selec.ted at random from the largest size in the heat. The purpose of these speci- j mens is to verify the capability of each heat. .
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i ZPS AMENDMENT 11 i
Sp1f ees of Reinforcement -
Provisions of ACI 318-63 Section 805 will be met for lapped splices for har sizes #11 and smaller. Bar sizes #14 and larger s will be spliced by mechanical connectors. The splice will be designed to develop 3 the specified minimum ultimate strength. Reinforcing spliced with mechanical {
{
connectors shall conform to the following specification in Paragraph 12.4.4.1
( 12.4.4.1 Specification for Splicing Reinforcing Bar Using the Cadweld Process l
1.0 SCOPE This specification covers the mechanical splicing of deformed concrete reinforcing bar for full tensile loading. The tensile strength of the splices f i
shall equal or exceed the minimum ultimate tensile strength for each grade of re -
inforcing steel as specified in the appropriate ASIM standard.
}
2.0 PROCESS l
{
i All splices shall be made by the Cadweld Process using clamping devices, j sleeves, charges, etc. , as specified by the Cadweld Instruction Sheets for "B" and "T" series connections. "C" I series materials shall not be permitted. 7 3.0 QUALIFICATION OF OPERATORS Prior to the production splicing of reinforcing bars, each operator or crew, including the foreman or supervisor for that crew, shall prepare and test two (2) joints for each of the positions to be used in production work. These i splices shall be made and tested in strict accordance with this specification. 7 To qualify, the completed splices shall meet the acceptance standards of Para-graph 7.0 for workmanship, visual quality and minimum tensile strength. A list containing the names of qualified operators and their qualification test results shall be maintained at the job site.
3 4.0 PROCEDURE SPECIFICATION All joints shall be made in accordance with the manufacturer's in-struction sheets "Rebar Instructions for Vertical Column Joints", plus the fol-lowing additional requirements:
4.1 A manufacturer's representative, experienced in Cadweld splicing of rein-forcing bar, shall be present at job site at the outset of the work to demonstrate the equipment and techniques used for making quality splices.
He shall also be present for at least the first 25 production splices to l observe and verify that the equipment is being used correctly and that quality splices are being obtained.
4.2 The splice sleeves, exothermic powder, and graphite molds shall be I stored in a clean dry area with adequate protection from the elements I I
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{- 1 4 -
12.4-3
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l ZPS AMEh*DMENT 11 12.2.1.1-2 (AEC - October l's . 1970. Ques eion 12.3)
QUESTION ll The matter of interconnected Class I and ' Class II strt.ctures is dis-cussed in Paragraph 12.2.1.1. Describe the method of design which will be en-I . ployed to assure that the failure of a Class II structure will not impair the l integrity of the Class I structure or its contents.
ANSWER The method that will be utilized for the design of interconnecting Class j 1 and Class II structures is explained in Paragraph I.2.6 entitled "Intercon-necting Class I and Class II Structures" of Appendix 1.0.
l ZPS A)GNDMENT 11 12.3.7-1 (AEC - October 13, 1970, Question 5.4)
QUESTION In regard to potential turbine missiles, which may be generated in the unlikely event of a turbine disc breakup, and potential missiles which may be generated by a tornado, discuss the protection proposed for the y rimary contain-ment, control room, and fuel pool if postulated trajectories of the missiles could damage these areas. Describe the design provisions to allow safe shutdown of the plant in the event of a postulated turbine disc breakup or tornado strike.
ANSWER See Paragraphs 12.3.7.2 and 12.3.7.3 Amendment 3 and Paragraph 12.3.8.3 yy of the Wm. II. Zimmer PSAR Amendment 11.
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ZPS AMENDMEhT 11 12.4.4 1 (AEC - October 13. 1970, Question 12. 7)
QUESTION Describe the user tests that will be required for the reinforcing steel,
^
specifying types of tests, frequency of tests and criteria for the test results.
Indicate whether full size bar samples will be used in the tests.
ANSWER The answer can be found in Paragraph 12.4.3.3 of the Wm. H. Zinaner PSAR Amendment 11.
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ZPS
{
l AMENDME!!T 1 12.5.1-1,(AEC - October 13. 1970. Question 12.18)
QUESTION i Submit additional details on the containment and internal structural con- 11 crete that define all structural elements, including:
- a. Interior column sizes and layout plan. l
- b. Fully developed elevation of containment to include all penetration 11 locations or areas.
- c. Major dimensions such as wall thicknesses, and diameter of reactor support structure.
ANSWER
- a. The amended Figure 5.2-1 of Section 5.0 (Amendment 3) of the PSAR pre-sents further details of containment layout and dimensions.
- b. See Paragraph 5.2.3.4.1 and Figure 5.2-15 of the Wm. H. Zimmer PSAR Amendment 11,
- c. inter) i i
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l 2PS AMENDMENT 11 1
' INSTRUCTIONS FOR UPDATING YOUR PSAR VOLUME 4 -
SECTION 13.0 - CONDUCT OF OPERATIONS
' This section has been revised to information new information and provide answers to AEC DRL questions, i l I
i All changes have been indicated by a vertical line and the amendment number (11) in the right margin of the page.
All pages (text, tables, figures) with changes have also been marked r
in the upper right corner of the page with " AMENDMENT 11".
Figures that have been altered in any way are indicated by the amend-ment number in the upper right corner of the figure; note that there are no I other marks that would indicate changes in figure. On the page marked " LIST i OF FIGURES", figures that have changed in any way are designated by a vertical p line with the amendment number alongside the title of the figure. See example below: l b FIGURE NUMBER TITLE 2.2-1 i t
Station Site Area Topography 11 '
To update your copy of the Wm. H. Zimmer Nuclear Power Station PSAR,
, please us the following procedure:
- 1. In Volume 4, SECTION 13.0- CONDUCT OF OPERATIONS, remove and destroy Table of Contents Pages 13.0-iv and 13.0-v and replace with Amended Pages 13.0-iv and 13.0-v.
s i 2. In Volume 4, SECTION 13.0- CONDUCT OF OPERATIONS, remove and destroy I
the following pages and replace with the appropriate pages listed j below:
i REMOVE PAGE REPLACE WITH AMENDED PAGE 13.6-2 13.6-2 l 13.6-3 13.6-3 through 13.6-15
- 3. Volume 4, SECTION 13.0- CONDUCT OF OPERATIONS, behind the red tabbed divider page titled " Amendments to Section 13.0":
- a. After Page 13.0-2 insert Page 13.0-3.
- b. After Page 13.3-1 insert Page 13.6.4-1.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___I
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ZPS AMENDMENT 11 TABLE OF CONTENTS. (Continued)
PAGE i 13.6.4.1 General 13.6-2 l 13.6.4.2 General Description of Site and Surrounding Terrain 13.6-3 13.6.*e.3 Emergency Organization 13.6-4
.f 13.6.4.4 Liaison: Local and Civil Authorities and Other Agencies 13.6-7
! 13.6.4.5 Protective Hessures : On and Off Site 13.6-9 l 13.6.4.6 Emergency Treatment; Decontamination; Transportation 13.6-14 13.6.4.7 Training 13.6-15
- 13.6.4.8 Recovery and Re-Entry 13.6-15
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i 13.7 RECORDS 13.7-1 13.7.1 Initial Tests and Operations 13.7-1 f 13.7.2 Normal Operations 13.7-1 j l 13.7.3 Maintenance And Testing 13.7-1 13.7.4 other Records 13.7-2 13.8 OPERATIONAL Fr.?IEW AND AUDITS 13.8-1 13.8.1 Administrative Control 13.8-1 13.8.2 Routine Reviews 13.8-1 13.8.3 Operations Review Committee 13.8-1 13.9 REFUELING OPERATIONS 13.9-1
(
13.9.1 General 13.9-1 l 13.9.2 Training for Refueling Operations 13.9-1 l i
4-13.9.3 Inspection Procedures 13.9-1 e
{ 13.9.4 Emergency Procedures 13.9-1 f 13.9.4 Emergency Procedures 13 9-1 l 1
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13.0-iv
< ZPS AMENDMENT 11 SECTION 13.0 - CONDUCT OF OPERATIONS LIST OF TABLES TABLE NUMBEh TITLE ,PAGE 13,3-1 Pre-operational Nuclear Train- 13.3-2 ing Tentative Course Outline 13.6-1 Letter of Agreement 13.6-10 l between CG&E and the 11 1 Ries Manufacturing Company 11 l
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ZPS 4 AMENDMENT 11 Subsec tion 13.8. The Plant Superintendent or his Assistant, acting within tl technical specifications of the facility, will approve all operating pro-cedures and/or changes.
13.6.3 Maintenance and Testing Specific written procedure manuals also will be developed for planned !
routine maintenance and testing. Selected personnel responsible for this phs of plant service will work with ' General Electric Company engineers in the preparation of these manuals. T. reliminary draf t of the maintenance and test manual will be available noc ::- than two (2) months prior to fuel loading and, like the power operation procedure (Paragraph 13.6.2), many of these methods will have been checked and approved during the programs outline j e
in Subsections 13.4 and 13.5. Non-routine maintenance and testing will be !
performed according to properly approved temporary procedures if the complexi> l of the job (s) warrants it.
The Plant Superintendent or Assistant Superintendent will approve all maintenance and test procedure manuals. Periodic review as previously men-tioned and detailed in Subsection 13.8, will be carried out for this phase of operation as well.
I 13.6.4 Emergency Procedures 13.6.4.1 General This section will outline the emergency procedures necessary to protect lif e, property and equipment. By necessity however, these objectives will be presented in basic format only; the final, detailed version will clearly deffncl the proper course of action to be taken should an accident occur that could ;
affect the safety and health of plant personnel or of the general public.
During operation of a power plant, hazards arise from'such causes as fire, electrical or mechanical failures, and from malfunction of equipment. j Because the prime concerns during the operation of a nuclear station involve ;
radiation and contamination, this plan of emergency procedures will anply {
particularly to these concerns. Moreover, general emergency and acci*ent categories will include fire, explosion, rescue attempts, natural disasters, sickness, injury, and civil disturbance.
The design objectives of the Wm. H. Zimmer Nuclear Power Station are to ensure that the plant will remain safe at all tLmes, that normal radioac tive releases to the environs as governed by the 10CFR20 guidelines are kept as low as; practicable, and that should a major reactor accident occur, that the conse-quences' of the occurrence will be such asito hold any radioactive releases to l wi thin the 10CFR100 guidelines.
13.6-2
E ZPS
- 1 AMENDMENT 11 the Studies of safety have been and will continue to be carried out to analyze and assess the many combinations and interrelationships of the reactc -
and its associated systems. The applicant will cooperate in every way possibl ~
to further the technical advancement in this area. When operating-procedure' changes are necessary or when design or equipment modifications are required, the emergency plans and procedures will be altered accordingly according to ied established practices.
ihese *
..h e Notice will be taken of all construction personnel on site during fuel loading and while the power test program for Unit 1 is progressing. Members.o; the public will pass thru the exclusion zone during power operation. Special consideration will be given to the Moscow Elenuncary School which is about
.ned 2650 f t. from the reactor building.
xity Additional personnel are expected to be assigned to the station for startup duty; at times temporary maintenance personnel will supplement the permanent plant maintenance staff. Temporary personnel of this type will be gi instructed in radiation-protection practices that include emergency plans and '
procedures.
)f 13.6.4.2 General Description of Site and Surrounding Terrain The Wm. H. Zimmer Nuclear Power Station utilizes a light-water-cocied-and-moderated, single cycle, forced-circulation, boiling-water reactor with j
! a maximum licensed power level of 2436 HWt. The station is located on the I
! Ohio side of the Ohio River, 24 mi. southeast of Cincinnati, Ohio and approxi-ect mately 2700 ft. north of Moscow, Ohio. About 69 persons will permanently staff
,e the station when Unit 1 becomes' operational.
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The surrounding countryside in the immediate 5-mile radius of the site is mostly forest cover and land used for farming. Water for local household purposes is primarily ground water; but the cities of Cincinnati, Ohio and Newport and Covington, Kentucky obtain water from the Ohio River about 20 mi. downstream from the site. The closest industrial water intake is the W. C.
Beckjord Power Station, owned and operated by the CG6E Co. located about 8 mi.
NNW of the site.
, Industrial activity within a 5-mile radius of the station is limited to 11 j the Ries Manuf acturing Company and The Black River Mining Company. The Ries
', mufacturing Company, located on the west side of U.S. Route 52, is situated to on about 4.2 acres of land and employs 20 to 25 individuals. This company is about 1,000 f t. from the reactor building at the closest boundary. The Black as River Mining Company operates a limestone quarry just north of Carntown, Kentucky. The main entrance, an 18 f t. x 20 f t. opening with an 18-degree sloped shaf t extending to a vertical depth of 630 f t., is about 2 mi.SSW of the site. At present, 157 men are employed in this deep mining operation, i
13.6-3 9
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2PS AMENDMENT 11 l
Recreational activity in the area consists of pleasure kating on the Ohio River during the seasonal months and public tours of a historic monument marking U.S. Grant's birthplace, in Point Pleasant, Ohio 2 mi. north of the site .
l This monument' is open all year. Sport fishing is limited along this reach of j
.the river.
U. S. Route 52 passes in a NNW be SSE direction approximately 1200 f t.
east of the center line of the reactor building. . General access into the station will be via this route. Moscow-Laurel Road, generally running in a NNE by SSW direction, enters U. S. Route $2 about 400 f t. north of the inter-section of the east-west center-line of the reactor building with U. S. 52.
A county road along the northern property line of the Ries Manufacturing Company, ends at the west boundary of the Ries Company. Kentucky State Route 8, which paralicis the Ohio River on the Kentucky side, is about 4200 f t. from the plant at its closest point. The Chesapeake and Ohio Railroad has a main line also running parallel to the River on the Kentucky side approximately 2700 f t. from the plant.
The Wm. H. Zimmer Station is in an area of low population. The 1970 census indicates that certain regions in the innaediate vicinity of the Zimmer plant have shown a population decrease in the past decade. Figure 2.2-5, Present6 Future Population Distribution 0-5 miles, shows the estimated sector population based on the 1960 census and the available 1970 population pro- )
{
jections. Moreover, the main access roads and highways, as well as all neigh- '
boring towns, villages and boroughs, are indicated in Figure 2.2-5. The village of New Richmond, Ohio with a 1970 census count of 2650 is located 6 mi.
NNW of the site and is the closest city with a population of over 1000 persons within a 10-mile radius.
The exclusion area calculated for the Wm. H. Zimmer Nuclear Station is shown on Figure 2.2-9. This area f alls wholly within the property owned and controlled by The Cincinnati Cas & Electric Company with the exception of that 4.2 acre parcel of land owned by the Ries Manufacturing Company. The exclusion bound sry'is a circle with a 380-meter radius with its center located at the station gaseous release vent.
The low population zone (LPZ) for the Zinuner Station is that area sur-rounding the plant within a radius of 4828 meters (3 miles) in accordance with the definition of LPZ in 10CFR100. Paragraphs 100.3(b) and 100.11(a)2. The total population within this zone is estimated to be about 1800 and is based on projected 1960 census values and on available 1970 counts.
13.6.4.3 Emergency Organization
- a. Cencral - As described in Paragraph 13.2.1.1, the Plant Superin-tendent has the overall responsibility for the operation and maintenance of the W:a. H. Zimmer Nuclear Power Station; this includes the preparation and i
13.6-4
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zes AMENDMENT.11 implementation of the emergency plans and procedures. All operating pro-cedures, including the emergency plansi and procedures, are subject to review by the Operations Review Committee, described in Appendix B.0, Technical Specifications.
l The station will be operated on a continuous, 24-hr-per-day, seven-day per- )
week basis and will employ appropriate operating crews with a shif t supervisor j in charge of each crew. The qualifications of those who operate and those j who supervise the operation of the reactor, are examined by the AEC and l
granted reactor operators' licenses and senior reactor operators' licenses respectively. As part of the examination, a significant knowledge of radia-tion protection, normal operating procedures, and the written emergency plans and procedures must be demonstrated. Figure 13.2-1 shows the normal plant j staffing structure.
I It is believed that emergency situations will be localized and only station property and personnel should be exposed to calculated hazards. How-ever, it is believed further that in some situations, both on-site personnel and, to a lesser degree, members of the general public, could be exposed to radiological hazards that require protective measures, evacuation, or both.
Action levels will be chosen
- conservatively to avoid undue exposure hazards to individuals both on and off the site.
- b. Normal Operating Crew - All abnormal occurrences, accidents and/or emergency situations at the Wm. H. Zimmer Nuclear Power Station will be initial- 11 ly handled by the shif t supervisor on duty. He will be alerted immediately of the situation by his own observation, inter-plant telephone contact, public address system, appropriate alarm, or instrument response, and then he will assume control of the situation. The shif t supervisor's immedia te response to abnormal occurrences, accidents, or emergencies will be prescribeJ by written operating and emergency procedures, but some judgment will be required af ter assessing the individual situation by means of instrument responses, alarms, radiation surveys, personal reports, etc.
The licensed operators on duty in the reactor control room, will respond to an abnormal occurrence, accident, or emergency according to pre-scribed written procedures. These detailed procedures will be developed as described in Paragraph B.6.3, Technical Specifications, and will include specific action levels and delineate the proper course of action to pursue in response to each action level. Other operating personnel will respond as outlined in the procedures or as assigned and directed by the shif t super-visor.
The normal operating crew (shif t personnel) will be trained in basic radiation protection and survey techniques. Assistance in radiological matters will be provided by the chemical radiation protection engineer and technicians under his jurisdiction.
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j ZPS :
l AMENDMENT 11 l l
- c. Emergency Organization - An emergency organization will be designated )
to supplement the normal operating crew for conditions that exceed the capa-bilities of the shif t crew. The emergency procedures will be developed to r provide for those emergencies that may be rastricted to a local plant area, that may affect the entire station, or that could extend beyond the site boundary exclusion area and into the low population zone (LPZ). The procedures requiring emergency organization will consider the generalized situations outlined above.
Because it is not practical to list all the personnel required for each emer-gency situation, an emergency coordinator will be designated to be in complete I charge of the emergency. A duty roster, headed by the plant superintendent, will be established to ensure that responsible persons are always available '
to act as emergency coordinators. Furthermore, the procedures will take into consideration that the emergency staff may not always be on site; therefore, will provide for responsible back up or call out availability will be provided.
The emergency procedure will prescribe the reporting channels required and will be so developed that the shif t supervisor, as the initial party responsive to an emergency, will need to devote a minimum of time to making contact with the energency coordinator.
I Should additional assistance be required during emergency situations, the full resources of The Cincinnati Gas & Electric Company and backup from corporate management is available thru the Electric Production General Office. ,
Details concerning the possible use of personnel and equipment from other 11 l departments within the company will be developed at a later date, but the plan will include those individuals from the General Engineering Departauntt who actively participated in the plant design and layout, as well as other i persons knowledgeable in specialty areas such as communication, purchasing i and stores, transportation, community relations, radiation survey,' decon-l tamination, and medical aid.
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' d. Communication - The usual form of contact and communication with individuals outside the plant will be by commercial telephone which is operated f by the Bell system. This is normally a reliable method of communication'.
The Cincinnati Gas & Electric Company presently is studying several methods of radio communication for use during emergencies. The final design of this emergency radio network will take into consideration the necessity of com-munication with off-site agencies, such as state and local police.
Walkie-talkie transceivers with reliable 5-mile ranges from set to set are assigned to the plant for use during startup and routine testing operation, and those will be available for emergency purposes if the need arises.
Inter-plant communication will be by an internal PX telephone sys tem '
and a public address installation with loud speaker horns' located at strategic points thru out the station.
13.6-6 e .l
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ZPS l
AMENDMENT 11
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- e. Alarms - Area radiation monitors in locations as outlined in Section 7.0 will have alarms that sound in the main control room. Certain monitors will have integral alarms that will sound in the inmediate area as well. Action levels and responses to pre-determined alarm settings will be established and will be part of the documented procedures.
The station liquid and vent discharges will be equipped with radio- )
activity monitors with readout and- alarm transmitted to the control room. The ;
nuclear boiler and associated support systems and processes will be instru-c p
mented so that the Itcensed operators are kept apprised of the status of the system at all times. Action levels and appropriate corrective responses that may be required will be developed as part of the written procedures, f An evacuation alarm system will be installed so that all plant per-t sonnel, regardless of their location, will be alerted to evacuate the station in case of high radiation levels.
j Station fire and first-aid alarms will be completely distinct from the evacuation alarm.
, During those periods when the construction headquarters is maintained and any construction forces are on site, an evacuation warning system will be provided to ensure adequate coverage of the entire construction site.
If unforeseen situations require the installation of additional radia-tion detectors, alarms, or signal horns, they will be installed af ter appraisal 11 of the conditions.
13.6.4.4 Liaison: Local and civil Authorities and Other Agencies General In the unlikely event of radiological health hazard situations in-
, volving the general public, the emergency coordinator will decide on the need and extent for communication and assistance from outside sources and agencies. The emergency coordinator will have the complete complement of i control-room instrument data available as well as other calculational aids and portable instruments necessary to make required decisions. The preliminary planning indicates that these organizations would be coordinated to perform l such tasks as: blocking or restricting traffic flow, c1 caring and preventing boating activity, determining the need and directing the evacuation of off-site personnel, medical assistance, logistics support, and performing radiation j surveys. Up-to-date listings of key persons , responsible for activating j agency support including method of contact, will be part of the emergency procedures; notification during off-duty ' hours will be included.
13.6-7 c
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1 ZPS j l
AMENDMENT 11 I
1
__ Contacts and Arrangements
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l Preliminary contacts have been initiated and informal discussions held with'seveal agencies and local authorities. For the past six years, pro-cedures particularly involving the evacuation and quartering of residents in l the areas of New Richmond, Ohio and Moscow, Ohio during Ohio River flood con- l ditions have been in use or are in development. The Red Cross and local i citizens have participated actively in the development of this procedure; ]
the CG&E Company has taken an active role in carrying out the plan. Law l enforcement agencies, National Guard, Clermont County Department of Health, medical support, the f acilities of the village of New Richmond, the Ohio State Highway Department and the Clermont County Civilian Defense Agency, j among others, have cooperated in implementing the plan. Basically, tasks such l as communication, providing transportation and manpower, medical support, ;
logistics support, police protection, traffic direction, and fire protection are all incorporated in the prx " dure. A recent decision of the New Richmond Flood Comittee through the n c;m. ation of the Red Cross has been to increase j the scope of the flood plan te "clude all types of disasters and to invite several of the local commuaittes to participate and enter into agreement for mutual assistance.
The CG&E Company's preliminary discussions with the Red Cross and this Flood Committee have revolved around the inclusion of the Wm. H. Zimmer Nuclear Power Station in the overall emergency plan for this locale. A1'though this 1 plan is still in the elementary stage, the experience in enforcing the flood I,H 11 plan will benefit the eventual emergency procedure involving the LPZ and off-site agencies for the spectrum of postulated nuclear events for the Wm. H. l Zimmer Station. The CG&E Company will work actively within this group and l will participate in procedure preparation. '
Because the local chapter of the Red Cross is intimately involved in dir. aster planning, and actual procedural implementation, and because juris-tictional territory includes the Kentucky counties of Cambel'1 and Pendleton, the CG6E Company believes that the basic concepts of the emergency plan presently under consideration for the areas on the Ohio side of the Ohio River, will be practicable, through the cooperation of similar Kentucky agencies in those areas of the LPZ on the Kentucky side of the river. Discussions with the appropriate Kentucky agencies will be scheduled soon. I J
Other contacts to be made will include: l i
- a. Law Enforcement Agencies : Ohio State Highway Patrol; Kentucky State Police; Clermont County Sheriff; Campbell County Sheriff; Pendleton 3 County Sheriff; Local Sheriff or Police. I I
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ZPS ,
I luiENDNENT 11 i
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- b. Fire Departments : Washington Township Volunteer Fire Department; New Richmond Fire Department & Life Squad.
- c. U.S. Coast Guard
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- e. AEC Regional Compliance Office i f. Medical Support: Local hospitalg physicians, and necessary medical
! supplies will be coordinated thru the Cincinnati Gas & Electric Company I Medical Director.
} Final arrangements and agreement as well as well as the scope of responsibility of each agency or organization will be included in the FSAR.
The necessary procedures for notification plus the action levels required for protection and/or evacuation will be included in the detailed procedure manuals. Check-off lists will be used as part of the emergency procedures
, to ensure, for example, that all necessary and required contacts with off-site l , organizations or agencies have not been overlooked.
I
.As pointed out in Paragraph 13.6.4.2, the Ries : Manufacturing Company 11 l, is located on the outer boundary of the exclusion area. This company employs approximately 20 to 25 persons and specializes in the design and manufacture 1:
of hospital and surgical supplies. The normal work-day is from 8:00 AM to 4:30 PM, five days per week with selected Saturday and/or Sunday work sched- ;
uled as required. Arrangements with the Ries Company officers, stating that j they would be willing to cooperate in establishing the necessary control pro-cedures in the event of an incident at the Zimmer Station, was reached during a meeting in' the Ries Company offices held on April 6,1971. Table 13.6-1 is a copy of this agreement.
13.6.4.5 Protective Measures : On and Off Site
- a. On-Site - Within The Plant Restricted Area - Generally, radiological situations involving plant personnel will caused by spills, leaks, or changes, in process conditions. The operating and emergency procedures covering these occurrences will outline those operator functions required to bring the ab-normal condition under control or the proper steps to ensure a safe and orderly shutdown of processes or equipment.
The development of permissible doses, levels, concentrations, exposures in restricted areas, personnel monitoring, signs, labels, and signals will be t based on those requirements outlined in 10CFR20, and will be included in the i emergency plans. Action levels will be chosen conservatively so that undue hazards to individuals will not result. In responding to certain action 13.6-9
a-j AMENDMENT 11
, gg TABLE 13.6-1 LETTER OF AGREEMENT BEWEEN CC&E i AND THE RIES MANUFACTURING COMPANY .-
3,
( ;r THE CINCINNATI GAS & ELECTRIC COMPANY
... .... u ...., ,. ....
Cip0CINNATI. CHIO 451
- U$.".".i."?"
May 10, 1971 Ries Manufacturing Company 125 West Central Parkway '
Cincinnati, Ohio 45202 s Attention: Miss A. Schonhoft, President 4
, Gentlemen:
i This letter will confirm the agreement reached at the i meeting held in the offices of your company in the afternoon of April 6, 1971. Present were Miss A. Schonhoft and Mr. and Mrs. i Henry Cushard of the Ries Manufacturing Company (Ries) and Messrs. Wm. V. van Gilse and James R. Schott of The Cincinnati Gas & Electric Company (CG&E) .
As part of the application of this Company for a con-k struction permit for the Wm. H. Zimmer Nuclear Power Station,
, we have filed with the U. S. Atomic Energy Commission a Pre-S1 i liminary Safety Analysis Report (PSAR) that describes the pro-posed nuclear plant in great detail. Part 100 of Title 10 of the Code of Federal Regulations provides that one of the evalua-
- tion factors to be considered in plant siting is an Exclusion i Ar.ea surrounding the nuclear facility. The property plat l presented to you during the meeting shows the location of the
- Wm. H. Zimmer Nuclear Power Station, the Ries property, and the Exclusion Area. As indicated on the plat, the radius of the Exclusion Area is 380 meters (1246.7 feet) and passes through
, your facility. CG&E must demonstrate to the AEC at this stage of the licensing process that CG&E has authority to determine activities within the Exclusion Area, including exclusion or removal of personnel and property from the Area.
f To this end, the parties agree:
- - (i) CG&E will, at its expense, install and maintain on the Ries property such monitoring and
~, communication equipment as is necessary and satisfactory to the AEC to alert persons on the Ries property in the event evacuation of the
. property is necessary; I
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ZPS TABIE 13.6-1 (Continued)
"" Ries Manufacturing Company Page 2 May 10,'1971
)St (ii) Ries agrees that CG&E will have the right to require evacuation of any persons from the Ries prc perty when conditions, in CG&E's judgment, warrant the evacuation of persons from within the Exclusion Area-(iii) CG&E will assume responsibility for losses pertinently sustained, including losses on account of lost production, attributable to any such evacuation; and (iv) to enter into a definitive written agreement respecting the above as and when required by I
the Atomic Energy Commission.
If you agree to the above, please execute three copies of this letter in the space below, retain one copy and i return the other copies to me.
Sincerely, l l
THE CINCINNATI GAS & ELECTRIC COMPANY
/s/ W. H. Dickhoner W. H. Dickhoner Vice President We agree to the foregoing RIES MANUFACTURING COMPANY By /s/ A. Schonhoft President Date May 10, 1971
, 13.6-11
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i ZPS I
AMENDMM 11 ;
levels, it may be necessary to send radiation surveyors into restricted areas ;
to assess the situation. Trained personnel will be utilized for the. '
surveys, and they will be supplied with the proper protective clothing and instrumentation to ensure their personal protection. A complete comple.aent of portable radiation-monitoring instruments will be kept functional and will be calibrated for use during emergencies as well as during routine operation.
Procedures to prevent the spread of contamination will be developed.
i Evacuation of plant personnel from local areas or complete station I evacuation will be accomplished if situations develop that are likely to expose individuals to doses and airborne activity concentrations exceeding those limits specified in 10CFR20 Paragraphs 20.101 and 20.103, Possible '
evacuation routes from areas within the plant are being studied as the plant design is developed. Radiation levels throughout the station also are in pro-cess of development with particular attention given to adequate shielding and alternate access routes so that plant personnel are not exposed unduely during routine operation or when responding to abnormal or emdegency situations.
A primary rally point will be chosen that in the event local or major i plant areas must be evacuated, all personnel can be rapidly assesabled and accounted for. Procedures for personnel accountability and/or rescue to be administered if appropriate,will be part of the emergency manual. A stock of supplies including several portable radiation-monitoring instruments, will be maintained in the primary assembly location.
- b. Within The Exclusion Area - All members of the general public or other persons not subject to occupational radiation exposure but who are located within the exclusion area of the plant will be innediately warned, 11 protected, and/or evacuated, as the emergency situation dictates, if as a result of the emergency, they are likely to be exposed to doses in excess of those limits specified in 10CFR20 Paragraph 20.105.
As described in Section 7.0, Instrumentation & Control, certain systems will be automatically isolated af ter pre-established limits are reached; whereas, other processes will not require automatic isolation because routine batch-and-grab sampling provide the control necessary before release. Ikv-ever, should an emergency arise, alarms and instrument read-out, together with the appropriate action levels, provide the licensed operators with adequate information so that necessary procedures will be initiated.
Protection and/or evacuation of the personnel employed by and head-quartered in the Ries Manufacturing Company will be accomplished according to written agreement upon notification by suitable alarm or verbal announcement.
As discussed in Paragraph 13.6.4.4, preliminary agreement has been made be-tween The Cincinnati Gas & Electric Company and the Ries Manufacturing Company.
13.6-12
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ZPS AMENDMENT 11 1
Evacuation of construction workers who may possibly be on site will t carried out thru mutually agreeable procedures to be developed between CG&E and Kaiser Engineers, Inc. The plans will require that the construction workers evacuate to an assembly point for head count and instruction, l
- c. Outside The Exclusion Area - Within The Low Population Zone - Shou a najor accid.ent situation arise that could result in abnormal radioactive hazards outside the exclusion area and extending into the low population zone (IEZ), the previously mentioned alarms, instrument responses, and action derels will provide the information required so that necessary procedures may be initiated. In addition, a fully instrumented meteorological tower will provide the data necessary to record wind direction and speed and atmospheric stability. The shif t supervisor and/or emergency coordinator will, through the use of the meteorological data, overlays, nomographs, local area maps, population data, radiation surveys and other available tools, be able to evaluate the off-site effects of radioactive releases. Protective actions wil.1 be developed for the general populace within the LPZ to enable evacuation l or Protective procedures to begin if doses in excess of specified limits, to be developed later, are likely to result.
The Cincinnati Cas & Electric Company is fully awcre that the Moscow Elesmentary School is located approximately 2650 feet from the Unit I reactor 11 building . In 1971 the enrollment of the school was about 170 with estimated ages of the children ranging from 5 to 13. Protective action levels will be conservatively chosen so they do not present a radiological hazard to the school children. A written evacuation procedure will be developed in co-operation with the school officials and shall include practice drills similar to those established for use in case of fire or other emergencies.
The Camp Meacham facility located about 2-1/2 miles north of the site on the Kentucky side of the Ohio River, (about midway between the communities of Mentor tsnd California, Kentucky) is used primarily as summer boy's camp. '
Arrangements will be made with the camp officials so suitable protective
, action or evacuation can be accomplished if req. tired.
- d. General - The Control room will be designed for continuous occupancy, the design criteria being 10CFR100 (see Paragraph 12.6.1). In the case of a major accident, equipment or process malfunctions affecting local plant areas, operating personnel working in the control room, will be able to perform any operator function required and keep the emergency co- j ordinator appraised of the situation.
An off-site, secondary emergency headquarters location will be es-tablished at a later date. This headquarters will be stocked with necessary -
emercency equipment and supplies, shall be so situated that it will be un-affected by the emergency and shall be accessible to the emergency personnel that may be called upon to respond to the emergency. Any local, government,
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AMENDMENT 11 i i
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ba or outside agency responding to a call from the emergency coordinator for aid would use this control point as headquarters.
CG&E management will be kept informed of any situation thru the Electric Production Department General Office. The emergency coordinator ould :
will keep the Superintendent or the Manager of Electric Production appraised of the situation and he will notify appropriam members of upper management.
as-All press releases or news. media releases will be coordinated through the CG&E Public Relations Department.
ay The reporting and notification requirements as set forth in the ic Technical Specifications will be part of the procedures and check off lists, 13.6.4.6 Emergency Treatment: Decon tamination: Transpor tation Working agreements will be made with local physicians to provide emer-
.on gency medical service. The CG&E Co. has had relations with physicians whose practice is near existing power plants and have enjoyed exdcllent service and cooperation.
An emergency first aid room and personnel decontamination room will be provided on the mezzanine floor elevation of the service building. These rooms will be stocked with supplies and equipment as directed by the CG&E Co. Medical Department. Station personnel will be trained in the use of such equipment. Sink and shower drains from these rooms will be routed to the 11 liquid radwaste system, r .
There will be several means of emergency transportation. During
" normal" hours, a company passenger vehicle will be available to transport 3
those individuals whose injuries might be considered minor but for whom a : professional medical advice or assistance is required. The procedures
' prescribed for minor injuries,will also c'ontain an outline of the steps to be taken if contamination is detected on the patient and/or in a wound. Other private vehicles could be used in an extreme emergency if it were necessary to transport the patient off-site.
If the case is serious, disabling, or considered ambulatory, the Washington Township Volunteer Fire Department, or in case they are unable to respond, the Clermont County Sheriff, will be contacted for ambulance service.
The necessary decontamination procedures (if applicable) will be implemented and the Company Medical Department will be notified of the accident.
f f Selected station personnel will be assigned to the emergency first l
aid squad and they will respond to cases such as these. Radiation protection personnel will be contacted if the patient is contaminated and they will initiate the proper procedures to assure adequate protection and safety to all concerned.
13.6-14
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AMENDMENT 11 f.
-i sid .
e 13.6.4.7 Training .
All plant personnel will be tsakned in radiation fundamentals and pro-tection. Additionally, those persons requiring AEC operator's licenses will d j be required to demonstrate a knowledge of radiation protection as well as a thorough. knowledge of the emergency' plans and procedures. An emergency squad will be formed. It will consist of men from each operating crew and will be kept up to date on current procedures for rescue, fire fighting and first aid measures. Senior staff personnel will all be thoroughly trained in the above and the radiation and lab technicians will be trained in even more de-
, tail in radiation protection and decontamination procedures.
Periodic drills will be conducted to simuiste emergency conditions and to help eliminate 'any procedural problems.
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0ther personnel from outside agencies such as police, firemen, hos-i pital personnel, nurses, ambulance drivers, etc'. will be trained as required l and as requested by their governing bodies. CG&E personnel will assist in every way possible in this training, including aid in preparing procedures be j and training drills. )
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! 13.6.4.8 Recovery and Re-Entry Il i l It is felt that automatic safety system operation and operator function following detailed procedure manuals will prevent abnormal releases of radio-activity to the environs. However, should a major accident occur, the control room, as previously described, is designed for continuous occupancy. Auto-matic safety systems and operator action will assure a safe and orderly shutdown of the station.
Af te,r the accident situation is under complete control, detailed plans will be developed dependent upon the specific occurrence to return the plant to normal operation. The principal criteria for these recovery plans will be to keep exposure hazards within the 10CFR20 guidelines. The plans will also specify specific re-entry criteria so that relief, maintenance or other individuals can supplement the operating crew. Cooperation with all regulatory bodies and compliance with all applicable codes, licenses and regulations will be strictly adhered to.
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13.6-15
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ZPS AMENDMLTT 11 1
13.0-3 (zps . February 23, 1971, AEC Question 13.7)
QUESTION Indicate to what extent a review has been or will be performed of the plant layout and design to assure that critical equipment necessary for safe operation and/or shut-downis adequately protected from acts of industrial sabotage or civil disorder.
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ANSWER Plant layout, with regard to general access control to the main plant structures, has had extensive review to date. Several schemes acre presently being studied and evaluated. Although our plans are ntt finalfr M , the main thrust of protective requirements will be general access control and monitoring techniques both in and outside the building. We are keeping abreast of surveil-lance methods and equipment currently being uti112ed for these poorposes and if appropriate, will factor these developments into out design.
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f ZPS AMENDMENT 11 13.6.4-1 (ZPS - February 23, 1971, AEC Question 13.6) 1 QUESTION Discuss your plans, as appropriate for a PSAR, for coping with emergen- q cies. The emergency plans provided should be in accordance with the require-l t
ments of the proposed Paragraphs 50.34(a)(10) and 50.34(b)(6)(V), of 10 CFR Part 50 dated May 12, 1970 and the Commission-developed paper entitled " Guide for Emergency Planning" which was referenced in the notice for the new Appendix E published in the Federal Register on December 24, 1970.
ANSWER The plans for coping with emergencies is decribed in Paragraphs 13.6.4 through 13.6.4.8 of the Wm. H. Zimmer Nuclear Power Station PSAR, Amendment 11.
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ZPS AMENDMENT 11 INSTRUCTIONS FOR UPDATING YOUR PSAR VOLUME 4 SECTION 14.0 - PLANT SAFETY ANALYSIS l This sectf.on is amended with the inclusion of an answer to an AEC question.
All changes have been indicated by a vertical line and the amendment number (11) in the right margin of the page.
All pages (text, tables, figures) with changes have also been marked in the upper right corner of the page with " AMENDMENT 11".
Figures that have been altered in any way are indicated by the amend-ment number in the upper right corner of the figure; note that there are no other marks that would indicate changes in figure. On the page marked " LIST
! 0F FIGURES", figures that have changed in any way are designated by a vertical line with the amendment number alongside the title of the figure. See example below:
FIGURE NUMBER TITLE 2.2-1 Station Site Area Topography 11 To update your copy of the Wm. H. Zimmer Nucicar Power Station PSAR, please use the following procedure:
- 1. In Volume 4, SECTION 14.0 - PLANT SAFETY ANALYSIS, behind the red tabbed divider page titled " Amendments to Section 14.0" af ter Page 14.6.3-13 insert Pages 14.9.1-1 and 14.9.1-2.
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! AMENDMENT 11 14.9.1-1 (ZPS - February 23, 1971 AEC Question 14.12)
QUESTION For electrical and mechanical equipment of the reactor protection system and engineered safety features located in the primary containment or elsewhere in the plant, state the design criteria which take into account the potential effects on these components of radiation resulting from both normal operation and accident conditions superimposed on long-term normal operation. Describe the analysis and testing performed to verify compliance with these design criteria.
l ANSWER l
All equipment provided for the Wm. H. Zimmer Nuclear Power Station will be purchased through individual procurement specifications. The specifications
. for electrical and mechanical equipment of the reactor protection system and engineered safety features will require equipment operation af ter having re-ceived the expected life-time dose due to normal operation plus the integrated ,
l dose which the equipment would receive under accident conditions. When l possible, these specifications will list the radiation lev-1s (gamma and/or neutron) during normal operation at all places where equipment is to be in-stalled. The radiation levels will be given in terms of Rad (carbon) per hour maximum for gansna and neutrons./cm2/see maximum for neutron dose rates.
The equipment will then be required to function under the influence of a total integrated dose consistent with the total expected life of the equipment.
Other specifications will list the expected life-time dose under which the equipment is required to function.
In order to specify the dose that protection system and engineered safeguards equipment would receive under accident conditions, a TID 14844 l fission product source model will be assumed (see Paragreph 14.9.1) . This j source term model has the following features:
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l a. Activity in suppression pool: --50 percent of the core halogen in- l l ventory and 1 percent of the core particulate inventory are in-g
. stantaneously released to the suppression pool.
f b. Activity airborne in primary containment: --100 percent of the core noble gas activity, 25 percent of the core halogen activity, and 1 l percent of the core particulate activity are airborne in the i primary containment.
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- c. Activity airborne in the secondary containment: a time-dependent
[ activity level in the secondary containment is calculated
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ZPS AMENDMENT 11 assuming a 0.635%/ day primary containment leak rate, an 80%
building mixing efficiency, :mi a standby gas treatment system flow rate of one (1) reactor b.d! ding air volume per day.
em
- d. Filter activity and heat loading: --the activity and. heat loading oc the filters in the standby gas treabnent system are calculated assuming 100% removal efficiency for both the particulate and charcoal filters.
i Use of the above assumptions will result in conservative values for the doses received under accident conditions. Sufficient design margins will be .
incorporated into the specifications so that the equipment will be capable of I withstanding accident conditions and still perform their intended function.
The materials that manufacturers use to meet the specification are well known and have been used industrially for radiation service in environ-sens similar to those encountered in a BWR. Most of the materials that are used have been tested by industry and published data are available so that testing as a condition of the purchase order is not required. However, materials that have no history of successful radiation experience or testing i are unacceptable for BWR application until their radiation capability is l established by test.
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1 Actual radiation testing of assembled equipment is limited to those devices involved in the sensing of radiation and intended to be installed I in potential high radiation zones. The devices referred to are neutron
, detection and cable assemblies, gamma chambers, scintillation counters and
, area radiation monitor sensor converter units.
i 14.9.1-2
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l ZPS l
AMENDMENT 11 l INSTRUCTIONS FOR UPDATING YOUR PSAR VOLUME 5 All changes have been indicated by a vertical line and the Amendment Number (11) in the right margin of the page.
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- 1. At the beginning of Volume 5 remove and destroy Page 7 and replace l with amended Page 7. Remove and destroy Pages 17 through 19 and re-l place with amended Pages 17 through 19.
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! AMENDMENT 11 VOLUME 4 TABLE OF CONTEWS, (Continued)
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PACE
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l 10.1
SUMMARY
DESCRIPTION 10.1-1 f J0.2 NEW FUEL STORAGE 10.2-1
, IJ.3 SPENT FUEL STORAGE 10.1-1 )
i j 10.4 TOOLS AND SERVICING EQUIPMEW I
10.4-1 I
1o.5 FUEL POOL COOLING AND CLEANUP SYSTEM 10,5-1
{ 10.6 REACTOR BUILDING CLOSED COOLING WATER SYSTEM 10.6-1 i
- 10.7 TURBINE BUILDING CLOSED COOLING WATER SYSTEM i
10.7-1
-l 10.8 SERVICE WATER SYSTEM 10.8-1 l
i 10.9 FIRE PROTECTION SYSTEM 10.9-1 10.10 HEATING, VEffrILATION, AND AIR CONDITIONING SYSTEMS 10.10-1 10.11 MARE-UP WATER TREATMENT SYSTEM 10.11-1 10.12 INSTRUMEP(T AND SERVICE AIR SYSTEMS 10.12-1 10.13 POTABLE AND SANITARY WATER SYSTEM 10.13-1 10.14 EQUIPMENT AND FLOOR DRAINAGE SYSTEMS 10.14-1 10.15 PLANT PROCESS SAMPLING' SYSTEM 10.15-1 10.16 C0mUNICATION SYSTEM 10.16-1
, 10.17 LIGliTING SYSTEM 10.17-1 10.18 HEATING BOILERS 10.18-3 h
10.19 PRIMARY CONTAINMENT MONITORING SYSTEM 10.19-1 .
[ 10.20 PRIMARY CONTAINMENT HYDROGEN, OXYGEN AND FISSION
' )
PRODUCTS SAMPLING 10.20-1 11
' 11.0 STEAM AND POWER CONVERSION SYSTEM TABLE OF CONTENTS 11.0-1
! 11.1
SUMMARY
DESCRIPTION
. 11.1-1
! 7
ZPS AMENDMENT 11 -
l LIST OF ZPS, FEBRUARY 23, 1971 AEC QUESTIONS AEC QUESTION . RENUMBERED VOLLME NUMBER AS QUESTION PAGE OF PSAR i
2.12 2.2.3-2 2.2.3-12 1 2.13 2.3.2.1-2 2.3.2.1-2 1 2.14 2.3.2.1-3 2.3.2.1-3 1
, 2.15 Later Later Later 2.16 2.3.8-1 2,3.8-1 1 4.9 4.7-2 4.7-2 2 4.10 4.7-1 4.7-1 2 4.11 4.9-1 4.9-1 2 4.12 4.0-1 4.0-1 2 5.11 5.2.3.7-1 5.2.3.7-1 2 5.12 10.19-1 10.19-1 2 5.13 Later Later Later S.14 Iater Later Later 5.15 Later Later ,
Later 5.16 Iater Later Later 5.li' 5.2.3-1 5.2.3-1 2 7.1 5.3.3.3.3-1 5.3.3.3.3-1 s 2
7.2 5.3.3.3.2-1 5.3.3.3.2-1 2 7.3 5.3.3.3.3-2 5.3.3.3.3-2 2 7.4 7.1-1 7.1-1 3 7.5 Ieter Later Later 7.6 4.4-1 4.4-1 2 7.7 later Later Later
'l 7.8 7. 2. 3. 6 - 1 7.2.3.6-1 3 7.9 7.2-1 7.2-1 3 7.10 7.2.3.9-1 7.2.3.9-1 3 i l
7.11 Later Later Later
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l AMENDMENT 11 j' LIST OF ZPS, FEBRUARY 23, 1971
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AEC QUESTIONS, (Continued)
AEC QUESTION RENUMBERED VOLUME NUMBER AS QUESTION PACE OF PSAR 7.12 7.2-2 7.2-2 3 7.13 Later Later Later 7.14 Later later La te r
! 7.15 7.4.3-1 7.4.3-1 3 l 7.16 7.5.7.3.3-1 7.5.7.3.3-1 3 7.17 7.8.5-1 7.8.5-1 3 9
7.18 7.5.8-1 7.5.8-1 3 7.19 7.6.3-1 7.6.3-1 3 9
7.20 7.8.5.2-1 7.8.5.2-1 3 7.21 Later Iater I4ter 7.22 Later Later La ter 7.23 Later 14ter Later 7
7.24 Later Later Later l
7.25 D.0-1 D.0-1 5 l 7.26 i 10.10.3-1 10.10.3-1 4 7.27 7.2-3 7.2-5 3 f
7.28 Later La ter Later 7.29 ,
10.19-2 10.19-2 4 7.30 Later Later Later l 7.31 Later Later Later 11
] 8.1 8.3.2.1-1 8.3.2.1-1 4 8.2 8.3.2-1 8.3.2-1 4 8.3 8.3.3-1 8.3.3-1 4 8.4 8.4.3-1 8.4.3-1 4 gg 8.5 8.5.4-1 8.5.4-1 4 8.6 8.4.3-2 8.4.3-2 4 8.7 8.5.3.1-1 8.5.3.1-1 4 8.8 8.0-1 8.0-1 4 18 9 l
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ZPS AMENDMENT 1:
LIST OF ZPS. FEBRUARY 23, 1971 AEC QUESTIONS, (Con tinued)
AEC QUESTION RENUMBERED VOLUME NUMBER AS QUESTION PAGE 11 OF PSA:
8.9- 8.0-2 8.0-2 4 8.10 8.9-1 8.9-1 4' 8.11 8.10-1 8.10-1 4
. 9.1 9. 2.4 -1 9.2.4-1 4 9.2 9.2.4.6-1 9.2.4.6-1 4 11 i 9.3 9.4-1 9.4-1 4 9.4 9.4.6-1 9.4.6-1 4 l 9.5 9.2.4.7-1 9.2.4.7-1 4 9.6 9.4.3-1 9.4.3-1 4 10.1 Later Later Later i 10.2 10.5-1 10.5-1 4 10.3 10.0-1 10.0-1 4 l' 10.4 10.11.2-1 10.11.2-1 4
] 11 12.22 12.6.1-1 12.6.1-1 4 12.23 12.5.6-1 12.5.6-1 4 4 11 i
' 13.1 13.0-1 13.0-1 4 i 13.2 13.2.1.6-1 13.2.1.6-1 4 l
1 1 13.3 13.2.1.2-1 13.2.1.2-1 4 13.4 13.0-2 13.0-2 4 13.5 13.3-1 13.3-1 4 13.6 13.6.4-1 13.6.4-1 4 13.7 13.0-3 13.0-3 4
-1 14.12 14.9.1-1 14.9.1-1 4 14.13 Later Later Later
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19 9
2PS AMEKDMENT 11 i
l INSTRUCTIONS FOR UPDATING YOUR PSAR VOLUME 5 APPENDIX D.O. - QUALITY CONTROL SYSTEM
\
question.
This appendix is amended with the inclusica of answer to an AEC All changes have been indicated by a vertical line and the amendment number (11) in the right margin of the page. {
l All pages (text, tables, figures) with changes have also bee:n usarked in the upper right corner of the page with "AMENDNENT 11".
Figures that have been sitered in any way are indicated by the amend- j ment number in the upper right corner of the figure; note that there are no j other marks that would indicate changes in figure. On the page marked " LIST !
0F FIGURES", figures that have changed in any way are designated by a vertical line with the amendment number along side the title of the figure. See example below: j FIGURE NUMBER TITLE }
2.2-1 Station Site Area Topography 11 To update your copy of the Wm. H. Zinsner Nuclear Power Station PSAR, please use the following procedure:
- 1. In Volume 5, '.MENDIX D.O - QUALITY CONTROL SYSTEM, behind the red tabbed divider page titled " Amendments to Appendix D.0'".
- a. In front of Page D.5-1 insert Page D . 0-1.
- b. Remove and deucroy Pages D.6-12 and D.6-13 and replace with amended Pages D.6-12 and D.6-13.
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ZPS AMENDMENT 11 D.0-1 (ZPS - February 23, 1971 AEC Question 7.25) f QUESTION .
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! Describe the quality ar orance procedures which apply to the equipment
.in the reactor protection system , engineered safety feature circuits, and the emer8ency electric power system. This description should include the quality
- assurance procedures used during equipment fabrication,. shipment, field storage, l field installation, and system component checkout and the records pertaining to j each of these. l f ANSWER j The overall quality assurance program for the Wm. H. Zimmer Nuclear Power Station is detailed in Volume 5, Appendix D.O of the PSAR. This program includes individual programs for General Electric Company, Sargent & Lundy, Kaiser Engi-
- neers, Inc., and the Cincinnati Cas & Electric Company.
i l Appropriate interfaces exist between these organizations and programs; l and the final responsibility for the overall quality assurance program rests
! with Cincinnati Cas & Electric Company.
5
, The procedures covering the quality assurance requirements for components and systems pertaining to fabrication, acceptance testing, shipping, field stor-age, installation and checkouts are described in Appendix D.O of the PSAR. All components and systems are classified as QA & S Classes 1, 2, and 3 with appro-priate requirements for quality assurance procedures and documentation, c
The reactor section system, engineered safety feature circuits and the emergency electric power system are included in QA & S Classes 1, 2, and 3 and therefore the quality assurance procedures and record retention for these systems are described in the PSAR.
D.0-1
ZPS AMENDMEhT 11 HGA TRIP LOGIC REIAYS The HCA relay is used in auxiliary functions as slaves to other relays where basic safety operation of the concerned sys tem is not af f ected by mal- 7 function of this relay. The relay showed acceptable performance in the vertical axis (5g) and the axis parallel to face of mounting panel (4g) but contacts changed states at 1.lg at 32 Hz in axis nonnal to face of relay (and mounting panel) with coil de-energized. With coil energized the contacts remained stable out to 6g's over the test frequency range. These relays are mounted in control panels that are in turn mounted rigidly to the control room floor. Test and/or analyses shows that the floor response spectrum for the control room floor 'and the rigidity of the control room panels are such that the devices will be sub-jected to less than 1.0g acceleration during the Seismic I condition. The HCA relays are therefore considered to be acceptably qualified for the application.
CRD SCRAM DISCHARGE VOLUME TANK SWITCH 7 The switches are float activated mercury magnetic switches that remain stable out to an acceleration level of 0.5g at all test frequencies. These switches are solidly mounted to the volume chamber which in turn is mounted to pipe and supported from the reactor building floor or the drywell wall. The acceleration level at the switch during a design basis earthquake is not ex-pected to reach 0.5g. The failure mode of the switch under vibration, like the reactor level switch, is such as to cause a reactor scram, should the accelera-tion level become 0.5g or greater.
, LEVEL SUITCH (CONDENSATE STORAGE TANK LEVEL)
The level switches are mercury magnetic switches that remain stable out to an acceleration level of 0.5g at ill test frequencies.
The switches are mounted near the bottom of the tank close to ground level where they are expected to see less than lg acceleration during a Design Basis Earthquake, s
The use of the mercury switches in the condensate storage tank level application does not compromise plant safety because the switch at worst condi-fi tion will merely switch HPCS suction from condensate atorage tank to the suppres-sion pool without loss-of-coolant flow.
i f BALANCE OF PLANT I The method of seismic analysis for all components not listed above is
, explained in Appendix I.O. 11 i
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l ZPS AMENDMENT 11 l
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)~ Class I structures, systems, and cceponents for the balance of plant are
- 7' defined in amended Subsection 12.1 and amended Appendix D, Table D.6-2 (Amendment
- 3) of the PSAR. The method of seismic analysis for structures and components not .
.part of the Nuclear Steam Supply Sys tem is described as follows:
A. Piping
[
All essential piping systems are listed in the amended Appendix D (Amendment 3), Table D.6-2. All other piping systems are listed -
in Appendix A.0, Table A.2-1 of the PSAR. The method of analysis is described in Subsection A.3 of Appendix A .0 and further amplified in' Appendix I.O." titled Procedures for the Seismic Analysis of gg ,
Critical Nuclear Power Plant Structures, Systems and Equipment." l 7
B. S truc tures i All Class I structures are listed in amended Subse ion 12.1, Amend- (
ment 3 of the PSAR. The method of seismic analysis performed by Sargent & Lundy for the above structures is described in amended Subsection 12.3, Amendment 3 and in Appendix I.O. titled "Proce-dures for the Seismic Analysis of Critical Nuclear Power Plant 11 Structures, Systems, and Ecuipment". All stresses will satisfy the .
the stress criteria as defined in amended Subsection 12.2, Amendment j 3, of the PSAR. l 1
C. Louipment Class I systems and components within these systems are listed in amended Appendix D 0, Table D.6-2. The stress criteria and method of seismic analysis performed by Sargent & Lundy for the balance of I plant equipment as listed is described in Appendix C.0 and further amplified Apoendix I.O. titled " Procedures for the Seismic Analysis g of Critical Nuclear Power Plant Structures, Systems, and Equipment".
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ZPS i
AMENDMENT 11 INSTRUCTIONS FOR UPDATING YOUR PSAR -
VOLUME 5 i
j if APPENDIX I.O.-PROCEDURES FOR THE SEISMIC ANALYSIS G* CRITICAL NUCLE Q, POWER PLANT STRUCTURES. SYSTEMS AND EQUIPMENT
- i[ This appendix has been amended to incorporate new information.
l$
.g All changes have been indicated by a vertical line and the amendment number (11) in the right margin of the page.
- A p
,Ig All pages (text. tables, figures) with changes have also been marked in the upper right corner of the page with " AMENDMENT 11".
b bW Figures that have been altered in any way are indicated by the h amendment number in the upper right corner of the figure; note that there are no other marks that would indicate changes in figure. On the page I marked " LIST OF FIGURES", figures that have changed in any way are de-l signated by a vertical line with the amendment number alongside the title
. of the figure. See example below:
b FIGURE NUMBER a
TITLE l.
g' 2.2-1 ,
Station Site Area Topography , 1 To update your copy of the Wm. H. Zimmer Nucicar Power Station PSAR,
- l. ' ,
please use the following procedure:
t- 1
[f.
- 1. In Volume 5, APPENDIX I.O. remove and destroy the Table of Contents Pag'e I.0-1 and replace with Amended Page I.0-1.
)
- }
- 2. In Volume 5, APPENDIX I.O. remove and destroy text Pages I.2-8 l and I.10-4 and replace with Amended Pages I.2-8 and I.10-4.
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- I ZPS AMENDMENT 11 APPENDIX I.O - PROCEDURES FCR THE SEISMIC ANALYSIS OF CRITICAL NUCLEAR POWER PLANT STRUC1URES. SYSTEMS AND EQUIPMENT TABLE OF CONTENTS i PAGE I.1 INTRODUCTION I.1-1 I.2 ANALYSIS CF BUILDINGS AND MAJOR SIRUCTURES I.2-1 l
l I.2.1 Dynamic Seismic Analysis of Shear Structures (DSASS) I.2-2 I.2.2 Matrix Analysis of Seismic Stresses (MASS IV) I.2-5 I.2.3 Dynamic Analysis of Structures (DYNAS) I.2-6 j' I.2.4 Applications of Programs I.2-7 i
' l- I.2.5 Damping I.2-8 I.2.6 Interconnecting Class I and Class II Structures I.2-8 l 1:
I.3 DEVELOPMENT T EQUIPMENT DESIGN CRITERIA I.3-1 I.4 ANALYSIS OF COMPONENTS SUPPLIED BY MANUFACTURER I.4-1 I.5 SHELL STRUCTURES I.5-1 I.6 DYNAMIC SOIL PRESSURES I.6-1 9' I.6.1 References I.6-2 I.7 DESIGN OF STRUCTURES AND COMPONENTS FOR RELATIVE MOVEMENT EFFECTS DUE TO SEISMIC EXCITATION I 7-1 I.8 SEISMIC DESIGN OF CLASS I CABLE PAN SUPPORTS I.8-1 I.9 SPRING-SLAB ANALYSIS - I.9-1 j I.9.1 Introduction I.9-1 i I.9.2 Program Description I.9-1 I.10 SEISMIC DESIGN CRITERIA FOR CLASS I SYSTEMS AND EQUIPMENT I.10-1 I.10.1 Introduc tion I.10-1 1.10.1.1 Scope I.10-1 t
I.10.1.2 Definitions I.10-1 I.0-1
ZPS AMENDMENT 11 <
response in each mode is calculated independent of time. Whereas actual moda] l
. responses are nearly independent functions of time and maximum repsonses in different modes do not necessarily occur a. simultaneously. Therefore the maximu possible response of the system is given by the sum of the maximum. nodal
. responses without regard to sign. It has been shown that the probable maximum !
response is ~ about equal to the square root of the sums of the squares of the !
modal maximums.
This root-mean-square criteria is used in combining the modal l responses in the res'ponse spectrum method of analysis except in combining closely-spaced in-phase modes of vibration.
f These closely-spaced in-phase modes of vibration are detected by l computing the model's modal responses and then using both the root-mean-square criteria and the absolute sum criteria in combining modes. In maany location it f a complex model, both criteria give nearly equal results which means a single l
I mode is contributing to the response. If the two criteria give results which i differ by a large amount, more than one mode is contributing to the response.
The modes, modes which contribute are checked, and if they are closely-spaced in-phase they are combined using the absolute sum criteria and treated as a sing 1.
mode when combined with the rest of the modes using the root-mean-square criteria.
, I.2.5 Damping The degree of damping present in a structure is a function of several variables. Among these variables are the method of construction, the materials of which the structures are composed and the stress levels attained in the dynamic response of the structure. In addition, there is little test data avail-9 ab1'e to enable one to determine with a bich degree of certainty the appropriate damping to use in the analysis of structures peculiar to nucisar power plants.
A table of damping values thought to be rmsonable yet suitably conservative are presented in Table I.2-1.
I.2.6 Interconnection Class I and Class II ' Structures _
If a structure designated as Class II is connected to a Class I structure, the class II structure will be included in the dynamic analysis of the Class I structure. This. requirement is necessary in that the two structures represent a coupled system for which the true response can only be determined by considering the system as a whole. The Class II structure will not, however, be designed to the same criteria as the class I structure but it will be shown that the distortion of the Class II area will not affect the integrity of the system as a whole during the earthquake event. j i
i I.2-8
ZPS AMENDMENT 9 l
I.10.3.3 Design Basis Earthquake lal i l
The DBE acceleration response spectra plots presented in this document mun shall be applied in designing Class I equipment located at specified eleva-tions of the power plant. The system shall be subjected concurrently to ex-BE citation along two mutually perpendicular horizontal axes and the vertical j direction.
31 l I.10.4 Compliance Requirements j i
{ I.10.4.1 General ee ,
The design of the system may be verified by either an analytical in method, a test method, or a combination of both. The method used will be that 4
I which is most appropriate for the system under consideration. In the event the 9 ,
computational analyses or tests indicate the fragility level of the system is ]
exceeded, isolation of the systea shall be considered. Where isolation is i pe , utilized, the dynamic analyses shall consider the effects of the isolation his syatem.
I.10.4.2 Seismic Analysis A seismic analysis shall be performed to determine the forces and/or ;
displacements resulting from two postulated earthquakes. The dynamic analysis shall adequately account for the effects of higher modes of vibration on the a
dynamic responses of the system and for the appropriate damping in the system. 1 l
'il- , I.10.4.2.1 Simple Systems g t }
1
- For equipment that can be idealized as a single-degree-of-freedom Fe (SDOF) system, the equipment can be analyzed directly by using the appropriate
]
i response spectrum curve. Af ter determining natural frequency, the proper design acceleration value is read off the spectrum curve. The analysis that follows
- } is a standard static-type of stress analysis. j
?I l
- ! I.10.4.2.2 Complex Systems '
t as 11 For equipment that is too complex to be idealized as a SDOF system, a more sophisticated multi-degree-of-freedom model shall be used for determining dynamic loads with the response spectrum method of analysis.
{ The analytical model of the complex system should consist of an ideal- j l ized discrete mass representation of the actual system. A sufficient number of <
masses shall be used to predict the true response of the system. The discrete lumped masses shall be interconnected by weightless, linear elastic springs representing the stiffness of the actual system. However, if it can be justi-i fied that the equipment's structure can be more realistically modeled by using nonlinear springs or as a continuous system, then this method of analysis will f also be acceptable.
I I.10-4
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7PS 1
d AMENT) MENT 11
.INSTRIXTIONS FOR UFDATING YO!!R PSAR VOLLHE 5, APPENDIX J.0 - ANALYSIS OF UNDERGROUND SER'/ ICE PIPES AND SUPPORTING PILES To update your copy of the Wm. H. Zimmer Nuclear Power Station PSAR, please use the following procedure:
- 1. In Volume 5, APPENDIX J.0 af ter Figure J.8-2 insert the red ' tabbed divider page titled " Amendment to Appendix J.0,";
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