ML20202E054
| ML20202E054 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 01/30/1998 |
| From: | Maine Yankee |
| To: | |
| References | |
| NUDOCS 9802180056 | |
| Download: ML20202E054 (492) | |
Text
._ . . .
Proc. No. 0-06 1
., Rev. No. 4 Page 5 of 6
[ COPY NUMBER: 38 - NRC HEADQUARTERS - DOCUMENT CONTROL DESK ATTACHMENT A MAINE YANKEE CONTROLLED DOCUMENT TRANSMITTAL FORM DOCUMENT: DEFUELED SAFTY ANALYSIS REPORT - REVISION 14 TRANSMITTAL ISSUE DATE: 02-04-98 _ TRANSMITTAL RETURN DATE*: 03-05-98 Enclosed is your controlled copy of the DEFUELED SAFl:TY ANALYSIS REPORT.
To acknowledge receipt, please sign this transmittal and return it to Document Control within twenty working days.
L The above listed document has been Iriserted into the assigned manual / file and all superseded pages have been destroyed.
MANUAUFILE UPDATED BY:
[ Please Print Name DATE:
Signature CAUTION
- Manual Holders who do not sign and return this transmittal wtm to Document Control on or before the required return date may be required to return their controlled manual (s) to Document Control. Reissuance shall require Department Manager or higher management approval.
Please return to: MAINE YANKEE ATOMIC POWER COMPANY Document Control Center P.O. Box 408 /
- [](j(;g j Wiscasset, Maine 04578
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. DEFUELED SAFETY ANALYSIS REPORT
's Defueled Safety Analysis Report (DSAR) is derived from Amendment 13 of the
The DSAR has been prepared, and is intended, as the principallicensing basis document deemed more germane than the FSAR to the Maine Yankee plant's permanently defueled condition. The DSAR is additionally applicable to the Maine
, Yankee plant during decommissioning operations activities. This DSAR is intended to supersede the content of previous revisions of the FSAR.
The format and content of the DSAR generally follow that of the FSAR. Sections and information not pettinent to the operation of a permanently defueled plant have been omitted;information relevant to the decommissioning process has been added where appropriate.
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DSAR Revision 14 l L J
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' f] MYAPC Q LIST OF EFFECTIVE PAGES PAGE . RE REMARKS PAGE REV" REMARKS PAGE REV REMARKS.
4 TABLE OF GONTENIA SECTION 1.0 SECTION 2.0 1 14 11 14 Table of Contents 2-1 14 Table of Contents 11 ' 14 1 11 14 List of Tables 2 - 11 14 Table of Contents ill 14 1 ill 14 List of Figures 2-lll 14 List of Tables iv 14 11 14 2-iv 14 Listof Figures v 14 12 14 21 14 vi 14 13 14 22 14 vii . 14 14 14 2-3 14 vili 14 15 14 2-4 14 lx 14 16 14 2-5 14 Table 2.1,1 17 14 2-6 14 Table 2.1.2 18 14 2-7 14 Figure 2.1 1 19 14 2-8 14 Figure 2.12 1 10 14 Table 1.3.1 29 14 Figure 2.13 1 11 14 Table 1.3.1 2 10 14 Figure 2.14 1 12 14 Table 1.3.1 2 11 14 Figure 2.1-5 1 13 14 Table 1.3.1 2-12 14 Figure 21-6 1 14 14 Table 1.3.1 2 13 14 Figure 2.17 1 15 .14 Table 1.3.1 2 14 14 Figure 2.18 1 16 14 Table 1.3.1 2-15 14 1 17 14 Teble 1.3.1 2 16 14 1 18 14 Figure 1.31 2 17 14 O
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1 19 1 20 14 14
. Figure 1.3-2 Figure 1.3-3 2 18 2 19 14 14
- \ '/ 1-21 14 Figure 1.3-4 2 20 14 1 22 14 2-1. 14 1 23 14 2-22 14 2-23 14 2-24 14 2-25 14 2-26 14 Table 2.2.1 2-27 14 Table 2.2.2 2-28 14 Table 2.2.3 2-29 14 Table 2.2.4 2-30 14 Table 2.2.5 4
2-31 14 Table 2.2.6 2-32 14 Table 2.2.7 2-33 14 Table 2.2.8 2 34 14 Table 2.2.9 2-35 14 Table 2.2.10 2-36 14 Table 2.2.10 -
2 37 14 Figure 2.2-1
'2-?8 14 Figure 2.2-2 2-39 14 Figure 2.2-3 2-40 14 Figure 2.2-4 2-41 14 Figure 2.2-5 2-42 14 Figure "!.2-6 2-43 14 Figure 2.2-7 2-44 14 2-45 14
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MYAPC LIST OF EFFECTIVE PAGES PAGE REV REMARKS PAGE REV REMARKS PAGE REV REMARKS 2-47 -14 SECTION 3.0 3-45 14 Figure 3.2-7 2-48 14 3-46 14 Figure 3.2-8 2-49 14 31 14 Table of Contents 3-47 14 Figure 3.2-9 2 50 14' 3 -11 14 Table of Contents 3-48 14 Figure 3.210 2-51 14 3-lil 14 Table of Contents 3-49 14 Figure 3.211 2-52 14 3-lv 14 List of Tables 3 50 14 Fir = 3.212
- 2-53 14 Table 2.3.1 3-v 14 List of Figures 3-51 '4 FI 93.213 2-54 14 Table 2.3.2 3 vi 14 List of Figures 3 14 2-55 14 Figure 2.3-1 3-1 14 3-53 14 2-56 14 Figure 2.3-2 3-2 14 3-54 14 2 57 14 Figure 2.3-3 3-3 14 3-55 14 2 14 Figure 2.3-4 3-4 14 3-56 14 2-59 14 Figure 2.3 5 3-5 14 3-57 14 2-60 14 Figure 2.3-6 3-6 14 3-58 14 2-61 14 Figure 2.3 7 3 7 -- 14 3-59 14 2 62 14 Figure 2.3-8 3-8 14 3-60 14 63 14 3-9 14 3-61 14 2-64 14 3 10 14 3-62 14 2 65 14 Figure 2.41 3 11 14 3-63 14
- 2-66 14 3-12 '14 3-64 '14 2-67 14 3-13 14 3-65 14 2-68 14 3-14 14 3-66 14 i
2-69 -14 Figure 2.51 - 3-15 14 3-67 14 2 70 14 Figure 2.5-2 3 16 14 3-68 14 g 2-71 14 Figure 2.5-3 3-17 14 3-69 14 2-72 14 Figure 2.5-4 3-18 14 3-70 it 2-73 14 Figure 2.5 5 - 3 19 14 3 71 14 3-20 14 3-72 14 3-21 14 3-73 14 3-22 14 3 74 14 3-23 14 3-75 14 3-24 14 3 76 14 3 25 14 Table 3.1.1 3-77 14 26 14 Figure 3.1 1 3-78 14 3 14 Figure 3.12 3-79 14 3-28 14 3-80 14 3-29 11 3-81 14 Figure 3.3-1 3-30 14 3-82 14 Figure 3.3-2 3-31 14 3-83 14 . Figure 3.3 3 3-32 14- 3-84 14 Figure 3.3-4 3-33 14 3-85 14 Figure 3.3-5 3-34 14 3-86 14 Figure 3.3-6 3-35 14 3-87 14 Figure 3.3-7 3-36 14 3-88 14 3 14 3-89 14 3 38 14 3-90 14-3-39 14 Figure 3.21 3-91 14 3-40 14 Figure 3.2-2 3-92 14 3-41 t4 Figure 3.2 3 3-93 14 Figure 3.3-8 3-42 14 Figure 3.2 4 3-94 14 3-43 14 Figure 3.2 5 3-95 14 pg 3-44 14 Figure 3.2-6 3-96 14 V DSAR Rev.14
N MYAPC LIJT OF EFFECTIVE PAGES PAGE - REV REMARKS PAGE REV REMARKS PAGE REV- REMARKS 3 97 14 3-98 14 SECTION 4.0 SECTION 5.0 3-99 14 Figure 3.3-9 3 100 14 Figure 3.3-10 41 14 . Table of Contents 5-1 14 Table of Contents 3 101 14 Figure 3.311 4 - 11 14 Table of Contents 5 - 11 14 List of Tables 3-102 14 4-lil 14 List of Tables 5-lil 14 List of Figures
< 3-103 14 4-IV 14 List of Figures 51 14 3 104 14 41 14 52 14 3-105 14 4-2 14 53 14 3 106 14 Figure 3.3-12 43 14 5-4 14 3-107 14 4-4 14 55 14 3 108 14 45 14 5-6 14 3 109 14 4-6 14 57 14 3-110 14 47 14 5-8 14 Table 5.2.1 3-111 14 48 14 5-9 14 3-112 14 49 14 5-10 14
- 3 113 14 4 10 14 Table 14 5 11 14 3-114 14 4-11 14 5 12 14 3-115 14 Table 3.3.1 4 12 14 5-13 14 Table 5.3.1 3 116 14 Figure 3.3-13 4 13 14 5 14 14 Table 5.3.2 3 117 14 Figure 3.3-14 4 14 14 5 15 14 Table 5.3.3 3 118 14 Figure 3.315 4 15 14 5-16 -14 3-119 14 Figure 3.3 4 16 14 Table 4.6.1 5 17 14 3-120 14 Figure 3.3-17 4 17 14 Table 4.6.1 5-18 14 3-121 14 Figure 3.318
[ 3 122 14 5 19 14 5-20 14 3 123 14 5-21 14 3 124 14 5-22 14 3-125 14 5-23 14 3 126 14 5 14 3 127 14 Figure 3.3-19 5 25 14 3-128 14 Figure 3.3 20 5-26 14 3-129 14 5-27 14 Table 5.5.1 4- 3 130 14 5-28 14 Table 5.5.2
. 3 131 _14 5-29 14 Table 5.5.2
-3 132 14 5 30 14 Table 5.5.2 3 133 14 Figure 3.3-21 5-31 14 Table 5.5.2 3 134' 14 Figure 3.3-22 5-32 14 Table 5.5.3 3 135 14 - 5-33 14 - Figure 5.5-1 3-136 14 5-34 14- Figure 5.5-2 3-137 14 5-35 14 Figure 5.5-3 3-138 14 Figure 3.3-23 36 14 Figure 5.5-4 3-139 14 5 37 14 3-140- 14 5 38 14 3-141 14 5-39 14 3 142 14 5 14 Table 5.6.1 3-143 14 5 14 Table 5.6.2 3-144 14 Figure 3.3-24 5-42 14 3-145 14 Figure 3.3-25 5-43 14 3-146 14 5A-1 14 3-147 14 5A-2 14 3 148 14 SA-3 14
{ 3-149 14 g SA-4 14 Table 5.A.1 A '
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MYAPC s
LIST OF EFFECTIVE PAGES
-PAGE REV REMARKS PAGE REV REMARKS PAGE REV REMARKS 50 1 14 SECTION 8.0 SECTiON 7.0 58 2- 14 58-3 14 Table 5.B.1 6 14 Table of Contents 71 14 Table of Contents 6 - 11 14 List of Tables 7 il 14- List of Tables 6 - 111 14 List of Figures 7 lii 14 List of Figures 61 14 71 14'-
6-2 14 72 14 6-3 14 73 14 6-4 14 7-4 14 6-5 14 75 14 6-C 14 7-6 14 67 14 Figure 6.1 1 77 14 6-8 14 78 14 6-9 14 79 14 6 10 14 7 10 14 6 11 14 7 11 14 6 12 14 7 12 14 6 13 14 7 13 14 7 14 14 7 15 14
-7 16 14 7 17 14 O
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MYAPC y LIST OF EFFECTIVE PAGES PAGE REV REMARKS PAGE REV- REMARKS APIND.lKA A-47 14 Table A.2.3 A-48 14 Table A.2.3 A-l 14 Table of Contents A-49 14 Table A.2.3 A il 14 List of Tables A-50 14 Table A.2.3 A-lil 14 List of Figures A-51 14 Table A.2.3 A1 14 A-52 14 Table A.2.3
, A2 14 A 53 14 Table A.2.4 A3 14 A 54 14 Table A.2.5 '
A-4 14 A-55 14 Table A.2.6 A5 14 - Table A.1.1 A-56 14 Table A.2.7 A-6 14 Table A.1.2 A 57 - 14 Figure A.21.
A7 14 Table A.1.3 A-58 14 Figure A.2-2 A8 14 Table A.1.4 A-59 14 . Figure A.2 3 A-9 14 Table A.1.5 A-60 14 Figure A.2-4 A-10 14 Table A.1.6 A-61 .14 Figure A.2 5
, - A-11 14 Table A.1.7 A-62 14 Figure A.2-6 A 12 _14 Table A.1.8 A-63 14 ' Figure A.2-7 A 13 14 Tabie A.1.9 A-64 14 Figure A.2-8 A-14 14 Table A.1.10 A-65 ~ 14 Figure A.2-9 A-15 14 Table A.1.11 A-66 14 Figure A.2-10 A 16 14 - Table A.1.12 A-67 14 Figure A.211 A 17 14 Table A.1.13 A-18 14 Table A.1.14
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A-19 A 20 14 14 Table A.1.15 Table A.1.16 A-21 14 Table A.1.17 A-22 14 Figure A.11 A-23 14 Figure A.12 A-24 14 Figure A.13 A-25 14 - Figure A.1-4 A-26 14 Figure A.1-5 A-27 14 Figure A.1-6 A-28 14 Figure A.1-7 A-29 14- Figure A.1-8 A-30. 14 Figure A.1-9 A-31 14 Figure A.110 A-32 14 Figure A.111
- A-33 14 Figure A.112 A-34 14 A-35 14 A-36 14 Table A.2.1 A-37 14 Table A.2.2 A-38 14 Table A.2.2
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A-39 14 Table A.2.2 A-40 14 Table A.2.2 A-41 14 Table A.2.2 A-42 14 Tahle A.2.2 A-43 14 Table A.2.2 A-44 14- Table A.2.2 A-45 14 Table A.2.3 A-46 14 Table A.2.3 i\
DSAR Rev.14
MYAPC U.A Defueled Safety Analysis Report
. TABLE OF CONTENTS 1
1.0 INTRODUr lON AND
SUMMARY
J 1.1 Introduction 1.2 General Plant Descriotion 1.2.1 Design Criteria 3
1.2.2 Fuel Handling System 1.2.3 Fuel Storage System 1.2.4 Radiological Waste Treatment System 1.2.5 Radiological Waste Storage and Disposal System 1.3 Facility Deslan Overview 1.3.1 Plant Site and Population 1.3.2 Structures 1.3.3 Chemical Treatment 1.3.4 Process Instruments 1.3.5 Shielding 1.3.6 Electrical Equipment 1.3.7 System Flow D'igrams 1.4 Identification of Aaents and Contractors 4 1.5 Material Incorocrated by Reference 4
1 i Rev.14 i
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MYAPC sO" Defueled Safety Analysis Reoort TABLE OF CONTENTS (continued) 2.0 SITE CHARACTERISTICS 2.1 Location and Area 2.1.1 Population 2.1.2 Land Use 2.2 Meteorolooy 2.2.1 General 2.2.2 Onsite Meteorological Field Programs 2.2.3 Coastal Fog 2.2.4 Temperature 2.2.5 Precipitation 2.2.6 Tornadoes, Hurricanes, and Severe Thunderstorms i
2.2.7 Environmental Monitoring Program (O 2.3 Hydrology 2.3.1 Surface Hydrology 2.3.2 Oceanographic Features 2.3.3 Probable Maximum Flood 2.3.4 Ice Loading, Oil Spill, and Debris Blockage 2.3.5 Groundwater 2.4 Geolooy 2.5 Seismology 2.5.1 Tectonics 2.5.2 Tsunamis
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k MYAPC
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'd Defueled Safety Analysis Report TABLE OF CONTENTS (continued) 1 3.0 FACillTY DESIGN AND OPERATION 1
4 3.1 Desian Criteria 5 3.1.1 Conformance with 10 CFR 50 Appendix A General Design Criteria 3.1.2 Classification of Structures, Systems, and Components 3.1.2.1 SSCs important to the Defueled Condition 3.1.2.2 Wind, Missile, and Tornado Loadings 3.1.2.3 Water Level (Flood) Design 3.1.2A Seismic Design 3.2 Structures.
3.2.1 Fuel Building 3.2.1.1 General 3.2.1.2 Fuel Unloading Arez 3.2,1.3 New Fuel Storage s
, 3.2.1.4 Spent Fuel Pool 3.2.1.5 Fuel Storage Racks 3.2.2 Storage Buildings 3.2.2.1 Underground RCA Storage Bunker 3.2.2.2 Radiation Controlled Area (RCA) Storage Building 3.2.2.3 LSA Storage Building 3.2.2.4 Ware;.ouse 3.2.2.5 Low Level Waste Storage Building 3.2.3 Service Building 3.2.3.1 Control Room Area 3.2.4 Turbine Building 3.2.5 Primary Auxiliary Building 3.2.6 Service Water Intake Structure 3.2.7 Fire Pump House 3.2.8 Masonry Walls 4
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A MYAPC r i V
Defueled Safen* Analysis Report l TABLE OF CONTENio (continued) 3.0 FACILITY DESIGN AND OPERATION (continued) 3.3 Systems 3.3.1 Fuel Storage 3.3.1.1 Design Basis 3.3.1.2 System Descriptio.1 3.3.1.3 Design Evaluation 3.3.1.4 System Operation 3.3.1.5 Monitoring and Instrumentation 3.3.2 Fuel Handling System 3.3.2.1 Design Basis 3.3.2.2 System Dese.dption 3.3.2.3 Design Evaluation 3.3.2.4 System Operation 3.3.2.5 Monitoring and Instrumentation
[ '3.3.3 Primary Component Cooling Water I
3.3.3.1 Design Basis 3.3.3.2 System Description 3.3.3.3 Design Evaluation 3.3.3.4 System Operation 3.3.3.5 Monitoring and Instrumentation
, 3.3.4 Service Water 3.3.4.1 , Design Basis 3.3.4.2 System Description 3.34.3 Design Evaluation 3.3.4.4 System Operation 3.3.4.5 Monitoring and instrumentation 3.3.5 Ventilation Systems 3.3.5.1 Fuel Building Ventilation System 3.3.5.2 Control Room Ventilation System 3.3.5.3 Auxiliary Ventilation Systems DSAR IV Rev.14
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s MYAPC Defueled Safety Analysis Report
. TABLE OF CONTENTS (continued)
- 3.0 FACILITY DESIGN AND OPERATION (continued) 3.3.6 Auxiliary Systems
- 3.3.6.1 Compressed Air-3.3.6.2 Boric Acid Makeup 3.3.6.3 Primary Water System 3.3.6.4 Primary Vent and Drain System
, 3.3.6.5 Radiological Waste Processing System 3.3.6.6 Fire Protection System 3.3.6.7 Meteorological Instrumentation 3.3.7 Electric Power 3.3.7.1 Ofhite Power System
, 3.3.7.2 Onsite Power System a
3,4 Control of Heaw Loads O
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MYAPC V .Defueled Safety Analysis Report TABLE OF CONTENTS (continued) 4.0 RA91ATION PROTECTION 4.1 Source Terms 4.2 Radiation Protection Prooram 4.2.1 Radiation Protection Design Features 4.2.1.1 Shieldling, Radiation Zoning, and Access Control 4.2.2 Health Physics 4.2.3 Radioactive Materials Safety 4.2.3.1 Materials Safety Program 4.2.3.2 Facilities and Equipment 4.2.3.3 Personnel and Procedures 4.2.3.4 Required Materials 4.2.4 Decommissioning Activities 4.3 ALARA Proaram 4.3.1 Policy Considerations
.] 4.3.2 Design Considerations 4.3.3 Operational Considerations 4.4 Llauld Waste Treatment 4.4.1 Design Bases 4.4.2. System Description 4.4.3 Design Evaluation 4.5 Solid Waste Treatment 4.5.1 Design Bases 4.5.2 System Description 4.5.2.1 Spent Resin Transfer 4.5.2.2 Filter Handling 4.5.2.3 Solid Wastes 4.5.3 Design Evaluation 4,6 Radiation Monitc*ina_ Systems 4.6.1 Design Bases 4.6.2 System Description 4.6.2.1 Process Monitoring 4.6.2.2 Area Monitoring
[~N, 4.6.3 Design Evaluation N_)
DSAR vi Rev.14
r MYAPC V] Defueled Safety Analysis Report TABLE OF CONTENTS (continued) 5.0 ACCIDENT ANALYSIS 5.1 letroduction 5.2 Soent Fuel Criticality Analvses '
5.2.1 Misplaced Assembly 5.2.2 Dropped Assembly 5.2.3 Assembly Adjacent to the Racks 5.2.4 Assembly in the Corner of the Racks 5.2.5 Boron Dilution
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5.3 Fuel Handlina_ Accident 5.4 Soent Fuel Cask Dron 5.5 Soent Fuel Pool Accidents 5.5.1 Loss of Spent Fuel Pool Cooling ,
5.5.1.1 Blocked / improper Cell Flow 4 '
5.5.1.2 Loss of Forced Flow
'k 5.5.1.3 Loss of Heat Sink 5.5.2 Loss of Spent Fuel Pool Inventory 5.6 Low Level Waste Release Incident 5.6.1 Radioactive Wate Gas System Leaks and Failures 5.6.2 Radioactive Liquid Waste System Leaks and Failures 5.6.3 Low Level Waste Storage Building Accident l
APPENDICES APPENDIX 5.A Summary of Parameters Used for Evaluating the Radiological Effects of Accidents APPENDIX 5.8 Atmospheric Transport and Diffusion Characteristics for Accident Analysis
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MYAPC i
Defueled Safety Analysis Report TABLE OF CONTENTS (continued) ;
1
. 6.0 CONDbOT OF OPERATIONS 6.1 Resoonsibility and Organization 6.1.1 Duties and Responsibilities of the Operating Staff Personnel 6.1.2 Duties and Responsibilities of the Support Staff 6.2 Technical Soecifications 6.3 Iraining 6.4 Procedures 6.5 Prograrrs 6.5.1 Emergency Plan 6.5.2 Security Plan 6.5.3 Fire Protection Program 6.5.4 Fitness For Duty (FFD) 6.5.5 Offsite Dose Calculation Manual (ODCM)
O 6.5.6 Quality Assurance Program V 6.5.7 Process Control Program 6.6 Review and Audit 6.6.1 Genera!
6.6.2 Plant Operations Review Committee 6.6.3 Nuclear Safety Audit and Review Committee O
DSAR viii Rev.14
l A MYAPC U Defueled Safety Analysis Report TABLE OF CONTENTS (continued)
- 7.0 DECOMMIE310NING 7.1 Summa. / of Activities 7.1.1 Decommissioning Approach 7,1.2 Storage of Radioactive Warte 7.1.2.1 High Level Waste 7.1.2.2 Low Level Waste 7.1.3 Radiation Exposure Monitoring 7.2 Estimate of Radiological Exoosures 7.2.1 Nuclear Worker 7.2.2 General Public
- 7.2.3 Normal Transportation
- 7.3 CQatrol of Radiation Releases Associated With Decommissioning Events 7.3.1 in Plant Events 7.3.2 Transportation Accidents
-7.4 Non-Radiolooical Safety Evaluation APPENDIX A. Meteorolooical Data Summaries 1
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SECTION
1.0 INTRODUCTION
AND
SUMMARY
TABLE OF CONTENTS Sacron Illla P_aga 1.1 I n trod u ctio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . . .1-1 1.2 Ge neral Plant Descriotion . . ... ... .... . ..... ... ..... . . . .. . . .. .. .. .. . . . . .... .. .. . .. . 1-2
-1.2.1 Design Criteria 1.2.2 Fuel Handling System
!. 1.2.3 Fuel Storage System I- 1.2.4 Radiological Waste Treatment System
!T 1.2.5 Radiological Waste Storage and Disposal System 4 O 1.3 Fa cilitv De sian Overview ... .. ..... . . .... .. .. . .. . .. . . ... .. .. ... ... .. . ..... . ..... . . . .,, 1-6
'%M - 1.3.1 Plant Site and Population 1.3.2 Structures 1.3.3 ' Chemical Treatment 1.3.4 Process instruments 1.3.5 Shiciding 1.3.6 Electrical Equipment 1.3.7 System Flow Diagrams 1.4. Identification of Acents and Contractors ..................................... 1-22 1.5 - Material Incorocrated bv Reference ............................................ 1-23 b 'DSAR 1-1 Rev.14
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MYAPC SECTION 10 LIST OF TABLES Table No. .TJ11a 1.3.1 Maine Yankee Design Characteristics i
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MYAPC SECTION 1.0 LIST OF FIGURES
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- Floure No. Ilt[g:
1.3-1 Site Plan
'1.3-2 . Plot Plan
! 1.3-3 Standard Symbols for Flow D.isgrams 1.3-4 F!ow Diagram Symbols i-1 1
I DSAR 1-iii Rev.14
O MYAPC
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V SECTION
1.0 INTRODUCTION
AND
SUMMARY
1.1 Introduction The Maine Yankee nuclear electrical generation plant is located in the Town of Wiscasset, within the midcoast region of the state of Maine, approximately 27 miles northeast from the city of Portland. The plant, a pressurized light water moderated nuclear reactor, is owned by a consortium of 11 New England electric utilities representing consumers in Maine, New Hampshire, Vermont, Massachusetts, Connecticut, and Rhode Island. During the 24 year operating li'etime of the Maine Yankee plant, more than 125 million megawatt-hours of electrical power were generated and distributed to these consumers.
The Defueled Safety Analysis Report (DSAR) is devc!oped as the principal licensing source document describing the pertinent equipment, structures, systems, operational constraints and practices, accident analyses, and decommissioning activities associated with the existing defueled condition of the Maine Yankee plant. As such, the DSAR is intended to serve in the same role as 7S the Final Safety Analysis Report of Maine Yankee during the periods of power operation between
) 1972 and 1997. The DSAR is applicable throughout the decommissioning of Maine Yankee.
The predecessor to the DSAR, the Final Safety Ana!ysis Report (FSAR), was developed to apply for a license under Section 104(b) of the Atomic Energy Act of 1954, as amended, and the regulations of the Atomic Energy Commission (AEC) as set forth in Title 10 of the Code of Federal Regulations (CFR), to construct and operate the Maine Yankee nuclear electrical generation plant.
The application for this license was submitted by the Maine Yankee Atomic Power Company (MYAPC)in 1967.
The construction permit was issued on October 21,1968. The Operating License (OL) was issuod on September 15,1972. This OL authorized power operation of the facility until October 21,2008.
Additionally, the OL authorized power levels up to 75% rated thermal power. Commercial operation of the plant commenced on December 28,1972. The AEC granted a license to operate the facility at 100% rated thermal and authorized power levels u;; *o and including 2440 megawatts thermal (MWt) in June,1973.
Various amendments to the operating license were subsequently issued and, for a period, authorized the power station to operate at power levels up to and including 2,700 MWt. This power p)
DSAR 1-1 Rev.14
MYAPC level corresponds to a nuclear steam supply system (NSSS) output of 2,715 MWt and a gross electrical output of approximately 931 MWe, On January 3,1996, the successor to the AEC, the Nuclear Regulatory Commission (NRC), restricted power oporatien at the Maine Yankee plant to
- 2440 MWt (90% of tho currently rated licensed power) pending reviews and assessments regarding use of Small Break Loss of Coolant Accident analysis methods utilizing the computer code RELAP5YA.
Maine Yankee ceased power production on December 6,1996 to address cable separation and other issues. By June 20,1997, the reactor had been completely defueled and all spent fuel was resident in the spent fuel pool. _ On August 6,1997, the Maine Yankee Atomic Power Company Board of Directors voted to permanently cease power operations and initiate the decommissioning process. On August 7,1997, Maine Yankee provided written certification to the Nuclear Regulatory Commission, purcuant to 10 CFR 50.82 (a)(1)(l) and(li), that the Maine Yankee Atomic Power Company permanently ceased operations of tne Maine Yankee Atomic Power Station and that all nuclear fuel had been permanently removed from the reactor.
The issuance of this certification fundamentally changes the licensing basis of the Maine Yankee
, _p plant in that the NRC-issued 10 CFR 50 license no longer authorizes operation of the reactor or 1
emplacement or retention of fuelin the reactor vessel. Therefore, as of August 7,1997, only those cond.hons or activities associated with the safe storage of fuel and radiological protection (including waste handling, storage and disposal) are applicable to the defualed Maine Yankee plant.
4 1.2 General Plant Description 1.2.1 Design Criteria in the permanently defueled condition, principal structures and equipment are required to safely store new and spent nuclear fuel, prevant the uncontrolled release of radioactb e effluents, and provide shielding to maintain occupational radiological exposures As Low As Reasonably Achievable (ALARA). Additionally, certain structures and equipment function to confine an 2ncontrolled release of radioactive materials and effluents and prevent or mitigate accidents and their consequences. Structures and equipment required to safely store and handle new and spMt fuel are designed, fabricated and erected in accordance with applicable codes, standards, and regulations. Likewise, liquid and gaseous radioactive waste treatment, storage and disposal systems are designed, fabricated and erected in accordance with applicable codes, standards, and
'( D DSAR 1-2 av.14 l
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'f]i MYAPC V
regulations. The systems and components of the facility are designed to enable the facility to withstand the traditionally defined external forces that may be imposed by natural phenomena, without ioss of the copability to protect the public. The protected area is enclosed within a security fence with all access controlled through a guardhouse, in addition, intrusion barriers prevent breech of the fence by vehicles.
In the permanently defueled condition, two power sources are available to provide for spent fuel cooling / makeup, fire protection, security and emergency preparedness functions.
1.2.2 Fuel Handling System Fuel handling and storage facilities are provided for the safe handling, storage and shipment of both new and spent fuel and are designed to proclade accidental criticality. The fuel building houses a new-fuel unloading area, a new-fuel storage room, a spent fuel pool and the necessary cranes required for the handling of the fuel assemblies. The equipment decontamination area and the spent fuel pool support systems are also located in the fuel building.
(3 The spent fuel movable platform and hoist is a traveling bridge which spans the spent fuel pool and moves on re'Is over useable spent fuel storage locations. The hoist hook suppcrts handling tools for grappling and moving fuel assemblies at a safe depth below the operational water level. The design of the fuel handling system precludes the exposure of operating personnel to abnormal radiation fielos as the result of operational transients in the spent fuel pool. Likewise, the design and operation of the fuel handling system prevents the accidental exposure of a spent fuel assembly to a position above or near the water surface.
New or unirradiated fuel assemblies have been removed from the Maine Yankee site and therefore do not require a handling system. The remaining 6 unirradiated fuel rods stored in the new fuel storage room may be handled safely by hand or with the existing fuel building crane.
1.2.3 Fuel Storage System The spent fuel pool,37 feet wide by 41 feet long by 38 feet deep, is located in the south end of the fuel building adjacent to the reactor containment. The pool is cor.structed of reinfc rced concrete with a wall and floor lining of 1/4-inch thick stainless steel. The walls of the spent fuel pool are approximately 6 feet thick. The floor of the spent fuel pool rests on bedrock. The fuel transfer tube, which is 36 inches in diameter, connects the containment refueling cavity with the fuel pool. The
! I DSAR 1-3 Rev.14
(3 MYAPC
\'") transfer tube is closed by a manual isolation gate valve located in the spent fuel pool and by a blind -
flange located inside the containment.
Following the cessation of power operation and the removal of fuel from the vessel, the spent fuel inventory consists of a total of 1432 complete spent fuel assemblies,2 partially consolidated assemblies, and 2 partially full failed fuel rod containers residing in the spent fuel pool. Additionally, the spent fue' pool contains a number of trash baskets containing *Srs from prior pool and refueling cavity vacuuming efforts, reactor startuo neutron sources, Control Element Assemblies (CEAs), CEA plugs, incore Instrumentation (ICl) thimbles, and ICI assembly tips. The new fuel inventory is composed of the 6 unirradiated fuel rods located in the new fuel storage room.
The spent fuel is stored in a single tier rectilinear array of free standing modules. Each fuel assembly, failed fuel rod container, or trash basket is contained within an individual cell, Ce!!s are grouped together to form the free standing modules or racks. These high density spent fuel storage racks are designed for the passive reactivity control of the spent fuelin storage through the use of a fixed neutron absorber material and the spacing between assemblies. Design accommodations far cooling the spent fuel while in storage in the racks is provided through the use p of flow holes in the bottom of each fuel storage cell. The spent fuel storage racks are designed to (j maintain the fuel in a coolable, subcritical geometry during accident conditions in the pool.
The spent fuel pool is filled with a borated water solution to provide a medium for cooling the spent fuel, shielding for workers and the public from normal and accidental radiation exposure, and as a means of controlling the spent fuel reactivity, The moent fuel pool cooling system removes the spent fuel der y heat stored in the spent fuel pool by circuating the borated pool water through a heat exchanger. Each spent fuel pool cooling pump takes suction from the fuel pool, circulates the water through a heat exchanger, and returns it to the fuel pool below normal water level. Primary component cooling (PCC) water flows through the shell side of the heat exchar.jer and cools the tube side borated pool water. The heated primary component cooling water flows through the shell side of the component cooling water heat exchanger. The Service Water (SW) System provides service water to the tube side of the component cooling water heat exchanger and cools the shell side primary component cooling water. In the event that PCC or SW are not available, fire protection system water may be substituted by manually connecting a hose to a flange connection on the shell side of the fuel pool heat exchanger.
(h I i DSAR 1-4 Rev.14 i
l
\
The fuel pool cooling system also has a purification loop consisting of a pump, two filters, and a Cemineralizer which may be operated independently of the fuel pool cooling system. Flow from the discharge of the fuel pool purification pump is directed through the fuel pool prefilter, demineralizer, and postfilter, in that order, while the return is through the fuel pool cooling return line. The purification pump can also be used for skimming operation or to circulate pool water through the ,
heat exchanger.
Cooling water make-up to the spent fuei poolis available from a variety of sources. These sources include the Primary Water Storage Tank through the refueling purification system, the primary water system, and, as emergency sources, the fire water system and potable water from the town of Wiscasset water supply system. Boron make-u;,is available from the Refueling Water Storage Tank.
1.2.4 Radiological Waste Treatment System The radioactive waste treatment system is designed so the discharge of radioactivity to the environment will be minimized and at all time will be in accordance with the requirements of 10 p CFR 20.
t i V
. ..e radioactive waste treatment system is designed to collect, store, process, monitor and dispose of solid, liquid and gaseous radioactive wastes from the plant. The principal design objective of this system is to ensure that the general public is protected from exposure to radioactive waste products in accordance with 10 CFR 20. The normal sources of radioactive waste in the defueled condition includes those wastes generated from decontamination and decommissioning activities and residual radioective precipitants/ contaminants resulting from the storage of spent fuel in the spent fuel pool. The radioactive waste treatment system is designed to safely process radioactive wastes.
1.2.5 Radiological Waste Storage and Disposal System Liquid v aste may be mixed with a solidification agent, containerized into an approved container, sn3dl Led, and sealed in approved shipping containers and allowed to solidify befora shipment to an approved waste disposal site. Altematively, liquid waste may be released in accordance with the limits set in the Off-Site Dose Calculation Manual, (p)
DSAR 1-5 Rev.14 i
(N MYAPC U Contaminated compressible inaterials are compressw by a compactor into strong, tight shipping containers as required by applicable regulations.
Noncompressible solid wastes directly'resulting from power operation (such as contaminated
' metallic materials and highly contaminated solid objects) are placed inside approved containers and stored until they are disposed of at an NRC approved disposal site.
Demineralizer resins from the radioac;ive waste treatment system are isolated and primary grade
-water is used to flush the spent resin into the waste resin storage tank. The waste resin storage tanks are located in the Radiation Controlled Area (RCA) storage area in the lower level of the fuel handling building. The slurry and flush liquid are pumped by the waste resin transfer pump into the shipping container and dewatered. The shipping container is then shipped, or made ready for storage and eventual shipping, Shipping containers are placed in shielded shipping casks or inside concrete shield rings and disposed of at one of the NRC approved disposal sites.
Expended filter cartridges are raised into a filter removal shield and moved by the yard crane to a concrete shipping cask filter container or to storage in the underground RCA storage bunker or the ?
low level waste and equipment tem; orary storage building. Small, low activity finer cartridges are Q packaged for ultimate disposalin approved containers. In each case, the procedure ccnforms to DOT regulations for shipping to an NRC approved disposal site.
A monitoring program is established to assure that radioactive effluents are monitored and-controlled prior to release to the environment.
1.3 Facility Deslan Overview Key deshn data for plant systems and components relevant to the defueled condition of the Maine Yankee plant are listed in Table 1.3.1. Data relevant to the plant while operating are retained in this table for historical purposes only.
1.3.1. Plant Site and Population The Maine Yankee plant is located on the west shore of the Back River approximately 3.9 miles south of the center of Wiscasset, Maine. This location is shown in Figure 2.1-3.
'O
\'J DSAR 1-6 Rev.14
l MYAPC The minimum exclusion radius for the site is slightly greator than 2,000 feet. The outer boundary of the Low Population Zone, as defined in the 10 CFR 100 regulations, is 6 miles from the plant.
Within a 5 mile radius of the plant site, the resident population density is estimated to average 72 persons per square mile (1990). The nearest population grouping within 5 miles is situated around the Town of Wiscasset,3.9 miles NNE of the site. The town of Wiscasset has a population of about 3,340 people (1990). The surrounding towns of Edgecomb, Boothbay, Woolwich, and Westport Island, wholly or partially within a 5 mile radius, have a population of 6874 persons (1990). The city of Lewiston,24 miles WNW from the plant site, is the nearest city with a population in excess of 25,000.
The site characteristics are discussed in detailin Section 2.
1.3.2 Structures The major structures on the :lte are the reactor containment, primary auxiliary building, fuel building, turbine building, service building and circulating water pump house. Figures 1.3-1 and (3 1.3-2 depict the major structures along with administrative office buildings, storage buildings and V other structures.
The reactor containment is a steel-lined reinforced concrete cylinder with a hemispherical dome and an essentially flat reinforced concrete foundation mat. It served as a confinement barrier during plant op3 rations and provides adeques radiation shielding for any defueled or decommissioning condition.
The turbine building houses the turbine generator, the component cooling water heat exchangers and pumps, and the two diesel generators. Tho turbine building, and related components, are discussed in detailin Section 3.
The service building contains the main control room, switchgear rooms, shops and employee facilities. The service building, and related components, are discussed in detail ln Section 3.
The primary auxiliary building hot.ses equipment used for purification and processing of water from the reactor coolant system and the Fuel Building ventilation / filtration systeni. The fuel handling ventilation / filtration system is discussed in detail in Section 3.
O
\)
DSAR 1-7 Rev.14
4 g MYAPC i
'v') The fuel building provides space for the storage of new and spent reactor fuel and a!so waste disposal equipment. The fuel building arrangement is discussed in detailin Section 3.
The circulating water pump house contains the circulating water and service water pumps. Its arrangement is discussed in detailin Section 3.
The seismic criteria to be used in the design of the structures and equipment in the station are described in Section 3.1.
1.3.3 Chemical Treatment in the permanently defueled condition, a self-contained portable batch (ank and pump or the RWST will be used to maintain boron concentration in the spent fuel pool. Spent fuel pool water is periodically tested to assure the desired quality.
1.3,4 Process Instruments
,- Temperature, pressure, flow and liquid level monitoring is provided as required to keep the s ) operating personnel informed of plant and/or fuel storage conditions and to provih information from which plant processes can be evaluated or regulated. Instrument signal transmission for the plant instruments is electric.
The plant gaseous and liquid effluents are monitored for radioactivity. The results of this monitoring are displayed and high values are annunciated.
Area radiation monitoring stations are provided to monitor radioactivity at selected locations around the plant.
1.3.5 Shielding Shielding is provided so that radiation exposure of personnel does not exceed the recommended limits cf 10 CFR Part 20. The design of radiatiw shielding is dependent both on the extent of access required to a particular location ar'd on the sources of radiation adjacent to that location.
In the defueled condition, ALARA principles apply to assure that personnel have taken the necessary precautions prior to accessing or working on contaminated equipment.
C ';
1 d DSAR 1-8 Rev.14
(N MYAPC The plant is provided with a control room having adequate shielding to permit occupancy during all credible accident situations. The radiation shielding in the plant, in combination with plant radiation control procedures, ensures that operating personnel do not receive radiation exposures in excess of the applicable limits of 10 CFR Part 20. The control room is shleided to permit continuous occupan:y following any accidentai reLase of radioactMty resulting from a design basis accident, it should be noted however that, in the defueled condition, control room shielding is not required due to the lack of a significant source term from any of the design basis accidents, in addition, control rcom ventilation is not credited in the safety analyses.
1.3.0 Eleettical Equipment The plant electrical supply is provided by the auxiliary station service transformer connected through oil circuit breakers to the 115 kV transmission lines from the Central Maine Power Company system, in the event that off site power is interrupted, an on-site diesel generator is available for standby power, Offsite and onsite electrical power is provided for the safe storage of spent fuel. Following
,q a loss of offsite power to the spent fuel pool cooling system, and considering the significantly Q diminished decay heat load of spent fuel in the pool, ample time is available for operators to initiate attemate means of cooling or makeup for the spent fuel pool prior to substantial heatup or inventory Icss.
Batteries are installed to supply any required de power.
1.3.7 System Flow Diagrams Fiow diagrams for opch plant system are incorporated in the apprcpriate section of this report. The symbols used in these diagrams are shown in Figures 1,3 3 and 1.3-4.
/^'\
DSAF 19 Rev.14 l
t MYAPC -
U TABLE 1.3.1 MAINE YANKEE DESIGN CHARACTERISTICS Note: Selected information in this, table is being retained for historical purposes onl+ Such information retention is noted.
Plant Outout The Information h this subsection is Leing retained for historical purposes only.
Net Electrical Power Output (MWE) @ 2,700 MWt 905 Gross Electrical Power Output (MWE) @ 2,700 MWt 931 Maximum Expected Gross Electrical Output (MWE) 931 NEGleatJteam Sunniv System The information in this subsection is being retained for historical purposes only.
Core Thermal Output (MWt) 2700 Ox Operating Pressure (psig) 2235 Design Pressure (psig) 2485 Reactor Coolant inlet Temperature (F) 500-552 Reactor Coolant Outlet Temperature (F) 500-603 Pipe Size: Outlet (ID, inches) 33 1/2 (Wall Thickness, inches) 31/4 Inlet ( D, inches) 33 1/2 (Wall Thickness, inches) 3-1/4 Flow per Loop (ib/hr) 44.73 x 10' Number of Loops 3 Number of Pumps 3 Type Vertical, Centrifugal Mechanical Seais Design Flow / Pump (gpm) 120,000 Reactor Core The information in this subsection is being retained for historical purposes only, Total Heat Output (Btu /hr) 9.215 x 10' Heat Generated in Fuel (%) 97.5 DSAR- 1-10 Rev.14
(] MYAPC TABLE 1.3.1 (continued)
MAINE YANKEE DESIGN CHARACTERISTICS DNB Ratio at Nominal C> iditions Minimum DNBR for Design Transients (YAEC 1 Correlation)
Core Power Density (kW/ liter) 83.01 Number of Fuel Assemblies 217 Number of Fuel / Poison Rods per Assembly 176 Fuel Rod Pitch (inches) 0.580 Fuel Clad Material Zircaloy-4 or ZlRLO Fuel Clad Nominal Thickness (inches) CE & W 0.028 Fuel Clad Nominal Thickness (inches) Exxon 0.031 Fuel Poison Materials Number of Control Rod Locations (maximum) 85 CEA Pitch (inches) 11.57 CEA Poison Materials B.C/B.C with Ag in Cd tips Stainless Steel
(~~ Control Rod Drive Type Magnetic Jack
( Equivalent Core Diameter (inches) 136 Total Uranium (MTU) 80-83 Reactor. Vessel The information in this subsection is being retained for historical purposes only.
Inside Diameter (inches) 172 Overall Height, including CEDM Nozzles 42-1 3/8 (feet -inches)
Wall Thickness, Minimum (inches) 8 5/8 Wall Material A 533 Grade B Class 1 Steel Cladding Thickness (inches) 5/16 Cladding Material Weld Deposited 304 SS Design Temperature (F) 650 Design Pressure (psig) 2485 Total Weight (tons) 472 DSAR 1 11 Rev.14
MYAPC TABLE 1.3.1 (continued)
MAINE YANKEE DESIGN CHARACTERISTICS Steam Generators The information in this subsection is being retained for historical purposes only.
Number of Units 3 Type Vertical "U" Tube Upper Shell Outside Diameter (feet inches) 15-7 1/2 Lower Shell Outside Diameter (feet inches) 11 - 8 3/4 Overall Height (feet inches) 58 9 15/16 Number of Tubes 5703 Tube OD (inches) 3/4 Tube Material Ni-Cr Fe Alloy Primary Side:
Tube Side Design Pressure (psig) 2485 Tube Side Design Temperature (F) 650 Tube Side Operating Pressure (psig) 2235' p Coolant inlet Temperature (F) 500-603 Coolant Outlet Temperature (F) 500-552 Bottom Head Clad Material 304 SS Secondary Side:
Shell Side Design Pressure (psig) 985 Shell Side Design Temperature (F) 550 Steam Generator Dome Pressure (psla) 750-877 Operating Temperature (F) ~511-529 Quality (%) 99.8 Steam Flow / Steam Generator (10 lb/hr) 4.023 Turbine Cycle The information in this subsection is being retained for historical purposes only.
Turbine Design Tanden Compound, i HP 2 LP Turbines Exhaust Pressure (in Hg absolute)- 1.5 Steam Atmospheric Dump (% rated steam flow) 2.5 Steam Bypass to Condenser (% normal steam 45
- flow to condenser)
Feedwater Heate.r Stages 6 Condensate Pumps - Number 3 Half Capacity
-d DSAR- 1-12 Rev.14
1 l
l l
l
(~] MYAPC TABLE 1.3.1 (cont!nued)
MAINE YANKEE DESIGN CHARACTE_Ri&IlGR Design Flow (gpm) 9060 Design Head (ft) 960 Feedwater Pumps - Electrical - Number 2 Ha;f Capacity Design Flow (gpm) 14,000 Design Head (ft) 2038 ;
Feedwater Pump Steam Driven Number 1 Full Capacity Design Flow (gpm) 28,000 Design Head (ft) 2200 :
Circulating Water Pumps Number 4 Quarter Capacity '
Design Flow (gpm) 100,500 Design Head (ft) 26 Flectrical Generator The Information in this subsection is being retained for historical purposes only, Design Rating (MVA) 900 t Power Factor .90 k Terminal Voltage (kV) 22 NARS Auxillarv P'istems (a) Chemical and Volume Centrol System Except for the demineralizers and filters design characteristics, the information in this subsection is being retained for historical purposes only.
Normal Letdown Flow Rate (gpm) 80 Maximum Letdown Flow Rate (gpm) 200 Charging Pumps - Number 3 Fixed Capacity Design Flow (gpm) 150 Design Pressure (psig) 2850 Auxiliary Charging Pump - Number 1 Variable Speed Design Flow (gpm) 10 to 30 Design Pressure (psig) 3700 Regenerative Heat Exchanger - Number 1 Full Capacity Design Heat Transfer (Btu /hr) 10.0 x 10'
(
DSAR 1 13 Rev.14
l MYAPC TABLE 1.3.1 l
(continued)
MAINE YANKEE DESIGN CHARACTERISTICS Letdown Heat Exchanger Number 1 Full Capacity ;
Design Heat Transfer (Blu/hr) 7.82 x 10' Domineralizers - Number 3 Purification 1 Deborating ,
Nominal Rating (gpm) 80 Maximum Flow (gpm) 200 Resin Volume (ft') 32 Filter - Number 2 Type Cartridge Design Rating (gpm) 200 Nominal Filter Size (microns) 2 (b) Safety inloction System The information in this subsection is being retained for historical purposes only.
Safety injection Tanks Number 3 O Volume Total (ft') 3500 V Borated Water (ft*)
Nitrogen @ 225 psig (ft')
1500 2000 Design Pressure (psig) 250 Design Temperature (F) 200 Low Pressure Pumps - Number 2 Full-Capacity Rating, Each (gpm) 3000 Head (ft) 350 (c) GDDialnment Soray System The information in this subsection is being retained for historical purposes only.
Spray Pumps Number 3 Rating, Each (gpm) 3700 Head (ft) 305 (d) Refueling Water Tank The information in this subsection is being retained for historical purposes only.
Maximum Indicated Volume (gals) 331,000 Maximum Allowed Fluid Volume (gals) 338,400 p Boron Concentration 1720 ppm k)
DSAR 1-14 Rev.14
MYAPC TABLE 1.3.1 (continued)
MAINE YANKEE DESIGN CHARACTERISTICS (e)- Auxillary Feed System The information in this subsection is being retained for historical purposes only.
Steam Generator Emergency Feed Pumps (P 25A & C)
Motor Driven. Quantity: 2 Full Capacity l Rating (gpm) 500 l
. Head (ft) 2525 Steam Generator Auxiliary Feed Pump (P 25B)
Turbine Driven - Quantity: 1 Full Capacity Rating (gpm) 500 Head (ft) 2525 (f) Comnonent CoQHDg_ System The information regarding the Secondary Cooling Components is being retained for historical purposes only.
Component Cooling Pumps - Number 2 Primary 2 Secondary Rating, Each (gpm) 6000 Head (ft) 190 Heat Exchangers - Number 2 Primary 2 Secondary Vendor data sheet info. (Btu /hr)
E-4A. Primary 72.5 x 10' E-48, Primary 51.3 x 10' E SA, Secondary 51.3 x 10' E 58, Secondary 72.5 x 10' (g) Snent Fuel Coolina System Spent Fuel Pool Storage Capacity (design) 1758 assemblies
. Spent Fuel Pool Storage Capacity (actual) 1432 assembiles 2 consolid, assemb.
. 8 2 failed rod cages
- Volume (ft ) 59,116 Pumps - Number 2
~ Rating, Each (gpm) 772 O DSAR 1 15 Rev.14
= _ = _
MYAPC TABLE 1.3.1 (continued)
MAINE YANKEE DESIGN CHARACTERISTICS j l
Head (ft) 120 Heat Exchanger Number 1 -
Rating (Blu/hr) 22.3 x 10' Filter- Number 2 Type Cartridge Rating (gpm) 200 '
Nominal Size (microns) 2 Demineralizer - Number 1 Resin Type mixed bed Bed Size (ft') 32 Nominal Flow (gpm) 200 Conventional Plant Auxiliary SystemE (a) Service Water Svstem p Service Water Pumps Number 4 Full Capacity
(
V) Rating (gpm)
Head (ft) 10,000 66 (b) Compregned Air System Compressors - Number 3 Rating (scfm) 300-
, Discharge Pressure (psig) 100 Gentainment Structure The information in this subsection is being retained for historical purposes only.
Type Reinforced Concrete Diameter (ft inches) 135-0 Height (ft inches) _ _ 169-6 Liner- Material Thickness (inches) ASTM A516 Grade 60 Wall 3/8 Dome %
Floor 1/4 Design Pressure (psig) 55 Design Temperature (F) 280 Leak Rate (percent per day) 0.1 U DSAR 1-16 Rev.14
O MYAPC TABLE 1.3.1 (continued)
MAINE YANKEE DESIGN CHARACTERISTIC 1 ElectricaLEgulpment Station Batteries Number 6 Type Lead Calcium
- Rating (8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> rate in ampere hours)
Battery Nos.1 and 3 1760 Battery Nos. 2 and 4 560 Battery No. 5 1752 Battery No. 6 160 Chargers - Number 7 i Inverters Number 7 l
The information in this subsection below is being retained for historical purposes only, Main Transformer Number 2 Capacity (MVA) 600 Voltage (kV) 345 Diesd Generators - Number 2 Full Capacity Rating (kW) 2500 V Fuel Oil Capacity 1 Week Diesel Generator (Appendix R) Number i Rating (kVA) 250 O DSAR 1-17 .Rev.14
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l r~'s MYAPC ("l 1.4 [dentification of Agents and Contractqta Maine Yankee Atomic Power Company is the solo petitioner for the operating license. The Company was organized by eleven New England utility companies for the purpose of constructing and operating the Malne Yankee nuclear generating plant in Wiscasset, Maine. Maine Yankee, as owner and operator of the station, is responsible for the design, construction, fabrication of components, operation and quality assurance. The principal contractor organizations associated with the construction and operation of the Maine Yankee plant are: Entergy Nuclear, Inc. provides operating and management services. The Nuclear Services Division of the Yankee Atomic Electric Company (now Duke Engineering and Services)ln the performance of engineering studies, design reviews, plant modifications, and construction coordination. Combustion Engineering (CE), Inc.(now ABB-CE), as the designer and supplier of the (Vj Nuclear Steam Supply System (NSSS). The NSSS includes the reactor coolant system, reactor auxillary system components, nuclear and certain process instrumentation, and reactor control and protection system. The Stone & Webster Engineering Corporation as the designer and supplier of the balance of the plant equipment and structures. Additionally, Stone & Webster constructed the balance of plant with technical advice provided by CE for installation of the reactor plant components. The Westinghouse Electric Ccrporation as the original supplier and erector of the turbines and electrical generator. The low pressure turbines were replaced during Maine Yankee operations with equipment from Asea Brown Bovari. Combustion Engineering, Westinghouse, and EXXON Nuclear (now Siemens Power Corporation) were suppliers of nuclear fuel at various time during the operating lifetime of Maine Yankee. 7
'" DSAR 1 22 Rev.14
MYAPC 1.5 Materialincorporated By Reference i Certain program documents and associated top; cal reports or analyses have been incorporated
- into the DSAR by reference and are listed in each section as appropriate. This documentation may !
include information developed by Maine Yankee, as well as Yankee Atomic, ABB CE, ! Westinghouse,' Stone and Webster, and other organizations, j Some documentation that is incorporated by reference continues to be updated to assure that the information used is the latest available. These documents include the following:
- 1. Quality Assurance Program
- 2. Emergency Plan
- 3. Security Plan
- 4. Fire Protection Program
- 5. Fitness for Duty Program
- 6. Off Site Dose Calculation Manual
- 7. Process Control Program ;
- 8. Post Shutdown Decommissioning Activities Report t
- 9. Technical Specifications Each of these programs and plans may be modified as necessary in accordance with the regulatory and Malna Yankee requirements identified in section 6. ;
I h-DSAR 1 23 Rev 14_
MYAPC
\
SECTION 2.0 SITE CHARACTERISTICS TABLE OF CONTENTS Section TMa Eaga 2.1. Location And Area . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 ........................... 2.1.1 Population i
-2.1.2 Land Use 2.2 Ma' r:oroloay . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-15 2.2.1 General 2.2.2 On Site Meteorological Field Programs 2.2.3 Coastal Fog 2.2.4 Temperature 2.2.5 Precipitation 2.2.5.1 Snowfall, Snow and Ice Loading 2.2.6 Tornadoes, Hurricanes and Severe Thunderstorms 2.2.7 Environmental Monitoring Program 2.2.7.1 Program Description 2.2.7.2 Program Scope 2.2.7.3 Program Sample Media 2.2.7.4 Emergency Surveillance 2.2.7.5 Program Evaluation and Reports 2.2.7.6 Land Use Census 2.2.7.7 Sample Locations 2.3 H yd roloay . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , , , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-44 2.3.1 Surface Hydrology 2.3.2 Oceanographic Features 2.3.3 Probable Maximum Flood 2.3.3.1 Maximum Water Surface Elevation 2.3.3.2 Wave Runup and Wave Forces 2.3.3.3- Extreme Low Water 2.3.4 Ice Loading, Oil Spill and Debris Blockage 2.3.5 - Ground Water Hydrology \
DSAR 2-1 Rev.14
~ . _ _ . . . - -
.1 T a
- e MYAPC f
i S SECTION 2.0 1 SITE CHARACTERISTICS i TABLE OF CONTENTS (continued) 4 s i Sectiori IWs East i ,i 2.4 Geoloa_v_.............................................................................................. 2 63 v ] j 2.5 Selsmoloav........................................................................................ 2 66
! 2.5.1 Tectonics 4
- 2.5.2 Tsunamis i,
/ 8 i 4 b 1 i i h i . J 4 e i e 4 t .A 4 i s N 1 i DSAR 2 il Rev.14 f 5
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. MYAPC SECTlON 2.0 LIST.OF TABLES l Table No. Titla 2.1.1 Resident Population of Towns Within Ten Miles 2.1.2 Agricultural Statistics for Principal Counties Within 50 Miles 2.2.1 Wind Direction Versus Wind Speed 2.2.2 Temperature Data for Portland and Brunswick Naval Air Station 2.2.3 - Precipitation Statistics Brunswick 2.2.4 Precipitation Statistics Portland 2.2.5 Portland, Maine 1940-1965 Snowfallin Inches
-2.2.6 Wind Data for Brunswick Naval Air Station 2.2.7 Portland, Maine 1940-1965 + Maximum Wind Speeds 2.2.8 Environmental Monitoring Program - Profile Media 2.2.9 Environmental Monitoring Program - Functk>nal Media 2.2.10 Radiological Environmental Surveillance Locations q 2.3.1 Currents -
,G 2.3.2 Tides E s DSAR 2-lil Rev.14
MYAPC U SECT 10N 2.0 LIST OF FIGURES Floure No. Illig 2.1 1 Site Boundary 2.1 2 Local Topographical and Low Population Zone 2.1 3 Location and General Topography 2.14 Pau!ation Distribution Based on 1990 U.S. Census 0-5 Miles 2.15 Population Distribution Based on 1990 U.S. Census 0 50 Miles 2.1 0 Cities: Population Over 5,000 0-50 Miles Based on 1990 Census 2.1 7 Cities: Population Over 25,000 0100 Miles Based on 1990 U.S. Census 2.1 8 Projected Population Distribution Year 2000 5 50 Miles 2.2 1 Return Period of Extreme Short Interval Rainfall, Portland, Maine 2.2 2 Environmental Radiological Sampling Locations Within i Kilometer of MY 2.2-3 Environmental Radiologicai Sampling Locations Within 12 Kilometers of MY 2.2-4 Environmental Radiological Sampling Locations Outside 12 Kilometers of MY n 2.2 5 Direct Radiation Monitoring Locations Within 1 Kilometer of MY 2.2-6 Direct Radiation Monitoring Locations Within 12 Kilometers of MY 2.2 7 Direct Radiation Monitoring Locations Outside 12 Kilometers of MY 2.3 1 Plant Site and Adjacent Waters 2.3-2 SheepscotWatershed Area 2.3-3 Maximum Probable Flood for Sheepscot River at W;scasset 2.3-4 Plan Topographical Profiles 2.5 5 Tepographical Profiles 2.3 6 Profile - Maximum Wave Runup 2.3 7 Contour - Maximum Wave Runup 2.3-8 Percent of Waves During Critical Period 2.4 1 Boring Location Plan 2.5 1 Tectonic Map of New England 2.5 2 Compilation of Earthquakes Intensity VI or Greater Northern New England 2.5-3 Compilation of Earthquakes - Southern Maine 2.5-4 Earthquake Intensity Attenuation - Northeastern United States 2.5 5 Earthquake MM Intensity ( )
\ DSAR 2-iv Rev.14
p MYAPC
't SECTION 2.0 1HE SITE AND ENVIRONS 2.1 LQGation And Areia The Maine Yankee Atomic Power Station is located in the town of Wiscasset, Lincoln County, Maine. Site coordinates are approximately 43 degrees 57 minutes 5 seconds north latitude and 69 degroes 41 minutes 45 seconds west longitude. The plant site is bounded by Back River on the east, mainland on the north, and Birch Point Road on the west. Maine Yankee has purchased this land in fee. It has also purchased Foxbird Island and Little Oak Island. The plant is located on a peninsula known as Bailey Point which extends towards the south to Montsweag Bay, as shown by the site area map in Figure 2.1-1 and topographical map in Figure 2.12. Figure 2.13 locates the site relation to nearby towns and cities.
Since lands within the site boundaries are owned in fee by Maine Yankee, they are subject to its control. The immediate plant area is enclosed on all sides by a chain link fence. Guard service is maintained at all times to prevent unauthorized entry into this fenced area. The 345 kV
.m - substation is located north of the station. It is enclosed with chain link fence and locked gates, f I v
The waters of Back River, Montsweag Bay, and its tributaries are tidal and open to boating, both commercial and recreational. Regulation of this boating is the responsibility of the United States Coast Guard and the state of Maine. The site area and adjacent Back River provide a minimum exclusion radius of 2,000 feet (Figure 2.1 1). A low population zone (LPZ) has been established with a radius of six miles, as shown in Figure 2.1-2. Also indicated in Figure 2.12 are the schools within the LPZ. There are no hospitals or other institutions within this area. The plant site itself is located on a ridge of bedrock running northeast to southwest to form Bailey Point. The maximum elevation of this rock is a knob 75 feet above mean sea levellocated about 700 feet northeast of the plant. The general elevation of Bailey Point varies from sea level to 40 feet above mean sea level. The plant area is graded to elevation 20 feet. A layer of glacial till aas been deposited above the bedrock and has an average depth of 15 to 20 feet. A detailed description of site geology is given in Section 2.4. \ 13 )
'v' DSAR 2-1 Rev.14
~'
MYAPC
'v' 2.1.1 Population '
The concentration of population in the vicinity of the site is low. Within a 5 mile radius of the site, the resident population density is estimated to average 72 people per square mile in 1990. Figure 2.14 Indicates the approximate distribution of the 1990 resident population within 5 miles. The nearest population grouping within 5 miles is situated around the center of the town of Wiscasset, 3.9 miles NNE of the site. Scattered housing marks the nature of the balance of the population in the immediate site area. For towns within the 5-mile radius, the resident population has shown a 17% increase between 1980 and 1990, as compared with a 31% increase for the period 1970 to 1980. Table 2.1,1 lists the resident population for all towns within 10 miles of the site for the census dates 1960,1970,1980, and 1990. For those towns wholly or partially within a 10-mile radius, the resident population experienced a 15% growth between 1980 and 1990, and a 15% growth for the period 1970 to 1980. There is a summer seasonal increase in population associated with recreational activities which takes place along the Maine coast. Based upon Reference 1, there is about a 50% increase in the population within a 5-mile radius of the site during the summer period. () The surrounding area to a distance of 50 miles has a 1990 population distribution, as shown in Figure 2.15. The total resident 1990 population within 50 miles is estimated to be 649,895 (83 people per square mile). This represents about a 13% growth in population from 1980 when the 50-milo cumulative population was estimated at 577,300. Many of the population groups within 50 miles of over 5,000 persons are shown on Figure 2,1-6. The major population centers of over 25,000 people are shown on Figure 2.17. The outer boundary of the LPZ, as defined in 10 CFR 100,is 6 miles. Lewiston,24 miles NW from the plant, is the nearest city with a population in excess of 25,000, and thence lies well beyond the outer boundary of the LPZ. Population changes for the areas in the vicinity of the site were estimated in the original FSAR through the year 2000. The fundamental basis for estimating tho year 2000 statistics was the U.S. Census Forecast Il D (Reference 2) for the whole state of Maine. This information was extrapolated beyond 1985. On the basis of the change in population between 1940 and 1960, it was assumed that one-half the population increase in the future for the entire state would occur in the following seven counties: Androscoggin, Cumberland, Kennebec, Knox, Lincoln, Sagadahoc and Waldo. A linear extrapolation of the 1960 and 1970 population figures was used to determine O \ )
's DSAR 2-2 Rev.14
_._- -- - -- - -- - -.-. ._~_- - - . _ __- - - n MYAPC the year 2000 distributions. Within 5 miles of the site, no major increase in population is expected. The original FSAR 50-mile radius population estimate for the year 2000 was approximately 816,000 (Figure 2.18). If the rate of population growth over the 10 year period from 1980 to 1990 (13%) . for the seven principal counties located within 50 miles of the site is used as the basis to project the 1990 population within 50 miles (Figure 2.15) over the next decade, the year 2000 population would be approximately 734,000, or 82,000 below the estimate in the original FSAR. There Is no historical basis for predictions of industrial growth in the area. Commercial and service , trades will probably continue to be responsive to needs of residents and tourists. The recreational I opportunities with respect to boating, fishing, and local features of historical Interest will probably i stimulate seasonal trade. 2.1.2 Land Use Within 5 miles of the site, land use is largely home sites, small businesses, summer houses, idle farmland and forest. There is one small dairy within this area, with several other locations having a few milk cows for private use. Housing is scattered along principal roads and is concentrated
- only in the center of Wiscasset.
O The waters near the plant are reported to be relatively low in productivity of fish and shellfish. Some lobstering is carried out in Montsweag Bay and the Back River. The primary type of boating in the Mordaweag Bay - Back River area is shallow craft pleasure boats. With no commercial traffic
*n the area, there is no hazard to the plant from potential accidents with commercial barges or i boats carrying toxic or explosive materials, i
, The Wiscasset Municipal Airport is the nearest airport to the site and is located over a mile northwest of the Maine Yankee containment, as shown on Figure 2.1 1. It consists of one runway , 3,400 feet long and 75 feet wide. The single runway (7-25) is oriented such that takeoffs and landings are on headings of 070 or 250. Approximately 80% of the time, runway heading 250 is used, . . The airport is used almost exclusively by private aircraft such as the Piper Colt, Piper Cherokee, fi l Cessna 150 and 172, King Air, and Queen Air. This type of light aircraft accounts for about 500 takeoffs and landings per month at the facility. The largest aircraft that typically lands at Wiscasset is the King Air 200 type aircraft. This larger type aircraft is estimated to account for about 5 takeoffs or landings per month. b V DSAR 2-3 Rev.14 _ _._ .__.m._ . _ _ . . _ ._- - .
MYAPC \ The normal takeoff pattern is a straight climb to 700 feet followed by a 45 degree turn to the laft. When the aircraft reaches an altitude of 1,000 feet, it is considered to be out of the pattem. During landing, the planes fly a downwind course parallel to and to the right of the runway at an altitude of approximately 1,000 feet. Upon passing the end of the runway, the aircraft would execute two 90 degree tums to the left to bring it around to the proper heading for landing. The distance between the runway and the airplane during the downwind leg depends on the size of the aircraft. There oro no active plans for expansion of the Wiscasset Altport. Several navigational aid addit'ons to the airport have been made. A non directional beacon (NDB) approach was installed - at ti e Wiscasset Airport in 1977 and two Global Positioning System approaches were installed in 19f 6. Neither of these instrument flight rule (IFR) approach capabilities is expected to change the tratfic count or type of aircraft that frequent the alrport. The land use within a 10-mile radius of the Maine Yankee Nuclear Plant is also mainly farmland, with recreational activities taking place on a series of peninsulas jutting into the Gulf of Maine. Because of its unique coastal terrain and many M., the area is a summer recreational center for boating and other water related activities. This summer recreation and its supportive businesses, motels, restaurants, shops, etc., provide much of the economic base for the area. Another activity utilque in many respects is the marine worm industry. The marine worm industry harvests two species, the sand worm (the Nevels virens) and the blood worm (Glycora dibranchiata), The worm digging is confined to mudflats in the intertidal areas. These worms are sold for bait to commercial and sports fishermen along the Atlantic coast. Industrial activity within the 10-mile radius is somewhat limited, with ship building in the city of Bath on the Kennebec River the largest industrial facility in the state. Agricultural production for the seven principal counties within 50 miles of the site is summarized on Table 2.1.2. Section 2.1 References .
- 1. Evacuation Time Estimates for the Maine Yankee Plume Exposure Pathway Emergency Planning Zone. HMM Associates, Inc., April 1992
- 2. US Bureau of Census, Current Population Reports, Series P 25, #L.6, Illustrative Projection of Population of State,1960-1985, US Govemment Printing Office, Washington, D.C.,1965.
DSAR 2-4 Rev.14
I L MYAPC O TABLE 2.1.1 i RESIDENT POPULATION OF TOWN WITHIN TEN MILESJ4 ! Maine 1960 1970 1980 1990-Wiscasset' 1800 2244 2830- 3339 Edgecomb' 453 549 832 993 Boothbuy* 1617- 1814 2299 2648 Boothbay Harbor 2252 2320 2193 2347 Woolwich* 1417 1710 2169 2570 Dresden 766 787 982 1332 Alna 347 315 428 571 Newcastle 1101 1076 1233 1538 Damariscotta 1093 1264 1498 1811 Bristol 1441 1721 2094 2326 South Bristol 610 664 800 825 Southport 416 473 590 645 Phlppsburg 1121 1229 1529 1815 i Bowdoinham 1131 1294 1828 - 2192 Bath 10717 9679 10222 9799 West Bath - 766 836 1310 1716 Brunswick 15797 16195 17392 20906 Westport Island' 133 228 418 - 663 Arrowsic 177- 188 305 498 Georget;wn 790 464 731 914 .; Total 43945 45050' 51683 59448 Towns that are wholly or partially within a 5 mile radius.
. (1) 1990 Census of Population and Housing, Summary Tape File ia, New England Division, U.S. Department of Commerce, Bureau of the Census.
'DSAR 2-5 Rev.14
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MYAPC (o) k/ TABLE 2.1.2 AGRICULTURAL STATISTICS FOR THE PRINCIPAL COUNTIES WITHIN 50 MILES 01 Sagadahoc Uncoln Knox Waldo Kennebec Androscoggin Cumberland fdillLCaws $51 798 662 4520 7845 4553 2121 Numberof Head Cattle & Calves 589 815 787 3933 5500 3847 3163 Number of Head Sold Sheep & Lambs 379 968 1300 1191 751 509 977 inventory Hoon & Plas 133 213 72 382 381 295 713 Number of Head Sold Numbu. of Chickens 401 D (2) D(2) D(2) 487,026 0(2) D(2) Number of Drollers D(2) D (2) 0(2) 252.085 D(2) 190 327 Vagg'thics 199 177 994 125 321 442 816 Acrea6 Orchards 40 86 85 276 830 1478 340 Acreage l (1) Source: 1992 Cencus of Agiculture, Volume 1, Geographic Area Series, Part 19, Maine State and Coanty Data, Bureau of the Census (Table 1, County Summary Hi9 hlights). (2)(D) Withheld to avoid disclosing data for individual farms. C'\ DSAR 26 Rev.14
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(N MYAPC k- 2.2 Meteorology 2.2.1 General The Bailey Point site is located in the mid-coastal region of Maine. This coastal region is characterized by many inlets, bays, channels, harbors, rocky, islands, and promontories. The area adjacent to the site has many small forested hills. The general climatic regime is maritime with its cool air moving in from the North Atlantic. Of special importance, from an engineering standpoint, are the extremes in annual snowfall for the coastal region; the occasional heavy rains, the coastal storm or "Nor' easter" with its resultant strong winds and heavy rain or snow; and sometimes glaze or " ice storms". These and other pertinent meteorological data which have been compiled by the TRC Corporation of Hartford, Connecticut, are presented in the following sections. 2.2.2 On-Site Meteorological Field Programs (q
)
An initial data collection program was undertaken at the site of the Maine Yankee Atomic Power Station to provide information on meteorological conditions for dispersion analyses for the original FSAR. Data for one year, from July 1,1967 to June 30,1968, were evaluated and formed the bases for those analyses. Appendix A contains a discussion of the July 1967 - June 1968 data l collected from the initial monitoring program. ! An upgraded on-site monitoring program which meets the intent of Revision 0 to Regulatory Guide 1.23 was installed in late 1976 and is currently in operation. A description of this upgraded system is also presented in Appendix A, along with wind and stability data summaries for one year of operation, from January 1,1979 to December 31,1979. This one-year period -!-record also forms the basis for the off-site dose accident analyses presented in Section 5 of inis report. 2.2.3 Coastal Fog Heavy fog is frequent and sometimes persistent along the coast, and may occur on one day in six during certain portions of the year. Data for the 11-year period (1951-1962) from the Brunswick Naval Air Station located 13 miles from the site indicate that 4.1% of the time (3,855 out of 95,073 observations), fog conditions exist. A fog condition is said to exist when the visibility is 0 to % mile. DSAR 2-15 Rev.14
t MYAPC \-) The breakdown of fog coaditions for the various wind speed classifications is as follows: 0- 2 mph 1.8% 3-14 mph 2.2% 15-23 mph 0.2% 24+ moh <0.1% Total 4.1% During fog conditions, the time frequency of wind direction versus wind speed is illustrated in Table 2.2.1 From Table 2.2.1, it can be seen that when fog is present, the wind speed is equal to or greater than 3 mph for approximately 60% of the time. These winds are adequate for dispersing radioactive effluent during fog conditions. This also indicates that the fog is of the advection type (formed ay moving warm moist air over a cold surface) rather than a radiation type fog which is associated with very little air movement. That the fog is of the advection type is further substantiated in Table 2.2.1 by the preponderance of onshore winds (south to southwest) during fog conditions. f,,T 2.2.4 Temperature Q,) The temperature of the coastal region tempered by the Atlantic Ocean is not subject to the wider extremes of the inland areas. The average annual temperature is about 45'F, with the frequency of temperature above 90*F being very small. The average January temperature is about 22'F with between 10 and 20 days of sub-zero temperatures occurring yearly, Temperature data for 11 years for Brunswick Naval Air Station (13-14 miles southwest), and 30 years of data for Portland (39-40 miles southwest) are shown in Table 2.2.2 and give representative values for the site. 2.2.5 Precipitation Precipitation along the Maine coast is influenced by the Atlantic Ocean. Summer thunderstorm activity.is somewhat suppressed by the effects of the cool ocean, while winter precipitation is increased by coastal storms or "Nor' easters". These corrbined effects give this area more precipitation in the winter months than in the summer months. Monthly totals are about 4 inches during the winter as compared to 3 inches in summer. Total precipitation (Reference 1) averages nearly 46 inches for the coastal areas. Winter precipitation occurs mostly as rain or wet snow. Also, this area, more than further inland, is subject to occasional glazing or " ice storm" conditions. A e ! V DSAR 2-16 Rev.14
q MYAPC Summaries of precipitation statistics (References 2 through 4) for Brunswick and Portland are shown in Tables 2.2.3 and 2.2.4. Intense rainfall may be produced by the occasional severe thunderstorm, hurricane, or "Nor' easter." The maximum recorded point rainfall (References 5 and 6) of short time intervals for Portland, Maine, (period 1893-1961) is given below. Short Time 'nterval Precioitation Portland. Maine Minutes Hours Inches Inches 5 10 15 30 60 2 3 6 12 24 0.51 0.78 1.09 1.49 2.11 3.40 4.51 5.84 7.09 7.71 The return period of extreme short-interval rainfallis a useful design and planning guide. The 1
]p nearest location for available retum period data which should be representative for the Bailey Point area is Portland, Maine. This data is illustrated in Figure 2.21, 2.2.5.1 Snowfall, Snow and Ice Loading The average seasonal snowfall has a marked variation along the coastal region, which may be as little as 29 inches or as much as 119 inches (References 2 and 3). Along the coast, the snow cover may entirely melt once or more in the midwinter to be replaced by new snow. The average numbar of days with 1 inch or more of snowfall for Portland is 20 per season. Table 2.2.5 shows average snowfall statistics for Portland which are considered to be representative of the site.
Snow-load data for the Bailey Point area, from a HHFA study (Reference 7) conducted in 1952, are as follows: Wt of Estimated Wt of Seasonal Wt of Max Max Accumulation Snowpack Equated or Snowpack on on Ground Plus Wt of Exceeded 1 Yr. in 10 Record Max Possible Storm 40 lbs ft2 60 lb ft2 80 lb ft2 7 DSAR 2-17 Rev.14
MYAPC V Data relating to freezing rain and resultant formation of glaze ice on highways and utility lines have been obtained from the following studies: American Telephone and Telegraph Company,1917-18 to 1924-25 Edison Electric Institute,1926-27 and 1937-38 Association of American Railroads,1928-29 to 1936-37 Quartermaster Research and Engineering Command, U.S. Army,1959 A polar front wave with an active warm front moving in a north or northeasterly direction toward MaSe is the most typical synoptic condition for the formation of glaze or freezing rain. A quasi-stationary high pressure area north of New 'ingland, with the center of the ridge or cell usually located somewhere northeast of Newfoundland, causes a flow of continental-polar air over the area from the south or east behind the warm front. If the over running maritime-tropical air or modified continental-polar air is warmer than 32*F, while the continental-polar air beneath the front has temperatures of 20*F to 30'F, then freezing rain or drizzle may result. Glaze and ice storms of winter are usually of brief duration, although a few widespread and gy prolonged ice storms have occurred. The following data for glaze storms (Reference 8) will apply: k
- 1. Time of occurrence - October through April
- 2. Average frequency without regard to ice thickness,1-3 storms per year
- 3. Return periods for freezing rain storms producing ice of various thicknesses are:
Ice 0.25 every year 0.50 every year 0.75 at least every 3 years The extreme radial thickness of glaze on utility wires for the period 1928-29 to 1936-37 for the Bailey Point area was between 1.75 inches to 1.99 inches. A U.S. Weather Bureau summary for the years 1939-1948 gives the actual number of days with freering precipitation (without regard to ice formation) for Portland, Maine, as follows: Total Days Nov Dec Jan Feb Mar Apr in 10 Years 1 27 24 20 12 2 86 /^\ v i s UI DSAR 2-18 Rev.14
c MYAPC
. 2.2.6 L Tornadoes, Hurricanes and Severe Thunderstorms Coastal regions are sometimes seriously affected by a variety of storms. They generate strong _
winds, heavy rain or snow and, occasionally, a glaze of " ice storm". In winter, these storms produce some of the heavier snowfalls in the coastal area, in summer or fall, a storm of tropical
. origin mayialso affect the coastal area. Usually these are similar to the "Nor' easter".but -
Loccasion'ally a few may attain near or full hurricane force.- Extreme wind data for Brunswick Naval Air Station for the period of December 1951 through' October 1962 are shown in Table 2.2.6. The highest wind speeds recorded for the period 1940-1965 for each month at Portland is given in Table 2.2.7. L Collins and Howe (Reference 9) have developed indices of relative storm damage for different parts - of the United States using storm data from the 1954-1963 decade, in the study, an "index damage _ potential" is defined in units of 1,000ths of 1 percent of residential property values per year for various types of storms. The "index damage potential" which excludes tomadoes, hurricanes,
. tropical storms and hail for the Bailey Point area is 8, compared to a value of 16 for the Oklahoma-Kansas area.
.- (
- Storms of hurricane origin do not affect Maine in most years. - In 1954, two hurricanes affected Maine within a period of less than 2 weeks. The first, " Carol", traveled northward along the Maine- -
= New Hampshire border on August 31. --Wind speeds were no longer full hurricane force, but--
[ = substantial property and crop damage resulted in westem Maine. -Then "Edna" entered the 1
. coastline near Eastport on September 11. In this case, the prf:1cipal damage was due to heavy flooding and washing rains." Two hurricanes in 1 year should be expected probably less than 1 year
. in 10. The "index of hurricane and tropical storm damage potential"(Reference 9)(defined in units of 1,000ths of 1 percent of residential property values per year) for the Bailey Point area is 95 as compared to 337 for the Cape Cod area,606 for the Cape Hatteras/ North Carolina areas, and 633
. for the Miami, Florida area.
Tomadoes are not a common phenomenon in Maine. Yet they do occur occasionally. Of the few, about 80% occur between May 15 and September 15. About 90% strike between 1:00 and 7:00 p.m. The peak month is July and the peak hour of occurrence.ls 2:00 to 3:00 p.m.
-- s)
DSAR 2-19 Rev.14 e b
MYAPC i
'd Fifty year totals (1916-1965) for tornadoes listed by month for the state of Maine and other New England states are as follows:
_50-Year Tornado Record 50 yr Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec Total Maine 0 0 1 0 5 6 15 12 8 2 0 0 49 Mass 0 0 0 0 12 15 26 9 6 6 3 2 79 Vt 0 1 1 0 3 7 8 2 0 1 0 0 23 NH 0 0 0 0 6 8 14 6 2 0 0 1 37 Conn 0 0 0 0 7 5 8 4
, 2. 2 0 0 28 A National Severe Storms Forecast Center (NSSFC) listing of tomadoes within a 125 nautical mile radius of the site indicates that 120 tomadoes occurred during the period 1950 through 1993, with a mean path area of 0.101 square miles (Reference 10). Thom (Reference 11) has developed a procedure for estimating the probability of a tomado striking any point from an analysis of mean
("j tomado path area and the frequency of tomado occurrence in the region around the site. Applying Thom's procedure to the NSSFC data gives an annual probability of about 1x10-5 of a tomado , striking any point within 125 nautical miles (144 miles) of the site. This calculation accounts for the water area within the 125 nautical mile region. l Thunderstorms and hail storms occur most frequently from mid-spring to early fall. Thunderstorms occur on about 10 to 20 daya a year along the coast. The most severe may be attended by hail. The mean number of days that Portland experiences thunderstorms for the period 1940-1965 is as follows: Jan I Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec Total 0 1 2 S 4 3 2 *
- 1 19
*Less than one-half day
'p)
'd DSAR 2-20 Rev.14
n MYAPC-l' 1 2.2.7 Environmental Monitoring Program 2.2.7.1 Program Description in the operation and decommissioning of Maine Yar m, small quantities of radioactive gases and liquids are released to the environment in a controlled manner in accordance with the applicable federal and state regulations. - An environmental monitoring program was established to demonstrate the adequacy of safeguards inherent in the plant design, and the effectiveness of in-plant monitoring of controlled releases of radioactive materials. The program consists of three phases, pre-operational, operational, and post operational each having specific objectives. The pre-operational phase was started approximately two years prior to operation of Maine Yankee to establish background radiation levels and concentrations at selected locations, to determine variability between sample locations, and to observe any cyclical or seasonal trends in the environmental sample media. The operational phase of the program had the following objectives:
- 1. To assure those organizations responsible for public health and safety that concentrations of p!y radionuclides in the environment resulting from plant operation met the applicable federal, state, and license regulations.-
i
- 2. - To make possible the prompt recognition of any significant increase in environment level of radioactivity related to plant operation.
- 3. To differentiate plant releases from other abnormal trends in environmental radiation due to fallout, other nuclear facilities and seasonal changes in natural background.
- 4. To obtain information on the critical radionuclides and pathways leading to the quantitative evaluation of the dose to man resulting from the the operation of the plant if significant trends in positive environmental concentration are Identified.
A description of the radiological environmental monitoring program, in tabular form, is shown on Tables 2.2.8 and 2.2.9. O DSAR 2-21 Rev.14 j
p MYAPC
'v 2.2.7.2 Program Scope To determine whether or not there is a significant effect by the plant on the surrounding environment, a two-zone sample collection method is employed for most media. The zones have been designated as follows:
Zone 1 - The area within approximately a 5-mile radius from the plant site. This area is considered under the influence of the plant. Zone II - The area outside approximately a 5-mile radius from the plant site. This area provides background data for the environmental surveys. Since data from each of the two zones are correlated, there is simultaneous monitoring in both areas to establish and facilitate a statistical analysis of the survey data. This allows Maine Yankee to differentiate between plant releases and other abnormal trends in environmental radioactivity due to fallout from atmospheric nuclear weapons tests or other sources of radioactivity. Direct radiation monitoring locations are grouped into an inner ring and an outer emergency response ring, as well p as a control group. 2.2.7.3 Program Sample Media The program monitors four pathway categories. They are the direct radiation, airborne, waterbome, and ingestion pathways. Each of these categories is monitored by the collection of one or more sample media, which are listed below, and are described in more detail in this section. Additional sampling media may at times supplement those included in the basic pathway categories. Airborne Pathway Waterborne Pathway Air Particulate Sampling Estuary Water Sampling Charcoal Cartridge (Radioiodine) Sampling Ground Water Sampling Shoreline Sediment Sampling inoestion Pathway Direct Radiation Pathway Milk Sampling TLD Monitoring Fish and invertebrate Sampling Y/ DSAR 2-22 Rev.14
MYAPC
) 1. Altborne Monitoring Air monitoring stations are established at a total of five locations, four of which are Zone I locations. Alrborne particulates and radiolodines are collected by passing air through a 47 mm fiberglass filter in series with an iodine adsorption medium.
Air sampler pumps operate continuously and a dry gar meter is incorporated into the sampling stream to measure the total amount of air sampled in a given interval. The air particulate filters are collected weekly and analyzed for gross beta radioactivity. Weekly air particulate filters from each location are composited and analyzed for gamma emitting radionuclides. Charcoal filters are collected weekly and analyzed for lodine-131 radioactivity.
- 2. Waterborne Pathways A composite sampler at the plant outfall area collects an aliquot of estuary water at least every two hours. On a monthly basis, this composited sample, as well as a grab sample from the control location, are analyzed for gamma-emitting radionuclides. The samples are composited at the analytical laboratory for p quarterly tritium analyses.
V The collection of fresh or ground water samples is not required since no source of water used for drinking or irrigation purposes is in an area where the hydraulic gradient or recharge properties are suitable for contamination. However, samples may be periodically collected to provide additional information regarding potential contamination of ground water media. Sediment cores are collected at both the former and the current discharge areas on a semiannual schedule. Each sample is analyzed for gamma-emitting radioactivity.
- 3. Ingestion Pathways Milk samples are collected monthly from three locations. The samples are analyzed for gamma-emitting radionuclides and low level lodine-131 radioactivity.
Samples of food crops from indicator and controllocations are needed only when milk sampling is not being performed since both assecs radioactivity levels in the same pathway. At least tyr commercially or recreationally important marine biological specimens (such a. rish, mussels, crabs, and lobsters) are collected at (a 'd
)
DSAR 2-23 Rev.14
p MYAPC the Long Ledge (discharge) area as well as at a control location on a semiannual or seasonal schedule. All samples are analyzed for gamma-emitting radionuclides.
- 4. Direct Radiation Pathway Direct gamma radiation measurements are obtained at 38 stations, which consist of inner ring, outer ring, and control locations. Thermoluminescent Dosimeters (TLDs) are employed to record the integrated gamma radiation t.xposures over a quarterly period.
2.2.7.4 Emergency Surveillance The radiological environmental monitoring program surveillances required when the Emergency Plan has been activated are addressed in the Emergency Plan. 2.2.7.5 Program Evaluation and Reports A report on the radiological environmental monitoring program is submitted annually to the USNRC. The report contains a summary, interpretations and an analysis of trends for the results of the !p
) radiological environmental surveillance activities for the report period. Included are comparisons with operational controls and previous environmental surveillance reports, plus a description of the radiological environmental program and maps of all sampling locations. An assessment of the observed impacts of the station on the environment is also included.
1 2.2.7.6 Land Use Census A census of the relevant land use activities surrounding Maine Yankee is performed annually to ensure that the optimum milk sampling locations are being monitored (or food crop locations if milk sampling is not being done). Specifically, the location of the nearest milk animal, the nearest garden of greater than 50 square meters, and the nearest residence in each of the 16 meteorological sectors within five miles of the plant is identified. Dose calculations are done to determine the optimum milk or food sampling locations. 2.2.7.7 Sample Locations The sampling and monitoring points for the measurements involved in this surveillance program are presented in Table 2.2.10, and locations are shown on Figures 2.2-2 through 2.2-7. A
!'o) DSAR 2-24 Rev.14
----__ J
(3 MYAPC
'b] Section 2.2 Refetences
- 1. " Rainfall Intensity - Duration - Frequency Curves", Techrkal Paper No. 25,1955, US Weather Bureau.
- 2. " Climatic Summary of the United States - Supplement for 1951-1960", New England, US Weather Bureau.
- 3. " Climates of the States - Maine", September 1959, US Weather Bureau.
- 4. " Local Climatological Data - With Comparative Data",1965, Portland, Maine, US Weather Bureau.
- 5. " Maximum 24-Hour Precipitation in the United States", Technical Paper No.16, US Weather Bureau.
- 6. " Maximum Recorded United States Point Rainfall for 5 Minutes *o 24 Hours", Technical Paper No. 2, Revised to 1961, US Weather Bureau.
G 7, " Snow Load Studies", Housing Research Paper No.19, Housing and Home Finance Agency,1952.
- 8. " Glaze, its Meteorology and Climatology, Geographical Distribution and Economic Effects", Quartermaster Research and Engineering Center,1959 U.S. Army.
- 9. " Weather and Eytended Coverage", George F. Collins and Ge'irge M. Howe, TRC Service Corporation.1964.
- 10. National Severe Storms Forecast Center, A Listir g t' Tornadoes for Period 1950-1993,@ NCAA, Kansas City, MO.
- 11. "Tomado Probabilities", H.C.S. Thom, Monthly Weather Review, October / November / December 1963.
/ \
DSAR 2-25 Rev.14 I
. l
_ _ _ _ - = _ _ _ _.___- .- ___ - _ __ - __ L MYAPC-TABLE 2.2.1 WIND DIRECTION VERSUS WIND SPEED - buRING FOG CONDITIONS - DIRECTION 0-2 mph 3-14 mph 15-23 mph - 24+ mph N- 2,4% 0.2% 0.1% NNE 3.5 . 0.7 0.2 NE 5.7 : 0.6
- I 1
ENE 2.2 - *
- E 2.2 0.3 ESE. 1.8 0,1 SE 2.5 0,1 i SSE 2.1 0.2 0.1 l S 8.6 - 0.4 0.1
;. -SSW 10,3 1.7 0.2 SW- - 8.1 1.3
- WSW 1.3 --*
.W 1.1
.WNW 0.5
- NW - - 0.8 -
s NNW 0.9
- 8 CALM- 39.6 TOTAL' -39.6 54.0~ 5.7 0.7 1 -
- Less than 0.1%--
O DSAR- 2-26 Rev.14 u
1,1 ' i MYAPC. TAE LE 2.2.*4 TEMPERATURE DATA FOR PORTLAh D AND BRUNSWICK NAVAL. AIR STATION Portiond Temperatures ('F)
'l .!an Feb Mar Apr May Jun - Jul Aug Sep Oct Nov Dec f Mean Daily ~
Maximum 32- 34 41 52 - 64 73 80 78 70 60 48 35 Mean Daily; Minimum 12 12 '22 32 42 51- 57 55 47 37 29 16 Nkan 22 ' - 23 .31- 42 53 62 68 67 59. 49 38 26 l Extreme - Maximum 64 64 86- 85 92 97 98 100 95 88 73 62 Extreme Minimum 21 39 21 8 23 33 40 33 23 18 6 21 Brunswick Naval Air Station Temperature ('F) Jan Feb Mar Apr - May Jun Jul Aug Sep: Oct Nov Dec-Mean 22 25- 32 43 53 62 68 66 59 49 40 27
' The mean number of days with temperatures >90*F or.<32'F for Por'Jand is as -
follows: Portland Temperatures. Jan Feb Mar - Apr May Jun' Jul - Aug Sep Oct Nov- Dec Ann Above-90*F :0: 0 0 0 *'
.1 '2 2
- O -0 0 6 Below 32*F - 30 27- 27 15 3 0 0 0 2 9 20 29 162 l
- More than'0 but less than 1 DSAR 2-27 Rev.14
- __ _ _ _ _ - . _ . _ . . _ _ _ _ _ _ . _ , _ _ . . _ _ . _ _ _ _ , . _ _ _ . . . . - . . ._. . . _ _ _ ~ _ . _ . _ _ . _ _ .
O MYAPC TABLE 2.2.3 PRECIPITATION STATISTICS BRUNSWICK [ Mean No. Of Days with Mean No, Of Days with Mean onth 0.10 inches or More 0.50 inches or More (Inches) k Dec 8 3 4.56 Jan. 9 3 4.40 Feb 8 3 4.27 WINTER 25 9 4 Mar - 7 4 5.28 Apr 7 3 3.97 May 7 3 3.59 SPRING 21 10 June 5 2 2.64 t July 6 2 2.79 Aug 5 2 3.00 SUMMER 16 6 Sept 6 2 3.81 Oct 6 3 3.78 Nov 7 3 -4.51 FALL 19 8 ANNUAL 81 33 46.59 + 1 4 DSAR 2-28 Rev.14
g MYAPC TABLE 2.2.4 PRECIPITATION STATISTICS PORTLAND Mean No. of Mean No. of Extreme Extreme l Days with 0.10 Days with Monthly. Monthly Maximum inches or More 0.50 inches Mean Minimum Maximum in 24 Hrs Month or More (Inches) (Inches) (Inches) (inches) Dec 6 3 3.85 0.98 6.56 3.46 Jan 7 3 4.37 0.76 9.41 3.57' , Feb 6 3 3.80 1.56 6.38 3.21 WINTER 19 9 Mar 7 3 4.34 0.81 9.97 3.48 Apr 6 3 3.73 0.71 6.48 3.82 May 6 2 3.41 0.49 7.74 4.86 SPRING 19 8 June 6 2 3.18 0,70 5.94 4.96 July 6 2 2.86 0.61 5.87 4.23 Aug 5 2 2.42 0.27 8.30 4.18 SUMMER 17- 6 Sept 7 2 3.52 0.30 9.81 7.49 Oct 6 3 3.20 0.26 12.27 7,71 Nov 6 3 4.17 2.10 9.71 3.76 FALL 20 8 ANNUAL 75 31 42.85
; o DSAR 2-29 Rev.14
MYAPC
\
TABLE 2.2.5 PORTLAND. MAINE 19401965 - SNOWFALL IN INCHES Jan Feb Mar Apr May Jun - Jul Aug Sep Oct Nov Dec Ann Average Monthly 18.9 19.0 14.0 2.6 0.3 0.0 0.0 0.0 T 0.2 2.9 ' 13.1 71,0 Snowfall Maximum - Monthly 35.3 35.3 46.6 11.2 7.0 0.0 0.0 0.0 T 3.6 13.4 35.5 46.6 Minimum Monthly - 3.6 0.3 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.6 - Maximum 24 Hr, 14.2 20.8 11.5 6.1 '7.0 0.0 0.0 0.0 T 3.6 8.5 21.0 21.0 Number of Days >1.0 5 5 4 1 0 0 0 0 1 3 20.0 inches l, i
-( T - Trace j
* -Less than one-half inch
(' DSAR- 2-30 Rev.14 '
T
-MYAPC O-- _ _
TABLE 2.2.6 WIND DATA FOR BRUNSWICK NAVAL AIR STATION Max Winds (1)(from hourly observations) Peak Gusts (2) Month Direction Speed (Knots) Direction Speed (Knots)
-Jan NNW. 36 NNW - 51 Feb' S 43
-f W' 59=
Mar - NW. 40 NW 53 Apr NW 35 WNW - 50 ' May NNE 32 W 50
-Jun NW 30 NW- 68 -
Jul NW~ 28 NW 45 Aug SSW 56- S 73 h Sept: S- 32 -N~ 72
- Oct SSE~ 31 SE. _48
'Nov SSW 36 -SSE' 55
-Dec - SSW -. -35 'SSE 52
- (1) Observed wind taken from hourly observations.
(2) Peak gusts taken from recording instruments. DSAR' 2-31 Rev.14
, MYAPC 'd TABLE 2.2.7 P_ORTLAND. MAINE 1940-1965 - MAXIMUM WIND SPEEDS Month Speed (mph) Direction Year Jan. 50 SE 1951 Feb. 58 N 1952 Mar. 76 NE 1947 Apr. 57 S 1946 May 49 NW 1950 June 37 E 1947 July 44 W 1941 Aug. 69 E 1954 Sept. 62 SE 1960 Oct. 45 N 1963 (m
(y\ ! Nov. 76 NE 1945 Dec. 62 SE 1957 b'kl DSAR 2-32 Rev.14
. . m , _. _ . . ._ _,. _ - . _ . . . _ . . _ . . _ . . _ . _ _ _ . _ _ . . _ . _ _ _ _ _ _ . _ _ _ _ _ _
4 L i, MYAPC I' g, ,
' TABLE 2.2.8 -
g ENVIRONMENTAL MONITORING PROGRAM - PROFILE MEDIA Number of ? Sample Type - Sampling Frequency Required Analyses Sample Locations Air Particulate Weekly Gross-beta 5 Quarterly Composite -. Gamma spectroscopy ' Charcoal Filter Weekly 1131 5
- Ground Water
- Quarterly H 3, 2 7 i gamma spectroscopy Estuary Water -
Montnly Ccmposite Gamma spectroscopy .2 , Quarterly Composite H-3 I Direct Radiation . Quarterly Integrated gamma 38 dose 3 $ [ *
- Groundwater samples shall be taken when this source is tapped for drinking or irrigation -
j- purposes in areas where hydraulic gradient or recharge properties are suitable for ' !- contamination, p 1 4 6 4 p o 5 I ~ DSAR 2-33 Rev.14 L i-
p .MYAPC TABLE 2.2.9 - RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM - FUNCTIONAL MEDIA - Number of Sampling Sample Sample Type Frequency Required Analysl Locations Milk Monthly Gamma spectroscopy 3 1131 Food Crop
- Monthly Gamma spectroscopy 3-(3 types of broad leaf I-131 l vegetation)
Sediment Semiannually Gamma spectroscopy 2 Fish and Invertebrates Semiannually Gamma spectroscopy 2 (2 commercially or. or in Season recreationalimportant . species)
' Perfo med only if milk sampling is not done.
O- DSAR 2-34 Rev.14 tl *
/ MYAPC TABLE 2.2.10 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LOCATIONS unstance t-rom utraction Station Designation Location Plant (km) From Plant Air Particulate and Charcoal Filter ,
AP/CF-11 Montsweag Brook 2.7 NW AP/CF-13 Bailey Farm (ESL) 0.7 NE AP/CF-14 Mason Steam Station 4.8 NNE F/CF-16 Westport Firehouse 1.8 S AP/CF-29 Dresden Substation ~20.1 N Estuary Water WE-12 Plant Outfall 0.3 SW
~WE-20 Kennebec River 9.5 SW Ground Water WG-13 Bailey Farm (ESL) 0.7 NE WG-24 Morse Well 9.9 W l Fish and Invertebrates
. /' FH/MU/CA/HA-11 Long Ledge Area 0.9 S !' FH/MU/CA/HA-24 Sheepscot River 11.1 S Sediment SE-16 Old Outfalt Area 0.6 S l SE-18 Foxbird Island 0.6 S Milk TM 15 Mitman Farm 5.5 S TM-18 Chewonki Farm 1.9 WSW TM-25 Hanson Farm 18.3 W Direct Radiation TL-1 Old Ferry Road 0.9 N TL-2 Old Ferry Road 0.8 NNE TL-3 Bailey House 0.7 NE TL-4 Westport Island , Rt.144 1.3 ENE TL-5 MY Information Center 0.2 E TL-6 Rt.144 and Greenleaf Rd 1.0 E TL-7 Westport Island, Rt.144 0.9 ESE TL-8 MY Screenhouse 0.2 SE TL-0 Westport Island, Rt.144 0.8 SE n V DSAR 2-35 Rev.14
, MYAPC
.'ABLE 2.2.10 (continued)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LOCATIQNS Distance From Ulf ection - Station Designation Location From Plant Plant (km) Direct Radiation (continued) TL 10 Bailey Point 0.3 SSE TL 11 Mason Steam Station 4.8 NNE TL 12 Westport Firehouse 1.7 S TL 13 Foxbird Island 0.3 SSV! TL 14 Eaton Farm 0.7 SW TL 15 Eaton Farm 0.8 WSW TL 16 Eaton Farm 0.7 W TL 17 Eaton Farm Read 0.0 WNW TL 18 Eaton Farm Road 0.8 NW TL 19 Eaton Form Road 0.9 NNW TC20 Bradford Road, Wiscasset 6.4 N TL 21 Federal Street, Wiscasset 7.1 NNE j ,q TL 22 Cochran Road, Edgecomb 8.3 NE j ) TL 23 Middle Road, Edgecomb 6.4 ENE TL 24 River Road, Edgecomb 7.8 E
% 3'~' River Road, and Rt. 27 7.7 ESE TL 26 Rt. 27 and Boothbay RR 7.9 SE Museum X 27 ~ Barters Island 7.2 SSE TL 28 Westport Island, Rt.144, and 7.9 S East Shore Road TL 29 Harrison's Trailer 6.2 SSW T L-30 Lerman Farm, Woolwich 7.8 SW TL 31 Barley Neck Road, Woolwich 6.8 WSW TC'2 Baker Farm Woolwich 7.3 W TGM Rt.127 , Woolwich Y.4 WNW TL 34 Rt.127. Woolwich 7.9 NW TL-35 Rt.127, Dresden 9.1 NNW TL 36 Boothbay Harbor Pre Station 12.2 SSE TL-37 Bath Fire Station 10.7 WSW TL 38 Dresden Substation 20.1 N \
_DSAR 2-36 Rev.14
. J'
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O\ V FIGURE 2.2-6 Direct Radiation Monitorinp. Loc *ttions Within 12 Kilometers of Maine Yankee
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4 4rt4vrre eg.w r 4' / 0 in 10 81LCHETtas ggt S U f!GURE 2,:.y U LOC 3tions t* de . lomete s T02 flaine Yankee
i l A MYAPC 2.3 Hy.dtelegy 2.3.1 Surface Hydrology l The site is located at the south end of a peninsula (Bailey Point) which is bount'ed on the east by l Back River and on the west by an arm of Montsweeg Bay. These two contiguous bodies of water, together with Hockomock and Knubble Bays on the south, are a part of the Sheepscot River estuarial system. Hockomock Bay is connected to the Kennebec River on the west by the Sasanoa River. The principal hydrographic features of this estuarial system are shown on Figure 2.31. t Surface drainage in the area is generally from no1h to south, roughly parallel to the strike of the regional bedrock formation. Runoff averages about 50% of the total rainfall on an annual basis, but this ratio varies seasonably from a maximum of 140% in April to a minimum of 10% in August. Due to the moderately sloping terrain, nearly complete vegetational cover and some natural storage, runoff is not excessive and dry period flows are not unusually low. The principal streams in the vicinity of the site are the Sheepscot River and Montsweag Brook, which have the following watershed areas and approximate flows: Q Drainage Area, Discharge,. Square Miles cfs (19381960) (Head of Tidewater) Maximum Average Minimum Sheepscot River 190 6,750 307 6.4 Montsweag Brook 10.8 394 18 0.4 Runoff from the power station site is conveyed by the underground storm drain system or flows overland directly to Back River or to the cove west of Bailey Point. The fresh water discharge of the Sheepscot River and Montsweag Brook are small when compared with tidal movement of the receiving estuarial waters, and for this reason, have no significant effect on tide levels, tidal flows, or water temperature. The Wiscasset town water system and Montsweag Brook is the source of fresh water supply for industrial usage at the plant. For this purpose, a storage reservoir having a usable capacity of f
-- about 185 acre feet has been built at a point on this stream about three fourths of a mile above v DSAR 2-44 Rev.14
q MYAPC ticewater and two miles northwest of the plant site. Water is pumped to the plant through an 8-inch bued pipeline Natural runoff plus draft on storage provides a firm flow of 200 gpm for a period of 26 weeks during the driest period on record. 2.3.2 Oceanographic Features The Back River extends in a northerly direction from a point known as Long Ledge, which is at the northem limit of Montsweag Bay, a distance of about 4 miles, to a confluence with the Sheepscot River at the northem tip of Cushman Point. It varies in width from a maximum of 1,500 feet at Berry Island to a minimum of 500 feet at Cowseagan Narrows. Channel depths vary between 10 to over 60 feet at mean low water, with a maximum depth at the plant site of approximately 36 feet. Montsweag Bay extends southward from Back River in the vicinity of Long Ledge a distance of about 4 miles to Phipps and Hubbard Points, where it connects with Hockomock Bay, Montsweag t Bay varies in width from approximately 2,000 feet at its northem and southern limits, to about 8,000 feet midway between these points and has mean tide level area of about 1,800 acres. Except for j a relatively narrow central channel, the bay is quite shallow, with mean low water depths generally less than 2 feet. Accordingly, extensive intertidal mud flats are exposed at low tide and especially
;VO so during spring low tides. The central channel varies in depth from 13 to 50 feet. Montsweag Brook enters the bay from the northwest. Tidal flows enter and leave the Back. River Montsweag
} Bay area at the Cowsoagan Narrows on the north and through the passage separating Phlpps and Hubbard Points to the south, in order to comply with the August 23,1972, Department of Environmental Protection Order, which imposed a 25 acre mixing zone on Maine Yankee's discharge, the Cowseagan Narrows Causeway which connected Westport Island to the mainland was removed in November 1974 after a high level bridge was constructed in its place. As a result of this action, natural circulation that existed prior to 1950 has been restored. The net southerly flow into and out of Montsweag Bay through Cowseagan Narrows has increased greatly, thereby inducing currents upwards of 3 to 4 knots (5 to 7 fps). Tidal currents at selected stations are published annually by the National Ocean Survey (NOS) of the National Oceanic and Atmospheric Administration (NOAA). Predicted time differences, velocity ratios and maximum currents based on values taken at Portsmouth, New Hampshire, are given in Table 2.3.1 for 4 stations in the area, in addition to changes in currents, the tidal range has increased both in Montsweag Bay and Back River. Data collected since the causeway was removed show a mean tidal range of 9.44 feet. This DSAR 2-45 Rev.14
MYAPC l' value coincides closely to the value of 9.1 feet measured in 1943 by the U.S. Coast and Geodetic Survey (now NOS). Mean low tide values are now reported at 4.57 feet below mean sea level. Such a change represents a 0.6 to 1 foot reduction in mean low water. It has also been calculated that as a result of the increased water level fluctuations, approximately 270 391 additional acres of Intertidal flats are now exposed at low tide. The tidal datum for the area is 4.5 feet below mean sea level, measured at Portland, Maine. Tidal height and time differences for a number of stations situated throughout the area are referenced to this Portland data (Table 2.3.2). It should be pointed out that the above differences and ranges are for astronomical tides on!/. Astronomical tides are frequently affected by meteorological I conditions which must be considered separately. The location of the various current and tidal stations throughout the area are shown in Figure 2.31. The monthly and annual estimated mean and mean maximum ocean water temperatures in the area, based on temperatures published by NOS for Portland and Bar Harbor, Maine, are tabulated below to the nearest half degree Fahrenhelt. n Mean Mean l i Month Mean Maximum Month Mean Maximum ( January 33.5 37.0 August 59.0 62.0 February 32.5 34.5 September 57.5 61.0 March 34.5 38.0 October 52.0 56.0 April 40.0 44.0 November 45.5 49.5 May 47.0 52.0 December 38.0 42.5 June 53.5 58.0 July 58.0 62.0 Annual 46.0 The above temperatures are for coastal waters and estuaries exposed directly to the ocean. Water temperatures in confined estuaries such as Montsweag Bay and Back River average a degree or two higher due to the increased heating of the relatively shallow waters and more widely exposed beaches and flats. ,A \ U) DSAR 2-46 Rev.14
- , . - . L_ .. * -- _ ._ w - -
A MYAPC (v) 2.3.3 Probable Maximum Flood 2.3.3.1 Maximum Water Surface Elevation An investigation was made to predlet the probable maximum flood level which could occur at the site of the Maine Yankee Atomic Power Station on the Sheepscot River estuary when the probable maximum hurricane is taken as the design basis meteorological event. The investigation is based upon the parameters of the probable maximum hurricane as defined by U.S. Weather Bureau Report HUR 7 97, Interim Report . Meteorological Characteristics of Probable Maximum Hurricane, Atlantic and Gulf Coasts of the United States (Reference 1). This investigation shows that the maximum water levels at the Maine Yankee Power Station due to the probable maximum hurricane are predicted to be at Elevation 19.9 feet and Elevation 21.4 feet on the plant site and screen well structure, respectively. These levels are based upon the simultaneous occurrence of the maximum storm surge, maximum predicted astronomical tide, an Initial rite in mean sea level, estuarian amplification, the probable maximum flood in the Sheepscot watershed, maximum waves in Montsweag Bay and existence of a channet res'riction at the former
,- 3 Cowseagan Narrows Causeway.
I \ Removal of the causeway and mplacement with a bridge increases the degree of conservation of this original work because the causoway acted as a dam to the hurricane surge and caused " water pile-up" at its face, This dam effect is included in the maximum water elevation mentioned above. As a result of the causeway removal, the surge resulting from the probable maximum hurricane can now travel well up the estuary, thereby lowering the maximum water levels at the plant site. Thus, the maximum water levels presented above are higher than what could occur during a probable maximum flood caused by a probable maximum hurricane. Safety measures have been implemented in the design of the plant regarding this design basis flooding. The t.creen well is protected up to Elevation 22 feet 0 inches, while the floor grade of the principal power station structures is Elevation 21 feet 0 inches. There will be no significant risk of flood at the site since the minimum shore protection Elevation of 20 feet 0 inches and site grade which varies from Elevation 20 to 21 feet should preclude water from entering. The maximum probable flood flow in the Sheepscot River (Figure 2.3 2) was determined by means of a triangular hydrograph method (Reference 2). This results in a flow at Wiscasset of 126,500 cfs (Figure 2.3-3) Due to its vast size, the water levels in the Sheepscot Bay will remain essentially O'0 DSAR 2-47 Rev.14
MYAPC unaffected by this flood flow. Therefore, a backwater curve was calculated for the Sheepscot River from the open coast to Wiscasset, assuming conservatively that none of the flow passes through Back River but that the water surface at the plant site will be the same as Wiscasset, it was assumed that the highest water levels would occur if the peak of the runoff coincided with the peak of the astronomical high tide and storm surge Uniform flow was assumed with an n= 0.03, and the area and hydraulic radius at Doggett Castle, as taken from USC&GS Chart No. 314, was used as the mean river section. This resulted in a water slope of 0.035 feet per mile, which is equal to a total increase in water levels of only 0.4 feet. 2.3.3.2 Wave Runup and Wave Forces Wave runup on the slope above still water level is dependent on the roughneri and porosity of the material composing this slope as well as period and height. As given in Reference 3 a sandy slope is considered smooth while a rubble mound structure or a riprap covered structure is considered rough. The slope for which wave runup was determined is covered with trees and brush. Since the trees will break up and retard the waves in the same manner as rubble, the wave runup was determined using a slope roughness equal to the average of smooth and rough as shown in i Reference 3. Figures 2.3 4,2.3 5,2.3 6, and 2.3-7 give plans and profiles for the plant and surrounding area. Wave runup from the significant wave at the Maine Yankee site was determined in the probable maximum hurricane to be 5.11 feet on the slope south of the turbine building and 6.68 feet on the circulating water pump house. Varying the period of the significant wave affects the extent of wave __ runup on the structures, As seen on Graph B, on Figure 2.3-8, the wave runup on the slope increases with greater wave period while the wave runup on the screen well increases with shorter wave period. This condition is due to the wave runup at the screen well being a standing wave while the wave runup on the slope.ls due to a breaking wave. The: probability of any wave occurring in a spectrum of waves can be determined using Bretschneider's Joint Distribution as found in Reference 3. Assuming that the design wave occurs during the period of 2.2. hours when the winds in the probable maximum hurricane exceed 110 mph, approximately 1,840 waves can occur. This period of maximum winds we define as the
- critical period" of the probable maximum hurricane. Using Bretschneider's Distribution, the
. distribution of waves with a height of 5.63 feet and a varying period is shown in Graph A of Figure
. 2.3 8. This graph shows that 0.6% of the waves that occur when the winds exceed 110 mph have a period of 1.7 to 2.6 seconds; accordingly, the wave runup on the circulating water pump house DSAR' 2-48 Rev.14
MYAPC V or slope as plotted in Graph B would have the same frequency of occurrence. Accumulative frequency of wave runup during the " critical period" has been obtained by plotting the sum of the Individual frequencies as shown in Graph A versus the corresponding wave runup as shown by Graph B. This accumulative frequency of wave runup is shown in Graph C. ! The wave runup frequency curve for the pump house shows that approximately 10% of the waves during the period of maximum winds would result in a wave runup equal to or greater than 6 feet. For reference purposes, design wave runup for both the slope and the pump house is Indicated in Graph C. Wave runup equal to or greater than the design runup would occur during approximately 4% of the " critical period." Should any wave runup occur that would overtop the slope or pump house, the flow rate due to overtopping would be such that the site pump house could drain, it is not considered credible that these waves would be consecutive in the wave spectrum. Provisions are made in the design of the circulating water pump house to accept the splash and flow of wave runup to Elevation 22 feet 0 inches. This is accomplished with plates which seal the Interior deck. For water to overtop the 12 inch kick plate / floodgate system inside the building, water would have to stand above that elevation on the outside of the building; however, this is not possible. Only one door of the pump house will be exposed to the runup, in addition to the precautions designed into the structure, the exposed door can always be sandbagged if runup is d a problem. Equipment located in the circulating water pump house and the structure itself have been designed to be unaffected by the maximum water height described. Pump seals, castings and auxillary piping all are instead with due consideration of the maximum water height. An investigation has been conducted to determine the forces on the outside curtain wall and the supporting floor for the service water. This investigation includes the dynamic forces of a breaking wave on the structure, The first part of the calculation assumes a wave train approaching the front face of the structure at a 30 degree angle. This angle is considered the flattest angle at which a wave of 5.63 feet could be generated due to the estuarial geometry, We have conservatively assumed the wave break at the structure. The Minikin formula (Reference 3), as modified by the approach angle and wallinclination, was used to calculate the static and dynamic forces on the exterior curtain wall of the structure. The water levelin the in3rior of the structure was assumed to oscillate with the same period and ainplitude as the incident wave. The dynamic force of this free surface rising to strike the underside of the pump well floor was calculated by determining the momentum change in the vertical direction. The underside of the pump well floor is flat, and thus offers no resistance to horizontal fluid flow, which is the direction to which the velocity is converted once impingement occurs. A factor of two design margins between the strength of the wall and the force of a breaking wave was found.
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m\
\ DSAR 2-49 Rev.14
c MYAPC An investigation has been carried out to examine the characteristics of waves that could be generated across Back River from the east. Significant wave heights of 3.6 feet or less were i predicted for conditions of a 2,000 foot fetch length, a sustained wind speed of 110 mph, and 4 different water surface elevations ranging from 7 feet to +15 feet (MSL). The Thijsse and Schljf ; method (Reference 4) formed the basis of these computations, Each of the four wave heights was i examined according to wave breaking criteria summartzed in Reference 4 and found to be smaller than a breaking wave. Consequently, none of the waves generated under these conditions could 4. break. Therefore, they pose no threat to the structuralintegrity of the screen well curtain wall. ' The new Fuel Oil Storage Facility is well back from the river bank, Wave run up on tha slope will ! drain back between waves and not flow around the fuel oil facility's two doors. The doors sills are i at Elevation 25 feet,1 inch, which is 5 feet,1 inch, above the highest calculated run up. One of two vent pipe centerlines penetrates a wall at Elevation 35 feet,71/4 inches, while the other vent pipe centerline penetrates a wall at Elevation 35 feet, g inches. The two fill pipe centerlines, which have locking caps, penetrate a wall at Elevation 21 feet, O inches. The fill pipe penetrations are grouted and sealed. All other buildings are even more remote from the river edge and less susceptible to flooding from any cause. Nevertheless, they all are capable of being sandbagged if conditions indicated the i need, Computations for these improbable combinations of events show that no structural damage or jeopardy to the plant would result. Even if the waves could break against the pump house structure with appreciably more energy, the curtain wall would not fall catastrophically, but might crack and be held in place by the reinforcing steel. The curtain wallis not required for the Integrity of the remainder of the structure, 2,3,3,3 Extreme Low Water In support of original service water system design a study of the most severe combination of tide influencing conditions was made by coupling the effects of a low t,pring tide and the winds of the probable maximum hurricane blowing from a northerly direction rather than southerly,
. The low spring tide level is 5.2 feet below MSL at Lower Hell Gate. Wind setup calculations proceeded from the method summarized in Reference 3. An additional change in the water surface of 2.6 feet was predicted which brings the water surface elevation to .7.8 feet at the service water .
V DSAR 2 50 Rev.14 i _ . . _ . . ,. , , . . ~ , , , . . _ _ . _ _ , . , . _ - . . . . _ _ . . . _ . . , , , . _ . . .-
O MYAPC I l V pumps. The design extreme low river water levelis minus 7 feet at the service water pumps. This level has been experienced occasionally during the plant operating history. The recommended minimum submergence of the service water pump suction bell to prevent vortex formation and flow i interruption is 4 feet. The elevation of the service water pump suction bell mouth is minus 14 feet, l 4 inches. Assuming a maximum lovel drop of 0.5 feet across the traveling water screeris and the design extreme low river water level of minus 7 feet, the minimum service water pump suction bell submergence would be 6 feet,10 inches which exceeds minimum recommended submergence with a margin of about 3 feet. 2.3.4 Ice Loading, Oil Spill and Debris Blockage The intake structure is protected by a concrete curtain wall which extends across the face of the intake structure and from Elevation 21 feet 0 inches down to 7 feet 0 inches as shown in Figure 5.4-6. The curialn walls provide protection against floating objects causing damage to the water intakes. Submerged objects could damage the intake trash racks or restrict flow though one or more of the intake channels, (q j One service water pump is required to support spent fuel pool cooling, it is highly improbable that submerged objects would exist in the "ack River which could simultaneously block all intake channels. Ice will not form at the intake ' innels to a depth to obstruct intake flow. Frazile ice will i not form to block the intake flow. The Central Maine Power Company Mason Station has three oil storage tanks located at Pirch Point in Wiscasset. One tank has the capacity of 100,000 barrels of oll; the other two har capacities of 132,000 barrels each, making a total oil storage of 364,000 barrels at the site, it is most unlikely that any oil contained in these tanks could reach the water and have an influence on the safe operation of Maine Yankee. All of the tanks are enclosed by dikes which would contain any oil coming from any tank within the dike area, Manually-operated valves " - the bottom of the dikes for the purpose of draining off water are maintained in the closed position. Should a tanker spill oil during delivery, there is the possibility that oil could advance toward the area of Maine Yankee. However, the Maine Yankee intake structure provides a curtain wall which projects 7 feet below the mean sea level. This 'Yould prevent oil or surface borno debris from entering the service water system. Therefore, it is extremely unlikely that an oil spill of any kind will cause an operational problem at Maine Yankee. V DSAR 2-51 Rev.14
p MYAPC t
'v') 2.3.5 Groundwater Hydrology Groundwater in the region occurs as free groundwater within the clay silt soll mantle and joints in the underlying bedrock. The peninsula on which the plant is located widens and increases in elevation toward the north so that the general groundwater movement in the area is from north to south, with a gradual shift toward the east and west in the direction of the adjacer,t tidal waters.
Gradients are expected to be roughly parallel to the surface topography. Percolation rates are low due to the low permeability of the local soils and limited bedrock Jcinting. Water wells in the area are either dug wells, usually less than 25 feet deep, or drilled wells penetrating the bedrock for depths of 100 feet or more. Such wells are for domestic or farn i use. Although adequate for the purpose, they seldom exceed 5 or 10 gpm for short term pumping and l even less with sustained pumping. Since the yield of a dug wellis sensit;ve to groundwater levels, those in unfavorable locations can dry up during drought periods. The drilled wells are consistently reliable because water table levels have little effect on their yield. There are no municipal or other important well water supply systems in the area. p) Precipitation at the power station site will percolate downward to the water table and then move with the normal groundwater flows toward the adjoining salt water areas. All wells within the 2,000 foot exclusion radius are controlled by Maine Yankee Atomic Power Company. Section 2.3 References
- 1. U.S. Department of Commerce,1968, Interim Report Meteorological Characteristics of the Probable Maximum Hurricane, Atlantic and Gulf Coasts of the United States, Weather Bureau Memorandum HUR 7 97.
- 2. U.S. Department of Interior,1965, Design of Small Dams, Bureau of Reclamation.
- 3. U.S. Army,1966, Shore Protection, Planning and Design, Coastal Engineering Research Center, Technical Report No. 4.
- 4. A. T. Ippen,1966, Estuary and Coastline Hydrodynamics, McGraw-Hill.
l ( i O DSAR 2-52 Rev.14
,- MYAPC ' v' TABLE 2.3.1 CURRENTS Time Differences
- Velocity Maximum Currents auos' Flood Ebb Station Slack Max Max Max Direction Average Direction Average Water Current Flood Ebb Velocity (True) (True) Velocity l ,
- h. m. h. m. deg. knots deg. knots Sheepscot River, off 1 00 0 50 0.7 0.6 5 0.8 200 1.1 Batter Island 2 Lower Pt., NE.
I of Sasanoa 0 45 0 10 1.4 1.0 325 1,7 150 1.8 River ,7 3 3 Lower Holl t,
', ) Gate, Knubble 0 35 +0 20 2.5 1.9 200 3.0 155 3.5 Bay 4 Upper Hell Gate, Sasanoa " "
0.8 0.5 305 1.0 140 0.8 River Based on Portsmouth Harbor entrance. Flood begins, +3 h 30m; maximum flood, +2 h 50 m; ebb begins, +1 h 20 m; maximum ebb, +2 h 05 m.
- 2. Georgetown due south of Westport South tip, as Lowe Point which is Southeast of Sassnoa
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DSAR 2-53 Rev.14 l
p MYAPC TABLE 2.3.2 TIDES Differences Based on Portland Data Time Height Ranges
- High : Low High Low - Mean Spring Mean Tide Station Water Water Water Water feet feet Level h m. h. m. feet feet feet i Isle of Springs Due East, -0 0; -0 01 0.1 0.0 8.9 10.3 4.4 South and Westport across Sheepscot Cross River Entrance . +0 07 +0 04 +0.1 0.0 9.1 10.5- 4.5 North of Barlers Island Wiscasset +0 16 +0 04 +0.4 0.0 9.4 10.8 4.7 O Sheepscot (below rapids) 3 Back Rivers .
+0 20
+0 34
+0 20 -
+0 31
+0.6 -
+0.1 0.0 0.0 0.6 9.1 11.0 10.5 4.8 4.5 East . Barters latand Between Arrowsic &
Georgetown >
, z Vic Hant Robinhood, Sasanoa +0 14 +0 14 0.2 00 8.8 10.1 4.4 River South end Knubble Bay Mill Point, Sasanoa River - +0 35 +0 43- 0.2 0.0 8.8 10.1 4.4 South End to Sasanoa Upper Hell Gate, Sasanoa - +1 11 +1 31 2.0 0.0 7.0 8.0 3.5 River N
DSAR- 2 54 Rev.14
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O MYAPC 2.4 Gsdyg2 Site geogrr.phical information was developed from study of the available literature, surface reconnais:ance and field study, and examination of samples of soll and rock core taken from borings on the site. Refroction eelsmic surveys were also made to determine the rock surface elevations. References 1,2 and 3 contaire detailed reports on these r,tudies. Twelve borings were made on the site in the initial studies and these have buer' supplemented by 12 more borings in the southem portion of the site where the plant is located (Figure 2,4-1). Overburden in thh area consists of medium soft to medium stiff silty clays with occasional sandy lenses and pebb'y sands. Overburden varies 15 to 20 feet in thickness. Bedrock is of Silurian-Devonian age characterized by steeply dipping schistose rocks of the Cape Elizabeth formation, interlayered with lenticular masses of granite and coarse crystalline pegmatite. Joints in the bedrock, as they appear in outcrops and cores, are medium spaced, ranging from 1 to 5-foot intervals and less. Variation from quartzite to mica schist, to granite and pegmatite, is a common geologic feature. Though the mica schist is relatively weaker than the other rocks, it is fresh, sound, and not materially weathered. Both borings and geologic surveys show the presence Ci of tight folds, mostly in the schistose formations. No major faults have been recorded in the area h during the site studies or in previous studies by other geologists. Two minor faults, long since healed, have been recorded in the outcrops along the shore. They are located approximately 900 - feet northeast of the plant site. These do not show any appreciable brecciation or gouge formation. The studiea show that, geologically, the site is suitable for an atomic power plant. No major or active faults have been detected or . ire suspected in the vicinity of the site. Beorock is sound, within reach and provides good foundation support for the structures and equipment. The major structure.s ere founded directly on hard, crystalline bedrock. Minor structures are founded either on rock or on compacted granular fill above the rock. The seismic surveys at the site show average values of compression wave velocity of 13,000 to 15,000 fps and shear wave velocity of 7,000 fps. From these, values of Poisson's ratio of 0.33 and Young's modulus of 5 x 108 psi were calculated. From work by Eristov, 8th Intemational
- Conference on Large Dams, Young's modulus for the bulk rock would be equal to 5.0 x 10' = 2.9 x 10' psi and the shear modulus,1.1 x 10' psi.
1.7 The values of Young's modulus range from 4.94 x 10' psi to 5.67 x 105 psi. The shear modulus DSAR 2-63 Rev.14
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l 1 1 MYAPC varies from 1.80 x 10' to 2.06 x 10' psi. Variation? of this amount in the modull have been shown 4 to have no effoM on the containment vibration rpodes.
. Section 2.4 References 1.- " Geological Considerations, Maine Yankee Atomic Power Project," John Rand, Consulting Geologist, November 15,1967.'
, .2. " Seismic Survey, Proposed Nuclear Power Plant, Bailey Point, Wiscasset, Maine," Weston l Geophysical Research, Inc., November 4,1966. 3, " Seismic Survey, Maine Yankee Atomic Power Project," Weston Geophysical Research, i . Inc., June 1967. i i-1 $-( i t i 5 1
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, MYAPC V 2.5 Salamoloav 2.5.1 Tectonics Despite rather complex interrelations between rocks of various ages and types, New England is generally characterized by competent, unweathered bedrock and stable geologic conditions. The regional bedrock geology, and known major faults and fault zones, are shown on Figure 2.5-1. The nearest significant fault is located about 75 miles to the west. In the geological investigation, two small inactive faults were found about 900 feet from the structures.
Since Triassic time,180 million years ago, no major geologic changes other than those produced by glaciation have occurred. The southern part of Maine, in vicinity of the site, is composed of three general types of Paleozoic rocks, North of the area, the bedrock consists of consolidated early Paleozoic (350 million years old) sediment which has been metamorphosed through intense folding. Some middle and late Paleozoic (250 to 350 million years old) igneous intrusives are also present. During the late Paleozoic Era (250 to 300 million years old), sedimentary material was deposited and lithified in a series of basins extending from the Narragansett Bay area of Rhode Island through eastem Massachusetts into southwestem Maine almost as far north as the site area. , O The bedrock of southem Maine was last affected by orogenic movements at the close of Paleozoic V time by the Appalachlan orogeny. Recional Seismicity Seismic activity in New England is small and typified by infrequent shallow focus earthquakes of low magnitude and intensity. Historical records of New England earthquakes date back 300 years; more than 30 years of instrumental data exist for the New England area. Earthquakes of a Modified Mercalli Intensity VI, or greater, for the northern New England area are shown on Figure 2.5-2. This plot shows a total of 15 earthquakes which have occurred since the middle of the 18th century, with epicenters within about 150 miles of the site. The largest of these earthquakes have been previously assigned a Modified Mercalli Intensity of Vill. There were three such earthquakes, all of which occurred before 1800. Historical records of earthquakes in this area, especially the older ones, must be evaluated with considerable caution. Much of colonial construction was of poor quality, and this is especially true for chimneys. Population centers tended to be clustered along and at the mouths of rivers on soft, recent alluvium and, in many cases, epicenters have been assigned to population centers. The records are of variable quality, a few indicating careful observation, but many showing obvious exaggerations. g A review indicates the intensity assigned to these earthquakes is questionable. There is evidence DSAR 2-66 Rev.14 1
MYAPC 3 that all the intensities were less than Vill, in one case probably VI, and in the two other cases, probably VI or Vil. Instrumental observations over a period of more than 30 years have indicated 4 areas in New England where concentrations of shocks have been noted. These are the Milo, Maine area, the v Ossipee-Lake Winnipesaukee area of New Hampshire, the Cape Ann Massachusetts Bay area, and Augusta, Maine. The closest earthquake to the plant site was a magnitude 4.0 mb earthquake that occurred on April 17,1979, approximately four miles from the site. The largest earthquakes to occur in New England since the installation of seismic instrumentation occurred on December 20 and 24,1940, in the Ossipee, New Hampshire area. Both of these earthquakes were of Intensity Vil with a Richter magnitude of 5.8, which is the largest magnitude determined for any New England earthquake. Magnitude determinations for most of the larger New England earthquakes range between 4 and 5 on the Richter scale. The St. Lawrence River Valley, lying more than 200 miles north of the site, is an area of significant tectonic activity. Earthquakes occurring along the St. Lawrence are felt throughout New England. A number of large earthquakes of Intensities Vill, IX, and X have occurred in the St. Lawrence f'} River Valley. Nearly all of these earthquakes occurred early in colonial times. The most recent of V these occurred on February 28,1925, and was Intensity IX. Reports of this earthquake in the site area Indicate that its intensity was IV, although at Brunswick, Maine the intensity may be estimated t as high as V. An earthquake on October 20,1870, also of Intensity IX, reportedly broke a few windows (Intensity V) at Portland, Maine. Seismicity of the Wiscasset Area Seismicity of the southem Maine area surrounding Wiscasset is shown in Figure 2.5-3. This figure shows the approximate epicentrallocation of all earthquakes on record which had epicenters within 75 miles of the site. Nearby small earthquakes include the earthquake at Brunswick in 1881, approximately 15 miles from the site, which had an epicentral intensity of approximately IV or V, based on historical data, with a probable Intensity IV at the plant site, and the April 17,1979, magnitude 4.0 mb event about 4 miles from the site with a site intensity of about V. The earthquake of April 26,1957, located 25 miles to the south, was offshore, but has been estimated to have had an epicentral Intensity of VI. All of the other earthquakes which have occurred within a 20-mile radius of the site were of Intensity lli or less. DSAR 2-67 Rev.14 i
MYAPC The site may be expected to be subjected to earthquake vibrations from relatively nearby, local-quakes of small magnitude or from major earthquakes along the St. Lawrence River Valley. Based on observed earthquake attenuations for this portion of the United States, as shown on Figure - 2.5-4, an Intensity X earthquake in the nearest portion of the St. Lawrence River Valley would result in an intensity at the site of approximately V, A number of studies have been made relating maximum ground acceleration and earthquake intensity. As shown on Figure 2.5-5, that given by TID.7024 (Curve H)is the most conservative. These studies have largely been based on experience outside of New England. However, recent comparisons of instrumentally determined magnitudes and intensity in small earthquakes in this region have aprced well with the relations given by TID-7024. Accordingly, these relationships, although based primarily on soil-supported structures, have been accepted as reasonable and conservative for small earthquakes of the range ofintensities anticipated. Based upon a relatively long historical record, on the attenuation of earthquake intensity with distance in the northeastern United Stctes and eastem Canada region, on the geology of the area and on the character of the component bedrock on which the facility is founded, an estimate of the probable maximum intensity at the site is a Modified Mercalli Intensity V to low Intensity VI, corresponding to a ground acceleration of 0.04g. b 2.5.2 Tsunamis it is considered that tsunamis would not have any measurable effect on this site. The only tsunami on record on the East Coast is from the Grand Banks earthquake of November _18,1929, 'or which a water level height of 1/2 inch was noted at Atlantic City. There was little or no tsunami effect to the Maine coast from this earthquake, nor has there been a record of tsunami effect from any other earthquake. DSAR 2-68 Rev.14
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I ( t MYAPC SECTION 3.0 FACILITY DESIGN AND OPERATION TABLE OF CONTENTS Section Iltla Eaga 3.1 De sig n C rite ria . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.1.1 Conformance with 10CFR 50 Appendix A Gc*eral 1 Design Criteria 3.1.2 Classification of Structuros, Systems, and Components 3.1.2.1_ SSCs important to the Defueled Condition 3.1.2.2 Wind, Missile, and Tomado Loadings j '3.1.2.3 Water Level (Flood) Design ! 3.1.2.4 Seismic Design 3.2 Structures................................................................................... 3-28 3.2.1 Fuel Building 3.2.1.1 General 3.2.1.2 Fuel Unloading Area >.b ~3.2.1.3 New Fuel Storsge 3.2.1.4 Spent Fuel Pool 3.2.1.5 Fuel Storage Racks 3.2.2 Storage Buildings 3.2.2.1 Underground RCA Storage Bunker 3.2.2.2 Radiation Controlled Area (RCA) Storage Building 3.2.2.3 LSA Storage Building 3.2.2.4 Warehouse 3.2.2.5 Low Level Waste Storage Building 3.2.3 Service Building 3.2.3.1 Control Room Area 3.2.4 Turbine Building 3.2.5 Primary Auxiliary Building 3.2.6 Service Water intake Structure 3.2.7 Fire Pump House 3.2.8 Masonry Walls DSAR 3-1 Rev.14
i
, M).JC
\
L SECTION 3.0 FACILITY DESIGN AND OPERATION TABLE OF CONTENTS Section Illla ' East 3.3 Systems 3.3.1 F uel Storage. . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-5 2 3.3.1.1 Design Basis 3.3.1.2 System Description 3.3.1.3 Design Evaluation 3.3.1.4 System Operation 3.3.1.5' Monitoring and Instrumentation 3.3.2 Fuel Handling System.. . ...... ... .... ... . .. .. .. . . .... . ... . .... .. . ... .. ... ..... . . . . . . . . . 3-89 3.3.2.1 Design basis 3.3.2.2 System Description 3.3.2.3 Design Evaluation {m i 3.3.2.4 Inspection and Testing 3.3.2.5 ~ Monitoring and Instrumentation 3.3.3 Primary Compcaent Cooling Water.............................................. 3295 3.3.3.1 Design Basis -
'3.3.3.2' System Description 3.3.3.3 Design Evaluation 3.3.3.4 System Operation 3.3.3.5 Monitoring and Instrumentation 3.3.4 - Service Water System..... ..... .. . . ... .. .. ... . .. . .. . ..... . . . . . .. .... . .... . . ....... ... 3-103 3.3.4.1 Design Basis 3.3.4.2 System Description 3.3.4.3 Design Evaluation
-3.3.4.4 System Operation 3.3.4.5 Monitoring and Instrumentation DSAR 3 ii Rev.14
MYAPC ( SECTION 3.0 FACILITY DESIGN AND OPERATION TABLE OF CONTENTS Section Illla Eagn , 3.3.5 Ventilation Systems...... . ..... .... . .. . . .... . ........ ........... . ...... . . . . ... . . ... ... . .. 3 108 3.3.5.1 Fuel Building Ventilation System 3.3.5.2 Control Room Ventilation System 3.3.5.3 Auxiliary Ventilation Systems 3.3.5.3.1 Turbine Building Ventilation 3.3.5.3.2 Primary Auxiliary Building Ventilation
-3.3.6 Auxilia ry Systems. .... . . . .. . . .. . . . . .. . .. . . . . . .. . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . 3- 1 15 3.3.6.1 Compressed Air l
3.3.6.2 Boric Acid Makeup l 3.3.6.3 Primary Water System p 3.3.6.4 Primary Vent And Drain System ( 3.3.6.5 Radioactive Waste Processing System 3.3.6.6 Fire Protection System 3.3.6.7 Meteorological Instrumentation 3.3.7 Ele ctrical Syste ms. . . . . . . . . ..... ... . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . .. . . . . . . . . . . . . . . .. 3- 13 7 3.3.7.1 Offsite Power 3.3.7.2 Onsite Power 3.4 Control of He aw Loada. . . ... . . . . . . . . . . . . . . . . . . .. . . . . , , . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . .3-147 ........... DSAR 3-iii Rev.14
- . --...-. - - . _ .~ .. - . - --. - - - . - . - - - . - - _ .~ - . - - .-. _ . - - - - . ...
1 l MYAPC SECTION 3.0 ' f FACILITY DESIGN AND OPERATION LIST OF TABLES l ) 1 Table No.- Illig ' 3.1.1 Earthquake Damping Factors 3.3.1 Compressed Air System - Equipment Description and Ratings i I l 1 1 i t i l 1 l i 7 l 1 j t i.
- DSAR 3-iv Rev.14 1
1 2 i k.
l MYAPC SECTION 3.0 FACILITY DESIGN AND OPERATION LIST OF FIGURES Flaure No. Illig
- 3.1 1 Response Spectra for 0.05g Maximum Ground Acceleration 3.1 2 Response Spectra for 0.10g Maximum Ground Acceleration 3.2-1 Fuel Building Arrangement 3.2-2 Fuel Building Arrangement 3.2-3 Service Building 3.2-4 Service Building 3.2-5 Main Control Room 3.2-6 Turbine Building - Elevation 21' 3.2-7 Turbine Building - Elevation 39' 3.2-8 Turbine Building - Operating Floor 3.2-9 Primary Auxiliary Building Arrangement
, p, 3.2-10 Primary Auxiliary Building Arrangement V 3.2-11 Service Water intake Structure -
3.2.12 Fire Pump House 3.2-13 Fuel Transfer Tube
.3.3-1 Typical Fuel Assembly 3.3-2 Typical Fuel Spacer Grid ;
3.3-3 Control Element Assembly 3.3-4 Control Element Assembly , 3.3-5 Spent Fuel Pool Assembly Placement Limitations 3.3-6 High Dens;ty Spent Fuel Pool Layout For The Two Region Pool 3.3-7 Fuel Pool Cooling System
-3.3-8 Refueling Equipment Arrangement. ,
3.3-9 Primary Component Cooling Water 3.3-10 Primary Component Cooling Water 3.3-11 Primary Component Cooling Water 3.3-12 Service Water System 3.3-13 Compressed Air System 3.3 Compressed Air System
'v -DSAR 3-v Rev.14
. ---. _ ___- _... _ . _..._ . . _ _ . - . . _ . _ _ _ _ _ _ - . _ - . _ _ . ~ . _ _ _ - _ . _ _ . . _ _ . _ . =-
MYAPC SECTION i - FAC!LITY DESIGN AND OPERATION 4 LIST OF FIGURES Figure No. . Title 3.3-15 Compressed Air System - Turbine Building 3.3 16 Compressed Air System - Turbine Building 3.3-17 Compressed Air System - Auxiliary Building ) 3.3-18 Compressed Air System- Auxiliary Building . 3.3 19 Primary Vent And Drain System 3.3-20 Primary Vent and Drain System
- 3.3 21 Fire Protection System 3.3 22- Fire Protection System 3.3-23 One Line Diagram - Transmission and Utility interconnections with MY 3.3-24 One Line Diagram - Auxiliary Power System 3.3-25 One Line Diagram - 125 Volt DC Emergency Buses and 120 Volt AC Vital Buses Q
i O DSAR 3-vi Rev.14 NY
MYAPC SECTlON 3.0 FACILITY DESIGN AND OPERATION Section 3.0 discusses the design and operation of the structures, systems, and components I required to safely store fuel, it also discusses supporting systems such as ventilation and auxiliary systems used to safely store fuel or support decontamination and decommissioning activities. 3.1 Design Criteria 3.1.1 Conformance with 10CFR 50 Appendix A General Design Criteria in July of 1967, the Atomic Energy Commission issued the proposed general design criteria for nuclear power plants. These 70 criteria were '4 sued for comment by the industry but had not yet been adopted as a regulatory requirement. Nevertheless, as the following d;scussion shows, the Maine Yankee plant has been designed and constructed in accordance with the intent of these criteria. In the following paragraphs, each criterion is stated and its conformance indicated. On September 18,1992, the USNRC confirmed that plants with construction permits issued prior to O V May 21,1971 were evaluated on a plant-specific basis, and backfitting the current General Design Criteria of Appendix A to 10CFR 50 would provide little or no safety benefit. (Reference 1, SECY-92-223) On August 7,1997,' Maine Yankee certified in accordance with 10CFR 50.82 that the company had permanently ceased power operation and that all fuel was removed from the reactor vessel (Reference 2, MN-97-89). With the docketing of these certifications, the Maine Yankee license no longer allowed operation of the reactor or placing of fuelin the reactor vessel. The AEC 1967 Design Criteria listed below are relevant to 'he permanently defueled plant condition. ANALOGOUS 1971-CRITERIA NUMBER CRITERIA Group I- Overall Plant Requirements Quality Standards 1 1 Performance Standards 2 2 Fire Protection 3 3 Sharing of Systems 4 5 Record Requirement 5 1 A ij DSAR 3-1 Rev.14
MYAPC' V Group 111 Nuclear and Radiation Controls Control Room 11 19 Instrumentation and Control Systems 12 13 Monitoring Radioactive Releases 17 64 Monitoring Fuel and Waste Storage 18 63 Group Vill - Fuel and Waste Storage System Prevention of Fuel Storage Criticality 66 62 Fuel and Waste Storage Decay Heat 67 61 Fuel and Waste Storage Radiation Shielding 68 61 Protection Against Radioactivity Release from Spent Fuel and Waste Storage 69 -- Group IX - Plant Effluents Control of Release of Radioactivity to the Environment 70 60 Group I Overall Plant Requirements O Criterion 1 - Quality Standards Those systems and components of reactor facilities which are essential to the prevention of accidents which could affect the public health and safety, or to mitigation of their consequences, shall be identified and then designed, fabricated and erected to quality standards that reflect the importance of the safety function to be performed. Where generally recognized codes or standards - on design, materials, fabrication, and inspection are used, they shall be identified. Where adherence to such codes or standards does not suffice to assure a quality product in keeping with the safety function, they shall be supplemented or modified as necessary. Quality assurance programs, test procedures, and inspection acceptance levels to be used shall be identified. A showing of sufficiency and applicability of codes, standards, quality assurance programs, test procedures, and Inspection acceptance levels used is required. (This Criterion is directly analogous to Criterion 1 of 10CFR 50, Appendix A,1971, except that the 1971 version addresses records retention. Maine Yankee record retention requirements are addressed in Section 5.10 of the Technical Specifications or the Quality Assurance Program. Design, fabrication, and construction records are addressed by Criterion 5 of this section.) O DSAR 3-2 Rev.14
n I \ MYAPC ( _./ Resoonse: Those systems and companents which are essential to the prevention of accidents which could affect the public health and safety, or to mitigation of their consequences, have been designed, fabricated, and erected to quality standards that reflect the importance of the safety function to be performed. Generally recognized codes and standards on design, matenals, fabrication, and inspection have been used. These were supplemented to reflect current practices. The descriptioris of the systems and components to which this criterion applies include the codes and other standards met by these systems. The quality assurance program is submitted to the ' regulator for review and approval of any proposed revisions which would result in a reduction of previous commitments.
References:
Sections 3 and 4 Criterion 2 - Performance Standards Those systems and components of reactor facilities which are essential to the prevention of accidents which could affect the public health and safety, or to mitigation of their consequences,
'") shall be designed, fabricated, and erected to performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tomadoes, flooding conditions, winds, Ice, and other local site effects. The design bases so established shall reflect: (a) appropriate consideration of the most severe of these natural phenomena that have been recorded for the site and the surrounding area, and (b) an appropriate margin for withstanding forces greater than those recorded to reflect uncertainties about the historical data and their suitability as a basis for design.
(This Criterion is directly analogous to Criterion 2 of 10CFR 50, Appendix A,1971, except that the 1971 version addresses consideration of appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena and the importance of the safety functions to be performed.) Resoonse: Systems and components which are essential to the prevention of accidents '.vhich could affect the public health and safety, or to mitigation of their consequences, are designed, fabricated and erected to performance standards that enable the facility to withstand, without loss of the capability (~) DSAR 3-3 Rev.14
,^ MYAPC k
to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects. The design bases so established reflect appropriate consideration for the most severe natural phenomena that have been recorded for the site and surrounding area, and appropriate margin for withstanding forces greater than those recorded to reflect uncertainties about the historical data and their suitability as a basis for design. On April 17,1979, an earthquake of approximate magnitude 4 occurred about 10 kilometers west of the plant. On January 9,1982, an earthquake of approximate magnitude 5.75 occurred in Central New Brunswick. Subsequently, the Licensee joined with the Regulator in a program to assess the seismic ruggedness of the Maine Yankee plant. This program is referred to as the Seismic Design Margins Program (SDMP). This program is fully described in NUREG/CR-4826, dated March 1987. On March 26,1987, the Regulator issued a Safety Evaluation Report which concluded that all issues associated with the seismic design were considered resolved (Ref. 3).
References:
Sections 2 and 3. !q Criterion 3 - Fire Protection () The reactor facility shall be designed (1) to minimize the probability of events such as fires and explosions, and (2) to minimize the potential effects of such events to safety. Noncombustible and fire resistant materials shall be used whenever practical throughout the facility, particularly in areas containing critical portions of the facility such as containment, control room, and components of engineered safety features. (This Criterion is directly analogous to Criterion 3 of 10CFR 50, Appendix A,1971, except that the 1971 version addresses fire protection and fire fighting considerations.) Resoonse: As described in Section 3, the materials and layout used in the station design have been chosen to minirnize the possibility and to mitigate the effects of fire. Sufficient fire protection systems and equipment have been provided to minimize the adverse effects of fire on structures, systems, and components important to safety taking into account the decommissioning plant condition and activities. (3
)
DSAR 3-4 Rev.14
,m MYAPC
/ j j Criteilon 5 - Record Reouirement Records of the design, fabrication, and construction of essential components of the plant shall be maintained by the reactor operator or under its control throughout the life of the reactor. (This Criterloa, in combination with Crlferion 1 of this section, is analogous with Criterion 1 of 10CFR 50, Appendix A 1971.) Resoonse: Records of design, fabrication, and construction of components are being maintained for the duration of the license. Design calculations are in the possession of Stone & Webster and Yankee Atomic. All other required design, fabrication, and construction information is in the possession of Maine Yankee.
References:
Section 6 and the Quality Assurance Program
- GROUP lli- NUCLEAR AND RADIATION CONTROLS Criterion 11 - Control Room The facility shall be provided with a control room from which actions to maintain safe operational status of the plant can be controlled. Adequate radiation protection shall be provided to permit access, even under accident conditions, to equipment in the control room or other areas as necessary to shut down and maintain safe control of the facility without radiation exposures of personnel in excess of 10CFR 20 limits, it shall be possible to shut the reactor down and maintain it in a safe condition if access to the control room is lost due to fire or other cause.
Resnonse: This criterion is met only to the extent that radiation exposures of personnel in excess of 10CFR 20 limits cannot occur based on the available source term from any credible accident. Sufficient time is available to allow operators to restore cooling or makeup to the spent fuel pool and maintain exposures well below the limits of 10CFR 20. Controls for spent fuel pool cooling, makeup and purification are located near the equipment and are not in the control room.
,n
References:
Sections 3,5, and 7 i / w/ DSAR 3-5 Rev.14
_ , _ . .- - . ~. - n MYAPC I \ V Criterion 12 - Instrumentation and Control Systems Instrumentation and centrols shall be provideo as required to monitor and maintain variables within prescribed operating ranges. Resoonse: Instrumentation is provided as required to monitor and maintain significant variables. Controls are provided for the purpose of maintaining these variables within the limits prescribed for safe - operation. In the permanently defueled condition, the principal variables to be monitored include fuel poollevel, and temperature. Boron concentration is determined via sampling.
References:
Sections 2,3, and 4 Criterion 17 - Monitorina Radioactive Releases Means shall be provided for monitoring the containment atmosphere, the facility effluent discharge n paths, and the facility environs for radioactivity that could be released from normal operations, from ( .
' anticipated transients, and from accident conditions, (This Criterion is analogous to Criterion 64 of 10CFR50, App. A,1971, except that the 1971 version addresses spaces containing components for recirculation of loss-of-coolant accident fluids.)
Resoonse: The n,eans for monitoring radiation levels in the spent fuel pool and concentrations on-site are provided. The sensitivity and range of this equipment is adequate for operating and anticipated transients in the permanently defueled condition. Effluent discharge paths where a potential release of radioactive material exists are monitored. Monitoring equipment has sufficient sensitivity and range for operating and anticipated transients in the permanently defueled condition, instrumentation is provided for monitoring radiation levels in the spent fuel pool area, PAB and RCA. A I \ DSAR 3-6 Rev.14
( (] MYAPC
)
L/ An environmental surveillance program has been established. The facility's environmental surveillance program provides effect!ve monitoring of radioactive material released from the plant.
References:
Section 4. Criterion 18 - Monitoring Fuel and Waste Storaog Monitoring and alarm instrumentation shall be provided for fuel and waste storage and handling areas for conditions that might contribute to loss of continuity in decay heat removal and to radiation exposuies. ResDonse: The spent fuel is monitored by an area detector with aucible and visual alarms activated at the detector location and in the control room, if the permanently installed demineralizers or waste storage tanks in the PAB and RCA storage buildings contain irradiated resin beds, they are monitored by an area detector with audible and visual alarms activated at the detector location end 7- in the control room. For waste handling operations or decontamination operations, portable l 1 detectors with audible alarms may be used to monitor rcdiation exposure. The spent fuel poolis equipped with high and low liquid level alarms, and a high temperature alarm which will indicate loss of coritinuity in decay heat removal capacity of the fuel pool cooling system. (
References:
Section 3 and 4 GROUP Vill - FUEL AND WASTE STORAGE SYSTEM
. Criterion 66 - Prevention of Fuel Storage Criticali'v Criticality in new and spent fuel storage shall be prevented by physical systems or processes.
Such means as geometrically safe configurations shall be emphasized over procedural controls. Resoons3 Criticality is prevented by geometrically safe c.onfigurations. New fuel is stored dry in a steel and concrete building at an elevation at which flooding with water is impossible. DSAR 3-7 Rev.14
p MYAPC-
' irradiated spent fuelis stored under water in a reinforced concrete cool, lined with stainless steel.
Fuel assemblies are spaced and the racks are so fabricated that criticality is precluded. Although the water in the poolis generally borated, neither soluble boron nor control rods are required to keep even unirradiated fuel assemblies suberitical. The boron concentration required as a result of the analyzed " misplaced assembly" incident is discussed in section 3.3 and 5.2.
Reference:
Section 3.3. Criterion 67 - Fuel and Wastc Storage Decav Heat Reliable decay heat removal systems shall be designed to prevent damage to the fuel in storage facilities that could result in radioactivity release to plant operating areas or the public environs. Resoonse: The spent fuel pool is designed to maintain the water level safely above the spent fuel assemblies at all times. The fuel pool outlet pipe, which serves as the suction pipe to the fuel pool cooling g pumps, is siphon protected to prevent draindown of the pool. The retum piping is located such that (] siphoning by the cooling system is limited to no less than 10 fest above the active fuel. Pool water level may be restored through diverse (offsite and onsite) supplies of fresh or demineralized water. Emergency makeup water and cooling is also available from the fire pond.
Reference:
Section 3. Criterion 68 - Fuel and Waste Storage Radiation Shielding Shielding for radiation protection shall be provided in the design of spent fuel and waste storage facilities as required to meet the requirements of 10CFR 20.
-(Also refer to Criterion 67 of this DSAR section.)
Resoonse; The radiation shielding (water) of the spent fuel pool is designed to provide adequate personnel protection. The waste storage and processing facilities are shielded as required to protect . ,a
'd
)
DSAR 3-8 Rev.14
p MYAPC personnel from exposures in excess of regulatory requirements, in addition to shleiding design, implementation of the Radiation Protection Plan requirements assures that doses to personnel performing work in these areas are maintained ALARA.
Reference:
Section 3 and 4. Criterion 69 - Protection Aaalnst Radioactivity Release from Soent Fuel and Waste Storace Containment of fuel and waste storage shall be provided if accidents could lead to release of undue amounts of radioactivity to the public environs. Resoonse: Fuel and waste storage areas are designed to preclude the inadvertent release of undue amounts of radioactivity, Considering the significantly diminished source terms in the permanently defueled condition, there are no credible accidents resulting in doses to the public approaching the 10CFR 100 limits. All spent fuel and waste storage systems are conservatively designed with ample p margin to prevent the possibility of gross mecht.nical failure which could release significant (~j amounts of radioactivity, Backup systems such as floor and trench drains are provided to collect 4 potentialleakages to preclude the release of radioactive materials to the environment. Personnel are rigorously trained and administrative procedures are strictly followed to reduce the potential for human error, in addition, radiological limits on waste storage systems are established to assure that credible accident conditions will not result in doses to the public which approach 1 rem (whole body) exposure over a 2 hour-limited duration accident. The consequences of a fuel handling incident are presented in Sections 3 and 5. In this analysis, it is demonstrated that undue amounts of radioactivity are not released to the public.
References:
Sections 3,4,5 and 7.
/ \
i i DSAR 3-9 Rev.14
MYAPC U GROUP IX a PLANT EFFLUENTS Criterion 70 - Control of Release of Radioactivity to the Environment Ths facility design shall include those means necessary to maintain control over the plant radioactive effluents, whether gaseous, liquid, or solid. Approprl@ holdup capacity shall be provided for retention of gaseous, liquid, or solid effluents, particularly where unfavorable environmental conditions can be expected to require operationallimitations upon the release of radioactive effluents to the environment, in all cases, the design for radioactivity control shall be justified (a) on the basis of 10CFR 20 requirements for normal operation and for any transient
. situation that might reasonably be anticipated to occur and (b) on the basis of 10CFR 100 dosage level guidelines for ootential reactor accidents of exceeding!y low probability of occurrence except that reduction of 'he recommended dosage levels may be required where high population densities or very large cities can be affected by the radioactive effluents.
Resoonse: The plant radioactive waste control systems (which include the liquid, gaseous and solid radwaste t,
) systems) are designed to limit the off site radiation exposure during normal operation to levels below limits set forth in 10CFR 20.
References:
Sections 4. I t.
- /~Y DSAR 3-10 Rev.14
MYAPC l O\ 3.1.2 Classification of Structures, Systems, and Components Structure Classification The p! ant structures and process systems are classified according to their function and the degree of integrity wquired to protect the public from unenntrolled releases of radioactive byproducts.
- Structural design criteria include two classes of buildings: '
- 1. Class l Structures Class I structures were designed in accordance with the " Building Code Requirements for Reinforced Concrete," ACl 318-63, including increases allowed for stresses- produced by earthquake loads in combination with other appropriate loads. Where steel is utilized, it is designed in accordance with NSC " Specification for the Design, Fabrication and Erection of Structural Steel for Buildings " A Class I structure will maintain its Integrity during the hypothetical earthquake where the combination of the normal operating loade. and the seismic stresses do not exceed 90%
of the yield strength. These structures are primarily constructed with massive reinforced concrete p and the earthquake loads are not a major factor in the design, b Class I structures and equipment are designed to remain functional during an operational basis j- earthquake (ground acceleration 0.05g) and maintain fuel pool integrity during the more severe design basis of the hypothetical earthquake (ground acceleration 0.10g). In addition, some of the Class I structures are designed so that damage will not result from tornado Winds or missiles. ! The structures, ayatems and components which have been designed to Class I seismic requirements are listed below; Reactor Containment Reinforced concrete substructure Reinfoced concrete superstructure Reactor containment fuel transfer tube Soent Fuel Pool Fuel assemblies Reinforced concrete structure, partly below ground Steel superstructure h - DSAR 3 11 Rev.14
r MYAPC ( Spent fuel storage rack Fuel building yard crane steel support structure, portion within spent fuel pool building only I
- Fuel 1;U(! ding . yard crane l'oel handling platform and hoist !
Antl41 phon device Flow Hmit6r. an the lir,er leakage detection system
- 2. Class il Structures Class ll structuros are usually designed to the requirements of the Uniform Building CWe. Class ll structures were not designed for OBE and DBE. They were designed for dead load plus live load plus wind in accordance with AISC " Specification for the Design, Fabrication and Erection of Structural Steel for Buildings" and "Bullding Code Requirements for Reinforced Concrete", ACl-318-63, Part 1%A, utilizing working stress design methods.
Where Class I structures are connected directly to the Class 11 structures, the interac3on between the structures is taken into account in the design of both, in addition, shake space between all adjacent Class I structures is provided for in the design. This rattle space is a minimum of 3 inches which le conservative with respect to the actual seismic requirements to prevent impact between
'bu%ngs in the event of a disturbance, Djamantlement of Seismically Deslaned Structures. Systems and Comoorants For SSCs formerly designed to Seismic Class I requirements but not credited for performing a Seismic Class I function in the defueled condition, the following criteria apply prior to pe forming disniantlement operatiens:
- a. Declassification of components shall be performed in accordance with appropriate engineering and design procedures and processes,
- b. When declassifying an SSC, a 10CFR 50.59 evaluation shall be performed if:
- 1) the safety classification is described in the FSAR, or
- 2) It's failure in a seismic event could affect a Seismic Class I component described in the FSAR in such a manner as to cause an unanalyzed incident or an accident with offsite doses exceeding 230 mrem (whole body) or 260 mrem (organ dose),
- c. When declassifying an SSC, a 10CFR 50.54 evaluation shall be performed if the classification is described in the OOAP.
4 DSAR' - 3 12-- Rev.14
, .- ~ _
p MYAPC
- d. SSCs shall be designed to Mimic Class I requirements if, during a seismic event, its failure has the potential tr 4 I the fuel pool water level lower than 10 feet above the active fuel.
3.1.2.1 Structures. Systems and Components important to The Defueled Condition (ITDC) General On August 7,1997, Maine Yankee certified per 10CFR 50.82 that the company had permanently ceased power operation and that all irradiated fuel had been permanently removed from the reactor vessel (Reference 2). This is a permanont, non revocable certification that changed Maine Yankee's licensing basis by no longer allowing fuel in the reactor vessel and no longer allowing power operation. The license basis for the majority of Structures, Systems and Components (SSCs) associated with nuclear safety has been changed. Those SSCs which only performed a reactor safety function (i.e., SSCs which do not support a spent fuel or radiation protection safety function) need no longer m be maintained under nuclear grade controls. I \ SSC classification involves a determination that an SSC is, or is not, safety related'. SSCs classified as safety related are treated differently by regulation than other SSCs.8 For a plant undergoing decommissioning, the only SSCs which meet the definition of safety-related8 are the spent fuel pool structure, storage racks and fuel assemblies. This results in two areas of interest: 1. safety related sSCs are those relied upon to remain funcbonal during and following design basis events to ensure: a) the integrity of the reactor molai .t pressure boundary; b) the capatdlity to shut down the reactor and maintain it in a tafe shutdown condibon; and c) the capabihty to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guidehnes of 10 CFR 100. 2. 10CFR s0 Appendit B notes that 'The pertinent requirements of this appendix apply to all activibes affecting the safety-related functions of..? ssCs. 3. The first two parts of the safety-related definition (reactor coolant pressure boundary, and capabihty to achieve and maintain safe shutdown) do not apply to a decommissioning plant. given the limnSe restricbons of 10cFR s0.82. The tW part of the safety.retated definsbon (accident consequences comparable to 10CFR 100 guidelines) also does not app y. At Maine Yankee, the consequences associated with the de86prbiscense basis events apphcable to decommissioning are nearty three N orders of magnitude lower than Part 100 guidelines and lower than the EPA protective achon guide limit. DSAR 3-13 Rev.14 -
I p MYAPC (
- 1) Maine Yankee's ' nuclear grade" processes are based largely upon quality assurance (10CFR 50 Appendix B) requirements. Reclassifying all SSCs as non safety related could lead to ihe elimination of most management controls in situations where maintaining rigorous management controls is intended,
- 2) Maine Yankee recognizes that certain functions remain important to safety in the defueled condition, it is necessary to reclassify SSCs in order to proceed with decommissioning. Strictly following regulatory requirements in reclassification will result in elimination of most of the current management controls, which is contrary to management's intent.
These concerns are addressed by introducing an artificial classification scheme that goes beyond regulatory requirements: SSCs which support a fuel safety or radiation protection safety function are designated as important to the Defueled Cond$on (ITDC). s . Enhanced management ccntrols are maintained on SSCs classified as ITDC. The current license basis for ITDC SSCs is reviewed and revised if appropriate using the applicable change mechanism (e.g.,10CFR 50.54,50.90,50.59, etc). This includes their safety related or non safety related designations. SSCs which cre not ITDC are eliminated from the license basis. The following criteria are used to determine which SSCs are designated as ITDC: Criterion 1. Is the SSC associated with storage, control, or maintenance of nuclear fuel in a safe condition; or handling of radioactive waste? This incuder direct as well as indirect effects.* Criterion 2. Is the SSC associated with Radiological Safety?'
- 4. Includes the following:
- Those ssCs required to safely store and handle radioactive waste and spent fuel, a
Those ssC's required to protect Wers and the public from the consequences of the remaining (or new) DBA's.
- Ybose Ssc's required to safety store new fuel until it is shipped offsite.
* . 'tse systems which rnonitor or control radiological effluent release paths.
Ch 5. Preserve ssCs associated with maintaining exposures ALARA. DSAR 3-14 Rev.14
l MYAPC Criterion 3. Is the SSC associated with an outstanding commitment to the regulators which remains applicable to storage, control, or maintenance of nuclear fuel in a safe condition; or handling of radioactive waste? Criterion 4. Does the SSC satisfy a requirement based in regulations? This includes any SSC which is Independently required by Technical Specification.' l A positive response to any criterion Indicates that an SSC is ITDC. A negative response to all l criteria identifies a non-lTDC SSC. When applying these criteria to determine the effect on an SSC, the scope of the criteria below must be carefully applied. Authorizations. Restrictions and Limitations on use of the SSC redamalfication criteria. The SSC reclassificatN criteria will be used as a basis to change various Maine Yankee processes, provided that the change involves an SSC that is non-lTDC and, provided that plant procedures contain an acceptable method for approving the change. The following kinds of
" software" changes associated with non-lTDC SSCs are allowed:
s e _ SSC classifications
. drawings,
.- calculations, e procedures
+- nonconforming items and corrective actions (Learning Bank) e extemalindustry operating experience repons e
commitments e open work orders (in process at the time the decision was made to decommission the plant)
+
the application of 10CFR 50 Appendix B criteria provided it does not represent a reduction in
. commitment.
Use of these criteria does not authorize;
- a. - Activities creating new hazards or initiators not already recognized as part of the current license basis (e.g., decontamination or decommissioning of major components defined in 10CFR 50.82) 6.
TNs evaluaton assures that the appropriate regulatory change mechanism is used for effecting the change This indudes dianges associated with the Securtty Plan, operatonal Quality Assurance Program. Ernemency Planning, operator Training, O Permits, Ucense Condibons, rules, etc. DSAR- 3 __ Rev.14
j - MYAPC )
- b. The physical removal / disassembly of existing SSCs, or the Installation of new SSCs.
l However, it may provide the basis for initiating a hardware change. I
- c. Changes to Technical Specification requirements applicable to the current mode of operation.
- d. Changes to regulations, license conditions, rules, and permits until such time that relief is granted from the regulating authority. However, it may provide the basis for requesting relief from the regulations, license conditions, rules, and permits,
- e. Changes to commitments. Application of the commitment change process is required to change commitments.
- f. Changes to the OOAP. However, it may provide the basis for initiating a change to the l OQAP.
- g. Changes to the ODCM. However, it may provide the basis for initiating a change to the ODCM.
- h. Changes to the Emergency Plan. However, it may provide the basis for initiating a change to the Emergency Plan,
- l. Changes to the Security Plan. However, it may provide the basis for initiating a change to the Security Plan.
J. . Changes to the Fire Protection Plan. However, it may provide the basis for initiating a change n to the Fire Protection Plan. i k. Changes to the Radiation Protection Program. However, it may provide the basis for initiating a change to the Radiation Protection Program. So_undaries and Interfaces for ITDC SSCs SSCs identified as ITDC that require
- availability", must meet the following criterion:
*A system, subsystem, train, component, or device is "available" or will have " availability" when it i; capable of performing its specified function (s)."
Implicit in this definition is the assumption that the necessary attendant instrumentation, controls, power sources or equipment, or other auxiliary equipment that are required to support the available SSC, are capable of performing their support function, as necessary. Enaineered Reauirements for ITDC SSCs A higher level of quality is maintained for ITDC components to assure that the capability exists to reliably meet the performance expectations and requirements. The controlled list of ITDC
]
DSAR 3 16. Rev.14
l MYAPC components is governed by plant procedures. ITDC components are not safety related components and are not required to satisfy 10CFR50 Appendix B requirements. Although not required by regulation, the following criteria is developed and applied, as appropriate, to ITDC SSCs to assure continued reliability:
- a. Design Control Measures will be invoked to assure applicable regulatory requirements Ilcense basis, and design basis information is correctly translated into specifications, drawings, procedures and Instructions. These measures shall include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled. Design changes, including field changes will be subjected to engineered design control measures commensurate with the importance of the SSC.
- b. Procurement Document Control Measures will be invoked to assure that applicable regulatory requirements, design basis, and other requirements which are necessary to assure adequate quality are suitably included or referenced in the documents for procurement of material, equipment, services, q c. Instructions, Procedures, and Drawings Activities affecting SSCs will be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and will be accomplished in accordance with these instructions, procedures, and drawings. Instructions procedures, and drawings will include appropriate quantitative or qualitative acceptance criteria for determining that
. Important activities have been satisfactorily accomplished,
- d. Control of Purchased Material, Equipment, and Services Measures will be invoked to assure that material, equipment, and services conform to the procurement documents. These measures shall include provisions, o s appropriate, for source evaluation and selection, objective evidence of quality furnished, inspection at the source, and examination upon delivery,
- e. Inspection Inspection of activities affecting quality will be invoked and executed to verify conformance with the documented instmetions, procedures, and drawings for accomplishing the activity.
.DSAR 3-17 Rev.14
)
MYAPC U f. Handling, Storage and Shipping Measures will be invoked to control the handling, storage, shipping, cleaning and preservation of material and equipment in accordence with work and inspection instructions to prevent damage or deterioration,
- g. Test Control Surveillance testing will be established for SSCs to ensure that the SSCs perform satisfactorily commansurate with the importance of its intended safety function.
- h. Measuring and Test Equipment Appropriate controls will be invoked to assure that measuring and test devices used on SSCs are properly controlled, calibrated and adjusted at specified periods to maintain accuracy within necessary limits,
- l. Corrective Action Measures will be invoked to assure that conditions adverse to quality are promptly identified and corrected, in the case of significant conditions adverse to quality, the measures will
,q , assure that the cause of the condition is determined and corrective action is taken to preclude ( / repetition. 3.1.2.2 Wind, Missile, and Tornado Loadings The Maine Yankee facility is capable of withstanding the effects of severe winds or tornadoes without loss of capability of the safety systems to perform their safety functions. Section 2.2.6 discusses wind and tornado data for this region. The design tornado has a rotational velocity of 300 mph, a velocity of advance of 60 mph, and an external vacuum of 3 psig developed in 5 seconds. Thus the total effective velocity is 360 mph. Missiles may travel with the tornado equivalent to (1) a utility pole 35 ft long,14 inches in diameter, weighing 50 lb/cu. ft. (1850 lbs total), and traveling 150 mph, or, (2) a 1 ton automobile traveling at 150 mph. The structures housing spent fuel and adjacent structures are designed to resist the combined effects of tomado wind load, pressure drop and missile loads to produce the most critical loading condition. The original design code for the spent fuel pool and other Class I structures was ACI 318-63. The allowable stresses for shear and flexure are defined by the criteria in sections 1600 and 1700 of ACI 318-63. This specifically included the appropriate capacity reduction factor and d DSAR 3-18 Rev.14
f ew MYAPC t V) allowable stresses. The Fuel Building is designed for protection against wlad and tornado as follows: Reinforced concrete structure Steel superstructure Fuel building - yard crane steel support structure, portion within spent fuel pool building only The 6 foot thick reinforced concrete walls, which extend from 12' 6" below ground grade to 26' above ground grade, are designed to withstand the effects of tornado and missiles. A very substantial degree of added tornado and missile protection is afforded by the below-grade construction of the fuel pool. The nominal grade level (20' elev) is approximately at the same elevation as the top of the active fuel. The steel framing above the poolis designed for tomado loadings such that it will not fall into the pool and damage fuel assemblies. The loss of water from the storage pool under the effect of a tornado would result in a maximum calculated loss of five feet of water, primarily due to vortexing. Given that the normallevel of the spent fuel pool is at the 44 ft. elevation and approximately 23 ft above the active fuel, this water g loss is not significant as it is bounded by the siphoning incident. The fire pump structure is also designed to withstand the effects of wind and tornado. The fire pumps, located at ground grade near the 1,500,000 gallon fire protection reservoir are tornado protected and screened from the full effects of the tornado by the dike. This is a very reliable makeup source of water to the pool. 3.1.2.3 Water Level (Flood) Design 3.1.2.3.1 Hurricane An investigation was made to predict the probable maximum flood level which could occur at the site of the Maine Yankee Atomic Power Station on the Sheepscot River estuary when the probable maximum hurricane is taken as the design basis meteorological event. The investigation is t,ased upca the parameters of the probable maximum hurricane as defined by U.S. Weather Bureau Report HUR 7 97, Interim Report Meteorological Characteristics of Probable Maximum Hurricane, Atlantic and Gulf Coasts of the United States and discussed in section 2.2.6.
/'N i
DSAR 3-19 Rev.14
o MYAPC (\ This investigation shows that the maximum water levels at the Maine Yankee Power Station due to the probable maximum hurricane are predicted to be at Elevation 19.9 feet and Elevation 21.4 feet on the plant site and scmen well structure, respectively. These levels are based upon the simultaneous occurrence of the maximum storm surge, maximum predicted astronomical tide, an initial rise in mean sea level, estuariar amplification, the probable maximum flood in the Sheepscot watershed, maximum waves in Montsweag Bay and existence of a channel restriction at the formor Cowseagan Narrows Causeway. 3.1.2.3.2 Snow Section 2.2.5.1, Table 2.2.5 shows average snowfall statistics for Portland which are considered to be representative of the site. Structuralloading and capacity reduction factors for buildings required in the defueled condition provide ample margin to accommodate snow loading. I l 3.1.2.3.3 Ice, Glaze, and Temperature Extremes leo loading is discussed in Sections 2.2.5 and 2.3.4 of this report. Glaze and Ice storms usually q occur in the months October through April with an averago frequency of 1 to 3 storms per year, l ) Ice thicknesses of .25 inches or .5 inches will likely occur every year whereas .75 inches is likely to occur at least every three years. Structuralloading and capacity reduction factors for buildings required in the defueled condition provide ample margin to accommodate ice loading. The circulating water intake structure is used as the intake structure for both service water and circulating water systems. The intake structure is protected from ice as it will not form at the intake channels to a depth to obstruct intake flow. Circulating water provides no spent fuel storage cooling or makeup functions. It may be used, if elected, to dilute planned and controlled waste discharges. The service water system performed a safety class function during power operation but does not support a safety class function in the defueled condition. The service water does provide the normal heat sink function for the primary component cooling water system and one service water pump is more than adequate to support spent fuel pool cooling functions. However, in the event that service water is not available, attemate cooling methods can be employed to effect pool cooling. As discussed in section 2.2.4, the average January temperature is about 22*F with between 10 and 20 days cf sub zero temperaturcs occurring yearly. Temperature data representative for the site is provided in Table 2.2.2. During extended periods of freezing temperatures,it is possible that [n) DSAR -3 20 Rev.14
MYAPC freeze damage could occur to water filled piping in buildings no longer maintained due to the defueled condition. The primary concem regarding freezing is the affect on the in egrity of the pool. Pool water freezing is not possible provided that the aggregate decay heat load of the stored assemblies is reasonably high. Assurance is provided through routine operator rounds which monitor and log the temperature of the pool water. Adjacent buildings and rooms containing i significant water sources which could potentially affect the safe storage of fuel are either ; temperature controlled to preclude freezing, or the water source is appropriately heat traced or ; dra;ned, as necessary. This includes adequate administrative or design controls for protecting the integrity of the fuel transfer tube. 3.1.2.4 Selsmic Design 3.1.2.4.1 Design Basis All structures and elements of the plant are designed in accordance with sound engineering practice and are considered capable of withstanding seismic forces corresponding to a ground acceleration of at least 0.03g, in addition to normalloads, without damage or loss of function. In addition, all structures and components of the plant which are important from the standpoint of nuclear safety and damage which could affect the health and safety of the public, i.e.," Class l" portion, are declgned to meet the following criteria:
- 1. The design earthquake is based on a ground acceleration of 0.0Sg, and this portion of the plant shall be capable of operating through such earthquake.
- 2. The hypothetical earthquake is based on a ground acceleration of 0.1g, and this portion of the plant shall be capable of performing its intended safety function under an earthquake.
- 3. A spectrum analysis is used, with appropriate conservative damping factors.
3.1.2.4.2 Design Data While 0.04g was found to be the maximum probable ground acceleration at the base of the structures on the site, it was decided to round this upward to 0.0Sg horizontal for design purposes. The values for seismic design of Class I structures and components are as stated under " Design Bases." In Class I structures and components, stresses due to normal loads plus the design
- earthquake do not exceed those design values permitted in the applicable codes, while stresses P ' DSAR -
3 21 Rev.14
l 1 MYAPC due to normalloads plus the hypothetical earthquake do not exceed the yield stress of the affected materials. Earthquake stresses are based on a horizontal ground acceleration and a vertical ground acceleration of two thirds of the horizontal, with the two acting simultaneously. Design for Class I structures and components used the "rssponse spectrum" approach in the analysis of the dynamic loads imparted by earthquakes. The seismic design is based on the acceleration response spectrum curves shown in Figure 3.1 1 for the design earthquake and Figure . 3.12 for the hypothetical earthquake. The curves are derived from the "Housner Spectrum" I normalized to 0.0$g for the design earthquake and 0.10g for the hypothetical earthquake. , i l The design response spectra (Figures 3.1 1 and 3.12) are a specification of the level of seismic ( design acceleration, or displacement, an a function of natural period of vibration and damping level.
; The response spectrum analysis is applied to all category I structures and components and groups thereof whose responses may be interdependent, considering their natural period and using appropriate damping factors as listed on Table 3.1.1. >
Class I structures and components are designed in the following general manner:
- 1. An analysis is made to determine the natural periods of vibration of the structure using equivalent lump mass systems or distributed mass systems as is considered appropriate, in these analyses, periods and mode shapes are determined for each lumped mass mode. These data then define participation factors for each structure. Where structures are supported on their own foundations, foundation displacements are considered in determining natural periods and participation factors. It should be noted, however, that Class I structures at this site are founded on granite gneiss. Accordingly, foundation yielding will be very small and may in many cases be neglected without introducing significant error. ;
- 2. The earthquake design acceleration value for the specific natural period of the structure or component being considered is determined from Figure 3.11 using appropriate damping factors. The horizcatal component of the ground acceleration is taken directly and the vertical component is taken as two-thirds of the horizontal value. These components are considered as acting simultaneously, f
- 3. For certain structures, and especially for vibratory systems of a highly complex nature, such as a piping system, ese of the maximum response value (peak of the curve) corresponding to the appropriate damping factor may be elected in performing the stress analysis of the system.
O, DSAR- 3 22 Rev.14
i MYAPC O 4. A tabulation of typical damping factors which are used for various vibratory systems important to nuclear safety is presented in Table 3.1.1. Conservative values are shown for various materials, methods of construction, and location with respect to the ground.
- 5. The design is then checked to verify that stresses are within acceptable limits for the hypothetical earthquake using Figure 3.12.
Refer to section 3.1.2 for a listing of SSCs required to be designed to Seismic Class I requirements in the defueled condition. SSCs which were previously designed to Class I criteria but are not listed in section 3.1.2, are not credited for performing a safety function (Class I ) in the permanently defueled condition, and therefore, are no longer required to be designed to Class I requirements. 3.1.2.4.3 Seismic Design, Qualification, and Instrumentation As of March 24,1988, new Class I systems, structures, and ccaponents will be designed and qualified to the seismic demands as defined by a 0.18g NUREG/CR-0098 50th percentile Ground Response Spec,tra (GRS). Analytical qualification will be to the SEP allowable stress levels and o damping values listed in Reference 4 (i.e., for piping, damping s 3% or PVRC, allowable stress = 2.4Sh, no OBE) up to any Interface with existing structures, systems, and components. Seismic adequacy m6y also be demonstrated 'brough similarity by comparison to the documented performance of equipment in natural earthquakes (Reference 5), or simulated earthquakes on testing machines. Maine Yankee's original seismograph installation consists of two triaxial strong motion accelerographs. A vertical seismic trigger, set at approximately 0.01g will have the device operating fully within 0.1 seconds of the initial"P" wave and will continue to operate 10 seconds after the last motion of a seismic event is detected. Each device is operated by a salf contained battery supply and contains enough 70mm photographic film to operate for 25 minutes. One accelerograph is located on bed rock adjacent to the containment base mat at elevation .16'- 9". The other instrument is bolted to a heavy shelf which in turn is bolted to the outside of the reactor containment wall at approximately elevation 34'. The two instruments will be in the same vertical plane. The vertical separation of 50' constitutes a significant fraction of the structure's height. DSAR 3-23 Rev.14
w+1an 3.1 References!
- 1. Memorandum; SECY 92 223 Resolution of Deviations ;dentified During The Systematic Evaluation Program: S J. Chilk, Secretary, USNRC to J.M. Taylor, Sept.18,1992.
- 2. MY Letter to the NRC M +97-89
- Certification of Cessation of Power Operation and Permanent Removal of Fuel from the Reactor, dated August 7,1997.
- 3. Letter: " Seismic Design Margins Program," P.M. Sears, USNRC to J.B. Randazza, Maine Yankee Atomic Power Company, March 26,1987; NMY 87 029
- 4. ~ USNRC Letter to MYAPCO, dated March 26,1987.
- 5. USNRC Letter to MYAPCO, dated February 19,1987, (Generic Letter 87-02).
O
- -DSAR: 3 24 Rev.14
MYAPC \p \v) TABLE 3.1.1 DAMPING FACTORS Percent of Critical Damping DESIGN HYPOTHETICAL EARTHQUAKE EARTHQUAKE Reactor Containment 2.0 5.0 Reinforced concrete structure, other than 2.0 5.0 containment, founded on soll or rou, Reinforced concrete supporting structure, not 2.0 0.0 founded on soil or rock Steel framed structures, including supporting structures and foundat;ons Bolted or Riveted 3.0 5.0 Welded 1.0 2.0 o Reactor vessel, internals and control rod drives Welded Assemblies 1.0 1.0 Bolted Assemblies 3.0 3.0 Mechanical equipment, including pumps, fans and 2.0 2.0 similar items Piping Systems 1.0 2.0 \
'v DSAR 3 25 Rev.14
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i MYAPC
\ 3.2 Structures "
3.2.1 Fuel Building 3.2.1.1 - General The principal function of the fuel building is to provide a location for the safe storage of new and spent fuel assemblies. The building houses a new fuel unloading area, a new fuel storage room, a spent fuel pool and the necessary cranes required for the handling of the fuel assemblies. The spent fuel pool cooling system heat exchanger, the fuel pool cooling pumps and the fuel pool purification pump are locatea in the area adjacent to the solid weste disposal equipment. The equipment decontamination area and the spent fuel pool support systerr.s are also located in the fuel building. The fuel building arrangement is shown on Figures 3.21 and 3.2 2. ; 3.2.1.2 Fuel Unloading Area e New fuel was shipped to the site in two element shipping casks. A five ton overhead crane in the fuel building was used to unload the shipping casks. The spent fuel pool purification system filters >
/ and domineralizer are located in shielded cubicles below the fuel unloading area. Shield slabs are ,
removed from the fuel unloading floor to replace expended filter cartridge elements. 3.2.1.3 New Fuel Storage Area
- The new fuel storage room is designed for storage of 160 fuel assemblies. The fuel room is located over the solid waste disposal area and the spent fuel pool cooling pumps and heat exchanger. The fuel rack consists of guide sleeves symmetrically located on the floor at Elevation ,
31 ft.1 1/2 in, and through the ceiling of the new-fuel room at Elevation 44 ft 6 in. The fuel room floor has a drain opening located over the spent fuel pool cooling equipment cubicle. The floor opening prevents flooding of the new fuel storage area. The spent fuel pool new fuel elevator winch is also located in the new-fuel storage room. 3.2.1.4 Spent Fuel Pool Cooling of the spent fuel assemblies during the radioactive decay period is accomplished in a stainless steel lined reinforced concrete pool filled with borated water. Space is provided in the . pool to place the spent fuel shipping cask. The pool is serviced by means of the yard crane, as well t DSAR - 3 28 Rev.14
MYAPC as a moveable platform with holst. A new fuel area adjoins the spent fuel pool. The pool is designed to safely resist the hypothetical earthquake or tomado, as well as the applied loads of the water and fuel. The pool has a reinforced concrete floor founded on rock and sidewalls 6 feet thick which extend from 12 feet 6 inches below ground grade to 26 feet above ground grade. The concrete is reinforced with #11 bars at 12 inch center to center spacing with a yield strength of 40,000 psl. The concrete has a 28 day minimum compressive strength of 3,000 psl. The reinforced spent fuel pool was originally designed in accordance with ACl 318-63 to resist the appropriate dead, live, hydrostatic and maximum hypothetical seismic loadings. The structure was reanalyzed, in support of EDCR 92111, to demonstrate the acceptability of installing the new high density spent fuel storage racks. As part of the preliminary decommissioning activities, the structural evaluations have been performed which demonstrate the adequacy of the SFP concrete and liner to withstand the effects of dead, live and hydrostatic forces in conjunction with an elevated pool water temperature of 212'F. Complete details of this evaluation era contained in References 3.21 and 3.2 2. The poolis completely lined with plates of stainless steel which have test channels behind each weld. The test channels are piped to the spent resin pit sump through four 1 inch tell tale pipes, each with a flow limiter at the end of the pipe. In the event of a rupture of the liner, the combined leakage through the bedrock and the tell tale is no greater than 5 gpm. The liner is designed as a ASME Section lil, Division 2, Paragraph CC-3720, Liner, Table CC. 3720-1, Service Category, Membrane. The plate materialis ASTM A240, Type 304 stainless steel. Liner Anchors are designed to ASME Section Ill, Division 2, Paragraph CC 3730 and are constructed of ASTM A 36 steel. The weld rods used to weld the vertical stiffener flanges to the liner wall liner were ASTM E309 (carbon to stainless steel) with a minimum tensile strength of 81,000 psl. The fuel transfer tube was originally designed as safety class 2; however, since the containment integrity design basis is not applicable in the defueled condition, it has been reclassified as safety class 3. It consists of a 36 inch OD,3/8 inch thick, ASTM A312 TP304, stainless steel pipe installed inside a 40-inch OD stainless steel sleeve as shown in detail on Figure 3.2-13. The inner pipe acts as the transfer tube and connects the containment refueling canal with the spent fuel pool and is welded to the fuel pool stainless steelliner. The outer pipe is fitted with bellows expansion C t
\ DSAR- 3 Rev.14
O MYAPC
\ \
V joints, backed up by a packed slip joint to compensate for any differential movement. Structural steel supports a superstructure of protected metal siding which encloses the pool. The steel framing above the pool is designed for earthquake and tornado to prevent it from falling into the pool and damaging fuel assemblies. The masonry wall at the south end of the fuel building is not designed for certain wind or earthquake loadings, and, therefore, an evaluation of the consequences of a well collapse was performed. The analysis demonstrated adequate spent fuel pool cooling capability and structural rack integrity. 3.2.1.5 r iStorage Racks The new and spent fuel pool structures including fuel racks are designed to withstand the
- anticipated earthquake loadings as Class I structures in accordance with the guidance of Regulatory Guide 1.2g. Analyses show that the racks will perform their intended function under both seismic and load drop loadings in accordance with Rogulatory Guide 1,124 and NUREG-0800.
The design ensures that during the event, rack to-rack and rack to-wallinteraction is appropriately considered. Structural material used in the rack design is ASME Section 11, SA 240, Type 304 p stainless steel. The design considered thermalloads induced by an operating temperature of Q 154'F. Subsequently an evaluation was performed which documented the acceptability of the racks at a temperature of 212'F. The ANSYS version 4.4A program was used for all computer aided mechanical analysis, t The design considered impact loads from a fuel element dropped from 18 inches above a module, a fuel element hangup during removal, ano the load induced if an assembly hit the top of a rack while moving at the maximum horizontal velocity of the crane. The dropping of objects over the storage array was conservatively analyzed by assuming that the dropped object is twice the weight of a standard assembly. Subcriticality and a coolable geometry are maintained and damage to the stored fuelis minimized. The racks consist of individual storage cells joined into a rack module. The racks are a single tier, rectilinear array of free standing modules, not anchored to the pool walls, floor or adjoining racks. Each rack module is provided with adjustable support feet. Each fuel rack is a folded metal plate assembly of 14 gage metal, approximately 180 inches high,117 inches wide and 128 inches deep. The folded metal plate assembly is welded to a baseplate, which is supported by adjustable supported feet. Region I contains 5 racks, spaced on a minimum of 10.5 inch centers. Region ll contains 21 racks spaced on a minimum of 9 inch centers. Spent fuel storage racks may be moved q DSAR -3 30 Rev.14
p)
\
V MYAPC only in accordance with written procedures which ensures that no rack modules are moved over fuel assemblies. 3.2.2 Storage Buildings 3.2.2.1 Underground RCA Storage Bunker The underground RCA storage bunker, also referred to as the high rad bunker, is located within the protected area about 120 feet northwest of containment. The bunker is a reinforced concrete structure and is partially buried below yard grade. The bunker is 27.5 feet by 16 feet and approximately 12 feet high. The top of the bunker is about 5 feet higher than the surrounding yard grade. The bunker is divided into five internal compartments each separated by 18 inch thick concrete walls. The exterior bunker walls vary from 12 to 18 inches in thickness. The roof of the bunker consists of six removable 18 Inch thick concrete roof plugs. A floor drain system directs any liquids collected in the bunker to a sump. Liquids collected in the sump would be pumped to the spent resin pit sump in the nearby RCA storage building through an p underground pipe. The spent resin pit sump discharges to the Aerated Drain System where any Q liquids would be collected for processing. The bunker provides temporary storage for radioactive wastes before they are moved to the low level waste storage building, described in Section 3.2.2.5, for longer term storage. The outside yard crane is available to move waste containers in and out of the bunker. Waste is typically stored in the bunker for up to about one year to allow for some decay before being placed in the low level waste storage building until arrangements are made for permanent off site disposal. 3.2.2.2 Radiation Controlled Area (RCA) Storage Building This RCA storage building, adjacent to the fuel building, houses the Duratek skid waste processor, a decontamination area, and the waste solidification system. The waste solidification system includes: the waste resin storage tank (TK 85), waste holdup tank (TK 95), resin holdup tank (TK. 109), the resin transfer pump (P-144), resin holdup tank dewatering pump (P 145), and dewatering pump (P-123). Waste storage and holdup tanks are appropriately shielded. The area provides space for solidification agent stora0e and a shipping container filling station. The solidification agent is thoroughly mixed to a homogeneous mixture which is then sealed in approved shipping containers and allowed to solidify before shipment to an approved waste disposal site. This area (3 g DSAR 3 31 Rev.14 l
,o MYAPC V) ( is also used as an equipment decontamination facility. The ventilation system can be operated to maintain the waste processing and equipment decontamination area under a slightly negative pressure. 3.2.2.3 Low Specific Activity (LSA) Storage Building This building houses the LSA compactor and serves as the storage building for LSA containers to keep them free of damage or deterioration. It extends from the south wall of the RCA building to the containment, but is not structurally attached to either. The building is designed non nuclear safety. The LSA compactor compresses LSA material. It is vented into the RCA filtered ventilation system. The LSA sump pump discharges to the spent resin pit sump in the RCA Building. 3.2.2.4 Warehouse This building is located outside the RCA. The building is used to store replacement componerts and equipment. gs 3.2.2.5- Low Level Waste Storage Building The low level waste storage building is located on the plant site, outside of the protected area. The building is 154 feet by 04 feet by approximately 25 feet high from ground to roof line with floor elevation at 26 feet. The outer walls of the building consist of steel siding with one foot thick concrete shield walls inside the siding to a height of 16 feet. A truck bay is included to allow access for shipping waste containers. The building provides for interim on site storage of low level waste and for storage of contaminated equipment. 3.2.3 Service Building The service building floor plan \ showa ~ ' gures i 3.2 3 and 3.2 4. This building houses the warehouse and tool crib, the instNment y ca t offices, the secondary chemistry laboratory, and the RCA machine shops. The control room, the personnel service area, lockers, washrooms, lunchroom, are also located on the first floor of the service building. The building floor plan shows the layout of the offices and shops servicing within the service building. The RCA and secondary services are separated in such a way so as to effectively control traffic through a health physics check point. A section of the building is built up to provide two floor areas that house the cable tray and the switchgear areas at Elevation 35 feet 0 inches and 45 feet 6 inches. (n) U DSAR 3-32 Rev.14
,ew MYAPC \
v' 3.2.3.1 Control Room Area The main control room, located in the service building, contained many of the controls and instrumentation necessary to monitor and control various areas and equipment in the plant. The main control room is designed to be available at all times. Safe occupancy of the main control room during an abnormal condition is provided in the design of the service building. Adequate shielding and ventilation are used to maintain tolerable radiation and temperature levels in the main control room. The decayed source term in the defueled condition, in conjunction with the location and design of the control area, provides sufficient protection to ensure that control room personnel will not be subjected to doses which would exceed 10CFR20 limits. Equipment in this area has been designed to mlnlmize the possibility of a condition which could lead to possible inaccessibility or evacuation. In the event that this area becomes inaccessible, the controls for spent fuel pool cooling and makeup, water treatment and waste disposal, are located at control stations remote from the main control room. The main control room is shown in Figure 3.2 5. The main control room arrangement gives consideration to the fact that certain systems normally l_ require more operator attention than do others. The main control board is the centralitem in the main control room. The control board has a bench section and a vertical section. Annunciators o) are provided so that the control room operator is made aware of any deviation from normal conditions at remote control stations. Radiation monitoring system audiMe alarms are provided in the control room. Output information from the computer is displayed in the main control room for use of the plant operating personnel. The control room operators interface with the plant process computer through a standard computer terminal keyboard and color monitor. The keyboard also contains "Special Function" keys which allow the operator to access frequently used applications by depressing a single key. Other applications are accessed through a variety of menu options so that the operators only have to select an option and depress the transmit key to execute programs. Terminals of equal functionality are also located in the TSC and EOF for use by support personnel in the event of any situation which required activation of these centers. Printers are used to provide permanent operating records. One printer is used to record alarmed parameters and sequence-of-events logs (other devices provide trending capability), and one printer will print various plant logs as they are requested by the operators. Power to operate the plant computer system is supplied by a static inverter which is connected to
'd DSAR 3 33 Rev.14 1
m MYAPC
\
V one of the station batteries. The inverter is Independent of ttation service AC power sources and thus the availability of computer functions is assured during plant translents or a loss of station service power. A continuous indication of noble gas fission product concentrations in the primary vent stack is l provided in the control room and meets the following criteria:
- 1. Measurement and indication is in epm or mR/hr convertible to pCl/cc;
- 2. The systems are not designed to meet the redundancy requirements as defined in IEEE 279;
- 3. Radiation levels are continuously recorded in the control room during an accident. The major equipment associated with the plant computer system is located in the computer room. This computer equipment consists of termination cabinets, signal conditioning components, analog to digital converters, central processing units and bulk data storage devices. Additional support devices (console terminals, line printers, communication controllers and display generators) are also located in the computer room. Additionally, two cabinets for special terminations and the mounting of special hardware, including power supplies, converters, t ! latching contacts and chassis for resistance temperature detector bridges are also located in the computer room.
The main control room has three independent communication systems. One system consists of telephones which are leased from the local telephone company. These telephones and several - outside trunk lines service the station for outside calls. This system may or may not be available under emergency conditions. The second system is a communication and voice paging system and interconnects the entire station. The third system is sound-powered, with telephone Jacks and interconnecting wires at each major control point for test purposes. This system does not rely on any power source; therefore, it is available at all times, 3.2.4 Turbine Building The turbine building houses the secondary plant components and systems. Though the turbine building is not a Class I seismic structure, it is designed for wind forces which are greater than the seismic forces as determined from a combined seismic analysis of the service building, control room and the turbine hall. The turbine hall will remain intact during the DBE. The building is proviAd with three operating levels: C\ DSAR 3 34 Rev.14
rx MYAPC (V) First Floor Elevation 21 feet 0 Inches The main equipment located on this level as shown Figure 3.2-6 consists of the main condenser, condensate, heater drain and steam generator feed pumps, vertical feedwater heaters and the water treatment plant. None of the aforementioned systems or components are required in the defueled condition of the plant. Enclosures for the auxillary boilers, diesel generatore and for the control room are adjacent to the turbine building first floor operator level. The first floor levelis extended to include the service building area. Menanine Floor Elevation 39/35 Feet 0 inches The heater drain receivers, gland steam condenser and air ejectors are located on the mezzanine level. The principal function of this floor is h provide convenient access to the operating ec,ulpment and valve operating stations. None of the uforementioned systems or components are required in the defueled condition of the plant. The mezzanine floor plan is shown in Figurr 3.2 7. Turbine Generator Floor Elevation 61 Feet 0 inches p
) The third floor area is occupied with the turbine generator, the moisture separators and the second I
point feedwater heaters. The feedwater vertical heat exchangers project through and above the floor Hvel. None of the aforementioned systems or components are required in the defueled condition of the plant. The third level of the turbine building is shown in Figure 3.2 8. 3.2.5 Primary Auxiliary Building (PAB) Function The PAB is located within the radiation control area (RCA) and is shown on Figures 3.2 9 and 3.210. Descriotion The PAB has a concrete fcundation on bedrock 9 feet below ground grade, with 2 foot thick reinforced concrete exterior walls extending 16 feet above ground level. Radioactive equipment and tanks within the structure are shielded as required with concrete walls. The concrete cubicle provided to shield the volume control tank extends above the roof of the building. All axterior and interior concrete walls and slabs are designed to safely resist the hypothetical earthquake. DSAR- 3 35 Rev.14 l
l MYAPC
'b Primarv AL xillary Buildina Eaulomant The building contains the aerated drain and the hydrogenated drain tanks, the waste gas, and liquid ,
waste disposal system desc.ibed in Sections 3.3.6 and 4.4 of this report. The larger tanks of the I waste disposal system are the boron waste storage and the test tanks which are located on the north
- and south side of the PAB. The containment pipe tunnel connects to the building at the 11 foot 0-Inch floor elevation, and the associated piping and valves occupy most of the floor area. The lower level of the PAB contains the liquid transfer pump *, the auxiliary charging pump and the boric acid mix tank. The three charging pumps of the charging and volume control system are located on the second or ground floor of the building at Elevation 21 feet 0 inches. The upper floor plan view ls shown in Figure 3.2 9. The containment purge fan, the PAB ventilation fan and the particulate and charcoal filter of the ventilation system are located on this floor. The waste gas compressors, the waste gas decay drums and the volume control tank are also located on the upper floor level. This floor is provided with removable slabs over the filtor and demineralizer cubicles which are built up from the lower PAB floor at Elevation 11 feet 0 inches. The upper floor operating area is also provided with a monorail for the handling of expended filter elements. ^ 3.2.6 Sarvice Water Intake Structure The circulating water pumps and seMee water pumps are located in this structure. The building floor plans at Elevation 7 feet 0 inches and 21 feet 0 inches are shown on Figures 3.211.
A concrete barrier separates the service water pumps from the circulating water pumps and piping. This barrier, in conjunction with high water alarms and circulating water pump trips, protects the service water systems from the effects of flooding should a rupture of the circulating water system occur. Attemate sources of heat sink are available for fuel pool cooling in the event that the service water pumps are not available. 3.2.7 Fire Pump Building
- The fire pump house is shown on Figure 3.212. It is located near the water storage pond and houses the equipment servicing the liquid portion of the fire protection system. The building houses two fire pumps, a pressure maintenance pump and a hydro-pneumatic tank. A diesel engine drives one pump while the other is motor-driven. The building houses the diesel fuel tank, the batteries and control board required for the diesel operation.
,0 DSAR 3 36 Rev.14
s MYAPC s 3.2.8 Masonry Walls 1 Masonry walls in the vicinity of safety-related equipment have been evaluated to determine whether they will withstand all postulated design loads (selsmic, tomado, hurricane, attached equipment, etc.). , Table 3.2.1 lists the walls evaluated and notes those which are fully qualified and those which are assumed to fail, but such failure will not result in unacceptable damage to equipment. Bulletin 8011 Category Definitions: Category 3: Walls whose collapse will not affect safety related equipment. i t ' This list has been revised to address only those walls that are required in the peimr.nently defueled condition. ROOM REFEP.ENCE WALL EQUIPMENT BUILDING WALL 10 LOCATION DRAWING LOCATION PROTECTED CATEGORY Fuel FB 44 1 Spent Fuel 11550-FA 12A South Wall Spent Fuel 3 Pool Spent Fuel Purification Elev 44' 6" Building and Cooling Return Lines and spent fuel storage racks. b ( I a t DSAR- 3-37 Rev.14
.. _ _ . __ _ ~ . _ _ __._ _..._._ -
MYAPC Radian 3.2 - Referenr==:
- 1. YNSD l.etter to R. Fraser from D.L. Magnarelli/W.E. Henries,
- Review of AES Analysis of SFP for Elevated Temperatures," dated Nov. 25,1997.
- 2. YNSD Calculation No. MYC-2001. " Analysis of SFP Structure for Elevated Pool Water Temperature," dated November 25,1997.
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