ML14191A257

Defueled Safety Analysis Report, Revision 11.2

ML14191A257
Person / Time
Site:	Millstone
Issue date:	06/30/2014
From:	Dominion Nuclear Connecticut
To:	NRC/FSME
Shared Package
ML14191A199	List: ML14191A257
References
14-310
Download: ML14191A257 (204)

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{{#Wiki_filter:ONE POWER STATION UNIT 1 LED SAFETY ANALYSIS REPORT ON 11.2 CUMENT INCORPORATES APPROVED CHANGES MPS-1 DSAR. O THIS DOCUMENTS REVISION HISTORY FOR LARS.

REPORTING PERIOD 2009 - 2010 FSC PKG Document Number DATE SECTION Summary Description of Changes - 04/09 As identified in the 2009 NRC Submittal List Administrative (FSAR content not affected). Change indicator (s) and of Changed Pages and submitted Summary of page change identification (s) present in the 2009 NRC Submittal Change. removed in preparation for the 2010 NRC Submittal. This forms the base line for changes incorporated under the Revision 7 series. Revision level of the authoring files are unchanged. This supports Revision/Change traceability. Revision 8 (2010 - 2011 Reporting Period) sion Document Elements Affected Change Activity Summary Description of Changes Date (Sections, Tables, Figures) 011 MP1-DFCR-2010-001 S3.2.7.2 Reflects change to substation nomenclature (Northeast Utilities Distribution Project) for off site power source. Page 1 of 3

Changed Pages and submitted Summary of Change. change identification (s) present in the 2011 NRC Submittal removed in preparation for the 2012 NRC Submittal. This forms the base line for changes incorporated under the Revision 9 series. Revision level of the authoring files are unchanged. This supports Revision/Change traceability. Revision 10, (2012-2013 Reporting Period) sion Document Elements Affected Change Activity Summary Description of Changes Date (Sections, Tables, Figures) 0 Administrative As identified in the previous NRC Submittal List of Administrative (FSAR content not affected). Change indicator (s) and page Changed Pages and submitted Summary of Change. change identification (s) present in the previous NRC Submittal removed in preparation for the 2013 NRC Submittal. This forms the base line for changes incorporated under the Revision 10 series. Revision level of the authoring files are unchanged. This supports Revision/Change traceability. 2 MP1-DFCR-2011-001 S1.2.3.2.1, S4.1, S4.4 Reflects updates for compliance with NPDES permit. Page 2 of 3

Changed Pages and submitted Summary of Change. change identification (s) present in the previous NRC Submittal removed in preparation for the 2014 NRC Submittal. This forms the base line for changes incorporated under the Revision 11 series. Revision level of the authoring files are unchanged. This supports Revision/Change traceability. Administrative List of Figures Reflects administrative clarification of revision status for engineering controlled drawings that are coincidently FSAR Figures. FSAR figures corresponding to Controlled P&IDs are updated on a periodic basis by the FSAR Coordinator. MP1-DFCR-2013-001 List of Figures Reflects administrative clarification of revision status for engineering controlled drawings that are coincidently FSAR Figures. FSAR figures corresponding to Controlled P&IDs are updated on a periodic basis by the FSAR Coordinator. Page 3 of 3

RODUCTION ...................................................................................... 1.1-1 NERAL PLANT DESCRIPTION ......................................................... 1.2-1 ANT SITE AND ENVIRONS ............................................................... 1.2-1 ation and Site ........................................................................................ 1.2-1 Ownership............................................................................................ 1.2-1 ess to the Site ........................................................................................ 1.2-1 cription of the Environs ........................................................................ 1.2-1 logy....................................................................................................... 1.2-1 mology and Design Response Spectra ................................................. 1.2-1 rology ................................................................................................... 1.2-2 eorology................................................................................................ 1.2-2 Environmental Radioactivity Monitoring Program ............................. 1.2-2 MMARY PLANT DESCRIPTION ....................................................... 1.2-3 STEMS .................................................................................................. 1.2-3 l Storage and Fuel Handling ................................................................. 1.2-3 ioactive Waste Processing System ....................................................... 1.2-3 iation Monitoring and Control.............................................................. 1.2-4 iliary Systems....................................................................................... 1.2-5 ion Communication System.................................................................. 1.2-5 ion Water Purification, Treatment and Storage System ....................... 1.2-6 NTIFICATION OF AGENTS AND CONTRACTORS...................... 1.3-1 PLICANTS SUBSIDIARIES............................................................... 1.3-1 CLEAR STEAM SUPPLY SYSTEM SUPPLIER............................... 1.3-1 CHITECT/ENGINEER ......................................................................... 1.3-1 RBINE-GENERATOR SUPPLIER ...................................................... 1.3-1 i Rev. 3.2

NFORMANCE TO NRC REGULATORY GUIDES .......................... 1.5-1 MMARY DISCUSSION ....................................................................... 1.5-1 FERENCE.............................................................................................. 1.5-2 CHAPTER 2- SITE CHARACTERISTICS CATION AND AREA .......................................................................... 2.1-1 PULATION ........................................................................................... 2.1-2 ulation Distribution Within 50 Miles.................................................... 2.1-3 nsient Population ................................................................................... 2.1-3 Population Zone.................................................................................. 2.1-3 ulation Center ....................................................................................... 2.1-4 ND USE................................................................................................. 2.1-5 cription of Facilities.............................................................................. 2.1-5 elines...................................................................................................... 2.1-8 terways .................................................................................................. 2.1-8 ports ....................................................................................................... 2.1-8 hways .................................................................................................... 2.1-9 lroads ..................................................................................................... 2.1-9 jections of Industrial Growth............................................................... 2.1-10 TERMINATION OF DESIGN BASIS EVENTS ............................... 2.1-11 ECTS OF DESIGN BASIS EVENTS ............................................... 2.1-12 FERENCES ......................................................................................... 2.1-12 TEOROLOGY ...................................................................................... 2.2-1 GIONAL CLIMATOLOGY.................................................................. 2.2-1 CAL METEOROLOGY........................................................................ 2.2-1 ii Rev. 3.2

al Meteorological Conditions for Design and Operating es. .......................................................................................................... 2.2-1 SITE METEOROLOGICAL MEASUREMENTS PROGRAM ......... 2.2-1 ORT TERM (ACCIDENT) DIFFUSION ESTIMATES ...................... 2.2-2 ective ..................................................................................................... 2.2-2 culations ................................................................................................ 2.2-2 ults......................................................................................................... 2.2-2 NG-TERM (ROUTINE) DIFFUSION ESTIMATES........................... 2.2-2 ective ..................................................................................................... 2.2-2 culations ................................................................................................ 2.2-3 FERENCES ........................................................................................... 2.2-3 DROLOGIC ENGINEERING .............................................................. 2.3-1 DROLOGIC DESCRIPTION ............................................................... 2.3-1 E AND FACILITIES ............................................................................ 2.3-1 OODS ..................................................................................................... 2.3-1 od History .............................................................................................. 2.3-1 od Design Considerations...................................................................... 2.3-1 ct of Local Intense Precipitation .......................................................... 2.3-1 OBABLE MAXIMUM FLOOD (PMF) ON STREAMS D RIVERS............................................................................................. 2.3-2 TENTIAL DAM FAILURE, SEISMICALLY INDUCED................... 2.3-2 OBABLE MAXIMUM SURGE AND SEICHE FLOODING.............. 2.3-2 bable Maximum Winds and Associated Meteorological ameters................................................................................................... 2.3-2 ge and Seiche Water Levels .................................................................. 2.3-3 ve Action ............................................................................................... 2.3-3 iii Rev. 3.2

bable Maximum Tsunami Flooding ..................................................... 2.3-4 EFFECTS ............................................................................................ 2.3-4 OLING WATER CANALS AND RESERVOIRS ............................... 2.3-4 ANNEL DIVERSIONS......................................................................... 2.3-4 OODING PROTECTION REQUIREMENTS ...................................... 2.3-4 W WATER CONSIDERATIONS ........................................................ 2.3-4 Flow in Rivers and Streams ............................................................... 2.3-4 Water Resulting from Surges, Seiches, or Tsunamis ......................... 2.3-4 PERSION, DILUTION, AND TRAVEL TIMES OF CIDENTAL RELEASES OF LIQUID EFFLUENTS RFACE WATERS. ................................................................................ 2.3-4 OUNDWATER ..................................................................................... 2.3-5 CHNICAL SPECIFICATION AND EMERGENCY ERATION REQUIREMENTS .............................................................. 2.3-5 FERENCES ........................................................................................... 2.3-5 OLOGY, SEISMOLOGY, AND GEOTECHNICAL GINEERING ......................................................................................... 2.4-1 SIC GEOLOGIC AND SEISMIC INFORMATION............................ 2.4-1 RATORY GROUND MOTION .......................................................... 2.4-1 e Fuel Storage Earthquake..................................................................... 2.4-1 RFACE FAULTING ............................................................................. 2.4-1 logic conditions of the Site................................................................... 2.4-1 dence of Fault Offset ............................................................................. 2.4-1 thquakes Associated with Capable Faults ............................................. 2.4-1 estigation of Capable Faults .................................................................. 2.4-1 relation of Epicenters with Capable Faults ........................................... 2.4-2 iv Rev. 3.2

ults of Faulting Investigation ................................................................ 2.4-2 ABILITY OF SUBSURFACE MATERIALS AND UNDATIONS ........................................................................................ 2.4-2 ABILITY OF SLOPES .......................................................................... 2.4-2 BANKMENTS AND DAMS ............................................................... 2.4-2 FERENCES ........................................................................................... 2.4-2 CHAPTER 3 - FACILITY DESIGN AND OPERATION SIGN CRITERIA .................................................................................. 3.1-1 NFORMANCE WITH 10 CFR 50 APPENDIX A NERAL DESIGN CRITERIA............................................................... 3.1-1 mary Discussion .................................................................................. 3.1-1 tematic Evaluation Program and Three Mile Island luations of General Design Criteria ...................................................... 3.1-1 ASSIFICATION OF STRUCTURES, SYSTEMS, AND MPONENTS ........................................................................................ 3.1-1 mic Classification................................................................................. 3.1-1 ety Related Classification ...................................................................... 3.1-3

 -Safety Related Plant Functions Maintained in the ueled Condition..................................................................................... 3.1-4 s Important to the Defueled Condition ................................................ 3.1-4 ND AND TORNADO LOADINGS ...................................................... 3.1-8 TER LEVEL DESIGN ......................................................................... 3.1-8 SILE PROTECTION ........................................................................... 3.1-8 rnally Generated Missiles ..................................................................... 3.1-8 siles Generated by Natural Phenomena ................................................ 3.1-9 siles Generated by Events Near the Site ............................................... 3.1-9 v                                                                Rev. 3.2

mparison of Measured and Predicted Responses ................................. 3.1-10 SIGN OF CLASS I AND CLASS II STRUCTURES......................... 3.1-10 ign Criteria, Applicable Codes, Standards and cifications............................................................................................ 3.1-10 ds and Loading Combinations ............................................................ 3.1-10 ctural Criteria for Class II Structures ................................................. 3.1-12 mic Class I and II Structures .............................................................. 3.1-13 SMIC QUALIFICATION OF SEISMIC CATEGORY I TRUMENTATION AND ELECTRICAL EQUIPMENT ................. 3.1-16 VIRONMENTAL DESIGN OF ELECTRICAL EQUIPMENT ........ 3.1-16 FERENCES ......................................................................................... 3.1-16 STEMS .................................................................................................. 3.2-1 EL STORAGE AND HANDLING ....................................................... 3.2-1 w Fuel Storage........................................................................................ 3.2-1 nt Fuel Storage ...................................................................................... 3.2-1 nt Fuel Pool Cooling System ................................................................ 3.2-3 l Handling System ................................................................................ 3.2-5 NITORING AND CONTROL FUNCTIONS ..................................... 3.2-6 CAY HEAT REMOVAL (DHR) SYSTEM ......................................... 3.2-6 ign Bases .............................................................................................. 3.2-6 tem Description ..................................................................................... 3.2-7 ety Evaluation ........................................................................................ 3.2-7 ting and Inspection ............................................................................... 3.2-7 rumentation .......................................................................................... 3.2-7 KEUP WATER SYSTEM.................................................................... 3.2-7 mineralized Water ................................................................................. 3.2-7 vi Rev. 3.2

ign Bases............................................................................................... 3.2-8 tem Description ..................................................................................... 3.2-8 ety Evaluation ........................................................................................ 3.2-9 ting and Inspection ............................................................................... 3.2-9 ECTRICAL SYSTEMS ......................................................................... 3.2-9 oduction................................................................................................. 3.2-9 Site Source............................................................................................ 3.2-9 ntionally Deleted................................................................................... 3.2-9 Site Electric System .............................................................................. 3.2-9 CONDITIONING, HEATING, COOLING AND NTILATION SYSTEMS..................................................................... 3.2-11 ctor Building and SFPI Heating and Ventilation System ................... 3.2-11 waste Building Ventilation System .................................................... 3.2-13 ntionally Deleted................................................................................. 3.2-14 bine Building Heating and Ventilation ............................................... 3.2-14 E PROTECTION SYSTEMS ............................................................. 3.2-15 ign Bases ............................................................................................ 3.2-15 tem Description .................................................................................. 3.2-16 ety Evaluation and Fire Hazards Analysis........................................... 3.2-19 ection and Testing .............................................................................. 3.2-21 sonnel Qualification and Testing......................................................... 3.2-22 FERENCES ........................................................................................ 3.2-23 vii Rev. 3.2

URCE TERMS ...................................................................................... 4.1-1 DIATION PROTECTION DESIGN FEATURES ............................... 4.2-1 CILITY DESIGN FEATURES ............................................................. 4.2-1 ign Basis ............................................................................................... 4.2-1 tilation .................................................................................................. 4.2-1 DIATION PROTECTION PROGRAM................................................ 4.2-1 anization................................................................................................ 4.2-1 ARA PROGRAM .................................................................................. 4.3-1 LICY CONSIDERATIONS ................................................................. 4.3-1 ign Considerations ................................................................................ 4.3-1 rational Considerations......................................................................... 4.3-1 UID WASTE MANAGEMENT SYSTEMS ....................................... 4.4-1 LID WASTE MANAGEMENT ............................................................ 4.5-1 SIGN BASES ....................................................................................... 4.5-1 STEM DESCRIPTION.......................................................................... 4.5-1 FERENCES ........................................................................................... 4.5-2 LUENT RADIOLOGICAL MONITORING AND SAMPLING ....... 4.6-1 SIGN ..................................................................................................... 4.6-1 ign Basis ............................................................................................... 4.6-1 tem Design Description......................................................................... 4.6-1 EA RADIATION MONITORING INSTRUMENTATION ................ 4.6-2 ign Bases............................................................................................... 4.6-2 tem Description ..................................................................................... 4.6-2 viii Rev. 3.2

CHAPTER 5 - ACCIDENT ANALYSIS RODUCTION ...................................................................................... 5.1-1 CIDENT EVENT EVALUATION ....................................................... 5.1-1 cceptable Results for Design Basis Accidents (DBAs)........................ 5.1-1 l Handling Accident Assumptions ....................................................... 5.1-1 ults......................................................................................................... 5.1-1 iological Consequences ........................................................................ 5.1-1 FERENCES ........................................................................................... 5.1-2 EL HANDLING ACCIDENT ............................................................... 5.2-1 EL HANDLING ACCIDENT SCENARIOS IN THE NT FUEL POOL.................................................................................. 5.2-1 DIOLOGICAL CONSEQUENCES...................................................... 5.2-2 FERENCES ........................................................................................... 5.2-3 CHAPTER 6 - CONDUCT OF OPERATIONS GANIZATIONAL STRUCTURE ....................................................... 6.1-1 NAGEMENT AND TECHNICAL SUPPORT GANIZATION ...................................................................................... 6.1-1 hnical Support for Operations............................................................... 6.1-1 anizational Arrangement....................................................................... 6.1-1 ERATING ORGANIZATION ............................................................. 6.1-1 nt Organization ..................................................................................... 6.1-1 nt Personnel Responsibilities and Authorities ....................................... 6.1-1 rating Shift Crews ................................................................................ 6.1-1 ix Rev. 3.2

FERENCES ........................................................................................... 6.1-2 CHNICAL SPECIFICATIONS ............................................................ 6.2-1 OGRAMS ............................................................................................. 6.3-1 AINING ................................................................................................. 6.3-1 ERGENCY PLAN ................................................................................ 6.3-1 YSICAL SECURITY PLANS............................................................... 6.3-1 ALITY ASSURANCE PROGRAM DESCRIPTION (QAPD) PICAL REPORT ................................................................................... 6.3-1 FERENCES ........................................................................................... 6.3-2 OCEDURES ......................................................................................... 6.4-1 VIEW AND AUDIT.............................................................................. 6.5-1 SITE REVIEW...................................................................................... 6.5-1 EPENDENT REVIEW ........................................................................ 6.5-1 DITS ..................................................................................................... 6.5-1 CHAPTER 7 - DECOMMISSIONING MMARY OF ACTIVITIES .................................................................. 7.1-1 COMMISSIONING APPROACH ....................................................... 7.1-2 nning ..................................................................................................... 7.1-2 Characterization................................................................................... 7.1-3 ontamination ......................................................................................... 7.1-3 or Decommissioning Activities ............................................................ 7.1-4 er Decommissioning Activities............................................................. 7.1-5 x Rev. 3.2

ORAGE OF RADIOACTIVE WASTE................................................. 7.1-6 h Level Waste ....................................................................................... 7.1-7 Level Waste ........................................................................................ 7.1-7 ste Management..................................................................................... 7.1-7 DIATION EXPOSURE MONITORING.............................................. 7.1-7 FERENCES .......................................................................................... 7.1-7 IMATE OF RADIATION EXPOSURE.............................................. 7.2-1 CLEAR WORKER .............................................................................. 7.2-1 NERAL PUBLIC .................................................................................. 7.2-1 RMAL TRANSPORTATION .............................................................. 7.2-2 NTROL OF RADIATION RELEASES ASSOCIATED TH DECOMMISSIONING EVENTS .................................................. 7.3-1 PLANT EVENTS ................................................................................. 7.3-1 ANSPORTATION ACCIDENTS ......................................................... 7.3-1 N-RADIOLOGICAL ENVIRONMENTAL IMPACTS ..................... 7.4-1 DITIONAL CONSIDERATIONS ........................................................ 7.4-1 xi Rev. 3.2

Revision 10, (2012-2013 Reporting Period) Millstone Unit No.1 Licensing Milestones This Table has been Intentionally Deleted 1990 Population and Population Densities - Cities and Towns within 10 miles of Millstone Population Growth 1960 - 1990 Population Distribution within 10 miles of Millstone - 1990 Census Population Distribution Within 10 Miles of Millstone 2000 Projected Population Distribution Within 10 Miles of Millstone 2010 Projected Population Distribution Within 10 Miles of Millstone 2020 Projected Population Distribution Within 10 Miles of Millstone 2030 Projected Population Distribution Within 50 Miles of Millstone - 1990 Census Population Distribution Within 50 Miles of Millstone - 2000 Projected Population Distribution Within 50 Miles of Millstone - 2010 Projected Population Distribution Within 50 Miles of Millstone - 2020 Projected Population Distribution Within 50 Miles of Millstone - 2030 Projected Transient Population Within 10 Miles of Millstone 1991-1992 School Enrollment Transient Population Within 10 Miles of Millstone - Employment Population Distribution Within 50 Miles of Millstone - 2030 Projected Low Population Zone Permanent Population Distributions Low Population Zone School Enrollment and Employment Metropolitan areas Within 50 Miles of Millstone 1990 Census Population Population Centers within 50 Miles of Millstone Population Density Within 10 Miles of Millstone 1990 (People per Square Mile) xii Rev. 2

Population Density Within 50 Miles of Millstone 1990 (People per Square Mile) Population Density Within 50 Miles of Millstone 2030 (People per Square Mile) Cumulative Population Density Within 50 Miles of Millstone 1990 (People per Square Mile) Cumulative Population Density Within 50 Miles of Millstone 2030 (People per Square Mile) Description of Facilities List of Hazardous Materials Potentially Capable of Producing Significant Missiles Comparison with NRC General Design Criteria Allowable Stresses for Class I Structures Effluent Radiation Monitors Area Radiation Monitoring System Sensor and Converter Locations for Millstone Unit No. 1 Assumptions and Input Conditions for Fuel Handling Accident at Millstone Unit No. 1 xiii Rev. 2

eneral Arrangement RAD Waste Buildings - Plans eneral Arrangement Buildings RAD Waste Buildings - Sections eneral Site Location eneral Vicinity te Layout te Plan owns Within 10 Miles opulation Sectors for 0 - 10 Miles opulation Sectors for 0 - 50 Miles oads and Facilities in the LPZ PZ Population Sectors Distribution strument Landing Patterns at Trumbell Airport ir Lanes Adjacent to Millstone Point ew London County - State Highways and Town Roads opography in the Vicinity of Millstone Point eactor Building Seismic Loads cceleration Diagram Under Seismic Loads 5 Percent Damping hear Diagram Under Seismic Loads oment Diagram Under Seismic Loads isplacement Diagram Under Seismic Loads adwaste Building - Mathematical Model &ID: SFPI, Fuel Pool Cooling System &ID: SFPI, Fuel Pool Cooling System &ID: SFPI, Fuel Pool Cooling System &ID: Reactor Building and HVAC Room SFPI Secondary Cooling (DHR) ystem xiv Rev. 11.2

&ID: HVAC B.O.P. System Composite rough 3.2-11 Intentionally Deleted &ID: HVAC Balance Of Plant System Composite &ID: HVAC System (Radwaste Storage Building) re Protection Composite xv Rev. 11.2

of Millstone Unit Number 1. rinciple licensing source document describing the pertinent equipment,

, operational constraints and practices, accident analyses, and activities associated with the existing defueled condition of Millstone Unit
 , the DSAR is intended to serve in the same role as the Final Safety Analysis e Unit Number 1 during the periods of power operation between 1970 and s applicable throughout the decommissioning of Millstone Unit Number 1. The process is dynamic. The issuance of the DSAR does not alleviate the licensee follow all required surveillances, procedures, technical specifications or until those documents are officially modified using approved processes.

res of structures, systems, or components (SSCs) included or referenced in the d within the licensing basis of the facility only to the extent that they show ribed in the text of the DSAR. Other contents of drawings and figures may not configuration of the facility and are not maintained. illstone Unit Number 1 was authorized by a provisional construction permit 19,1966, in AEC Docket 50-245. Millstone Unit Number 1 was completed and ing during October 1970. The plant went into commercial operation on

0. On July 21, 1998, pursuant to 10 CFR 50.82(a)(1)(i) and 10 CFR licensee certified to the NRC that, as of July 17, 1998, Millstone Unit Number ceased operations and that fuel had been permanently removed from the issuance of this certification fundamentally changes the licensing basis of mber 1 in that the NRC-issued 10 CFR 50 license no longer authorizes actor or emplacement or retention of fuel in the reactor vessel. Therefore, as of those conditions or activities associated with the safe storage of fuel and tion (including waste handling, storage and disposal) are applicable to the Unit Number 1 plant.

mber 1 was a single cycle, boiling water reactor with a Mark I containment d, furnished and constructed by General Electric Company as prime contractor e General Electric Company engaged Ebasco Services Incorporated as Millstone Unit Number 1 had a reactor thermal output of 2011 megawatts and put of 652.1 megawatts. The Millstone site is located in the town of Waterford, ty, Connecticut, on the north shore of Long Island Sound. 1.1-1 Rev. 2

November 1, 1968 ing License Issued October 7, 1970 ng License Issued October 31, 1986 e October 7, 1970 October 26, 1970 he Grid November 1970 r January 6, 1971 tion December 28, 1970 ed Operations July 21, 1998 Page 1 of 1 Rev. 2

town of Waterford, Connecticut on the north shore of Long Island Sound and Niantic River Estuary. It is located 3.2 miles west-south-west of New London,

-east of Hartford, Connecticut. The site is bounded on the west, south, and sides by Long Island Sound. The nearest residential boundary is 855 meters ajor structures of Millstone Unit Number 1. Chapter 2 contains more detailed site and surrounding areas.

ership by Dominion Nuclear Connecticut, Inc. the Site a around the station, excluding the intake and discharge canal, is completely rity fence. This fence establishes the protected area boundary of the station. n is controlled by Security Personnel. on of the Environs to the north and west is cultivated land with residential dwellings. The village ng of a small commercial complex and attendant residential development, is st of the Reactor Building. Other residential areas adjoin the site at the end of ad and at distances of 1 to 3 miles. miles ENE of the Reactor Building, is the nearest urban complex and includes commercial, and industrial uses. derlain by Monson gneiss and Westerly granite. The Westerly granite intrudes , is more resistant to weathering and therefore forms ridges. Seismic surveys al or extreme subsurface conditions. Chapter 2 contains more detailed logy and seismic qualities. gy and Design Response Spectra t site area is placed in Zone 2 (zone of moderate damage) on the seismic the 1964 Uniform Building Code. 1.2-1 Rev. 10

ral grade level is at an elevation of approximately 14 feet above mean sea level. tours of the land and ground strata, and the distance of the reactor from water accidentally released from the plant can reach industrial or drinking water ns more detailed information on hydrology. ogy f the site area is basically that of a sea-coast location with relatively favorable n conditions prevailing. The inland terrain in Connecticut is not pronounced any significant local modifications of synoptic conditions at the shoreline. The however, experience local modifications of synoptic patterns because of the nces between air over land and air over water. in an area occasionally traversed by hurricanes. The design basis hurricane for mph maximum gradient winds and a 17 mph speed of translation. This is intense than the worst on record (hurricane of 1938). ed that a tornado can be expected to strike a point on the Millstone site about In spite of this low probability, the features of the plant important to the safe d fuel have been designed to withstand 300 mph winds. from the viewpoint of site meteorology, the site is suitable for the station as r 2 contains more detailed information concerning meteorology.) ronmental Radioactivity Monitoring Program radioactivity monitoring program was initiated and has been conducted at the

67. Data are collected to measure radioactivity present in the environs. The ing in order to assure prompt detection and evaluation of any changes in 1.2-2 Rev. 10

The overall arrangement of this building is shown in Figures 1.2-2 and 1.2-3. age and Fuel Handling orage and Handling Equipment age pool holds fuel assemblies, control rods, and small vessel components. The ns provisions to maintain water cleanliness and instrumentation to monitor p water is available from the Unit 2 demineralized water system and the fire racks in which fuel assemblies are placed are designed and arranged to ensure pool. ent fuel is performed within the Reactor Building. This employs a refueling water fuel transport, storage racks for fuel and control rods in a storage pool, eparation stations, and floor mounted jib cranes. Control rods can be stored in or on hooks on the side of the pool. f the fuel storage and equipment storage facilities meets all requirements for For additional information, refer to Chapter 3. ol Cooling System ng system provides cooling for the spent fuel pool water when required. ng system consists of a circulating pump, heat exchanger, skimmer surge g, valves, and instrumentation and controls. Pool cleanup is provided by an in-and filter. For additional information, refer to Chapter 3. ve Waste Processing System ste processing system is designed to control the release of plant-produced l to within the limits specified in 10 CFR 20 and Appendix I to 10 CFR 50. lection, transfer, and evaporation. 1.2-3 Rev. 10

gical monitoring will be conducted using an in-line Liquid Effluent Monitor or to discharge through DSN-001A (Emergency Service Water discharge canal), dilution flow requirements will be established by crediting Unit 2 Flow to the common discharge canal. Alternatively, this system could be e process liquids from the Reactor Building sumps to containers which would liquid to be processed onsite or offsite. adwaste Handling ating from nuclear system equipment maybe stored in the spent fuel storage for off site shipment in approved shipping containers. llected and appropriately prepared for off site shipment. Examples of these ter residue, spent resins, paper, air filters, rags, and used clothing. For tion, refer to Chapter 4. Monitoring and Control on Monitoring and Sampling ol Island ventilation exhaust is monitored for gaseous radiation and iculate sampling skid is provided for Unit 1 Balance of Plant (BOP) exhaust to r any significant changes. For additional information, refer to Chapter 4. adiation Monitors are provided to monitor for abnormal radiation at selected locations on the ors actuate alarms when abnormal radiation levels are detected. Radwaste Processing System Control e system is designed to safely and economically collect, store, process, and cle, all radioactive or potentially radioactive liquid waste generated. The a batch basis. adwaste Control be transferred to high integrity cask containers for shipment. 1.2-4 Rev. 10

ring and Control Functions t 2 Control Room is continuously manned, and serves as the control room for Millstone Unit 2 Operations personnel are responsible for the monitoring and 1 spent fuel pool island (SFPI) and auxiliary systems via a computer console stone Unit 2 Control Room. otection System detection systems are provided at Millstone Unit Number 1 to protect

, and components important to the defueled condition of the unit.

system includes a fire water supply system that consists of two fire water mps and a distribution system that delivers fire water to all parts of the plant. within the plant protect individual hazards and include sprinkler systems and al Power System Power Supply system includes the electrical equipment and connections required to supply xiliaries. Power Supply C system is provided via rectified AC at the point of use. In addition, a separate 125V DC system powered by batteries and a battery charger provides a source decommissioning electrical system. provided by power supplies within the SFPI Programmable Logic Controller ommunication System ication system provides for reliable on site and off site communications both contingency conditions. 1.2-5 Rev. 10

1.2-6 Rev. 10 STEAM SUPPLY SYSTEM SUPPLIER ompany was the nuclear steam system supplier for the plant. CT/ENGINEER corporated was the Architect/Engineer for Millstone Unit Number 1. GENERATOR SUPPLIER tor was manufactured by General Electric Company. 1.3-1 Rev. 2

1.4-1 Rev. 2 illstone Unit Number 1 began operation with Provisional Operating License ued October 7, 1970. mber 1 submitted summaries of compliance to these guides in the early 1970s pplication for a full-term operating license (Reference 1.5-1). is application, the NRC (formerly AEC) initiated the Systematic Evaluation 1977 to review the designs of older operating nuclear reactor plants in order to ent their safety. Millstone Unit Number 1 was identified as an SEP plant. s were: blish documentation that shows how the criteria for each operating plant d compare with current criteria on significant safety issues and to provide a for acceptable departures from these criteria. ide the capability to make integrated and balanced decisions with respect to uired backfitting. ide for early identification and resolution of any significant deficiencies. ss the safety adequacy of the design and operation of currently licensed nuclear lants. available resources efficiently to minimize requirements for additional es by NRC or industry. re that the safety assessments were adequate for conversion of provisional ng licenses to full-term operating licenses. f the SEP program report included the status of all applicable generic activities cluding those that formed the basis for the Integrated Safety Analysis Program emented by the Licensee. Based upon the acceptable conclusions reached in ed the full-term operating license for Millstone Unit Number 1 on October 31, 1.5-1 Rev. 2

1.5-2 Rev. 2 0 miles southeast of Hartford. t Number 1 containment structure is located immediately south of Millstone 2 hical coordinates of the centerline of the reactor is as follows: mber 1 Latitude and Longitude Northing and Easting N 41° 18'32" N 173, 800 W 72° 10'04" E 759, 965 y Dominion Nuclear Connecticut, Inc. Figures 2.1-1 through 2.1-4 identify the area is considered the restricted area. The restricted area has been ed and administrative procedures, including periodic patrolling, have been access to the area. For the purpose of radiological dose assessment of usion area boundary (EAB) was considered the actual site boundary for xcept in the Fox Island / discharge channel area on the south end of the site. For he nearest land site boundary distance was used. rmal releases are discharged to the atmosphere via the Unit Number 1 BOP he SFPI ventilation exhaust point. The distance from the Unit Number 1 BOP he SFPI ventilation exhaust point to the nearest residential property boundary int Colony development (Point A on Figure 2.1-3) is greater than 2,800 feet. adjacent to the eastern site boundary, consists of single family homes on 104 of the conditions of the sale of the site to the Hartford Electric Light Company t Light and Power Company was that permanent dwellings would never be ach area of the development. Because of this restriction, normal release doses oint A rather than at the nearest point on the site boundary. mplete control of activities within the exclusion area, except for the passage of ovidence & Worcester (P&W) / Amtrak Railroad track which runs east-west y of people within the exclusion area during an emergency, an emergency plan n prepared. The plan includes provisions for alarms both inside and outside eates the evacuation routes and assembly areas to be used. The State of 2.1-1 Rev. 3

clusion area is leased to the Town of Waterford for public recreation and is soccer and baseball games. Figure 2.1-3 shows the general location of these pt is made to restrict the number of persons using these facilities. Estimates of ce indicate that about 2,000 visitors could be within the exclusion area at any cer and baseball fields. The licensee's Emergency Plan provides for removal of e site. The number and configuration of roads and highways assure ready as described above (Figures 2.1-2, 2.1-3, and 2.1-4). ION ulation within 10 miles of the station was estimated to be 120,443. This cted to increase to about 129,846 people by the year 2000 and to a total of

,277 people by the year 2030 (New York State Department of Economic 9 (Reference 2.1-1); State of Connecticut Office of Policy and Management,

.1-2); US Department of Commerce, Bureau of the Census, 1990 Census of nce 2-1-3)). The 10 mile area includes portions, or all of, New London and s in Connecticut and a small portion on Suffolk County of Fishers Island which of Southold, New York. Figure 2.1-5 shows counties and towns within the 10 pulations and population densities are provided in Table 2.1-2. rford, in which Millstone Unit Number 1 is located, contained a total 30 people in 1990 at an average density of 547 people per square mile (US mmerce Bureau of the Census 1991) (Reference 2.1-3). The population growth mall with the 1990 total representing only a 0.5 percent increase over its 1980 red to towns immediately surrounding it, with the exception of New London, lowest increase in population between 1980 and 1990 (US Department of of the Census, 1991 (Reference 2.1-3)). has been consistently slowing down over the past 30 years, as shown in Table owth is projected by state demographers to continue at a low rate through the h time the population is expected to reach 18,480. After that, it is projected to tion. By the year 2010 (the last year of projections), the town's population is 080 (Connecticut Office of Policy and Management, Interim Population Reference 2.1-2)). Population distribution by sector for the area within 10 Unit Number 1 is shown for the years 1990, 2000, 2010, 2020 and 2030 in gh 2.1-8, that are keyed to the population sectors identified in Figure 2.1-6. 2.1-2 Rev. 3

n Distribution Within 50 Miles miles of Millstone Unit Number 1 includes portions, or all, of eight counties in ounties in Rhode Island and one county in New York. Figure 2.1-7 shows within the 50 mile area. In 1990, the 50 mile area contained approximately U.S. Department of Commerce), 1990 Census of Population and Housing This population is projected to increase to about 3,223,654 by the year 2030 e of Policy and Management, 1991 (Reference 2.1-2); New York State nomic Development, 1989 (Reference 2.1-1); Rhode Island Department of 89 (Reference 2.1-5); US Department of Commerce, 1990 Census of using, 1991 (Reference 2.1-4)). Population distribution by sector for the area Millstone Unit Number 1 is shown for the years 1990, 2000, 2010, 2020 and -9 through 2.1-13, which are keyed to the population sectors identified in tion and projections within the 50 mile region surrounding Millstone Unit culated based on population by municipalities and were assigned to sectors allocation. Projections for the 50 mile area were based on country-wide Population n increases resulting from an influx of summer residents total approximately many of the beaches and recreation facilities in the area are used by residents, ot represent any increase in population but instead a slight shift in population. r, a number of schools, industries, and recreation facilities which create daily ions in sector populations. Tables 2.1-14 through 2.1-16 show annular sector ns resulting from school enrollments, industrial employment, and recreation umented attendance). ulation Zone n zone (LPZ) surrounding Millstone Unit Number 1 encompasses an area ance of about 2.4 miles. The distance was chosen based on the requirements of gure 2.1-8 shows topographical features, transportation routes, facilities, and the LPZ. 2.1-3 Rev. 3

variations due to transient population are minimal within the LPZ. Several d within the area; however, they are predominantly used by local residents and facilities for parking or accommodation of large groups. Three schools, Great nd Southwest Elementary in Waterford, and Niantic Elementary in East Lyme, the LPZ. Major employment consists of the Camp Rell Military Reservation um. The New London Country Club is also located within the LPZ. n Center tion center to Millstone Unit Number 1 (as defined by 10 CFR 100 to contain esidents) is the city of New London which contained a 1990 population of n average population density of 5,189 people per square mile (US Department au of the Census 1991). The distance between Millstone Unit Number 1 and rporate boundary is about 3.3 miles to the northeast, just beyond the minimum nt set by 10 CFR 100. 50 miles of Millstone Unit Number 1 includes portions, or all, of 11 tical Area's. The populations of these areas are shown in Table 2.1-19. ulation centers within 50 miles of Millstone Unit Number 1, containing 25,000 1990. They are listed in Table 2.1-20 with the populations indicated. the area within 50 miles of Millstone was approximately 2,800,000 in 1990, nsity of 361 people per square mile. This density is lower than the NRC of 500 people per square mile (NRC Regulatory Guide 1.70, Revision 3, Within 30 miles of Millstone, the population density is considerably less, at an ple per square mile. By 2030, the 50 mile population is projected to increase to erage population density of about 410 people per square mile, considerable C comparison figure for end-year plant life of 1,000 people per square mile. e average density will be 223 persons per square miles by the year 2030. s by sector for 1990 and 2030 are shown for within 10 miles of Millstone in 2.1-23 respectively, which are keyed to Figure 2.1-6, and for within 50 miles of s 2.1-23 and 2.1-24, respectively, which are keyed to Figure 2.1-7. Cumulative s 1990 and 2030 are shown in Tables 2.1-25 and 2.1-26, respectively. 2.1-4 Rev. 3

erstate highway (Interstate 95), passenger and freight railroad lines, gas bove ground gas and oil storage facilities and two major waterways (Long mes River) in the vicinity of the Millstone site. gas transmission lines, oil transmission or distribution lines, under ground gas rilling or mining operations, or military firing, or bombing ranges near the site. d routes are shown of Figures 2.1-10 and 2.1-11. Figure 2.1-12 shows the road m in the area of the Millstone site. on of Facilities significant industrial, transportation, military, and industrial related facilities, aterials used, is shown in Table 2.1-27 as listed below. ical Corporation of Allen Point, Ledyard, Connecticut is located on the east Thames River approximately 10 miles north-northeast of the site. Dow mploys approximately 115 people and produces organic compounds, such as rofoam, and a base product of latex paints. All materials are moved to and from y by truck and/or railroad. oration of Eastern Point Road, Groton, Connecticut is located on the east bank es River, approximately 4.9 miles east-northeast of the site. Pfizer Corporation proximately 3,000 persons and produces organic compounds and tical materials, such as citric acid, antibiotics, synthetic medicines, vitamins

e. All materials are moved to and from Pfizer corporation by truck and/or at Division of General Dynamics of Eastern Point Road, Groton, Connecticut pproximately 5 miles east-northeast of the site. Electric boat employs ely 12,000 persons, and is a producer of submarines and oceanographic for commercial industry and the U.S. Navy. The nature of products produced at at requires that they handle substantial amounts of nuclear material which is der the Naval Reactors Division. All material is moved by truck, railroad, and/

and from the company with the exception of completed ships which leave own power. 2.1-5 Rev. 3

Thames River, is approximately 4 miles northeast of the site. Approximately are employed there on a full-time basis. The New London Transportation large complex in downtown New London in the City Pier area. It encompasses acilities, including a train station, several ferry companies, commercial and t slips, an interstate bus terminal, local bus inter-changers, and commercial ortation facilities. It serves as the prime entrance and exit for New London for commercial travel. Submarine Base, Groton, Connecticut is located on the east bank of the ver, approximately 7 miles northeast of the site. The base population includes ely 8,500 military personnel. In addition, there are about 1,800 civilian at the base. The U.S. Navy Submarine Base provides logistics as well as d operation of the base and its ships (nuclear and non-nuclear). All materials are ruck, railroad, barge and / or ship, to and from this government installation. oast Guard Academy, New London, Connecticut is located on the west bank of River, approximately 5.6 miles northeast of the site. Approximately 900 d the academy, while approximately 360 military and civilian personnel are ere. located approximately 2 miles northwest of the site, is a training headquarters necticut Army National Guard. It is owned and operated by the Military t of the State of Connecticut. On a full-time basis, it employs 16 persons d civilian), including the headquarters for the Connecticut Military Academy, ions personnel, and 745th Signal Company. On a part-time basis, during ekends, Camp Rell is occupied by varying numbers of troop units for ive training maneuvers, billeting, and supply functions for the Connecticut onal Guard. During the training maneuvers there may be from 300 to 1,200 e facility. Camp Rell is an administrative training center for troops of the t Army National Guard. Because of the solely administrative nature of its the camp's operation has no effect on the Station's operation. to Camp Rell, the Military Department of the State of Connecticut also field training facility known as Stone's Ranch Military Reservation, located ely 7 miles northwest of the site. Fourteen persons are employed there full-o regional motor vehicle and equipment maintenance shops. It is also occupied me basis by varying numbers of troop units for periods of field training for the t Army National Guard. During some weekend training sessions there may be eople at the facility. 2.1-6 Rev. 3

operations. Because of its distance from the site, the limited quantity of ored and used, and the type of aircraft operations occurring at the facility, ch Military Reservation does not pose any hazard to the Millstone station. orporation of Eastern Point Road, Groton, Connecticut is located on the east Thames River, approximately 5 miles east-northeast of the site. It is located zer Corporation, and south of General Dynamics-Electric Boat Division and a fuel storage facility. There are about 14 persons employed there on a full time Oil Corporation operates a fuel distribution and storage facility for home and kerosene. There are large above ground tanks capable of storing heating l fuel oil, and kerosene. The fuel arrives by ships or barges and is distributed by e medium-sized propane storage area in the proximity of the Millstone site. roleum Company, is located in Waterford, approximately 2.5 miles northeast of Great Neck Road, and employs about 75 people. Hendel Petroleum Company uel distribution facility for commercial and residential use. There are 5 above ks (3-30,000 gallons and 2-16,000 gallons) which are capable of storing llons total of propane gas. The facility also stores 40,000 gallons of gasoline, gallons of Number 2 fuel oil. The propane for the facility arrives by train and s distributed by truck. lstone site, at the Fire Training Facility located approximately 2,800 feet to the protected area are two 1,000 gallon propane cylinders. The two cylinders are ply propane to the fire simulator.The Fire Training Facility was constructed in e purpose of training fire brigade members. The Training Facility consists of n "mock-ups" which replicate nuclear power plant fire hazards. Propane is used e "fireplaces." The two storage cylinders are positioned such that their ends are ay from the Millstone site. Both cylinders are above ground domestic storage esigned per ASME Code for Pressure Vessels, Section VIII Division 1-92. tation is a Fossil Fuel powered electric generating plant operated by t Light & Power Company in Montville, Connecticut. It is located on the west Thames River, approximately 9.5 miles north-northeast of the site. tely 67 people are employed there. It is capable of providing 498 MW of wer. The fuel is stored in three large above ground tanks, capable of storing ely 175,000 barrels of fuel each; two medium above ground tanks, capable of roximately 12,000 barrels of fuel each; and two small above ground tanks, 2.1-7 Rev. 3

miles from the site, located along Rope Ferry Road in Waterford. This 35 psi e is a 6-inch plastic pipeline, buried approximately 3 feet deep. The control s located at the intersection of Clark Lane and Boston Post Road in Waterford. tribution line, ends at and serves the shopping center complex, near the and Parkway North, approximately 4 miles north of the site. This 35 psi gas an 8 inch plastic pipeline buried approximately 3 feet deep. The control valve ted at the complex where it intersects with Parkway North. nsmission or distribution lines within 5 miles of the Millstone site. ys he site in the shipping channels of Long Island Sound are of two types: general ually partially unloaded, with drafts of 20 to 25 feet, and deep draft tankers 38 feet. Both of these classes of ships must remain at least 2 miles offshore to round on Bartlett Reef. to the shore side of Bartlett Reef, and since there are no tank farms in Niantic pass with 2 miles of the site. vicinity of the site has been diminishing over the past several years due to the ount of oil used by area facilities. Barge traffic is heaviest during the winter ges only 1 barge per day during these months. On the average of once a month,

 ,000 barrels of sulfuric acid is towed past the site outside of Bartlett Reef.

ships per day traverse the Reef in the vicinity, 6 miles of the site. it is concluded that shipping accidents would not adversely affect Millstone 3 ities. don Airport, approximately 6 miles east-northeast of the site, handles regularly cial passenger flights. It is served by U.S. Air Express. It has two runways: 0 feet long; and 15-33, which is 4,000 feet long. Both runways are illuminated. ower at Groton / New London, with ILS (Instrument Landing System) and requency Omni Range) navigation aides located on the airfield. The ILS is way 5. As shown on Figure 2.1-10, the landing patterns used do not direct lstone site. 2.1-8 Rev. 3

approximately 4,490 military flights, approximately half of which were

s. Millstone Station is not in the flight path of these flights, and pilots are e site.

e 2.1-11, the air lane nearest the site is V58 which is approximately 4 miles

e. Other adjacent air lanes include V16, which is approximately 6 miles te, and V308, which is approximately 8 miles east of the site.The nearest high-121-581, passes approximately 9 miles southeast of the site. A second jet route, imately 12 miles northwest of the site.

s e Millstone site is served by interstate, state and local roads. These are shown he nearest major highway which would be used for frequent transportation of s is U.S. Interstate 95, which is located 4 miles from the Millstone site. Other which pass near the site include U.S. Highway 1 which is located 3 miles from Highway 156, located 1.5 miles from the site. istances exceed the minimum distance criteria given in Regulatory Guide 1.91, vide assurance that any transportation accidents resulting in explosions or toxic k size shipments of hazardous materials would not have a significant adverse peration or shutdown capability of the unit. d from east to west by a Providence & Worcester (P&W)/Amtrak railroad mainline tracks are more than 2,000 feet from the Millstone Unit Number 1 tructure. or the rail stock is both diesel and electric locomotives. Overhead electric lines These lines affect neither the site nor the overhead transmission lines leaving ing the railroad right-of-way above the tracks. Transportation and P&W/Amtrak have been contacted for information fic on the mainline tracks. Approximately eighteen scheduled passenger trips the tracks near the Millstone site. e freight train per day passes by the site. Hazardous material shipped on the ine, anhydrous ammonia, carbon dioxide, propane, ethyl alcohol, rosin, 2.1-9 Rev. 3

erves the Millstone Nuclear Power Station exclusively. The switch for that spur through traffic. In order to reach any station facility, a train car must also pass witch, which is normally set to direct traffic past the station to a dead end near re, the possibility of unauthorized transport of hazardous materials on the spur crossings on or adjacent to the site at which hazardous materials might be the tracks. ns of Industrial Growth cilities is presently planned in the area for oil distribution within the n of Connecticut. The gas distribution line along Rope Ferry Road ends at hool, approximately 2.9 miles from the Millstone site. The gas distribution line y North ends at, and serves the shopping complex approximately 4 miles from tioned, ship and barge traffic in the area of Millstone site has decreased over ars. No new ship or barge traffic is anticipated at this time in the Niantic Bay d Sound near location of the intake structures. cilities at Groton / New London Airport is proposed although some he facility, such as expansion of the approach lights, and upgrading of the ays in planned. Southeastern Connecticut Regional Planning Agency (SCRPA) master plan be prepared for the airport before any major physical made. The agency has previously adopted the policy that Groton / New London ain a small feeder airport providing connection to larger airports and direct number of cities with a 500 mile radius. 2.1-10 Rev. 3

about 0.5 miles from the Millstone Unit Number 1 Reactor building and lant. Traffic on the spur of the mainline track which extends onto the site is mize the possibility of railroad traffic-related accidents.

 /Amtrak railroad serves the Millstone Nuclear Power Station exclusively. The is normally set for through traffic. To reach any station facility, the locomotive a second switch, which is normally set to direct traffic past the station to a dead
 . Therefore, the possibility of unauthorized transport of hazardous materials he spur.

ls that are shipped on the track which crosses the site between New Haven and de chlorine, anhydrous ammonia, carbon dioxide, propane, ethyl alcohol, rosin, and hydrochloric acid. Among these materials, only the shipment of propane per year) is in the frequently shipped quantities of hazardous material d in Regulatory Guide 1.78. highway which would be used for frequent transportation of hazardous terstate 95, which is located at a distance of 4 miles from the Millstone site. tance exceeds the minimum distance criteria given in Regulatory Guide 1.91, erefore, provides assurance that any transportation accidents resulting in size shipments of hazardous materials will not have an adverse effect on the e plant. e of Groton / New London airport and the location of flight paths, the impact of stone Unit Number 1 is highly unlikely. gas transmission lines within 5 miles of the site. The nearest low pressure gas 2.9 miles from the site and is located near Waterford High School on Rope smission line is approximately 5 miles from the site in Groton Connecticut. miles or more away from the site, both the major gas and oil transmission lines t to the safe conduct of activities associated with storage of irradiated fuel or of Millstone Unit Number 1 or to the site in general. 2.1-11 Rev. 3

pography of the site is about the same grade as the rail line and therefore would flow of the cloud toward the plant site. CES State Department of Economic Development, Interim County, MSA and jections, 1980 - 2010, 1989. t Office of Policy Management, Interim Population Projections Series 91.1, tment of Commerce, Bureau of the Census, 1990 Census of Population, P.L. nts by Census Block, 1991. tment of Commerce, Bureau of the Census, 1990 Census of Population and Connecticut, 1990 CPH-1-8, 1991. nd Department of Administration, Projections by County, 1990 - 2020, 1989. gical Survey, 7.5-Minute Quadrangle Maps. ar Regulatory Commission, Regulatory Guide 1.70, Revision 3. Number RA-01-2-7, 1972. Association of American Railroads and Railway stitute Final Phase 01 Report on Summary of Ruptured Tank Cars Involved in ents, Revised July 1972. Chicago, IL. Number RA-02-2-18, 1972. Association of American Railroads and Railway stitute Final Phase 02 Report on Accident Review, Chicago, IL. ubber Company, 1972. Handbook of Chemistry and Physics 44th and 53rd

80. Hazardous Materials Link Report between New Haven and New London, t from January 1978 through June 1979.

A. et al 1980. An Assessment of the Risk of Transporting Propane by Truck and rt prepared for the U.S. Department of Energy by Pacific Northwest Battelle Memorial Institute. 2.1-12 Rev. 3

ort 3023, 1978. Workbook for Estimating the Effects of Accidental Explosions nt Ground Handling and Transport Systems. R-72-6, 1971.National Transportation Safety Board Railroad Accident Report n, TX. R-1, 1972.National Transportation Safety Board Accident Report for East St. R-75-7, 1974. National Transportation Safety Board Railroad Accident Report n, TX. R-79-11, 1979. National Transportation Safety Board Railroad Accident Report ew, FL. R-81-1, 1980. National Transportation Safety Board Railroad Accident Report ugh, KY. 800, 1981. Standard Review Plan: Evaluation of Potential Accidents (Section ilton 1973. Chemical Engineers Handbook, 5th Edition McGraw-Hill, Inc. ommunication between S.N. Bajpai and Robert Folden, Federal Railroad tion, Office of Safety, February 17, 1982. nd Special Programs Administration, U.S. Department of Transportation,

 , D.C. 1981. Computer Printout of Incidents Involving Deaths, Injuries, reater than $50,0000 or Evacuations. Run Dated March 26, 1981., Covering ember 22, 1970 to September 5, 1980.

nd Special Programs Administration, U.S. Department of Transportation,

 , D.C. 1981. Computer Printout of Incidents Involving Fire and Explosions
. Run dated 4/15/81 Covering Period June 6, 1973 through November 1, 2.1-13                                       Rev. 3

ntal Criteria and Assessment Office. U.S. EPA-600/8-80 p 6-150. tment of Transportation. Incidents Involving LPG and Ammonia, Computer red for Stone & Webster, 1981. etts Institute for Social and Economic Research, Revised Projections of the of Massachusetts Cities and Towns to the year 2000, 1991. ment of Commerce, Bureau of the Census, State and Metropolitan Area Book tistical Abstract Supplement, 1991. ment of Commerce, Bureau of the Census, 1990 P.L. 94-171 Counts by ty - New York, 1991. ment of Commerce, Bureau of the Census, 1990 Census P.L. 94-171 Counts by ty - Rhode Island, 1991. ment of Commerce, Bureau of the Census, Number of Inhabitants: Connecticut, 1971; PC80-1-A8, 1981. 2.1-14 Rev. 3

Page 1 of 1 Rev. 2 1990 Population Density 1980-1990 Total (People/Square Mile) Change (%) 15,340 451 10.6 45,144 1,442 9.9 14,913 391 8.6 1,949 61 7.0 16,673 397 1.3 28,540 5,189 -1.0 6,535 283 6.1 9,552 637 2.9 17,930 547 0.5 rk 19,836 394 3.5 Census of Population and Housing. population of all municipalities totally or partially within 10 miles of the site. Page 1 of 1 Rev. 2

6,782 11,399 13,870 15,340 68.1 21.7 10.6 29,937 38,523 41,062 45,144 28.7 6.6 9.9 5,395 14,558 13,735 14,913 169.8 -5.7 8.6 1,183 1,484 1,822 1,949 25.4 22.8 7.0 7,759 15,662 16,455 16,673 101.9 5.1 1.3 34,182 31,630 28,842 28,540 -7.5 -8.8 -1.0 3,068 4,964 6,159 6,535 61.8 24.1 6.1 5,274 8,468 9,287 9,552 60.6 9.7 2.9 15,391 17,227 17,843 17,930 11.9 3.6 0.5 pulation, Number of Inhabitants, Connecticut, PC80-1-A8, 12/81. pulation, Number of Inhabitants, Connecticut, PC10-A8, 4/71. ion and Housing Counts, Connecticut, PHC80-V-8, 3/81. pulation and Housing, Connecticut, CPH-1-8, 7/91. Page 1 of 1 Rev. 2

722 866 784 116 213 542 209 536 1,717 5,721 359 1,146 1,978 1,861 1,622 2,242 2,242 2,192 3,142 16,221 455 839 3,888 10,584 7,752 8,164 8,129 911 1,961 42,646 455 292 4,963 971 7,186 3,748 3,748 1,008 2,662 24,354 636 413 1,804 193 552 0 63 1,434 904 5,999 143 36 0 0 0 0 0 115 214 508 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 14 0 0 0 0 0 0 0 14 0 489 91 86 312 472 158 0 74 1,682 178 1,061 1,014 440 763 476 562 881 408 5,782 476 1,165 1,946 346 239 211 1,654 509 4-17 6,981 634 873 1,192 1,140 644 599 101 209 81 5,473 314 892 522 646 918 221 429 456 314 4860 4,372 8,086 18,200 16,383 20,201 16,098 16,594 8,251 11,894 120,443 Page 1 of 1 Rev. 2

778 932 845 126 230 582 225 578 1,852 6,166 387 1,234 2,131 2,006 1,749 1,796 2,415 2,366 3,389 17,487 489 905 4,191 11,441 7,359 8,802 8,765 983 2,115 46,203 492 314 5,352 1,045 7,746 4,041 3,285 1,087 2,870 26,256 685 444 1,944 208 597 0 68 1,546 975 6,467 154 39 0 0 0 0 0 125 233 551 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 14 0 0 0 0 0 0 0 14 0 528 98 92 336 509 169 0 78 1,810 192 1,144 1,093 473 821 513 606 950 436 6,228 514 1,255 2,118 373 258 227 1,783 548 448 7,524 684 940 1,285 1,229 695 646 108 226 88 5,901 304 961 564 696 990 238 462 491 339 5,239 4,715 8,710 19,621 17,663 21,781 17,354 17,886 8,900 12,823 129,846 Page 1 of 1 Rev. 2.1

803 961 871 129 237 600 230 595 1,908 6,352 399 1,272 2,197 2,068 1,804 1,853 2,492 2,437 3,495 18,301 504 930 4,321 11,767 8,617 9,074 9,036 1,013 2,180 47,626 507 324 5,518 1,078 7,988 4,166 3,387 1,119 2,960 27,072 707 458 2,005 215 616 0 70 1,593 1,005 6,669 159 41 0 0 0 0 0 138 255 593 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 15 0 0 0 0 0 0 0 15 0 54 102 95 346 525 175 0 79 1,867 198 1,179 1,126 440 488 847 530 625 443 6,417 529 1,294 2,184 385 266 234 1,838 566 461 7,757 705 969 1,325 1,267 716 666 111 232 90 6,081 350 992 582 718 1,021 245 476 506 350 5,403 4,861 8,980 20,231 18,210 22,458 17,893 18,440 9,180 13,226 133,883 Page 1 of 1 Rev. 2

828 990 899 133 243 620 236 613 1,968 6,549 411 1,310 2,264 2,132 1,860 1,909 2,569 2,513 3,602 18,584 519 960 4,455 12,134 8,885 9,355 9,318 1,044 2,247 49,105 523 333 5,689 1,220 8,236 4,296 3,492 1,151 3,052 27,907 728 472 2,067 222 635 0 72 1,642 1,036 6,874 162 41 0 0 0 0 0 144 268 615 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 15 0 0 0 0 0 0 0 15 0 562 105 98 356 541 180 0 80 1,922 205 1,216 1,161 504 874 546 644 1,011 450 6,611 544 1,226 2,252 398 274 242 1,895 583 476 8,000 727 998 1,365 1,308 738 687 114 239 93 6,269 361 1,023 600 738 1,053 253 491 523 362 5,572 5,008 9,256 20,857 18,777 23,154 18,449 19,011 9,463 13,634 138,023 Page 1 of 1 Rev. 2

855 1,021 927 136 250 638 242 631 2,027 6,746 425 1,351 2,334 2,196 1,916 1,968 2,650 2,590 3,712 19,156 535 990 4,592 12,510 9,160 9,644 9,606 1,075 2,315 50,620 539 343 5,866 1,145 8,492 4,428 3,598 1,188 3,147 28,772 751 487 2,132 229 655 0 73 1,692 1,068 7,087 167 43 0 0 0 0 0 151 281 642 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 15 0 0 0 0 0 0 0 15 0 580 108 101 366 558 185 0 81 1,979 212 1,254 1,197 520 901 561 663 1,043 458 6,809 560 1,377 2,323 409 281 249 1,956 602 490 8,247 748 1,029 1,407 1,349 761 708 116 246 95 6,459 371 1,055 618 761 1,085 261 507 539 374 5,745 5,163 9,545 21,504 19,356 23,867 19,015 19,596 9,757 14,048 142,277 Page 1 of 1 Rev. 2

1 22,283 26,357 32,610 18,658 105,629 21 34,824 23,730 27,465 35,598 137,838 48 9,444 11,334 29,987 199,334 292,947 54 23,914 16,498 43,001 99,721 207,488 9 10,712 7,992 10,920 0 35,623 0 0 836 0 1,344 0 807 0 0 807 0 2,420 0 0 2,420 1,614 13,541 0 0 15,155 2,443 12,569 14,807 4,498 34,317 938 22,042 8,252 143,933 175,179 2 2,471 0 0 20,389 24,542 2 27,956 34,384 184,723 267,465 520,310 1 12,474 27,895 148,259 259,824 455,433 3 6,215 31,331 191,767 365,578 600,364 0 8,809 17,850 115,424 78,820 225,762 ,443 164,097 248,750 808,051 1,493,818 2,835,159 Page 1 of 1 Rev. 2

0-10 10-20 20-30 30-40 40-50 Total 6 24,028 28,707 35,404 20,273 114,578 87 37,551 25,721 29,926 38,135 148,820 03 10,183 12,196 31,611 206,940 307,133 56 25,744 17,633 45,998 105,848 221,509 7 11,497 8,553 11,687 0 38,204 0 0 895 0 1,446 0 878 0 0 878 0 2,635 0 0 2,635 1,759 14,742 0 0 16,501 2,660 13,688 16,122 4,897 37,367 1,022 24,000 8,985 156,725 190,746 0 2,641 0 0 22,201 26,652 8 29,887 36,343 195,006 281,709 549,173 4 13,340 29,762 156,623 273,153 480,402 1 6,660 33,435 200,205 380,339 626,540 9 9,492 19,194 121,620 83,732 239,277 ,846 176,464 267,517 854,082 1,573,952 3,001,861 Page 1 of 1 Rev. 2

0-10 10-20 20-30 30-40 40-50 Total 2 24,773 300,056 36,785 21,101 119,067 31 38,716 26,730 31,421 39,720 154,618 26 10,499 12,626 32,221 210,368 313,340 72 26,652 18,530 48,258 109,494 230,006 9 11,986 8,981 12,272 0 39,908 0 0 940 0 1,533 0 920 0 0 920 0 2,761 0 0 2,761 1,847 15,445 0 0 17,292 2,788 14,344 16,896 5,132 39,160 1,073 25,151 9,416 164,248 199,903 7 2,689 0 0 23,267 27,823 7 30,426 37,096 199,100 286,889 559,928 7 13,590 30,311 159,776 278,156 489,590 1 6,807 34,052 202,762 384,902 634,604 3 9,778 19,778 123,964 85,735 244,658 ,883 181,624 276,781 873,811 1,609,012 3,075,111 Page 1 of 1 Rev. 2

0-10 10-20 20-30 30-40 40-50 Total 9 24,541 31,470 38,219 21,963 123,742 84 39,916 27,784 32,989 41,349 160,622 05 10,825 13,051 32,748 213,221 318,950 07 27,557 19,336 50,343 112,285 234,428 4 12,452 9,376 12,811 0 41,513 0 0 981 0 1,596 0 965 0 0 965 0 2,894 0 0 2,894 1,939 16,184 0 0 18,123 2,922 15,033 17,707 5,379 41,041 1,127 26,355 9,869 172,131 209,497 2 2,737 0 0 24,383 29,042 1 30,974 37,863 203,283 292,190 570,921 0 13,844 30,871 162,992 283,254 498,961 9 6,957 37,678 205,354 389,518 642,776 2 10,070 20,382 126,369 87,794 250,187 ,023 186,861 286,242 893,665 1,643,467 3,148,258 Page 1 of 1 Rev. 2

0-10 10-20 20-30 30-40 40-50 Total 6 26,332 32,953 39,716 22,860 128,607 56 41,155 28,879 34,637 43,058 166,885 20 11,159 13,494 33,286 219,112 324,671 72 28,495 20,176 52,519 115,158 245,120 7 12,937 9,789 13,375 0 43,188 0 0 1,024 0 1,666 0 1,011 0 0 1,011 0 3,033 0 0 3,033 2,036 16,957 0 0 18,993 3,062 15,755 18,558 5,637 43,012 1,183 27,619 10,342 180,394 219,553 9 2,787 0 0 25,554 30,320 9 31,532 38,647 207,551 297,607 582,146 7 14,102 31,441 166,276 288,449 508,515 9 7,110 35,317 207,981 394,192 651,059 5 10,373 21,003 128,835 89,919 255,875 ,277 192,263 296,074 914,100 1,678,940 3,223,654 Page 1 of 1 Rev. 2

0 0 374 897 2,073 174 0 0 444 3,962 0 636 210 697 1,352 1,542 534 0 0 4,971 0 0 2,501 0 888 0 1,043 1,609 266 6,307 181 0 0 0 1,330 0 0 183 0 1,805 0 0 0 0 0 0 0 0 0 68 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 263 0 864 0 1,127 0 345 0 0 0 0 0 0 0 345 0 0 843 0 0 0 0 0 0 843 0 0 298 1,250 0 0 0 0 0 1,548 602 981 4,226 2,844 5,643 1,979 1,651 2,656 1,191 21,773 rollment only. tment of Education listing of schools: Telephone survey conducted in March 1992. Page 1 of 1 Rev. 2

0 0 300 0 0 0 0 0 200 500 0 0 0 0 0 375 375 109 277 1,134 0 375 80 831 0 375 375 0 0 2,036 0 0 0 8,800 5,500 820 0 0 0 15,120 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 68 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 256 0 256 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 125 125 0 0 250 500 0 843 0 0 125 125 0 0 750 0 0 0 0 0 0 0 0 0 0 500 375 380 9,631 5,500 1,820 1,000 363 477 20,046 0 employees or more. Excludes plant employee population. e suvey conducted in March 1992. Page 1 of 1 Rev. 2

Y LOCATION ATTENDANCE ATTENDANCE ENE/E 6-8 97,641 490

• ENE 5-6 58,965 200
• ENE/E 7-9 11,675 60
• morial E 2-3 157,962 790
• W 3-5 412,495 2,360 **

WNW/NNW 7-10 81,146 400

• ttendance based on 90% of yearly attendance from April through September.

rs from April 15 to September 15. ut DEP - Office of Parks and Forests, 1990 Park Attendance. Page 1 of 1 Rev. 2

1,298 1,536 903 1,065 1,144 1,351 768 909 760 899 179 212 0 0 0 0 0 0 0 0 3 3 429 506 1,025 1,211 1,046 1,233 1,167 1,377 1,124 1,327 9,846 11,629 pulation and Housing of Policy and Management, Interim Population Projections Series 91.1, 4/91 Page 1 of 1 Rev. 2

NE 0 0 NE 0 75 NE 0 0 E 292 0 SE 0 0 SE 0 0 SE 0 0 S 0 0 SW 0 0 SW 0 0 SW 0 0 W 0 0 NW 345 0 W 0 500 NW 0 0 AL LPZ 947 0 Enrollment loyees or more. conducted in March 1992; Connecticut Department of Education school listing. Page 1 of 1 Rev. 2

ord, CT PMSA 443,722 79,488 I PMSA 157,272 SA 767,899 iden, CT MSA 530,240 NY PMSA 2,609,212 PMSA 148,188 rwich, CT-RI MSA 266,819 MSA 654,869 AS 221,629 PMSA 90,320 etropolitan Statistical Area. n Statistical Area. metropolitan areas completely or only partially within 50 miles of the site. Page 1 of 1 Rev. 2

Bristol 60,640 Cheshire 25,684 East Hartford 50,452 East Haven 26,144 Enfield 45,532 Glastonbury 27,901 Groton 45,144 Hamden 52,434 Hartford 139,739 Manchester 51,618 Meriden 59,479 Middletown 42,762 Milford 49,938 Naugatuck 30,625 New Britain 75,491 New Haven 130,474 New London 28,540 Newington 29,208 Norwich 37,371 Shelton 35,418 Southington 38,518 Stratford 49,389 Vernon 29,841 Wallingford 40,822 Waterbury 108,961 West Hartford 60,110 West Haven 54,021 Wethersfield 25,651 Page 1 of 2 Rev. 2

Cranston 76,060 Johnston 26,542 Newport 28,227 Warwick 85,427 West Warwick 29,268 Brookhaven 407,779 Southampton 44,976 ies with 25,000 people or more. Municipalities completely or only partially Census of Population and Housing. Page 2 of 2 Rev. 2

610 1,168 1,440 1,054 751 653 762 657 843 827 772 855 2,830 5,993 3,591 3,200 2,761 273 526 2,183 772 298 3,612 550 3,328 1,469 1,035 302 714 1,241 1,080 421 1,313 109 256 0 21 430 242 306 243 37 0 0 0 0 0 34 57 26 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 14 0 0 0 0 0 0 0 1 0 498 66 49 145 185 54 0 20 86 302 1,082 738 249 353 186 191 264 109 295 808 1,118 1,429 196 111 83 562 153 112 356 1,076 890 868 646 298 235 34 63 22 279 533 909 380 366 425 87 146 137 84 248 464 515 828 580 585 394 352 155 199 384 us of Population Page 1 of 1 Rev. 2

722 1,377 1,700 1,243 887 771 900 776 995 976 908 1,009 3,345 7,084 4,243 3,780 3,263 322 621 2,579 915 350 4,272 648 3,933 1,736 1,222 356 844 1,466 1,275 496 1,553 130 303 0 25 507 286 361 284 44 0 0 0 0 0 45 75 33 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 15 0 0 0 0 0 0 0 1 0 591 79 57 170 219 63 0 22 101 360 1,278 872 294 417 220 225 313 123 347 951 1,404 1,692 232 130 98 664 180 131 420 1,270 1,049 1,025 764 352 278 39 74 25 329 630 1,075 450 431 503 102 172 162 100 293 548 608 979 685 691 466 416 183 235 453 of Policy and Management, Interim Population Projections Series 91.1, 4/91. Page 1 of 1 Rev. 2

NNE 827 591 242 200 202 281 NE 2,183 160 116 218 1,129 597 ENE 1,241 406 168 313 564 423 E 306 182 81 79 0 73 ESE 26 0 0 6 0 3 SE 0 0 8 0 0 2 SSE 0 0 25 0 0 5 S 0 27 138 0 0 31 SSW 0 41 128 108 25 70 SW 1 16 225 60 815 357 WSW 86 42 0 0 115 50 W 295 475 350 1,345 1,514 1,061 WNW 356 212 284 1,079 1,471 928 NW 279 106 319 1,396 2,070 1,224 NNW 248 150 182 840 446 460 Average 384 174 158 368 528 361 us of Population and Housing. Page 1 of 1 Rev. 2

976 699 294 252 244 340 2,579 190 138 242 1,224 662 1,466 484 206 382 652 499 361 220 100 97 0 88 33 0 0 7 0 3 0 0 10 0 0 2 0 0 31 0 0 6 0 35 173 0 0 39 0 52 161 135 32 88 1 20 81 75 1,021 447 101 47 0 0 145 62 347 536 394 1,511 1,685 1,187 420 240 320 1,210 1,633 1,036 329 121 360 1,514 2,232 1,327 293 176 214 938 509 522 453 204 189 416 594 410 of Management, interim Population projections, Series 91.1, 4/91. Page 1 of 1 Rev. 2

292 378 269 237 106 215 827 591 242 200 202 281 2,183 160 116 218 1,129 597 1,241 406 168 313 564 423 306 182 81 79 0 73 26 0 0 6 0 3 0 0 8 0 0 2 0 0 25 0 0 5 0 27 138 0 0 31 0 41 128 108 25 70 1 16 225 60 815 357 86 42 0 0 115 50 295 475 350 1,345 1,514 1,061 356 212 284 1,079 1,471 928 279 106 319 1,396 2,070 1,224 248 150 182 840 446 460 384 174 158 368 528 361 us of Population and Housing. Page 1 of 1 Rev. 2

976 768 505 394 340 2,579 787 426 346 662 1,466 730 438 414 499 361 255 169 138 88 33 8 4 5 3 0 0 6 3 2 0 0 17 10 6 0 26 108 60 39 0 39 107 119 88 1 15 163 125 447 101 61 27 15 62 347 488 436 906 1,187 420 285 305 701 1,036 329 173 277 818 1,327 293 205 210 529 522 453 226 223 307 410 Page 1 of 1 Rev. 2

Employed or Site Miles Stationed Corp Ledyard 115 10+ NNE ion Groton 3,000 4.9 ENE Division Groton 12,000 5 ENE amics) ondon Groton 153 6 ENE ull) New London 20 4 NE Center marine Groton 10,300 7 NE d New London 1,260 5.6 NE East Lyme 16 2 NW Military East Lyme 14 7 NW Facilities oration Groton 14 5 ENE eum Co. Waterford 75 2.5 NE ion Montville 67 10 NNE tion Plant Page 1 of 1 Rev. 2

Containing Hazardous Approx. No. of Cars per Material Materials Year 2.20 44 monia 0.266 5 2.466 49 Page 1 of 1 Rev. 2

of the Millstone 3 Final Safety Analysis Report of Reference 2.2-1). ETEOROLOGY of the Millstone 3 Final Safety Analysis Report of Reference 2.2-1). Influence of the Plant and Its Facilities on Local Meteorology mber 1 used a once-through cooling water system, discharging its cooling water arry into which Units 2 and 3 also discharge and thence into Long Island of steam fog occasionally form over the quarry and less frequently over the uring the winter months, depending on tidal conditions and temperature n air and water. This fog dissipates rapidly as it moves away from the warm e the maximum discharge plume (defined by the 1.5°F isotherm of temperature ll three Millstone units were at full power) is approximately an ellipse of 1500 rs, the extent of the steam fog is negligible. With the permanent shutdown of mber 1, this maximum discharge plume size is further reduced. teorological Conditions for Design and Operating Bases. Basis Tornado for the Millstone Unit Number 1 design basis tornado are: velocity 300 mph al velocity 60 mph ure drop 2.25 psi ssure drop 1.2 psi/sec ETEOROLOGICAL MEASUREMENTS PROGRAM is served by a common meteorological tower, located south of Millstone Unit teorological tower is capable of measuring wind speed, direction, and air ous heights. For details regarding the capability of the On Site Meteorological gram, see Section 2.3.3 of the Millstone 3 Final Safety Analyses Report ith the exception that Millstone Unit Number 1 no longer has the data and data recording capability to display parameters transmitted by modem/ 2.2-1 Rev. 2

sult in short-term releases of radioactivity from several possible venting points. sion factors (/Q) based on site meteorological data are calculated at the ndary (EAB) and low population zone (LPZ) for each downwind sector for The diffusion factors are calculated for different release time periods ength of the release. These diffusion factors are used in the calculation of quences of the releases. ons Point and Receptor Locations o be 3860 m in all sectors from any release point. re calculated using the basic methods of Regulatory Guide 1.145. /Q values nit 2 Control Room due to ground level releases were calculated using the y and Campe. (Reference 2.2-2). s used in design basis accident (DBA) radiological consequence calculations the list of assumptions in Chapter 5. RM (ROUTINE) DIFFUSION ESTIMATES oactivity are routinely released on a continuous basis from the Unit Number 1 and the SFPI ventilation exhaust point. Atmospheric diffusion factors (/Q) orological data are calculated for various downwind receptor locations of orological data is used to calculate the dose consequences to the public from fluents. The calculated doses are submitted periodically to the Nuclear ission (NRC). 2.2-2 Rev. 2

se rformed on a periodic basis using the actual meteorology for this period. und level dispersion factors, and releases are modeled using a conventional odel. CES nit 3, Final Safety Analysis Report, Section 2.3-Meterorology. G., and Campe, K. M. Nuclear Power Plant Control Room Ventilation System Meeting General Criterion 19, 13th AEC Air Cleaning Conference, 1973. 2.2-3 Rev. 2

of the Millstone 3 Final Safety Analysis Report, Reference 2.3-1). FACILITIES ocated on the north shore of Long Island Sound. To the west of the site is the east is Jordon Cove. Figure 2.3-1 shows the general topography of the e site grade elevation for Millstone Unit Number 1 varies from 14 feet to above vel (MSL). Section 2.3.3.2 discusses the probable maximum hurricane used to water levels. s the flood history in the vicinity of Millstone Point, flood design the effects of local intense precipitation. story ite has historically been caused by hurricanes. The maximum historical sult of a hurricane on September 21, 1938, which produced a flood level of 9.7 ondon, Connecticut. f flooding that could affect Millstone Unit Number 1 are direct rainfall and sign Considerations ent for flooding at the Millstone site is a storm surge resulting from the bable maximum hurricane (PMH) (see Section 2.3.6). The maximum 18.11 feet MSL, and the associated wave run up is +22.3 feet MSL. s the flooding protective features at Millstone Unit Number 1. Local Intense Precipitation e development of the probable maximum precipitation (PMP) for the site may n 2.3.2 of Reference 2.3-1. 2.3-1 Rev. 2

that the area east of Millstone Unit Number 1, north of the radwaste truck bay, -enclosed area just east of the Unit 2 Control Room would have maximum er of 15.5 to 16.2 feet. MSL. Further, these studies show that areas west of mber 1 and 2, south of Millstone Unit Number 1, extending around the gas the east side of Millstone Unit Number 1 north of the radwaste truck bay ess ponding on the order of 14.6 to 14.9 feet. MSL. Ponding at the intake negligible since runoff would flow directly to the adjoining Niantic Bay. nario, in-leakage through door openings could occur once the flood depths evations. Secured external and internal doors will have a tendency to limit or of in-leakage. E MAXIMUM FLOOD (PMF) ON STREAMS AND RIVERS of Reference 2.3-1, the Millstone Unit 3 Final Safety Analysis Report). AL DAM FAILURE, SEISMICALLY INDUCED of Reference 2.3-1, the Millstone Unit 3 Final Safety Analysis Report). E MAXIMUM SURGE AND SEICHE FLOODING Maximum Winds and Associated Meteorological Parameters characteristics used to calculate the probable maximum storm surge at the e are those associated with the PMH as reported by the U.S. National Oceanic dministration (NOAA) in their unpublished report HUR 7-97. HUR 7-97 as a hypothetical hurricane having that combination of characteristics the most severe that can probably occur in the particular region involved. The pproach the point under study along a critical path and at an optimum rate of lly, nine different PMH storm patterns can be constructed using wind speed, ward speed parameters given in HUR 7-97 in various combinations. The storm, the maximum surge buildup at the entrance to Long Island Sound is one with aximum wind and a slow speed of translation. Pertinent parameters are ssure Index 2.3-2 Rev. 2

eed

s. This is the rate of forward movement of the hurricane center.

Wind mph. This is the absolute highest surface wind speed in the belt of maximum Pressure inches. This is the surface atmospheric pressure at the outer edge of the here the hurricane circulation ends. ametric combinations give a higher wind speed, this particular combination urge. d Seiche Water Levels orms and squall lines cause tidal flooding in the Millstone Point area, by far the ng has resulted from hurricanes. For this reason, the PMH as defined in Section compute the design storm surge level at the site. The calculated total surge r level considers the wind setup, the water level rise due to barometric pressure ical tide and forerunner or initial rise. water level is +18.11 feet, and the associated wave run up elevation is +22.3 tion cs are dependent upon wind speed and duration, fetch length, and water depth. sheltered from the direct onslaught of open ocean waves by Long Island. peak surge, the wind is from the southeast direction and the wave attack would axis of the point. Thus the intake structure, and the southeast portions of the e Generator Buildings are primarily involved. 2.3-3 Rev. 2

xis of the point. Thus, the southeast portions of the Reactor Building would be Maximum Tsunami Flooding of the Millstone Unit 3 Final Safety Analysis Report Reference 2.3-1). CTS le history of ice or ice jams in Niantic Bay. WATER CANALS AND RESERVOIRS of the Millstone Unit 3 Final Safety Analysis Report Reference 2.3-1.) DIVERSIONS of the Millstone Unit 3 Final Safety Analysis Report Reference 2.3-1.) G PROTECTION REQUIREMENTS ER CONSIDERATIONS w in Rivers and Streams it Number 1 does not depend on either rivers or streams as a source of cooling is not applicable. er Resulting from Surges, Seiches, or Tsunamis one Unit 1. ON, DILUTION, AND TRAVEL TIMES OF ACCIDENTAL RELEASES OF FFLUENTS SURFACE WATERS. 2 of the Millstone Unit 3 Final Safety Analysis Report Reference 2.3-1.) 2.3-4 Rev. 2

or flood conditions. CES nit 3, Final Safety Analysis Report, Section 2.4 - Hydrologic Engineering. J. J. Shea to W. G. Counsil, Millstone Nuclear Power Station, Unit 1 - Safety Report on Hydrology SEP Topics II-3.A, II-3.B, II-3.B.1, and II-3.C, dated 82. 2.3-5 Rev. 2

RY GROUND MOTION ing vibratory ground motion at the Millstone site is presented in Section 2.52 . With the exceptions given below, that information is incorporated herein by Storage Earthquake lant is such that spent fuel pool remain intact during a ground motion of 0.17 g. FAULTING conditions of the Site Reference 2.4-1 discusses the stratigraphy, structural geology, and geologic re in detail. of Fault Offset logic maps which include the site area do not indicate the presence of faulting. lts discovered during excavation of the Millstone Unit Number 3 site can be 5.3.2 of Reference 2.4-1. This discussion can be considered typical for the e. kes Associated with Capable Faults ce of capable faults within the five-mile radius of the site. The majority of the activity has been associated with the White Mountain Plutonic Province. Some ssociated with the Ramapo fault system (Reference 2.4-2); however, the fault is able (Reference 2.4-3). tion of Capable Faults le faults within the site area. The faults uncovered in the excavation are n 2.5.3.2 of Reference 2.4-1. 2.4-1 Rev. 2.1

le faults within five miles of the site. uiring Detailed Faulting Investigation ault zones were uncovered during excavation at the Millstone Unit Number 3 ave been mapped in detail and are discussed in Section 2.5.3.2 of Reference f Faulting Investigation ce of capable faulting within the five mile radius of the site. The faults at the he rifting associated with the Triassic-Jurassic Period or older, with the last approximately 142 million years ago. Y OF SUBSURFACE MATERIALS AND FOUNDATIONS the stability of subsurface materials and foundations is available from the lstone Unit Number 1. A discussion of this subject for the Millstone Unit on can be found in Reference 2.4-1. This information can be considered typical te. Y OF SLOPES pes at the Millstone site was evaluated in Reference 2.4-4, wherein it was e are no natural or man-made slopes at the site that could be or become affect safety related structures, systems or components. MENTS AND DAMS or dams have been constructed at the Millstone site. CES nit 3, Final Safety Analysis Report, Section 2.5, Geology, Seismology, and al Engineering. 2.4-2 Rev. 2.1

gulatory Commission. Letter from J. Shea to W. G. Counsil dated June 30, Review Topic II-4, D, Stability of Slopes, Millstone Nuclear Power Station 2.4-3 Rev. 2.1

n Criteria (GDC) for Nuclear Power Plants as listed in Appendix A to 10 CFR May 21, 1971 and subsequently amended July 7, 1971. mber 1, was issued a provisional operating license (POL) on October 7, 1970, d to comply with the GDC (Reference 3.1-3). Therefore, Millstone Unit equired to seek exemptions for those areas where it does not comply with the n of the design bases of the Millstone Nuclear Unit Number 1, as compared to formed in support of the application for a full term operating license (FTOL),

1. It was concluded therein that Millstone Unit Number 1 satisfies and is in e intent of the GDCs. Nevertheless, it should be noted that this comparison and t a commitment to meet all of the current GDCs or even to meet the intent of Instead, the Reference 3.1-1 comparison determined the degree of compliance hat time. Also, compliance is demonstrated based upon those interpretations in he specific licensing question, or issue, was being addressed.

ic Evaluation Program and Three Mile Island Evaluations of General Design atic evaluation program (SEP) initiated by the NRC in 1977, a large number of pecific safety concerns were addressed and resolved (Reference 3.1-2). Many s, and later issues which arose from the Three Mile Island (TMI) accident, ration of the NRC GDC affected by a specific issue and how the plant design iteria. A compilation of this more recent evaluation of specific safety concerns DC are listed in Table 3.1-1. CATION OF STRUCTURES, SYSTEMS, AND COMPONENTS lassification al Regulations requires that structures, systems, and components important to gned to withstand the effects of earthquakes without loss of capability to safety functions. 10 CFR 100, Appendix A further defines a safe shutdown nd the structures, systems and components required to remain functional, as s necessary to ensure: The integrity of the reactor coolant pressure boundary, 3.1-1 Rev. 2.1

1.29, Revision 3, describes an acceptable method for identifying and lant features that should be designed to withstand the effects of an SSE. s and equipment, including their foundations and supports, are divided into two tegories: ructures and equipment whose failure could cause significant release of oactivity or which are vital to the removal of decay heat. tructures and equipment which are not essential to the containment of oactivity or removal of decay heat. uilding at and below elevation 108 feet 6 inches , the fuel pool liner and the main Seismic Class I in the permanently defueled condition. The Reactor rotects and supports the spent fuel pool. It supports maintenance of the fuel e fuel pool, provides protection from external hazards and supports ter in the fuel pool to a depth necessary to ensure the irradiated fuel is always l storage racks are designed to assure subcriticality in the fuel pool and are and the anticipated earthquake loadings as Class I structures. ing structure above elevation 108 feet 6 inches (enclosure) is classified as the permanently defueled plant condition. The Reactor Building above 6 inches provides a weather enclosure for the spent fuel pool and supports the erhead crane. However, it has no structural function in providing support for . Since the enclosure is no longer credited to provide secondary containment

 .2.2) and since its failure during a seismic event could adversely affect the its contents or adjacent safety related SSCs, the seismic design of the rized as Seismic II/I and is further discussed in Section 3.1.6.

Seismically Designed Structures, Systems and Components ed as Seismic Class I in the permanently defueled condition, the following to performing dismantlement operations: Downgrading seismic classification of components shall be performed in accordance with appropriate engineering and design procedures and processes. 3.1-2 Rev. 2.1

incident or an accident with offsite doses exceeding the doses from the design basis accident. When downgrading seismic classification of an SSC, a 10 CFR 50.54 evaluation shall be performed if the structure classification is described in the Quality Assurance Program (QAP). When downgrading seismic classification of an SSC, a 10 CFR 50.59 evaluation shall be performed if, during a seismic event, its failure has the potential to drain the fuel pool water level lower than 9 feet above the active fuel. lated Classification have traditionally been classified as safety related in accordance with 10 0 CFR 100, Appendix A, Section III, if they are relied upon to remain nd following design basis events to assure: ty of the reactor coolant pressure boundary, lity to shut down the reactor and maintain it in a safe shutdown condition, lity to prevent or mitigate the consequences of accidents which could result in f site exposures comparable to the applicable guideline exposures set forth in .34(a)(1) or 10 CFR 100.11. o parts of the safety related definition (reactor coolant pressure boundary, and ve and maintain safe shutdown) do not apply to a permanently defueled plant, estrictions of 10 CFR 50.82. The third part of the safety related definition nces comparable to 10 CFR 100 guidelines) is dependent on the results of new nt analysis assumptions and results developed to address the existing defueled C that are required to protect workers and the public from the consequences of may need to remain classified as safety related. ences of potential accidents were reanalyzed and it was concluded that the only is a fuel handling accident. This accident was analyzed assuming no ment or standby gas treatment system in operation, with a puff ground level 3.1-3 Rev. 2.1

d not change and that the fuel pool water remained inplace. This implies that of either the fuel pool structure or the fuel racks. Since these are passive sumption of failure is not required as long as the items are safety related and nd these loads. Therefore, the fuel pool and supporting structure, fuel pool acks must be considered as safety related to support the assumptions made in is. No other components, systems or structures meet this criterion. ty Related Plant Functions Maintained in the Defueled Condition afety Related criterion above, other non-safety related plant functions must be efueled condition. The following criteria were used to determine which SSC Is the SSC associated with storage, control or maintenance of nuclear fuel in a safe condition; or handling of radioactive waste? Is the SSC program associated with radiological safety? 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 or radiological safety? Does the SSC satisfy a requirement based on regulations governing management of nuclear fuel or radioactive materials, including any SSC which is independently required by the License or Technical Specifications? applied to all Millstone Unit Number 1 SSC. A positive response to any the Safety Related criterion, results in the group of those SSC which must portant to the Defueled Condition remain functional will be maintained in accordance with applicable Millstone cedures or quality processes. Commitments exist for augmented quality related FPQA), and Radwaste (RWQA). These requirements would apply to the s of the SSC which meet the criteria above. 3.1-4 Rev. 2.1

ntrol, maintenance or handling of nuclear fuel ntrol, maintenance or handling of radioactive waste, if not already RWQA al safety onents were reviewed for these functions. Note that application of these om the 4 criteria in that requirements apply only to the primary SSC and are pporting systems, equipment or structures. The intent of the ITDC augmented se reliable operation of the system(s) primarily responsible for performing each le performance of the supporting SSC are demonstrated during routine eriodic testing of the ITDC SSC. in regulatory requirements to which the licensee made a licensing commitment functional scope of an SSC (e.g., Emergency Plan, Security Plan, Quality

 , etc.). These commitments and legal requirements were also considered in the cess.

strictions and Limitations on use of the SSC reclassification criteria cation criteria is used as a basis to change various Millstone Unit Number 1 grams, provided that the change involves an SSC that is non-ITDC and, procedures contain an acceptable method for approving the change. The soft changes associated with non-ITDC SSCs are allowed: ications, s, ing items and corrective actions, ustry operating experience reports,

nts, 3.1-5 Rev. 2.1

reating new hazards or initiators not already recognized as part of the current s (e.g., decontamination or decommissioning of major components defined in .82) al removal/disassembly of existing SSCs, or the installation of new SSCs. t may provide the basis for initiating such a change. Technical Specification requirements. regulations, license conditions, rules, and permits until such time that relief is the regulating authority. However, it may provide the basis for requesting relief gulations, license conditions, rules, and permits. commitments. Application of the commitment change process is required to mitments. the QAP. However, it may provide the basis for initiating a change to the QAP. the Radiological Effluent Monitoring and Offsite Dose Calculation Manual M). However, it may provide the basis for initiating a change to the M. the Emergency Plan. However, it may provide the basis for initiating a change gency Plan. the Security Plan. However, it may provide the basis for initiating a change to y Plan. the Fire Protection Plan. However, it may provide the basis for initiating a he Fire Protection Plan. the Radiation Protection Program. However, it may provide the basis for change to the Radiation Protection Program. erfaces for ITDC SSCs ITDC that require availability shall be maintained in a state such that the al capability is maintained. 3.1-6 Rev. 2.1

trol: Measures will be invoked to assure applicable regulatory requirements, s, and design basis information is correctly translated into specifications, rocedures and instructions. These measures shall include provisions to assure riate quality standards are specified and included in design documents and that from such standards are controlled. Design changes, including field changes jected to engineered design control measures commensurate with the of the SSC. nt Document Control: Measures will be invoked to assure that applicable requirements, design basis, and other requirements which are necessary to uate quality are suitably included or referenced in the documents for nt of material, equipment, services. Procedures, and Drawings: Activities affecting SSCs will be prescribed by d instructions, procedures, or drawings, of a type appropriate to the ces and will be accomplished in accordance with these instructions, , and drawings. Instructions procedures, and drawings will include appropriate e or qualitative acceptance criteria for determining that important activities atisfactorily accomplished. Purchased Material, Equipment, and Services: Measures will be invoked to material, equipment, and services conform to the procurement documents. ures shall include provisions, as appropriate, for source evaluation and bjective evidence of quality furnished, inspection at the source, and n upon delivery. Inspection of activities affecting quality will be invoked and executed to verify ce with the documented instructions, procedures, and drawings for ing the activity. torage, and Shipping: Measures will be invoked to control the handling, pping, cleaning and preservation of material and equipment in accordance with nspection instructions to prevent damage or deterioration. l: Surveillance testing will be established for SSCs to ensure that the SSCs isfactorily commensurate with the importance of their intended function. 3.1-7 Rev. 2.1

ction is taken to preclude repetition. D TORNADO LOADINGS al Design Criteria 2), as implemented by Standard Review Plan (SRP) Sections Regulatory Guides (RG) 1.76 and 1.117, requires that the plant be designed to ts of natural phenomena such as wind and tornadoes. Number 1 capability to withstand wind and tornado loadings was evaluated in luation Program (SEP) (Reference 3.1-4) as Topic III-2. Several submittals

 .S. Nuclear Regulatory Commission (NRC) to address issues raised under that 3.1-6, 3.1-7, 3.1-8, and 3.1-9). In an evaluation dated November 25, 1985 the NRC concluded that the proposal will provide adequate protection against EVEL DESIGN n basis water level at Millstone Unit Number 1 is the probable maximum flood 0 feet above mean sea level (MSL). In the defueled condition, flooding of Unit ptable. The spent fuel is stored in the upper elevations of the Reactor Building, uately protected from the PMF. The intake structure itself which was originally c Class 1, is designed to withstand a water level of elevation 32.4 feet MSL.

s for an assumed 13.4 feet MSL still water level and for non-breaking waves they strike the structure. ROTECTION onents have been examined to identify and classify potential missiles. Generated Missiles ies of systems and components are reviewed to determine the potential for

 ; pressurized components and high speed rotating machinery. Only designs ure could lead to a missile ejection were considered.

there are no highly pressurized components or high speed rotating machines ng significant missile hazards in the permanently defueled condition. nally generated missiles are postulated. 3.1-8 Rev. 2.1

Generated by Events Near the Site is assessment (Reference 3.1-19) is to assure that the integrity of the safety systems, and components will not be impaired and that they will perform their the event of a site proximity missile. azardous activities in the vicinity of the Millstone site are addressed in Chapter cludes that the generation of missiles at these facilities does not pose a credible one site. Therefore, no specific protection is required other than that described ted missiles. ne Unit Number 1 does not present an undue risk to the health and safety of the f proximity missile hazards. Hazards one small commercial airport approximately 6 miles east-northeast of the site. on Airport handles regularly scheduled commercial passenger flights but is dling large jets. The licensee has determined that the probability of an aircraft ted structures of Millstone Unit Number 1 is sufficiently low that it does not cant hazard. DESIGN t Number 1 plant was designed for an earthquake (equivalent to the operating r OBE) with a horizontal peak ground acceleration (HPGA) of 0.07g and rthquake (equivalent to the safe shutdown earthquake or SSE) with a PGA of d design response spectrum recommended by John Blume and Associates and component of the 1952 Taft earthquake record normalized to the specified as seismic input for the analyses and design. The vertical component of ground ed to be two-thirds of the horizontal components. For the dynamic analyses of ctures, the buildings (or structures) were modeled as lumped mass-spring base to simulate the rock founded foundations. nses of the Reactor Building and Radwaste Building/Control were analyzed by ach. used for the analysis of safety related equipment: 3.1-9 Rev. 2.1

, summarizes the details of the original analysis and design.

Reactor Building enclosure (structure above elevation 108 feet 6) is capable of E with a peak ground acceleration of 0.17g without adversely affecting nearby s (Seismic II/I criterion), this portion of the structure is analyzed for the entered in-structure accelerations developed by Vectra Technologies for use in the USI A-46 (SQUG) program evaluations of equipment in the of the Reactor Building. These floor accelerations and spectra are considered e they incorporate the variabilities of the input motion at a rock site and the rs (mass and stiffness). The SSE floor accelerations at elevation 82 feet 9 vation evaluated in the Vectra report) are approximately 80% of the r accelerations obtained from the EDS Report (Reference 3.1-23). Therefore ons at the operating floor and at the roof level are conservatively taken as 80% ng accelerations from Reference 3.1-23. on of Measured and Predicted Responses ave been developed for abnormal operational events such as earthquakes. If etected, plant walkdowns are initiated to determine plant capability. F CLASS I AND CLASS II STRUCTURES riteria, Applicable Codes, Standards and Specifications tructures and facilities (Class I and II) conformed to the applicable general ions in effect at time of design. d Loading Combinations nts for the design of all structures and equipment include provisions for es resulting from dead loads, live loads and wind or seismic loads with impact s part of the live load. The treatment of equipment stresses are generally limited by non-operating loads such as the effect of building motion due to earthquake r support for a piece of equipment. However, the loads resulting from operating ratures on equipment are considered where they would increase the stresses. in the foundation were not considered in the design. 3.1-10 Rev. 2.1

drostatic, temperature loads or operating pressures and live loads expected to n the plant is operating. uake load. thquake load. have been followed for all Class I structures with respect to stress levels and for the postulated events are noted below: Reactor Building and Radwaste Building mal allowable code stresses are used (AISC for structural steel, ACI for forced concrete). The customary increase in design stresses, when earthquake s are considered, is not permitted. sses are limited to the minimum yield point as a general case. However, in a cases, stresses may exceed yield point. In this case, an analysis, using the it-Design approach, will be made to determine that the energy absorption acity exceeded the energy input. This method has been discussed in the AEC lication TID-7024, Nuclear Reactor and Earthquakes, Section 5.7. The lting distortion is limited to assure no loss of function and adequate factor of ty against collapse. mal allowable code stresses (AISC for structural steel, ACI for reinforced crete) with the customary increases in stresses when wind loads are considered. wable stresses used for various loading conditions are given for Class I 3.1-2. re established based upon equipment and operating loads and applied to the

, which is recommended to the boroughs by the State of Connecticut. Roof live m of 60 psf for Class I buildings and 40 psf for Class II buildings.

es will withstand the maximum potential loadings resulting from a wind es per hour with gusts up to 140 miles per hour. Although some damage to 3.1-11 Rev. 2.1

gs are not generally structurally connected, torsional effects are likely to be of oncrete structures, the design modulus of 3x106 psi is in accordance with the requirements for reinforced concrete (ACI 318-63), Section 1102, which is actice. However, it is recognized that the modulus of elasticity of concrete following the 28 day period, but it is difficult to evaluate the amount of wing factors affect the strength of concrete. Curing temperature Initial temperature Variations in mixes Amount of hydration s is not directly proportional to the strength of concrete: nevertheless, the the strength causes an increase in the modulus. However, the increase in the e is not believed to be significant in the light of all the uncertainties affecting crete. l change in the modulus may be, this effect is partially accounted for by cracks cture due to shrinkage and temperature. Such cracks tend to make the structure ch tends to compensate for the increased modulus. Also, the percent change in ll compared to other inputs in the analysis such as dimensions, areas, cross uping, etc. Hence, the effect of a small modulus change on the validity of the s considered to be negligible. l Criteria for Class II Structures and equipment are designed following the normal practice for the design of State of Connecticut, but as a minimum, this was not less than given in the Code for Zone 2. The usual practice of determining the stress due to lying a static load based on a specified seismic coefficient was followed. The II portion of the Reactor Building is addressed in Section 3.1.6. 3.1-12 Rev. 2.1

e Reactor Building are to enclose the spent fuel pool and associated equipment the weather. It supports maintenance of the fuel configuration in the fuel pool, from external hazards and supports maintenance of water in the fuel pool to a ensure the irradiated fuel is always immersed. ng at and below elevation 108 feet 6 inches is retained as seismic Class I in the led condition. Above elevation 108 feet 6 inches, the building enclosure is Class II with the requirement that the enclosure is capable of sustaining an SSE eismic II/I criterion). ing completely encloses the spent fuel pool. This building is a cast-in-place e structure. At the 108 foot 6 inch elevation, internal steel frame lateral bracing support the crane and the roof of the Reactor Building. The Reactor Building is ith adequate strength at an elevation of minus 32 feet 0 inches, with a orced concrete 142 feet 6 inches square. ge vault and the spent fuel storage pool are located in the Reactor Building. The refueling area is serviced by an overhead bridge crane. A refueling service ssary handling and grappling fixtures services the spent fuel storage pool. s a reinforced concrete structure, completely lined with seam-welded stainless welded to reinforcing members embedded in concrete. between elevations 65 feet 9 inches and 108 feet 6 inches. The fuel pool sits hick reinforced concrete slab which is supported by the reactor building rimary containment drywell wall. The pool stainless steel liner prevents unlikely event the concrete develops cracks. gned considering thermal stress, and the welds were dye penetrant inspected to tegrity. Construction materials used in the construction of the spent fuel ludes 4000 psi, 28 day strength concrete, 40 ksi deformed bar reinforcing steel,

, Type 304 stainless steel.

3.1-13 Rev. 2.1

ed to an excursion through the north 69° west component of the 1952 Taft applied factor of 7/17. The resulting maximum shears, moments and e used for design. elopes of building design shears, moments and displacements are presented res 3.1-3 through 3.1-5, respectively. These curves have been used in the he Reactor Building. Loads and shears from reactor pressure vessel and nd equipment are transferred to the drywell structure and pedestal and then to

. Careful grouting between the drywell and mat ensures direct transfer of to the mat. Shears are transferred to the mat by friction and bearing.

ing was designed to resist the seismic shears and moments presented herein ncrease in stress for short-term loadings. In addition, the structure was that it can resist 2.4 times the postulated seismic shears and moments without e structure. In addition to the horizontal accelerations, a vertical building (and ation was used for design. ing enclosure structure (above elevation 108 feet 6 inches) is analyzed for a ntered SSE, as described in Section 3.1.6, and is shown to resist the resulting the accelerations with no loss of structural integrity. Room and Radwaste Treatment Building nt facility is north of and adjacent to the reactor building. The building includes kage space below grade with the plant control room above grade. The area einforced concrete construction with shielded compartments provided for the adwaste equipment. The control room above grade is of reinforced concrete ot thick reinforced concrete roof. The control room and radwaste facility are Class II. The analytical model used in the seismic analysis of the control room ing is shown in Figure 3.1-6 and is similar to those for the Reactor Building. lding is seismically analyzed consistent with Regulatory Guide 1.143. Features ructure is founded on rock. The maximum bearing pressure on the rock is 10

t. The exterior walls are of cast-in-place concrete and designed for an earth foot at any depth equal to the depth in feet times 90 pounds. The exterior walls 3.1-14 Rev. 2.1

floor at elevation 14 feet 6 inches, including the concrete shielding plugs ways over equipment in the substructure, is designed for a uniform live load of of ductile metal and all sump pits are lined so that these containers can be ntial distortion without rupture. massive reinforced concrete, not subject to fracturing. Even in the event

 , seepage would be into the building rather than out, since the water table is vel.

Structure e is a reinforced concrete frame supported on a reinforced concrete is founded on rock. The building has a flat roof consisting of 10 gauge steel covered with insulation and a tar and felt roofing membrane. Hatches are f for removal of major pieces of equipment. The front wall of the intake d to resist the standing wave. Seismic stress levels were calculated using g at grade and 0.12 g at the roof level for design earthquake and 2.4 times maximum earthquake. The structure is capable of withstanding 300 mph wind internal pressure of 2.5 psi. However, the large number of hatches in the roof essure. Although originally design as seismic Class I, the intake is considered the permanently defueled condition. e is located west of the main plant and has five 11 foot 2 inch wide bays. Each th manually raked trash racks and stop log guides. ce and cooling water strainers is made in a separate covered pit adjacent to the Building ing is a Class II structure. The Turbine Building foundation consists of a mat supported on rolled structural steel H section bearing piles. All piles were refusal in the dense strata immediately above rock. Reinforced concrete shield up to the operating deck at elevation 54 feet 6 inches. 3.1-15 Rev. 2.1

UALIFICATION OF SEISMIC CATEGORY I INSTRUMENTATION AND AL EQUIPMENT rocess instrumentation provides safety related functions in conjunction with dling of irradiated fuel or radioactive waste, or is credited with any function in ns performed to ensure that no undue risk to the health and safety of the public trumentation or electrical systems are required for mitigation of the design accident. Seismic qualification of plant instrumentation and electrical quired. MENTAL DESIGN OF ELECTRICAL EQUIPMENT ted to qualification of the electrical portion of the engineered safety features to ded functions in the combined normal, accident and post accident re are no non-structural engineered safety features related to the safe storage e irradiated fuel or radioactive waste, or credited in the safety evaluations e that no undue risk to the health and safety of the public exists. No non-ed safety features are credited in accident analysis to prevent or mitigate the e current design basis fuel handling accident. CES uclear Power Station Unit Number 1 Application for Full Term Operating ptember 1, 1972. 824, Integrated Plant Safety Assessment, Systematic Evaluation Program, uclear Power Station, Unit Number 1, February 1983. hilk (Nuclear Regulatory Commission) memo to J. M. Taylor (Nuclear Commission), SECY-92-233 Resolution of Deviations Identified during the Evaluation Program dated September 18, 1992. Plant Safety Assessment, Systematic Evaluation Program, Millstone Nuclear on, Unit Number 1, NUREG-0834, Supplement Number 1, November 1985. Grimes (NRC) to J.F. Opeka, subject: IPSAR Sections 4.4 Wind and Tornado nd 4.7 Tornado Missiles. 3.1-16 Rev. 2.1

ion Requirements, II-4.F Settlement of Foundations and Buried Equipment, and Tornado Loadings, III-3.A Effects of High Water Level on Structures, III-6 sign Considerations. ober 7, 1983, from W. G. Counsil to D. M. Crutchfield (NRC),

Subject:

uclear Power Station Unit Number 1, SEP Topic III-2 Wind and Tornado ember 3, 1982, from W. G. Counsil to D. M. Crutchfield (NRC),

Subject:

uclear Power Station Unit Number 1, SEP Topic III-2 Wind and Tornado ruary 4, 1986, from J. F. Opeka to C.I. Grimes (NRC),

Subject:

Millstone wer Station Unit Number 1 ISAP Topic 1.19, Integrated Structural Analysis. G. Counsil to D. M. Crutchfield (NRC), dated March 16, 1984, Millstone wer Station, Unit Number 1, SEP Topic II-3.B, Flooding Potential and Requirements, SEP Topic II-4.F, Settlement of Foundations and Buried , SEP Topic III-2, Wind and Tornado Loadings, SEP Topic III-3.A, Effects of Level on Structures SEP Topic III-6, Seismic Design Considerations. , Appendix A, General Design Criterion 4. ne 29, 1982, W.G. Counsil to D. M. Crutchfield,: Millstone Nuclear Power t Number 1, SEP Topic III-4.A, Tornado Missiles. arch 9, 1982, W.G. Counsil to D. M. Crutchfield,: Millstone Nuclear Power t Number 1, SEP Topic III-4.A, Tornado Missiles. ovember 19, 1981, W.G. Counsil to D. M. Crutchfield: Millstone Nuclear on Unit Number 1, SEP Topic III-4.A, Tornado Missiles. ugust 31, 1981, W.G. Counsil to D. M. Crutchfield,: Millstone Nuclear Power t Number 1, SEP Topic III-4.A, Tornado Missiles. ctober 16, 1985, J. F. Opeka to C. I. Grimes, Millstone Nuclear Power Station er 1, Integrated Safety Assessment Program. 3.1-17 Rev. 2.1

ptember 17, 1981, W. G. Counsil to D. M. Crutchfield, Millstone Nuclear on Unit Number 1, SEP Topic III-4.D, Site Proximity Missiles. EG/CR-2024 Report, Seismic Review of the Millstone-1 Nuclear Power 1981. Topics III-6, Seismic Design Considerations and III-II, Component Integrity - uclear Power Station Unit Number 1, SAR dated 6/30/82. t Number 02-0240-1094, Generation of In-Structure Seismic Response llstone Unit Number 1, dated June 1982. Site Specific Ground Response Spectra for SEP Plants Located in the Eastern es, June 17, 1981. EG/CR-1582 Report, Seismic Hazard Analysis, Vols. 2-4, October 1981. Opeka to C.I. Grimes, Millstone Nuclear Station, Unit Number 1 ISAP Topic rated Structural Analysis, dated January 6,1986. D. G. Eisenhut, NRC, to W. G. Counsil, dated January 1, 1980. D. M. Crutchfield, NRC, to W. G. Counsil, dated July 28, 1980. W. G. Counsil to D. M. Crutchfield, NRC, dated October 16, 1985. nit 2 Final Safety Analysis Report Section 5.8.6. nit 3 Final Safety Analysis Report Section 3.7.4.2. nologies Report Number 0024-00099-RB-1, Rev. 1, Reactor Building A-46 ated June 10, 1996. 3.1-18 Rev. 2.1

TABLE 3.1-1 COMPARISON WITH NRC GENERAL DESIGN CRITERIA SEP AND TMI SAFETY ISSUES WHICH LISTED THE SPECIFIED GDC GENERAL DESIGN CRITERIA AS PART OF CONCERN AFFECTED REGULATORY GUIDE I OVERALL REQUIREMENTS 1 QUALITY STANDARDS AND RECORDS SEP II-3.A, II-3.B, II-3.C, III-3.A AND III-7.B 1.27, 1.59 2 DESIGN BASES FOR PROTECTION AGAINST SEP II-2.A, II-3.A, II-3.B, II-3.C, II-4.E, II-4.F, II-4.3, III-19 III-2, III-3.A, 1,27, 1.,32, 1.59, 1.60, 1.61, 1.68, 1.75, 1.76, 1.92, 1.102, 1.117, NATURAL PHENOMENA III-3.B, III-3.C, III-6, III-7.B, III-8.C, III-11, VIII-3.A, VIII-3.B, TMI II.B.1 1.120, 122, 1.127, 1.129, 1.132 3 FIRE PROTECTION (SEE DSAR Section 3.2.9) 4 ENVIRONMENTAL AND MISSILE DESIGN BASES SEP II-1.C, II-3.A, II-3.B, II-3.C, III-1, III-4.B, III-5.A, III-5.B, III-7,B, III-11, 1.3, 1.4, 1.7, 1.20, 1.27, 1.29, 1.32, 1.35, 1.45, 1.46, 1,59, 1.68, 1.75, V-5, VIII-3.A, VIII-3.B, TMI II.B.2, II.B.3, 2.1.6.A, 2.1.8.A, III-4,A 1.115, 1.12, 5 SHARING OF STRUCTURES, SYSTEMS AND SEP III-1, VIII-3.A AND VIII-3.B 1.32, 1.75, 1.129 COMPONENTS II PROTECTION BY MULTIPLE FISSION PRODUCT BARRIERS 10 REACTOR DESIGN NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION 11 REACTOR INHERENT PROTECTION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION 12 SUPPRESSION OF REACTOR POWER NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED OSCILLATIONS CONDITION 13 INSTRUMENTATION AND CONTROL NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION 14 REACTOR COOLANT PRESSURE BOUNDRY NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION 15 REACTOR COOLANT SYSTEM DESIGN NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION 16 CONTAINMENT DESIGN NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION 17 ELECTRIC POWER SYSTEMS SEP III-1, VII-7, VIII-2, VIII-3.A VIII-3.B, TMI II.E.3.1, II.G.1 1.6, 1.9, 1.32, 1.75, 1.129 18 INSPECTION AND TESTING OF ELECTRIC POWER NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED SYSTEMS CONDITION 19 CONTROL ROOM NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION Page 1 of 4 Rev. 2

TABLE 3.1-1 COMPARISON WITH NRC GENERAL DESIGN CRITERIA SEP AND TMI SAFETY ISSUES WHICH LISTED THE SPECIFIED GDC GENERAL DESIGN CRITERIA AS PART OF CONCERN AFFECTED REGULATORY GUIDE III PROTECTION AND REACTIVITY CONTROL SYSTEMS 20 PROTECTION SYSTEM FUNCTIONS NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION 21 PROTECTION SYSTEM RELIABILITY AND NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED TESTABILITY CONDITION 22 PROTECTION SYSTEM INDEPENDENCE NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION 23 PROTECTION SYSTEM FAILURE MODES NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION 24 SEPARATION OF PROTECTION AND CONTROL NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED SYSTEMS CONDITION 25 PROTECTION SYSTEM REQUIREMENTS FOR NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED REACTIVITY CONTROL MALFUNCTIONS CONDITION 26 REACTIVITY CONTROL SYSTEM REDUNDANCY NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED AND CAPABILITY CONDITION 27 COMBINED REACTIVITY CONTROL SYSTEMS NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CAPABILITY CONDITION 28 REACTIVITY LIMITS NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION 29 PROTECTION AGAINST ANTICIPATED NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED OPERATIONAL OCCURRENCES CONDITION IV FLUID SYSTEMS 30 QUALITY OF REACTOR COOLANT PRESSURE NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED BOUNDARY CONDITION 31 FRACTURE PREVENTION OF REACTOR COOLANT NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED PRESSURE BOUNDARY CONDITION 32 INSPECTION OF REACTOR COOLANT NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED PRESSURE BOUNDARY CONDITION 33 REACTOR COOLANT MAKEUP NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION Page 2 of 4 Rev. 2

TABLE 3.1-1 COMPARISON WITH NRC GENERAL DESIGN CRITERIA SEP AND TMI SAFETY ISSUES WHICH LISTED THE SPECIFIED GDC GENERAL DESIGN CRITERIA AS PART OF CONCERN AFFECTED REGULATORY GUIDE 34 RESIDUAL HEAT REMOVAL NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION 35 EMERGENCY CORE COOLING NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION 36 INSPECTION OF EMERGENCY CORE COOLING NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED SYSTEM CONDITION 37 TESTING OF EMERGENCY CORE COOLING NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED SYSTEM CONDITION 38 CONTAINMENT HEAT REMOVAL NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION 39 INSPECTION OF CONTAINMENT HEAT NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED REMOVAL SYSTEM CONDITION 40 TESTING OF CONTAINMENT HEAT REMOVAL NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED SYSTEM CONDITION 41 CONTAINMENT ATMOSPHERE CLEANUP NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION 42 INSPECTION OF CONTAINMENT ATMOSPHERE NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CLEANUP SYSTEMS CONDITION 43 TESTING OF CONTAINMENT ATMOSPHERE NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CLEANUP SYSTEMS CONDITION 44 COOLING WATER NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION 45 INSPECTION OF COOLING WATER SYSTEM NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION 46 TESTING OF COOLING WATER SYSTEM NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION V REACTOR CONTAINMENT 50 CONTAINMENT DESIGN BASIS NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION 51 FRACTURE PREVENTION OF CONTAINMENT NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED PRESSURE BOUNDARY CONDITION Page 3 of 4 Rev. 2

TABLE 3.1-1 COMPARISON WITH NRC GENERAL DESIGN CRITERIA SEP AND TMI SAFETY ISSUES WHICH LISTED THE SPECIFIED GDC GENERAL DESIGN CRITERIA AS PART OF CONCERN AFFECTED REGULATORY GUIDE 52 CAPABILITY FOR CONTAINMENT LEAKAGE RATE NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED TESTING CONDITION 53 PROVISIONS FOR CONTAINMENT INSPECTION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED AND TESTING CONDITION 54 SYSTEMS PENETRATING CONTAINMENT NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION 55 REACTOR COOLANT PRESSURE BOUNDARY NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED PENETRATING CONTAINMENT CONDITION 56 PRIMARY CONTAINMENT ISOLATION NOT APPLICABLE TO THE PERMANENTLY DEFUELEDCONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION 57 CLOSED SYSTEMS ISOLATION VALVES NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION VI FUEL AND RADIOACTIVITY CONTROL 60 CONTROL OF RELEASES OF RADIOACTIVE SEP II.2.C, XI-1, XI-2, TMI II.B.2, II.B.3, 2.1.6.A, 2.1.8.A 1.3, 1.4 MATERIALS TO THE ENVIRONMENT 61 FUEL STORAGE AND HANDLING AND SEP XI-1, XI-2 RADIOACTIVITY CONTROL 62 PREVENTION OF CRITICALITY IN FUEL STORAGE AND HANDLING 63 MONITORING FUEL AND WASTE STORAGE SEP XI-1, XI-2 64 MONITORING RADIOACTIVITY RELEASES SEP II-2.C, XI-1, XI-2; TMI II.B.2, II.B.3, 2.1.6.A, 2.1.8.A Page 4 of 4 Rev. 2

TABLE 3.1-2 ALLOWABLE STRESSES FOR CLASS I STRUCTURES CONCRETE STRUCTURAL STRUCTURAL REINFORCING MAX. CONCRETE CONCRETE STEEL STRUCTURAL STEEL STEEL MAX. ALLOWABLE MAX. MAX. TENSION ON STEEL SHEAR COMPRESSION STRUCTURAL ALLOWABLE COMPRESSION ALLOWABLE ALLOWABLE THE NET ON GROSS ON GROSS STEEL LOADING CONDITIONS STRESS STRESS SHEAR STRESS BEARING SECTION SECTION SECTION BENDING DEAD LOADS PLUS LIVE LOADS,* PLUS 0.5 Fy 0.45 f c 1.1 f c 0.25 f c 0.60 Fy 0.40 Fy VARIES WITH 0.66 Fy OPERATING LOAD PLUS SEISMIC LOADS (0.07g) SLENDERNESS TO RATIO 0.60 Fy DEAD LOADS PLUS LOADS,

• PLUS OPERATING 0.667 Fy 0.60 f c 1.467 f c 0.333 f c 0.80 Fy 0.53 Fy VARIES WITH 0.88 Fy LOADS PLUS WIND LOADS SLENDERNESS TO RATIO 0.80 Fy DEAD LOADS PLUS LIVE LOADS,* PLUS GROSS STRUCTURAL INTEGRITY CAN BE OPERATING LOADS, PLUS SEISMIC LOADS 0.17g MAINTAINED (SEE NOTE 1 BELOW) of live loads were considered concurrent with the seismic loads

= Minimum Yield Point of the Material.

= Compressive Strength of Concrete.

TE 1: The structure was analyzed to assure that gross structural integrity can be maintained during ground motion having 17/7 the intensity of the operating basis earthquake described in SECTION 3.1.6, even though stresses in e of the materials may exceed the yield point. Page 1 of 1 Rev. 2

el Storage Bases or the storage of spent fuel are as follows: ge pool for the underwater storage of 2959 fuel assemblies. keff of less than 0.95 at all times, including postulated criticality accidents. re worst case results, considering maximum variation in the position of the fuel within the storage rack, neutron absorber variation (where credited), seismic flections and calculation uncertainty. Boraflex is not credited. te shielding walls are designed as part of the Class 1 portion of the Reactor ructure. The thickness of the walls and the standards of design are such as to ructural damage or loss of function of the walls. esign of the fuel storage and equipment storage facilities meets all ts for Class I structures. orage racks for the fuel are designed to assure subcriticality in the fuel pool. racks are an interconnected honeycomb array of square stainless steel boxes ividual cells for fuel storage. 1045 storage cells contain Boraflex sheets (not n four sides, and 2184 storage cells contain B4C plates for neutron absorption. 5 storage cells with Boraflex, only 775 cells are allowed to contain fuel. Accident Requirements. Millstone 1 has chosen to comply with

.68(b).

es Description ains water which is not borated. The fuel storage pool is a reinforced concrete ly lined with seam-welded, stainless steel plate (11 gauge) which is welded to rs (channels, I-beams, etc.) embedded in concrete. The liner is reinforced by s and suitable insert strips in areas subject to heavy loading such as the cask concrete shielding walls are two or more feet thick and are designed as part of of the Reactor Building structure. 3.2-1 Rev. 8

are suitably grouped to indicate the area of leakage. To avoid unintentional l, there are no penetrations that would permit the pool to be drained below e feet above the top of the active fuel, and all lines extending below this level suitable valving to prevent backflow. The passage between the fuel storage ing cavity above the reactor vessel is provided with two gates. The refueling d in a drained down state. The gate adjacent to the refueling cavity is welded to rming a permanent pressure boundary for the fuel storage pool. The double t to the fuel storage pool is removable but normally maintained in the closed ly open drain line between the gates permits detection of leaks from the gate storage pool. The drain line may be isolated and the volume between the gates removal of the gate for repairs in the event of such leakage. NRC I.E. Bulletin 84-03, augmented leak detection capability has been nt fuel pool to indicate high/low level in the pool. ol is cooled and filtered as required by the spent fuel pool cooling and in-pool scribed in Subsection 3.2.1.3. designed to hold 20 fuel channels. mately seven feet by seven and one half feet is reserved for loading a spent fuel g irradiated reactor vessel internals and other materials classified per 10 CFR class C (GTCC) waste are stored in the fuel storage pool adjacent to the fuel Evaluation l bundles in the spent fuel storage pool, the presence of neutron absorbing dited) in the fuel storage racks, not placing fuel in prohibited locations chnical Specifications, and the design of the fuel bundles maintains keff less

5. This is assured by limiting the fuel assemblies in the pool to those that have 1.24 in the normal reactor configuration at cold conditions, and an average of 3.8 weight percent or less. The criticality analysis confirms acceptable f the spent fuel pool temperature.

3.2-2 Rev. 8

cks are analyzed to withstand the impact of a dropped fuel assembly and a combined dry weight of 1675 pounds from the maximum lift height of the telescoping mast. The analyses performed (References 3.2-9 to 3.2-12) e spent fuel racks remain functional and that the spent fuel remains in a ged and coolable condition. mitter, monitoring pool water level, is provided to detect loss of water from the mitter, monitoring the skimmer surge tank, is provided to permit water loss ing a low level alarm and provide level indication in the Millstone Unit 2 of spent fuel can be found in References 3.2-1, 3.2-2, 3.2-3., 3.2-4, 3.2-5, and el Pool Cooling System ol Cooling System has been analyzed to remove the maximum heat load from Bases cture, pool liner, fuel racks, and external cooling system have been designed for proximately 150°F. However, all of these structures and components have to be structurally adequate for abnormal temperature excursions to 212°F. ss of external cooling and a closed airspace above the pool, it would take days for the pool temperature to rise to 212°F from an initial SFP bulk water

°F, or approximately 7.5 days to rise to 212°F if starting from the TRM upper f 140°F. The spent fuel pool cooling system and secondary DHR cooling qualified for satisfactory operation with pool temperatures as high as 170°F.

the maximum anticipated pool water temperature, following loss of cooling, al ventilation within the reactor building is established within approximately 5 m an initial SFP bulk water temperature of 100°F, or 2.5 days if starting from mperature limit of 140°F. are available to add water to the pool and adequate time is available to repair, or line up the system used for pool water cooling. Most significantly, if this to cool the pool water, no fuel damage would result and the potential off site t approach the guidelines established in 10 CFR 50.34(a) or 10 CFR 100.11 3.2-3 Rev. 8

high clarity water to the fuel pool using the in-pool cleanup system. radioactivity released to the pool water using the in-pool cleanup system. uel Pool Heat Load mber 1 has permanently ceased power operation and all irradiated fuel has been ved from the reactor vessel. There are 2885 irradiated fuel assemblies in the luding one segmented bundle, consisting of 19 fuel rods. A decay heat load rformed utilizing the computer program ORIGEN2, an industry standard for erence 3.2-13). The results show that total heat load in the pool was 1.781 . The spent fuel pool secondary cooling system (DHR) has been sized to uel decay heat load of approximately 1.5 Mbtu/hr, projected to exist on 6/1/00. Fuel Pool Cooling pool heat load established, a second calculation (Reference 3.2-14) was mine the transient and steady state spent fuel pool and reactor building ut active cooling to the spent fuel pool. Several cases were analyzed with n configurations such as forced ventilation, natural ventilation and no the building. Steady state and transient calculations were performed to pool and building temperatures and evaporation rates, as well as time frames tor actions. All analyses were performed using the GOTHIC computer valuated was during summer conditions (92°F, 50% Relative Humidity) of active spent fuel pool cooling and without the reactor building HVAC system s case the time to reach 212°F in the spent fuel pool is approximately 7.5 days TRM upper temperature limit of 140°F. This calculation also establishes a tive loss of 3.8 gpm under the above conditions. If natural ventilation is ning the reactor building truck bay doors, equipment hatch garage doors and rs on the reactor building roof, the maximum calculated pool temperature is imum evaporation rate is 3.0 gpm. Description l cooling system cools water in the fuel pool on an as needed basis to maintain An in-pool demineralizer and filter maintain purity and water quality. Water is 3.2-4 Rev. 8

of the fuel storage pool, and a local temperature indicator. The transmitter d in the Millstone Unit 2 Control Room via the Programmable Logic Controller des both indication of bulk temperature and notification of a high and low conditions within the fuel storage pool. ol demineralizer operates on an as needed basis to maintain pool water pool filter operates on an as needed basis to maintain pool water clarity. The ks are shielded with concrete. ng system is controlled and operated locally and from the Millstone Unit 2 e system is provided with indicators and alarms for system flow, water level, kimmer surge tank level, and component operating status. Evaluation r acts passively to transfer decay heat from the fuel and will protect the fuel out human intervention as long as the fuel is completely immersed in water. If stopped, the pool water temperature would gradually increase, resulting in no most severe case of a closed airspace, with the current decay heat load in the mber 1 Fuel Pool and no external cooling, the pool temperature would only stop rising) when the pool water boils, which is the natural limit of water ace at atmospheric pressure. The fuel pool structure, pool liner, fuel racks, and stem have been demonstrated to be adequate for abnormal temperature F. With a complete loss of external cooling and a closed airspace above the approximately 10 days for the pool temperature to rise to 212°F from an initial mperature of 100°F, or approximately 7.5 days to rise to 212°F if starting from mperature limit of 140°F. This is significantly longer than required to reinstate the water. If natural ventilation is established, by opening the reactor building uipment hatch garage doors and the tornado dampers on the reactor building calculated pool temperature is 163°F. dling System Bases or the fuel handling system are as follows: of contamination or exposure of personnel to radiation will exceed the 10 CFR 3.2-5 Rev. 8

ing the fuel storage pool. The operating floor is serviced by the Reactor ich is equipped with a 110 ton main hoist and a seven-ton auxiliary hoist. ach any major equipment storage area on the operating floor. Evaluation ge and other fuel handling equipment are required for movement of fuel and in the fuel pool into storage/shipping containers. The reactor building crane is torage and shipping casks in the reactor building. These functions are required defueled condition, but are not safety related. ING AND CONTROL FUNCTIONS t 2 Control Room serves as the control room for Millstone Unit 1, and is ed. It is described in Section 7.6 of the Millstone Unit 2 Final Safety Analysis Unit 2 Operations personnel are responsible for the monitoring and control of el pool island (SFPI) and auxiliary systems via a computer console located in 2 Control Room. The computer console in the Millstone Unit 2 Control Room rogrammable Logic Controller (PLC) for data acquisition and trending. The he Millstone Unit 1 Central Monitoring Station (CMS). The CMS is located ance Shop. t 1 CMS is not manned. It contains two computer consoles that may only be ecause they are normally in a locked supervisory mode. oring stations in the original Unit 1 Control Room. The original Unit 1 Control rforms any Unit 1 function. EAT REMOVAL (DHR) SYSTEM ases s designed to provide cooling to the spent fuel pool cooling system. The system re: 170°F (maximum) 625 gpm per pump 3.2-6 Rev. 8

rovides a supply of cooling water to the shell side of the spent fuel pool heat is circulated in a closed loop by the DHR pumps. Heat is removed from the DHR air-water heat exchangers located outside on the roof above the H&V guration may vary depending on heat load. The remainder of the system g water expansion tank, an air separator, piping and valves, and controls and demineralizer maintains system activity below established limits. The flow tem is shown in Figure 3.2-4. aluation upplies cooling water to the fuel pool heat exchangers. Fuel pool cooling is a uired for the permanently defueled condition, but is not safety related. ction of the DHR system is not safety related. nd Inspection nents and instrumentation are tested periodically as necessary to ensure ss. ntation mentation and controls are located locally and in the Millstone Unit 2 Control WATER SYSTEM lized Water Description l makeup system will supply and store demineralized water to makeup for akage in the pool. The primary source will be from the Unit 2 Primary Makeup pplied from the onsite water treatment facility. A 5,000 gallon storage tank and nstalled in the reactor building to provide makeup water to the spent fuel pool n the normal makeup from Unit 2 is unavailable. A connection to the pool provided near the reactor building truck bay door to allow makeup to be er truck or fire water if necessary. 3.2-7 Rev. 8

l makeup water system provides demineralized makeup water to the spent fuel pool cooling system. This function supports fuel pool cooling, but is not safety and Inspection akeup system is on demand at intermittent intervals to replenish water in the keup water storage tank and the skimmer surge tanks. The equipment is periodically. Sampling of the makeup water storage tank is a standard ure. entation switch for the makeup water transfer pump is located locally at the pump. age tank level indication is also provided. NALLY DELETED SAMPLING SYSTEM ases pling process gases is to provide representative samples for testing to obtain e performance of the plant equipment and systems are determined. escription 1 BOP ventilation exhaust flow is continuously sampled for radioactive mple is taken from the exhaust duct which runs along the north exterior wall of ng. A single point sample nozzle is positioned to obtain a representative sample xhaust air. The sample passes through a particulate filter and is then expelled ust duct. on exhaust flow is continuously monitored for gaseous radiation and ample is taken from the exhaust duct near the reactor building exhaust plenum. ple nozzle is positioned to obtain a representative sample of the turbulent and t air. The sample passes through a particulate filter and a gas monitor and is into the exhaust duct. 3.2-8 Rev. 8

nd Inspection re performed after installation. Routine use substitutes for subsequent periodic ception of calibration and maintenance. AL SYSTEMS ion al systems include the equipment and facilities which provide power to desired strumentation and controls. The system is designed to provide reliable power y defueled condition. The power system is designed with a sufficient source, ntrol, and necessary switching. ource is through the emergency station service transformer (ESST), which steps ce from the Waterford Substation 36F2 circuit to 4160V. system is designed to provide a reliable source of power to the on site AC system. ally Deleted. lectric System ction vailable to operators following a loss of offsite power to assure the continued without reliance on emergency sources of power. ded through the emergency station service transformer. The emergency station r has adequate capacity to supply all normal auxiliaries required to support the led condition. Power for the SFPI and other decommissioning related activities ia Bus 14H. issioning related 125V DC power is obtained from rectified AC power at the separate 125V DC source consisting of a 125V DC battery, a battery charger, and distribution panel. 3.2-9 Rev. 8

ally when over-current conditions exist. The control power to the 4160 volt bus om the decommissioning 125 volt DC system. ent of the 4160 volt power system is described below. Station Service Transformer tion service transformer is an outdoor, 27,750-4160 volt three phase, 60 Hz., 12.5 MVA OA/FA 55°C, and 14 MVA, FA 65°C, transformer. t System olt bus 14H is stepped down through transformers energizing the 480 volt FAC-B2. ply breakers are opened and closed locally. All breakers will trip automatically ditions exist. t Systems l system utilizes its own dedicated 120V AC power derived from the SFPI AC nt AC system is provided by the SFPI 120V AC distribution system and of use UPS equipment. The SFPI PLC system has an integral 24V DC power wer System Design Criteria g Capacity - The switchgear, load centers, motor control centers, and panels are sized for interrupting capacity based on maximum short circuit at their location. Low voltage metal enclosed breakers at load centers and e breakers at motor control centers are adequately sized for these maximum ort circuit currents. ystem Protection - Electrical system protection is provided by protective elays which monitor the electrical characteristics of the equipment and/or em to assure operation consistent with design parameters, as follows: 3.2-10 Rev. 8

DC utilizes rectified AC power. The rectifiers are located at the SFPI 480V AC addition, the decommissioning 125V DC system consists of a 125V DC sconnect switch and distribution panel. nally Deleted Evaluation defueled condition portions of the electrical systems are required for power equired non-safety related equipment in other systems. Since none of the d by these systems is safety related (Class 1E), all of the electrical systems are Although single failure criteria still apples to the unit, it need not be applied to ment that are non-safety related. Since none of the electrical systems or y related or required for Regulatory Guide 1.97 (post accident monitoring) EEQ program need not be applied. General Design Criteria Number 17 stems) includes certain requirements for availability of offsite power to support Since the reactor cannot be made critical under allowed plant conditions in the led condition, no power source is required to be operable or available. ITIONING, HEATING, COOLING AND VENTILATION SYSTEMS uilding and SFPI Heating and Ventilation System Bases ing and SFPI heating and ventilation systems are operated to maintain a freezing within the areas of that building. aintain a slightly negative pressure when compared to the outside atmosphere. o ensure that there will be no inadvertent unmonitored release to the site area ilding. quiescent evaporation of liquid waste may be released into the ventilation s allows only the distillate vapor into the ventilation system, assuring positive ecies and concentration of radionuclides released with Reactor Building 3.2-11 Rev. 8

ncludes supply and exhaust fans installed in modular units. Description ing and SFPI HVAC systems provide for the protection of personnel and rborne radioactive contaminants and excessive thermal conditions. Air flow is progressively greater radioactive contamination prior to exhaust. ing is provided with supply and exhaust ventilation to ensure proper air flow ve heat generated from equipment. ncludes variable speed supply and exhaust fans to maintain space temperature imits while also maintaining a negative pressure within the SFPI envelope ide and to Reactor Building areas outside the SFPI envelope. the Reactor Building HVAC system is given in Figure 3.2-12. The SFPI HVAC Figure 3.2-6. VAC nt of the system provides fresh air to all levels in the Reactor Building outside Outside air passes through fixed louvers, a damper, filters, and electric heating ailable to deliver air flow. Electric unit heaters are provided inside the drywell

n. Exhaust air flow combines in a common duct and continues on to the main s, in addition to those mentioned above, include screens, filters, ductwork with utlets, return and exhaust intakes, heating coils, and instrumentation and ctuation, indication, and alarm instrumentation are incorporated in a central rol panel.

m nt of the system provides fresh air to the operating floor of the Reactor of the 82 feet 9 inches elevation and the spent fuel pool pump area. Outside air d louvers in the side of the reactor building wall, filters, and electric heating able speed 100% capacity fan is available to deliver air flow. 3.2-12 Rev. 8

s, in addition to those mentioned above, include ductwork with dampers, rn and exhaust intakes, and instrumentation and controls. Control actuation, rm instrumentation are incorporated in a local control panel. Indication and provided in the Millstone Unit 2 Control Room. cooling capability is also provided by opening the Reactor Building truck bay atch garage doors and the tornado dampers located on the Reactor Building uld be used following an extended loss of all spent fuel pool cooling capability. Evaluation ing and SFPI heating and ventilation systems maintain environmental ing spaces (to support personnel comfort or operation of equipment located on ct ventilation air from areas of low radioactive contamination to areas of er contamination (to minimize the spread of contamination), and vent inated exhaust air. Natural ventilation cooling capability is also provided for ling following an extended loss of all active pool cooling capability. The nd SFPI heating and ventilation systems are not safety related, but are required defueled condition because they house SSCs that are associated with the safe ng of irradiated fuel or radioactive waste. Building Ventilation System Bases lding ventilation system operates to supply filtered air to this building's areas.

d. The presence of dust particles potentially increases the spread of radioactive lters the exhaust air prior to its discharge, to limit the release of any radioactive e environment.

is routed to areas of progressively greater radioactive contamination potential st. Back-draft dampers are provided to prevent reverse flow between areas of ation potential. 3.2-13 Rev. 8

creening. The air is drawn through a filter designed to remove dust. A header o various areas of the building. y is located in the clean areas of the building while the inlets to the exhaust here the rate of contamination is the highest. passed through the filtering system before discharge through the main exhaust Evaluation lding ventilation directs ventilation air from areas of low radioactive reas of progressively greater contamination (to minimize the spread of d vents potentially contaminated exhaust air. The Radwaste Building is only required, in the permanently defueled condition, to support personnel ally Deleted uilding Heating and Ventilation Bases ing ventilation system is operated to maintain a slight negative pressure in the any radioactive out-leakage, as well as, to provide fresh air to support Description d to the Turbine Building through louvers in the walls and roof. tem is arranged with one supplementary transfer fan and connecting ductwork he north end of elevation 14 feet 6 inches. ing exhaust system collects air from various areas into an exhaust air header nto a plenum which also receives air from the Reactor Building and Liquid

. One exhaust fan is furnished to handle the combined exhaust from these three discharges into a duct which runs along the north wall of the Reactor Building e exhaust air to the environment. Potentially contaminated areas in the Turbine 3.2-14                                         Rev. 8

Evaluation ing ventilation system directs ventilation air from areas of low radioactive reas of progressively greater contamination (to minimize the spread of d vents potentially contaminated exhaust air. The Turbine Building ventilation ired, in the permanently defueled condition, to support personnel access to the TECTION SYSTEMS lear Plant Fire Protection Program has been developed to ensure that any cause an unacceptable risk to public health and safety, and will not se the risk of radioactive release to the environment. rogram has been established at Millstone Unit Number 1. This program protection policy for the protection of structures, systems, and components fueled condition of the unit and the procedures, equipment, and personnel ent the program. ases intain a high level of confidence for the Fire Protection Program, it has been ministered using the defense-in-depth concept. The defense-in-depth concept level of fire protection fails, another level is available to provide the required tection terms, this defense-in-depth concept consists of the following levels; fires from starting, tion of fires that do start, and and/or extinguishing them quickly so as to limit their damage. ls can be perfect or complete, but strengthening any one level can compensate or weaknesses, known or unknown, in the others. 3.2-15 Rev. 8

services individually valved lines feeding fixed pipe water suppression

, waterspray, and standpipes) throughout the plant and hydrants located around plant.

t Number 2 and 3 fire pumphouses contain three, 2,000 gpm at 100 psi, fire ly the yard loops; two with electric-motor drives and one with diesel engine e Unit Number 3 pumphouse contains one electric driven pump (M7-8), fed it Number 3 power, and the diesel-driven fire pump (M7-7). The Millstone contains one electric driven pump (P-82) fed from Unit 2 power. All three dual connections to the underground supply system. Maximum system flow ements can be met with any one of the three pumps out of service. s such that a 50 gpm electric jockey pump (M7-11) maintains system pressure arting when line pressure drops to 105 psig and will run until pressure reaches ed by a line pressure switch. A hydro-pneumatic tank is provided in the system cling of the jockey pump. At pressures below 105 psig, the MP2 P-82 electric 98 psig to maintain system pressure and flow. The Millstone Unit Number 3 p then will start at 85 psig and it is fed 480 VAC from MCC-CD-6 (MCC mpartment number 1A). This pump is auto-started by a pressure switch set at 85 hile the M7-7 diesel-driven fire pump is auto-started by a separate pressure g decreasing. The diesel pump is started by its own self-contained battery harger is provided for recharging. Both Millstone Unit Number 3 electric and umps deliver 2000 gpm at 100 psi discharge pressure and remain in operation ally shut down. Electrical interlocks stop the jockey pump when either of the Number 3 fire pumps start. supplied from two 250,000 gallon ground level tanks. The tanks are d through a water line fed from city water. y location of the MP-1 site should occur, the combined water tank and makeup uld provide an adequate water supply for MP-1. The necessary pressure and ntained through the use of any two simultaneously operating 2,000 gpm rated uppression Systems features for the Unit 2 Control Room are discussed in Section 9.10 of the inal Safety Analysis Report. 3.2-16 Rev. 8

tic Deluge Waterspray System (ESST Deluge System) e Manual Sprinkler Systems (Condenser Bay, Turbine Building Truck ing Area, and Reactor Building Rail Airlock Sprinkler Systems) concept for the fixed fire water suppression systems will use automatic ystems for the heated plant area (Maintenance Shop/CMS) and the ESST side the east wall of the Maintenance Shop. For the unheated plant areas, a uation concept will be used. The design will be to operate with dry pipes in the eas (Turbine, Reactor, and Radwaste Buildings) and flood up the piping activate the suppression system by opening a single isolation valve in the e Shop (Valve 1-Fire-37). This valve will be accessible to the plant operators ng fire department members outside the fire areas being protected by the dry er systems and deluge waterspray system have been designed using the f the National Fire Protection Association (NFPA) Standard Number 13 for the n of Sprinkler Systems or NFPA Standard Number 15 for Waterspray Fixed The dry manual operating concept is not in conformance with NFPA but has mined to be acceptable for the hazards of the decommissioned plant. utomatic Operating Sprinkler System tic, closed head, wet pipe design sprinkler system has been provided for the e Shop/Central Monitoring Station (CMS) area. This system has an alarm e which actuates an electric pressure switch to transmit a waterflow signal to he system is provided with an outside screw and yoke (OS&Y) isolation valve e supply connection and the system distribution piping. Sprinkler heads are t actuated type sprinkler heads. Operating Deluge Waterspray Systems tic, open head, deluge type waterspray system has been provided for the Station Services Transformer (ESST). This system has a deluge valve that on an input from a heat detection circuit located around the transformer. Upon n electric alarm switch actuates and transmits an actuation signal to the PLC lows into the distribution piping and discharges from all open spray heads. The an OS&Y isolation valve located between the supply header and the 3.2-17 Rev. 8

kler systems are provided in the unheated portion of the facility. These systems Condenser Bay, the Turbine Building Truck Unloading Area, and the Reactor ail Airlock. Sprinkler systems in the unheated portion of the plant are operated manual sprinkler systems. Each system has an isolation valve that separates the m the supply header. The systems have closed fusible type sprinkler heads. waterflow alarm provided. System piping has been arranged to facilitate raining during cold weather conditions. These systems would be charged with anually opening isolation valve 1-Fire-37 located in the Maintenance Shop rea as part of a fire fighting strategy for the facility. e Suppression Capabilities m Coverage m coverage is available to all fire areas of the plant from stand pipe connections inch hose stations or by use of 2.5 inch diameter hose with gated wye s available from outside hose houses. ations in the Maintenance Shop/CMS area are fed by the wet header piping ilable for immediate fire suppression use. The hose stations in the Turbine eactor Building, and Liquid Radwaste Building are fed off of the dry fire er and will be available for fire fighting following the flood-up of the header he opening of valve 1-Fire-37 in the Maintenance Shop. Hose stations in the aste Building are fed directly off a connection to the yard fire main and are wet with heat tracing on the piping and valves to prevent freezing in this ea. n locations are shown in the FHA (Reference 3.2-19). tinguishers nd placement of portable fire extinguishers are in accordance with the intent of es of NFPA Standard Number 10, Standard for Portable Fire Extinguishers. ishers utilized are Underwriters Laboratories (UL) listed. 3.2-18 Rev. 8

ems are used for early warning detection and in some cases may have the e fixed fire suppression systems. consist of fixed temperature detectors and smoke detectors. Smoke detectors

, employing the ionization principle. Specific application of these detectors in tailed in the FHA (Reference 3.2-19).

allation of detector units is in accordance with the intent of the guidelines set dard Number 72E, Standard on Automatic Fire Detectors. rs, as with waterflow indicators, and valve tamper devices are arranged to local alarm panels and a fixed suppression system control panel, if applicable. re also transmitted through the local alarm panels to control panels in the Station (CMS). A Fire Alarm panel located in the CMS monitors those areas rt the Spent Fuel Pool Island. Trouble signals for these devices are transmitted

r. A general alarm is provided in the Unit 2 Control Room. Identification of the ble signals must be performed locally in the Unit 1 CMS.

also monitors other miscellaneous fire protection system features. ion Systems and Smoke Removal ducts of combustion from any specific plant area requires the use of the ation system, which is designed to handle the expected normal environment or the use of portable exhaust fans by the fire brigade. There are no cable r other unventilated areas that pose any special venting problems. Removal of e waste either from plant processes or airborne particulates requires the use of d filtration systems of potential radiation release areas are discussed in detail rbine, Radwaste, Radwaste storage, and Screenhouse Buildings in the FHA, aluation and Fire Hazards Analysis ion Criteria e overall Fire Protection Program as indicated by the FHA, 3.2-19 Rev. 8

nit Number 1. BTP APCSB 9.5.1 provides the guidelines acceptable to the ementing the following criteria: sign Criterion 3 (10 CFR 50, Appendix A) - Fire Protection. -Depth Criterion: For each fire hazard, a suitable combination of fire fire detection and suppression capability, and ability to withstand safely the fire is provided. Both equipment and procedural aspects of each are ure Criterion: No single active failure shall result in complete loss of protection primary (fix installed systems) and backup fire suppression capability extinguishers). ssion System Capacity and Capability: Fire suppression capability is provided, ty adequate to extinguish any fire that can credibly occur and have adverse quipment and components important to safety. e Suppression Capability: Total reliance for fire protection is not placed on a matic fire suppression system. Appropriate backup fire suppression capability in the form of portable fire extinguishers or hose stations. pecific guidance of the BTP, the evaluation considered the adequacy of the Fire on the effects of potential fire hazards throughout the plant based on sound ineering practices and judgments. zard Analysis Methodology s evaluated by conducting a fire hazard analysis of individual fire areas and fire ant. The analysis methodology is described in the Fire Hazards Analysis zard Analyses Results alysis results for each fire area are contained in the FHA (Reference 3.2-19). 3.2-20 Rev. 8

irements found in Millstone Unit Number 1 Technical Requirements Manual g condition for operation and surveillance requirements for the fire protection nical requirements ensure the fire protection system is properly maintained and equipment and systems are subject to periodic inspections and tests in e intent of National Fire Codes and the Fire Protection Program. protection features will be subjected to periodic tests and inspections: alarm and detection systems pipe automatic sprinkler systems er spray systems rior fire water supply headers pumps barriers (walls, fire doors, penetration seals, fire dampers) nual suppression (fire hoses, hydrants, extinguishers) ervice including fire suppression, detection, and barriers will be controlled strative program and appropriate remedial actions taken. The program requires fire protection systems to be identified and appropriate notification given to the or evaluation. ant, remedial actions would include compensatory measures to ensure an ire protection in addition to timely efforts to effect repairs and restore ce. 3.2-21 Rev. 8

igade and Training de and Nuclear Training are a site (Units 1, 2, and 3) organizations. The Brigade consists of a minimum of a Shift Leader and four Fire Brigade upplies an advisor, who is at a minimum a fully qualified Unit 1 Plant or, to the Fire Brigade Shift Leader. The advisor will provide direction and plant operations and priorities. re Brigade are trained by the Nuclear Training Department. ersonnel are responsible for responding to all fires, fire alarms, and fire drills. ity, a minimum of a Shift Leader and four Fire Brigade personnel remain in the rea and do not engage in any activity which would require a relief in order to

.g., continuous fire watch).

ded to fight a fire, additional equipment and manpower is supplied by the off rtments. Within a 5 mile radius of the plant there are numerous local volunteer tters of commitment to supply public fire department assistance have been e fire companies. oordinates the Site Fire Brigade activities, and ensures proper communications f support for the local fire department chief or officer in charge once on site, ctivities (e.g., Chemistry, Health Physics, and Security). oordinates with the Site Fire Marshal and periodically familiarizes local fire nel with the Stations layout and fire fighting equipment. The Site Fire Marshal e Site Fire Brigade Personnel and all Unit Shift Managers, informing them of e fire protection equipment, should equipment become inoperable or ls are planned and critiqued by Nuclear Training and members of the responsible for plant fire protection. Performance deficiencies of the Fire idual Fire Brigade personnel are remedied by scheduling additional training igade or individuals. 3.2-22 Rev. 8

mber 50-245, LS05-82-03-060, J. Shea to W.G. Counsil, 'SEP Topic IX-1, Fuel illstone 1)," March 9, 1982. mber 50-245, B10301, W.G. Counsil to D.M. Crutchfield, 'Millstone Nuclear on, Unit Number 1, SEP Topic IX-1, Fuel Storage,' August 31, 1981. mber 50-245, B10346, W.G. Counsil to D.M. Crutchfield, 'Millstone Nuclear on, Unit Number 1, SEP Topic IX-1, Fuel Storage,' December 14, 1981. mber 50-245, B12961, M.L. Boyle to E.J. Mroczka, 'Millstone Nuclear Power it Number 1, Issuance of Amendment Number 40 (TAC No. 68157),' 27, 1989. mber 50-245, A08680, M.L. Boyle to E.J. Mroczka, 'Millstone Nuclear Power it Number 1, Issuance of Amendment Number 43 (TAC No. 72183)," March mber 50-245, J.W. Andersen to J.F. Opeka, 'Millstone Nuclear Power Station, er 1, Issuance of Amendment Number 89 (TAC No. M93080)," November 9, deleted. Dominion) letter to U.S. NRC, Millstone Power Station, Unit Number 1, mber 50-245, Fuel Storage Requirements, Technical Specification 4.2, Letter 8972, dated Sept. 18, 2003. ort Number H1-971914, Revision 1, Analysis Of 1675 Pound Fuel Assembly p Onto The Irradiated Fuel Assembly. ort Number AH1-971691, Revision 0, Criticality Safety Analysis Of The With A Dropped Fuel Assembly. ort Number H1-971698, Revision 0, Flow And Temperature Field Analysis ed Cell Blockage In The Millstone Unit Number 1 Spent Fuel Pool. ort Number H1-971675, Revision 1, Analysis Of Tetrabor And Boraflex er 1675 Pound Fuel Assembly System Impact. 3.2-23 Rev. 8

on, Unit Number 1, SEP Topic IX-3, Station Service and Cooling Water November 24, 1981. mber 50-245, NUREG-0824, Integrated Plant Safety Assessment, Systematic Program, Millstone Nuclear Power Station, Unit Number 1, February 1983, , "Station Service and Cooling Water Systems. mber 50-245, B10292, W.G. Counsil to D.M. Crutchfield Millstone Nuclear on, Unit Number 1, SEP Topic IX-5, Ventilation Systems, November 19, mber 50-245, LS05-82-09-043, J. Shea to W.G. Counsil, SEP Topic IX-5, Systems, Millstone Nuclear Power Station, Unit Number 1, September 14, Analysis Millstone Unit Number 1, Revision 6, July 2000. uclear Power Station Fire Protection Program Manual. 3.2-24 Rev. 8

all amounts of solid waste as evaporator bottoms or contaminated materials rolled. Planned liquid effluent releases will be evaluated prior to release, and 11-1 s (e.g., monitoring) will be established. The Radiological Effluent Monitoring alculation Manual ensures that Unit 1 complies with 10 CFR 50, Appendix I. 4.1-1 Rev. 10

mber 1 is permanently shutdown and many installed components which are lding, are no longer required to safely store irradiated fuel. However, many of ponents continue to contain radioactive material or remain radioactive ing that was originally designed to shield these components while they peration, continues to provide shielding from the residual activity in the own condition. a drained down condition, a concrete shielding package is installed over the and reactor cavity floor to provide shielding from activated reactor vessel asis conditions determined the major portion of the original plant shielding design exceptions to this were the Control Room where shielding was determined by duced during the loss-of-coolant accident and the shutdown cooling system s determined by shutdown conditions. Although these conditions are no longer ere the bases for the unit shielding. n tilation systems is contained in Chapter 3. N PROTECTION PROGRAM tion ction program is established to provide an effective means of radiation anent and temporary employees and for visitors at the station. The radiation is developed and implemented through the applicable guidance of Regulatory on 0; 8.8, Revision 3; and 8.10 Revision 1. ction department and line function management implement and enforce the n program. sible for implementing the radiation protection program is defined in the QAP. 4.2-1 Rev. 5

4.2-2 Rev. 5 onsiderations e of facility radiation shielding is to reduce external dose to plant personnel in program of radiologically controlled personnel access and occupancy in evels which are both ALARA and within the regulations defined in 10 CFR 20. utdown and all fuel stored in the spent fuel pool, the number and magnitude of sources have been reduced substantially from the original bases for the n design features. al Considerations have been performed and will continue to be performed to ensure that plant posted and barricaded. 4.3-1 Rev. 2

d for disposing of this waste water will be using the Waste Water Processing ocated in the Unit 1 Reactor Building. The Waste Water Processing System 10,000 gallon Sample Tanks, recirculation pump, demineralizer, filters and The A RBFD sump will pump to the WWPS sample tanks, where the water ulated and sampled before subsequent discharge. Radiological monitoring will an in-line Liquid Effluent Monitor (RE-MG-110). Prior to discharge through gency Service Water discharge piping to discharging canal), dilution flow be established by crediting Unit 2 Circulating Water Flow to the common the future, the WWPS will be used to process, sample and discharge Unit 1 ater after all spent fuel assemblies are removed from the Spent Fuel Pool. thod for processing waste water will be using an eight (8) gallon per hour rator. Waste water collected in the A RBFD sumps will be pumped to a cted liquids may be surveyed for activity and pumped to the evaporator. The l be diluted in the Balance of Plant (BOP) Reactor Building Exhaust flow and d level release. Radiological monitoring will be conducted by a particulate P ventilation exhaust or by screening a grab sample of the process liquid. bottom of the atmospheric evaporator will be collected as required, and pecific Activity (LSA) trash. iquid process methods are available, the RBFD sumps can be pumped to ould permit the collected liquids to be processed at a later date, or sent offsite 4.4-1 Rev. 10

age facilities accept waste from Millstone Units 1, 2 and 3. Information esign criteria is presented in Section 11.4 of the Millstone Unit 3 Final Safety ASES bjective of solid waste management is to provide for processing, packaging and wastes, and to allow for radioactive decay and/or temporary storage prior to nd subsequent disposal. dling at Millstone Unit 1 ensures compliance with the following regulations ides: , Standards for Protection Against Radiation , Appendix I .55, Classification of Waste for Near Surface Disposal 6, Waste Characteristics , Quality Assurance Criteria for Shipping Packages of Radioactive Material Guide 1.143, Design Guidance for Radioactive Waste Management Systems, and Components Guide 8.8, ALARA Provisions DESCRIPTION anagement process is designed to accommodate the following radioactive ypical for BWR power plants: which consist of contaminated clothing, tools and small pieces of equipment omically decontaminated; miscellaneous paper, rags, etc., from contaminated m radioactive ventilation systems; used reactor equipment such as control rod control curtains, fuel channels and in-core ion chambers - Radioactivity levels ow enough to permit handling by contact, it is processed and stored in ers to allow for off site shipment. Used radioactive equipment may be stored 4.5-1 Rev. 2.1

CES uclear Power Station Unit Number 1, Docket Number 50-245, Annual e Effluents Report. 4.5-2 Rev. 2.1

e means for compliance with Nuclear Regulatory Commission (NRC) 20, 10 CFR 50 Appendix A General Design Criteria (GDC) 60, 63 and 64, dix I and Regulatory Guides (RG) 1.21, 4.15 and 8.8. esign Description xhaust Monitor on exhaust radiation monitor is designed with the capability to monitor, the discharge of gaseous radioactivity. Capability for sampling of particulate

 . Annunciation in the Millstone Unit 2 Control Room occurs if setpoints are tor cannot determine the individual activity level of the radionuclides in the vides the overall level and a basis for correlation with laboratory analyses of ple activities.

le is taken from the exhaust duct near the reactor building exhaust plenum. A e nozzle is positioned to obtain a representative sample of the turbulent and t air. The monitor is located in a heated enclosure on the 65 foot elevation of ng directly below the exhaust duct. The sample passes through a particulate d detection chamber (fixed volume) and is then expelled back into the exhaust te filters are periodically removed for detailed radiological quantitative ut is sent to the PLC for display and recording. The range of indication is x 100 ci/cc (Kr-85). xhaust Monitor 1 BOP ventilation exhaust flow is continuously sampled for radioactive mple is taken from the exhaust duct which runs along the north exterior wall of ng. A single point sample nozzle is positioned to obtain a representative sample ust air. The particulate sample skid is located in an insulated enclosure on the north wall, of the Reactor Building. The sample passes through a particulate pelled back into the exhaust duct. The particulate filter is periodically removed gical quantitative analysis. 4.6-1 Rev. 2.1

escription monitoring system detects, measures, and indicates ambient gamma radiation ed locations in the SFPI. It provides audible and visual alarms in the Millstone m (locally at some locations) when radiation levels exceed pre-selected values has operational failure. Table 4.6-2 lists the area radiation monitor locations ea Radiation Monitor ARM is a 3 channel digital unit. Each detector is a gama sensitive GM tube d in Table 4.6-2. Each channel is provided with a failsafe High, Warn and as well as an analog output. The alarms and analog output are sent to the PLC larm. Each unit has a built in check source and local audible and visual alarm CE W.G. Counsil to D.G. Eisenhut dated July 1, 1981, Haddam Neck Plant, uclear Power Station, Unit Numbers 1 and 2, Post TMI Requirements - o NUREG-0737, Docket Numbers 50-213, 50-245, 50-336. 4.6-2 Rev. 2.1

xhaust (1) Beta Sinctillator 10-6 to 100 ci/cc None Page 1 of 1 Rev. 2

REACTOR BUILDING er SENSOR AND CONVERTER LOCATION Range mR/hr 1 West Refuel Floor 0.01-102 2 East Refuel Floor 0.01-102 3 West Refuel Floor Hi Range 10.0-106 Page 1 of 1 Rev. 2

al mode, reactor related accidents are no longer a possibility. lyzed accident that is in this chapter is the fuel handling accident. Conservatism n, conformance to high standards of material and construction, the control of essure loads, and strict administrative control over plant operations all serve to of the fuel in the spent fuel pool. initiators, and new accidents that may challenge offsite guideline exposures, as a result of certain decommissioning activities. These issues will be scope and type of the decommissioning activities are finalized. T EVENT EVALUATION able Results for Design Basis Accidents (DBAs) considered to be unacceptable safety results for DBAs: e material release that results in dose levels that exceed the guideline values of 0. tem stresses in excess of those allowed for the accident classification by ndustry codes. xposure to plant operations personnel in the Millstone Unit 2 Control Room in REM whole body, 30 REM inhalation, and 75 REM skin. dling Accident Assumptions dent analysis assumptions are listed on Table 5.2-1. uel Handling Accident analytical evaluation are provided in Section 5.2. ical Consequences adioactivity release during a fuel handling accident are presented in Section 5.1-1 Rev. 2

5.1-2 Rev. 2 longer part of the plants design and licensing basis. Several fuel handling are still possible in the spent fuel pool. These scenarios are identified later in sequences of a fuel handling accident in the spent fuel pool are described in nservatism, a bounding analysis was made to calculate the radiological release l fuel rods in four (4) fuel assemblies in the spent fuel pool. Other assumptions ation are described later in this Section. The off site radiological consequences from 4 failed fuel assemblies or, for example, 248 fuel rods for 8x8 fuel stantially less than the 10 CFR Part 100 limits and are tabulated in this section. NDLING ACCIDENT SCENARIOS IN THE SPENT FUEL POOL of the following postulated fuel handling drop events were evaluated: pool gate (1200 lbs.) drop onto irradiated fuel and fuel storage racks in the ool. ssembly drop (600 lbs.) onto irradiated fuel and fuel storage racks in the spent Tri-Nuc Filter skid (965 lbs.) into the spent fuel pool and potential drop onto uel and fuel storage racks. drop of items (pumps, boxes, filters, stellite containers and tables) temporarily e spent fuel pool equipment rail onto irradiated fuel and fuel storage racks. irradiated fuel assembly onto other irradiated fuel in the spent fuel pool. ized two sophisticated elasto-plastic finite-element models. The first represents omponents, the second represents the rack with its pedestals, liner and ced concrete structure. The LS-DYNA3D computer code was used. Conservative assumptions and restrictive inputs were utilized to ound estimate of the calculated damage for the postulated drop event. mptions were utilized in the analysis: 5.2-1 Rev. 2

rag force opposed to the impactor movement is proportional to its velocity drag force is conservatively neglected. act mechanism transmission: or makes first contact with the fuel assembly handle which is located above the on. Furthermore, the handle is conservatively considered as a prefect rigid ut deformability or energy absorption capacity. riteria: n individual fuel rod is assumed to occur when the irradiated zircaloy material postulated failure stress (strain). For additional conservatism, the entire length l rod is assumed irradiated to the state where the brittle material behavior is of the lower guide ends (between the lower end of the fuel rod and the bottom ot considered as a failure of the supported rod. se additional accident scenarios has determined that the limiting event is the uel pool gate, which can result in extensive damage of the fuel assemblies, 54 ruptured fuel rods. The drop of the new fuel assembly resulted in damage to semblies, but no ruptured fuel rods were recorded for either the impactor or the rradiated fuel assembly results in failure of all 64 guide ends, but no rupture of hese results bounded all fuel types stored within the Millstone Unit Number 1 the analyses performed to date. GICAL CONSEQUENCES has certified to the NRC that there is a permanent cessation of operations of mber 1 and that fuel has been permanently removed from the reactor vessel, a ing the radiological consequences of a fuel handling accident in the spent fuel d and eventually chosen as the new bounding accident (Reference 5.2-2). t the actual source term of the fuel in the spent fuel pool (i.e., appropriate , the reanalysis assumed four fuel assemblies (e.g., 248 rods in an 8x8 the spent fuel pool and resulted in an unfiltered, i.e., no Standby Gas 5.2-2 Rev. 2

se at the exclusion area boundary 5.44E-04 REM se at the low-population zone 1.69E-05 REM y dose (calculated as TEDE) at the exclusion area boundary 1.03E-03 REM y dose (calculated as TEDE) at the low-population zone 3.20E-05 REM ell within the limits of 10 CFR 100, and are therefore acceptable. lculated to the Millstone Unit Number 2 Control Room. The results of this as follows: se to the Millstone Unit Number 2 Control Room 7.65E-02 REM y dose to (calculated as TEDE) the nit Number 2 Control Room 8.67E-02 REM ose to the Millstone Unit Number 2 om 2.19E+01 REM s than the limits specified in GDC 19. Doses were not calculated for the mber 3 control room since the atmospheric dispersion factor (/Q) is imes less that the (/Q) to the Millstone Unit Number 2 control room. e to the Millstone Unit Number 3 control room would be approximately 50 Millstone Unit Number 2 control room dose. CES 3D, Version 932, Livermore Software Technology Corporation, May 1, 1995. Package NUC-197, MP1 Defueled State - Radiological Analysis of a Fuel ccident, Duke Engineering and Services, October 11, 1999. 5.2-3 Rev. 2

fy conservative results based on actual burnup. Regulatory Guide 1.25 See Ref. 5.2-3. fy conservative results based on actual burnup Regulatory Guide 1.25 See Ref. 5.2-3. Factor = 60 Extrapolation of Regulatory Guide 1.25 DF to MP1 conditions. See Ref. 5.2-3. of Iodine above pool: Regulatory Guide 1.25 See Ref. 5.2-3. mental anic emblies in Core: 580 Technical Specifications l dose assessment: Number of fuel assemblies assumed to fail = 4 DSAR Section 5.2.2 ns from fuel rods: Regulatory Guide 1.25 & conservative assumption le Gases nes n for secondary containment Technical Specifications peration an unfiltered ground release

= 3.47 x 10-4 m3/sec                                                 Regulatory Guide 1.25 dispersion factor (/Q):                                              SEP Topic 11-2.c, Docket Number 50-245

= 6.10 x 10-4 sec/m3 1.90 x 10-5 sec/m3 or fuel = 3.8 years Based on the MP1 shutdown on November 4, 1995. Page 1 of 1 Rev. 2

g 100 percent of the Millstone Unit Number 1 nuclear plant, is Dominion ut, Inc.. MENT AND TECHNICAL SUPPORT ORGANIZATION ing the management and technical support organization is presented in Section .1-1. That information is incorporated herein by reference. Support for Operations ing the technical support for operations is presented in Section 1.0 of hat information is incorporated herein by reference. tional Arrangement ing the organizational arrangement is presented in Section 1.0 of Reference ation is incorporated herein by reference. NG ORGANIZATION anization tion is as shown in Reference 6.1-1. sonnel Responsibilities and Authorities ing the plant personnel responsibilities and authorities is presented in Section .1-1. That information is incorporated herein by reference. Shift Crews t crew composition is contained in the Administrative Controls section of the mber 1 Technical Specifications. 6.1-1 Rev. 3.2

Operations Manager or Assistant Operations Manager shall be a Certified l Handler. Radiation Protection Manager shall meet or exceed the qualifications of ulatory Guide 1.8, Rev. 1. CES surance Program Description Topical Report. National Standards Institute, ANSI N 18.1-1971, Selection and Training of wer Plant Personnel. 6.1-2 Rev. 3.2

uirements Manual (TRM) contains clarifications for certain technical a central location for other documents which place operating limits on the he TRM are controlled pursuant to the 10 CFR 50.59 process. 6.2-1 Rev. 2.1

the equipment manufacturers or other vendors is utilized as necessary. nuing basis is used to maintain a high level of proficiency in the staff. CY PLAN Millstone Nuclear Power Station Emergency Plan (Reference 6.3-1) addresses h in NUREG-0654, FEMA-REP-1, Criteria for Preparation and Evaluation of gency Response Plans and Preparedness in Support of Nuclear Power Plants, ber 1980 and NUREG-0737, Supplement 1. As such, the Emergency Plan eptable state of emergency preparedness and meets the requirements of 10 ppendix E thereto. L SECURITY PLANS Reference 6.3-2) states the security measures to be employed by the licensee f Units 1, 2 and 3 at the Millstone Nuclear Power Station, Waterford, st radiological sabotage. The plans have been submitted in accordance with 10 on 73.55, Requirements for Physical Protection of Licensed Activities in actors Against Radiological Sabotage. e measures to deter or prevent malicious actions that could result in the release rials into the environment though sabotage. This protection is provided armed guards, physical barriers, monitors, personnel access controls alarms, esponse to security contingencies, and liaison with appropriate law ies. ASSURANCE PROGRAM DESCRIPTION (QAPD) TOPICAL REPORT eveloped and implemented a comprehensive Quality Assurance Program nformance with established regulatory requirements as set forth by the y Commission, and accepted industry standards. The participants in the QAP gn, procurement, construction, testing, operation, maintenance, repair, and of nuclear power plants are performed in a safe and effective manner. l Report complies with the requirements set forth in Appendix B of 10 CFR applicable sections of the Safety Analysis Report. 6.3-1 Rev. 3.2

g Revision 6 to the Millstone Nuclear Power Station, Unit Numbers 1, 2, and cy Plan, dated November 4, 1991 [and subsequent revisions thereto submitted al basis]. letter to U.S. Nuclear Regulatory Commission, Millstone Nuclear Power it Numbers 1, 2, and 3, Physical Security Plan, Revision 15, dated December d subsequent revisions thereto. 6.3-2 Rev. 3.2

6.4-1 Rev. 2 EVIEW uties, areas of review responsibility, and requirements of both the plant and ew committees are described in the Quality Assurance Program Description eport (Reference 6.1-1). DENT REVIEW w of activities affecting the unit's safety is performed by the Management mmittee as described in the QAPD Topical Report (Reference 6.1-1). for activities affecting safety related systems, structures, or components is as APD Topical Report (Reference 6.1-1). 6.5-1 Rev. 3.2

ently cease further operation of the plant. Certification to the NRC of the n of operation and permanent removal of fuel from the reactor vessel, in 0 CFR 50.82 (a)(1)(i) & (ii), was filed on July 21, 1998 (Reference 7.1-1), at no longer authorized operation of the reactor or placement of fuel in the reactor licensee is to decommission the plant safely and in a cost effective manner. ntained in this section of the DSAR is based upon the best information . The plans discussed herein may be modified as additional information or conditions change. which are unique to the multi-unit Millstone Station require that certain mber 1 decommissioning activities be delayed and performed concurrently sioning of Millstone Unit Numbers 2 and 3. Other considerations may dictate certain decommissioning activities. Therefore, the approach to Millstone Unit Number 1 can best be described as a modified SAFSTOR. In ntamination and dismantlement activities may be undertaken early in the wherever it makes sense from a safety or economic viewpoint. For instance, certainty over access to a low level waste disposal site, early shipment of s will occur. The amount of decommissioning work completed prior to a depends upon a number of factors currently under evaluation. and the SAFSTOR options are approaches found acceptable to the NRC in its ronmental Impact Statement (GEIS) (Reference 7.1-2). decommissioning schedule is contingent upon three key factors: ed access to licensed low level waste (LLW) disposal sites, l of spent fuel from the site, and unding of the decommissioning activities. e Unit Number 1 has access to Chem-Nuclear Systems Barnwell, S.C. the Envirocare disposal site in Tooele County, Utah. Escalation costs for the ave been incorporated into financial planning. Additionally, the licensee has sibility that during the decontamination and dismantlement phases, access to evel waste disposal site could be denied or that the facility could be closed. 7.1-1 Rev. 5

nd 3. Currently, after spent nuclear fuel is removed from the Unit 2 and Unit 3 fely stored in the existing SFPs. Capacity of these pools was designed with the E high level waste repository would provide permanent storage. However, the truction and licensing of such a repository have been delayed. As is the case facilities as the SFPs approach full capacity, spent fuel from Millstone Unit ill be stored in the ISFSI. A description of the ISFSI is contained in the Unit fety Analysis Report. lity such as unavailability of a LLW disposal site, temporary shortfall in unding, or other unforeseen circumstances, 10 CFR 50.82 requires the licensee ability to suspend decontamination and dismantlement. ISSIONING APPROACH nning on decommissioning Millstone Unit Number 1 using a modified ch in which the decontamination and dismantlement of the systems, structures and facilities (i.e., DECON) are completed prior to and following a In this plan, an ISFSI may be constructed and the transfer of spent fuel from (SFP) could be completed during the SAFSTOR period. The SAFSTOR period mination and dismantlement of any remaining systems, structures, and ence in coordination with Millstone Unit Number 2 and Millstone Unit issioning. ts from the ISFSI to DOE are scheduled, when practicable, following the cing operations. Delays in the operation of the repository limits the transfer of the cost of long term spent fuel storage. ussion provides an outline of the current decommissioning plan activities and the remaining significant activities. The planning required for each activity, including the selection of the process to perform the work, is the start of work for that activity. des implementation of a site characterization plan, preparation of a detailed plan, and the engineering development of task work packages. The detailed ed to support the decontamination and dismantlement of systems, structures, e performed prior to the start of field activities. 7.1-2 Rev. 5

internals segmentation, including the upper core grid. the reactor cavity and reactor vessel. a radiation shielding package over the reactor vessel head and cavity floor. vities remain: d choose a dry fuel storage system, if pursued. Investigate and prepare for the licensing of an ISFSI and prepare procurement specifications for a fuel canister ancillary equipment. acterization ortion of the planning period a detailed site characterization was undertaken logical, regulated and hazardous wastes were identified, categorized, and s were conducted to establish the contamination and radiation levels throughout Number 1 portion of the site. This information is used in developing re that hazardous, regulated or radiologically contaminated materials are sure that worker exposure is maintained as low as reasonably achievable d surveys of the outdoor areas in the vicinity of Millstone Unit Number 1 may ough a detailed survey of the environs would likely be deferred pending of Millstone Unit Numbers 2 and 3. It is worthwhile to note that site a process that continues throughout decommissioning. As decontamination and rk proceed, surveys are conducted to maintain current characterization and that activities are adjusted accordingly. lysis of the reactor internals, the reactor vessel, and the biological shield wall a part of the site characterization. Using the results of this analysis, these lassified in accordance with 10 CFR 61 and form the basis for the detailed aging and disposal. The interior grid portion of the top guide structure was reater than class C (GTCC) material, was segmented from the reactor vessel, spent fuel pool in canisters sized to be compatible with ISFI dry storage ination he decontamination effort are two fold. First, to reduce the radiation levels lity in order to minimize personnel exposure during dismantlement. Second, to erial as possible to unrestricted use levels, thereby permitting non radiological 7.1-3 Rev. 5

the radiation sources reduces the radiation levels by significant amounts. mination of the reactor recirculation system may provide value through reduced valuation is performed to determine whether the expected reduction in the force exposure is justified by the costs associated with the decontamination. ults are sensitive to the amount and type of work to be performed prior to a Any decontamination method used employs established processes with well-al interactions. The resulting waste is disposed of in accordance with plant plicable regulations. ve of the decontamination effort is achieved by decontaminating structural ing steel framing and concrete surfaces. The method used to accomplish this is quires the removal of the surface or surface coating. This process is used ial and contaminated sites. commissioning Activities FR 50.2 a "major decommissioning activity" is any activity that results in l of major radioactive components, permanently modify the structure of the ults in dismantling components for shipment containing GTCC waste in 0 CFR 61.55. oning activities completed to date include the removal of the drywall head and tor vessel internals by segmentation. The drywall head was sectioned and sent

r. The reactor vessel internals, classified as GTCC, are limited to the interior uide structure, which has been segmented from the reactor vessel and is stored ol. The reactor cavity and reactor vessel have been drained. Without the GTCC everal options are available for later removal and disposal of the reactor vessel:

ioning into pieces, or disposal as an intact package. ation of activity levels, ease of execution, personnel exposure, schedule al facility availability, and cost, segmentation of the internals may be postponed is removed from the SFP. ctor vessel follows the removal of the reactor internals and may not occur until period. It is likely that the vessel would be removed by sectioning or l sectioning or segmenting permits a substantial portion of the waste to be sent ssor instead of a near surface disposal site. The dismantling of the drywell and er is undertaken as part of the reactor building demolition. 7.1-4 Rev. 5

pent fuel management program, pursuant to 10 CFR 50.54(bb) ajor decommissioning activities listed above, the following decommissioning and regulated materials (e.g., asbestos, lead, mercury, PCBs, oil, chemicals) are uring characterization and plans are developed for the removal of these onents removed from the Turbine Building include the Turbine Generator, Feedwater Heaters, Moisture Separators and miscellaneous system and ipment. ous solid waste removed include: control rod blades, local power range pent resins and filters, the Reactor Pressure Vessel Head Insulation assembly, oner platform, and the Refuel Floor shield plugs. The larger components may ed and packaged for removal through the Reactor Building hatchway. tes are processed and discharged using plant procedures in accordance with regulatory requirements as the liquid waste inventories become available. inventories of the plant water systems are processed. Upon completion of the on and packaging of the reactor vessel internals, the reactor cavity and reactor ined and the waste inventory processed. When the spent fuel is removed, the ned and the water processed. Systems are then isolated and deactivated in a ompatible with the operations previously described. Spent fuel pool systems after removal of the spent fuel. aminated or activated materials are removed from the site as necessary to allow ed for unrestricted access. Low level waste is processed in accordance with nd existing commercial options, and sent to licensed disposal facilities or waste her volume reduction. Wastes may be incinerated, compacted, or otherwise rized and licensed contractors, as appropriate. Mixed wastes, if any, are g to all applicable federal and state regulations. Mixed wastes are transported and licensed transporters and shipped only to authorized and licensed 7.1-5 Rev. 5

, the final site survey using Reference 7.1-4 may proceed in two phases: 1) surveyed as decontamination and dismantlement are completed, and 2) external onjunction with completion of the Unit 2 decontamination and dismantlement.

uired to prepare a License Termination Plan (LTP) for Millstone Unit Number the details of the final radiological survey to be performed once the ctivities are completed. The LTP conforms to the format defined in Reference s the limits of 10 CFR 20 using the pathways analysis defined in Reference 7.1-ance ensures that survey design and implementation is conducted in a manner h degree of confidence that applicable NRC criteria are satisfied. Once the , the results are provided to the NRC in a format that can be verified. oration he Millstone Unit Number 1 area of the Millstone site will be undertaken when license for Millstone Unit Number 1 is terminated. This event may coincide t Numbers 2 and 3 license terminations. Buildings, structures, and other not currently known to be radiologically contaminated, such as the Strainer e, and the Discharge Structure are dismantled, as part of the building fter the final license termination survey for Millstone Unit Number 1 is uildings can be removed late in the building demolition phase since there is no operational need to remove them earlier. Site restoration requires that all ed to an elevation 3 feet below grade or to an elevation consistent with the essary amounts of contaminated material. OF RADIOACTIVE WASTE GEIS (Reference 7.1-2) provides an estimate for low-level waste disposal from g water reactor (BWR) of 18,975 cubic meters (669,817 cubic feet) for both the TOR options. The licensee estimates the low-level waste burial volume for 1, will be at or below this value for the modified SAFSTOR alternative. The includes, by a reduction of approximately 40 percent (industry standard), the nt-day volume reduction techniques. For waste requiring deep geological waste, the licensee estimates that the volume for Millstone Unit Number 1 is at ubic meters anticipated for a reference BWR discussed in Section 5.4 of the ates support the conclusion that the previously issued environmental nding since the disposal of waste requires fewer resources, i.e., less waste ea, than what was considered in the GEIS. 7.1-6 Rev. 5

igh-level waste repository or some interim storage facility will not be least 2010. Shipments of fuel and GTCC waste to DOE are planned to be SFSI. urrently stored in the SFP. The licensee may license a dry, ISFSI. Fuel will be e pool and stored temporarily on site using licensed canisters. For the period of will be stored in the SFP, the systems necessary for SFP operations will be n Island concept and configured for SFP clean-up and cooling. el Waste aminated or activated materials are removed to allow the site to be released for . Low level waste is processed in accordance with federal and state regulations, nd existing commercial options, and transported to license disposal facilities. anagement t of the total cost of decommissioning Millstone Unit Number 1 is the cost of osing of systems, components and structures, contaminated soil, water and liquids. A waste management plan incorporates the most cost effective onsistent with regulatory requirements for each waste type. The waste will be based on the evaluation of available methods and strategies for ing, and transporting radioactive waste in conjunction with the available tions and associated waste acceptance criteria. N EXPOSURE MONITORING exposure is maintained ALARA and monitoring is conducted in accordance protection program described in Chapter 4. Exposure specifically related to activities is identified and tracked. Exposure monitoring is used to identify ausing excessive exposure and to initiate corrective actions to reduce personnel CES 388 from Bruce D. Kenyon to U. S. Nuclear Regulatory n,Certification of Permanent Cessation of Power Operations and that Fuel ermanently Removed from the Reactor, dated July 21, 1999. 7.1-7 Rev. 5

ar Regulatory Commission report NUREG-1575, Multi-Agency Radiation and Investigation Manual (MARSSIM), Final Report. ar Regulatory Commission report NUREG-1700, Standard Review Plan for Nuclear Power Reactor License Termination Plans," (Currently in Draft form). 7.1-8 Rev. 5

ve waste that is removed from the site occupies only a small portion of the proved waste disposal sites. The non-radiological environmental impacts are significant. ose exposure for decommissioning Millstone Unit No. 1 is less than described e of two main reasons. First, the licensee initiated a zinc injection program for 1 in 1987 that significantly reduced the buildup of contaminated corrosion e remaining plant operation period. Second, with the plant shutdown since y of leading radionuclides have reduced overall plant general dose levels time decontamination and decommissioning activities occur. tified in this chapter resemble the DECON option. Therefore, the modified tional and public dose exposure is compared to the DECON option dose in the ional and public dose effects for a modified SAFSTOR alternative is bounded tion. The exposure from decontamination and dismantlement activities and the ansportation of the low-level wastes is included in this dose estimate. NUREG-1-2), Table 5.3-2, estimates a total occupational dose of 18.74 person-Sv (1874 DECON alternative for the reference BWR plant. The values estimated by the or below this value. WORKER for external occupational radiation exposure that accumulate dose for workers during the dismantlement program are developed based on a task by rsonnel hours and expected radiation dose rates associated with each task. e based on the following: ARA principles are implemented. iation exposure is monitored to identify jobs that are causing excessive osure and corrective actions are taken to reduce the severity. PUBLIC he public is maintained below comparable levels when the plant was operating ued application of radiation protection and contamination controls combined ource term available in the facility. 7.2-1 Rev. 2.1

drivers during a 500 mile trip would probably spend no more than 12 hours ab and 1 hour outside the cab at an average distance of 6 feet from the truck. ck servicing en route would require that two garage men spend no more than 10 out 6 feet from a shipment. from the general public might be exposed to radiation when a truck stops for he drivers to eat. The onlooker dose is calculated on the basis that 10 people erage of 3 minutes each at a distance of about 6 feet from a shipment. ative dose to the general public from truck shipments is based on population x 10-6 man-rem per km. , Table 11.4-2, provided a generic estimate of the routing radiation doses from n of radioactive wastes. The doses are based on the maximum allowable dose ment in exclusive use trucks and are conservatively high, on the number of nd on the shipping distances. The estimated external radiation dose for routing ations is 110 man-rem to transportation workers and 10 man-rem to the general ates the volume of both high level and low level wastes to be less than the UREG/CR-0672. The total number of shipments of radioactive wastes is less determine the exposure in the NUREG/CR, and therefore the exposure to the kers and the general public is less than those identified above. 7.2-2 Rev. 2.1

EVENTS tone Unit Number 1 is the fuel handling accident and a detailed discussion can Chapter 5. The acceptance criteria for all other potential events at the plant are t the potential consequences of any postulated event are less than 1 REM at the RTATION ACCIDENTS idents have a wide range of severities. Most accidents occur at low speeds and or consequences. In general, as speed increase, accident severity also r, accident severity is not a function of vehicle speed only. Other factors, such ent, the equipment involved, and the location can have an important bearing on ge in a transportation accident is not directly related to accident severity. In a of the same severity, or in a single accident involving a number of packages, s may vary from none to extensive. In relatively minor accidents, serious s can occur from impacts on sharp objects or from being struck by other cargo. n very severe accidents, damage to packages may be minimal. f truck accidents used in the NUREG/CR-0672 study were based on accident e DOT. Accidents are classified into five categories as functions of speed and ive categories in order of increasing severity are: minor, moderate, severe, xtreme. Table N.5-3 of NUREG/CR-0672 provides the probabilities of h classification. frequencies, release amounts and radiation doses to the maximum exposed cted accidents for transportation of radioactive material are discussed in of NUREG/CR-0672. The frequencies are calculated by multiplying the total rt with the total probability of accident per distance traveled for each accident osed individual is assumed to be located 100 meters from the point of a dent. The calculated dose values provided in Table N.5.6 of NUREG/CR-0672 ose and the fifty year dose commitment to the bone, lung, thyroid and whole 7.3-1 Rev. 2.1

7.3-2 Rev. 2.1 decommissioning. The primary environmental effects of the decommissioning include small increases in noise levels and dust in the immediate vicinity of the eases in truck traffic to and from the site for hauling equipment and waste. imilar to those experienced during normal refueling outages and certainly less resent during the original plant construction. No significant socioeconomic to local culture, terrestrial or aquatic resources have been identified. NAL CONSIDERATIONS tive, the following considerations are also relevant to concluding that activities do not result in significant environmental impacts not previously of effluents continues to be controlled by plant license requirements and plant rocedures throughout the decommissioning. ct to radiological releases, Millstone Unit No. 1 continues to operate in with the Offsite Dose Calculation Manual during decommissioning. non-radiological effluents continues to be controlled per the requirements of and State of Connecticut permits. ed to treat or control effluents during power operation are either maintained or temporary or mobile systems for the decommissioning activities. rotection principles used during plant operations remain in effect during ioning to ensure that protective techniques, clothing, and breathing apparatus appropriate. econtamination and source term reduction prior to dismantlement are to ensure that occupational dose and public exposure do not exceed those n the Final Generic Environmental Impact Statement (Reference 7.1-2. e radiological surveys are performed prior to starting the waste campaigns to burial volume of low-level radioactive waste and highly activated components ire deep geological disposal. f radioactive waste is in accordance with plant procedure, applicable Federal , and the requirements of the receiving facility. 7.4-1 Rev. 2.1

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15 LA N O dJ" 50 Miles cot'_- '{ORI< 0  :,o 5 5to J 10 15 I __ ~E'fI L. ft N SCALE - MILES IS SCALE -MILES 2.1-1 FIGURE2.1-I FTGURE General SiteLocation GeneralSite Location Millstone Nuclear Power Station Millstone Nuclear Power Station September1999 September 1999 Rev. 2 Rev. 2

MPS-I DSAR MPS-l DSAR FIGUR E 2.1-2 FIGURE GENERAL 2.1_2GENER VICINITY AL VICINI TY I MNP S-1D SAR ) NIA~TIC o BAY

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                                                  &.~<t,~ .       :    FIGURE 2.1-2 LON G ISL AND  SOu NO                                                 General Vicinit y Millsto ne Nuclear power station September 1999 September      1999 Rev. 2

MPS-1 DSAR MPS-I DSAR FIGURE 2.1-3 FTGURE SITE LAYOUT 2.1-3 SITE LAYOUT

                                                                                                                     *+_:

JOROf JORDAN COVE COVE ISFSIAREA lara a tcr ntr to ElÉl ldÉat aa ltrttf ftat I fKffi

**r**i LEGEN' {-r; m y#i,"ff*?.----           a ffilnr 0        500     1000 I          I       J                                                              qICT
       . SCALE-FEET sc--FET (rrll(t 0            zfi 2S()         SCI!)                                              ttüddÉ,

I BOP & SFPI SCALE-METERS SCA.8-lrETEr Ventilation Exhaust Ventilation Exhaust tttAtttc NIANTIC 8.ât BAY Rev. 2.3 Rev. 2.3

MPS-l MPS-I DSAR FIGURE 2.1-4 2.I_4 SITE PLAN PLAN

                                               /-zOrS4A68 l craE!

LE~ND LGFr{o PRrvrrLr OWNED FI4&B! PR(VAT£L1'

                                                                .                oürrD DIIQttá R[CR(ATIOU ncRrlor. AREA aRE!

I oo 250 250 l I 500

                                                                                           ,l l-
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SCALE~FEET SCÂLE=FIET scRvrcE r tfR Sts "udot LOtttG t5 L A ttto soulv D Rev. 2.3

MPS-IDSAR MPS-l DSAR FIGUR FIGURE E 2.1-5 2.1_5TOWNTOWNS WITHIN S WITHI MILES N 10IOMILES MNPMNPS-1 S-1DDSAR SAR Cf( clf

              .EAST EAST H AOOAM HADDA M I
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                                                                              ,I LONG LONG            ISLATVD ISLAN D...-      .souilo
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                                                                                 'lr FIGURE 2.1-5 FIGURE 2.I-5 TOWN S WITHIN 10 MILES BAY TOWNS      WITHIN 10 MILES MILLSTONE NUCLEAR        POWER STATION MILLSTONE NUCLEAR POWER        STATION

-W* Septe mber1999 1999 September Rev. 2 Rev. 2

MPS-IDSAR MPS-l DSAR FIGURE 2.I-6POPULATION FIGURE2.1-6 POPULATIONSECTORS SECTORSFOR FOR0 O-10

                                                                                             - IOMILES MILES
                                           -I                                                                                              MNPS-1DSAR t~

I I I I I I N I AsL I aooAfi fiEAST HAODAM _i 'NNw- MOTTVILLE I

        ------1/

LEGEND

                                                                                                                 ;  LEGENO

____ TOWN BOUNDARY TOWN AOUNOARY _ _ _ COUNTY BOUNDARY _ _ _ STATE BOUNDARY I STATE AOUNoARY STO,NINGTON OLD SAYBROOK 024 t i l , I SCALE- MILES SCAL- IILES FIGURE 2'.1-6 FIGURE 2.1-6 Population Sectors for 0 - 10 Miles Population Sectors for 0 - 10 Miles s

                 ,,","~""R' Millstone Nuclear Power Station Millstone Nuclear Power Station sourHoLo
                          ,4

[-J,( 6'ánowRs DR' September 1999 September 1999 Rev,2 Rev. 2

MPS-I DSAR MPS-l DSAR FIGUR FIGURE 2.I-7 POPUL E 2.1-7 POPULATION ATION SECTO SECTORS FOR 0O -- 50 RS FOR MILES 50 MILES MNP MNPS.I S-1D SAR ac ct^ NW LEGEND LEGEND:: _ - COUNTY @UflTY BOUNDAR tT

                                                                                                   û t

Bo('TIOÂESIES _ _ _ STATE BOUNDARIES STATE 8üJNOARIES

                                                                         ~

N" f'()PU(.ATlO PoPuL,ÂftolN carER 8Q(JNOARY camR sû.n{ofY C' ENE E o 5 10 I 15 I 30 ESE I I SCALE-MI LES SCAL-lIILES FIGURE FIGURE 2.I-72.1-7 popula Populationtion Sectors Sectors for 0 - 50 Miles Miles

                                            ,50 !/liles Millsto  ne    Nuclea Millstone Nuclear         r  Power Power      Station SE ssw s

September 1999 September Rev. 22 Rev.

MPS-I DSAR MPS-l DSAR FIGURE 2.1-8 FIGURE 2.1_8 ROADS ROADS AND AND FACILITIES FACILITIES IN THE LPZ IN THE LPZ (,D cf MNPS-1 MNPS.I DSAR . t .LEGEND LEGEND _ - - TOWN TOWH BOUNDARY SOUNDARY EA E ASS T LY'y. L - PRIMARY PRIIIARY ROADS ROAOS f

                                                                                                                 ~ P&W          I AMTRAK RAILROAD PAW/A.TR.AK   RALFOAD
                                                                                                                   @      STATE STATE ROUTES ds ROUTES OJ tr     NIANTIC NIAiITIC ELEMENTARY ELEMENTARY SCHOOL SCHOOL tr (g)    SOUTHWEST SOUTHWEST ELEMENTARY ELEMENTARY SCHOOL SCHOOL
                                                                                                                   @)

E NEW NEw LONDON LONoON COUNTRY COuIITRY CLUB E

                                                                                                                   @]GREATGRAT NECKNECK ELEMENTARY ELEMENTARY SCHOOL SCHOOL lID El     BA'NIEW   NURSING HOME BAwrwNURsrNGnoiltE E                  .SEASIDE
                                                                                                                              . SEASIOE REGIONAl.

REGIOL CENTER nt SCENT SEACH NIANT/C N IANTIC PLEASURE 8AY BAY ATTAWAN MILLSTONE POINT.

                                            .      /"              .

BLCK POINT ACH CLUE LONG LONG ts'íau'oi ISLANO' 1.!Y SOUN SOUNO

                                                                '!y'
                                                              "'tr s\)                            FIGURE 2.1-8
                                               ,./          I                                                   Roads and Facilities in the LPZ Roads o                        1/2                                       Millstone Nuclear Power Power Station Station LPZ BOUNDARY                                  SCALE-MILES SCAL-MILES (2.4 Miles)

September 1999 1999 Rev. 22 Rev.

MPS-1 DSAR MPS-I FIGURE 2.1-9 FIGURE 2.1-9 LPZ POPULATION SECTORS SECTORS DISTRIBUTION DISTRIBUTION MNPS-1 DSAR l,U N {

                                                                                                                                 ¿ E AS E A S T  LYM L   Y                          I b           W    AT       E"R l

I a C) A 3 n r4 I I l I

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E. SCENT i PARK BAY ATTAWAN

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                                                  / S L-'                     SOU N      0 t\
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                                                              .\                                                FIGURE 2.1-9
                                                          /      l\

l Distribution LPZ Population Sectors Distribution Millstone Nuclear Power Millstone Nuclear Station Power Station SW 1/2 1/2 PZ BOUNDARY LPZ EOUNDARY SCALE-MILES SCALE-MILE S SE (2.4 Miles)

(2.4 Miles)

                                                          ;\sse\         I ssw SSW                    SCI                  'SSE Septembl 1999 September         tgsg_

Rev. 2

MPS-I DSAR MPS-l DSAR FIGURE 2.1-IO INSTRUMENT FIGURE 2.1-10 INSTRUMENT LANDING LANDING PATTERNS PATTERNS AT AT TRUMBELL TRUMBELL AIRPORT AIRPORT MNPS-1DSAR MNPS-1 DSAR lts RwY 5 tüfllr ,u . cqccDirr VOR RWY 5 VORRWY 5 rrAuu.l. cor*crcúI voR RWY 23 VORRWY.23 a.totc iar,aru ofûo(accnclrf

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GaOTON. <OrCCnCrrI VORRWY 5 l'tta-tz.'t' 41 '200N - n'03'w 214 C~C\JI ruruuu TRUM8UU 'lr VOR RWY 23 r.o.H-r2.0!.rv 41'2O'N -72 '03'W c"oron. 216 NOTE: NOTE: PAGES FROM PAGES FROM DOO DOD FLIGHT FLIGHT INFORMATION PUELICATION- PUBLICATION- FIGURE 2.,l-102.1-10 FTGURE LOW ALTITUOE LOW ALTITUDE INSTRUMENT INSTRUMENT APPROACH APPROACH PROCEOURESPROCEDURES Instrument Landing Landing Patterns Patterns NORTHEAST UNfTEO NORTHEAST UNITED STATES STATES VOL-7 VOL-7 lnstrument at Trumbull at Trumbull Airport Airport Millstone Nuclear Millstone Nuclear Power Power Station Station ) September 1999 September 1999 Rev. 2

MPS-I MPS -l-- DSA DSARR FIGU FIGURE 2.I-II RE 2.1-1 1 AIR AIR LAN ADJACENT ES ADJA LANES CEN T TO MILLSTONE TO MILL POII{T STO NE POIN T MNP IT'INPS.1 S-1 DSADSAR R B el ftL.- ilJ \o "f

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MPS-IDSAR MPS-l DSAR FIGURE2.1-12 FIGURE 2.1_12NEW NEWLONDON LONDON COUNTY COUNTY - -STATE HIGHWAYSAND STATEHIGHWAYS ANDTOWN TOWNROADS ROADS MNPS-1 DSAR

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                             'Pt i

Highways and and Town Town Roads Roads i o SCALE-K SCALE-KI 2 ILOMET 3 4 Highways Millstone Nuclear Millstone NuclearPower PowerStation Station ) September 1999 September 1999* Rev. 2 Rev. 2

MPS-I-MPS- I-DSA DSAR R FIGURE 2.3_I lOPOTOPOGRAPHY GRAP HY IN THE VICIN VICINITY ITY OFOF MILL STON E POIN MILLSTONE POINT T FIGU RE 2.3-1 fts 150 .

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                                '5 .S       0 September Sep        temb er 1999 il .0 Rev.2 Rev. 2

MPS-IDSAR MPS-1 DSAR FIGURE FIGTJRE 3.1-1 3.1-1REACTOR REACTORBUILDINGBUILDINGSEISMIC SEISMICLOADS LOADS BUILDING WEIGHT BUILDING ITEGHTAND ND SECTION SECTIONPROPERTIES PROPERTES tç Fr.al Ac GT.z) GT.z) Elo H7 FT. - 2~IN. 324KK .

                                                       )S24 1                            ~

N 60.0 K-SEC ~l. 60.0 K-SEC kT. 1,464,266 r,464¿66 527.0 5n.o 271.0

                                    ....                                                                             nt.o LL EL. I29 FT.

El.129 FT. - 0 IN. a> 2244K 2244K 2 z W

                                    ....               6S.7 5s-7 K-SEC    ~l.

K-SEC?TI. ),464,266 1,464266 . 527.0 527.0 271.0 27t.0 LL 0 N tO8 FT. EL. 108 Elo FT. - 6 IN. 17.751.05 t7.751.05 K K 4 3

                                     ~                 571.6   K-S[C ~T.

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                                     .........          445.9  K-SEC ~T.

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        !!EL. -26   rT. -- 00 IN.
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• EEfTEClIVf TECIJYT S}EÂRSf£AR ÂR AREA FIGURE 3.1-1 FIGURE Reactor Building 3.1-1 Reactor Building Seismic Seismic Loads Loads September 1999 September 1999 Rev. 2 Rev.2

MPS-l MPS-I DSAR FIGURE 3.I-2 ACCELERATION FIGURE 3.1-2 ACCELERATION DIAGRAM UNDER SEISMIC LOADS 5 PERCENT PERCENT DAMPING DAMPING fL. 147 FT. - 2~lN. 150 150 l.129 Fr.-- 00 IN. 129 FT. ln 120 to8 FT.- 6 llt h90 I-w U w

                   ~

EI.. 82 FT. - 9 IN. fllJIPIJENi EqTPUENT SElSMIC CURVE CLNVE FOR STISMIC COEFfJCENT FOR RIGID COEFFICEN' RGID EQUlPMENT EQI,PIENT IN IN BUllDlNG BI,LOING IINCLUDES ONCLUDES flEXURAL FLTRJRAL

             ~                                           ANO AND ROCKING ROCKING \.lODES)

UODES) z=

z. EL 65 FI. - 9 IN.

0 PGo i=

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              .-J          EL. 42 Fr. - 6    N.

U w NOTE: N0TE: FOR CRITICAI- EQUIP~ENT FOR CRITICAl TUJIPMENT HAVING HAVING A PERIOO PERIM OF Of VIBRA lION VIBRÂTION 30 GREATER GREATER THANTHAN 0.05 O.O5 SECCtlDS SECO{DS A A DYNAMIC DYNAMIC ANALYSIS WAS ANALYSS TAS EL t4 FT. -- 66 IN. EI. 14 ]N. PERFOOIlEO PERFMUED CONSIDERING CONSIDERING BUILDING 8I'ILDING INTERACT1!li INERACTIOi fL. 00FT.-0!N. Fl. - 0 IN. 0 _ _- L_ _ __- L_ _ ____________ _~~~ L-~ ~ ~ o0 .10

                          .r0 .20 20 .30  .30 .40.40 .50 .50 .60 .60 ACCELERA
                             ÂCCELERAT!0NliON INlN *'tE.

• UNJTS

                                                         'UNITS FIGURE 3.1-2 Acceleration Diagram FIGURE 3.1-2                         Diagram UnderUnder Seismic Loads 5 Percent Damping September September 1999 1999 Rev. 2

MPS-IDSAR MPS-l DSAR FIGURE 3.I_3SHEAR FIGURE 3.1-3 SHEARDIAGRAM DIAGRAM UNDER UNDERSEISMIC SEISMICLOADSLOADS

    ]50                 - ZrN.

EL t47 rT. EL.129 FT.-- 00 IN. EL. t29 rT. I.r. EL IOBrT. EL 108 FT,-- 65 IN. IN. n90 I-wU w4 90 EL. B? n. EL. 82 FT.-- 99 IN. IN.. u.. ~

=

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    -30 0     2      4      68t01?14 6           8      10      12      14       16 SHEAR SHIAR IN  IN 1000 IOOO KIPS KIPS FIGURE 3.1.3 FIGURE    3.1-3 Shear Shear DiagramDiagram Under          Seismic Loads Under Seismic        Loads September 1999 September   1999 Rev.22 Rev.

MPS-IDSAR MPS-1 DSAR FIGURE 3.I-4 MOMENT FIGURE 3.1-4 MOMENT DIAGRAM DIAGRAM UNDER SEISMIC LOADS UNDER SEISMIC TOADS rl EL 147 1:1) ri- - 2~IN. raz n. zjrH. EL IZ3 n. EL lZg FT- - 0O ]N.

                              ]N.

120 1 tL. r08 FT. - 6 tN.

...whso9:l        tL. 82 FT.   - 3 tN.
 ~

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