ML11231A618

Millstone Power Station, Unit 1, Defueled Safety Analysis Report, Revision 8.0

ML11231A618
Person / Time
Site:	Millstone
Issue date:	06/30/2011
From:	Dominion Nuclear Connecticut
To:	NRC/FSME, Office of Nuclear Reactor Regulation
References
Download: ML11231A618 (202)

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{{#Wiki_filter:REVISION 8.0 M ILLSTONE P OWER STATION U NIT 1 D EFUELED S AFETY A NALYSIS R EPORT REVISION 8.0THIS DOCUMENT INCORPORATES APPROVED CHANGES TO THE MPS-1 DSAR.REFER TO THIS DOCUMENT'S

SUMMARY

OF CHANGE FOR PARTICULARS. MPS-1 DSAR Page 1 of 1 Revision History The NRC's "Guidance for Electroni c Submissions to the NRC", in Section 2.2, Living Document Update s, requires that submittals: "...indicate the part(s) (e.g., chapter, section, or graphic) that has been changed as well as the general scope of the change." The Millstone Unit 1 DSAR's Revision History fulfil ls this requirement. REPORTING PERIOD 2009 - 2010REVISIONFSC PKG Document NumberDATESECTIONSummary Description of Changes7-- 04/09As identified in the 2009 NRC Submittal List of Changed Pages and submitted Summary of Change.Administrative (FSAR content not affected). Change indicator (s) and page change identification (s) present in the 2009 NRC Submittal removed in preparation for the 2010 NRC Submittal. This forms the base line for changes incorporated under the Revision 7series. Revision level of the authoring files are unchanged. This supports Revision/Change traceability.Revision 8 (2010 - 2011 Reporting Period) Revision Revision Release Date Change Activity Document Elements Affected (Sections, Tables, Figures) Summary Description of Changes8March 2011MP1-DFCR-2010-001S3.2.7.2Reflects change to substation nomenclature (Northeast Utilities Distribution Project) for off site power source. MPS-1 DSARiRev. 3.2 TABLE OF CONTENTSSection Title PageCHAPTER 1- INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

1.1INTRODUCTION

......................................................................................1.1-11.2GENERAL PLANT DESCRIPTION.........................................................1.2-11.2.1PLANT SITE AND ENVIRONS...............................................................1.2-11.2.1.1Location and Site........................................................................................1.2-1 1.2.1.2Site Ownership............................................................................................1.2-1 1.2.1.3Access to the Site........................................................................................1.2-11.2.1.4Description of the Environs........................................................................1.2-1 1.2.1.5Geology.......................................................................................................1.2-1 1.2.1.6Seismology and Design Response Spectra.................................................1.2-1 1.2.1.7Hydrology...................................................................................................1.2-2 1.2.1.8Meteorology................................................................................................1.2-2 1.2.1.9Site Environmental Radioactivity Monitoring Program.............................1.2-21.2.2

SUMMARY

PLANT DESCRIPTION.......................................................1.2-3 1.2.3SYSTEMS..................................................................................................1.2-3 1.2.3.1Fuel Storage and Fuel Handling.................................................................1.2-3 1.2.3.2Radioactive Waste Processing System.......................................................1.2-31.2.3.3Radiation Monitoring and Control..............................................................1.2-41.2.3.4Auxiliary Systems.......................................................................................1.2-51.2.3.5Station Communication System..................................................................1.2-5 1.2.3.6Station Water Purification, Treatment and Storage System.......................1.2-61.3IDENTIFICATION OF AGENTS AND CONTRACTORS......................1.3-11.3.1APPLICANT'S SUBSIDIARIES...............................................................1.3-11.3.2NUCLEAR STEAM SUPPLY SYSTEM SUPPLIER...............................1.3-11.3.3ARCHITECT/ENGINEER.........................................................................1.3-1 1.3.4TURBINE-GENERATOR SUPPLIER......................................................1.3-1 MPS-1 DSARTABLE OF CONTENTS (CONTINUED) Section Title PageiiRev. 3.21.4MATERIAL INCORPORTED BY REFERENCE....................................1.4-11.5CONFORMANCE TO NRC REGULATORY GUIDES..........................1.5-1 1.5.1

SUMMARY

DISCUSSION.......................................................................1.5-11.5.2REFERENCE..............................................................................................1.5-2CHAPTER 2- SITE CHARACTERISTICS2.1LOCATION AND AREA..........................................................................2.1-12.1.1POPULATION...........................................................................................2.1-22.1.1.1Population Distribution Within 50 Miles....................................................2.1-3 2.1.1.2Transient Population...................................................................................2.1-3 2.1.1.3Low Population Zone..................................................................................2.1-3 2.1.1.4Population Center.......................................................................................2.1-42.1.2LAND USE.................................................................................................2.1-5 2.1.2.1Description of Facilities..............................................................................2.1-52.1.2.2Pipelines......................................................................................................2.1-8 2.1.2.3Waterways..................................................................................................2.1-8 2.1.2.4Airports.......................................................................................................2.1-8 2.1.2.5Highways....................................................................................................2.1-9 2.1.2.6Railroads.....................................................................................................2.1-9 2.1.2.7Projections of Industrial Growth...............................................................2.1-102.1.3DETERMINATION OF DESIGN BASIS EVENTS...............................2.1-11 2.1.4EFFECTS OF DESIGN BASIS EVENTS...............................................2.1-12 2.

1.5REFERENCES

.........................................................................................2.1-122.2METEOROLOGY......................................................................................2.2-1 2.2.1REGIONAL CLIMATOLOGY..................................................................2.2-12.2.2LOCAL METEOROLOGY........................................................................2.2-1 MPS-1 DSARTABLE OF CONTENTS (CONTINUED) Section Title PageiiiRev. 3.22.2.2.1Potential Influence of the Plant and Its Facilities on Local Meteorology......................................................................................2.2-12.2.2.2Local Meteorological Conditi ons for Design and Operating Bases...........................................................................................................2.2-12.2.3ON SITE METEOROLOGICAL MEASUREMENTS PROGRAM.........2.2-12.2.4SHORT TERM (ACCIDENT) DIFFUSION ESTIMATES......................2.2-22.2.4.1Objective.....................................................................................................2.2-2 2.2.4.2Calculations................................................................................................2.2-2 2.2.4.3Results.........................................................................................................2.2-2 2.2.5LONG-TERM (ROUTINE) DIFFUSION ESTIMATES...........................2.2-22.2.5.1Objective.....................................................................................................2.2-2 2.2.5.2Calculations................................................................................................2.2-3 2.

2.6REFERENCES

...........................................................................................2.2-32.3HYDROLOGIC ENGINEERING..............................................................2.3-12.3.1HYDROLOGIC DESCRIPTION...............................................................2.3-12.3.2SITE AND FACILITIES............................................................................2.3-1 2.3.3FLOODS.....................................................................................................2.3-1 2.3.3.1Flood History..............................................................................................2.3-12.3.3.2Flood Design Considerations......................................................................2.3-12.3.3.3Effect of Local Intense Precipitation..........................................................2.3-12.3.4PROBABLE MAXIMUM FLOOD (PMF) ON STREAMS AND RIVERS.............................................................................................2.3-22.3.5POTENTIAL DAM FAILURE, SEISMICALLY INDUCED...................2.3-22.3.6PROBABLE MAXIMUM SURGE AND SEICHE FLOODING..............2.3-22.3.6.1Probable Maximum Winds a nd Associated Meteorological Parameters...................................................................................................2.3-22.3.6.2Surge and Seiche Water Levels..................................................................2.3-3 2.3.6.3Wave Action...............................................................................................2.3-3 MPS-1 DSARTABLE OF CONTENTS (CONTINUED) Section Title PageivRev. 3.22.3.6.4Resonance...................................................................................................2.3-42.3.6.5Protective Structures...................................................................................2.3-42.3.6.6Probable Maximum Tsunami Flooding .....................................................2.3-42.3.7ICE EFFECTS............................................................................................2.3-4 2.3.8COOLING WATER CANALS AND RESERVOIRS...............................2.3-4 2.3.9CHANNEL DIVERSIONS.........................................................................2.3-4 2.3.10FLOODING PROTECTION REQUIREMENTS......................................2.3-42.3.11LOW WATER CONSIDERATIONS........................................................2.3-4 2.3.11.1Low Flow in Rivers and Streams ...............................................................2.3-42.3.11.2Low Water Resulting from Surges, Seiches, or Tsunamis.........................2.3-42.3.12DISPERSION, DILUTION , AND TRAVEL TIMES OF ACCIDENTAL RELEASES OF LIQUID EFFLUENTS SURFACE WATERS.................................................................................2.3-42.3.13GROUNDWATER.....................................................................................2.3-5 2.3.14TECHNICAL SPECIFICATION AND EMERGENCY OPERATION REQUIREMENTS..............................................................2.3-52.3.15REFERENCES...........................................................................................2.3-52.4GEOLOGY, SEISMOLOGY, AND GEOTECHNICAL ENGINEERING.........................................................................................2.4-12.4.1BASIC GEOLOGIC AND SEISMIC INFORMATION............................2.4-12.4.2VIBRATORY GROUND MOTION..........................................................2.4-1 2.4.2.1Safe Fuel Storage Earthquake.....................................................................2.4-1 2.4.3SURFACE FAULTING.............................................................................2.4-12.4.3.1Geologic conditions of the Site...................................................................2.4-1 2.4.3.2Evidence of Fault Offset.............................................................................2.4-12.4.3.3Earthquakes Associated with Capable Faults.............................................2.4-12.4.3.4Investigation of Capable Faults..................................................................2.4-12.4.3.5Correlation of Epicenters with Capable Faults...........................................2.4-2 MPS-1 DSARTABLE OF CONTENTS (CONTINUED) Section Title PagevRev. 3.22.4.3.6Description of Capable Faults.....................................................................2.4-22.4.3.7Zone Requiring Detailed Faulting Investigation........................................2.4-22.4.3.8Results of Faulting Investigation................................................................2.4-2 2.4.4STABILITY OF SUBSURFACE MATERIALS AND FOUNDATIONS........................................................................................2.4-22.4.5STABILITY OF SLOPES..........................................................................2.4-2 2.4.6EMBANKMENTS AND DAMS...............................................................2.4-2 2.

4.7REFERENCES

...........................................................................................2.4-2CHAPTER 3 - FACILITY DESIGN AND OPERATION3.1DESIGN CRITERIA..................................................................................3.1-13.1.1CONFORMANCE WITH 10CFR50 APPENDIX A GENERAL DESIGN CRITERIA..................................................................................3.1-13.1.1.1Summary Discussion..................................................................................3.1-13.1.1.2Systematic Evaluation Progr am and Three Mile Island Evaluations of General Design Criteria......................................................3.1-13.1.2CLASSIFICATION OF STRU CTURES, SYSTEMS, AND COMPONENTS ........................................................................................3.1-13.1.2.1Seismic Classification.................................................................................3.1-13.1.2.2Safety Related Classification......................................................................3.1-3 3.1.2.3Non-Safety Related Plant Functions Maintained in the Defueled Condition.....................................................................................3.1-43.1.2.4SSCs Important to the Defueled Condition................................................3.1-4 3.1.3WIND AND TORNADO LOADINGS......................................................3.1-8 3.1.4WATER LEVEL DESIGN.........................................................................3.1-83.1.5MISSILE PROTECTION...........................................................................3.1-83.1.5.1Internally Generated Missiles.....................................................................3.1-8 3.1.5.2Missiles Generated by Natural Phenomena................................................3.1-93.1.5.3Missiles Generated by Events Near the Site...............................................3.1-9 MPS-1 DSARTABLE OF CONTENTS (CONTINUED) Section Title PageviRev. 3.23.1.5.4Aircraft Hazards..........................................................................................3.1-93.1.6SEISMIC DESIGN.....................................................................................3.1-9 3.1.6.1Comparison of Measured and Predicted Responses.................................3.1-103.1.7DESIGN OF CLASS I AND CLASS II STRUCTURES.........................3.1-103.1.7.1Design Criteria, Applicable Codes, Standards and Specifications............................................................................................3.1-103.1.7.2Loads and Loading Combinations............................................................3.1-10 3.1.7.3Structural Criteria for Class II Structures.................................................3.1-123.1.7.4Seismic Class I and II Structures..............................................................3.1-133.1.8SEISMIC QUALIFICATION OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT.................3.1-163.1.9ENVIRONMENTAL DESIGN OF ELECTRICAL EQUIPMENT........3.1-163.1.10REFERENCES.........................................................................................3.1-163.2SYSTEMS..................................................................................................3.2-13.2.1FUEL STORAGE AND HANDLING.......................................................3.2-13.2.1.1New Fuel Storage........................................................................................3.2-1 3.2.1.2Spent Fuel Storage......................................................................................3.2-1 3.2.1.3Spent Fuel Pool Cooling System................................................................3.2-33.2.1.4Fuel Handling System ................................................................................3.2-53.2.2MONITORING AND CONTROL FUNCTIONS .....................................3.2-6 3.2.3DECAY HEAT REMOVAL (DHR) SYSTEM.........................................3.2-6 3.2.3.1Design Bases ..............................................................................................3.2-6 3.2.3.2System Description.....................................................................................3.2-7 3.2.3.3Safety Evaluation........................................................................................3.2-73.2.3.4Testing and Inspection ...............................................................................3.2-73.2.3.5Instrumentation ..........................................................................................3.2-7 3.2.4MAKEUP WATER SYSTEM....................................................................3.2-7 3.2.4.1Demineralized Water .................................................................................3.2-7 MPS-1 DSARTABLE OF CONTENTS (CONTINUED) Section Title PageviiRev. 3.23.2.5INTENTIONALLY DELETED.................................................................3.2-83.2.6PROCESS SAMPLING SYSTEM.............................................................3.2-83.2.6.1Design Bases...............................................................................................3.2-8 3.2.6.2System Description.....................................................................................3.2-8 3.2.6.3Safety Evaluation........................................................................................3.2-9 3.2.6.4Testing and Inspection ...............................................................................3.2-93.2.7ELECTRICAL SYSTEMS.........................................................................3.2-9 3.2.7.1Introduction.................................................................................................3.2-93.2.7.2Off Site Source............................................................................................3.2-9 3.2.7.3Intentionally Deleted...................................................................................3.2-9 3.2.7.4On Site Electric System..............................................................................3.2-93.2.8AIR CONDITIONING, HE ATING, COOLING AND VENTILATION SYSTEMS.....................................................................3.2-113.2.8.1Reactor Building and SFPI Heating and Ventilation System...................3.2-113.2.8.2Radwaste Building Ve ntilation System....................................................3.2-133.2.8.3Intentionally Deleted.................................................................................3.2-14 3.2.8.4Turbine Building Heating and Ventilation...............................................3.2-143.2.9FIRE PROTECTION SYSTEMS.............................................................3.2-153.2.9.1Design Bases ............................................................................................3.2-15 3.2.9.2System Description ..................................................................................3.2-163.2.9.3Safety Evaluation and Fire Hazards Analysis...........................................3.2-193.2.9.4Inspection and Testing..............................................................................3.2-213.2.9.5Personnel Qualification and Testing.........................................................3.2-22 3.2.10REFERENCES ........................................................................................3.2-23 MPS-1 DSARTABLE OF CONTENTS (CONTINUED) Section Title PageviiiRev. 3.2CHAPTER 4 - RADIOACTIVE WASTE MANAGEMENT4.1SOURCE TERMS......................................................................................4.1-14.2RADIATION PROTECTION DESIGN FEATURES...............................4.2-1 4.2.1FACILITY DESIGN FEATURES.............................................................4.2-14.2.1.1Design Basis...............................................................................................4.2-1 4.2.1.2Ventilation..................................................................................................4.2-14.2.2RADIATION PROTECTION PROGRAM................................................4.2-14.2.2.1Organization................................................................................................4.2-14.3ALARA PROGRAM..................................................................................4.3-14.3.1POLICY CONSIDERATIONS .................................................................4.3-14.3.1.1Design Considerations................................................................................4.3-1 4.3.1.2Operational Considerations.........................................................................4.3-14.4LIQUID WASTE MANAGEMENT SYSTEMS.......................................4.4-14.5SOLID WASTE MANAGEMENT............................................................4.5-14.5.1DESIGN BASES .......................................................................................4.5-14.5.2SYSTEM DESCRIPTION..........................................................................4.5-14.

5.3REFERENCES

...........................................................................................4.5-24.6EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING.......4.6-14.6.1DESIGN .....................................................................................................4.6-14.6.1.1Design Basis...............................................................................................4.6-14.6.1.2System Design Description.........................................................................4.6-14.6.2AREA RADIATION MONITORING INSTRUMENTATION................4.6-24.6.2.1Design Bases...............................................................................................4.6-2 4.6.2.2System Description.....................................................................................4.6-2 MPS-1 DSARTABLE OF CONTENTS (CONTINUED) Section Title PageixRev. 3.24.6.3REFERENCE..............................................................................................4.6-2CHAPTER 5 - ACCIDENT ANALYSIS

5.1INTRODUCTION

......................................................................................5.1-15.1.1ACCIDENT EVENT EVALUATION.......................................................5.1-15.1.1.1Unacceptable Results for Design Basis Accidents (DBAs)........................5.1-15.1.1.2Fuel Handling Accident Assumptions .......................................................5.1-15.1.1.3Results.........................................................................................................5.1-1 5.1.1.4Radiological Consequences........................................................................5.1-1 5.

1.2REFERENCES

...........................................................................................5.1-25.2FUEL HANDLING ACCIDENT...............................................................5.2-15.2.1FUEL HANDLING ACCIDE NT SCENARIOS IN THE SPENT FUEL POOL..................................................................................5.2-15.2.2RADIOLOGICAL CONSEQUENCES......................................................5.2-2 5.

2.3REFERENCES

...........................................................................................5.2-3CHAPTER 6 - CONDUCT OF OPERATIONS6.1ORGANIZATIONAL STRUCTURE .......................................................6.1-16.1.1MANAGEMENT AND TECHNICAL SUPPORT ORGANIZATION......................................................................................6.1-16.1.1.1Technical Support for Operations...............................................................6.1-1 6.1.1.2Organizational Arrangement.......................................................................6.1-1 6.1.2OPERATING ORGANIZATION .............................................................6.1-1 6.1.2.1Plant Organization .....................................................................................6.1-1 6.1.2.2Plant Personnel Responsibilities and Authorities.......................................6.1-16.1.2.3Operating Shift Crews................................................................................6.1-1 MPS-1 DSARTABLE OF CONTENTS (CONTINUED) Section Title PagexRev. 3.26.1.3QUALIFICATIONS OF NUCLEAR PLANT PERSONNEL ..................6.1-26.1.3.1Qualification Requirements........................................................................6.1-26.

1.4REFERENCES

...........................................................................................6.1-26.2TECHNICAL SPECIFICATIONS ............................................................6.2-1 6.3PROGRAMS .............................................................................................6.3-16.3.1TRAINING.................................................................................................6.3-16.3.2EMERGENCY PLAN................................................................................6.3-1 6.3.3PHYSICAL SECURITY PLANS...............................................................6.3-1 6.3.4QUALITY ASSURANCE PROGR AM DESCRIPTION (QAPD) TOPICAL REPORT...................................................................................6.3-16.

3.5REFERENCES

...........................................................................................6.3-26.4PROCEDURES .........................................................................................6.4-16.5REVIEW AND AUDIT..............................................................................6.5-1 6.5.1ONSITE REVIEW......................................................................................6.5-16.5.2INDEPENDENT REVIEW........................................................................6.5-1 6.5.3AUDITS .....................................................................................................6.5-1CHAPTER 7 - DECOMMISSIONING7.1

SUMMARY

OF ACTIVITIES ..................................................................7.1-17.1.1DECOMMISSIONING APPROACH .......................................................7.1-27.1.1.1Planning .....................................................................................................7.1-2 7.1.1.2Site Characterization...................................................................................7.1-3 7.1.1.3Decontamination.........................................................................................7.1-37.1.1.4Major Decommissioning Activities............................................................7.1-47.1.1.5Other Decommissioning Activities.............................................................7.1-5 MPS-1 DSARTABLE OF CONTENTS (CONTINUED) Section Title PagexiRev. 3.27.1.1.6Final Site Survey and Termination of License...........................................7.1-67.1.1.7Site Restoration...........................................................................................7.1-6 7.1.2STORAGE OF RADIOACTIVE WASTE.................................................7.1-6 7.1.2.1High Level Waste.......................................................................................7.1-7 7.1.2.2Low Level Waste........................................................................................7.1-7 7.1.2.3Waste Management.....................................................................................7.1-7 7.1.3RADIATION EXPOSURE MONITORING..............................................7.1-77.

1.4REFERENCES

..........................................................................................7.1-77.2ESTIMATE OF RADIATION EXPOSURE..............................................7.2-17.2.1NUCLEAR WORKER ..............................................................................7.2-17.2.2GENERAL PUBLIC ..................................................................................7.2-1 7.2.3NORMAL TRANSPORTATION..............................................................7.2-27.3CONTROL OF RADIATI ON RELEASES ASSOCIATED WITH DECOMMISSIONING EVENTS ..................................................7.3-17.3.1IN PLANT EVENTS .................................................................................7.3-17.3.2TRANSPORTATION ACCIDENTS.........................................................7.3-17.4NON-RADIOLOGICAL ENVIRONMENTAL IMPACTS .....................7.4-17.4.1ADDITIONAL CONSIDERATIONS........................................................7.4-1 MPS-1  DSARxiiRev. 2 List of Tables Number TitleReporting PeriodRevision 8 (2010 - 2011 Reporting Period)TABLE 1.1-1Millstone Unit No.1 Licensing MilestonesTABLE 2.1-1This Table has be en Intentionally DeletedTABLE 2.1-2  1990 Population and Population Densities - Cities and Towns within 10 miles of MillstoneTABLE 2.1-3 Population Growth 1960 - 1990 TABLE 2.1-4 Population Distribution with in 10 miles of Millstone - 1990 CensusTABLE 2.1-5Population Distri bution Within 10 Miles of Millstone 2000 ProjectedTABLE 2.1-6Population Distri bution Within 10 Miles of Millstone 2010 ProjectedTABLE 2.1-7 Population Distribution Within 10 Miles of Millstone 2020 ProjectedTABLE 2.1-8 Population Distribution Within 10 Miles of Millstone 2030 ProjectedTABLE 2.1-9 Population Distribution With in 50 Miles of Millstone - 1990 CensusTABLE 2.1-10 Population Distribution Within 50 Miles of Millstone - 2000 ProjectedTABLE 2.1-11Population Distribu tion Within 50 Miles of Millstone - 2010 ProjectedTABLE 2.1-12Population Distribu tion Within 50 Miles of Millstone - 2020 ProjectedTABLE 2.1-13 Population Distribution Within 50 Miles of Millstone - 2030 ProjectedTABLE 2.1-14 Transient Population Within 10 Miles of Millstone 1991-1992 School EnrollmentTABLE 2.1-15 Transient Population Within 10 Miles of Millstone - EmploymentTABLE 2.1-16 Population Distribution Within 50 Miles of Millstone - 2030 ProjectedTABLE 2.1-17Low Population Zone Pe rmanent Population DistributionsTABLE 2.1-18 Low Population Zone Sc hool Enrollment and EmploymentTABLE 2.1-19Metropolitan areas Within 50 Mi les of Millstone 1990 Census PopulationTABLE 2.1-20Population Centers wi thin 50 Miles of MillstoneTABLE 2.1-21Population Density Within 10 Miles of Millst one 1990 (People per Square Mile)TABLE 2.1-22Population Density Within 10 Miles of Millst one 2030 (People per Square Mile)TABLE 2.1-23Population Density Within 50 Miles of Millst one 1990 (People per Square Mile)

MPS-1 DSAR List of Tables (Continued) Number Title xiiiRev. 2TABLE 2.1-24Population Density Within 50 Miles of Millst one 2030 (People per Square Mile)TABLE 2.1-25Cumulative Population Density Within 50 Miles of Millstone 1990 (People per Square Mile)TABLE 2.1-26Cumulative Population Density Within 50 Miles of Millstone 2030 (People per Square Mile)TABLE 2.1-27Descripti on of FacilitiesTABLE 2.1-28List of Hazardous Materials Po tentially Capable of Producing Significant MissilesTABLE 3.1-1Comparison with NR C General Design CriteriaTABLE 3.1-2Allowable Stresses for Class I Structures TABLE 4.6-1Effluent Radiation Monitors TABLE 4.6-2Area Radiation Monitoring System Sensor and Converter Locations for Millstone Unit No. 1TABLE 5.2-1Assumptions and Input Conditions fo r Fuel Handling Accident at Millstone Unit No. 1 MPS-1 DSARxivRev. 3.3 List of Figures Number TitleFIGURE 1.2-1Plot PlanFIGURE 1.2 - 2A General Arrangeme nt RAD Waste Buildings - PlansFIGURE 1.2 - 2B General Arrangeme nt RAD Waste Buildings - PlansFIGURE 1.2 - 3General Arra ngement Buildings RAD Wa ste Buildings - SectionsFIGURE 2.1-1General Site Location FIGURE 2.1-2General Vicinity FIGURE 2.1-3Site Layout FIGURE 2.1-4Site Plan FIGURE 2.1-5Towns Within 10 Miles FIGURE 2.1-6Population Sectors for 0 - 10 Miles FIGURE 2.1-7Population Sectors for 0 - 50 Miles FIGURE 2.1-8Roads and Facilities in the LPZFIGURE 2.1-9LPZ Populati on Sectors DistributionFIGURE 2.1-10Instrument Landing Pa tterns at Trumbell AirportFIGURE 2.1-11Air Lanes Adjacent to Millstone Point FIGURE 2.1-12New London County - State Highways and Town Roads FIGURE 2.3-1Topography in the Vi cinity of Millstone PointFIGURE 3.1-1Reactor Building Seismic Loads FIGURE 3.1-2Acceleration Diagram Under Seismic Loads 5 Percent Damping FIGURE 3.1-3Shear Diagram Under Seismic Loads FIGURE 3.1-4Moment Diagram Under Seismic LoadsFIGURE 3.1-5Displacement Di agram Under Seismic LoadsFIGURE 3.1-6Radwaste Buildi ng - Mathematical ModelFIGURE 3.2-1P&ID: SFPI, Fuel Pool Cooling System FIGURE 3.2-2P&ID: SFPI, Fuel Pool Cooling System FIGURE 3.2-3P&ID: SFPI, Fuel Pool Cooling System FIGURE 3.2-4P&ID: Reactor Building and HV AC Room SFPI Secondary Cooling (DHR) System MPS-1 DSAR List of Figures (Continued) Number Title xvRev.3.3FIGURE 3.2-5P&ID: Reactor Building SFPI, Make-Up Water SystemFIGURE 3.2-6P&ID SFPI HVAC System Composite FIGURE 3.2-7P&ID: HVAC B.O.P. System Composite FIGURE 3.2-8through 3.2-11 Intentionally Deleted FIGURE 3.2-12 P&ID: HVAC Balanc e Of Plant System CompositeFIGURE 3.2-13P&ID: HVAC Syst em (Radwaste Storage Building)FIGURE 3.2-14Fire Protection Composite MPS-1 DSAR1.1-1Rev. 2CHAPTER 1- INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

1.1INTRODUCTION

This Defueled Safety Analysis Report (DSAR) is submitted by the licensee in support of the decommissioning for Millstone Unit Number 1 at the Millstone Nuclear Power Station in Waterford, Connecticut. Dominion Nuclear Connectic ut, Inc. owns and is responsible for the decommissioning of Millstone Unit Number 1. The DSAR is the principle licensing source document describing the pertinent equipment, structures, systems, operational constraints and practices, accident analyses, and decommissioning activities associated with the existing defueled condition of Millstone Unit Number 1. As such, the DSAR is intended to serve in the same role as the Final Safety Analysis Report of Millstone Unit Number 1 during th e periods of power operation between 1970 and 1998. The DSAR is applicable throughout the decommissioning of Millstone Unit Number 1. The decommissioning process is dynamic. The issuance of the DSAR does not alleviate the licensee from continuing to follow all required surveilla nces, procedures, technical specifications or similar documents, until those documents are of ficially modified using approved processes. Drawings and figures of structur es, systems, or components (SSCs) included or referenced in the DSAR, are included within the licensing basis of the facility only to the extent that they show SSCs that are described in the text of the DSAR. Other contents of drawings and figures may not reflect the current confi guration of the facility and are not maintained. Construction of Millstone Unit Number 1 was au thorized by a provisiona l construction permit CPPR-20, on May 19,1966, in AEC Docket 50-245. Millstone Unit Number 1 was completed and ready for fuel loading during October 1970. Th e plant went into commercial operation on December 28, 1970. On July 21, 1998, pursuant to 10CFR50.82(a)(1)(i) and 10CFR50.82(a)(1)(ii), the licensee certified to the NRC that, as of July 17, 1998, Millstone Unit Number 1 had permanently ceased operations and that fuel had been permanently removed from the reactor vessel. The issuance of this certification f undamentally changes the licensing basis of Millstone Unit Number 1 in that the NRC-issued 10CFR50 license no longer authorizes operation of the reactor or emplacement or retention of fuel in the reactor vessel. Therefore, as of July 21, 1998, only those conditions or activities associated with the safe storage of fuel and radiological protection (including waste handling, storage and disposal) are applicable to the defueled Millstone Unit Number 1 plant.Millstone Unit Number 1 was a single-cycle, boiling water reactor with a Mark I containment which was designed, furnished a nd constructed by General Electric Company as prime contractor for the licensee. The General Electric Company engaged Ebasco Services Incorporated as architect-engineer. Millstone Unit Number 1 had a reactor thermal output of 2011 megawatts and a net electrical output of 652.1 megawatts. The Millstone site is located in the town of Waterford, New London County, Connecticut, on the north shore of Long Island Sound. MPS-1 DSARPage 1 of 1Rev. 2TABLE 1.1-1MILLSTONE UNIT NO .1 LICENSING MILESTONES EVENTDATEConstruction Permit IssuedMay 19, 1966FSAR FiledNovember 1, 1968 Provisional operating License IssuedOctober 7, 1970Full-Term Operating License IssuedOctober 31, 1986 Full Power LicenseOctober 7, 1970 Initial CriticalityOctober 26, 1970Synchronized to the GridNovember 1970 100 Percent PowerJanuary 6, 1971Commercial operationDecember 28, 1970Permanently Ceased OperationsJuly 21, 1998 MPS-1 DSAR1.2-1Rev. 3.41.2GENERAL PLANT DESCRIPTION1.2.1PLANT SITE AND ENVIRONS1.2.1.1Location and SiteThe site for the Millstone Nuclear Power Station c onsists of a tract of land of approximately 500 acres located in the town of Waterford, Connec ticut on the north shore of Long Island Sound and on the east side of Niantic River Estuary. It is located 3.2 mile s west-south-west of New London, and 40 miles south-east of Hartford, Connecticut. The site is bounded on the west, south, and portions of the east sides by Long Island Sound. Th e nearest residential boundary is 855 meters north-east of the major structures of Millstone Unit Number 1. Ch apter 2 contains more detailed information on the site and surrounding areas.1.2.1.2Site Ownership

The site is owned by Domini on Nuclear Connecticut, Inc.1.2.1.3Access to the Site The immediate area around the station, excluding the intake and discharge canal, is completely enclosed by a security fence. This fence establ ishes the protected area boundary of the station. Access to the station is cont rolled by Security Personnel.1.2.1.4Description of the EnvironsAdjacent to the site to the north and west is cultivated land with residential dwellings. The village of Niantic, consisting of a small commercial co mplex and attendant residential development, is

1.5 miles

north-west of the Reactor Building. Other re sidential areas adjoin the site at the end of the plant access road and at distances of 1 to 3 miles. New London, 3.2 miles ENE of the Reactor Building, is the nearest urban complex and includes mixed residential, commer cial, and industrial uses.1.2.1.5Geology The site area is underlain by Monson gneiss and Westerly granite. The Westerly granite intrudes the Monson gneiss, is more resi stant to weathering and therefore forms ridges. Seismic surveys disclosed no unusual or extreme subsurface conditions. Chapter 2 contains more detailed information on geology and seismic qualities.1.2.1.6Seismology and Design Response Spectra The Millstone Point site area is placed in Zone 2 (zone of moderate damage) on the seismic probability map of the 1964 Uniform Building Code. MPS-1 DSAR1.2-2Rev. 3.4 The seismic design for critical it ems for this station is based on dynamic anal ysis of acceleration or velocity response spectrum curves which are based on a ground motion of 0.07g.The preceding design criteria are for critical items only, that is, for Class I items. Class I items are defined in Chapter 3.1.2.1.7Hydrology The plant site natural grade level is at an elevation of approximately 14 feet above mean sea level. Because of the contours of the land and ground strata, and the dist ance of the reactor from water supplies, no water accidentally re leased from the plant can reac h industrial or drinking water supplies.Chapter 2.0 contains more detailed information on hydrology.1.2.1.8Meteorology The meteorology of the site area is basically that of a sea-coast location wi th relatively favorable atmospheric dilution conditions pr evailing. The inland terrain in Connecticut is not pronounced enough to produce any significant lo cal modifications of synoptic conditions at the shoreline. The shoreline areas do, however, experi ence local modifications of synopt ic patterns because of the temperature differences between air over land and air over water.The site is located in an area occasionally traversed by hurricanes. The design basis hurricane for Millstone has 124 mph maximum gradient winds and a 17 mph speed of translation. This is significantly more intense than the worst on record (hurricane of 1938). It has been estimated that a tornado can be expect ed to strike a point on the Millstone site about every 1,804 years. In spite of this low probability, the features of the plant important to the safe storage of irradiated fuel have been designed to withstand 300 mph winds. It is concluded that from the viewpoint of site meteorology, the site is suitable for the station as described. (Chapter 2 contains more detailed information concerning meteorology.)1.2.1.9Site Environmental Radi oactivity Monitoring Program An environmental radioactivity monitoring program was initiated and has been conducted at the site since April 1967. Data are co llected to measure radioactivit y present in the environs. The program is continuing in order to assure prompt detecti on and evaluation of any changes in radioactivity. MPS-1 DSAR1.2-3Rev. 3.41.2.2

SUMMARY

PLANT DESCRIPTION The plot plan (Figure 1.2-1) shows the general arrangement of Millstone Unit Number 1 on the Millstone Point site. The reactor building houses fuel storage facilities, refueling equipment and other auxiliary equipment.The Radioactive Waste Building, located northeas t of the Reactor Building, is a two-story concrete structure. The overall arrangement of this building is shown in Figures 1.2-2 and 1.2-3.1.2.3SYSTEMS1.2.3.1Fuel Storage and Fuel Handling 1.2.3.1.1Fuel Storage and Handling EquipmentThe spent fuel storage pool holds fuel assemblies, control rods, and small vessel components. The pool system contains provisions to maintain water cleanliness a nd instrumentation to monitor water level. Makeup water is available from the Unit 2 demineralized water system and the fire water system. The racks in which fuel assemblies are placed are designed and arranged to ensure subcriticality in the pool. The handling of spent fuel is performed within the Reactor Building. This employs a refueling platform for underwater fuel tran sport, storage racks for fuel and control rods in a storage pool, underwater fuel preparation stati ons, and floor mounted jib cranes. Control rods can be stored in the fuel pool racks or on hooks on the side of the pool.Structural design of the fuel st orage and equipment storage facilities meets all requirements for Class I structures. For additional information, refer to Chapter 3.1.2.3.1.2Fuel Pool Cooling System

The fuel pool cooling system pr ovides cooling for the spent fu el pool water when required. The fuel pool cooling system co nsists of a circulating pump, heat exchanger, skimmer surge tanks, system piping, valv es, and instrumentation and controls. Pool cleanup is provided by an in-pool demineralizer and filter. For additional informa tion, refer to Chapter 3.1.2.3.2Radioactive Waste Processing System The radioactive waste processing system is desi gned to control the release of plant-produced radioactive material to within the limits specified in 10CFR20 and Appendix I to 10 CFR50. This is done by collection, transfer, and evaporation. MPS-1 DSAR1.2-4Rev. 3.41.2.3.2.1Liquid Radwaste System Liquid waste drained or transferred to the Re actor Building sumps may be processed in an atmospheric evaporator. The distil late vapor will be diluted in the Reactor Building exhaust and released as a ground level release. Radiological monitoring will be conducted by a particulate monitor in the ventilation exhaust or by scre ening a grab sample of the process liquid. Concentrates in the bottom of the Reactor Buildi ng atmospheric evaporator will be collected as required, and disposed as Low Specific Activity (LSA) trash. Alternatively, this system could be utilized to pump the process liquids from the Reactor Building sumps to containers which would permit the process liquid to be processed onsite or offsite.1.2.3.2.2Solid Radwaste Handling

Solid wastes originating from nuclear system equipment maybe stored in the spent fuel storage pool and prepared for off site shipme nt in approved shipping containers. Solid wastes are collected and appropriately prepared for off site shipment. Examples of these solid wastes are filter residue, spent resins, paper, air fi lters, rags, and used clothing. For additional information, refer to Chapter 4.1.2.3.3Radiation Monitoring and Control 1.2.3.3.1Radiation Monitoring and Sampling The Spent Fuel Pool Island ventilation exha ust is monitored for gaseous radiation and particulates. A particulate sampling skid is provided for Unit 1 Balance of Plant (BOP) exhaust to permit sampling for any significant changes. Fo r additional information, refer to Chapter 4. 1.2.3.3.2Area Radiation Monitors Radiation monitors are provided to monitor for abnormal radiation at selected locations on the SFPI. These monitors actuate alarms when abnormal radiation levels are detected.1.2.3.3.3Liquid Radwaste Processing System Control The liquid radwaste system is designed to safely and economically collect, store, process, and dispose of, or recycle, all ra dioactive or potentially radioa ctive liquid waste generated. The system operates on a batch basis.1.2.3.3.4Solid Radwaste Control Solid radwaste can be transf erred to high integrity cask containers for shipment. MPS-1 DSAR1.2-5Rev. 3.41.2.3.4Auxiliary Systems1.2.3.4.1Decay Heat Removal (DHR) System The DHR system provides cooling water to the spent fuel pool cooling system. The system consists of circulating pumps, air-water heat exchangers, an expansion tank, air separator and associated piping valves and c ontrols, and a portable ethylene glycol addition pump and tank.1.2.3.4.2Monitoring and Control Functions

The Millstone Unit 2 Control Room is continuous ly manned, and serves as the control room for Millstone Unit 1. Mill stone Unit 2 Operations personnel are responsible for the monitoring and control of the Unit 1 spent fuel pool island (SFPI) and auxiliary systems via a computer console located in the Millston e Unit 2 Control Room.1.2.3.4.3Fire Protection System Fire protection and detection systems are provided at Millst one Unit Number 1 to protect structures, systems, and components importa nt to the defueled condition of the unit. The fire protection system includes a fire water supply system that consists of two fire water tanks, fire water pumps and a distri bution system that delivers fire water to all parts of the plant.Fire water systems within the pl ant protect individual hazards and include sprinkler systems and deluge systems.1.2.3.4.4Electrical Power System 1.2.3.4.4.1AC Power Supply The electric power system includes the electric al equipment and connecti ons required to supply power to station auxiliaries.1.2.3.4.4.2DC Power SupplyThe SFPI 125 V DC system is provided via rectified AC at the point of use. In addition, a separate decommissioning 125V DC system powered by batteries and a battery charger provides a source of DC power to the decomm issioning electrical system. SFPI 24V power is provided by power supplies within the SFPI Programmable Logic Controller (PLC) panels.1.2.3.5Station Communication System The plant communication system provides for reliable on site and off site communications both under normal and contingency conditions. MPS-1 DSAR1.2-6Rev. 3.41.2.3.6Station Water Purification, Treatment and Storage System This system provides deminerali zed makeup water to Millstone Unit Number 1 for use in the spent fuel pool. MPS-1 DSAR1.3-1Rev. 21.3IDENTIFICATION OF AGENTS AND CONTRACTORS1.3.1APPLICANT'S SUBSIDIARIES Dominion Nuclear Connecticut, Inc. is responsible for the deco mmissioning of Millstone Unit Number 1. Dominion Nuclear Connecticut, Inc. is a wholly owned subsidiary of Dominion Energy, which is wholly owned by Dominion Resources, Inc..1.3.2NUCLEAR STEAM SUPPLY SYSTEM SUPPLIER

General Electric Company wa s the nuclear steam system supplier for the plant. 1.3.3ARCHITECT/ENGINEEREbasco Services Incorporated was the Architect/Engineer for Mill stone Unit Number 1.1.3.4TURBINE-GENERATOR SUPPLIER The turbine generator was manufactured by General Electric Company. MPS-1 DSAR1.4-1Rev. 21.4MATERIAL INCORPORTED BY REFERENCESpecific sections of the Millstone Unit Number 2 and Number 3 FSARs are incorporated into the Unit 1 DSAR by reference. These sections are identified within the text of the DSAR. MPS-1 DSAR1.5-1Rev. 21.5CONFORMANCE TO NRC REGULATORY GUIDES1.5.1

SUMMARY

DISCUSSION The AEC issued Appendix A 'General Design Criteria' to 10CFR50 in July 1971. In November 1970, Safety Guides, later to become Regulator y Guides, began to be published. These guides provided acceptable means for complying with specified general design criteria. They were not in effect at the time Millstone Unit Number 1 began operation with Provisional Operating License (POL) DPR-21, issued October 7, 1970. Millstone Unit Number 1 submitte d summaries of compliance to these guides in the early 1970s in support of the application for a full-term operating license (Reference 1.5-1). Before acting on this application, the NRC (former ly AEC) initiated the Systematic Evaluation Program (SEP) in 1977 to review the designs of older operating nuclear reactor plants in order to confirm and document their safety. Millstone Unit Number 1 was identified as an SEP plant. The SEP objectives were:*To establish documentation that shows how the criteria for each operating plant reviewed compare with current criteria on si gnificant safety issues and to provide a rational for acceptable departures from these criteria.*To provide the capability to make integrat ed and balanced decisi ons with respect to any required backfitting.*To provide for early identification and re solution of any significant deficiencies.*To assess the safety adequacy of the desi gn and operation of currently licensed nuclear power plants.*To use available resources efficiently to minimize require ments for additional resources by NRC or industry.*To ensure that the safety assessments we re adequate for conve rsion of provisional operating licenses to full-term operating licenses.The final version of the SEP program report included the status of all applicable generic activities (TMI and USIs), including those that formed the basis for the Integrated Safety Analysis Program (ISAP) being implemented by th e Licensee. Based upon the acceptable conclusions reached in SEP, the NRC issued the full-term operating lice nse for Millstone Unit Number 1 on October 31, 1986. MPS-1 DSAR1.5-2Rev. 21.5.2REFERENCE1.5-1Millstone Nuclear Power Station Unit 1 Application for Full-Term Operating License, September 1, 1972. MPS-1 DSAR2.1-1Rev. 3CHAPTER 2- SITE CHARACTERISTICS2.1LOCATION AND AREAThe Millstone site is located in the Town of Waterford, New London County, Connecticut, on the north shore of Long Island Sound. Th e 524 acre site occupies the ti p of Millstone Point between Niantic Bay on the west and Jordan Cove on the east and is situated 3.2 miles west-southwest of New London and 40 miles southeast of Hartford. The Millstone Unit Number 1 containment structure is located immediatel y south of Millstone 2 and 3. The geographical coordinates of the centerline of the reactor is as follows:Millstone Unit Number 1The site is owned by Dominion Nuclear Connecticut, Inc. Figures 2.1-1 through 2.1-4 identify the site.The site protected area is considered the rest ricted area. The restricted area has been conspicuously posted and administrative proce dures, including periodic patrolling, have been imposed to control access to the area. For th e purpose of radiologica l dose assessment of accidents, the exclusion area boundary (EAB) wa s considered the actu al site boundary for overland sectors, except in the Fox Island / discharge channel area on the south end of the site. For all water sectors, the nearest la nd site boundary distance was used. Any significant normal releases are discharged to the atmosphere via the Unit Number 1 BOP exhaust point and the SFPI ventil ation exhaust point. The distance from the Unit Number 1 BOP exhaust point and the SFPI vent ilation exhaust point to the ne arest residential property boundary in the Millstone Point Colony development (Point A on Figure 2.1-3) is greater than 2,800 feet. This development, adjacent to the eastern site boundary, consists of single family homes on 104 half acre lots. One of the conditions of the sale of the site to the Hartford Electric Light Company and the Connecticut Light and Power Company wa s that permanent dwellings would never be permitted in the beach area of the development. Because of this restrict ion, normal release doses are calculated at Point A rather than at the nearest point on the site boundary. The licensee has complete control of activities within the exclusion area, except for the passage of trains along the Providence & Worcester (P&W) / Amtrak Railroad track which runs east-west through the site.To ensure the safety of people within the exclusion area during an emergency, an emergency plan for the site has been prepared. The plan include s provisions for alarms both inside and outside buildings and delineates the ev acuation routes and assembly areas to be used. The State of Latitude and LongitudeNorthing and Easting N 41° 18'32"N 173, 800 W 72° 10'04"E 759, 965 MPS-1 DSAR2.1-2Rev. 3Connecticut Emergency Plan also provides for the control of acti vities in that portion of the exclusion area extending offshore through a written agreement between the licensee and the U.S. Coast Guard at their station in New London, Connecticut. The owners have encouraged public use of por tions of the site. Ownership rights have not, however, been relinquished, and the owners can, a nd have provision to, fu lfill their obligations with respect to 10CFR20, "Standards for Protection Against Radiation." A portion of the exclusion area is leased to the Town of Waterford for public recreation and is used primarily for soccer and baseball games. Figure2.1-3 show s the general location of these activities. No attempt is made to restrict the number of persons using these facilities. Estimates of maximum attendance indicate that about 2,000 visitors could be within the exclusion area at any one time at the soccer and baseball fields. The licensee's Emergency Plan provides for removal of the visitors from the site. The number and conf iguration of roads and highways assure ready egress from the areas described a bove (Figures 2.1-2, 2.1-3, and 2.1-4).2.1.1POPULATION The total 1990 population within 10 miles of th e station was estimated to be 120,443. This population is expected to increase to about 129,846 people by the year 2000 and to a total of approximately 142,277 people by the year 2030 (New York State Department of Economic Development, 1989 (Reference 2.1-1); State of Connecticut Office of Po licy and Management, 1991 (Reference 2.1-2); US Depart ment of Commerce, Bureau of the Census, 1990 Census of Population (Reference 2-1-3)). The 10 mile ar ea includes portions, or all of, New London and Middlesex Counties in Connecticut and a small portion on Suffolk County of Fishers Island which is part of the town of Southold, New York. Figure2.1-5 shows counties and towns within the 10 mile area. Town populations and population densities are provided in Table2.1-2.The Town of Waterford, in which Millstone Unit Number 1 is located, contained a total population of 17,930 people in 1990 at an average density of 547 people per square mile (US Department of Commerce Bureau of the Census 1991) (Ref erence 2.1-3). The population growth of Waterford was small with the 1990 total repres enting only a 0.5 percent increase over its 1980 population. Compared to towns immediately su rrounding it, with the exception of New London, Waterford had the lowest increase in popul ation between 1980 and 1990 (US Department of Commerce Bureau of the Census, 1991 (Reference 2.1-3)).Waterford's growth has been consistently slowing down over the past 30 years, as shown in Table 2.1-3. This slow growth is projected by state dem ographers to continue at a low rate through the year 2000, at which time the population is expected to reach 18,480. After that, it is projected to decrease in population. By the year 2010 (the last year of projections), the town's population is projected to be 18,080 (Connecticut Office of Policy and Manageme nt, Interim Population Projections, 1991 (Reference 2.1-2)). Population di stribution by sector for the area within 10 miles of Millstone Unit Number 1 is shown for the years 1990, 2000, 2010, 2020 and 2030 in Tables 2.1-4 through 2.1-8, that are keyed to th e population sectors identified in Figure 2.1-6. MPS-1 DSAR2.1-3Rev. 3 Population distribution within 10 miles is based on 1990 US Ce nsus data by Census Block (Reference 2.1-3). The population wi thin a Census Block was assumed to be distributed evenly over its land area, unless USGS

7.5 minute

quadrangle maps i ndicated the population to be concentrated in only on portion of the Block. Th e proportion of each Block area in each grid sector was determined and applied to the Block total population, yielding the population in each grid sector. Population projections, by municipality, supplied by Connecticut's Office of Policy and Management provided grow th factors for projection of Projections, 1991 (Reference 2.1-2).2.1.1.1Population Distribution Within 50 Miles The area within 50 miles of Millstone Unit Number 1 includes portions, or all, of eight counties in Connecticut, four coun ties in Rhode Island and one county in New York. Figure 2.1-7 shows counties and towns within the 50 mile area. In 1990, the 50 mile area contained approximately 2,835,159 people (U.S. Department of Commerce), 1990 Census of Population and Housing (Reference 2.1-4). This populat ion is projected to increas e to about 3,223,654 by the year 2030 (Connecticut Office of Policy and Management, 1991 (Reference 2.1-2); New York State Department of Economic Devel opment, 1989 (Reference 2.1-1); Rhode Island Department of Administration, 1989 (Reference 2.1-5); US De partment of Commerce, 1990 Census of Population and Housing, 1991 (Refer ence 2.1-4)). Population distri bution by sector for the area within 50 miles of Millstone Unit Number 1 is shown for the years 1990, 2000, 2010, 2020 and 2030 in Tables 2.1-9 through 2.1-13, which are keye d to the population sectors identified in Figure 2.1-9. Population distribution and projections within the 50 mile region surrounding Millstone Unit Number 1 were calculated based on population by municipalities and were assigned to sectors based on land area allocation. Projections for th e 50 mile area were based on country-wide projections.2.1.1.2Transient Population Seasonal population increases resulting from an in flux of summer residents total approximately 10,500. However, many of the beaches and recreation faciliti es in the area are used by residents, and therefore, do not represent a ny increase in population but instead a slight shift in population. There are, however, a number of sc hools, industries, and recreati on facilities which create daily and seasonal variations in sector populations. Tables 2.1-14 th rough 2.1-16 show annular sector population variations resulting from school enro llments, industrial empl oyment, and recreation facilities (with docum ented attendance).2.1.1.3Low Population Zone

The low population zone (LPZ) surrounding Millst one Unit Number 1 encompasses an area within a radial distance of about 2.4 miles. The distance was chosen base d on the requirements of 10CFR100.11. Figure 2.1-8 shows topogr aphical features, transporta tion routes, facilities, and institutions within the LPZ. MPS-1 DSAR2.1-4Rev. 3 The LPZ contained approximately 9,846 people in 1990, with an average density of 545 people per square mile. By the year 2030, the LPZ population is projected to increased to about 11,629, or an average density of 643 people per square mile (US Department of Commerce, Bureau of the Census, 1991 (Reference 2.1-3); Connecticut Office of Policy and Management, 1991 (Reference2.1-2); US Geological Survey (Reference 2.1-6)). The LPZ population distribution for 1990 and 2030 is shown in Table 2.1-17. Ta ble 2.1-18 shows the 1991-1992 school and employment distribution within the LPZ. Both tables are ke yed to Figure 2.1-9. Daily and seasonal variations due to transient population are mini mal within the LPZ. Several beaches are located within the area; however, they are predominantly used by local residents and generally have no facilities for parking or accommodation of large groups. Three schools, Great Neck Elementary and Southwest Elementary in Waterford, and Niantic Elementary in East Lyme, are located within the LPZ. Major employment c onsists of the Camp Rell Military Reservation and Hendel Petroleum. The New London Country Club is also located within the LPZ.2.1.1.4Population Center The closest population center to Millstone Unit Number 1 (as defined by 10CFR100 to contain more than 25,000 residents) is the city of New London which contained a 1990 population of 28,540 people at an average populati on density of 5,189 people per sq uare mile (US Department of Commerce Bureau of the Census 1991). The di stance between Millstone Unit Number 1 and the city's closest corporate boundary is about 3.3 miles to the northeast, just be yond the minimum distance requirement set by 10 CFR100. The region within 50 miles of Millstone Unit Number 1 includes portions, or all, of 11 Metropolitan Statistical Area's. The populations of these areas are shown in Table 2.1-19.There were 38 population centers within 50 miles of Millstone Unit Number 1, containing 25,000 or more people in 1990. They are listed in Table 2.1-20 with the populations indicated. The population of the area wi thin 50 miles of Millstone was approximately 2,800,000 in 1990, with an average density of 361 people per square mile. This density is lower than the NRC comparison figure of 500 people per square mile (NRC Regulator y Guide 1.70, Revision 3, Reference 2.1-7). Within 30 miles of Millstone, the population density is considerably less, at an average of 189 people per square mile. By 2030, the 50 mile population is projected to increase to 3,200,000 or an average population density of about 410 people per square mile, considerable lower than the NRC comparison figure for end-ye ar plant life of 1,000 people per square mile. Within 30 miles, the average density will be 223 persons per square mil es by the year 2030. Population densities by sector for 1990 and 2030 are shown for with in 10 miles of Millstone in Tables 2.1-21 and 2.1-23 respectively, which are keyed to Figure 2.1-6, and for within 50 miles of Millstone in Tables 2.1-23 and 2.1-24, respectively, which are ke yed to Figure 2.1-7. Cumulative population densities 1990 and 2030 are shown in Tables 2.1-25 and 2.1-26, respectively. MPS-1 DSAR2.1-5Rev. 32.1.2LAND USE The area around the Millstone site contains three major industria l facilities (Dow Chemical Corporation, Pfizer Cor poration, and Electric Boat division of General Dynamics Corporation); two transportation facilities (Groton/New London) Airport and the New London Transportation Center; and four military inst allations (U.S. Navy Submarine Base, U.S. Coast Guard Academy, Camp Rell, and Stone's Ranch Military Reservation).There is also an interstate hi ghway (Interstate 95), passenger and freight railroad lines, gas distribution lines, above ground gas and oil stor age facilities and two major waterways (Long Island Sound, Thames River) in the vicinity of the Millstone site. There are no major gas transmission lines, oil transmission or distribution lines, under ground gas storage facilities, drilling or mining operations, or military firing, or bombing ranges near the site.Aircraft patterns and routes are shown of Figures 2.1-10 and 2.1-11. Figure 2.1-12 shows the road and highway system in the area of the Millstone site.2.1.2.1Description of Facilities A summary of the significant indus trial, transportation, military, and industrial related facilities, and products and materials used, is shown in Table 2.1-27 as listed below.1.Dow Chemical Corporation of Allen Point, Ledyard, Connecticut is located on the east bank of the Thames River approximately 10 miles north-northeast of the site. Dow Chemical employs approximately 115 people and produces organic compounds, such as Styron, Styrofoam, and a base product of latex paints. All materials are moved to and from the company by truck and/or railroad.2.Pfizer Corporation of Eastern Point Road, Groton, Connecticut is located on the east bank of the Thames River, approximately 4.9 miles east-northeast of the site. Pfizer Corporation employs approximately 3,000 persons and produces organic compounds and pharmaceutical materials, such as citric acid, antibiotics, synthetic medicines, vitamins and caffeine. All materials are moved to a nd from Pfizer corpor ation by truck and/or railroad.3.Electric Boat Division of General Dynamics of Eastern Point Road, Groton, Connecticut is located approximately 5 miles east-north east of the site. El ectric boat employs approximately 12,000 persons, and is a pr oducer of submarines and oceanographic equipment for commercial industry and the U.S. Navy. The nature of products produced at Electric Boat requires that they handle subs tantial amounts of nuclear material which is licensed under the Naval Reactors Division. Al l material is moved by truck, railroad, and/or barge to and from the company with th e exception of completed ships which leave under their own power. MPS-1 DSAR2.1-6Rev. 34.Groton / New London Airport, approximately 6 miles east-northeast of the site, handles regularly scheduled commer cial passenger flights. A pproximately 13 persons are employed at Groton/New London Ai rport on a full-time basis, excluding airline and car rental employees. The National Guard has an aircra ft repair facility at the airport that has approximately 140 fu ll time employees.5.The New London Transportation Center, located at City Pier, New London on the west bank of the Thames River, is approximately 4 miles northeast of the site. Approximately 20 persons are employed ther e on a full-time basis. The New London Transportation Center is a large complex in downtown New London in the City Pier area. It encompasses numerous facilities, including a train station, several ferry companies, commercial and private boat slips, an interstate bus terminal, local bus inter-changers, and commercial land transportation facilities. It serves as the prime entran ce and exit for New London for civilian and commercial travel.6.U. S. Navy Submarine Base, Groton, Connect icut is located on the east bank of the Thames River, approximately 7 miles northea st of the site. The base population includes approximately 8,500 military personnel. In addition, there are about 1,800 civilian employees at the base. The U.S. Navy Submar ine Base provides logistics as well as training and operation of the base and its ships (nuclear and non-nuclear). All materials are moved by truck, railroad, barge and / or shi p, to and from this gove rnment installation.7.The U.S. Coast Guard Academy, New London, Connecticut is located on the west bank of the Thames River, approximately 5.6 miles northeast of the site. Approximately 900 cadets attend the academy, while approximate ly 360 military and civi lian personnel are employed here.8.Camp Rell, located approximately 2 miles northwest of the si te, is a training headquarters for the Connecticut Army National Guard. It is owned and operated by the Military Department of the State of Connecticut. On a full-time ba sis, it employs 16 persons (military and civilian), including the headquarters for the Connecticut Military Academy, post Operations personnel, and 745th Signal Company. On a part-time basis, during various weekends, Camp Rell is occupi ed by varying numbers of troop units for administrative training mane uvers, billeting, and supply f unctions for the Connecticut Army National Guard. During the training maneuvers there may be from 300 to 1,200 people at the facility. Camp Rell is an ad ministrative training center for troops of the Connecticut Army National Guard. Because of the solely administrative nature of its occupancy, the camp's operation has no effect on the Station's operation.9.In addition to Camp Rell, the Military Department of the State of Connecticut also maintains a field training facility known as Stone's Ranch Mi litary Reservation, located approximately 7 miles northwest of the site. Fourteen persons are employed there full-time for two regional motor vehi cle and equipment maintenance shops. It is also occupied on a part-time basis by varying numbers of troop units for periods of field training for the Connecticut Army National Gu ard. During some weekend tr aining sessions there may be up to 500 people at the facility. MPS-1 DSAR2.1-7Rev. 3 Limited quantities of munitions and explosives are stored in underground bunkers at this facility. These materials are used in quarry operations for the Connecticut Army Corps of Engineers. No live ammunition is used at the facility. All materials are moved to and from Stone's Ranch by truck. In addition, a small paved utility landing strip is located at Stone's Ranch. While capable of handling light, fixed-wing airc raft, the strip is not routinel y used except for occasional rotary-wing operations. Because of its dist ance from the site, the limited quantity of materials stored and used, and the type of aircraft operations occurring at the facility, Stone's Ranch Military Reservation does not pose any hazard to the Millstone station.10.Hess Oil Corporation of Eastern Point Roa d, Groton, Connecticut is located on the east bank of the Thames River, approximately 5 mi les east-northeast of the site. It is located north of Pfizer Corporation, and south of General Dynamics-Elect ric Boat Division and services as a fuel storage facility. There are about 14 persons employed there on a full time basis. Hess Oil Corporation operates a fuel distribution and storage facility for home heating oil and kerosene. There are large a bove ground tanks capable of storing heating oil, residual fuel oil, and kerosene. The fuel arrives by ships or barges and is distributed by trucks.11.There is one medium-sized propane storage area in the pr oximity of the Millstone site. Hendel Petroleum Company, is located in Waterford, approximately 2.5 miles northeast of the site on Great Neck Road, and employs about 75 people. Hendel Petroleum Company operates a fuel distribution fac ility for commercial and residential use. There are 5 above ground tanks (3-30,000 gallons and 2-16,000 gall ons) which are capable of storing 126,000 gallons total of propane gas. The facility also stor es 40,000 gallons of gasoline, and 40,000 gallons of Number 2 fuel oil. The pr opane for the facility arrives by train and truck, and is distributed by truck.On the Millstone site, at the Fire Training Facility located approxima tely 2,800 feet to the north of the protected area are two 1,000 gallo n propane cylinders. The two cylinders are used to supply propane to the fire simulator.The Fire Training Facili ty was constructed in 1994 for the purpose of training fire brigade members. The Training Facility consists of six live burn "mock-ups" which replicate nuclear power plant fire hazards. Propane is used to fuel these "fireplaces." The two storage cylinders are positioned such that their ends are pointed away from the Millst one site. Both cylinders are above ground domestic storage cylinders designed per ASME Code for Pressure Vessels, Section VIII Division 1-92.12.Montville Station is a Fossil Fuel pow ered electric generating plant operated by Connecticut Light & Power Company in Montvill e, Connecticut. It is located on the west bank of the Thames River, approximate ly 9.5 miles north-northeast of the site. Approximately 67 people are em ployed there. It is capable of providing 498 MW of electric power. The fuel is stored in three large above ground tanks, capable of storing approximately 175,000 barrels of fuel each; two medium above ground tanks, capable of storing approximately 12,000 ba rrels of fuel ea ch; and two small above ground tanks, MPS-1 DSAR2.1-8Rev. 3 capable of storing approximately 250 barrels of fuel each. The fuel arrives by barge or trucks.2.1.2.2Pipelines There are no major gas transmission lines within 5 miles of the site. There are two medium pressure gas distribution lines in near proximity of the site. The nearest gas distribution line is approximately 2.9 miles from the site, located al ong Rope Ferry Road in Waterford. This 35 psi gas distribution line is a 6-inch plastic pipeli ne, buried approximately 3 feet deep. The control valve for this line is located at the intersection of Clark Lane and Boston Post Road in Waterford. The second gas distribution line, ends at and serves the shopp ing center complex, near the intersection of I-95 and Parkway North, approximately 4 miles north of the site. This 35 psi gas distribution line is an 8 inch pl astic pipeline buried approximately 3 feet deep. The control valve for this line is located at the complex where it intersects with Parkway North.There are no oil transmission or distribution lines within 5 miles of the Millstone site.2.1.2.3Waterways Ships that pass by the site in the shipping channels of Long Island Sound are of two types: general cargo freighters, usually partiall y unloaded, with drafts of 20 to 25 feet, and deep draft tankers with drafts of 35 to 38 feet. Both of these classes of ships must remain at least 2 miles offshore to prevent running aground on Bartlett Reef.No oil barges pass to the shore side of Bartlett Reef, and since there are no tank farms in Niantic Bay, no oil barges pass with 2 miles of the site.Barge traffic in the vicinity of the site has been diminishing over the past several years due to the decrease in the amount of oil used by area facilities. Barge traffic is heav iest during the winter months, and averages only 1 barge per day during these months. On the average of once a month, a barge carrying 15,000 barrels of sulfuric acid is towed past the site outsi de of Bartlett Reef. Approximately 10 ships per day traverse the Reef in the vicinity, 6 miles of the site. For these reasons, it is concluded that shipping accidents would not adversely affect Millstone 3 safety related facilities.2.1.2.4AirportsGroton / New London Airport, approximately 6 miles east-northeast of the site, handles regularly scheduled commercial passenger fl ights. It is served by U.S. Ai r Express. It has two runways: 5-23, which is 5,000 feet long; and 15-33, which is 4,000 feet long. Both runways are illuminated. There is a control tower at Groton / New L ondon, with ILS (Instrument Landing System) and VOR (Very High Frequency Omni Range) navigation aides located on the airfield. The ILS is associated with runway 5. As shown on Figure 2.1-10, the landing patterns used do not direct traffic near the Millstone site. MPS-1 DSAR2.1-9Rev. 3The largest commercial aircraft to use Grot on / New London Airport on a regularly scheduled basis are Beechcraft 1900's which carry approxima tely 19 passengers. The only jets using the airport on a regular basis are two small chartered Cessna C itation which carry 10 passengers.The largest military aircraft to use Groton/New London Airport on an occasional basis are C-130's and C-23's. Additionally, there are several military helicopters st ationed at the airport. In 1995 there were approximately 4,490 military fl ights, approximately half of which were military helicopters. Millstone Station is not in the flight pa th of these flights, and pilots are briefed to avoid the site.As shown on Figure 2.1-11, the air lane nearest th e site is V58 which is approximately 4 miles northeast of the site. Other adjacent air lanes include V16, which is approximately 6 miles northwest of the site, and V308, which is approximate ly 8 miles east of the site.The nearest high-altitude jet route, J121-581, passes approximately 9 miles southeast of the site. A second jet route, J55, passes approximately 12 miles northwest of the site.2.1.2.5Highways

The area around the Millstone site is served by in terstate, state and local roads. These are shown on Figure 2.1-12. The nearest major highway which w ould be used for freque nt transportation of hazardous materials is U.S. Interstate 95, which is located 4 miles from th e Millstone site. Other principal highways which pass near the site include U.S. Highway 1 which is located 3 miles from the site, and State Highway 156, lo cated 1.5 miles from the site.These separation distances exceed the minimum distance criteria given in Regulatory Guide 1.91, Revision 1 and provide assurance that any transportation accidents resulting in explosions or toxic gas releases of truck size shipme nts of hazardous materials would not have a significant adverse effect on the safe operation or shutdown capability of the unit.2.1.2.6Railroads The site is traversed from east to west by a Providence & Worcester (P&W)/Amtrak railroad right-of-way. The mainline trac ks are more than 2,000 feet from the Millstone Unit Number 1 Reactor Building structure.The motive force for the rail stock is both diesel and electric locomotives. Overhead electric lines power the former. These lines affect neither the site nor the overhead transmission lines leaving the site and traversing the railroad right-of-way above the tracks.The Department of Transportation and P&W/Am trak have been contacted for information concerning rail traffic on the ma inline tracks. Approximately ei ghteen scheduled passenger trips per day pass along the tracks near the Millstone site. Approximately one freight train per day passes by the site. Hazardous ma terial shipped on the track include chlorine, anhydrous ammonia, carbon dioxide, propane, ethyl alcohol, rosin, MPS-1 DSAR2.1-10Rev. 3 ammonium nitrate, and hydrochloric acid. See Table 2.1-28 for a list of hazardous materials handled over this track which are potential ly capable of produci ng significant missiles. Records of hazardous materials incidents dating back a number of years show no incident occurring between East Lyme and New London, Co nnecticut, including th e trackage near the Millstone site. See Section 2.1.3 for a more detailed evaluation of potential accidents.The railroad spur serves the Millstone Nuclear Power Station exclusively. The switch for that spur is normally set for through traffic. In order to reach any station facility, a tr ain car must also pass through a second switch, which is normally set to direct traffic past the station to a dead end near the Sound. Therefore, the possibility of unauthorized transport of hazardous materials on the spur is very remote. There are no grade crossings on or adjacent to th e site at which hazardous materials might be transported across the tracks.2.1.2.7Projections of Industrial Growth Pipelines No expansion of facilities is presently planne d in the area for oil distribution within the southeastern region of Connecticut. The gas dist ribution line along Rope Ferry Road ends at Waterford high School, approximately 2.9 miles from the Millstone site. The gas di stribution line at I-95 and Parkway North ends at, and serves the shopping comple x approximately 4 miles from the Millstone site.Waterways As previously mentioned, ship and barge traffic in the area of Millstone site has decreased over the past several years. No new ship or barge traffic is anticipated at this time in the Niantic Bay area on Long Island Sound near locat ion of the intake structures. Airports No expansion of facilities at Groton / New London Airport is proposed although some improvements to the facility, such as expansi on of the approach light s, and upgrading of the terminal and runways in planned. Southeastern Connecticut Regional Planning Agency (SCRPA) recommends that a master pl an be prepared for the air port before any major physical improvements are made. The agency has previously adopted the policy that Groton / New London Airport should remain a small feeder airport providing connection to larger airports and direct service to a limited number of cities with a 500 mile radius. MPS-1 DSAR2.1-11Rev. 32.1.3DETERMINATION OF DESIGN BASIS EVENTS The area around the Millstone site was invest igated and found to contain no explosives, chemicals, airborne pollutants, flammable or dangerous gases, nor tanks or pipelines near enough to the site to pose a danger if they were to explode or burn. A railroad right-of-way of P&W/Am trak companies transverses the site from east to west. The mainline tracks are about 0.5 miles from the Millstone Unit Number 1 Reactor building and upgrade from the plant. Traffic on the spur of the mainline track which ex tends onto the site is controlled to minimize the possibility of railroad traffic-related accidents.A spur of the P&W/Amtrak railroad serves the Millstone Nuclear Power Station exclusively. The switch for that spur is normally set for through traffic. To reach a ny station facility , the locomotive must pass through a second switch, which is normally set to direct traffic past the station to a dead end near the Sound. Therefore, the possibility of unauthorized transport of hazardous materials does not exist on the spur.Hazardous materials that are sh ipped on the track which crosses the site between New Haven and New London include chlorine, anhydrous ammonia, carbon dioxide, propane, ethyl alcohol, rosin, ammonium nitrate, and hydrochlor ic acid. Among these ma terials, only the shipment of propane (about 44 carloads per year) is in the "frequently shipped quantities of hazardous material" category as defined in Regulatory Guide 1.78. The nearest major highway which would be used for frequent trans portation of hazardous materials is U.S. Interstate 95, which is located at a distance of 4 miles from the Millstone site. This separation distance exceeds the minimum dist ance criteria given in Regulatory Guide 1.91, Revision 1; and therefore, provides assurance that any transportation accidents resulting in explosions of truck size shipments of hazardous materials will not have an adverse effect on the safe operation of the plant.Based upon the size of Groton / New London airport and the location of flight paths, the impact of an airplane on Millstone Unit Number 1 is highly unlikely.There are no major gas transmission lines within 5 miles of the s ite. The nearest low pressure gas distribution line is 2.9 miles from the site and is located near Waterford High School on Rope Ferry Road. The closest oil transmission line is approximate ly 5 miles from the site in Groton Connecticut.Because they are 5 miles or more away from the site, both the ma jor gas and oil transmission lines constitute no threat to the safe conduct of activities asso ciated with storage of irradiated fuel or decommissioning of Millstone Unit Numb er 1 or to the site in general. MPS-1 DSAR2.1-12Rev. 32.1.4EFFECTS OF DESIGN BASIS EVENTS Propane gas is heavier than air and can form a potentially explosive mixture in air. However as shown on topographic maps of the area, the rail line on which propane is shipped through the site runs through an excavation in the hill which is approximately 20 feet below the natural contour of the ground immediately north of the reactor facilities. This railroad cut would channel a heavier-than-air propane cloud in an east-west direction away from the plant. The map indicates that the remainder of the topography of the site is about the same grade as the ra il line and therefore would not cause a gravity flow of th e cloud toward the plant site.2.

1.5REFERENCES

2.1-1New York State Department of Economic Development, Interim County, MSA and Region Projections, 1980 - 2010, 1989.2.1-2Connecticut Office of Polic y Management, Interim Populat ion Projections Series 91.1, 1991.2.1-3U.S. Department of Commerce, Bureau of the Census, 1990 Census of Population, P.L. 94-171 Counts by Census Block, 1991.2.1-4U.S. Department of Commerce, Bureau of the Census, 1990 Census of Population and Housing - Connecticut, 1990 CPH-1-8, 1991.2.1-5Rhode Island Department of Administration, Projections by County, 1990 - 2020, 1989.2.1-6U.S. Geological Survey, 7.5-Minute Quadrangle Maps.2.1-7U.S. Nuclear Regulat ory Commission, Regulatory Guide 1.70, Revision 3.2.1-8AAR-RPI Number RA-01 7, 1972. Association of Ameri can Railroads and Railway Progress Institute Final Phase 01 Report on Summary of Ruptured Ta nk Cars Involved in Past Accidents, Revised July 1972. Chicago, IL.2.1-9AAR-RPI Number RA-02 18, 1972. Association of Ameri can Railroads and Railway Progress Institute Final Phase 02 Report on Accident Review, Chicago, IL.2.1-10Chemical Rubber Company, 1972. Handbook of Chemistry and Physics 44th and 53rd Editions.2.1-11ConRail 1980. Hazardous Materials Link Report between New Haven and New London, Connecticut from January 1978 through June 1979.2.1-12Giffen, C. A. et al 1980. An Assessment of the Risk of Transporting Propane by Truck and Train. Report prepared for the U.S. Department of Energy by Pacific Northwest Laboratory, Battelle Memorial Institute. MPS-1 DSAR2.1-13Rev. 32.1-13Lotti, R. C.; Krotuik W. J.; and DeBoisblanc, D. R. 1973. Report of Topical Meeting on Water Reactor Safety. USAEC. Washington, D.C. Hazards to Nuclear Plants from a Near Site Gaseous Explosions. Paper, March 26-28, 1973.2.1-14Kaiser, G. D. and Griffiths, R. F. 1982. The Accidental Release of Anhydrous Ammonia: A Systematic Study of the Factors Influenc ing Cloud Density and Di spersion, Journal of the Air Pollution Control Association, Vol. 32, Number 1.2.1-15NASA Report 3023, 1978. Workbook for Estimating the Effects of Accidental Explosions in Propellant Ground Handling and Transport Systems.2.1-16NTSB-RAR-72-6, 1971.National Transportation Safety Board Railroad Accident Report for Houston, TX.2.1-17NTSB-RAR-1, 1972.National Transportation Safety Board Accident Report for East St. Louis, MO.2.1-18NTSB-RAR-75-7, 1974. National Transportation Safety Board Railroad Accident Report for Houston, TX.2.1-19NTSB-RAR-79-11, 1979. National Transportation Safety Board Railroad Accident Report for Crestview, FL.2.1-20NTSB-RAR-81-1, 1980. National Transportation Safety Board Railroad Accident Report for Muldraugh, KY.2.1-21NUREG-0800, 1981. Standard Review Plan: Ev aluation of Potential Accidents (Section 2.23).2.1-22Perry & Chilton 1973. Chemical Engineer s Handbook, 5th Edition McGraw-Hill, Inc.2.1-23Personal Communication between S.N. Ba jpai and Robert Folden, Federal Railroad Administration, Office of Safety, February 17, 1982.2.1-24Deleted.2.1-25Research and Special Programs Administration, U.S. Department of Transportation, Washington, D.C. 1981. Computer Printout of "Incidents Involving Deaths, Injuries, Damages Greater than $50,0000 or Evacuati ons." Run Dated March 26, 1981., Covering Period December 22, 1970 to September 5, 1980.2.1-26Research and Special Programs Administration, U.S. Department of Transportation, Washington, D.C. 1981. Computer Printout of "I ncidents Involving Fire and Explosions by ConRail." Run dated 4/15/81 Coveri ng Period June 6, 1973 through November 1, 1980. MPS-1 DSAR2.1-14Rev. 32.1-27Rhoads, R.E. et al 1978. An Assessment of Risk of Transporting Gasoline by Truck PNL-2133. Pacific Northwest Laboratory (Battelle Memorial Institute), Richland, Washington.2.1-28Siewert, R.D. 1972. Evacuation Areas for Transportation Accident s Involving Propellant Tank Pressure Bursts. NASA Technical Memorandum X68277.2.1-29Tilton, B.E. and Bruce, K.M. 1980. Review of Criteria for Vapor Phase Hydrocarbons, Environmental Criteria and Assessment Office. U.S. EPA-600/8-80 p 6-150.2.1-30U.S. Department of Tran sportation. Incidents Involving LPG and Ammonia, Computer Runs Prepared for Stone & Webster, 1981.2.1-31Massachusetts Institute fo r Social and Economic Researc h, Revised Projections of the Population of Massachusetts Cities and Towns to the year 2000, 1991. 2.1-32US Department of Commerce, Bureau of the Census, St ate and Metropolitan Area Book 1991, a Statistical Abstract Supplement, 1991.2.1-33US Department of Comm erce, Bureau of the Census, 1990 P.L. 94-171 Counts by Municipality - New York, 1991.2.1-34US Department of Commerce, Bureau of the Census, 1990 Census P.L. 94-171 Counts by Municipality - Rhode Island, 1991.2.1-35US Department of Commerce, Bureau of the Census, Number of Inhabitants: Connecticut, PC(1)-A8, 1971; PC80-1-A8, 1981. MPS-1 DSARPage 1 of 1Rev. 2TABLE 2.1-1THIS TABLE HAS BEE N INTENTIONALLY DELETED MPS-1 DSARPage 1 of 1Rev. 2NOTES: Based on 1990 US Census of Population and Housing. Includes total 1990 population of all m unicipalities totally or partiall y within 10 miles of the site.TABLE 2.1-2 1990 POPULATION AND POPULATI ON DENSITIES - CITIES AND TOWNS WITHIN 10 MILES OF MILLSTONE Municipality 1990 Population Total 1990 Population Density (People/Square Mile) 1980-1990 Change (%)East Lyme15,34045110.6

Groton (including City)45,1441,4429.9Ledyard14,9133918.6Lyme1,949617.0Montville16,6733971.3New London28,5405,189-1.0 Old Lyme6,5352836.1Old Saybrook9,5526372.9Waterford17,9305470.5 Southold, New York (Fishers Island)19,8363943.5 MPS-1 DSAR Page 1 of 1 Rev. 2SOURCES: 1980 Census of Population, Number of I nhabitants, Connectic ut, PC80-1-A8, 12/81. 1970 Census of Population, Number of I nhabitants, Connec ticut, PC10-A8, 4/71. 1980 Final Population and Housing Counts, Connecticut, PHC80-V-8, 3/81. 1990 Census of Population and H ousing, Connecticut, CPH-1-8, 7/91.TABLE 2.1-3 POPULATION GROWTH 1960 - 1990 MunicipalityTotal Population% Change19601970198019901960-1970 1970-1980 1980-1990 East Lyme6,78211,39913,87015,34068.121.710.6Groton29,93738,52341,06245,14428.76.69.9Ledyard5,39514,55813,73514,913169.8-5.78.6 Lyme1,1831,4841,8221,94925.422.87.0Montville7,75915,66216,45516,673101.95.11.3New London34,18231,63028,84228,540-7.5-8.8-1.0 Old Lyme3,0684,9646,1596,53561.824.16.1Old Saybrook5,2748,4689,2879,55260.69.72.9Waterford15,39117,22717,84317,93011.93.60.5 MPS-1 DSARPage 1 of 1 Rev. 2TABLE 2.1-4 POPULATION DISTRIBUTION WI THIN 10 MILES OF MILLSTONE - 1990 CENSUS Distance to PlantSector0-11-22-33-44-55-66-77-88-99-10TotalN167228667841162135422095361,7175,721NNE133591,1461,9781,8611,6222,2422,2422,1923,14216,221 NE1654558393,88810,5847,7528,1648,1299111,96142,646ENE224552924,9639717,1863,7483,7481,0082,66224,354E06364131,8041935520631,4349045,999 ESE01433600000115214508SE00000000000SSE00000000000 S00000000000SSW0000000000SW0014000000014 WSW0048991863124721580741,682W01781,0611,0144407634765628814085,782WNW04761,1651,9463462392111,6545094-176,981 NW06348731,1921,140644599101209815,473NNW1483148925226469182214294563144860Total3544,3728,08618,20016,38320,20116,09816,5948,25111,894120,443 MPS-1 DSARPage 1 of 1 Rev. 2.1TABLE 2.1-5POPULATION DISTRIBUTION WITHIN 10 MILES OF MILLSTONE 2000 PROJECTED Distance to PlantSector0-11-22-33-44-55-66-77-88-99-10TotalN187789328451262305822255781,8526,166NNE143871,2342,1312,0061,7491,7962,4152,3663,38917,487 NE1794899054,19111,4417,3598,8028,7659832,11546,203ENE244923145,3521,0457,7464,0413,2851,0872,87026,256E06854441,9442085970681,5469756,467 ESE01543900000125233551SE00000000000SSE00000000000 S00000000000SSW00000000000SW0014000000014 WSW0052898923365091690781,810W01921,1441,0934738215136069504366,228WNW05141,2552,1183732582271,7835484487,524 NW06849401,2851,229695646108226885,901NNW1583049615646969902384624913395,239Total3934,7158,71019,62117,66321,78117,35417,8868,90012,823129,846 MPS-1 DSARPage 1 of 1 Rev. 2TABLE 2.1-6POPULATION DISTRIBUTION WITHIN 10 MILES OF MILLSTONE 2010 PROJECTED Distance to PlantSector0-11-22-33-44-55-66-77-88-99-10TotalN188039618711292376002305951,9086,352 NNE143991,2722,1972,0681,8041,8532,4922,4373,49518,301NE1845049304,32111,7678,6179,0749,0361,0132,18047,626ENE255073245,5181,0787,988 4,1663,3871,1192,96027,072E07074582,005215616 0701,5931,0056,669ESE015941000 00138255593SE00000000000SSE00000000000S00000000000 SSW00000000000SW0015000000015WSW0054102953465251750791,867 W01981,1791,1264404888475306254436,417 WNW05291,2942,1843852662341,8385664617,757NW07059691,3251,267716666111232906,081NNW1633509925827181,0212454765063505,403 Total4044,8618,98020,23118,21022,45817,89318,4409,18013,226133,883 MPS-1 DSARPage 1 of 1 Rev. 2TABLE 2.1-7 POPULATION DISTRIBUTION WITHIN 10 MILES OF MILLSTONE 2020 PROJECTED Distance to PlantSector0-11-22-33-44-55-66-77-88-99-10TotalN198289908991332436202366131,9686,549NNE144111,3102,2642,1321,8601,9092,5692,5133,60218,584 NE1885199604,45512,1348,8859,3559,3181,0442,24749,105ENE255233335,6891,2208,236 4,2963,4921,1513,05227,907E07284722,067222635 0721,6421,0366,874ESE016241000 00144268615SE00000000000SSE00000000000 S00000000000SSW00000000000SW0015000000015 WSW00562105983565411800801,922W02051,2161,1615048745466441,0114506,611WNW05441,2262,2523982742421,8955834768,000 NW07279981,3651,308738687114239936,269NNW1683611,0236007381,0532534915233625,572Total4145,0089,25620,85718,77723,15418,44919,0119,46313,634138,023 MPS-1 DSARPage 1 of 1 Rev. 2TABLE 2.1-8 POPULATION DISTRIBUTION WITHIN 10 MILES OF MILLSTONE 2030 PROJECTED Distance to PlantSector0-11-22-33-44-55-66-77-88-99-10TotalN198551,0219271362506382426312,0276,746NNE144251,3512,3342,1961,9161,9682,6502,5903,71219,156 NE1935359904,59212,5109,1609,6449,6061,0752,31550,620ENE265393435,8661,1458,492 4,4283,5981,1883,14728,772E07514872,132229655 0731,6921,0687,087ESE016743000 00151281642SE00000000000SSE00000000000 S00000000000SSW00000000000SW0015000000015 WSW005801081013665581850811,979W02121,2541,1975209015616631,0434586,809WNW05601,3772,3234092812491,9566024908,247 NW07481,0291,4071,349761708116246956,459NNW1743711,0556187611,0852615075393745,745Total4265,1639,54521,50419,35623,86719,01519,5969,75714,048142,277 MPS-1 DSARPage 1 of 1 Rev. 2TABLE 2.1-9 POPULATION DISTRIBUTION WITHIN 50 MILES OF MILLSTONE - 1990 CENSUS Sector Distance to PlantTotal0-1010-2020-3030-4040-50N5,72122,28326,35732,61018,658105,629 NNE16,22134,82423,73027,46535,598137,838NE42,8489,44411,33429,987199,334292,947ENE24,35423,91416,49843,00199,721207,488 E5,99910,7127,99210,920035,623ESE5080083601,344SE0080700807 SSE002,420002,420S01,61413,5410015,155SSW02,44312,56914,8074,49834,317 SW1493822,0428,252143,933175,179WSW1,6822,4710020,38924,542W5,78227,95634,384184,723267,465520,310 WNW6,98112,47427,895148,259259,824455,433NW5,4736,21531,331191,767365,578600,364NNW4,8608,80917,850115,42478,820225,762 Total120,443164,097248,750808,0511,493,8182,835,159 MPS-1 DSARPage 1 of 1Rev. 2TABLE 2.1-10 POPULATION DISTRIBUTION WITHIN 50 MILES OF MILLSTONE - 2000 PROJECTED Sector Distance to PlantTotal0-1010-2020-3030-4040-50N6,16624,02828,70735,40420,273114,578NNE17,48737,55125,72129,92638,135148,820 NE46,20310,18312,19631,611206,940307,133ENE26,25625,74417,63345,998105,848221,509E6,46711,4978,55311,687038,204 ESE5510089501,446SE0087800878SSE002,635002,635 S01,75914,7420016,501SSW02,66013,68816,1224,89737,367SW141,02224,0008,985156,725190,746 WSW1,8102,6410022,20126,652W6,22829,88736,343195,006281,709549,173WNW7,52413,34029,762156,623273,153480,402 NW5,9016,66033,435200,205380,339626,540NNW5,2399,49219,194121,62083,732239,277Total129,846176,464267,517854,0821,573,9523,001,861 MPS-1 DSARPage 1 of 1Rev. 2TABLE 2.1-11POPULATION DISTRIBUTION WITHIN 50 MILES OF MILLSTONE - 2010 PROJECTED Sector Distance to PlantTotal0-1010-2020-3030-4040-50N6,35224,773300,05636,78521,101119,067NNE18,03138,71626,73031,42139,720154,618 NE47,62610,49912,62632,221210,368313,340ENE27,07226,65218,53048,258109,494230,006E6.66911,9868,98112,2720 39,908 ESE5930094001,533SE0092000920SSE002,761002,761 S01,84715,4450017,292SSW02,78814,34416,8965,13239,160SW151,07325,1519,416164,248199,903 WSW1,8672,6890023,26727,823W6,41730,42637,096199,100286,889559,928WNW7,75713,59030,311159,776278,156489,590 NW6,0816,80734,052202,762384,902634,604NNW5,4039,77819,778123,96485,735244,658Total133,883181,624276,781873,8111,609,0123,075,111 MPS-1 DSARPage 1 of 1Rev. 2TABLE 2.1-12POPULATION DISTRIBUTION WITHIN 50 MILES OF MILLSTONE - 2020 PROJECTED Sector Distance to PlantTotal0-1010-2020-3030-4040-50N6,54924,54131,47038,21921,963123,742NNE18,58439,91627,78432,98941,349160,622 NE49,10510,82513,05132,748213,221318,950ENE27,90727,55719,33650,343112,285234,428E6,87412,4529,37612,811041,513 ESE6150098101,596SE0096500965SSE002,894002,894 S01,93916,1840018,123SSW02,92215,03317,7075,37941,041SW151,12726,3559,869172,131209,497 WSW1,9222,7370024,38329,042W6,61130,97437,863203,283292,190570,921WNW8,00013,84430,871162,992283,254498,961 NW6,2696,95737,678205,354389,518642,776NNW5,57210,07020,382126,36987,794250,187Total138,023186,861286,242893,6651,643,4673,148,258 MPS-1 DSARPage 1 of 1Rev. 2TABLE 2.1-13 POPULATION DISTRIBUTION WITHIN 50 MILES OF MILLSTONE - 2030 PROJECTED Sector Distance to PlantTotal0-1010-2020-3030-4040-50N6,74626,33232,95339,71622,860128,607NNE19,15641,15528,87934,63743,058166,885 NE50,62011,15913,49433,286219,112324,671ENE28,77228,49520,17652,519115,158245,120E7,08712,9379,78913,375043,188 ESE642001,02401,666SE001,011001,011SSE003,033003,033 S02,03616,9570018,993SSW03,06215,75518,5585,63743,012SW151,18327,61910,342180,394219,553 WSW1,9792,7870025,55430,320W6,80931,53238,647207,551297,607582,146WNW8,24714,10231,441166,276288,449508,515 NW6,4597,11035,317207,981394,192651,059NNW5,74510,37321,003128,83589,919255,875Total142,277192,263296,074914,1001,678,9403,223,654 MPS-1 DSARPage 1 of 1 Rev. 2 Note: Includes student enrollment only. Sources: Connecticut Department of Education listing of schools: Telephone survey conducted in March 1992.TABLE 2.1-14 TRANSIENT POPULATION WITHIN 10 MILES OF MILLSTONE 1991-1992 SCHOOL ENROLLMENT Sector Distance to PlantTotal0-11-22-33-44-55-66-77-88-99-10N031000000740413797NNE0003748972,073174004443,962 NE006362106971,3521,542534004,971 ENE0002,501088801,0431,6092666,307 E01810001,3300018301,805 ESE000000000068 SE00000000000 SSE00000000000 S00000000000 SSW00000000000 SW00000000000 WSW00000000000 W000000263086401 ,127WNW003450000000345 NW000843000000843 NNW0002981,250000001 ,548TOTAL06029814,2262,8445,6431,9791,6512,6561,19121,773 MPS-1 DSARPage 1 of 1 Rev. 2 Note: Firms with 50 employees or more. Excludes plant employee population. Sources: Telephone suvey conducted in March 1992.TABLE 2.1-15 TRANSIENT POPULATION WITHIN 10 MILES OF MILLSTONE - EMPLOYMENT Sector Distance to PlantTotal0-11-22-33-44-55-66-77-88-99-10N00030000000200500NNE0000003753751092771 ,134NE00375808310375375002,036 ENE00008,8005,50082000015,120 E00000000000 ESE000000000068 SE00000000000 SSE00000000000 S00000000000 SSW000000002560256 SW00000000000 WSW00000000000 W00000000000 WNW00000012512500250 NW050008430012512500750 NNW00000000000 TOTAL05003753809,6315,5001,8201,00036347720,046 MPS-1 DSARPage 1 of 1Rev. 2 Notes:* Daily summer attendance based on 90% of yearly attendance from April through September.

• • Includes campers from April 15 to September 15.

Source:State of Connecticut DEP - Office of Parks and Forests, 1990 Park Attendance.TABLE 2.1-16 POPULATION DISTRIBUTION WITHIN 50 MILES OF MILLSTONE - 2030 PROJECTEDFACILITYLOCATIONTOTAL ANNUAL ATTENDANCESUMMER DAILY ATTENDANCEState Parks:Bluff PointENE/E 6-897,641490

• Fort GriswoldENE 5-658,965200 *Haley FarmENE/E 7-911,67560 *Harkness MemorialE 2-3157,962790
• Rocky NeckW 3-5412,4952,360 **State Forests:NehanticWNW/NNW 7-1081,146400
• MPS-1 DSARPage 1 of 1Rev. 2 Sources: 1990 Census of Population and HousingConnecticut Office of Policy a nd Management, Interim Populati on Projections Series 91.1, 4/91TABLE 2.1-17LOW POPULATION ZO NE PERMANENT POPULATION DISTRIBUTIONSDIRECTION1990 CENSUS2030 PROJECTEDN1,2981,536NNE9031,065NE1,1441,351 ENE768909E760899ESE179212 SE00SSE00S00 SSW00SW33WSW429506 W1,0251,211WNW1,0461,233NW1,1671,377 NNW1,1241,327TOTAL LPZ9,84611,629 MPS-1 DSARPage 1 of 1Rev. 2 Notes: 1991-1992 Student Enrollment Firms with 50 employees or more.

Sources: Telephone survey conducted in March 1992; Connecticut Department of Education school listing.TABLE 2.1-18 LOW POPULATION ZONE SCHOOL ENROLLMENT AND EMPLOYMENT DIRECTIONSCHOOLEMPLOYMENTN3100NNE00NE075ENE00E2920ESE00SE00SSE00S00SSW00SW00WSW00W00WNW3450NW0500NNW00TOTAL LPZ9470 MPS-1 DSARPage 1 of 1Rev. 2 Notes: PMSA - Primary Metropolitan Statistical Area.MSA - Metropolitan Statistical Area.Total population of metropolitan ar eas completely or only partiall y within 50 mile s of the site.TABLE 2.1-19METROPOLITAN AREAS WI THIN 50 MILES OF MILLSTONE 1990 CENSUS POPULATIONAREA1990 POPULATION Bridgeport - Milford, CT PMSA443,722Bristol, CT PMSA79,488Fall River, MA-RI PMSA157,272 Hartford, CT PMSA767,899New Haven - Meriden, CT MSA530,240Nassau - Suffolk, NY PMSA2,609,212 New Britain, CT PMSA148,188New London - Norwich, CT-RI MSA266,819Providence, RI PMSA654,869 Waterbury, CT MAS221,629Middletown, CT PMSA90,320 MPS-1 DSARPage 1 of 2Rev. 2TABLE 2.1-20POPULATION CENTERS WITHIN 50 MILES OF MILLSTONESTATEMUNICIPALITY1990 POPULATIONConnecticutBranford27,603Bristol60,640Cheshire25,684East Hartford50,452East Haven26,144 Enfield45,532Glastonbury27,901Groton45,144 Hamden52,434Hartford139,739Manchester51,618 Meriden59,479Middletown42,762Milford49,938 Naugatuck30,625New Britain75,491New Haven 130,474 New London28,540Newington29,208Norwich37,371 Shelton35,418Southington38,518Stratford49,389 Vernon29,841Wallingford40,822Waterbury108,961West Hartford60,110West Haven54,021 Wethersfield25,651 MPS-1 DSARPage 2 of 2Rev. 2 Notes: Municipalities with 25,000 people or more. Municipalities completely or only partially within 50 miles. Source: 1990 U.S. Census of Population and Housing.Windsor27,817Rhode IslandCoventry31,083Cranston76,060Johnston26,542Newport28,227 Warwick85,427West Warwick29,268New YorkBrookhaven407,779Southampton44,976TABLE 2.1-20POPULATION CENTERS WITHIN 50 MILES OF MILLSTONESTATEMUNICIPALITY1990 POPULATION MPS-1 DSARPage 1 of 1 Rev. 2 Source: 1990 Census of PopulationTABLE 2.1-21POPULATION DENSITY WITHIN 10 MILE S OF MILLSTONE 1990 (PEOPLE PER SQUARE MILE)Sector0-11-22-33-44-55-66-77-88-99-10AverageN821,226883571669921271161460292NNE666101,1681,4401,054751653762657843827 NE8427728552,8305,9933,5913,2002,7612735262,183 ENE1127722983,6125503,3281,4691,0353027141,241 E01,0804211,313109256021430242306 ESE02433700000345726 SE00000000000 SSE00000000000 S00000000000 SSW00000000000 SW001400000001 WSW0049866491451855402086 W03021,082738249353186191264109295 WNW08081,1181,42919611183562153112356 NW01,076890868646298235346322279 NNW7555339093803664258714613784248 Average116464515828580585394352155199384 MPS-1 DSARPage 1 of 1 Rev. 2 Source: CT Office of Policy and Management, Inte rim Population Projections Series 91.1, 4/91.TABLE 2.1-22POPULATION DENSITY WITHIN 10 MILE S OF MILLSTONE 2030 (PEOPLE PER SQUARE MILE) Sector0-11-22-33-44-55-66-77-88-99-10AverageN971,4521,0416757711625082189544344NNE717221,3771,7001,243887771900776995976 NE9859081,0093,3457,0844,2433,7803,2633226212,579 ENE1339153504,2726483,9331,7361,2223568441,466 E01,2754961,553130303025507286361 ESE02844400000457533 SE00000000000 SSE00000000000 S00000000000 SSW00000000000 SW001500000001 WSW00591795717021963022101 W03601,278872294417220225313123347 WNW09511,4041,69223213098664180131420 NW01,2701,0491,025764352278397425329 NNW8886301,075450431503102172162100293 Average136548608979685691466416183235453 MPS-1 DSARPage 1 of 1 Rev. 2 Source: 1990 Census of Population and Housing.TABLE 2.1-23POPULATION DENSITY WITHIN 50 MILE S OF MILLSTONE 1990 (PEOPLE PER SQUARE MILE)Sector0-1010-2020-3030-4040-50AverageN292378269237106215NNE827591242200202281 NE2,1831601162181,129597 ENE1,241406168313564423 E3061828179073 ESE2600603 SE008002 SSE0025005 S0271380031 SSW0411281082570 SW11622560815357 WSW86420011550 W2954753501,3451,5141,061 WNW3562122841,0791,471928 NW2791063191,3962,0701,224 NNW248150182840446460 Average384174158368528361 MPS-1 DSARPage 1 of 1 Rev. 2 Source: CT Office of Management, interim Po pulation projections, Series 91.1, 4/91.TABLE 2.1-24POPULATION DENSITY WITHIN 50 MILE S OF MILLSTONE 2030 (PEOPLE PER SQUARE MILE)Sector0-1010-2020-3030-4040-50AverageN344447336289129262NNE976699294252244340NE2,5791901382421,224662ENE1,466484206382652499E36122010097088ESE3300703SE0010002SSE0031006S0351730039SSW0521611353288SW12081751,021447WSW101470014562W3475363941,5111,6851,187WNW4202403201,2101,6331,036NW3291213601,5142,2321,327NNW293176214938509522Average453204189416594410 MPS-1 DSARPage 1 of 1 Rev. 2 Source: 1990 Census of Population and Housing.TABLE 2.1-25CUMULATIVE POPULATION DENSITY WITHIN 50 MILES OF MILLSTONE 1990 (PEOPLE PER SQUARE MILE)Sector0-1010-2020-3030-4040-50AverageN292378269237106215NNE827591242200202281 NE2,1831601162181,129597 ENE1,241406168313564423 E3061828179073 ESE2600603 SE008002 SSE0025005 S0271380031 SSW0411281082570 SW11622560815357 WSW86420011550 W2954753501,3451,5141,061 WNW3562122841,0791,471928 NW2791063191,3962,0701,224 NNW248150182840446460 Average384174158368528361 MPS-1 DSARPage 1 of 1 Rev. 2TABLE 2.1-26CUMULATIVE POPULATION DENSITY WITHIN 50 MILES OF MILLSTONE 2030 (PEOPLE PER SQUARE MILE)Sector0-100-200-300-400-50N344421374337262NNE976768505394340 NE2,579787426346662 ENE1,466730438414499 E36125516913888 ESE338453 SE00632 SSE0017106 S0261086039 SSW03910711988 SW115163125447 WSW10161271562 W3474884369061,187 WNW4202853057011,036 NW3291732778181,327 NNW293205210529522 Average453 226 223 307410 MPS-1 DSARPage 1 of 1Rev. 2TABLE 2.1-27DESCRIPTION OF FACILITIESFacilityLocationApprox. No. Persons Employed or StationedApproximate Distance From Site Miles Sector Industrial

1. Dow Chemical CorpLedyard11510+NNE2. Pfizer CorporationGroton3,0004.9ENE3. Electric Boat (Division of General Dynamics)Groton12,0005ENETransportation

4. Groton/New London Airport (Trumbull)Groton1536ENE

5. New London Transportation CenterNew London204NE Military 6. U.S. Navy Submarine BaseGroton10,3007NE

7. U.S. Cost Guard AcademyNew London1,2605.6NE8. Camp RowlandEast Lyme162NW9. Stone's Ranch Military ReservationEast Lyme147NW Industrial Related Facilities 10.Hess Oil CorporationGroton145ENE11. Hendel Petroleum Co.Waterford752.5NE12. Montville Station Electric Generation PlantMontville6710NNE MPS-1 DSARPage 1 of 1Rev. 2TABLE 2.1-28LIST OF HAZARDOUS MATE RIALS POTENTIALLY CAPABLE OF PRODUCING SIGNIFICANT MISSILESHazardous MaterialAvg. No. of Cars per Train Containing Hazardous MaterialsApprox. No. of Cars per Year1. Propane2.2044

2. Anhydrous Ammonia0.2665Total2.46649 MPS-1 DSAR2.2-1Rev. 22.2METEOROLOGY Information regarding meteorology is presented in Section 2.3 of the Millstone Unit 3 Final Safety Analysis Report (Reference 2.2-1). With the exceptions given below, that information is incorporated herein by reference.2.2.1REGIONAL CLIMATOLOGY (See Section 2.3.1 of the Millstone 3 Final Sa fety Analysis Report of Reference 2.2-1).2.2.2LOCAL METEOROLOGY (See Section 2.3.1 of the Mills tone 3 Final Safety Analysis Report of Reference 2.2-1).2.2.2.1Potential Influence of the Plant a nd Its Facilities on Local MeteorologyMillstone Unit Number 1 used a once-through cooling water system, discharging its cooling water into an existing quarry into which Units 2 and 3 also discharge and thence into Long Island Sound. Thin wisps of steam fog occasionally form over the quarry and less frequently over the discharge plume during the winter months, depending on tidal conditions and temperature differences between air and water.

This fog dissipates rapidly as it moves away from the warm water area. Because the maximum discharge plume (defined by the 1.5 °F isotherm of temperature differential when all th ree Millstone units were at full power) is approxi mately an ellipse of 1500 meter by 800 meters, the extent of the steam fog is negligible. With the permanent shutdown of Millstone Unit Number 1, this maximum discharge plume size is further reduced.2.2.2.2Local Meteorological Conditions for Design and Operating Bases.2.2.2.2.1Design Basis Tornado The specifications for the Millstone Un it Number 1 design basis tornado are:Rotational velocity 300 mphTranslational velocity 60 mphTotal pressure drop 2.25 psiRate of pressure drop 1.2 psi/sec2.2.3ON SITE METEOROLOGICAL MEASUREMENTS PROGRAM The Millstone Site is served by a common meteorological tower, located south of Millstone Unit Number 1. The meteorological tower is capable of measuring wind sp eed, direction, and air temperature at various heights. Fo r details regarding the capability of the On Site Meteorological Measurements program, see Sect ion 2.3.3 of the Millstone 3 Fi nal Safety Analyses Report Reference 2.2-1, with the exception that Mill stone Unit Number 1 no longer has the data recording systems and data recording capabilit y to display parameters transmitted by modem/ MPS-1 DSAR2.2-2Rev. 2 phone line from the instrument shack at the base of the tower to th e control room area for display on the plant process computer described in Section 2.3.3.3. 2.2.4SHORT TERM (ACCIDENT) DIFFUSION ESTIMATES 2.2.4.1Objective Accidents could result in short-term releases of radioactivity from several possible venting points. Atmospheric diffusion factors (/Q) based on site meteorological data are calculated at the exclusion area boundary (EAB) and low population zone (LPZ) for each downwind sector for each release point. The diffusion factors are calculated for different release time periods depending on the length of the release. These diff usion factors are used in the calculation of radiological consequences of the releases.2.2.4.2Calculations 2.2.4.2.1Venting Point and Receptor Locations The LPZ is taken to be 3860 m in al l sectors from any release point. 2.2.4.2.2Models

Accident /Q's were calculated using the ba sic methods of Regulatory Guide 1.145. /Q values for the Millstone Unit 2 Control Room due to ground level releases were calculated using the methods of Murphy and Ca mpe. (Reference 2.2-2).2.2.4.3Results

The calculated /Q's used in design basis accident (DBA) radiological conse quence calculations are presented with the list of assumptions in Chapter 5.2.2.5LONG-TERM (ROUTINE) DIFFUSION ESTIMATES2.2.5.1ObjectiveLow levels of radioactivity are routinely released on a continuous basis from the Unit Number 1 BOP exhaust point and the SFPI ventilation exhaust point. Atmospheric diffusion factors (/Q) based on site meteorological data are calculated for various downwind r eceptor locations of interest. The meteorological data is used to calculate the dose consequences to the public from routine airborne effluents. The calculated doses are submitted periodically to the Nuclear Regulatory Commission (NRC). MPS-1 DSAR2.2-3Rev. 22.2.5.2Calculations2.2.5.2.1Venting Point and Receptor Locations Routine releases of radioactivity in gaseous ef fluents are vented from the Unit Number 1 BOP exhaust point and the SFPI ventilation exhaust point. 2.2.5.2.2Database Calculations are performed on a periodic basis using the actu al meteorology for this period.2.2.5.2.3Models /Q values are ground level dispersion factors, and releases are modele d using a conventional Gaussian plume model.2.

2.6REFERENCES

2.2-1Millstone Unit 3, Final Safety Analysis Report, Section 2.3-Meterorology.2.2-2Murphy, K. G., and Campe, K. M. Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19, 13th AEC Air Cleaning Conference, 1973. MPS-1 DSAR2.3-1Rev. 22.3HYDROLOGIC ENGINEERING Information regarding hydrologic engineering is presented in Section 2.4 of the Millstone 3 Final Safety Analysis Report (Reference 2.3-1). With the exceptions given below, that information is incorporated herein by reference.2.3.1HYDROLOGIC DESCRIPTION (See Section 2.4.1 of the Mills tone 3 Final Safety Analys is Report, Reference 2.3-1).2.3.2SITE AND FACILITIES Millstone Point is located on the north shore of Long Island Sound. To the west of the site is Niantic Bay and to the east is Jordon Cove. Figure 2.3-1 shows the general topography of the Millstone area. The site grade elevation for Millstone Unit Number 1 varies from 14 feet to above 15 feet mean sea level (MSL). Section 2.3.3.2 discusses the probable maximum hurricane used to calculate maximum water levels.2.3.3FLOODS This section reviews the flo od history in the vicinity of Millstone Point, flood design considerations, and the effects of local intense precipitation. 2.3.3.1Flood History

Flooding near the site has hist orically been caused by hurrica nes. The maximum historical flooding was the result of a hurricane on September 21, 1938, which produced a flood level of 9.7 feet MSL at New London, Connecticut. The only sources of flooding that could affect Millstone Unit Number 1 are direct rainfall and storm surges. 2.3.3.2Flood Design Considerations

The controlling event for flooding at the Millstone site is a storm surge resulting from the occurrence of a probable maximum hurricane (PMH) (see Section 2.3.6). The maximum still water level is +18.11 feet MSL, and the associated wave run up is +22.3 feet MSL. Chapter 3 describes the flooding protective features at Millstone Unit Number 1.2.3.3.3Effect of Local Intense Precipitation

A discussion on the development of the probable maximum precipitation (PMP) for the site may be found in Section 2.3.2 of Reference 2.3-1. MPS-1 DSAR2.3-2Rev. 2 A study was performed to determine the impact of the PMP intensit y on the plant roof structures. The radwaste disposal building, intake structure, radwaste/contro l building, and southwest corner of the reactor building roofs can support the loads resulting from a PMP without cred iting the roof drains.The turbine building, reactor building, warehouse, and heating/ventilat ion area roofs credit scuppers to assure that the loads due to a PMP will remain below the roof design live loads.PMP studies show that the area east of Millstone Unit Number 1, north of the radwaste truck bay, including the semi-enclosed area just east of th e Unit 2 Control Room would have maximum ponding on the order of 15.5 to 16.2 feet. MSL. Further, these studies show that areas west of Millstone Unit Number 1 and 2, south of Millstone Unit Numb er 1, extending around the gas turbine building, to the east side of Millstone Unit Number 1 north of the radwaste truck bay would experience less ponding on the order of 14.6 to 14.9 feet. MSL. Ponding at the intake structure would be negligible since runoff would flow directly to the adjoining Niantic Bay. During a PMP scenario, in-leakage through door openings could occur once the flood depths exceed door sill elevations. Secure d external and internal doors will have a tendency to limit or control the amount of in-leakage.2.3.4PROBABLE MAXIMUM FLOOD (PMF) ON STREAMS AND RIVERS (See Section 2.4.3 of Referenc e 2.3-1, the Millstone Unit 3 Fi nal Safety Analysis Report).2.3.5POTENTIAL DAM FAILUR E, SEISMICALLY INDUCED (See Section 2.4.4 of Referenc e 2.3-1, the Millstone Unit 3 Fi nal Safety Analysis Report).2.3.6PROBABLE MAXIMUM SURGE AND SEICHE FLOODING2.3.6.1Probable Maximum Winds and Asso ciated Meteorological Parameters The meteorologic characteristics used to calculate the probable maximum storm surge at the Millstone Point site are those associated with the PMH as reported by the U.S. National Oceanic and Atmospheric Administration (NOAA) in their unpublished report HUR 7-97. HUR 7-97 described the PMH as "-a hypothetical hurricane having that combination of characteristics which will make it the most severe that can pr obably occur in the partic ular region involved. The hurricane should approach the poin t under study along a critical path and at an optimum rate of movement." Actually, nine different PMH storm patterns can be constr ucted using wind speed, storm size and forward speed parameters given in HUR 7-97 in various combinations. The storm, which would cause the maximum surge buildup at the entrance to Long Island Sound is one with a large radius to maximum wind and a slow sp eed of translation. Pert inent parameters are tabulated below: Central Pressure Index MPS-1 DSAR2.3-3Rev. 2 The minimum surface atmospheric pres sure in the eye of the hurricanes.Radius to Maximum Wind (R) at 48 nautical miles. This is the distance from the eye of the storm to the locus of maximum wind. Forward Speed (T) 15 knots. This is the rate of forward movement of the hurricane center.Maximum Wind(Vx) 115.5 mph. This is the absolute highest surface wind speed in the belt of maximum winds.Peripheral Pressure(Pn) 30.56 inches. This is the surface atmosphe ric pressure at the outer edge of the hurricane where the hurrica ne circulation ends. Although other parametric combinations give a higher wind speed, this particular combination yields the highest surge. 2.3.6.2Surge and Seiche Water LevelsAlthough frontal storms and squall lines cause tidal flooding in the Millstone Point area, by far the most severe flooding has resulted from hurricanes. For this reason, the PMH as defined in Section 2.3.6.1 was used to compute the design storm surge level at the site. The calculated total surge height or still water leve l considers the wind setup, the water level rise due to barometric pressure drop, the astronomical tide and forerunner or initial rise.The maximum still water level is +18.11 feet, and the associated wave r un up elevation is +22.3 feet MSL. 2.3.6.3Wave Action Wave characteristics are depende nt upon wind speed and duration, fetch length, and water depth. Millstone Point is sheltered from the dir ect onslaught of open ocean waves by Long Island. At the time of the peak surge, the wind is from the s outheast direction and the wave attack would be along the large axis of the point. Thus the in take structure, and the southeast portions of the Reactor and Turbine Generator Bu ildings are primarily involved. MPS-1 DSAR2.3-4Rev. 22.3.6.4Resonance (See Section 2.4.5 of the Millstone Unit 3 Final Safety Analysis Report Reference 2.3.1).2.3.6.5Protective StructuresAt the time of the peak surge, the wind is from the s outheast direction and the wave attack would be along the large axis of the point. Thus, the southeast portions of the Reactor Building would be primarily involved.2.3.6.6Probable Maximum Tsunami Flooding (See Section 2.4. 6 of the Millstone Unit 3 Final Safe ty Analysis Report Reference 2.3-1).2.3.7ICE EFFECTS

There is no available history of ice or ice jams in Niantic Bay.2.3.8COOLING WATER CANALS AND RESERVOIRS (See Section 2.4.8 of the Millstone Unit 3 Fi nal Safety Analysis Report Reference 2.3-1.) 2.3.9CHANNEL DIVERSIONS (See Section 2.4.9 of the Millstone Unit 3 Fi nal Safety Analysis Report Reference 2.3-1.) 2.3.10FLOODING PROTECTION REQUIREMENTS None.2.3.11LOW WATER CONSIDERATIONS 2.3.11.1Low Flow in Rivers and Streams Since Millstone Unit Number 1 does not depend on either rivers or streams as a source of cooling water, this section is not applicable.2.3.11.2Low Water Resulting from Surges, Seiches, or Tsunamis No effect at Millstone Unit 1. 2.3.12DISPERSION, DILUTION, AND TRAVEL TIMES OF ACCIDENTAL RELEASES OF LIQUID EFFLUENTS SURFACE WATERS.(See Section 2.4.12 of the Millstone Unit 3 Fi nal Safety Analysis Report Reference 2.3-1.) MPS-1 DSAR2.3-5Rev. 22.3.13GROUNDWATER (See Section 2.4.13 of the Millstone Unit 3 Fi nal Safety Analysis Report Reference 2.3-1.) 2.3.14TECHNICAL SPECIFICATION AND EMERGENCY OPERATION REQUIREMENTSStation Procedures address necessary precautions an d actions to take in the event of anticipated hurricane, tornado, or flood conditions.2.3.15REFERENCES2.3-1Millstone Unit 3, Final Safe ty Analysis Report, Section 2.4 - Hydrologic Engineering.2.3-2Letter from J. J. Shea to W. G. Counsil, "Millstone Nuclear Power Stat ion, Unit 1 - Safety Evaluation Report on Hydrology SEP Topics II -3.A, II-3.B, II-3.B.1, and II-3.C," dated June 30, 1982. MPS-1 DSAR2.4-1Rev. 2.12.4GEOLOGY, SEISMOLOGY, AND GEOTECHNICAL ENGINEERING2.4.1BASIC GEOLOGIC AND SEISMIC INFORMATION Information regarding the geologic and seismic qua lities of the Millstone site is presented in Section 2.5.1 of the Millstone Unit Number 3 Fi nal Safety Analysis Report (Reference 2.4-1). That information is incor porated herein by reference.2.4.2VIBRATORY GROUND MOTION

Information regarding vibratory ground motion at the Millstone site is presented in Section 2.52 of Reference 2.4-1. With the exceptions given below, that information is incorporated herein by reference.2.4.2.1Safe Fuel Storage Earthquake The design of the plant is such that spent fuel pool remain intact during a ground motion of 0.17 g.2.4.3SURFACE FAULTING2.4.3.1Geologic conditions of the Site

Section 2.5.1.2 of Reference 2.4-1 discusses the stratigraphy, structural geology, and geologic history of the site are in detail.2.4.3.2Evidence of Fault Offset The published geologic maps which include the site area do not i ndicate the presence of faulting. A discussion of faults discovered during excavation of the Millstone Unit Number 3 site can be found in Section 2.5.3.2 of Reference 2.4-1. This di scussion can be considered typical for the entire Millstone site.2.4.3.3Earthquakes Associated with Capable Faults There is no evidence of capable faults within the fi ve-mile radius of the site. The majority of the significant seismic activity has been associated with the White Mountain Plutonic Province. Some activity has been associated with the Ramapo fault system (Reference 2.4-2); however, the fault is not considered capable (Reference 2.4-3).2.4.3.4Investigation of Capable FaultsThere are no capable faults with in the site area. The faults uncovered in the excavation are discussed in Section 2.5.3.2 of Reference 2.4-1. MPS-1 DSAR2.4-2Rev. 2.12.4.3.5Correlation of Epicente rs with Capable Faults There has been no spatial correlation between earthquakes and folds in the site region. Some correlation has been suggested with the Ramapo fault in New York and New Jersey. However, the Ramapo is not considered capable (Reference 2.4-3).2.4.3.6Description of Capable Faults There are no capable faults within five miles of the site.2.4.3.7Zone Requiring Detailed Faulting Investigation Eleven incapable fault zones were uncovered dur ing excavation at the Millstone Unit Number 3 site. These faults have been mapped in detail and are discussed in Section 2.5.3.2 of Reference 2.4-1. 2.4.3.8Results of Faulting InvestigationThere is no evidence of capable faulting within the fi ve mile radius of the site. The faults at the site are related to the rifting associated with the Triassic-Jurassic Period or older, with the last activity occurring approxima tely 142 million years ago.2.4.4STABILITY OF SUBSURFACE MATERIALS AND FOUNDATIONSNo information on the stability of subsurface materials and foundations is available from the excavation for Millstone Unit Number 1. A discus sion of this subject for the Millstone Unit Number 3 excavation can be found in Reference 2.4-1. This information can be considered typical for the Millstone site.2.4.5STABILITY OF SLOPES

The stability of slopes at the Millstone site was evaluated in Reference 2.4-4, wherein it was concluded that there are no natura l or man-made slopes at the si te that could be or become unstable such as to affect safety re lated structures, systems or components.2.4.6EMBANKMENTS AND DAMS

No embankments or dams have been constructed at the Millstone site.2.

4.7REFERENCES

2.4-1Millstone Unit 3, Final Safe ty Analysis Report, Section 2.5, Geology, Seismology, and Geotechnical Engineering. MPS-1 DSAR2.4-3Rev. 2.12.4-2Aggarwal, Y.P. and Sykes, L.R. 1978. Earthqua kes, Faults, and Nuclear Power Plants in Southern New York and Northern New Jersey. Science, Vol. 200, Number 430, pages 425-429.2.4-3Nuclear Regulatory Comm ission 1977. (6 NRC 547 (1977) At omic Safety and Licensing Appeal Board Hearings on Indian Point, Units 1, 2, and 3 (Dockets Number 50-3, 50-247, and 50-285) ALAB-436.2.4-4Nuclear Regulatory Commission. Letter from J. Shea to W. G. Counsil dated June 30, 1982, "SEP Review Topic II-4, D, Stability of Slopes, Millstone Nuclear Power Station Unit 1." MPS-1 DSAR3.1-1Rev. 2.1CHAPTER 3 -FACILITY DESIGN AND OPERATION3.1DESIGN CRITERIA3.1.1CONFORMANCE WITH 10CFR50 APPE NDIX A GENERAL DESIGN CRITERIA3.1.1.1Summary Discussion The General Design Criteria (GDC) for Nuclear Power Plants as listed in AppendixA to 10CFR50 were effective May 21, 1971 a nd subsequently amended July 7, 1971. Millstone Unit Number 1, was issued a provisi onal operating license (POL) on October 7, 1970, and is not obligated to comply with the GD C (Reference 3.1-3). Therefore, Millstone Unit Number 1, is not required to seek exemptions for those areas where it does not comply with the GDC. An evaluation of the design bases of the Millstone Nuclear Unit Number 1, as compared to the GDCs, was performed in support of the application for a full term operating license (FTOL), see Reference3.1-1. It was concluded therein that Millstone Unit Number 1 satisfies and is in compliance with the intent of the GDCs. Nevertheless, it should be noted that this comparison and conclusion was not a commitment to meet all of th e current GDCs or even to meet the intent of the current GDCs. Instead, the Re ference 3.1-1 comparison determined the degree of compliance with the GDCs at that time. Also, compliance is demonstrated based upon those interpretations in effect at the time the specific licensing question, or issue, was being addressed.3.1.1.2Systematic Evaluation Program and Three Mile Island Evaluations of General Design CriteriaDuring the systematic evaluation program (SEP) initiated by the NRC in 1977, a large number of generic and plant specific safety concerns were addressed and resolved (Reference 3.1-2). Man of these SEP issues, and later issues which aros e from the Three Mile Island (TMI) accident, involved a consideration of the NRC GDC affect ed by a specific issue a nd how the plant design compared to the criteria. A compilation of this more recent evaluation of specific safety concerns and the affected GDC are listed in Table 3.1-1.3.1.2CLASSIFICATION OF STRUCTURES , SYSTEMS, AND COMPONENTS 3.1.2.1Seismic Classification

The Code of Federal Regulations requires that st ructures, systems, and components important to safety shall be designed to withstand the effects of earthquakes wi thout loss of capability to perform necessary safety functions. 10CFR10 0, Appendix A further de fines a safe shutdown earthquake (SSE) and the structures, systems and components required to remain functional, as those plant features necessary to ensure:(1)The integrity of the reactor coolant pressure boundary, MPS-1 DSAR3.1-2Rev. 2.1(2)The capability to shut down the reacto r and maintain it in a safe shutdown condition, and(3)The capability to prevent or mitigate the consequences of accidents which could result in potential off site exposures comp arable to the guideline exposures of 10CFR100.Regulatory Guide 1.29, Revision 3, describes an acceptable method for identifying and classifying those plant features that should be designed to withstand the effects of an SSE.The plant structures and equipment, including their foundations a nd supports, are divided into two structural safety categories:Seismic Class I: structures and equipment whose failur e could cause significant release of radioactivity or which are vital to the removal of decay heat.Seismic Class II

structures and equipment which are not essential to the containment of radioactivity or removal of decay heat.

Only the Reactor Building at and below elevation 108 feet 6 inches , the fuel pool liner and the spent fuel racks remain Seismic Class I in the permanently defueled condition. The Reactor Building houses, protects and supports the spent fu el pool. It supports maintenance of the fuel configuration in the fuel pool, provides prot ection from external hazards and supports maintenance of water in the fuel pool to a depth ne cessary to ensure the irradiated fuel is always immersed. The fuel storage racks are designed to assure subcriti cality in the fuel pool and are designed to withstand the anticipated eart hquake loadings as Class I structures. The Reactor Building structure above elevation 108 feet 6 inches (enclosure) is classified as seismic Class II in the permanently defuel ed plant condition. The Reactor Building above elevation 108 feet 6 inches provi des a weather enclosure for the sp ent fuel pool and supports the reactor building overhead crane. However, it has no structural function in providing support for the spent fuel pool. Since the enclosure is no l onger credited to provide secondary containment (DSAR Section 3.1.2.2) and since its failure during a seismic event could adversely affect the spent fuel pool and its contents or adjacent safety related SSCs, the seismic design of the enclosure is categorized as Seismic II/I and is further discussed in Section 3.1.6. Dismantlement of Seismically Designed Structures, Systems and Components For SSCs designated as Seismic Class I in the permanently de fueled condition, the following criteria apply prior to performing dismantlement operations:(1)Downgrading seismic classification of components shall be performed in accordance with appropriate engine ering and design procedures and processes. MPS-1 DSAR3.1-3Rev. 2.1(2)When downgrading seismic classification of an SSC, a 10CFR50.59 evaluation shall be performed if:a.the seismic classification is described in the DSAR, or b.it's failure in a seismic event could affect a Seismic Class I component described in the DSAR in such a manner as to cause an unanalyzed incident or an accident with offsite doses exceeding the doses from the design basis accident.(3)When downgrading seismic classification of an SSC, a 10CFR50.54 evaluation shall be performed if the st ructure classification is described in the Quality Assurance Program (QAP).(4)When downgrading seismic classification of an SSC, a 10CFR50.59 evaluation shall be performe d if, during a seismic event, its failure has the potential to drain the fuel pool wate r level lower than 9 feet above the active fuel.3.1.2.2Safety Related ClassificationNuclear plant SSC have traditionally been classified as "safety related" in accordance with 10CFR50.2 and in 10 CFR100, Appendix A, Secti on III, if they are relied upon to remain functional during and following design basis events to assure:*The integrity of the reactor coolant pressure boundary,*The capability to shut down the reactor a nd maintain it in a safe shutdown condition, or*The capability to prevent or mitigate the consequences of accidents which could result in potential off site exposures co mparable to the applicable guideline exposures set forth in 10CFR50.34(a)(1) or 10CFR100.11.Clearly, the first two parts of the safety relate d definition (reactor coolant pressure boundary, and capability to achieve and mainta in safe shutdown) do not apply to a permanently defueled plant, given the license restrictions of 10CFR50.82. The third part of the safety related definition (accident consequences comparable to 10CFR100 guidelines) is dependent on the results of new design basis accident analysis assumptions and results developed to address the existing defueled plant condition. SSC that are required to protect workers and the public from the consequences of design basis events may need to rema in classified as safety related.Types and consequences of potential accidents were reanalyzed and it was concluded that the only remaining accident is a fuel handling accident. This accide nt was analyzed assuming no secondary containment or sta ndby gas treatment system in operation, with a puff ground level MPS-1 DSAR3.1-4Rev. 2.1release. The resulting off site ra diological exposure was determined to be significantly less than the guideline exposures set forth in 10CFR50.34(a)(1) and 10CFR100.11. Therefore, no SSC is required to be safety related to prevent or mi tigate the consequences of the only remaining accident, except to account for assu mptions inherent in the analysis. The only damage assumed in the analysis is damage to a certain number of fuel bundles. Assumptions inherent to this conclusion are that the fuel configuration, other than the direct impact damage, did not change and that the fuel pool water remained inpl ace. This implies that there is no failure of either the fuel pool struct ure or the fuel racks. Since these are passive structural items, assumption of failure is not re quired as long as the items are safety related and designed to withstand these loads. Therefore, the fuel pool and supporti ng structure, fuel pool liner, and the fuel racks must be considered as safety related to support the assumptions made in the accident analysis. No other components, sy stems or structures meet this criterion.3.1.2.3Non-Safety Related Plant Functions Maintained in the Defueled ConditionIn addition to the Safety Relate d criterion above, other non-safety related plant func tions must be maintained in the defueled condition. The followin g criteria were used to determine which SSC were still required:Criterion 1Is the SSC associated with storag e, control or maintenance of nuclear fuel in a safe condition; or handl ing of radioactive waste? Criterion 2Is the SSC program associ ated with radiological safety?Criterion 3Is 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 radioac tive waste or radiological safety?Criterion 4Does the SSC satisfy a requi rement based on regulations governing management of nuclear fuel or radi oactive materials, including any SSC which is independently required by the License or Technical Specifications? These criteria were applied to all Millstone Unit Number 1 SSC. A positive response to any criterion, including the Safety Re lated criterion, result s in the group of those SSC which must remain functioning. 3.1.2.4SSCs Important to the Defueled ConditionAll SSC that must remain functional will be main tained in accordance with applicable Millstone Unit Number 1 procedures or quality processes. Commitments exist for augmented quality related to Fire Protection (FPQA), and Radwaste (R WQA). These requirements would apply to the appropriate portions of the SSC which meet the criteria above. MPS-1 DSAR3.1-5Rev. 2.1Non-nuclear safety standards apply to other SSC that do not fall under existing quality processes. However, to provide added assura nce of adequate reliability for non-safety related SSC that are important to safeguarding the heath and safety of the public and workers, another augmented quality classification is being developed. This classification, Important to the Defueled Condition (ITDC), implements management expectations but does not satisf y any regulatory requirement, and will apply to selected systems and com ponents which perform the following functions:*Storage, control, maintenanc e or handling of nuclear fuel*Storage, control, maintenance or handling of radioactive waste, if not already RWQA*Radiological safety

Systems and components were reviewed for thes e functions. Note that application of these functions differs from the 4 crit eria in that requirements apply only to the primary SSC and are not extended to supporting systems, equipment or structures. The intent of the ITDC augmented quality is to increase reliable operation of the system(s) primarily responsible for performing each function. Acceptable performance of the suppo rting SSC are demonstrated during routine operation and/or periodic te sting of the ITDC SSC. Additionally, certain regulatory requirements to which the licensee made a licens ing commitment may go beyond the functional scope of an SSC (e.g., Emergency Plan, Security Plan, Quality Assurance Program, etc.). These co mmitments and legal requirements were also considered in the reclassification process. Authorizations, Restrictions and Limitations on use of the SSC reclassification criteriaThe SSC reclassification criteria is used as a ba sis to change various Millstone Unit Number 1 procedures and programs, provided that the ch ange involves an SSC that is non-ITDC and, provided that plant pro cedures contain an acceptable met hod for approving the change. The following kinds of "soft" changes asso ciated with non-ITDC SSCs are allowed:*SSC classifications,

• drawings,*calculations,*procedures,
• nonconforming items a nd corrective actions,*external industry operating experience reports,*commitments, MPS-1 DSAR3.1-6Rev. 2.1*open work orders (in process at the time the decision was made to decommission the plant)*the application of 10CFR50 Appendix B criteria provided it does not represent a reduction in commitment.

Use of these criteria does not authorize: a.Activities creating new hazards or initiators not already recognized as part of the current license basis (e.g., decontamination or deco mmissioning of major components defined in 1 CFR 50.82)b.The physical removal/disassembly of existi ng SSCs, or the installation of new SSCs. However, it may provide the basis for initiating such a change.c.Changes to Technical Specification requirements.d.Changes to regulations, license conditions, rule s, and permits until such time that relief is granted by the regulating authority. However, it may provide th e basis for requesting relief from the regulations, license conditions, rules, and permits.e.Changes to commitments. Application of the commitment change pr ocess is required to change commitments.f.Changes to the QAP. However, it may provide the basis for initiating a change to the QAP.g.Changes to the Radiological Effluent Monitoring and Offsite Dose Calculation Manual (REMODCM). However, it may provide the basis for initiating a change to the REMODCM.h.Changes to the Emergency Plan. However, it may provide the basis for initiating a change to the Emergency Plan.i.Changes to the Security Plan. However, it may provide the ba sis for initiating a change to the Security Plan.j.Changes to the Fire Protection Plan. However, it may provide the basis for initiating a change to the Fire Protection Plan.k.Changes to the Radiation Protection Program. However, it may provide the basis for initiating a change to the Radiation Protection Program. Boundaries and Interfaces for ITDC SSCs SSCs identified as ITDC that require "availability" shall be maintained in a state such that the necessary functional capability is maintained. MPS-1 DSAR3.1-7Rev. 2.1 Engineered Requirements for ITDC SSCs A higher level of engineered quality is maintained for ITDC SSCs to assure that the capability exists to reliably meet performance expectations and requirements. However, ITDC SSCs are not safety related and are not required to satisfy 10CFR50 Appe ndix B requirements. Although not required by regulation, the follow ing criteria is developed and applied, as specified by engineering, to ITDC SSCs to assure continued reliability:a.Design Control: Measures will be invoked to assure applicable regulatory requirements, license basis, and design basis information is correctly translated into specifications, drawings, procedures and instructions. These measures shall include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controll ed. Design changes, including field changes will be subjected to engineered design control measures commensurate with the importance of the SSC.b.Procurement Document Control: Measures wi ll be invoked to assure that applicable regulatory requirements, design basis, and other requirements which are necessary to assure adequate quality are suitably incl uded or referenced in the documents for procurement of material, equipment, services.c.Instruction, Procedures, and Drawings: Activities affecting SSCs will be prescribed by documented instructions, procedures, or dr awings, of a type appropriate to the circumstances and will be accomplished in accordance with these instructions, procedures, and drawings. Instru ctions procedures, and drawi ngs will include appropriate quantitative or qualitative acceptance criteria for determ ining that important activities have been satisfactorily accomplished.d.Control of Purchased Material, Equipment, and Services: Measures will be invoked to assure that material, equipment, and serv ices conform to the procurement documents. These measures shall include provisions, as appropriate, for so urce evaluation and selection, objective evidence of quality furnished, inspection at the source, and examination upon delivery.e.Inspection: Inspection of activities affecting quality will be invoked and executed to verify conformance with the documented instru ctions, procedures, and drawings for accomplishing the activity.f.Handling, Storage, and Shipping: Measures will be invoked to control the handling, storage, shipping, cleaning and preservation of material and equipment in accordance with work and inspection instructions to prevent damage or deterioration.g.Test Control: Surveillance testing will be established for SSCs to ensure that the SSCs perform satisfactorily commensurate with the importance of th eir intended function. MPS-1 DSAR3.1-8Rev. 2.1h.Measuring and Test Equipmen t: Appropriate contro ls will be invoked to assure that measuring and test devices used on SSCs are pr operly controlled, calibrated and adjusted at specified periods to maintain accuracy within necessary limits.i.Corrective Action: Measures will be invoked to assure that conditi ons adverse to quality are promptly identified and corrected. In the case of signi ficant conditions adverse to quality, the measures will assure that th e cause of the condition is determined and corrective action is take n to preclude repetition.3.1.3WIND AND TORNADO LOADINGS10CFR50 (General Design Criteria 2), as implemented by Standard Review Plan (SRP) Sections 3.3.1 and 3.3.2 and Regulatory Guides (RG) 1.76 and 1.117, requires that the plant be designed to withstand the effects of natural phe nomena such as wind and tornadoes.The Millstone Unit Number 1 capability to withstand wind and tornado loadings was evaluated in the Systematic Evaluation Program (SEP) (Reference 3.1-4) as Topic III-2. Several submittals were made to the U.S. Nuclear Regulatory Commission (NRC) to address issues raised under that topic (References 3.1-6, 3.1-7, 3.1-8, and 3.1-9). In an evaluation dated November 25, 1985 (Reference3.1-5), the NRC concluded that the pr oposal will provide ade quate protection against tornado events. 3.1.4WATER LEVEL DESIGN The original design basis water level at Millstone Unit Number 1 is the probable maximum flood (PMF) level of 19.0 feet above mean sea level (MSL). In the defueled condition, flooding of Unit 1 structures is acceptable. The spent fuel is stored in the upper elevations of the Reactor Building, and as such is adequately protected from the PMF. The intake structure itself which was originally designed as seismic Class 1, is designed to with stand a water level of elevation 32.4 feet MSL. This level accounts for an assumed 13.4 feet MS L still water level and for non-breaking waves above this level as they strike the structure.3.1.5MISSILE PROTECTION

Systems and components have been examined to identify and classify potential missiles.3.1.5.1Internally Generated MissilesTwo broad categories of systems and components are reviewed to determine the potential for generating missiles; pressurized components and high speed rotating machinery. Only designs where a single failure could lead to a missile ejection were considered. It was determined there are no highly pressuri zed components or high speed rotating machines capable of generating significan t missile hazards in the pe rmanently defueled condition. Therefore, no internally generated missiles are postulated. MPS-1 DSAR3.1-9Rev. 2.13.1.5.2Missiles Generated by Natural PhenomenaThe effects of missiles generated by natural phenomena has been evaluated References 3.1-13, 3.1-14, 3.1-15, 3.1-16, and 3.1-17). On the basis of thos e evaluations, Millstone Unit Number 1 is adequately protected against such missiles inhibiting the ability to maintain safe storage of new and irradiated fuel.3.1.5.3Missiles Generated by Ev ents Near the Site The objective of this assessment (Reference 3.1-19) is to assure th at the integrity of the safety related structures, systems, and components will not be impaired and that th ey will perform their safety functions in the event of a site proximity missile.The potential for hazardous activities in the vicinity of the Millstone site are addressed in Chapter2. The licensee concludes that the generation of missiles at these facilities does not pose a credible threat to the Millstone site. Therefore, no specific protection is required other than that described for tornado-generated missiles.Therefore, Millstone Unit Number 1 does not present an undue risk to the health and safety of the public as a result of pr oximity missile hazards.3.1.5.4Aircraft HazardsThere is presently one small commercial airport approximately 6 miles east-northeast of the site. Groton/New London Airport handles regularly scheduled commerci al passenger flights but is inadequate for handling large jets. The licensee has determined that the proba bility of an aircraft striking safety related structures of Millstone Unit Number 1 is sufficiently low that it does not constitute a significant hazard.3.1.6SEISMIC DESIGN

The Millstone Unit Number 1 plant was designed for an earthquake (equivalent to the operating basis earthquake or OBE) wi th a horizontal peak ground a cceleration (HPGA) of 0.07g and reviewed for an earthquake (equivalent to the sa fe shutdown earthquake or SSE) with a PGA of 0.17 g. A smoothed design response spectrum reco mmended by John Blume and Associates and the north 69 ° west component of the 1952 Taft earthquake record normalized to the specified HPGAs were used as seismic input for the analyses and design. The vertical component of ground motion was assumed to be two-thirds of the hor izontal components. For the dynamic analyses of seismic Class I structures, the buildings (or structures) were modeled as lumped mass-spring systems with fixed base to simulate the rock founded foundations. The dynamic responses of the Reactor Building and Radwaste Building/Control were analyzed by time-history approach.Two methods were used for the anal ysis of safety related equipment: MPS-1 DSAR3.1-10Rev. 2.1(1)the response spectrum analysis approach with smoothed response spectrum recommended by John Blume and Associates as input, and(2)the equivalent static method using peak structural responses as input. Chapter 4 of the NRC NUREG/CR-2024 report, "S eismic Review of the Millstone 1 Nuclear Power Plant prepared for the NRC as part of the Systematic Evaluation Program" (Reference 3.1-21), summarizes the detail s of the original analysis and design.To assure that the Reactor Building enclosure (str ucture above elevation 108 feet 6) is capable of withstanding an SSE with a peak ground acceleration of 0.17g without adversely affecting nearby safety related SSCs (Seismic II/I criterion), this portion of the structure is analyzed for the realistic, median-centered in-structure acc elerations developed by Vectra Technologies (Reference 3.1-32) for use in the USI A-46 (SQU G) program evaluations of equipment in the Category I portion of the Reactor Building. These fl oor accelerations and sp ectra are considered more realistic since they incorporate the variabili ties of the input motion at a rock site and the structural parameters (mass and stiffness). The SSE floor accelerations at elevation 82 feet 9 inches (highest elevation evaluated in the Vectra report) are approximately 80% of the corresponding floor accelerations obtained from the EDS Report (Reference 3.1-23). Therefore the floor accelerations at the operating floor and at the roof level are conservatively taken as 80% of the corresponding accelerations from Reference 3.1-23.3.1.6.1Comparison of Measured and Predicted Responses Plant procedures have been developed for abnor mal operational events such as earthquakes. If ground motion is detected, plant walkdowns are initiated to determine plant capability.3.1.7DESIGN OF CLASS I AND CLASS II STRUCTURES3.1.7.1Design Criteria, Applicable Codes, Standards and Specifications The design of all structures and facilities (Class I and II) conformed to the applicable general codes or specifications in effect at time of design. 3.1.7.2Loads and Loading Combinations General requirements for the design of all stru ctures and equipment include provisions for resisting the stresses resulting from dead loads, live loads and wind or seismic loads with impact loads considered as part of the live load. The treatment of equipment stresses are generally limited to those produced by non-operating loads such as the ef fect of building moti on due to earthquake on the anchorage or support for a piece of equipment. However, th e loads resulting from operating pressures or temperatures on equipment are considered where they would increase the stresses. Thermal gradients in the foundation we re not considered in the design. MPS-1 DSAR3.1-11Rev. 2.1Selection of materials to resist the expected loads is based on standard practice in the power plant field. The use of these materials is governed by local building codes and the experience and knowledge of the designers and builders. The loadings of concern are the following: D = Dead load of structure and equipment plus any other permanent loads contributing stress, such as soil, hydrostatic, temperature loads or operating pressures and live loads expected to be present when the plant is operating. E = Design earthquake load.

E' = Maximum earthquake load.W = Wind load. The criteria which have been followed for all Class I structures wi th respect to stress levels and load combinations for the postulated events are noted below: Class I portions of Reactor Building and Radwaste Building(a) D + ENormal allowable code stresses are used (AISC for stru ctural steel, ACI for reinforced concrete). The customary incr ease in design stresses, when earthquake loads are considered, is not permitted.(b) D + E'Stresses are limited to the minimum yield point as a general case. However, in a few cases, stresses may exceed yield point. In this case, an analysis, using the Limit-Design approach, will be made to determine that the energy absorption capacity exceeded the energy input. This method has been discussed in the AEC publication TID-7024, "Nuclear Reactor and Earthquakes," Section 5.7. The resulting distortion is limited to assure no loss of function and adequate factor of safety against collapse.(c) D + WNormal allowable code stresses (AISC for structural steel, ACI for reinforced concrete) with the customary increases in stresses when wind loads are considered. The maximum allowable stresses used for vari ous loading conditions are given for Class I structures in Table 3.1-2. Floor live loads were established based upon e quipment and operating loads and applied to the basic building code, which is recommended to the boroughs by the State of Connecticut. Roof live loads are a minimum of 60 psf for Class I bu ildings and 40 psf for Class II buildings. All Class I structures will withstand the maxi mum potential loadings resulting from a wind velocity of 115 miles per hour with gusts up to 140 miles per hour. Although some damage to MPS-1 DSAR3.1-12Rev. 2.1these structures could occur, this damage w ould under no circumstances impair the functions for the capability of safe storage of irradiated fuel.Accidental torsion on the structures was not considered in the analyses. The Reactor Building is a box-like structure with heavy co lumns and thick walls which give it a high torsional rigidity. Accidental eccentricity would therefore produce negligible stresses and has been ignored. Although a lack of symmetry applie s to the arrangement of Class I structures, it is felt that because the buildings are not genera lly structurally connected, torsional effects are likely to be of little consequence. In the analysis of concrete stru ctures, the design modulus of 3x10 6 psi is in accordance with the ACI building code requirements for reinforced concrete (ACI 318-63), Section 1102, which is standard design practice. However, it is recogni zed that the modulus of elasticity of concrete increases with age following the 28 day period, but it is difficult to evaluate the amount of increase. The following factors af fect the strength of concrete.(1)Curing temperature (2)Initial temperature(3)Variations in mixes(4)Amount of hydration The elastic modulus is not dire ctly proportional to the strength of concrete: nevertheless, the effect of increasing the strength causes an increase in the modulus. However, the increase in the modulus due to age is not believed to be significant in the light of all the uncertainties affecting the modulus of concrete.Whatever the small change in the modulus may be, this effect is partially accounted for by cracks in the concrete structure due to shrinkage and temperature. Such cracks tend to make the structure more flexible, which tends to compensate for th e increased modulus. Also, the percent change in the modulus is small compared to other inputs in the analysis such as dimensions, areas, cross sections, mass grouping, etc. Hence, the effect of a small modulus change on the validity of the dynamic analysis is consid ered to be negligible.3.1.7.3Structural Criteria for Class II Structures Class II structures and equipm ent are designed following the nor mal practice for the design of power plants in the State of Connecticut, but as a minimum, this was not less than given in the "Uniform Building Code" for Zone 2. The usua l practice of determin ing the stress due to earthquakes by applying a static load based on a specified seismic coefficient was followed. The design of the Class II portion of the Reactor Building is addressed in Section 3.1.6. MPS-1 DSAR3.1-13Rev. 2.1 Allowable stresses for building mate rials in Class II structures are as specified in the Basic Building Code, which is recommended to the boroughs by the State of C onnecticut. A one-third increase is allowed for combinations including seismic or wind loads.3.1.7.4Seismic Class I and II Structures3.1.7.4.1Reactor Building

FunctionThe functions of the Reactor Building are to enclose the spent fuel pool and associated equipment and protect it from the weather. It supports maintenance of the fuel configuration in the fuel pool, provides protection from external hazards and supports maintenance of water in the fuel pool to a depth necessary to ensure the irra diated fuel is always immersed. The Reactor Building at and below elevation 108 feet 6 inches is retained as seismic Class I in the permanently defueled condition. Above elevation 108 feet 6 inch es, the building enclosure is revised to seismic Class II with the requirement that the enclosure is capable of sustaining an SSE without collapse (Seism ic II/I criterion).Description The Reactor Building completely encloses the sp ent fuel pool. This building is a cast-in-place reinforced concrete structure. At the 108 foot 6 inch elevation, internal steel frame lateral bracing has been placed to support the crane and the roof of the Reactor Building. The Reactor Building is founded on rock with adequate strength at an elevation of minus 32 feet 0 inches, with a foundation of reinforced concre te 142 feet 6 inches square.The new fuel storage vault and the spent fuel storage pool are located in the Reactor Building. The reactor service and refueling area is serviced by an overhead bridge crane. A refueling service platform with necessary handli ng and grappling fixtures servic es the spent fuel storage pool. Fuel storage pool is a reinforced concrete struct ure, completely lined with seam-welded stainless steel plate which is welded to rein forcing members embedded in concrete. The pool is located between eleva tions 65 feet 9 inches and 108 feet 6 inches. The fuel pool sits on a 5 feet 4 inch thick reinforced concrete slab which is supported by the reactor building perimeter and the primary contai nment drywell wall. The pool st ainless steel liner prevents leakage even in the unlikely event the concrete develops cracks. The liner was designed considering thermal stress , and the welds were dye penetrant inspected to ensure leak tight integrity. Construction material s used in the construction of the spent fuel storage facility includes 4000 psi, 28 day strength concrete, 40 ksi deformed bar reinforcing steel, and ASTM, A-167, Type 304 stainless steel. MPS-1 DSAR3.1-14Rev. 2.1 Reactor Building Seismic Design and Analysis Based on the recommended earthquake design criteria established for the station, envelopes of maximum acceleration, displacement, shear and ove rturning moment versus height have been developed and are presented for the two assu med earthquake directi ons. See Figures 3.1-1 through 3.1-5. Based on the data developed by J ohn A. Blume and Associates, engineers, the design criteria have been established as follow s for computerized analysis: the mathematical model was subjected to an excursion through the north 69 ° west component of the 1952 Taft earthquake with an applied factor of 7/17. The resulting maximum shears, moments and displacements were used for design. The maximum envelopes of building design shear s, moments and displacements are presented graphically in Figures 3.1-3 through 3.1-5, respectively. These curv es have been used in the seismic design of the Reactor Building. Loads and shears from reactor pressure vessel and associated piping and equipment are transferred to the drywell structure and pedestal and then to the foundation mat. Careful grouting between the drywell and mat ensures direct transfer of compressive loads to the mat. Shears are tr ansferred to the mat by friction and bearing. The Reactor Building was designed to resist the seismic shears and moments presented herein without the usual increase in stress for shor t-term loadings. In addition, the structure was reviewed to assure that it can resist 2.4 times the postulated seismic shears and moments without causing injury to the structure. In addition to the horizontal accelerations, a vertical building (and equipment) acceleration was used for design. The Reactor Building enclosure st ructure (above elevation 108 feet 6 inches) is analyzed for a realistic median-centered SSE, as described in S ection 3.1.6, and is shown to resist the resulting inertial loads from the accelerations with no loss of structural integrity.3.1.7.4.2Control Room and Radwaste Treatment BuildingDescription The waste treatment facility is north of and adjacent to the reactor building. The building includes equipment and tankage space below grade with th e plant control room above grade. The area below grade is of reinforced concrete construc tion with shielded compartments provided for the various pieces of radwas te equipment. The control room above grade is of reinforced concrete walls with a two foot thick reinforced concrete roof. The control room and radwaste facility are considered seismic Class II. The analytical model used in the seismic analysis of the control room and radwaste building is shown in Figure 3.1-6 and is similar to those for the Reactor Building. The Radwaste Building is seis mically analyzed consistent with Regulatory Guide 1.143.General Structural Features The building substructure is f ounded on rock. The maximum bearing pressure on the rock is 10 tons per square foot. The exteri or walls are of cast-i n-place concrete and designed for an earth pressure per square foot at any depth equal to the depth in feet times 90 pounds. The exterior walls MPS-1 DSAR3.1-15Rev. 2.1 and the base slab were originally designed to resi st hydrostatic pressure a nd uplift due to exterior flooding to elevation 19 feet 0 inches. In the de fueled condition, the below grade elevations are allowed to flood such that the uplift force is reduced.The interior walls of the substructure are of cast-in-place concrete and those for th e superstructure are either cast in place or made of concrete masonry units. With minor exceptions, all floors are poured-in-place concrete slabs. The east half grade floor at el evation 14 feet 6 inches, including the concrete shielding plugs which close hatchways over equipmen t in the substructure, is designe d for a uniform live load of 200 psf.All tanks are made of ductile metal and all sump pits are lined so that these containers can be subjected to substantial distortion without rupture. The substructure is massive reinforced concrete , not subject to fracturing. Even in the event fracturing occurred, seepage would be into the build ing rather than out, since the water table is above basement level.3.1.7.4.3Intake Structure The intake structure is a reinforced concre te frame supported on a reinforced concrete substructure which is founded on rock. The buildi ng has a flat roof cons isting of 10 gauge steel with concrete slab covered wi th insulation and a tar and felt roofing membrane. Hatches are provided in the roof for removal of major pieces of equipment. The front wall of the intake structure is designed to resist the standing wave. Seismic stress levels were calculated using coefficients of 0.07 g at grade and 0.12 g at th e roof level for design earthquake and 2.4 times these values for the maximum earthquake. The structure is capable of withstanding 300 mph wind but not the tornado internal pressure of 2.5 psi. However, the large number of hatches in the roof will release this pressure. Although originally desi gn as seismic Class I, the intake is considered seismic Class II in the pe rmanently defueled condition.The intake structure is located west of the main plant and has five 11 foot 2 inch wide bays. Each bay is provided with manually rake d trash racks and stop log guides. Provision for service and cooling wa ter strainers is made in a separate covered pit adjacent to the intake.3.1.7.4.4Turbine Building The Turbine Building is a Class II structure. The Turbine Building foundation consists of a reinforced concrete mat supported on rolled structural steel H section bearing piles. All piles were driven to rock or to refusal in the dense strata immediately above rock. Reinforced concrete shield walls are provided up to the operating de ck at elevation 54 feet 6 inches. MPS-1 DSAR3.1-16Rev. 2.1The remaining portions of the building have steel framing and metal siding. The Turbine Building ground floor consists of a reinforced concrete slab supported on sand fill over the foundation mats. The turbine generator pedestal is a massive reinforced concrete pedestal designed to support the turbine generator. It is supported on a six foot thick mat which forms an integral part of the remaining building mat foundations. The roof is covered with metal decking, insulation and roofing material flashed at the parapet walls. An overhead rolli ng door at the west end of the building provides rail car access into the building.3.1.8SEISMIC QUALIFICATION OF SEISMI C CATEGORY I INST RUMENTATION AND ELECTRICAL EQUIPMENT None of the plant process instru mentation provides safety relate d functions in conjunction with the storage and handling of irradiated fuel or radioactive waste, or is credited with any function in the safety evaluations performed to ensure that no undue risk to the health and safety of the public exists. No plant instrumentation or electrical systems are required for mitigation of the design basis fuel handling accident. Seismic qualificat ion of plant instrumentation and electrical equipment is not required.3.1.9ENVIRONMENTAL DESIGN OF ELECTRICAL EQUIPMENTThis section is related to qualification of the electrical portion of the engineered safety features to perform their intended functions in the comb ined normal, accident and post accident environments. There are no non-structural engineered safety features related to the safe storage and handling of the irradiated fuel or radioactiv e waste, or credited in the safety evaluations performed to ensure that no undue risk to the h ealth and safety of th e public exists. No non-structural engineered safety features are credited in accident analysis to prevent or mitigate the consequences of the current desi gn basis fuel handling accident.3.1.10REFERENCES3.1-1Millstone Nuclear Power Station Unit Number 1 Application for Full Term Operating License, September 1, 1972.3.1-2NUREG-0824, Integrated Plant Safety As sessment, Systematic Evaluation Program, Millstone Nuclear Power Station, Unit Number 1, February 1983.3.1-3Samual J. Chilk (N uclear Regulatory Commission) memo to J. M. Taylor (Nuclear Regulator Commission), "SECY-92-233 Resolution of Deviations Identified during the Systematic Evaluation Progr am" dated September 18, 1992.3.1-4Integrated Plant Safety Assessment, Syst ematic Evaluation Program, Millstone Nuclear Power Station, Unit Number 1, NUREG-0834, Supplement Number 1, November 1985.3.1-5Letter, C.I. Grimes (NRC) to J.F. Opeka, subject: IPSAR Sections 4.4 Wind and Tornado Loadings and 4.7 Tornado Missiles. MPS-1 DSAR3.1-17Rev. 2.13.1-6Letter, February 2, 1984, from W.G. Counsil to D. M. Crutchfield (NRC),

Subject:

Millstone Nuclear Power Station Unit Number 1, SEP Topics II-3.B Flooding Potential and Protection Requirements, III-2 Wind and Tornado Loadings, III-3.A Effects of High Water Level on Structures, III-7.B Design Codes, Design Criteria and Load Combinations.3.1-7Letter, March 16, 1984, from W. G. Counsil to D. M. Crutchfield (NRC),

Subject:

Millstone Nuclear Power Station Unit Number 1, SEP Topics II-3.B Flooding Potential and protection Requirements, II-4.F Settleme nt of Foundations and Buried Equipment, III-2 Wind and Tornado Loadings, III-3.A Effects of High Water Level on Structures, III-6 Seismic Design Considerations.3.1-8Letter, October 7, 1983, from W. G. Couns il to D. M. Crutchfield (NRC),

Subject:

Millstone Nuclear Power Station Unit Number 1, SEP Topic III-2 Wind and Tornado Loadings.3.1-9Letter, December 3, 1982, from W. G. Counsi l to D. M. Crutchfield (NRC),

Subject:

Millstone Nuclear Power Station Unit Number 1, SEP Topic III-2 Wind and Tornado Loadings.3.1-10Letter, February 4, 1986, from J. F. Opeka to C.I. Grimes (NRC),

Subject:

Millstone Nuclear Power Station Unit Number 1 ISAP Topic 1.19, Integrated Structural Analysis.3.1-11Letter, W. G. Counsil to D. M. Cr utchfield (NRC), dated March 16, 1984, Millstone Nuclear Power Station, Unit Number 1, SEP Topic II-3.B, Flooding Potential and protection Requirements, SEP Topic II-4.F, Settlement of Foundations and Buried Equipment, SEP Topic III-2, Wind and Tornado Loadings, SEP Topic III-3.A, Effects of High Water Level on Structures SEP Topi c III-6, Seismic Design Considerations.3.1-1210CFR50, Appendix A, General Design Criterion 4.3.1-13Letter of June 29, 1982, W.G. Counsil to D. M. Crutchfield,: Millstone Nuclear Power Station Unit Number 1, SEP Topic III-4.A, Tornado Missiles." 3.1-14Letter of March 9, 1982, W.G. Counsil to D. M. Crutchfield,: Mi llstone Nuclear Power Station Unit Number 1, SEP Topic III-4.A, Tornado Missiles." 3.1-15Letter of November 19, 1981, W.G. Counsil to D. M. Crutchfield: Millstone Nuclear Power Station Unit Number 1, SEP Topic III-4.A, Tornado Missiles." 3.1-16Letter of August 31, 1981, W.G. Counsil to D. M. Crutchfield,: Mi llstone Nuclear Power Station Unit Number 1, SEP Topic III-4.A, Tornado Missiles." 3.1-17Letter of October 16, 1985, J. F. Opeka to C. I. Grimes, Millstone Nuclear Power Station Unit Number 1, "Integrated Sa fety Assessment Program." MPS-1 DSAR3.1-18Rev. 2.13.1-18Letter of November 25, 1985, C. I. Grimes to J. F. Opeka, Integrated Plant Safety Assessment Report, Section 4.4, Wind and Tornado Loadings, Section 4.7, Tornado Missiles - Millstone Unit Number 1. 3.1-19Letter of April 29, 1981, W. G. Counsil to D. M. Crutchfield, "Millstone Nuclear Power Station Unit Number 1, SEP Topic III-4.D, Site Proximity Missiles." 3.1-20Letter of September 17, 1981, W. G. Counsil to D. M. Crutchfield, "Millstone Nuclear Power Station Unit Number 1, SEP Topi c III-4.D, Site Proximity Missiles." 3.1-21NRC NUREG/CR-2024 Report, "Seismic Revi ew of the Millstone-1 Nuclear Power Plant," July 1981.3.1-22SEP Safety Topics III-6, Seismic Design Considerations and III-II, Component Integrity - Millstone Nuclear Power Station Un it Number 1, SAR dated 6/30/82. 3.1-23EDS Report Number 02-0240-1094, "Generation of In-Structure Seismic Response Spectra Millstone Unit Number 1," dated June 1982.3.1-24NRC letter,"Site Specific Ground Response Spectra for SEP Plants Located in the Eastern United States," June 17, 1981.3.1-25NRC NUREG/CR-1582 Report, "Seismic Hazard Analysis," Vols. 2-4, October 1981.3.1-26Letter J. F. Opeka to C.I. Grimes, "Millstone Nuclear Station, Unit Number 1 ISAP Topic 1-19, Integrated Structural Analysis," dated January 6,1986.3.1-27Letter from D. G. Eisenhut, NRC, to W. G. Counsil, dated January 1, 1980. 3.1-28Letter from D. M. Crutchfield, NRC, to W. G. Counsil, dated July 28, 1980. 3.1-29Letter from W. G. Counsil to D. M. Crutchfield, NRC, dated October 16, 1985.3.1-30Millstone Unit 2 Fi nal Safety Analysis Report Section 5.8.6.3.1-31Millstone Unit 3 Fi nal Safety Analysis Report Section 3.7.4.2.3.1-32Vectra Technologies Report Number 0024-00099-RB-1, Rev. 1, "Reactor Building A-46 Spectra," dated June 10, 1996. MPS-1 DSAR Page 1 of 4 Rev. 2TABLE 3.1-1COMPARISON WITH NR C GENERAL DESIGN CRITERIAGENERAL DESIGN CRITERIASEP AND TMI SAFETY ISSUES WHICH LISTED THE SPECIFIED GDC AS PART OF CONCERNAFFECTED REGULATORY GUIDE I OVERALL REQUIREMENTS1QUALITY STANDARDS AND RECORDSSEP II-3.A , II-3.B, II-3.C, III-3.A AND III-7.B1.27, 1.592DESIGN BASES FOR PROTECTION AGAINST NATURAL PHENOMENA SEP II-2.A, II-3.A, II-3.B, II-3.C, II-4.E, II-4.F, II-4.3, III-19 II I-2, III-3.A, III-3.B, III-3.C, III-6, III-7.B, III-8.C, III-11, VI II-3.A, VIII-3.B, TMI II.B.11,27, 1.,32, 1.59, 1.60, 1.61, 1.68, 1.75, 1.76, 1.92, 1.102, 1.117, 1.120, 122, 1.127, 1.129, 1.1323FIRE PROTECTION(SEE DSAR Section 3.2.9) 4ENVIRONMENTAL AND MISSILE DESIGN BASESSEP II-1.C, II-3.A, II-3.B, II-3.C, III-1, III-4.B, II I-5.A, III-5.B, III-7,B, III-1 1,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 1.3, 1.4, 1.7, 1.20, 1.27, 1.29, 1.32, 1.35, 1.45, 1.46, 1,59, 1.68, 1.75, 1.115, 1.12,5SHARING OF STRUCTURES, SYSTEMS AND COMPONENTSSEP III-1, VIII-3.A AND VIII-3.B1.32, 1.75, 1.129II PROTECTION BY MULTIPLE FISSION PRODUCT BARRIERS10REACTOR DESIGNNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION11REACTOR INHERENT PROTECTIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION12SUPPRESSION OF REACTOR POWER OSCILLATIONSNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION13INSTRUMENTATION AND CONTROLNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION14REACTOR COOLANT PRESSURE BOUNDRYNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELE D CONDITION15REACTOR COOLANT SYSTEM DESIGNNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION16CONTAINMENT DESIGNNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION17ELECTRIC POWER SYSTEMSSEP III-1 , VII-7, VIII-2, VIII-3.A VIII-3.B, TMI II.E.3.1, II.G.11.6, 1.9, 1.32, 1.75, 1.12918 INSPECTION AND TESTING OF ELECTRIC POWER SYSTEMSNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION19CONTROL ROOMNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION MPS-1 DSAR Page 2 of 4 Rev. 2III PROTECTION AND REAC TIVITY CONTROL SYSTEMS20PROTECTION SYSTEM FUNCTIONSNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION21PROTECTION SYSTEM RELIABILITY AND TESTABILITYNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION22PROTECTION SYSTEM INDEPENDENCENOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION23PROTECTION SYSTEM FAILURE MODESNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION24SEPARATION OF PROTECTION AND CONTROL SYSTEMSNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION25PROTECTION SYSTEM REQUIREMENTS FOR REACTIVITY CONTROL MALFUNCTIONSNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION26REACTIVITY CONTROL SYSTEM REDUNDANCY AND CAPABILITYNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION27COMBINED REACTIVITY CONTROL SYSTEMS CAPABILITYNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION28REACTIVITY LIMITSNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION29PROTECTION AGAINST ANTICIPATED OPERATIONAL OCCURRENCESNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONIV FLUID SYSTEMS 30QUALITY OF REACTOR COOLANT PRESSURE BOUNDARYNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION31FRACTURE PREVENTION OF REACTOR COOLANT PRESSURE BOUNDARYNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION32 INSPECTION OF REACTOR COOLANT PRESSURE BOUNDARYNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION33REACTOR COOLANT MAKEUPNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONTABLE 3.1-1COMPARISON WITH NR C GENERAL DESIGN CRITERIAGENERAL DESIGN CRITERIASEP AND TMI SAFETY ISSUES WHICH LISTED THE SPECIFIED GDC AS PART OF CONCERNAFFECTED REGULATORY GUIDE MPS-1 DSAR Page 3 of 4 Rev. 234RESIDUAL HEAT REMOVALNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION35EMERGENCY CORE COOLINGNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION36INSPECTION OF EMERGENCY CORE COOLING SYSTEMNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION37TESTING OF EMERGENCY CORE COOLING SYSTEMNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION38CONTAINMENT HEAT REMOVALNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION39INSPECTION OF CONTAINMENT HEAT REMOVAL SYSTEMNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION40TESTING OF CONTAINMENT HEAT REMOVAL SYSTEMNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION41CONTAINMENT ATMOSPHERE CLEANUPNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFU ELED CONDITION42INSPECTION OF CONTAINMENT ATMOSPHERE CLEANUP SYSTEMSNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION43TESTING OF CONTAINMENT ATMOSPHERE CLEANUP SYSTEMSNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION44COOLING WATERNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION45INSPECTION OF COOLING WATER SYSTEMNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION46TESTING OF COOLING WATER SYSTEMNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEF UELED CONDITIONV REACTOR CONTAINMENT50CONTAINMENT DESIGN BASISNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION51FRACTURE PREVENTION OF CONTAINMENT PRESSURE BOUNDARYNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONTABLE 3.1-1COMPARISON WITH NR C GENERAL DESIGN CRITERIAGENERAL DESIGN CRITERIASEP AND TMI SAFETY ISSUES WHICH LISTED THE SPECIFIED GDC AS PART OF CONCERNAFFECTED REGULATORY GUIDE MPS-1 DSAR Page 4 of 4 Rev. 252 CAPABILITY FOR CONTAINMENT LEAKAGE RATE TESTINGNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION53PROVISIONS FOR CONTAINMENT INSPECTION AND TESTINGNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION54SYSTEMS PENETRATING CONTAINMENT NOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELE D CONDITION55REACTOR COOLANT PRESSURE BOUNDARY PENETRATING CONTAINMENTNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION56PRIMARY CONTAINMENT ISOLATIONNOT APPLICABLE TO THE PERMANENTLY DEFUELEDCONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION57CLOSED SYSTEMS ISOLATION VALVESNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITIONNOT APPLICABLE TO THE PERMANENTLY DEFUELED CONDITION VI FUEL AND RADIOACTIVITY CONTROL60CONTROL OF RELEASES OF RADIOACTIVE MATERIALS TO THE ENVIRONMENTSEP II.2.C, XI-1, XI-2, TMI II.B.2, II.B.3, 2.1.6.A, 2.1.8.A1.3, 1.461FUEL STORAGE AND HANDLING AND RADIOACTIVITY CONTROL SEP XI-1, XI-262PREVENTION OF CRITICALITY IN FUEL STORAGE AND HANDLING63MONITORING FUEL AND WASTE STORAGESEP XI-1, XI-2 64MONITORING RADIOACTIVITY RELEASESSEP II-2.C, XI-1, XI-2; TMI II.B.2, II.B.3, 2.1.6.A, 2.1.8.ATABLE 3.1-1COMPARISON WITH NR C GENERAL DESIGN CRITERIAGENERAL DESIGN CRITERIASEP AND TMI SAFETY ISSUES WHICH LISTED THE SPECIFIED GDC AS PART OF CONCERNAFFECTED REGULATORY GUIDE MPS-1 DSAR Page 1 of 1 Rev. 2 25% of live loads were considered concurrent with the seismic loads Fy = Minimum Yield Point of the Material.f ' c = Compressive Strength of Concrete. NOTE 1: The structure was analyzed to assu re that gross structural integrity can be maintained during ground motion having 17/7 the intensity of the operating basis earthquake described in SE CTION 3.1.6, even though stresses in some of the materials ma y exceed the yield point.TABLE 3.1-2ALLOWABLE STRESSES FOR CLASS I STRUCTURESLOADING CONDITIONS REINFORCING STEEL MAX. ALLOWABLE STRESSCONCRETE MAX. ALLOWABLE COMPRESSION STRESSCONCRETE MAX. ALLOWABLE SHEAR STRESSCONCRETE MAX. ALLOWABLE BEARINGSTRUCTURAL STEEL TENSION ON THE NET SECTIONSTRUCTURAL STEEL SHEAR ON GROSS SECTION STRUCTURAL STEEL COMPRESSION ON GROSS SECTION STRUCTURAL STEEL BENDING(1)DEAD LOADS PLUS LIVE LOADS,* PLUS OPERATING LOAD PLUS SEISMIC LOADS (0.07g) 0.5 Fy0.45 f ' c1.1 f ' c0.25 f ' c0.60 Fy0.40 FyVARIES WITH SLENDERNESS RATIO 0.66 Fy TO 0.60 Fy(2)DEAD LOADS PLUS LOADS,

• PLUS OPERATING LOADS PLUS WIND LOADS0.667 Fy0.60 f ' c1.467 f ' c0.333 f ' c0.80 Fy0.53 FyVARIES WITH SLENDERNESS RATIO 0.88 Fy TO 0.80 Fy(3)DEAD LOADS PLUS LIVE LOADS,* PLUS OPERATING LOADS, PL US SEISMIC LOADS 0.17gGROSS STRUCTURAL INTEGRITY CAN BE MAINTAINED (SEE NOTE 1 BELOW)

MPS-1 DSAR3.2-1Rev. 83.2SYSTEMS3.2.1FUEL STORAGE AND HANDLING3.2.1.1New Fuel Storage Since Millstone Unit 1 is a de-commissioned unit, new fuel will no longer be received.3.2.1.2Spent Fuel Storage 3.2.1.2.1Design BasesThe design bases for the storage of spent fuel are as follows:a.A fuel storage pool for the underwat er storage of 2959 fuel assemblies.b.Maintain a keff of less than 0.95 at all times, incl uding postulated criticality accidents. Assumed are worst case results, considering maximum variation in the position of the fuel assemblies within the storage rack, neutron absorber variation (where credited), seismic induced deflections and calculation uncertainty. Boraflex is not credited.c.The concrete shielding walls are designed as part of the Class 1 portion of the Reactor Building structure. The thickness of the walls and the standards of design are such as to preclude structural damage or loss of function of the walls.d.Structural design of the fuel storage a nd equipment storage facilities meets all requirements for Class I structures.e.The fuel storage racks for the fuel are designe d to assure subcriticality in the fuel pool. The storage racks are an interconnected hone ycomb array of square stainless steel boxes forming individual cells for fuel storage. 1045 storage cells contain Bo raflex sheets (not credited) on four sides, a nd 2184 storage cells contain B 4 C plates for neutron absorption. Of the 1045 storage cells with Boraflex, onl y 775 cells are allowed to contain fuel.f.Criticality Accident Requi rements. Millstone 1 has chosen to comply with 10 CFR50.68(b).3.2.1.2.2Facilities Description

The fuel pool contains water which is not borated. The fuel storage pool is a reinforced concrete structure, completely lined with seam-welded, stainless steel plate (11 gauge) which is welded to reinforcing members (channels, I-beams, etc.) embedded in concrete. The liner is reinforced by increased thickness and suitable insert strips in areas subject to heavy loading such as the cask handling area The concrete shielding walls are two or more feet thick and are designed as part of the Class I portion of the Reactor Building structure. MPS-1 DSAR3.2-2Rev. 8 Interconnecting drainage paths are provided behind the liner welds to:(1)Prevent pressure buildup behind the liner plates, (2)To control the loss of pool water and (3)To provide liner leak detection and measurement capability. The drainage paths are suitably grouped to indicate the area of leakage. To avoid unintentional draining of the pool, there are no penetrations th at would permit the pool to be drained below approximately nine feet above the top of the act ive fuel, and all lines extending below this level are equipped with suitable valving to prevent backflow. The passage be tween the fuel storage pool and the refueling cavity above the reactor vessel is provided with two gates. The refueling cavity is maintained in a drained down state. The gate adjacent to the refueling cavity is welded to the passage liner forming a permanent pressure boundary for the fuel storage pool. The double sealed gate adjacent to the fuel storage pool is removable but normally maintained in the closed position. A normally open drain line be tween the gates permits detect ion of leaks from the gate adjacent to the fuel storage pool. The drain line may be isolated and the volume between the gates flooded to support removal of the gate for repairs in the event of such leakage. In response to the NRC I.E. Bulletin 84-03, a ugmented leak detection capability has been installed in the spent fuel pool to indicate high/low level in the pool. The water in the pool is cooled and filtered as required by the sp ent fuel pool cooling and in-pool cleanup system descri bed in Subsection 3.2.1.3. The storage pool is designed to hold 20 fuel channels.An area of approximately seven feet by seven and one half feet is reserved for loading a spent fuel shipping cask.Canisters containing irradiated reactor vessel internals and other materials classified per 10CFR61 as greater than class "C " (GTCC) waste are stored in th e fuel storage pool adjacent to the fuel shipping cask area.3.2.1.2.3Safety Evaluation

The spacing of fuel bundles in the spent fuel storage pool, the presence of neutron absorbing poisons (where credited) in the fuel storage r acks, not placing fuel in prohibited locations identified in the Technical Specifications, and the design of the fuel bundles maintains keff less than or equal to 0.95. This is assured by limiting the fuel assemblies in the pool to those that have a maximum K of 1.24 in the norma l reactor configuration at co ld conditions, and an average U-235 enrichment of 3.8 weight percent or less. The criticality analysis confirms acceptable results regardless of the spent fuel pool temperature. MPS-1 DSAR3.2-3Rev. 8Irradiated fuel being moved in the fuel storage pool is covered by an eight foot minimum of water above the top of active fuel, which is sufficient for radiation shielding. Ra diation monitors in the fuel storage pool work area monitor the radi ation level and alarm upon excessive levels. Limit switches on the refueling platform hoists in terrupt power to the hoist s when raising fuel, at a point that ensures a minimum of eight feet of water above the top of active fuel. The brakes on the refueling platform equipment lock upon loss of power. The fuel storage racks are analyzed to withst and the impact of a dropped fuel assembly and handling tool with a combined dry weight of 1675 pounds from the maximum lift height of the refueling platform telescoping mast. The anal yses performed (References 3.2-9 to 3.2-12) demonstrate that the spent fuel racks remain f unctional and that the spent fuel remains in a subcritical, submerged and coolable condition.A liquid level transmitter, monitoring pool water level, is provided to detect loss of water from the pool. A level transmitter, monitoring the skimmer surge tank, is provided to permit water loss detection by initiating a low level alarm and pr ovide level indication in the Millstone Unit 2 Control Room.A safety evaluation of spent fuel can be found in References 3.2-1, 3.2-2, 3.2-3., 3.2-4, 3.2-5, and 3.2-8.3.2.1.3Spent Fuel Pool Cooling System The Spent Fuel Pool Cooling System has been an alyzed to remove the maximum heat load from the spent fuel pool.3.2.1.3.1Design BasesThe Fuel Pool structure, pool liner, fuel racks, and external cooling system have been designed for a temperature of approximately 150 °F. However, all of these st ructures and components have been demonstrated to be structurally adequa te for abnormal temperature excursions to 212 °F. With a complete loss of external cooling and a closed airspace above the pool, it would take approximately 10 days for the pool temperature to rise to 212 °F from an initial SFP bulk water temperature of 100 °F, or approximately 7.5 days to rise to 212 °F if starting from the TRM upper temperature limit of 140 °F. The spent fuel pool cooling sy stem and secondary DHR cooling system have been qualified for satisfactory operation with pool temperatures as high as 170 °F. This is greater than the maximum anticipated pool water temperature, following loss of cooling, provided that natural ventilation within the reactor building is established within approximately 5 days if starting from an initia l SFP bulk water temperature of 100 °F, or 2.5 days if starting from the TRM upper temperature limit of 140 °F.Multiple methods are available to add water to the pool and adequate time is available to repair, manually reinstate or line up the system used for pool water cooling. Most significantly, if this system is not used to cool the pool water, no fuel damage would result and the potential off site exposure would not approach the guidelines established in 10CFR50.34(a) or 10CFR100.11 MPS-1 DSAR3.2-4Rev. 8 provided makeup is initiated at a rate equal to or greater than the maximum evaporation rate at any time prior to fuel uncovery. Water above the fuel provides shielding and heat sink functions. For the permanently defueled condition, the design bases for the fuel pool cooling system is: a.To maintain the bulk water temperature for the spent fuel pool at a temperature less than or equal to 140 °F and greater than 68 °F.b.To provide high clarity water to the fu el pool using the in-pool cleanup system. c.To remove radioactivity released to th e pool water using the in-pool cleanup system.3.2.1.3.2Spent Fuel Pool Heat LoadMillstone Unit Number 1 has permanently ceased power operation and all irradiated fuel has been permanently removed from the reactor vessel. Ther e are 2885 irradiated fuel assemblies in the spent fuel pool including one se gmented bundle, consisting of 19 fuel rods. A decay heat load calculation was performed utiliz ing the computer program ORIGEN2, an industry standard for such analysis (Reference 3.2-13). The results s how that total heat load in the pool was 1.781 MBtu/hr on 1/1/99. The spent fuel pool secondary cooling system (DHR) has been sized to remove the spent fuel decay heat load of approximately 1.5 Mbtu/hr, projecte d to exist on 6/1/00. 3.2.1.3.3Loss of Fuel Pool CoolingWith the spent fuel pool heat load establ ished, a second calculation (Reference 3.2-14) was performed to determine the transient and stea dy state spent fuel pool and reactor building temperatures without active cooling to the spent fuel pool. Several cases were analyzed with different ventilation configurat ions such as forced ventilat ion, natural ventilation and no ventilation through the building. Steady state and transient cal culations were performed to establish maximum pool and building temperatures a nd evaporation rates, as well as time frames for potential operator actions. All analyses were performe d using the GOTHIC computer program.The limiting case evaluated wa s during summer conditions (92 °F, 50% Relative Humidity) following the loss of active spent fuel pool cooling and without the reactor building HVAC system in operation. In this case the time to reach 212 °F in the spent fuel pool is approximately 7.5 days if starting from the TRM upper temperature limit of 140 °F. This calculation also establishes a maximum evaporative loss of 3.8 gpm under th e above conditions. If natural ventilation is established, by opening the reactor building truc k bay doors, equipment hatch garage doors and the tornado dampers on the reactor building roof, the maximum ca lculated pool temperature is 163°F and the maximum evaporation rate is 3.0 gpm.3.2.1.3.4System Description The spent fuel pool cooling system cools water in the fuel pool on an as needed basis to maintain water temperature. An in-pool demineralizer and filter maintain purity and water quality. Water is MPS-1 DSAR3.2-5Rev. 8 circulated by either one or two pumps which take suction from the skimmer surge tanks. The adjustable spent fuel pool weir gates maintain pool level and skim water from the surface of the fuel pool. System lineups may vary due to decrea sing heat removal needs. The flow diagram for the spent fuel pool cooling system is shown in Figures 3.2-1 through 3.2-3. The bulk temperature monitoring system consists of a single RTD (combined temperature transmitter and level transmitter) with the sens ing element located appr oximately 9' below the normal water level of the fuel storage pool, and a local temperature indicator. The transmitter output is monitored in the Millstone Unit 2 Control Room via the Programmable Logic Controller (PLC) which provides both indication of bulk te mperature and notification of a high and low water temperature conditions within the fuel storage pool. The in-pool fuel pool demineralizer operates on an as needed basis to maintain pool water chemistry. The in-pool filter operates on an as needed basis to maintain pool water clarity. The skimmer surge tanks are shielded with concrete. The fuel pool cooling system is controlled and operated locally and from the Millstone Unit 2 Control Room. The system is provi ded with indicators and alarms for system flow, water level, and temperature, skimmer surge tank le vel, and component operating status.3.2.1.3.5Safety Evaluation

The fuel pool water acts passively to transfer decay heat from the fuel and will protect the fuel from damage without human interven tion as long as the fuel is completely immersed in water. If external cooling is stopped, the pool water temper ature would gradually increase, resulting in no fuel damage. In the most severe case of a closed airspace, with th e current decay heat load in the Millstone Unit Number 1 Fuel Pool and no exte rnal cooling, the pool te mperature would only reach equilibrium (stop rising) when the pool water boils, which is the natural limit of water temperature in a space at atmospheric pressure. The fuel pool structure, pool liner, fuel racks, and external cooling system have been demonstrat ed to be adequate for abnormal temperature excursions to 212 °F. With a complete loss of external cooling and a closed airspace above the pool, it would take approximately 10 days for the pool temperature to rise to 212 °F from an initial SFP bulk water temperature of 100 °F, or approximately 7.5 days to rise to 212 °F if starting from the TRM upper temperature limit of 140 °F. This is significantly longer than required to reinstate external cooling of the water. If natural ventilation is established, by opening the reactor building truck bay doors, equipment hatch garage doors a nd the tornado dampers on the reactor building roof, the maximum calculated pool temperature is 163 °F.3.2.1.4Fuel Handling System 3.2.1.4.1Design Bases

The design bases for the fuel ha ndling system are as follows:a.No release of contaminati on or exposure of personnel to radiation will exceed the 10CFR20 limits. MPS-1 DSAR3.2-6Rev. 8b.Limited work on irradiated com ponents will be possible at any time.3.2.1.4.2System Description The fuel handling system handles irradiated fuel. A refueling platform, equipped with a refueling grapple and two on e-half ton auxiliary hoists is provided for servicing the fuel storage pool. The operating floor is serviced by the Reactor Building crane, which is equipped with a 110 t on main hoist and a seve n-ton auxiliary hoist. These hoists can reach any major equipment storage area on the operating floor.3.2.1.4.3Safety Evaluation

The refueling bridge and other fuel handling eq uipment are required for movement of fuel and other items stored in the fuel pool into storage/ shipping containers. The re actor building crane is required to move storage and shi pping casks in the reactor buildi ng. These functions are required in the permanently defueled condi tion, but are not safety related.3.2.2MONITORING AND CONTROL FUNCTIONS The Millstone Unit 2 Control Ro om serves as the control room for Millstone Unit 1, and is continuously manned. It is descri bed in Section 7.6 of the Millstone Unit 2 Final Safety Analysis Report. Millstone Unit 2 Operat ions personnel are responsible fo r the monitoring and control of the Unit 1 spent fuel pool island (SFPI) and auxili ary systems via a computer console located in the Millstone Unit 2 Control Room. The computer consol e in the Millstone Unit 2 Control Room interfaces with a Programmable Logic Controll er (PLC) for data acquisition and trending. The PLC is located in the Millstone Unit 1 Central Monitoring Station (CMS). The CMS is located within the Maintenance Shop. The Millstone Unit 1 CMS is not manned. It cont ains two computer consoles that may only be used as monitors, because they are normally in a locked supervisory mode.There are no monitoring stations in the original Unit 1 Control Room. The original Unit 1 Control Room no longer performs any Unit 1 function.3.2.3DECAY HEAT REMOVAL (DHR) SYSTEM 3.2.3.1Design Bases The DHR system is designed to provide cooling to the spent fuel pool cooling system. The system design bases are:Design Temperature: 170 °FDesign Flow Rate (maximum) 625 gpm per pump MPS-1 DSAR3.2-7Rev. 8Design Pressure: 200 psigThe DHR system is normally in service to supply spent fuel pool cooling system cooling loads as needed. System lineups vary during the perman ently defueled condition due to reduced heat removal needs.3.2.3.2System Description

The DHR system provides a supply of cooling water to the shell side of the spent fuel pool heat exchangers. Water is circulated in a closed loop by the DHR pumps. Heat is removed from the system by the four DHR air-water heat exchange rs located outside on th e roof above the H&V area. System configuration may vary depending on heat load. The remainder of the system consists of a cooling water expansion tank, an air separator, piping and valves, and controls and instrumentation. A demineralizer maintains system activity below established limits. The flow diagram for the system is shown in Figure 3.2-4.3.2.3.3Safety Evaluation The DHR system supplies cooling water to the fuel pool heat exchangers. Fuel pool cooling is a function that is required for the permanently defueled condition, but is not safety related. Therefore, this function of the DHR system is not safety related. 3.2.3.4Testing and Inspection The system components and instrumentation are tested periodically as necessary to ensure operational readiness.3.2.3.5Instrumentation DHR system instrumentation and c ontrols are located locally and in the Millstone Unit 2 Control Room.3.2.4MAKEUP WATER SYSTEM 3.2.4.1Demineralized Water 3.2.4.1.1System Description The spent fuel pool makeup system will supply and store demineralized water to makeup for evaporation and leakage in the pool. The primary source will be from the Unit 2 Primary Makeup System which is supplied from the onsite water treatment facility. A 5,000 gallon storage tank and transfer pump are installed in the reactor building to provide makeup water to the spent fuel pool during period when the normal makeup from Unit 2 is unavailable. A connection to the pool makeup line is also provided n ear the reactor building truck ba y door to allow makeup to be provided by a tanker truck or fire water if necessary. MPS-1 DSAR3.2-8Rev. 8 The piping, tanks and other equipment of the spen t fuel pool water storage and makeup system and makeup system are of corrosion resistant metals which prevent contamination of the makeup water with foreign material. The flow diagram for the syst em is shown in Figure 3.2-5.3.2.4.1.2Safety Evaluation

The spent fuel pool makeup water system provides demineralized makeup wate r to the spent fuel pool and spent fuel pool cooling system. This function supports fuel pool cooling, but is not safety related.3.2.4.1.3Testing and Inspection Operation of the makeup system is on demand at intermittent intervals to replenish water in the spent fuel pool makeup water storage tank and the skimmer surge tanks. The equipment is visually inspected periodically. Sampling of the makeup water storage tank is a standard monitoring procedure.3.2.4.1.4Instrumentation The motor control switch for the makeup water transfer pump is located locally at the pump. Local makeup storage tank level indication is also provided. 3.2.5INTENTIONALLY DELETED 3.2.6PROCESS SAMPLING SYSTEM 3.2.6.1Design Bases The reason for sampling process gases is to provide representative sample s for testing to obtain data from which the performance of the pl ant equipment and systems are determined.3.2.6.2System Description The Unit Number 1 BOP ventilation exhaust fl ow is continuously sampled for radioactive particulates. The sample is taken from the exhaust duct which runs along the north exterior wall of the Reactor Building. A single point sample nozzle is positioned to obtain a representative sample of the well mixed exhaust air. The sample passes through a particulate filter and is then expelled back into the exhaust duct. The SFPI ventilation exhaust flow is conti nuously monitored for gaseous radiation and particulates. The sample is taken from the exhaust duct near the reactor building exhaust plenum. A single point sample nozzle is positioned to obtai n a representative sample of the turbulent and well mixed exhaust air. The sample passes through a particulate filter and a gas monitor and is then expelled back into the exhaust duct. MPS-1 DSAR3.2-9Rev. 8 Grab samples can be taken from the BOP and SFPI ventilation exhaust duc ts and analyzed for radioactive content.3.2.6.3Safety Evaluation

The BOP and SFPI ventilation systems are not safety related. 3.2.6.4Testing and Inspection Functional tests were performed af ter installation. Routine use substitutes for subsequent periodic testing, with the exception of calibration and maintenance.3.2.7ELECTRICAL SYSTEMS 3.2.7.1Introduction The station electrical systems include the equipment an d facilities which provide power to desired plant equipment, instrumentation and controls. The system is desi gned to provide reliable power for the permanently defueled condition. The power system is designed with a sufficient source, relay protection, control, and necessary switching.3.2.7.2Off Site Source The off site source is through the emergency st ation service transformer (ESST), which steps down a 23 kV source from the Waterford Substation 36F2 circuit to 4160V.The off site power system is designed to provide a reliable source of power to the on site AC power distribution system. 3.2.7.3Intentionally Deleted. 3.2.7.4On Site Electric System 3.2.7.4.1Introduction Sufficient time is available to operators following a loss of offsite power to assure the continued safe storage of fuel without reliance on emergency sources of power. AC power is provided through the emergency station service transformer. The emergency station service transformer has adequate capacity to su pply all normal auxiliaries required to support the permanently defueled condition. Power for the SFPI and other decommissioning related activities is from the ESST via Bus 14H. SFPI and decommissioning related 125V DC power is obtained from rectified AC power at the point of use, and a separate 125V DC source cons isting of a 125V DC battery, a battery charger, disconnect switch and distribution panel. 10-1 MPS-1 DSAR3.2-10Rev. 83.2.7.4.24160 Volt System The emergency station service transformer (15G-31S) steps down 23 kV to 4160 volts for the auxiliary buses. In the permanentl y shutdown condition, the plant will normally be operated with auxiliary electrical loads supplied from the emergency station service transformer. The circuit breakers located on 4160 V bus 14H are operated locally at the switchgear. Breakers will trip automatically when over-current conditions exist. The control power to the 4160 volt bus circuit breaker is from the de commissioning 125 volt DC system.The major component of the 4160 volt power system is described below.(1)Emergency Station Service Transformer The emergency station service transformer is an outdoor, 27,750-4160 volt three phase, 60 Hz., 200 kV BIL, 10.7/12.5 MVA OA/FA 55 °C, and 14 MVA, FA 65 °C, transformer.3.2.7.4.3480 Volt System

Power from 4160 volt bus 14H is stepped down through transformers energizing the 480 volt buses SFPI-B1 and FAC-B2. Bus and MCC supply breakers are opened and closed locally. All breakers will trip automatically when overload conditions exist. 3.2.7.4.4120 Volt Systems The SFPI electrical syst em utilizes its own dedi cated 120V AC power deri ved from the SFPI AC power system. The SFPI instrument AC system is provided by the SFPI 120V AC di stribution system and backed up by point of use UPS equipment. The SFPI PLC system has an integral 24V DC power supply. 3.2.7.4.5AC Power System Design Criteria (1)Interrupting Capacity - The switchgear, load centers, motor control centers, and distribution panels are size d for interrupting capacity based on maximum short circuit availability at their location. Low voltage me tal enclosed breakers at load centers and molded case breakers at motor control center s are adequately sized for these maximum available short circuit currents.(2)Electrical System Protection - Electrical system protecti on is provided by protective devices or relays which monitor the electrical characteristics of the equipment and/or power system to assure operation consiste nt with design parameters, as follows: MPS-1 DSAR3.2-11Rev. 8(a)Initiate removal from serv ice any piece of equipment which has sustained a fault.(b)Provide automatic supervision of manual and/or automatic operations which could jeopardize the safe operation of the plant.3.2.7.4.6125 V DC System SFPI related 125V DC utilizes rectified AC power. The rectifiers are located at the SFPI 480V AC switchgear bus. In addition, th e decommissioning 125V DC system consists of a 125V DC battery, charger, disconnect switch and distribution panel.3.2.7.4.7Intentionally Deleted 3.2.7.4.8Safety Evaluation In the permanently defueled condition portions of the electrical system s are required for power and/or control of required non-safety related e quipment in other systems. Since none of the equipment powered by these systems is safety related (Class 1E), all of the electrical systems are non-safety related. Although single failure criteria still apples to the unit, it need not be applied to systems and equipment that are non-safety rela ted. Since none of the electrical systems or equipment is safety related or required for Regulatory Guide 1.97 (post accident monitoring) commitments, the EEQ program need not be applied. General Design Criteria Number 17 (Electric Power Systems) includes certain requirements for availability of offsite power to support critical functions. Since the reactor cannot be ma de critical under allowed plant conditions in the permanently defueled condition, no power sour ce is required to be operable or available.3.2.8AIR CONDITIONING, HEATING, COOLING AND VENTILATION SYSTEMS 3.2.8.1Reactor Building and SFPI Heating and Ventilation System 3.2.8.1.1Design Bases The Reactor Building and SFPI heating and ven tilation systems are operated to maintain a temperature above freezing within the areas of that building.The systems also maintain a slightly negative pressure when compared to the outside atmosphere. This is performed to ensure that there will be no inadvertent unmonitored release to the site area from the reactor building.Water vapor from quiescent eva poration of liquid waste may be released into the ventilation system. The process allows only the distillate vapor into the ventilation system, assuring positive control over the species and concentration of radionuclides released with Reactor Building exhaust air. MPS-1 DSAR3.2-12Rev. 8Ventilating air flow is routed to areas of progressively greater ra dioactive contamination prior to final exhaust. Back-draft dampers are provided to prevent reverse flow between areas of different contamination potential. Filtering of supply air is provided to reduce the presence of dust particles. Reactor Building main supply and exhaust units consist of fan, motor, and their associated controls. The SFPI system includes supply and exha ust fans installed in modular units.3.2.8.1.2System Description The Reactor Building and SFPI HVAC systems provide for the protection of personnel and equipment from airborne radioact ive contaminants and excessive thermal conditions. Air flow is directed to areas of progressively greater radioactive contamination prior to exhaust. The Reactor Building is provided with supply and exhaust ventilation to ensure proper air flow direction and remove heat generated from equipment. The SFPI system includes variable speed supply a nd exhaust fans to maintain space temperature within acceptable limits while also maintaining a negative pressure within the SFPI envelope relative to the outside and to Reactor Bu ilding areas outside the SFPI envelope.A flow diagram of the Reactor Building HVAC system is given in Figure 3.2-12. The SFPI HVAC system is shown in Figure 3.2-6.Reactor Building HVAC The supply segment of the system provides fresh ai r to all levels in the Reactor Building outside the SFPI envelope. Outside air passes through fixed louvers, a damper, filters, and electric heating coils. One fan is available to deliver air flow. Electric unit heaters are provided inside the drywell for freeze protection. Exhaust air flow combines in a common duct and continues on to the main exhaust fan plenum. System components, in addition to those mentioned above, include screens, filters, ductwork with dampers, supply outlets, return and exhaust inta kes, heating coils, and instrumentation and controls. Control actuation, indication, and alarm instrumentation are incorporated in a central HVAC master control panel.SFPI HVAC System The supply segment of the system provides fresh air to the operating floor of the Reactor Building, portions of the 82 feet 9 inches elevation and the spent fuel pool pump area. Outside air passes through fixed louvers in the side of the re actor building wall, filter s, and electric heating coils. A single variable speed 100% capacity fan is available to deliver air flow. MPS-1 DSAR3.2-13Rev. 8A single variable speed exhaust fan discharges air from the SFPI envelope through a HEPA filter, and fixed louver in the reactor building wall. The exhaust fan is operated in conjunction with the supply fan to maintain space temperatures within acceptable limits while also maintaining a slight negative pressure in the SFPI envelope relative to the outside and to Reactor Building areas outside the SFPI envelope. Electric unit heaters are installed in all SFPI areas to maintain acceptable space temperatures. System components, in addition to those ment ioned above, include duc twork with dampers, supply outlets, return and exhaust intakes, and instrumentation and controls. Control actuation, indication, and alarm instrumentat ion are incorporated in a local control panel. Indication and alarm functions are provided in th e Millstone Unit 2 Control Room. Natural ventilation cooling capability is also provide d by opening the Reactor Building truck bay doors, equipment hatch garage doors and the to rnado dampers located on the Reactor Building roof. This path would be used fo llowing an extended loss of all spent fuel pool cooling capability.3.2.8.1.3Safety Evaluation

The Reactor Building and SFPI heating and ve ntilation systems maintain environmental conditions in building spaces (t o support personnel comfort or ope ration of equipment located on those spaces), direct ventilation air from areas of low radioact ive contamination to areas of progressively greater c ontamination (to minimize the spread of contamination), and vent potentially contaminated exhaust air. Natural ventilation cool ing capability is also provided for spent fuel pool cooling following an extended lo ss of all active pool cooling capability. The Reactor Building and SFPI heating and ventilation systems are not safety related, but are required in the permanently defueled condition because they house SSCs that are associated with the safe storage and handling of irradiat ed fuel or radioactive waste.3.2.8.2Radwaste Building Ventilation System 3.2.8.2.1Design Bases The Radwaste Building ventilation system operates to supply filtered air to this building's areas.Supply air is filtered. The presence of dust particles potentially increases the spread of radioactive contamination.This system also filters the exhaust air prior to its discharge, to limit the release of any radioactive contaminants to the environment.Ventilating air flow is routed to areas of progressively greater radioact ive contamination potential prior to final exhaust. Back-draft dampers are pr ovided to prevent reverse flow between areas of different contamination potential. MPS-1 DSAR3.2-14Rev. 83.2.8.2.2System Description Figure 3.2-13 shows the ventilating flow through the Radwaste Building. The ventilating system is designed to provide a passive supply of filtered air and exhaust it after filtration. Air is drawn through the building by the main exhaust fa

n. An exhaust filter unit is provided.

Outside air is drawn into the system through tw o inlets above the roof of the building and protected by bird screening. Th e air is drawn through a filter de signed to remove dust. A header conveys fresh air to vari ous areas of the building. The fresh air supply is located in the clean areas of th e building while the inlets to the exhaust ducts are located where the rate of contamination is the highest. The exhaust air is passed through the filtering system before discharge through the main exhaust fan.3.2.8.2.3Safety Evaluation The Radwaste Building ventilat ion directs ventilation air fr om areas of low radioactive contamination to areas of progressively greate r contamination (to mini mize the spread of contamination), and vents poten tially contaminated exhaust air. The Radwaste Building ventilation system is only requi red, in the permanently defueled condition, to support personnel access to the space.3.2.8.3Intentionally Deleted 3.2.8.4Turbine Building Heating and Ventilation 3.2.8.4.1Design Bases The Turbine Building ventilation system is operated to maintain a slight negative pressure in the building to prevent any radioactive out-leakage, as well as, to provide fresh air to support personnel access. 3.2.8.4.2System Description Fresh air is supplied to the Turbine Build ing through louvers in the walls and roof. The ventilation system is arranged with one supplementary transfer fan and connecting ductwork to induce flow to the north end of elevation 14 feet 6 inches.The Turbine Building exhaust system collects air fr om various areas into an exhaust air header then discharges it into a plenum which also re ceives air from the Reactor Building and Liquid Radwaste Building. One exhaust fan is furnished to handle the combined exhaust from these three buildings. This fan discharges into a duct which runs along the north wall of the Reactor Building before releasing the exhaust air to the environment. Potentially contaminated areas in the Turbine MPS-1 DSAR3.2-15Rev. 8 Building are maintained at a nega tive pressure by exhausting from these areas. The exhaust air is drawn from adjacent spaces. This arrangement controls the air flow pattern and prevents out leakage.The Turbine Building ventilation air is normally discharged to the atmosphere without treatment.A flow diagram of the Turbine Building area ventilation system is shown in Figure 3.2-7.3.2.8.4.3Safety Evaluation The Turbine Building ventilation system directs ventilation air from ar eas of low radioactive contamination to areas of progressively greate r contamination (to mini mize the spread of contamination), and vents potentially contaminated exhaust air. The Turbine Building ventilation system is only required, in the permanently defueled condition, to support personnel access to the space. 3.2.9FIRE PROTECTION SYSTEMS The licensee's Nuclear Plant Fi re Protection Program has been developed to ensure that any single fire will not cause an unacceptable risk to public health and safety, and will not significantly increase the risk of radioactive release to the environment. A Fire Protection Program has been establishe d at Millstone Unit Number 1. This program establishes the fire pr otection policy for the protection of structures, systems, and components important to the defueled condition of the uni t and the procedures, equipment, and personnel required to implement the program.3.2.9.1Design Bases To achieve and maintain a high le vel of confidence for the Fire Protection Program, it has been organized and is administered us ing the defense-in-depth concept. The defense-in-depth concept assures that if any level of fire protection fails, another level is available to provide the required defense. In fire protection terms, this defense-in-depth concept consists of the following levels; *Preventing fire s from starting,*Early detection of fires that do start, and *Controlling and/or extinguishing them qui ckly so as to limit their damage. None of these levels can be pe rfect or complete, but strengtheni ng any one level can compensate in some measure for weaknesses , known or unknown, in the others. MPS-1 DSAR3.2-16Rev. 83.2.9.2System Description 3.2.9.2.1Site Water Supply System The underground fire protection water supply system consists of a 12 inch cast and ductile iron, cement-lined pipe extending ar ound Millstone Unit Number 1, 2, and 3 in a loop arrangement. The supply system services individually valved lines feeding fixed pipe water suppression systems (sprinklers, waterspray, and standpipes) throughout the plant and hydrants located around the exterior of the plant. The Millstone Unit Number 2 and 3 fire pumphous es contain three, 2,000 gpm at 100 psi, fire pumps which supply the yard loops; two with elect ric-motor drives and one with diesel engine drive. The Millstone Unit Number 3 pumphouse contains one electric dr iven pump (M7-8), fed from Millstone Unit Number 3 power, and the diesel-driven fire pump (M7-7). The Millstone Unit 2 pumphouse contains one electric driven pump (P-82) fed from Unit 2 power. All three pumps have individual connections to the underground supply system. Maximum system flow and pressure requirements can be met with any one of the three pumps out of service.System operation is such that a 50 gpm electric jockey pump (M7-11) maintains system pressure by automatically starting when line pressure drops to 105 psig and will run until pressure reaches 120 psig as indicated by a line pressure switch. A hydro-pneumatic tank is provided in the system to prevent short cycling of the jockey pump. At pressures below 105 psig, the MP2 P-82 electric pump first starts at 98 psig to maintain system pressure and flow. The Millstone Unit Number 3 M7-8 electric pump then will start at 85 psig and it is fed 480 VAC from MCC-CD-6 (MCC number 22A-2 Compartment number 1A). This pump is auto-started by a pressure switch set at 85 psig decreasing, while the M7-7 diesel-driven fire pump is auto-started by a separate pressure switch set at 75 psig decreasing. The diesel pum p is started by its ow n self-contained battery system. A battery charger is provided for rechargi ng. Both Millstone Unit Number 3 electric and diesel-driven fire pumps deliver 2000 gpm at 100 psi discharge pressure and remain in operation until they are manually shut down. Electrical inte rlocks stop the jockey pump when either of the two Millstone Unit Number 3 fire pumps start. The fire pumps are supplied from two 250,000 gallon ground level tanks. The tanks are automatically filled through a water line fed from city water.If a major fire in any location of the MP-1 site should occur, the combined water tank and makeup water capacity would provide an adequate water supply for MP-

1. The necessary pressure and flow would be maintained through the use of any two simultaneously operating 2,000 gpm rated pumps.3.2.9.2.2Fixed Suppre ssion Systems The fire protection features for the Unit 2 C ontrol Room are discussed in Section 9.10 of the Millstone Unit 2 Final Sa fety Analysis Report.

MPS-1 DSAR3.2-17Rev. 8(1)Sprinkler and Waterspray Systems The fixed water suppression systems for the "cold and dark" stage of the decommissioned unit are designed as follows:*Wet Pipe Automatic Sprinkler System (Maintenance Shop/Central Monitoring Station (CMS) Sprinkler System)*Automatic Deluge Waterspray System (ESST Deluge System)*Dry Pipe Manual Sprinkler Systems (Condenser Bay, Turbine Building Truck Unloading Area, and Reactor Buildi ng Rail Airlock Sprinkler Systems) The design concept for the fixed fire wate r suppression systems will use automatic operating systems for the heated plant area (Maintenance Shop/CMS) and the ESST located outside the east wall of the Main tenance Shop. For the unheated plant areas, a manual actuation concept will be used. The design will be to operate with "dry pipes in the unheated areas (Turbine, Reactor, and Radwaste Buil dings) and "flood up" the piping systems to activate the suppression system by opening a single isolation valve in the Maintenance Shop (Valve 1-Fire-37). This valve will be acces sible to the plant operators or responding fire department members outside the fire areas being protected by the dry pipes.The sprinkler systems and deluge waterspr ay system have been designed using the guidance of the National Fire Protection Association (NFPA) Standard Number 13 for the "Installation of Sprinkler Systems" or NFPA Standard Number 15 for "Waterspray Fixed Systems." The dry manual operating concept is not in conformance with NFPA but has been determined to be acceptable for the hazards of the decommissioned plant.(2)Wet Pipe Automatic Operating Sprinkler System An automatic, closed head, wet pipe design sprinkler system has been provided for the Maintenance Shop/Central Monitoring Station (CMS) area. This system has an alarm check valve which actuates an electric pressure switch to transmit a waterflow signal to the PLC. The system is provided with an outside screw and yoke (OS&Y) isolation valve between the supply connection and the system distribution piping. Sprinkler heads are closed, heat actuated type sprinkler heads.(3)Automatic Operating Deluge Waterspray Systems An automatic, open head, deluge type wate rspray system has been provided for the Emergency Station Services Transformer (ESS T). This system has a deluge valve that actuates upon an input from a heat detection circuit located around the transformer. Upon actuation, an electric alarm switch actuates and transmits an actuation signal to the PLC and water flows into the distribution piping and discharges from all open spray heads. The system has an OS&Y isolation valve lo cated between the s upply header and the MPS-1 DSAR3.2-18Rev. 8 distribution piping. Automatic operation is initiated by a single zone heat detection circuit installed in the hazard area. Manual operation of the automatic deluge system is provided via a mechanical pushbutton operator located on the deluge valve in the Maintenance Shop Welding area.(4)Dry Pipe Manual Sprinkler Systems Three sprinkler systems are provided in the unheated portion of the facility. These systems protect the Condenser Bay, the Turbine Building Truck Unlo ading Area, and the Reactor Building Rail Airlock. Sprinkler systems in the unheated portion of the plant are operated as dry pipe manual sprinkler systems. Each system has an isolation valve that separates the system from the supply header. The systems ha ve closed fusible type sprinkler heads. There is no waterflow alarm provided. System piping has been arranged to facilitate complete draining during cold weather conditions. These systems would be charged with water by manually opening isolation valve 1-Fire-37 located in the Maintenance Shop Welding Area as part of a fire fighting strategy for the facility. 3.2.9.2.3Portable Suppre ssion Capabilities (1)Hose Stream Coverage Hose stream coverage is available to all fire areas of the plant from stand pipe connections to fixed 1.5 inch hose stations or by use of 2.5 inch diameter hose with gated wye connections available from outside hose houses. The hose stations in the Maintenance Shop/CM S area are fed by the "wet" header piping and are available for immediate fire suppression use. The hose stations in the Turbine Building, Reactor Building, and Liquid Radwaste Building are fed off of the "dry" fire water header and will be av ailable for fire fighting follo wing the flood-up of the header following the opening of valve 1-Fire-37 in the Maintenance Shop. Hose stations in the Solid Radwaste Building are fed directly off a connection to the yard fire main and are maintained wet with heat tracing on the piping and valves to prevent freezing in this unheated area. Hose station locations are show n in the FHA (Reference 3.2-19).(2)Portable Extinguishers Selection and placement of portable fire extinguishers are in accordance with the intent of the guidelines of NFPA Standard Number 10, "Standard for Portable Fire Extinguishers". All extinguishers utilized are Underw riters Laboratories (UL) listed. MPS-1 DSAR3.2-19Rev. 83.2.9.2.4Fire Detection and Alarm Systems The fire detection and alarm sy stems installed in the plant are designed in general compliance with NFPA Standard Number 72D, "Standard fo r the Installation, Maintenance, and Use of Proprietary Protective Signaling Systems," and with NFPA Standard Number 72, "National Fire Alarm Code."Fire detection systems are used for early warn ing detection and in some cases may have the capability to actuate fixe d fire suppression systems. Detection devices consist of fi xed temperature detectors and sm oke detectors. Smoke detectors are of the spot type, employing the ionization principle. Specific a pplication of these detectors in each fire area is detailed in the FHA (Reference 3.2-19). In general, the installation of de tector units is in acc ordance with the intent of the guidelines set forth in NFPA Standard Number 72E, "Standard on Automatic Fire Detectors". Fire/smoke detectors, as with waterflow indicators, and valve tamper devices are arranged to transmit signals to local alarm pa nels and a fixed suppression system control panel, if applicable. Actuation signals are also transm itted through the local alarm panels to control panels in the Central Monitoring Station (CMS). A Fire Alarm panel located in the CMS monitors those areas necessary to support the Spent Fuel Pool Island. Trouble signals fo r these devices are transmitted in a similar manner. A general alarm is provided in the Unit 2 Control Room. Identification of the exact alarm or trouble signals must be performed locally in the Unit 1 CMS.The alarm system also monitors other mis cellaneous fire protec tion system features.3.2.9.2.5Ventilation Systems and Smoke Removal Removal of the products of combustion from a ny specific plant area requires the use of the normal plant ventilation system, which is desi gned to handle the expected normal environment within a given area or the use of portable exha ust fans by the fire brigade. There are no cable tunnels, culverts, or other unvent ilated areas that pose any specia l venting problems. Removal of gaseous radioactive waste either from plant proce sses or airborne particul ates requires the use of charcoal filters. The ventilation and filtration syst ems of potential radiation release areas are discussed in detail for the Reactor, Turbine, Radwaste, Radwaste storage, and Screenhouse Buildings in the FHA, Reference 3.2-19.3.2.9.3Safety Evaluation and Fire Hazards Analysis 3.2.9.3.1Evaluation Criteria An evaluation of the overall Fire Prot ection Program as indicated by the FHA, MPS-1 DSAR3.2-20Rev. 8 (Reference 3.2-19), found that the program does provi de reasonable assurance that a fire will not cause an unacceptable risk to the public health and safety. The fire protection program accomplishes this by assuri ng a fire will not significantly increa se the risk of radioactive release to the environment. Therefore, the Fire Prot ection Program meets the basic requirements of General Design Criteria 3 and 5 as applicable to a permanently defueled facility. Branch Technical Position (BTP) APCS B 9.5.1, "Guidelines for Fire Pr otection for Nuclear Power Plants," provides the implementi ng criteria for GDC 3 and gives the general guidelines used to review Millstone Unit Number 1. BTP APCSB 9.5.1 provides the guidelines acceptable to the NRC staff for implementing the following criteria:a.General Design Criterion 3 (10CFR 50, Appendix A) - Fire Protection.b.Defense-in-Depth Criterion: For each fire hazard, a suitable combination of fire prevention, fire detection and suppression capability, and abil ity to withstand safely the effects of a fire is provide

d. Both equipment and procedural aspects of each are considered.c.Single-Failure Criterion: No single active failure shall result in complete loss of protection of both the primary (fix installed systems) and backup fire suppression capability (standpipe/extinguishers).d.Fire Suppression System Capacity and Capability: Fire suppression capability is provided, with capacity adequate to extinguish any fire that can credibly occur and have adverse effects on equipment and components important to safety.e.Backup Fire Suppression Capability: Total reliance for fire protection is not placed on a single automatic fire suppression system. Appr opriate backup fire s uppression capability is provided in the form of portable fire extinguishers or hose stations.In addition to the specific guidance of the BTP, the evaluation considered the adequacy of the Fire Protection Program on the effects of potential fire hazards throughout the plant based on sound fire protection engineeri ng practices and judgments.3.2.9.3.2Fire Hazard Analysis Methodology Fire Protection was evaluated by conducting a fire hazard analysis of individual fire areas and fire zones within the plant. The analysis methodol ogy is described in the Fire Hazards Analysis (Reference 3.2-19). 3.2.9.3.3Fire Hazard Analyses Results The fire hazards analysis results for each fire area are contained in th e FHA (Reference 3.2-19).

MPS-1 DSAR3.2-21Rev. 83.2.9.4Inspection and Testing Administrative controls are provided through existing Plant Administrative Procedures, Operating Procedures and the Quality Assurance Program to ensure that the Fire Protection Program and equipment is properly maintaine

d. This includes QA audits of the program implementation, conduct of periodic test inspec tions, and remedial actions for systems and barriers out of service.

The technical requirements found in Millstone Unit Number 1 Technical Requirements Manual describe the limiting condition for operation and surveillance requi rements for the fire protection system. These technical requirements ensure the fire protection system is properly maintained and operated.All fire protection equipment and systems are s ubject to periodic inspections and tests in accordance with the intent of National Fi re Codes and the Fire Protection Program. The following fire protection fe atures will be subjected to periodic tests and inspections:(1)Fire alarm and detection systems(2)Wet pipe automatic sprinkler systems (3)Water spray systems(4)Interior fire water supply headers(5)Fire pumps (6)Fire barriers (walls, fire doors, penetration seals, fire dampers)(7)Manual suppression (fire hoses, hydrants, extinguishers) Equipment out of service includi ng fire suppression, detection, a nd barriers will be controlled through the administrative program and appropriate remedial actions taken. The program requires all impairments to fire protection systems to be identified and appropriate notification given to the Site Fire Marshal for evaluation. As conditions warrant, remedial actions would include compensatory measures to ensure an adequate level of fire protection in addition to timely efforts to effect repairs and restore equipment to service. MPS-1 DSAR3.2-22Rev. 83.2.9.5Personnel Qualification and Testing 3.2.9.5.1Fire Protection Organization The officer responsible for the Fire Protection Progr am at Millstone Unit Number 1 is defined in the QAP. Formulation, and assessment of the effectiveness of the program are delegated as indicated in Reference 3.2-20, the Fire Protection Program Manual.3.2.9.5.2Fire Brigade and Training The Site Fire Brigade and Nuclear Training are a site (Units 1, 2, and 3) organizations. The Millstone Site Fire Brigade consists of a mini mum of a Shift Leader and four Fire Brigade personnel. MP-2 supplies an advisor, who is at a minimum a fully qualified Unit 1 Plant Equipment Operator, to the Fire Brigade Shift Leader. The advisor will provide direction and support concerning plant operations and priorities.Members of the Fire Brigade are trained by the Nuclear Training Department. Site Fire Brigade personnel are responsible for respo nding to all fires, fire alarms, and fire drills. To ensure availability, a minimum of a Shift Leader and four Fire Brigade personnel remain in the owner controlled area and do not engage in any acti vity which would require a relief in order to respond to a fire (e.g., c ontinuous fire watch). If assistance is needed to fight a fire, additional equipment and manpower is supplied by the off site local fire departments. Wi thin a 5 mile radius of the plan t there are numerous local volunteer fire companies. Letters of commitment to supply public fire department assistance have been obtained from these fire companies.The Shift Leader coordinates the Site Fire Brigade activities, and ensu res proper communications and coordination of support for the local fire department chief or officer in charge once on site, and other on site activities (e.g., Chemistry, Health Physics, and Security).Nuclear Training coordinates with the Site Fire Marshal and periodically familiarizes local fire department personnel with the Station's layout and fire fighting equipment. The Site Fire Marshal coordinates with the Site Fire Brigade Personnel and all Unit Shift Managers, informing them of the status of the site fire protection equi pment, should equipment become inoperable or unavailable. Fire Protection drills are planned and critiqued by Nuclear Training and members of the management staff responsible for plant fire pr otection. Performance defi ciencies of the Fire Brigade or of individual Fire Brigade personnel are remedied by scheduling additional training for the Site Fire Brigade or individuals. MPS-1 DSAR3.2-23Rev. 83.2.9.5.3Quality Assurance The QA Program has been applied via the Fire Protection Program Manual to the FPSs which provide a function fo r the operating units. 3.2.10REFERENCES 3.2-1Docket Number 50-245, LS05-82-03-060, J. Shea to W.G. Counsil, 'SEP Topic IX-1, Fuel Storage (Millstone 1)," March 9, 1982.3.2-2Docket Number 50-245, B10301, W.G. Counsil to D.M. Crutchfield, 'Millstone Nuclear Power Station, Unit Number 1, SEP Topic IX-1, Fuel Storage,' August 31, 1981.3.2-3Docket Number 50-245, B10346, W.G. Counsil to D.M. Crutchfield, 'Millstone Nuclear Power Station, Unit Number 1, SEP Topic IX-1, Fuel Storage,' December 14, 1981.3.2-4Docket Number 50-245, B12961, M.L. Boyle to E.J. Mroczka, 'Millstone Nuclear Power Station, Unit Number 1, Issuance of Amendment Number 40 (TAC No. 68157),' November 27, 1989.3.2-5Docket Number 50-245, A08680, M.L. Boyle to E.J. Mroczka, 'Millstone Nuclear Power Station, Unit Number 1, Issuance of Amendment Number 43 (TAC No. 72183)," March 30, 1990.3.2-6Docket Number 50-245, J.W. Andersen to J.F. Opeka, 'Millstone Nuclear Power Station, Unit Number 1, Issuance of Amendment Number 89 (TAC No. M93080)," November 9, 1995.3.2-7Reference deleted. 3.2-8J.A. Price (Dominion) letter to U.S. NRC, "Millstone Power Station, Unit Number 1, Docket Number 50-245, Fuel Storage Requirements, Technical Specification 4.2", Letter Number B18972, dated Sept. 18, 2003.3.2-9Holtec Report Number H1 -971914, Revision 1, "Analysis Of 1675 Pound Fuel Assembly System Drop Onto The Irradiated Fuel Assembly." 3.2-10Holtec Report Number AH1-971691, Revision 0, "Criticality Safety Analysis Of The MP1 Racks With A Dropped Fuel Assembly." 3.2-11Holtec Report Number H1-971698, Revision 0, "Flow And Temperature Field Analysis Of Localized Cell Blockage In The Millst one Unit Number 1 Spent Fuel Pool." 3.2-12Holtec Report Number H1-971675, Revision 1, "Analysis Of Tetrabor And Boraflex Racks Under 1675 Pound Fuel Assembly System Impact." MPS-1 DSAR3.2-24Rev. 83.2-13Holtec Report Number H1-98210, Revision 1, "Decay Heat Load Calculation for the Millstone Unit 1 Spent Fuel Pool." 3.2-14Holtec Report Number H1-992125, Revision 0, "Steady State Temperature of Millstone Unit 1 SFP and RB with No Active SFP Cooling." 3.2-15Docket Number 50-245, B10291, W.G. Counsil to D.M. Crutchfield, "Millstone Nuclear Power Station, Unit Number 1, SEP Topic IX-3, Station Service and Cooling Water Systems," November 24, 1981.3.2-16Docket Number 50-245, NUREG-0824, Integrat ed Plant Safety Asse ssment, Systematic Evaluation Program, Millstone Nuclear Power Station, Un it Number 1, February 1983, Topic IX-3, "Station Service and Cooling Water Systems."3.2-17Docket Number 50-245, B10292, W.G. Counsil to D.M. Crutchfield "Millstone Nuclear Power Station, Unit Number 1, SEP Topic IX-5, Ventilation Sy stems", November 19, 1981.3.2-18Docket Number 50-245, LS05-82-09-043, J. Shea to W.G. Counsil, SEP Topic IX-5, Ventilation Systems, Millstone Nuclear Power Station, Unit Number 1, September 14, 1982.3.2-19Fire Hazard Analysis Millstone Unit Number 1, Revision 6, July 2000.3.2-20Millstone Nuclear Power Station Fire Protection Program Manual. MPS-1 DSAR4.1-1Rev. 2.1CHAPTER 4 - RADIOACTIVE WASTE MANAGEMENT4.1SOURCE TERMS With the permanent defueled condition of Unit 1, fission, co rrosion, and activation products from operation are no longer produced. The radioactive inventory that remains is primarily attributable to activated reactor components and structural materials and residual radioactivity. The accumulation of small amounts of solid waste as evaporator bottoms or contaminated materials may easily be controlled. Unit 1 no longer has routine liquid effluent re leases. Future planned liquid effluent releases will be evaluated pr ior to release, and appropriate controls (e.g., monitoring) will be established. The Radiological Effluent Monitoring and Offsite Dose Calculation Manual ensures that Unit 1 complies with 10CFR50, Appendix I. MPS-1 DSAR4.2-1Rev. 54.2RADIATION PROTECTION DESIGN FEATURES4.2.1FACILITY DESIGN FEATURES Radiation shielding was provided to restrict radiation emanating from various sources throughout the plant. The primary objective of radiation shie lding is to minimize th e radiation exposure of plant personnel and the general public. Millstone Unit Number 1 is permanently shut down and many installed components which are provided with shielding, are no longer required to safely store irradiated fuel. However, many of these installed components continue to contain radioactive material or remain radioactive themselves. Shielding that was originally desi gned to shield these components while they supported reactor operation, cont inues to provide sh ielding from the resi dual activity in the permanently shutdown condition.With the vessel in a drained down condition, a c oncrete shielding package is installed over the reactor vessel head and reactor cavity floor to provide shielding from activated reactor vessel internals.4.2.1.1Design Basis Normal operating conditions determined the major portion of the original plant shielding design requirements. Two exceptions to this were the Control Room wh ere shielding was determined by radiation levels produced during the loss-of-coolant accident a nd the shutdown cooling system where shielding was determined by shutdown c onditions. Although these conditions are no longer applicable, these were the bases for the unit shielding.4.2.1.2Ventilation Information on ventilation system s is contained in Chapter 3.4.2.2RADIATION PROTECTION PROGRAM 4.2.2.1Organization The radiation protection program is established to provide an effective means of radiation protection for permanent and tem porary employees and for visitors at the station. The radiation protection program is de veloped and implemented through the applicable guidance of Regulatory Guides 8.2, Revision 0; 8.8, Revision 3; and 8.10 Revision 1. The radiation protection department and line function management implement and enforce the radiation protection program.The officer responsible for implementing the radiation protection program is defined in the QAP. MPS-1 DSAR4.2-2Rev. 5 The radiation protection manager meets or exceeds the qualifications fo r radiation protection manager in Regulatory Guide 1.8, Revision 1. Radiat ion protection technicians meet or exceed the qualifications specified in ANSI N18.1-1971. MPS-1 DSAR4.3-1Rev. 24.3ALARA PROGRAM4.3.1POLICY CONSIDERATIONS It is the policy of the licensee to maintain i ndividual and plant personnel total radiation exposure ALARA. The licensee's ALARA policy complies with 10CFR20 and 10CFR50.4.3.1.1Design Considerations The basic objective of faci lity radiation shielding is to reduce external dose to plant personnel in conjunction with a program of radiologically controlled personnel access and occupancy in radiation areas to levels whic h are both ALARA and within the regulations defined in 10CFR20. With the reactor shutdown and all fuel stored in the spent fuel pool, the number and magnitude of potential radiation sources have been reduced substantially fr om the original bases for the radiation protection design features.4.3.1.2Operational Considerations Radiation surveys have been performed and will co ntinue to be performed to ensure that plant areas are correctly posted and barricaded. MPS-1 DSAR4.4-1Rev. 3.44.4LIQUID WASTE MANAGEMENT SYSTEMS Liquid waste from the Reactor Building drain system and some othe r liquid wastes, such as Unit 1 Radwaste Building sumps, Unit 1 Containment su mps, and Units 2 and 3 miscellaneous similar waste are collected in the Reactor Building sumps. Liquid Waste collected in the Radwaste Building sumps is pumped to the Reactor Building sumps. The liquid waste collected in the "A" reactor building sump is pumped to a staging tank. Collected liquids may be surveyed for activity and pumped to an atmospheric evaporator. The distillate vapor will be diluted in the Reactor Building exhaust and released as a ground leve l release. Radiological monitoring will be conducted by a particulate monitor in the ventilation exhaust or by screening a grab sample of the process liquid. Concentrates in the bottom of the Reactor Building atmospheric evaporator will be collected as required, and disposed as Low Specific Activity (LSA) trash. Alternatively, this system could be utilized to pump the proces s liquids from the Reactor Building sumps to containers which would perm it the process liquid to be processed onsite or offsite. MPS-1 DSAR4.5-1Rev. 2.14.5SOLID WASTE MANAGEMENT The plant has no capability for processing concentr ated waste solutions to a solidified product. Dry Activated Waste (DAW) is processed and stored in appropriate containers to allow for offsite shipment.Interim on site storage facilities accept waste from Millstone Units 1, 2 and 3. Information regarding facility design criteria is presented in Section 11.4 of th e Millstone Unit 3 Final Safety Analysis Report.4.5.1DESIGN BASES

The design basis objective of solid waste management is to provide for processing, packaging and handling solid dry wastes, and to allow for radioactive decay and/ or temporary storage prior to shipment off site and subsequent disposal. Solid radwaste handling at Millst one Unit 1 ensures compliance with the following regulations and Regulatory Guides:(1)10CFR20, Standards for Protection Against Radiation (2)10CFR50, Appendix I (3)10CFR61.55, Classification of Wast e for Near Surface Disposal (4)10CF61.56, Waste Characteristics (5)10CFR71, Quality Assurance Criteria for Sh ipping Packages of Radioactive Material (6)Regulatory Guide 1.143, Design Guidance for Radioactive Waste Ma nagement Systems, Structures and Components (7)Regulatory Guide 8.8, ALARA Provisions 4.5.2SYSTEM DESCRIPTION

The solid waste management process is designe d to accommodate the following radioactive wastes, which are typical for BWR power plants: Dry active wastes, which consist of contaminated clothing, tools and small pieces of equipment that cannot be economically decontaminated; miscellaneous paper, rags, etc., from contaminated areas; air filters from ra dioactive ventilation syst ems; used reactor equipment such as control rod blades, temporary control curtains , fuel channels and in-core ion chambers - Radioactivity levels of most DAW are low enough to permit handling by contact, it is proc essed and stored in appropriate containers to allow for off site ship ment. Used radioactive e quipment may be stored MPS-1 DSAR4.5-2Rev. 2.1for sufficient time to permit decay before removal for interim storage or off site shipment. Equipment too large to be handled as described above is handled on a case-by-case basis with special procedures. Concentrated bottoms from the waste evapor ator may be shipped as low specific activity (LSA) dry active waste. Summaries of solid waste shipme nts, types, volumes, and radionuc lide composition are given in Reference 4.5-1.4.

5.3REFERENCES

4.5-1Millstone Nuclear Power Station Unit Number 1, Docket Number 50-245, Annual Radioactive Effluents Report. MPS-1 DSAR4.6-1Rev. 2.14.6EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING 4.6.1DESIGN 4.6.1.1Design Basis The effluent radiation monitoring system (R MS) provides nonsafety related functions. The system provides the means for compliance with Nuclear Regulatory Commission (NRC) regulations 10CFR20, 10CFR 50 Appendix A Ge neral Design Criteria (GDC) 60, 63 and 64,10CFR 50, Appendix I and Regulat ory Guides (RG) 1.21, 4.15 and 8.8.4.6.1.2System Design Description SFPI Ventilation Exhaust Monitor The SFPI ventilation exhaust radiation monitor is designed with the capability to monitor, indicate and record the discharge of gaseous radioactivity. Capability fo r sampling of particulate activity is provided. Annun ciation in the Millstone Unit 2 Cont rol Room occurs if setpoints are exceeded.Although the monitor cannot determine the individua l activity level of the radionuclides in the effluent gas, it provides the overall level and a ba sis for correlation with laboratory analyses of filter and grab sample activities. The SFPI gas sample is taken from the exhaust duct near the reactor building exhaust plenum. A single point sample nozzle is posi tioned to obtain a representative sample of the turbulent and well mixed exhaust air. The monitor is located in a heated enclosure on th e 65 foot elevation of the Reactor Building directly below the exhaust duct. The sample passes through a particulate filter and a shielded detection ch amber (fixed volume) and is then expelled back into the exhaust duct. The particulate filters ar e periodically removed for detail ed radiological quantitative analysis.The detector readout is sent to the PLC for di splay and recording. The range of indication is 1 x 10-6 µci/cc to 1 x 10 0 µci/cc (Kr-85).BOP Ventilation Exhaust Monitor The Unit Number 1 BOP ventilation exhaust fl ow is continuously sampled for radioactive particulates. The sample is taken from the exhaust duct which runs along the north exterior wall of the Reactor Building. A single point sample nozzle is positioned to obtain a representative sample of well mixed exhaust air. The part iculate sample skid is located in an insulated enclosure on the 65 foot elevation, north wall, of the Reactor Building. The sample passes through a particulate filter and is then expelled back into the exhaust duct. The particulate filter is pe riodically removed for detailed radiological quantitative analysis. MPS-1 DSAR4.6-2Rev. 2.14.6.2AREA RADIATION MONITORING INSTRUMENTATION 4.6.2.1Design Bases The purpose of the ARM system is to warn of abnormal radiation levels in the SFPI where radioactive material may be present, stored, handled, or inadvertently introduced. The system also provides information concer ning radiation at selected locations in the SFPI.4.6.2.2System Description

The area radiation monitoring system detects, measures, and indicates ambient gamma radiation dose rates at selected locations in the SFPI. It pr ovides audible and visual alarms in the Millstone Unit 2 Control Room (locally at some locations) when radiation levels exceed pre-selected values or when a monitor has operational failure. Table 4.6-2 lists the area radiation monitor locations and ranges. Refueling Floor Area Radiation Monitor The refueling floor ARM is a 3 ch annel digital unit. Each detector is a gama sensitive GM tube located as described in Table 4.6-2. Each channel is provided with a failsafe High, Warn and Failure alarm relay as well as an analog output. The alarms and analog output are sent to the PLC for recording and alarm. Each unit has a built in check source and local a udible and visual alarm indication.4.6.3REFERENCE4.6-1Letter from W.G. Counsil to D.G. Eise nhut dated July 1, 1981, "Haddam Neck Plant, Millstone Nuclear Power Station, Unit Numbers 1 and 2, Post TMI Requirements - Response to NUREG-0737," Dock et Numbers 50-213, 50-245, 50-336. MPS-1 DSAR Page 1 of 1 Rev. 2TABLE 4.6-1EFFLUENT RADIATION MONITORS MonitorDetectorRangeTrip FunctionSFPI ventilation exhaust(1) Beta Sinctillator 10-6 to 10 0 µci/cc None MPS-1 DSAR Page 1 of 1 Rev. 2TABLE 4.6-2AREA RADIATI ON MONITORING SY STEM SENSOR AND CONVERTER LOCATIONS FOR MILLSTONE UNIT NO. 1 REACTOR BUILDINGStation NumberSENSOR AND CONVERTER LOCATIONRange mR/hrRM-SFPI-01 CH1West Refuel Floor 0.01-10 2RM-SFPI-01 CH2East Refuel Floor 0.01-10 2RM-SFPI-01-CH3West Refuel Floor Hi Range 10.0-10 6 MPS-1 DSAR5.1-1Rev. 2CHAPTER 5 -ACCIDENT ANALYSIS

5.1INTRODUCTION

In July of 1998, the licensee certified to the NRC that Millstone Unit Number 1 had both permanently ceased operations and that all fuel had been removed from the reactor vessel and placed in the spent fuel pool (Reference 5.1-1). Si nce Millstone Unit Number 1 will never again enter any operational mode, reactor related accidents are no longer a possibility. The remaining analyzed accident that is in this chapter is the fuel handling accident. Conservatism in equipment design, conformance to high standard s of material and cons truction, the control of mechanical and pressure loads, and strict admini strative control over plant operations all serve to assure the integrity of the fuel in the spent fuel pool. New hazards, new initiators, and new accidents that may challenge offsite guideline exposures, may be introduced as a result of certain d ecommissioning activities. These issues will be evaluated when the scope and type of th e decommissioning activities are finalized.5.1.1ACCIDENT EVENT EVALUATION 5.1.1.1Unacceptable Results for De sign Basis Accidents (DBAs) The following are consider ed to be unacceptable safety results for DBAs:(1)Radioactive material release that results in dose levels that exceed the guideline values of 10CFR100.(2)Nuclear system stresses in excess of those allowed for the accident classification by applicable industry codes.(3)Radiation exposure to plant operations personnel in the Millstone Unit 2 Control Room in excess of 5 REM whole body, 30 REM inhalation, and 75 REM skin.5.1.1.2Fuel Handling Acci dent Assumptions Fuel handling accident analysis assumptions are listed on Table 5.2-1.5.1.1.3Results The results of the Fuel Handling Accident an alytical evaluation are provided in Section 5.2.5.1.1.4Radiological Consequences

Consequences of radioactivity release duri ng a fuel handling accident are presented in Section5.2. MPS-1 DSAR5.1-2Rev. 25.

1.2REFERENCES

5.1-1"Millstone Nuclear Power Station, Unit Number 1 Certification of Permanent Cessation of Power Operations and that Fuel has been Permanently Removed from the Reactor," July 21, 1998. MPS-1 DSAR5.2-1Rev. 25.2FUEL HANDLING ACCIDENT As the bounding accident analysis, an inadvertent release of radioactivity, as a result of a fuel handling accident in the spent fuel pool, was evaluated and is discussed below.With the permanent cessation of operations of Mi llstone Unit Number 1, the prior limiting fuel handling accident, i.e., a fuel assembly drop onto the top of the core during fuel-handling operations, was no longer part of the plant's de sign and licensing basis. Several fuel handling accident scenarios are s till possible in the spent fuel pool. Thes e scenarios are identified later in this Section. he radiological consequences of a fuel handling accident in the sp ent fuel pool are described in this section. For conservatism, a bounding analysis was made to ca lculate the radiological release from a failure of all fuel rods in four (4) fuel assemblies in the spent fuel pool. Other assumptions taken into consideration are described later in this Section. The off site radiological consequences of this release, i.e., from 4 failed fuel assemblies or, for exam ple, 248 fuel rods for 8x8 fuel assemblies, are substantially less than the 10CFR Part 100 limits and are tabulated in this section.5.2.1FUEL HANDLING ACCIDENT SCEN ARIOS IN THE SPENT FUEL POOL The consequences of the following postulate d fuel handling drop events were evaluated:*Spent fuel pool gate (1200 lbs.) drop onto irra diated fuel and fuel storage racks in the spent fuel pool.*New fuel assembly drop (600 lbs.) onto irradi ated fuel and fuel storage racks in the spent fuel pool.*Lifting of a Tri-Nuc Filter sk id (965 lbs.) into the spent fu el pool and potential drop onto irradiated fuel and fuel storage racks.*Postulated drop of items (pumps, boxes, filters, stellite containers and tables) temporarily stored on the spent fuel pool equipment rail onto irradiated fuel and fuel storage racks.*Drop of an irradiated fuel assembly onto other irradiated fuel in the spent fuel pool.These analyses utilized two sophisticated elasto-plastic finite-element models. The fi rst represents the fuel assembly components, the second repres ents the rack with its pedestals, liner and underlying reinforced concrete struct ure. The LS-DYNA3D computer code (Reference 5.2-1) was used. Cons ervative assumptions and restri ctive inputs were utilized to result in an upper bound estimate of the calcul ated damage for the postulated drop event. The following assumptions were utilized in the analysis: MPS-1 DSAR5.2-2Rev. 2 Regarding the impactor movement and the target:*Both the impactor and the target are submerged.*The target is in a stationary position prior to impact.*The trajectory of the impactor is vertical.*The form drag force opposed to the impactor movement is proportional to it's velocity squared.*The friction drag force is conservatively neglected.

Regarding the impact mechanism transmission:

*The impactor makes first contact with the fuel assembly handle which is located above the rack elevation. Furthermore, the handle is c onservatively considered as a prefect rigid body, without deformability or energy absorption capacity.

Regarding failure criteria:

*Failure of an individual fuel rod is assumed to occur when the irradiated zircaloy material reaches its postulated failure stress (strain). For additional conservatism, the entire length of each fuel rod is assumed irradiated to the state where the brittle material behavior is active.*Overstress of the lower guide ends (between the lower end of the fu el rod and the bottom fitting) is not cons idered as a failure of the supported rod.The analysis of these additional accident scenarios has determined that the limiting event is the drop of the spent fuel pool gate, which can result in extensive damage of the fuel assemblies, showing a total of 54 ruptured fuel rods. The drop of the new fuel assembly resulted in damage to the targeted fuel assemblies, but no ruptured fuel rods were recorded for either the impactor or the target. Drop of an irradiated fuel assembly results in failure of all 64 guide ends, but no rupture of fuel rods occurs. These results bounded all fuel t ypes stored within the Millstone Unit Number 1 spent fuel pool for the analyses performed to date.5.2.2RADIOLOGICAL CONSEQUENCES Since the licensee has certified to the NRC that there is a permanent cessation of operations of Millstone Unit Number 1 and that fuel has been permanently removed from the reactor vessel, a calculation evaluating the radiological consequences of a fuel handling accident in the spent fuel pool was performed and eventually chosen as the new bounding accident (Reference 5.2-2). Taking into account the actual source term of the fuel in the spent fuel pool (i.e., appropriate decay time of fuel), the rean alysis assumed four fuel asse mblies (e.g., 248 rods in an 8x8 assembly) failed in the spent fuel pool and resulted in an unfiltered, i.e., no Standby Gas MPS-1 DSAR5.2-3Rev. 2Treatment System (SGTS) availa ble and secondary containment not set, puff release. Additional assumptions and input parameters are given in Table 5.2-1. This reanalysis was performed using the guidelines of Standard Review Plan 15.7.4 Rev. 1 and Regulatory Guide 1.25. Doses were calculated using the TACT-III, OR IGEN-2, and ELISA computer codes.The results of this dose assessment for 4 failed fuel assemblies revealed the following radiological dose data:

Thyroid dose at the exclusion area boundary 5.44E-04 REM Thyroid dose at the low-population zone 1.69E-05 REM Whole-body dose (calculated as TEDE) at the exclusi on area boundary 1.03E-03 REM Whole-body dose (calculated as TEDE) at the low-popul ation zone 3.20E-05 REMThese doses are well within the limits of 10CFR100, and are therefore acceptable. Doses were also calculated to the Millstone Un it Number 2 Control Room. The results of this dose assessment is as follows: Thyroid dose to the Millstone Unit Number 2 Control Room 7.65E-02 REM Whole-body dose to (calcu lated as TEDE) the Millstone Unit Number 2 Control Room 8.67E-02 REM Beta skin dose to the Millstone Unit Number 2 Control Room 2.19E+01 REM These doses are less than the limits specifie d in GDC 19. Doses were not calculated for the Millstone Unit Number 3 control room since the atmos pheric dispersion factor (/Q) is approximately 50 times less that the (/Q) to the Millstone Unit Number 2 control room.Therefore, the dose to the Millstone Unit Numb er 3 control room would be approximately 50 times less than the Millstone Unit Number 2 control room dose.5.

2.3REFERENCES

5.2-1LS-DYNA3D, Version 932, Livermore Software Technology Corporation, May 1, 1995.5.2-2Calculation Package NUC-197, "MP1 Defueled State - Radiological Analysis of a Fuel Handling Accident," Duke Engineering and Services, October 11, 1999. MPS-1 DSARPage 1 of 1Rev. 2TABLE 5.2-1ASSUMPTIONS AND INPUT CO NDITIONS FOR FUEL HANDLING ACCI DENT AT MILLSTONE UNIT NO. 1 Assumption Basis 1. Core Power Level During Irradiation = 2011 MWtTechnical Specifications 2. Varied to identify conservative results based on actual burnup. Regulatory Guide 1.25 See Ref. 5.2-3.3. Varied to identify conservative results based on actual burnupRegulatory Guide 1.25 See Ref. 5.2-3.4. Pool Scrubbing Factor = 60Extra polation of Regulatory Guide 1.25 DF to MP1 conditions. See Ref. 5.2-3.

5. Chemical form of Iodine above pool:Regulatory Guide 1.25 See Ref. 5.2-3.*85 percent Elemental*15 percent Organic

6. Number of Assemblies in Core: 580Technical Specifications

7. For radiological dose asse ssment: Number of fuel assemblies assumed to fail = 4DSAR Section 5.2.28. Release fractions from fuel rods: Regulatory Guide 1.25

& conservative assumption*30 percent Noble Gases*12 percent Iodines9. No credit taken for secondary containmentTechnical Specifications*SGTS not in operation*Puff release is an unf iltered ground release

10. Breathing rate = 3.47 x 10

-4 m 3/sec Regulatory Guide 1.2511. Ground level dispersion factor (/Q): EAB (0-2 hr.) = 6.10 x 10 -4 sec/m 3 LPZ (0-4 hr.) = 1.90 x 10 -5 sec/m 3SEP Topic 11-2.c, Docket Number 50-24512. Decay Time for fuel = 3.8 yearsBased on the MP1 shutdown on November 4, 1995. MPS-1 DSAR6.1-1Rev. 3.2CHAPTER 6 -CONDUCT OF OPERATIONS6.1ORGANIZATIONAL STRUCTURE Information regarding the organizational structur e is presented in Section 1.0 of the Quality Assurance Program Description Topical Report (Reference 6.1-1). With the exception given below, that information is in corporated herein by reference.The owner, holding 100 percent of the Millston e Unit Number 1 nuclea r plant, is Dominion Nuclear Connecticut, Inc..6.1.1MANAGEMENT AND TECHNICAL SUPPORT ORGANIZATION Information regarding the management and technical support organiza tion is presented in Section1.0 of Reference 6.1-1. That informat ion is incorporated herein by reference.6.1.1.1Technical Support for Operations Information regarding the technical support fo r operations is presented in Section 1.0 of Reference 6.1-1. That information is incorporated herein by reference.6.1.1.2Organizational Arrangement Information regarding the organizational ar rangement is presente d in Section 1.0 of Reference6.1-1. That information is incorporated herein by reference. 6.1.2OPERATING ORGANIZATION 6.1.2.1Plant Organization The plant organization is as shown in Reference 6.1-1.6.1.2.2Plant Personnel Responsibilities and Authorities

Information regarding the plant personnel responsibil ities and authorities is presented in Section 1.0 of Reference 6.1-1. That information is incorporated herein by reference.6.1.2.3Operating Shift Crews The minimum shift crew composition is contained in the Administrative Controls section of the Millstone Unit Number 1 Te chnical Specifications. MPS-1 DSAR6.1-2Rev. 3.26.1.3QUALIFICATIONS OF NUC LEAR PLANT PERSONNEL 6.1.3.1Qualification Requirements Qualifications of plant manage rial and supervisory personnel ar e established by the American National Standards Institute (ANSI) N18.1 (Reference 6.1-2) except for the following:a.The Operations Manager or Assistant Op erations Manager shall be a Certified Fuel Handler.b.The Radiation Protection Manager shall meet or exceed the qualifications of Regulatory Guide 1.8, Rev. 1.6.

1.4REFERENCES

6.1-1Quality Assurance Program Description Topical Report. 6.1-2American National Standard s Institute, ANSI N 18.1-1971, Selection and Training of Nuclear Power Plant Personnel. MPS-1 DSAR6.2-1Rev. 2.16.2TECHNICAL SPECIFICATIONS Technical Specifications set fort h the limits, operating conditions and other requirements for the protection of the health and safety of the public. These specifications have been written in fulfillment of 10 CFR 50.36 and are controlled pursuant to 10CFR50.90, 50.91, and 50.92. Technical Specifications are maintained as Appendix A to the operating license. The Technical Requirements Manual (TRM) contains clarifications for certain technical specifications and a central location for other documents which place operating limits on the plant. Changes to the TRM are controlled pursuant to the 10CFR50.59 process. MPS-1 DSAR6.3-1Rev. 3.26.3PROGRAMS 6.3.1TRAINING Programs are credited to train plant personnel. Key technical operating personnel receive onsite classroom or guided self study and on-the-job training. Appropriate pl ant personnel receive instruction in emergency plan a nd radiation protection procedures. Specialized training in specific areas conducted by the equipment manufacturers or other vendors is utilized as necessary. Training on a continuing basis is used to maintain a high level of proficiency in the staff.6.3.2EMERGENCY PLAN The staff approved Millstone Nuclear Power Station Emergency Plan (Reference 6.3-1) addresses the criteria set forth in NURE G-0654, FEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Prep aredness in Support of Nuclear Power Plants, Revision 1, November 1980 and NUREG-0737, Supplement 1. As such, the Emergency Plan provides for an acceptable state of emergency preparedness and meets the requirements of 10CFRPart 50 and Appendix E thereto.6.3.3PHYSICAL SECURITY PLANS The security plan (Reference 6.3-

2) states the securi ty measures to be employed by the licensee for the protection of Units 1, 2 and 3 at the Millstone Nuclear Power Station, Waterford, Connecticut, against radiological sabotage. The plans have been submitted in accordance with 10CFR Part 73, Section 73.55, "Require ments for Physical Protection of Licensed Activities in Nuclear Power Reactors Against Radiological Sabotage."These plans include measures to deter or prevent malicious actions that could result in the release of radioactive materials into the environmen t though sabotage. This protection is provided through the use of armed guards, physical barrie rs, monitors, personnel ac cess controls alarms, communications, response to security conti ngencies, and liaison with appropriate law enforcement agencies.6.3.4QUALITY ASSURANCE PROGRAM DESCRIPTION (QAPD) TOPICAL REPORT The licensee has developed and implemented a comprehensive Quality Assurance Program (QAP) to ensure conformance with establishe d regulatory requirements as set forth by the Nuclear Regulatory Commission, and accepted industry standards. The participants in the QAP assure that the design, procurement, construction, testing, operation, maintenance, repair, and decommissioning of nuclear power plants are performed in a safe and effective manner.The QAPD Topical Report complies with the requirements set forth in Appendix B of 10CFR Part 50, along with applicable secti ons of the Safety Analysis Report.

MPS-1 DSAR6.3-2Rev. 3.2 This QAP is also established, maintained and executed with regard to Radioactive Material Transport Packages as allowed by 10CFR71.101(f). In addition, the QAPD Topical Report is submitted periodically to the NRC in accordance with 10CFR50.54(a).6.

3.5REFERENCES

6.3-1J. F. Opeka letter to U.S. Nuclear Re gulatory Commission Document Control Desk transmitting "Revision 6 to the Millstone Nuclear Power Station, Unit Numbers 1, 2, and 3, Emergency Plan," dated November 4, 1991 [and subsequent revisions thereto submitted on an annual basis].6.3-2J. F. Opeka letter to U.S. Nuclear Re gulatory Commission, "Millstone Nuclear Power Station, Unit Numbers 1, 2, and 3, Physical Se curity Plan, Revision 15," dated December 16, 1991 and subsequent revisions thereto. MPS-1 DSAR6.4-1Rev. 26.4PROCEDURES Written procedures are required for maintenance, repair, or operational activities related to the structures, systems and components which are safety related (Safety Class 1,2, or 3). Written procedures shall be established, implemented, and maintained in accordance with the Technical Specifications. MPS-1 DSAR6.5-1Rev. 3.26.5REVIEW AND AUDIT A program describing the review a nd audit of activities important to and affecting station safety, has been established and complies with Re gulatory Guide (RG) 1.33, "Quality Assurance Program Requirements (Operation) ." The program provides a system to ensure that these activities are performed in accordance with company policy, ru les, and approved procedures.6.5.1ONSITE REVIEW The membership, duties, areas of review responsibility, and requirements of both the plant and site operations review committees are described in the Quality Assurance Pr ogram Description (QAPD) Topical Report (Reference 6.1-1).6.5.2INDEPENDENT REVIEW Independent review of activities affecting the unit's safety is performed by the Management Safety Review Committee as described in the QAPD Topical Report (Reference 6.1-1).6.5.3AUDITS The audit program for activities af fecting safety related systems, structures, or components is as described in the QAPD Topical Report (Reference 6.1-1). MPS-1 DSAR7.1-1Rev. 5CHAPTER 7 -DECOMMISSIONING7.1

SUMMARY

OF ACTIVITIES Millstone Unit Number 1 was shutdown for a normal refueling outage on November 4, 1995, and has not operated since. On November 19, 1995, transf er of all fuel assemblies from the reactor vessel into the spent fuel pool (SFP) for st orage was completed. On July 17, 1998, the licensee decided to permanently cease further operation of the plant. Certification to the NRC of the permanent cessation of operation and permanent re moval of fuel from the reactor vessel, in accordance with 10CFR50.82 (a)(1)(i) & (ii), wa s filed on July 21, 1998 (Reference 7.1-1), at which time the 10CFR50 license no longer author ized operation of the reactor or placement of fuel in the reactor vessel.The mission of the licensee is to decommission the plant safely and in a cost effective manner. The information contained in this section of the DSAR is based upon the best information currently available. The plans discussed herein may be modified as additional information becomes available or conditions change. Specific conditions which are unique to the multi-unit Millstone Station require that certain Millstone Unit Number 1 decom missioning activities be delayed and performed concurrently with the decommissioning of Millstone Unit Numb ers 2 and 3. Other cons iderations may dictate early scheduling of certain decommissioning activities. Th erefore, the approach to decommissioning Millstone Unit Number 1 can best be described as a modified SAFSTOR. In this approach, decontamination and dismantlement activities may be undertaken early in the decommissioning wherever it makes sense from a safety or economic viewpoint. For instance, given the future uncertainty over access to a low level waste disposal si te, early shipment of certain components will occur. The amount of decommissioning work completed prior to a SAFSTOR period depends upon a number of factors currently under evaluation.Both the DECON and the SAFSTOR options are approaches found acceptable to the NRC in its Final Generic Environmental Impact Statement (GEIS) (Reference 7.1-2). Completion of the decommissioning schedul e is contingent upon three key factors:*continued access to licensed low level waste (LLW) disposal sites,*removal of spent fuel from the site, and*timely funding of the d ecommissioning activities. Currently Millstone Unit Number 1 has access to Chem-Nuclear Systems' Barnwell, S.C. disposal site and to the Envirocare disposal site in Tooele County, Utah. Escalation costs for the disposal of waste have been incorporated into financial planning. Additionally, the licensee has considered the possibility that during the decontamination and dismantlement phases, access to the Barnwell low level waste disposal site could be denied or that the facility could be closed. MPS-1 DSAR7.1-2Rev. 5 The unavailability of the DOE high level waste repository may affect th e decontamination and dismantlement schedule for Millstone Unit Number

1. Delays in the operation of the repository have resulted in a significant increase in the cost of decomm issioning and, may require the use of an independent spent fuel storage installation (ISFSI).

Although storage of the Millstone Unit Number 1 sp ent fuel in an ISFSI is presented in this DSAR as an option; an ISFSI ha s been contracted to ensure th e continued operation of Millstone Unit Numbers. 2 and 3. Currently, after spent nucle ar fuel is removed from the Unit 2 and Unit 3 reactor core; it is safely stored in the existing SFPs. Capacity of these pools was designed with the assumption the DOE high level waste repository would provide permanent storage. However, the site selection, cons truction and licensing of such a repositor y have been delayed. As is the case with other nuclear facilities as the SFPs approach full capacity, spent fuel from Millstone Unit Numbers 2 and 3 will be stored in the ISFSI. A de scription of the ISFSI is contained in the Unit Number 3 Final Safety Analysis Report. Under any eventuality such as unavailability of a LLW disposal site, temporary shortfall in decommissioning funding, or other unforeseen circumstances, 10CFR50.82 requires the licensee to maintain the capability to suspend decontamination and dismantlement.7.1.1DECOMMISSIONING APPROACH The licensee is planning on decommissioning Millstone Unit Number 1 using a modified SAFSTOR approach in which the decontamin ation and dismantlement of the systems, components, plant structures and facilities (i.e., DECON) are comp leted prior to and following a SAFSTOR period. In this plan, an ISFSI may be c onstructed and the transfer of spent fuel from the spent fuel pool (SFP) could be completed during the SAFSTOR period. The SAFSTOR period ends with decontamination and dismantlement of any remaining systems, structures, and components commence in coordination with Mills tone Unit Number 2 and Millstone Unit Number 3 decommissioning. Spent fuel shipments from the ISFSI to DOE ar e scheduled, when practicable, following the repository commencing operations. Delays in the operation of the repository limits the transfer of fuel and increases the cost of long term spent fuel storage. The following discussion provides an outline of the current decommissioning plan activities completed to date and the remaining significan t activities. The planning required for each decommissioning activity, including the selecti on of the process to perform the work, is completed prior to the start of work for that activity.7.1.1.1Planning The planning includes implementation of a site characterization pl an, preparation of a detailed decommissioning plan, and the engi neering development of task work packages. The detailed engineering required to support th e decontamination and dismantlem ent of systems, structures, and components are performed prior to the start of field activities. MPS-1 DSAR7.1-3Rev. 5 Significant activites performed to date include:*Establishment of a spent fuel pool island.*Sent fuel pool cleanup.

• Removal and disposal of leg acy resins and filter media.*Removal, processing, and disposal of irradiated hardware from the reactor vessel including control rod blades and in-core instrumentation.*Reactor vessel internals segmentatio n, including the upper core grid.*Drain down of the reactor cavity and reactor vessel.
• Installation of a radiation shielding package over the reactor vessel head and cavity floor.The following activities remain:
• Evaluate and choose a dry fuel storage system , if pursued. Investigat e and prepare for the design and licensing of an ISFSI and prepare procurement specifications for a fuel canister system and ancillary equipment.7.1.1.2Site Characterization

During the initial portion of the planning period a detailed site characterization was undertaken during which radiological, regulated and hazar dous wastes were identified, categorized, and quantified. Surveys were conducted to establish the contamination and radiation levels throughout the Millstone Unit Number 1 portion of the site. This information is used in developing procedures to ensure that ha zardous, regulated or radiologically contaminated materials are removed and to ensure that worker exposure is maintained as low as reasonably achievable (ALARA). Selected surveys of the outdoor areas in the vicinity of Millstone Unit Number 1 may be performed, although a detailed survey of the environs would likely be deferred pending decommissioning of Millstone Unit Numbers 2 a nd 3. It is worthwhile to note that site characterization is a process that continues throughout decommissioning. As decontamination and dismantlement work proceed, surveys are conducted to maintain current ch aracterization and that decommissioning activities are adjusted accordingly. The activation analysis of the reactor internals, the reactor vessel, and the biological shield wall was undertaken as a part of the site characteri zation. Using the results of this analysis, these components were classified in accordance with 10CFR61 and form the basis for the detailed plans for their packaging and disposal. The interi or grid portion of the top guide structure was determined to be greater than class "C" (GTCC) material, was segmented from the reactor vessel, and is stored in the spent fuel pool in canister s sized to be compatible with ISFI dry storage containers.7.1.1.3Decontamination

The objectives of the decontamination effort are two fold. First, to reduce the radiation levels throughout the facility in order to minimize personnel exposure during dismantlement. Second, to clean as much material as possi ble to unrestricted use levels, th ereby permitting non radiological MPS-1 DSAR7.1-4Rev. 5 demolition and minimizing the quantities of materi al that must be disposed of by burial as radioactive waste. The need to decontaminate structures, systems, and components are determined by the schedule to dismantle them and by plant conditions. Earl y dismantling of contaminated components and systems may benefit from decontamination activ ities by reducing the radiation exposure to the workforce. Late dismantling may not require the components and systems to be decontaminated since the decay of the radiation sources reduces the radiation levels by significant amounts.Chemical decontamination of the reactor recirculation system may provide value through reduced worker dose. An evaluation is performed to determine whether the expected reduction in the accumulated workforce exposure is justified by the costs associated with the decontamination. The evaluation results are sensitive to the amount and type of work to be performed prior to a SAFSTOR period. Any decontamin ation method used employs esta blished processes with well-understood chemical interact ions. The resulting waste is dispos ed of in accordance with plant procedures and applicable regulations.The second objective of the decontamination effo rt is achieved by decontaminating structural components including steel framing and concrete surfaces. The method used to accomplish this is mechanical and requires the removal of the surf ace or surface coating. This process is used regularly in industrial and contaminated sites. 7.1.1.4Major Decommis sioning ActivitiesAs defined in 10CFR50.2 a "major decommissioni ng activity" is "any activity that results in permanent removal of major radioactive component s, permanently modify the structure of the containment, or results in dismantling com ponents for shipment containing GTCC waste in accordance with 10CFR61.55."Major decommissioning activities completed to date include the removal of the drywall head and removal of the reactor vessel internals by segmentation. The drywall head was sectioned and sent to a metal processor. The reactor vessel internals, classified as GTCC, are limited to the interior portion of the top guide structure, which has been segmented from the reactor vessel and is stored in the spent fuel pool. The reactor cavity and reactor vessel have been drained. Without the GTCC internals present, several options are available for later removal and disposal of the reactor vessel: segmentation, sectioning into pieces, or disposal as an intact package. Based on an evaluation of activity levels, ease of execution, personnel exposure, schedule constraints, disposal facility availability, and cost, segmentation of the internals may be postponed until after the fuel is removed from the SFP.Removal of the reactor vessel follows the removal of the reactor internals and may not occur until after a SAFSTOR period. It is likely that th e vessel would be removed by sectioning or segmenting. Vessel sectioni ng or segmenting permits a substantial portion of the waste to be sent to a waste re-processor instead of a near surface disposal site. The dismantling of the drywell and suppression chamber is undertaken as pa rt of the reactor building demolition. MPS-1 DSAR7.1-5Rev. 57.1.1.5Other Decommissioning ActivitiesOther decommissioning activities include:*Preparation and submittal of the following documents:1.A license termination plan pursuant to 10CFR50.82 2.A spent fuel management program, pursuant to 10CFR50.54(bb)In addition to the major decommissioning activit ies listed above, the following decommissioning activities include:*Hazardous and regulated material s (e.g., asbestos, lead, mercury, PCBs, oil, chemicals) are identified during characterization and plan s are developed for the removal of these materials.*Plant components removed from the Turbine Building include the Turbine Generator, Condenser, Feedwater Heaters, Moisture Se parators and miscellaneous system and support equipment.*Miscellaneous solid waste re moved include: control rod blades, local power range monitors, spent resins and filters, the Reactor Pressure Vessel Head Insulation assembly, the de-tensioner platform, and the Refuel Floor shield plugs. The larger components may be segmented and packaged for removal through the Reactor Building hatchway.*Liquid wastes are processed and discharged using plant procedures in accordance with applicable regulatory requirements as the liquid waste inventories become available. Initially the inventories of the plant water systems are processed. Upon completion of the segmentation and packaging of the reactor vessel internals, the reactor cavity and reactor may be drained and the waste inventory proc essed. When the spent fuel is removed, the SFP is drained and the water processed. System s are then isolated and deactivated in a sequence compatible with the operations prev iously described. Spent fuel pool systems are isolated after removal of the spent fuel. Radioactively contaminated or activated materials are removed from the site as necessary to allow the site to be released for unrestricted access. Low level waste is processed in accordance with plant procedures and existing commercial options, and sent to licensed disposal facilities or waste processors for further volume reduction. Wastes may be incine rated, compacted, or otherwise processed by authorized and licen sed contractors, as appropriate. Mixed wastes, if any, are managed according to all applicable federal and state regulations. Mixed wastes are transported only by authorized and licensed transporters and shipped only to authorized and licensed facilities. MPS-1 DSAR7.1-6Rev. 57.1.1.6Final Site Survey and Termination of License Since Millstone Unit Number 1 and Millstone Unit Number 2 are contiguous and have common structural boundaries, the plans for building demolit ion and for the license termination survey are implemented as a coordinated evolution for the two units. Consequently, the schedule for the Millstone Unit Number 1 license te rmination is constrained by the need to terminate the Part 50 license coincident with that of Millstone Unit Nu mber 2. As a result of this delay in requesting license termination, the final site survey using Reference7.1-4 may proceed in two phases: 1) internal structures surveyed as decontamination and dismantlement are completed, and 2) external areas surveyed in conjunction with completion of the Unit 2 decontamin ation and dismantlement.The licensee is required to prepare a License Termination Plan (LTP) for Millstone Unit Number 1. The LTP defines the details of the final ra diological survey to be performed once the decontamination activities are completed. The LTP conforms to the format defined in Reference 7.1-5 and addresses the limits of 10CFR20 using the pathways analysis defined in Reference7.1-4. Use of this guidance ensures that survey design and implementation is conducted in a manner that provides a high degree of confidence that applicable NRC criteria are satisfied. Once the survey is complete, the results are provided to the NRC in a format that can be verified.7.1.1.7Site Restoration The restoration of the Millstone Unit Number 1 area of the Millstone site will be undertaken when the 10CFRPart 50 license for Millstone Unit Number 1 is terminated. This event may coincide with Millstone Unit Numbers 2 and 3 license terminations. Buildings, structures, and other facilities which ar e not currently known to be radiologically contaminated, such as the Strainer Pit, Intake Structure, and the Discharge Stru cture are dismantled, as part of the building demolition effort after the final license termin ation survey for Millstone Unit Number 1 is complete. These buildings can be removed late in the building demolition phase since there is no decommissioning operational need to remove them earlier. Site restoration requires that all buildings be removed to an elevat ion 3 feet below grade or to an elevation consistent with the removal of the necessary amount s of contaminated material.7.1.2STORAGE OF RADIOACTIVE WASTETable 5.4-1 of the GEIS (Reference 7.1-2) provides an estimate for low-level waste disposal from a referenced boiling water reactor (BWR) of 18,975 cubic meters (669,817 cubic feet) for both the DECON and SAFSTOR options. The licensee estimates the low-level waste burial volume for Millstone Unit No. 1, will be at or below this value for the modified SAFSTOR alternative. The licensee's estimate includes, by a reduction of approximately 40 per cent (industry standard), the utilization of present-day volume reduction techniques. For waste re quiring deep geological burial, i.e.,GTCC waste, the licensee estimates that the volume for Millstone Unit Number 1 is at or below the 11.5 cubic meters anticipated for a reference BWR discussed in Section 5.4 of the GEIS. These estimates support the conclusion th at the previously issued environmental statements are bounding since the disposal of wa ste requires fewer resources, i.e., less waste disposal facility area, than what was considered in the GEIS. MPS-1 DSAR7.1-7Rev. 57.1.2.1High Level Waste Congress passed the "Nuclear Waste Policy Act" in 1982, assigning the responsibility for disposal of spent nuclear fuel created by the commercial nuclear generating plants to DOE. This legislation also created a Nuclear Waste Fund to cover the co st of the program, which is funded, in part, by the sale of electricity from the Millstone Unit Number 1 plant. The current DOE estimate for startup of the federal waste management system is 2010. For planning purposes, the licensee has assumed that the high-level wast e repository or some interim storage facility will not be operational until at least 2010. Sh ipments of fuel and GTCC wast e to DOE are planned to be directly from the ISFSI.The spent fuel is currently stored in the SFP. The licensee may license a dry, ISFSI. Fuel will be transferred from the pool and stored temporarily on site using licensed canisters. For the period of time when the fuel will be stored in the SFP, the systems nece ssary for SFP operations will be consolidated into an "Island" concept a nd configured for SFP clean-up and cooling.7.1.2.2Low Level WasteRadioactively contaminated or activated materials are removed to allow the site to be released for unrestricted access. Low level waste is processed in accordance with federal and state regulations, plant procedures and existing comm ercial options, and transported to license disposal facilities. 7.1.2.3Waste Management A major component of the total co st of decommissioning Millstone Unit Number 1 is the cost of packaging and disposing of systems, components and structures, contamin ated soil, water and other plant process liquids. A waste management plan incorporates the most cost effective disposal strategy consistent with regulatory requirements for each waste type. The waste management plan will be base d on the evaluation of availabl e methods and strategies for processing, packaging, and transporting radioact ive waste in conjunction with the available disposal facility options and a ssociated waste acceptance criteria.7.1.3RADIATION EXPOSURE MONITORING

Personnel radiation exposure is maintained ALAR A and monitoring is conducted in accordance with the radiation protection program described in Chapter 4. Exposure specifically related to decommissioning activities is identified and track ed. Exposure monitoring is used to identify activities that are causing excessive exposure and to initiate corrective actions to reduce personnel exposure.7.

1.4REFERENCES

7.1-1Letter B17388 from Bruce D. Ke nyon to U. S. Nuclear Regulatory Commission,"Certification of Permanent Cessation of Power Operations and that Fuel Has Been Permanently Removed from the Reactor," dated July 21, 1999. MPS-1 DSAR7.1-8Rev. 57.1-2U. S. Nuclear Regulatory Commi ssion report NUREG-0586, "Final Generic Environmental Impact Statement on Decomm issioning of Nuclear Facilities," dated August, 1988.7.1-3Letter B17790 from R. P. Necci to U. S. Nuclear Regulatory Commission, "Post Shutdown Decommissioning Activiti es Report," dated June 14, 1999.7.1-4U. S. Nuclear Regulator y Commission report NUREG-1575, "Multi-Agency Radiation Site Survey and Investigation Manual (MARSSIM)," Final Report. 7.1-5U. S. Nuclear Regulatory Commission report NUREG-1700, "S tandard Review Plan for Evaluating Nuclear Power Reactor License Termination Plans," (Currently in Draft form). MPS-1 DSAR7.2-1Rev. 2.17.2ESTIMATE OF RADIATION EXPOSURE The decommissioning of Millstone Unit No. 1 is accomplished with no significant adverse environmental impacts, in that no Millstone Unit No. 1 site specific factors should alter the conclusions of the GEIS (Reference 7.1-2) or the Millstone Environmental Statement. The radiation dose to the public during decommi ssioning is typically mi nimal. Decommissioning workers receive a fraction of the dose which radiation workers receive in an operating plant. The low-level radioactive waste that is removed from the site occupies only a small portion of the burial volume at approved waste di sposal sites. The non-radiologi cal environmental impacts are temporary and not significant. The occupational dose exposure for decommissioning Millstone Unit No. 1 is less than described in the GEIS because of two main reasons. First, the licensee initiated a zinc injection program for Millstone Unit No. 1 in 1987 that significantly reduced the buil dup of contaminated corrosion products during the remaining plant operation pe riod. Second, with the plant shutdown since 1995, natural decay of leading radionuclides have reduced overall plan t general dose levels significantly by the time decontamination and decommissioning activities occur. The activities identified in this chapter rese mble the DECON option. Th erefore, the modified SAFSTOR occupational and public dose exposure is compared to the DECON option dose in the GEIS. The occupational and public dose effects for a modified SAFSTOR alternative is bounded by the DECON option. The exposure from decontamination and dism antlement activities and the exposure during transportation of the low-level wastes is included in this dose estimate. NUREG-0586 (Reference 7.1-2), Table 5.3-2, estimates a total occupational dose of 18.74 person-Sv (1874 person-rem) for the DECON alternative for the reference BWR plant. The values estimated by the licensee will be at or below this value.7.2.1NUCLEAR WORKER Detailed estimates for external occupationa l radiation exposure that accumulate dose for decommissioning workers during the dismantlemen t program are developed based on a task by task analysis of personnel hours and expected radi ation dose rates associated with each task. These estimates are based on the following:1.ALARA principles are implemented.2.Radiation exposure is monitored to id entify jobs that are causing excessive exposure and corrective actions are taken to reduce the severity.7.2.2GENERAL PUBLIC Radiation dose to the public is maintained below comparable levels when the plant was operating through the continued application of radiation protection and cont amination controls combined with the reduced source term available in the facility. MPS-1 DSAR7.2-2Rev. 2.17.2.3NORMAL TRANSPORTATION Shipments of spent fuel and radioactive wast es are performed by exclusive use vehicles. Shipments will be in accordance with the Department of Transporta tion (DOT) regulations. Generic industry estimates of the doses from routing transportati on of radioactive materials are based on the following assumptions:*Two truck drivers during a 500 mile trip would probably spend no more than 12 hours inside the cab and 1 hour outside the cab at an average distance of 6 feet from the truck.*Normal truck servicing en route would require that two garage men spend no more than 10 minutes about 6 feet from a shipment.*Onlookers from the general public might be e xposed to radiation when a truck stops for fuel or for the drivers to eat. The onlooker dos e is calculated on the basis that 10 people spend an average of 3 minutes each at a di stance of about 6 feet from a shipment.*The cumulative dose to the general public from truck shipments is based on population dose of 2.3 x 10 -6 man-rem per km.NUREG/CR-0672, Table 11.4-2, provided a generic esti mate of the routing radiation doses from truck transportation of radioactive wastes. Th e doses are based on the maximum allowable dose rates for each shipment in exclusive use trucks and are conservatively high, on the number of truck shipments, and on the shipping distances. Th e estimated external radiation dose for routing transportation operations is 110 man-rem to transportation workers and 10 man-rem to the general public.The licensee estimates the volume of both high leve l and low level wastes to be less than the volumes used in NUREG/CR-0672. The total number of shipments of radioactive wastes is less than those used to determine the exposure in the NUREG/CR, and therefore the exposure to the transportation workers and the general public is less than those identified above. MPS-1 DSAR7.3-1Rev. 2.17.3CONTROL OF RADIATION R ELEASES ASSOCIATED WITH DECOMMISSIONING EVENTS During the decommissioning, processes may concen trate source terms. Non-routine events may occur with the potential to disperse the source term. This section of the DSAR establishes controls and requirements to maintain potential consequences of such event to below analyzed accidents. 7.3.1IN PLANT EVENTS The DBA for Millstone Unit Number 1 is the fuel handling accident and a detailed discussion can be found in DSAR Chapter 5. The acceptance criteria for all other potential events at the plant are controlled such that the potential consequences of any postulated event are less than 1 REM at the exclusion area.7.3.2TRANSPORTATION ACCIDENTS Transportation accidents have a wi de range of severities. Most ac cidents occur at low speeds and have relatively minor consequen ces. In general, as speed incr ease, accident severity also increases. However, accident severity is not a f unction of vehicle speed only. Other factors, such as the type of accident , the equipment involved, and the location can have an important bearing on accident severity. Damage to a package in a transportation accident is not directly related to accident severity. In a series of accidents of the same severity, or in a single accident involving a number of packages, damage to packages may vary from none to exte nsive. In relatively minor accidents, serious damage to packages can occur from impacts on sharp objects or from being struck by other cargo. Conversely, even in very severe accident s, damage to packages may be minimal. The probabilities of truck accidents used in the NUREG/CR-0672 study were based on accident data supplied by the DOT. Accidents are classified into five categories as functions of speed and fire duration. The five categorie s in order of increasing severity are: minor, moderate, severe, extra severe, and extreme. Table N.5-3 of NUREG/CR-0672 pr ovides the probabilities of occurrence for each classification. Estimated accident frequencies, release amounts and radiation doses to the maximum exposed individuals for selected accidents for transportation of radioactive material are discussed in Appendix N.5.2.3 of NUREG/CR-0672. The frequencies are calculated by mu ltiplying the total distance of transport with the total probability of accident per distance traveled for each accident severity class. The maximum exposed individual is assumed to be located 100 meters from the point of a transportation accident. The cal culated dose values provided in Table N.5.6 of NUREG/CR-0672 are the first year dose and the fifty year dose commitment to the bone, lung, thyroid and whole body. MPS-1 DSAR7.3-2Rev. 2.1The licensee anticipates that site specific anal ysis on the expected num ber of shipments and the shipping distance will confirm its applicability to the ge neric analysis provided in NUREG/CR-0672. MPS-1 DSAR7.4-1Rev. 2.17.4NON-RADIOLOGICAL ENVIRONMENTAL IMPACTS The non-radiological environmental impacts from the Millstone Unit Number 1 decommissioning is temporary and not significant. The largest occupational risk associated with the decommissioning is the risk of industrial accidents. This risk is minimized by adherence to work controls during decommissioning similar to th e procedures followed during power operation. Procedures controlling work related to asbestos , lead, and other non-radi ological hazards remain in place during the decommissioning. The primary environmental effects of the decommissioning are temporary and include small increases in noise levels and dust in the immediate vicinity of the site, and small increases in truck traffic to and from the site for hauli ng equipment and waste. These effects are similar to those experienced dur ing normal refueling outa ges and certainly less severe than those present duri ng the original plant constructi on. No significant socioeconomic impacts or impacts to local cult ure, terrestrial or aquatic re sources have been identified.7.4.1ADDITIONAL CONSIDERATIONS While not quantitative, the follo wing considerations are also relevant to concluding that decommissioning activities do not result in signifi cant environmental impacts not previously reviewed:*The release of effluents continues to be controlled by plan t license require ments and plant operating procedures throughout the decommissioning.*With respect to radiological releases, Mill stone Unit No. 1 continues to operate in accordance with the Offsite Dose Calc ulation Manual during decommissioning.*Release of non-radiological effl uents continues to be controlled per the requirements of the NPDES and State of Connecticut permits.*Systems used to treat or control effluents during power operation are either maintained or replaced by temporary or mobile systems for the decommissioning activities.*Radiation protection principles used during plant operations remain in effect during decommissioning to ensure that protective techniques, clot hing, and breathing apparatus are used as appropriate.*Sufficient decontamination and source te rm reduction prior to dismantlement are performed to ensure that occupational dos e and public exposure do not exceed those estimated in the Final Generic Environmental Impact Statement (Reference 7.1-2.*Detailed site radiological surveys are perfor med prior to starting the waste campaigns to confirm the burial volume of low-level radioactive waste a nd highly activated components which require deep geological disposal.*Transport of radioactive waste is in accordance with plant pr ocedure, applicable Federal regulations, and the requirements of the receiving facility. MPS-1 DSAR7.4-2Rev. 2.1*Plant ventilation systems, or alternate, te mporary systems, are maintained as long as needed in areas they service.*Site access control during decommissioning ensures that residual contamination is minimized or eliminated as a radi ation release pathway to the public. MP$'l DSAR FTGURE 1.2-1 PLOT PLAN 3\* -r'\'..".\t -d&úFdà lu/ ffi 41@mV. i lmER rS' tz Il,', .1 'r-l'.1.-.d:,-....'.---: SYSTEM CATEGORIZATION- UEL LEGEND ELACK : UIASSESSEOELUE : VAILABLE R*D : ABANDCNEO. t:: .l {--- . í:i'ffrt^flsrr*_*{rqk_ \{ffi?c.ÉÉ G<-\-\3o

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LCn3 l ! -D Sosr?*rt[ ¿6tÉ cffifr sLsw i.ir ryE tra4t rut 4q- .' -_ ,t.. . f'.,? ¿,--l3 =- ll j 'i 2 F-B- t0 .Rev.3.3/ ,. :,< ./ , / / I \ ' " ............. ! ,.';" " .I ..... , \ " .-.,' 0' . ./ / ,.-." "-;. ... ... ." ........ \" ... .-., .. . .. ! \ MPS-l DSAR FIGURE 1.2-1 PLOT PLAN ---/ -----.--' , 'O""'}' ; .' . ...... " ;' .', / -r -----... --"'-1 14 I .----' SYSTEM CATEGORIZA nON PROCESS -DEC 1502 . LEGENO BLACK : UNASSESSEO BLUE : AVAILABLE REO : ABANOONED NCTE-S a:1!tl'Al Elttf_C COWIlH'f ... -.tfLLSt'OtIE NUCi.<':olR P(Nt£fc S':'AiqN KOT PLAH Rev, 3.3 MP$l DSAR FIGURE I.2 -2A GENERAL ARRANGEMENT RAD \ryASTE BUTLDINGS - PLANS f_-I ,, \d_a-AN ^bte ÉL.t4rø{.\)I4!--L'l Ii'.I I I L T OU:PIdENT L IST gJ FLooR PLN rr Er-.ar s 1 cl¿'-c RAOwge Ai,t^. tr! PLN , g3: (r u.REFERENC* ORATINCS m 0PIRAftoils cRilcAt.\ :- ij 1... , ..il-¿{.20 .,Qit'.,,;iL c FLOOR PLAN ATEL.14.6 gl Al+-r uD uAsrÉ çT4\éÉ. BJtLürJ6 F-E PLAN Al eL.t4-6OTES: r. Ei [çtrs e@tr4rr to*{ttrtcrtG ¿, f- ElE\ 4iltf *o d rnt3 m.rtsa;i6t^¿?.H"iJ itLF. ls(srtfs 252d{dr r.!. 6 dr?Ét5f* & sl^rtq..-tl srs rrn¿ ¿rur*.*.ft"rJn7 I -t"i-..1 lr7 tkut ill \l FOR 4RG1 d TIa AFEA sf,E M 6-t6!s-t q$6ll toAa El t offiif: s o*as.2g2bz- Aot+. 512t1 s t -.')l_ffi.Jffi T a c ts a(W'E.&¿alg 3faa aI I L L. ::.i---í-Rev.3.3. fLOOR PLAN AT £U-)/'-(O (.ll*1;; MP8-1 DSAR FIGURE 1.2 -2A GENERAL ARRANGEMENT RAD WASTE BUILDINGS -PLANS NOTES: 1. O£o.t':-(s ID[NT,FIC&tf('f. Z. ,,:lIt ECUIPlo'i'" IOF.KTIF lEO Ott THIS OR ..... :M'" IS :; SER .... 'CE. SE£ .u>Pt.IC'61E '1AO ...... STE liST cOIfG. SEAI(S2S2:)2-l60481. ,llr..R;ftGT. "'T\t1'50 owe. FLOOR PLAN AT ELl4'-G 17 Ie I'\..At.J AaOVE. Et...,4'*", I '. I I f l := -, .. .. I J.r.E4 A . ...:.,...-. R ... OW ..... Te. AAu..-l\QOf PIAN ....... -OPERA nONS CRilICAl .\ : J tI..;f*'o ..... i. *. "1)':./,:' ,:d.... ______ ** __ 11- .-.... _-_ .. --_.-_ ..... -. ___ es-PUHS ?!!_=:l Ci1lOi[S FIRE HOSE: STAtlON.-'d" ()(I4T<:S FIAE £'ltTltlCUl$UEA

• . ,,: __

__ Rev. 3.3 FLOOR PLAN AT EL.(-)¿O.O MPS-I DSAR FIGURE 1.2-28 GENERAL ARRANGEMENT RAD WASTE BUILDINGS - PLANS FLOOR PLAN AT EL.G) I .C C) ?:'C xciES:!- ffi ú6ES at{qr rdfflFlcf d r .d re4F drt.l& 3 tilg dtlÉ _ rrr t3 !r ú fthq

• frr<.48.ceJt sÉ[s [ol*r rf

Ñsr:al :a-¿sar.!. 6 dl?r(*r43.:4 ctE 5l.rro-.r< eenl rtc c.irørwe.BI AI--øAD WASrÉ *.1c.âÉ. BilLür 6 FLOOø PLN. A1 EL.r4-6.rl-rj.=+qff-F:* l r#" J q-u sa/*. eL.t4'-a l8ttf q.'c*cqød tL*.<fOR &RGNII AruÉtÉ ø-l5 Rev.3.3 FIGURE 1.2 -2B FLOOR PLAN AT EL'<-l(!O-O , ! MPS-l DSAR GENERAL ARRANGEMENT RAD WASTE BUILDINGS -PLANS .. -: ." ...... :;-'i. *. : NiHES: 1. O£NOTtS [OUIP'WE'IT 100tfTinCATlOII

l.

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l. C3!JttQtE5

.. SrIoTIClN.* tt 0£t;(lT['.i FlftE [ltlN;UISH£R. fOft 01'--nn" .... ftWUo. '5t.t 1>WGo G.-I679I!eo fLOOR PLAN AT El.I4'-6 ..... A . -: ___ Tr ________ U..;'-__ ...,. _____ -" ____ -,,..---____ -._-,-__ ...... __ --,:------ ..... -----"T------T"------"1I!lr'-------,1'"----Ii I.. Ii Ii '7 II 16 fIIJ "1:10).-:1.'\005 'Xc* .f;o. J. Rev. 3.3 MPS-I DSAR.;FTGURE 1.2 - 3 GENERAL ARR.NGEMENT BUILDINGS RAD WASTE BUILDINGS - SECTIONS:: fllto anRTIAL S*CTION crc (6. r.ror rs er s:^:É 16,i.,no1 E'øeÉ'¿l ¿ crra.ra¿ oq.i,d léaoq-tiód.:., 1 r--'l lQAl :- :' '\i.:.0etrllr0lls cllllcar.s*clroN J-Jancal-Ea ÉttÉatt-t Écn-ú-Rev.3.3 MPS-l DSAR FIGURE 1.2 -3 GENERAL ARRANGEMENT BUILDINGS RAD WASTE BUILDINGS -SECTIONS ____ _Jz£_ ____ ____ ______ ______ ... ______ .. R_ ______ B-6 z 3

• 7 i II 12 Ii ... I. Ii 17 Ii PA.RTlAL SECTION G-G (a*,&14101 (e,"1 1;r;" i'.'J* ---I *** 1J\S8UIlT 1'\00. ,., *. -( ... ,:7 ;J CAD XOTE' .:eo l g ..... Ut4ce. OIl4-.1.<df SEe oa. ...... o...,** "

... :' OP[RAIIONS CRIlfCAt -_ ................. . .... 1!1 ** ... _. tIC..-a.,.. _____ _ ----------- CO£HtAAI,. IRttAMO(W[NT flAD .-sTt kIIlJ)HOS-S(CTIQfIIS Rev. 3.3 MPS-I DSAR FIGURE 2.1_I GENER,A,L SITE LOCATION MNPS-1 DSAR c}f I ri3 5lci15 L A N O v\to" j dJ" 5to SCALE -MILES FTGURE 2.1-I GeneralSite Location Millstone Nuclear Power Station Rev. 2 September 1999 MPS-l DSAR FIGURE 2.1-1 GENERAL SITE LOCATION . -...J t;!!? O:x: 00:: . SITE \ ___ IS. ---------/?' . " _------0 ,,",PLUM IS. 8 (.. 0 C SOU N cot'_-'{ORI< 0 __ L. ft N IS 50 Miles o 5 10 :, J , MNPS-1DSAR 15 I SCALE -MILES FIGURE 2.1-1 General Site Location Millstone Nuclear Power Station September 1999 Rev. 2 MPS-I DSAR FIGURE 2.1_2 GENERAL VICINITY September 1999 I ) '. LON G BAY ISLAND MPS-l DSAR FIGURE 2.1-2 GENERAL VICINITY SOu NO MNPS-1DSAR o . , /

.* . : FIGURE 2.1-2 General Vicinity Millstone Nuclear power station September 1999 Rev. 2 I ) '. LON G BAY ISLAND MPS-l DSAR FIGURE 2.1-2 GENERAL VICINITY SOu NO MNPS-1DSAR o . , /

.* . : FIGURE 2.1-2 General Vicinity Millstone Nuclear power station September 1999 Rev. 2 MPS-I DSAR FTGURE 2.1-3 SITE LAYOUT*+_: tcr ntr to ElÉl ldÉat aa ltrttf ftat I JOROf COVE lara a BOP & SFPI Ventilation Exhaust LEGEN' {-r;m y#i,"ff*?.----

• • r**isc-*-F*ET zfifKffi a ffilnr qICT (rrll(t ttüddÉ, SCA.8-lrET*Er tttAtttc 8.ât Rev. 2.3 MPS-1 DSAR FIGURE 2.1-3 SITE LAYOUT JORDAN COVE ISFSIAREA 0 500 1000 I I J . SCALE-FEET 0 2S() SCI!) I SCALE-METERS Ventilation Exhaust NIANTIC BAY Rev. 2.3 MPS-I DSAR FIGURE 2.I_4 SITE PLAN zOrS4A68/- l craE!L*GFr{o FI4&B! PRrvrrLr oürrD DIIQttá n*cR*rlor. aRE!

I o 250 500 l- l lSCÂLE=FIET

• 'r7*'scRvrcE r tfR"udot Sts LOtttG t5 L A ttto soulv DRev. 2.3 MPS-l DSAR FIGURE 2.1-4 SITE PLAN . PR(VAT£L1' OWNED . R[CR(ATIOU AREA o 250 500 I ,

Rev. 2.3 MPS-I DSAR FIGURE 2.1_5 TOWNS WITHIN IO MILES LEOYARO f0 Miles.souilo MNPS-1 clf t t I I DSAR.EAST H AOOAM 1---úI LEGEND TOWN EOUNDRY COUTITY SOUNOARY STATE 8O{JIOARY sTc.{tNo TON i..oL 0 sarERooK.- ls\soÚH0 LONG SCALE- lllLES FIGURE 2.I-5 TOWNS WITHIN 10 MILES MILLSTONE NUCLEAR POWER STATION'lr OL.D L Y M E ISLATVD-â/r/ffi;;:V-W*September 1999 Rev. 2 MPS-l DSAR FIGURE 2.1-5 TOWNS WITHIN 10 MILES EAST HADDAM I \ 1 I \ I I , I lEAST \ __ J MONTVILLE , \ ---\ -\--:--D-\ I . \ L Y M E '\ W ATE R FOR D <;! \ ")) \ /'" / \ < \ __ L, r---\ _v l \ \ \ { \ \ \ o I , LONG ISLAND...- BAY / ( l HEW '\ LeNDOI. \. 'I \ \ \ I .SOUND 10 Miles i , -LEGEND MNPS-1DSAR Cf( t ____ TOWN BOUNDARY ___ COUNTY BOUNDARY ___ STATE BOUNDARY 024 I I I ! .J SCALE-MILES FIGURE 2.1-5 TOWNS WITHIN 10 MILES MILLSTONE NUCLEAR POWER STATION September 1999 Rev. 2 MPS-I DSAR FIGURE 2.I-6 POPULATION SECTORS FOR O - IO MILES I IIAsL I fi aooAfi _i I------1/'NNw-; LEGENO TOWN AOUNOARY I --- STATE AOUNoARY SCAL*- IILES FIGURE 2.1-6 Population Sectors for 0 - 10 Miles Millstone Nuclear Power Station Rev. 2 MOTTVILLE sourHoLo [-J,( 6'ánow*Rs ,4 September 1999 OLD SAYBROOK EAST HAODAM -----... MPS-l DSAR FIGURE 2.1-6 POPULATION SECTORS FOR 0 -10 MILES -I I I I I DR' N STO,NINGTON s MNPS-1DSAR LEGEND ____ TOWN BOUNDARY -___ COUNTY BOUNDARY ___ STATE BOUNDARY 024 til , I SCALE-MILES FIGURE 2'.1-6 Population Sectors for 0 -10 Miles Millstone Nuclear Power Station September 1999 Rev,2 NW MPS-I DSAR FIGURE 2.I-7 POPULATION SECTORS FOR O - 50 MILES ,50!/liles MNPS.I ct^t TûLEGEND:@UflTY Bo('TIOÂESSTATE 8üJNOARIES N" PoPuL,Âftol carER sû.n{ofY C'ENE 30 ESE SCAL*-lIILES FIGURE 2.I-7 Population Sectors for 0 - 50 Miles Millstone Nuclear Power Station SE September 1999 Rev. 2 MPS-l DSAR FIGURE 2.1-7 POPULATION SECTORS FOR 0 -50 MILES ENE E ssw s -MNPS-1DSAR ac t LEGEND: _-COUNTY BOUNDARIES ___ STATE BOUNDARIES o I f'()PU(.ATlON camR 8Q(JNOARY 5 I 10 I SCALE-MILES 15 I FIGURE 2.1-7 population Sectors for 0 -50 Miles Millstone Nuclear Power Station September 1999 Rev. 2 (,D MPS-I DSARFIGURE 2.1_8 ROADS AND FACILITIES IN THE LPZ PLEASURE MILLSTONE POINT.SOUN s ,./ I ts'íau'oi'!ytr cf tE AS T LY'LEGEND-;-f@tr tr E E MNPS.I DSAR TOWH SOUNDARY PRIIIARY ROAOS PAW/A.TR.AK RALFOAD STATE ROUTESNIAiITIC ELEMENTARY SCHOOL SOUTHWEST ELEMENTARY SCHOOL NEw LONoON COuIITRY CLUB GRAT NECK ELEMENTARY SCHOOL ds E El BAwrwNURsrNGnoiltE. SEASIOE REGIOL CENTERFIGURE 2.1-8 Roads and Facilities in the LPZ Millstone Nuclear Power Station nt SCENT SEACH ATTAWAN BLCK POINT ACH CLUE NIANT/C 8AY , LONG SCAL*-MILES September 1999 Rev. 2E A S T L y. NIANTIC LPZ BOUNDARY (2.4 Miles) BAY LONG MPS-l DSAR FIGURE 2.1-8 ROADS AND FACILITIES IN THE LPZ ..... . . . /" . ISLANO' 1.!Y " ' o 1/2 ! SOUNO \) SCALE-MILES MNPS-1 DSAR . . LEGEND _--TOWN BOUNDARY -PRIMARY ROADS P&W I AMTRAK RAILROAD @ STATE ROUTES OJ NIANTIC ELEMENTARY SCHOOL (g) SOUTHWEST ELEMENTARY SCHOOL @) NEW LONDON COUNTRY CLUB @]GREAT NECK ELEMENTARY SCHOOL lID BA'NIEW NURSING HOME .SEASIDE REGIONAl. CENTER FIGURE 2.1-8 Roads and Facilities in the LPZ Millstone Nuclear Power Station September 1999 Rev. 2 MPS-I DSAR FIGURE 2.1-9 LPZ POPULATION SECTORS DISTRIBUTION l,U E AS T{¿lA LYM ,--luO 7'sotl i:" t\.\ 1./ l\:l 1/2 i i I j rl E.i SW PZ EOUNDARY (2.4 Miles)SCALE-MILE S I;\sse FIGURE 2.1-9 LPZ Population Sectors Distribution Millstone Nuclear Power Station W A T E"R I b I a C)3nr4 I I l I. \--\., N/ANTIC SCENT PARK BAY ATTAWAN SSW Septembl tgsgE A S T L Y BLACK POINT CLUB SW LPZ BOUNDARY (2.4 Miles) ssw MPS-1 DSAR FIGURE 2.1-9 LPZ POPULATION SECTORS DISTRIBUTION N \ -..*. . 1 I / S L-' N 'l / SOU N 0 ,/i\ . 1/2 ! SCALE-MILES SE SCI \ 'SSE MNPS-1 DSAR ------FIGURE 2.1-9 LPZ Population Sectors Distribution Millstone Nuclear Power Station September 1999 _ Rev. 2 MPS-I DSAR FIGURE 2.1-IO INSTRUMENT LANDING PATTERNS AT TRUMBELL AIRPORT lts RwY 5 tüfllrcqccDirr VOR RWY 5rrAuu.l. cor*crcúI voR RWY 23 ,u a.totc ^f MNPS-1 DSAR. iar,aru ofûo(*accnclrf rroa.'r a(tíü, Millstone Unit No. 1 AIRPORT lt00ta4t -'tln. lraú JL\tln.lrd¿ lnJL. ll+a+!5 **q,Ch6* rf.ao^cl ø{nÉr to.t2t9: ,.n;;.---.- -Horr-r1.."S -.r?op_-(g* r( qü9ó "L4rliút .'wFÃr( Orú 9ó {.4r l

l. úr ?\ --t$l¿ó2d Z Millstone I/r\ ,,,*r). -&4tlii

\ tc HfO rO52\.ú lr &tûtUnit No. 1 ,/g¿tsfonÉ _\*/' ,//a-d roqrtqM g rtio'tor r toÚl /1 J_Efttrhb( -_-t{ l 5,ifrr'- lt09- oa----J- rr_r^-lriXi qDl.É:'.. taLt zrl./Mills rUnit t\\Itrdoroírs -,- ffi-,' j \P:-l \\ lroHa'/tone : No. 1 2a2;;o=ror -kWi ]t{-.fé/ t 4 :/ .rf'-/ 'e7 ,NEW LI'NU(,T AIRPORT i\'.'\-t@b tl: riabe L 'i ./-dW r-l9l o.dlt a4ézlti."l /./t¿,'ttx- /rss@ ártrorx C.1*rrl*tnb2qt0, -a{'.'rtot*ra j, r"- t:ffi" ffi"2'-jrooo av lO I C lll r*trí ?,¿\-at¿\a0::*CAIKi(Talo c 92t.ó4l óü(rütoa{}tI.ó2 (OO'l!ló,.(}llt dtt tt6tl cnctG'óGl *or(,ort. l"##*6&?ólo ttftltr'h d E d.ff.ûr

s. tril-r<1 rd.ljGd dlitr 6idd irF d aA'.

l.O .; (d.1 C6J Oft2, oûd Cd. (ord-tf ris'talil7 irs tlrlh.l{gl. i.*n d aridi¿ lrr 161t.YRWY 23 r.o.H-r2.0!.rv c"oron.fsco ^t?ro^ol OiíS l0O0 do il, aaó2 lulln H..d rcúf lofire,r.t ,q-Ot'5.tts Gcn r tÉdl8l tq 5 ,r¿ t-rr t1.tlli ffil É d.ídi ø oi¿m.a rd. ldód .dr.alirE aa hdü d dt..d OA'.la{t lcrr, trrø S{tS C¿, O r&dfr GOlO.r. <OrCCnCrrI ruruuul'tta-tz.*'t' 'lrNOTE: PAGES FROM DOO FLIGHT INFORMATION PUELICATION-LOW ALTITUOE INSTRUMENT APPROACH PROCEOURES NORTHEAST UNfTEO STATES VOL-7 FTGURE 2.,l-10 lnstrument Landing Patterns at Trumbull Airport Millstone Nuclear Power Station )September 1999 MPS-l DSAR FIGURE 2.1-10 INSTRUMENT LANDING PATTERNS AT TRUMBELL AIRPORT VORRWY 5 C!IJMl!IOL

• '20.' 279.2 /' / Millstone Unit No.1 / I \ \ \ \ \ \ "-'" IlS RWY 5 ./' IWS$EO APPaOAOl Oi_to 2000 M TMU 1.(162 to Mif_W_ hold. 60().2 590(600-21 41 '200N -n'03'w 214 / / / / 0"'5.5 _ __ Wood ... ISL 1wy5 MIlL 1"71 5-23. 'S*U_'O-21

"-' 60 90 '20 ,SO 180 ..... ,s.., 5,30 3,40 2,12 l,so GaOTON. TRUM8UU Millstone Unit No.1 .I / " . $8C).1 VORRWY 5 NOTE: PAGES FROM DOD FLIGHT INFORMATION LOW ALTITUDE INSTRUMENT APPROACH PROCEDURES NORTHEAST UNITED STATES VOL-7 -110 VORRWY.23 / / .,/ /' / Millstone I Unit No.1 \ \ '\ \ 0M0bi0v "'" ..... to 2000. ... TAW 1.(162 toMo_ w _hold. TKiVOI CItQIHG' VOR RWY 23 MNPS-1DSAR ISL Iwy 5 .' ...,. 5023 _ l50ll \ \ 1 / ...... 60 fO 120 ISO , 41'2O'N -72 '03'W 216 FIGURE 2.1-10 Instrument Landing Patterns at Trumbull Airport Millstone Nuclear Power Station ... September 1999 Rev. 2 MPS-I - DSAR FIGURE 2.I-II AIR LANES ADJACENT TO MILLSTONE POII{TB elftL.-ilJ I'offiifibt;fl t2S_W'Hagúure ;Old lyme NCTE.: T HIS A PORT ION SE CT IONA L c t-tA RT .?i-¿ll+:W&'-;IT'INPS.1 DSAR-iru K I LOM ETERS NAUI-CAT A,{rtfs p\ øTnr MAP RE PRESE N'I S TATUTE MILES of THE New voRx'A E RONAUT ICAL, plum*\o-l"f$0900 io 7.1^'iur 0R-IBU MU FGURE 2.1-11 Air Lanes Adjacent to Millstone point Millstone Nuclear Power Station September 1999 Rev.2 MPS-l-DSAR FIGURE 2.1-11 AIR LANES ADJACENT TO MILLSTONE POINT MNPS-1 DSAR FIGURE 2.1-11 Air Lanes Adjacent to Millstone Point Millstone Nuclear Power Station September 1999 Rev. 2 MPS-I DSARFIGURE 2.1_12 NEW LONDON COUNTY - STATE HIGHWAYS AND TOWN ROADS Psf l{@6r Pt. WiHG RGk B""n NIANTIc BAY 8v PT Porplnt MNPS-1 DSAR ,ii l'.-"'í::':iI I t, Irl*nr..*l .LON DO W LONQQ HARBo\-.1t!l I I l'*. I Elrn' Ouinnr$g Pr RGrs I i-ltu]b (lilMc I (t¿'?'\l' -ot'('¿*, Sbcr' Al_. t tú td3.t{ülrtÉi'Xf. .,¿R*k MirdRct. Comc Rat r\I J:=l-I G6nen r 'n t'Black Point Beach Club C I.t FIGURE 2.I-I2 New London County - State Highways and Town Roads Millstone Nuclear Power StationNest otd Shor s¿ft*d-Pr'Bi4 P;wt"i*'Hat(h*t'Pt)SCALE-K ILOMET September 1999 Rev. 2 John, *Fi:oct< i i MPS-l DSAR FIGURE 2.1-12 NEW LONDON COUNTY -STATE HIGHWAYS AND TOWN ROADS

• Long ,L __ , Blackboys o BAY o SCA LE -MILES 2 SCALE-KI Tvvorree 2 3 4 J FIGURE 2.1-12 New London County -State Highways and Town Roads Millstone Nuclear Power Station September 1999* Rev. 2 MPS-I- DSAR FIGURE 2.3_I TOPOGRAPHY IN THE VICINITY OF MILLSTONE POINT 150 1"1íf"l ,//¿I cr:t b t 1(4$b r<\fi)'car0=z 1t CN I.(]U)7 (b\/t{--'fl?01i tl'r,.'v a.\\iltAilrtc 8âr 0'5 il .0 E NUCLEAR R STATION 0 oVzt'i r I r I I I r r I SCALE-IILES ol2 L--J-JCAIE-fll,ollcrERS FIGURE 2.3-I Topography in the VicinitYof Millstone Point Millstone Nuclear Power Station September 1999fts.*l--i--rr, l' u-) n (.ií¿ \*.ir \\. I ,-t\-Jr-\

\\i, iir: , .\Ní,?\1\\'Stl' " \'/-\wifttrllrord n il-.t tio\u" v[*]\)5p l.f^ t..' r-, (,o,o I Rev.2 MPS-I-DSAR FIGURE 2.3-1 lOPOGRAPHY IN THE VICINITY OF MILLSTONE POINT . NIANTIC BAY .S o J "1",.1"" SCAt.E-MILES o 1 2 , , I SCAI.£-IC'!-OMUERS FIGURE 2.3-1 Topography in the Vicinity of Millstone Point Millstone Nuclear Power Station September 1999 Z "'0 en I c en >> ::0 Rev. 2 MPS-I DSAR FIGTJRE 3.1-1 REACTOR BUILDING SEISMIC LOADS BUILDING ITEGHT ND324 K 60.0 K-SEC kT.2244K SECTION PROPERTES tç Fr.al Ac GT.z) GT.z)5s-7 K-SEC?TI.1,464266 527.0 t7.751.05 K 57.5 K-S[C kT.u.rro3L_44oJ K-StC ?FT, t4.680.5K s5a2 K-SEC?FT.17,121.9K 54. K-SEC?FI.?5.344.t K 767. K-SEC 4I.s,r3q000 2925.0 8,063.880 8,758,388 627s FIGURE 3.1-1 Reactor Building Seismic Loads September 1999 r,464¿66 5n.o 2,28t,875 2842.0 2,856,030 35.0 nt.o 27t.0 t4zt.0 16i3.1 r82.0 1772 3l38 445.3 K-SEC ?FT.2,355.60s 23M.0 t7,8n.3K EL.42FI.-6IN. EL. t4 Fl. - 6 t.7 EL.O FI.- O IN.!! -26 rr. - 0 N. E TECIJYT S}EÂR ÂR*t463.0 3543 EL. I29 FT.EL. tO8 FT.E1.82 FI.EL.65 rT.Rev.2 MPS-1 DSAR FIGURE 3.1-1 REACTOR BUILDING SEISMIC LOADS BUILDING WEIGHT AND SECTION PROPERTIES Elo H7 FT. -)S24 K . 1 -... 60.0 K-SEC N 1,464,266 527.0 271.0 .... LL 2244K El.129 FT. -0 IN. a> 2 z W .... 6S.7 K-SEC LL ),464,266 . 527.0 271.0 0 Elo 108 FT. -6 IN. N 17.751.05 K 3 4 571.6 K-SEC a> .... 2,261,875 2842.0 1421.0 LL LO N 14.170.3 K Elo 82 rT. -9 IN. 4 1= 0 440.1 K-SEC .... 2,856,030 3205.0 16(13.1 LL 14.680.5 K Elo 65 FT. -9 IN. !:=:; 5 z M 445.9 K-SEC .... ..... 2,355,605 2364.0 1l82.0 n N El.42 rT. -6 IN. 4> *17,81l.3 K 6 0 553.2 K-SEC .... 5,138,000 2925.0 1463.0 LL a> N 17.421.9 K El. 14 Fl. -6 IN. z U 7 W 541.1 K-SEC .... 8,063,880 3543 1772 EL. 0 rT. -0 IN. u. 25.344.IK ..,. 8 -=. 787.1 K-S[C 0 .... 8,758,388 6275 3138 ..... tD EL. -26 rT. -0 IN. N

• EfTEClIVf Sf£AR AREA FIGURE 3.1-1 Reactor Building Seismic Loads September 1999 Rev. 2 MPS-I DSAR FIGURE 3.I-2 ACCELERATION DIAGRAM UNDER SEISMIC LOADS 5 PERCENT DAMPING 0 .r0 20 .30 .40 .50 .60ÂCCELERAT!0N lN 't 'UNITS FIGURE 3.1-2 Acceleration Diagram Under Seismic Loads 5 Percent Damping September 1999 150 ln h90 U=z.PGo U)U 30 0-129 Fr.- 0 IN.to8 FT.- 6 llt EI.. 82 FT. - 9 IN.EqTPUENT STISMIC COEFFICEN' CLNVE FOR RGID EQI,PIENT IN BI,LOING ONCLUDES FLTRJRAL AND ROCKING UODES)

N0TE: FOR CRITICAI-TUJIPMENT HAVING A PERIM Of VIBRÂTION GREATER THAN O.O5 SECO{DS A DYNAMIC ANALYSS TAS PERFMUED CONSIDERING 8I'ILDING INERACTIOi EL 65 FI. - 9 IN.EL. 42 Fr. - 6 N.EI. t4 FT. - 6 ]N.0FT.-0!N.Rev. 2 MPS-l DSAR FIGURE 3.1-2 ACCELERATION DIAGRAM UNDER SEISMIC LOADS 5 PERCENT DAMPING I-w w .... z 0 i= <0: :>-w .-J w 150 fL. 147 FT. - l.129 FT. -0 IN. 120 fllJIPIJENi SElSMIC COEFfJCENT CURVE FOR RIGID EQUlPMENT IN BUllDlNG IINCLUDES flEXURAL ANO ROCKING \.lODES) 60 NOTE: FOR CRITICAl 30 HAVING A PERIOO OF VIBRA lION GREATER THAN 0.05 SECCtlDS A EL 14 FT. -6 IN. DYNAMIC ANALYSIS WAS PERFOOIlEO CONSIDERING fL. 0 Fl. -0 IN. BUILDING INTERACT1!li 0 __ -L __ __ -L __ ____________ o .10 .20 .30 .40 .50 .60 ACCELERA liON IN

• E.
• UNJTS FIGURE 3.1-2 Acceleration Diagram Under Seismic Loads 5 Percent Damping September 1999 Rev. 2 n90 U 4=z.=Go U J U MPS-I DSAR FIGURE 3.I_3 SHEARDIAGRAM UNDER SEISMIC LOADS 68t01?14 SHIAR IN IOOO KIPS FIGURE 3.1.3 Shear Diagram Under Seismic Loads September 1999 EL t47 rT. - ZrN.EL. t29 FT. - 0 I.r.EL IOB FT, - 5 IN.EL. B? FT. - 9 IN..EL. 65 FT. - I t{-E1.42 FT.- 6 N.EL 14 FT, - 6 IN.O FT.- O It Rev. 2 MPS-l DSAR FIGURE 3.1-3 SHEAR DIAGRAM UNDER SEISMIC LOADS ]50 EL.129 rT. -0 IN. EL 108 rT. -6 IN. I-90 w w EL. 82 n. -9 IN. u.. :z C> 60 -l <>: >-w -l W 30 --; 0 I BASE El. -26 FT. -30 0 2 4 6 8 10 12 14 16 SHEAR IN 1000 KIPS FIGURE 3.1-3 Shear Diagram Under Seismic Loads September 1999 Rev. 2 rl 1hso UJ L z z YC^l-w uJ tJ J,l 0 MPS-IDSAR FIGURE 3.I-4 MOMENT DIAGRAM UNDER SEISMIC TOADS raz ri- - zjrH.EL IZ3 FT- - O ]N.tL. r08 FT. - 6 tN.tL. 82 FT. - 3 tN.l!-s!ll::-9-!H' EL.42FT.-6IN.

EL 14 F. - 6 ]N.T,LO FT.- O IN.BAST EL. -26 F.0406080100 I MoMENT rN roooo K'P-FT I ch' lo FIGURE 3.14 Momentiagram UnderSeismic Loads DECEMBER 2OOT-30 MPS-1 DSAR FIGURE 3.1-4 MOMENT DIAGRAM UNDER SEISMIC LOADS 1:1) EL 147 n. - EL lZg n. -0 ]N. 120 ... 9:l w z 0 60 i= < > W ..J W 30 MOMENT IN 10000 KIP-FT I Ch.10 FIGURE 3.1-4 Moment Diagram Under Seismic Loads DECEMBER 2001 Rev. 2 t50 t20 b90 U r==3Go 3 U 30 0'!MPS-I DSAR FIGURE 3.1-5 DISPLACEMENT DIAGRAM UNDER SEISMIC LOADS 0 60 90 lz0 t50 DISPLACEITNT IN MILS FIGURE 3.1-5 Displacement Diagram Under Seismic Loads September 1999-30 ei. Hu rr. - zlnr.EL. tzs FT. - 0 lN.ROCXING AND FLEXURAL HMES EL t08 FT. - 6 tN.EL. 82 FT. - 9 IN, EL. 65 FT. - I tit EL 42 FT. - 6 IN.ELI4 FT.- 6 ]N.ELO FT.- O IN.BÂSE ET. -26 F.Rev. 2 MPS-l DSAR 3.1-5 DISPLACEMENT DIAGRAM UNDER SEISMIC LOADS J50 J20 :z: § 60 < > w --' w 30 o 30 ROCKING ANO FLEXURAL NOOES 60 90 120 DISPlACBIENT IN MILS El. 147 n. - 150 FIGURE 3.1-5 Displacement Diagram Under Seismic Loads September 1999 Rev. 2 MPS-I DSAR FIGTIRE 3.I_6 RADWASTE BUILDING - MATHEMATICAL MODEL38.t4 K ll9.o9 K EL 49.75 F'r.66.40K -sEcz/FT ts L o o q o d c\,*l T FI rl=l EL 26.5 FT.EL I4.5 FT.59.ogK -sEcz/FT t934,93 K 6o.o9K -src2/rr tL 54.5 FT.EL 34.5 FT.lz.zsK.2 K -sc ?rr ts6.29K s.rsK -sec?rr zto.53x aYK -sic?ri 6246.22 4slssK -sEd/FT EL 38.75 F'r.EL 28.75 FT.EL I4.5 FT.September 1999 ts r o L t.N¿ó.¿ I l.65.42 FT.r5934.6r EL -1.50 FT,gq.as K -stc?/rt EL -20.0 FT.KEY PLAN FIGURE 3.1-6 Radwaste Building - Mathematical Model Rev.2 MPS-l DSAR FIGURE 3.1--{) RADWASTE BUILDING -MATHEMATICAL MODEL El 54.5 H. 2l3B.14 K lIS.OSK El 49.75 FT. 4 66.40" -SEC2/FT 3.7 K -SEC ?ttl ...: ...: ..... ..... 0 4 166.29 K 0 EL 38.75 FT. N EL 34.5 FT. 1902.75 K 5.16 K -SECqFT 2 t-..... ...: 59.09 K -SEC 2/FT 2 u.. 210.53 K EL 28.75 FT. EL 26.5 H. 1934.93 K 6.54 K -SECq FT 6O.09 K -SEC 2/FT 3 ...: ..... 6246.22 3 LO N EL 14.5 FT. 1193.98 K -SEW n EL 14.5 FT. t. -, 28.62 FT. 7 65.42 FT. ...: ..... 7 0 15934.61 EL -1.50 FT. 494.86 K -SECZ/FT 8 ...: 8 ..... 1 ll"!1 co, -1 x x I I Y -i--+ ----1-Y Y ----14---Y I I X x KEY PLAN FIGURE 3.1-6 Radwaste Building -Mathematical Model September 1999 Rev. 2 MPS-I DSAR FIGURE 3.2-1 P&ID: SFPI. FUEL POOL COOLING SYSTEM"r!lo f*l.u r ,--.r^ .t.e¿ _ii;;t ti*il tt I.'^.Ãr 1 t-v t-t:-l_tfiF \lr il IL-"'";.Ã-""fr*a'1-a-rI?;Ttr " l*l $ *'ií..,r,' -,1;=@ #/l'-'l'"k*i' I t.,l / p.r**. \, ji t"Å "g* \/--rE-q EE_e- üi'ü,..e.-¿ t[rl#.9c5rr6 irn&øs-&e!(¿ ¿offi@rÉsÉtø f4*@ Ea & rF _ __ _ - __ __--&.ms{É f,sr4 r4{f--_- _-___- _-----& ftn r4trA ffiúr-___---.G {Gn 3r4@ @ m6.__- -__&& *zq tr4@ ffi úi _---___---_2@ f6fi lr8@ Élnff t<< ú r.$ dt_- ffit a .d dr r.É E @E sÉre É É-er

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