ML22193A023

7 to Defueled Safety Analysis Report

ML22193A023
Person / Time
Site:	Millstone
Issue date:	06/23/2022
From:	Dominion Energy Nuclear Connecticut
To:	Office of Nuclear Reactor Regulation
Shared Package
ML22193A089	List: ML22193A023 ML22193A024 ML22193A025 ML22193A063 ML22193A064 ML22193A065 ML22193A066 ML22193A067 ML22193A068 ML22193A069 ML22193A070 ML22193A071 ML22193A072 ML22193A073 ML22193A074 ML22193A075 ML22193A076 ML22193A077 ML22193A078 ML22193A079 ML22193A080 ML22193A081 ML22193A121 ML22193A122 ML22193A123 ML22193A124 ML22193A125 ML22193A126 ML22193A127 ML22193A128 ML22193A130 ML22193A131 ML22193A132 ML22193A133 ML22193A134 ML22193A135 ML22193A136 ML22193A137 ML22193A138 ML22193A139 ML22193A140 ML22193A141 ML22193A142 ML22193A143 ML22193A144
References
22-195
Download: ML22193A023 (211)

Text

MILLSTONE POWER STATION UNIT 1 DEFUELED SAFETY ANALYSIS REPORT

Revision 1706/30/22 Document Elements Summary Description of Changes Revision Revision Change Activity Affected Made under the provisions of 10 CFR 50.59 excep Release Date (Sections, Tables, Figures) where indicated in brackets.

16.01 09/30/21 MP1-DFCR-2021-001 1.4, 6.3.4 Incorporate Quality Assurance Program Description Reference CA8380637 Revision 1606/30/21 Document Elements Summary Description of Changes Revision Revision Change Activity Affected Made under the provisions of 10 CFR 50.59 excep Release Date (Sections, Tables, Figures) where indicated in brackets.

15.01 01/21/21 MP1-DFCR-2014-002 6.1.3.1.a Reflects updated discussion of responsible individua for fuel handling activity.

License Amendment Request (LBDCR 13-MP1-001 Revision 1506/27/19 Document Elements Summary Description of Changes Revision Revision Change Activity Affected Made under the provisions of 10 CFR 50.59 excep Release Date (Sections, Tables, Figures) where indicated in brackets.

15 06/27/19 Administrative As identified in the previous Administrative (FSAR content not affected). Chang NRC Submittal List of Changed indicator (s) and page change identification (s) prese Pages and submitted Summary in the previous NRC Submittal removed in preparati of Change. for the 2019 (2018-2019) NRC Submittal. This form the base line for changes incorporated under the Revision 15 series. Revision level of the authoring fi are unchanged. This supports Revision/Change traceability.

Document Elements Summary Description of Changes Revision Revision Change Activity Affected Made under the provisions of 10 CFR 50.59 excep Release Date (Sections, Tables, Figures) where indicated in brackets.

14 06/29/17 Administrative As identified in the previous Administrative (FSAR content not affected). Chang NRC Submittal List of Changed indicator (s) and page change identification (s) prese Pages and submitted Summary in the previous NRC Submittal removed in preparati of Change. for the 2017 (2016-2017) NRC Submittal. This form the base line for changes incorporated under the Revision 14 series. Revision level of the authoring fi are unchanged. This supports Revision/Change traceability.

13.1 02/15/17 MP1-DFCR-2015-001 Sections 1.2.3.4.3, 3.2.7.2, Reflects installation of a new padmounted transform 3.2.7.4.1, 3.2.7.4.2, 3.2.9.2.2 to replace existing ESST.

Revision 1306/22/15 Document Elements Summary Description of Changes Revision Revision Change Activity Affected Made under the provisions of 10 CFR 50.59 excep Release Date (Sections, Tables, Figures) where indicated in brackets.

13 06/22/15 Administrative As identified in the previous Administrative (FSAR content not affected). Chang NRC Submittal List of Changed indicator (s) and page change identification (s) prese Pages and submitted Summary in the previous NRC Submittal removed in preparati of Change. for the 2016 (2015-2016) NRC Submittal. This form the base line for changes incorporated under the Revision 13 series. Revision level of the authoring fi are unchanged. This supports Revision/Change traceability.

Revision 1210/29/2014 Document Elements Summary Description of Changes Revision Revision Change Activity Affected Made under the provisions of 10 CFR 50.59 excep Release Date (Sections, Tables, Figures) where indicated in brackets.

12 10/29/14 Administrative As identified in the previous Administrative (FSAR content not affected). Chang NRC Submittal List of Changed indicator (s) and page change identification (s) prese Pages and submitted Summary in the previous NRC Submittal removed in preparati of Change. for the 2014 NRC Submittal. This forms the base lin for changes incorporated under the Revision 12 seri Revision level of the authoring files are unchanged.

This supports Revision/Change traceability.

11.2 11/15/13 MP1-DFCR-2013-001 List of Figures Reflects administrative clarification of revision statu for engineering controlled drawings that are coincidently FSAR Figures. FSAR figures corresponding to Controlled P&IDs are updated on periodic basis by the FSAR Coordinator.

11.1 9/13/13 Administrative List of Figures Reflects administrative clarification of revision statu for engineering controlled drawings that are coincidently FSAR Figures. FSAR figures corresponding to Controlled P&IDs are updated on periodic basis by the FSAR Coordinator.

Revision 11July 2013 Document Elements Summary Description of Changes Revision Revision Change Activity Affected Made under the provisions of 10 CFR 50.59 excep Release Date (Sections, Tables, Figures) where indicated in brackets.

11 July 2013 Administrative As identified in the previous Administrative (FSAR content not affected). Chang NRC Submittal List of Changed indicator (s) and page change identification (s) prese Pages and submitted Summary in the previous NRC Submittal removed in preparati of Change. for the 2014 NRC Submittal. This forms the base lin for changes incorporated under the Revision 11 seri Revision level of the authoring files are unchanged.

This supports Revision/Change traceability.

Revision 10July 2012 Document Elements Summary Description of Changes Revision Revision Change Activity Affected Made under the provisions of 10 CFR 50.59 excep Release Date (Sections, Tables, Figures) where indicated in brackets.

10 July 2012 MP1-DFCR-2011-001 S1.2.3.2.1, S4.1, S4.4 Reflects updates for compliance with NPDES permi 10 July 2010 Administrative As identified in the previous Administrative (FSAR content not affected). Chang NRC Submittal List of Changed indicator (s) and page change identification (s) prese Pages and submitted Summary in the previous NRC Submittal removed in preparati of Change. for the 2013 NRC Submittal. This forms the base lin for changes incorporated under the Revision 10 seri Revision level of the authoring files are unchanged.

This supports Revision/Change traceability.

Revision 9June 2011 Document Elements Summary Description of Changes Revision Revision Change Activity Affected Made under the provisions of 10 CFR 50.59 excep Release Date (Sections, Tables, Figures) where indicated in brackets.

9 June 2011 Administrative As identified in the previous Administrative (FSAR content not affected). Chang NRC Submittal List of Changed indicator (s) and page change identification (s) prese Pages and submitted Summary in the 2011 NRC Submittal removed in preparation of Change. the 2012 NRC Submittal. This forms the base line fo changes incorporated under the Revision 9 series.

Revision level of the authoring files are unchanged.

This supports Revision/Change traceability.

Revision 8March 2011 Document Elements Summary Description of Changes Revision Revision Change Activity Affected Made under the provisions of 10 CFR 50.59 excep Release Date (Sections, Tables, Figures) where indicated in brackets.

8 March 2011 MP1-DFCR-2010-001 S3.2.7.2 Reflects change to substation nomenclature (Northe Utilities Distribution Project) for off site power sour

Revision 7April 2009 Document Elements Summary Description of Changes Revision Revision Change Activity Affected Made under the provisions of 10 CFR 50.59 excep Release Date (Sections, Tables, Figures) where indicated in brackets.

7 04/09 -- As identified in the 2009 NRC Administrative (FSAR content not affected). Chang Submittal List of Changed indicator (s) and page change identification (s) prese Pages and submitted Summary in the 2009 NRC Submittal removed in preparation of Change. the 2010 NRC Submittal. This forms the base line fo changes incorporated under the Revision 7 series.

Revision level of the authoring files are unchanged.

This supports Revision/Change traceability.

Table of Contents tion Title Page CHAPTER 1- INTRODUCTION AND GENERAL DESCRIPTION OF PLANT INTRODUCTION ...................................................................................... 1.1-1 GENERAL PLANT DESCRIPTION ......................................................... 1.2-1 1 PLANT SITE AND ENVIRONS ............................................................... 1.2-1 1.1 Location and Site ........................................................................................ 1.2-1 1.2 Site Ownership............................................................................................ 1.2-1 1.3 Access to the Site ........................................................................................ 1.2-1 1.4 Description of the Environs ........................................................................ 1.2-1 1.5 Geology....................................................................................................... 1.2-1 1.6 Seismology and Design Response Spectra ................................................. 1.2-1 1.7 Hydrology ................................................................................................... 1.2-2 1.8 Meteorology................................................................................................ 1.2-2 1.9 Site Environmental Radioactivity Monitoring Program ............................. 1.2-2 2

SUMMARY

PLANT DESCRIPTION ....................................................... 1.2-3 3 SYSTEMS .................................................................................................. 1.2-3 3.1 Fuel Storage and Fuel Handling ................................................................. 1.2-3 3.2 Radioactive Waste Processing System ....................................................... 1.2-3 3.3 Radiation Monitoring and Control.............................................................. 1.2-4 3.4 Auxiliary Systems....................................................................................... 1.2-5 3.5 Station Communication System.................................................................. 1.2-5 3.6 Station Water Purification, Treatment and Storage System ....................... 1.2-6 IDENTIFICATION OF AGENTS AND CONTRACTORS...................... 1.3-1 1 APPLICANTS SUBSIDIARIES............................................................... 1.3-1 2 NUCLEAR STEAM SUPPLY SYSTEM SUPPLIER............................... 1.3-1

tion Title Page 3 ARCHITECT/ENGINEER ......................................................................... 1.3-1 4 TURBINE-GENERATOR SUPPLIER ...................................................... 1.3-1 MATERIAL INCORPORATED BY REFERENCE ................................. 1.4-1 CONFORMANCE TO NRC REGULATORY GUIDES .......................... 1.5-1 1

SUMMARY

DISCUSSION ....................................................................... 1.5-1 2 REFERENCE.............................................................................................. 1.5-1 CHAPTER 2- SITE CHARACTERISTICS LOCATION AND AREA .......................................................................... 2.1-1 1 POPULATION ........................................................................................... 2.1-2 1.1 Population Distribution Within 50 Miles.................................................... 2.1-3 1.2 Transient Population ................................................................................... 2.1-3 1.3 Low Population Zone.................................................................................. 2.1-3 1.4 Population Center ....................................................................................... 2.1-4 2 LAND USE................................................................................................. 2.1-4 2.1 Description of Facilities.............................................................................. 2.1-5 2.2 Pipelines...................................................................................................... 2.1-8 2.3 Waterways .................................................................................................. 2.1-8 2.4 Airports ....................................................................................................... 2.1-8 2.5 Highways .................................................................................................... 2.1-9 2.6 Railroads ..................................................................................................... 2.1-9 2.7 Projections of Industrial Growth............................................................... 2.1-10 3 DETERMINATION OF DESIGN BASIS EVENTS ............................... 2.1-10 4 EFFECTS OF DESIGN BASIS EVENTS ............................................... 2.1-11 5 REFERENCES ......................................................................................... 2.1-12

tion Title Page METEOROLOGY ...................................................................................... 2.2-1 1 REGIONAL CLIMATOLOGY.................................................................. 2.2-1 2 LOCAL METEOROLOGY........................................................................ 2.2-1 2.1 Potential Influence of the Plant and Its Facilities on Local Meteorology...................................................................................... 2.2-1 2.2 Local Meteorological Conditions for Design and Operating Bases. .......................................................................................................... 2.2-1 3 ON SITE METEOROLOGICAL MEASUREMENTS PROGRAM ......... 2.2-1 4 SHORT TERM (ACCIDENT) DIFFUSION ESTIMATES ...................... 2.2-2 4.1 Objective ..................................................................................................... 2.2-2 4.2 Calculations ................................................................................................ 2.2-2 4.3 Results......................................................................................................... 2.2-2 5 LONG-TERM (ROUTINE) DIFFUSION ESTIMATES........................... 2.2-2 5.1 Objective ..................................................................................................... 2.2-2 5.2 Calculations ................................................................................................ 2.2-2 6 REFERENCES ........................................................................................... 2.2-3 HYDROLOGIC ENGINEERING .............................................................. 2.3-1 1 HYDROLOGIC DESCRIPTION ............................................................... 2.3-1 2 SITE AND FACILITIES ............................................................................ 2.3-1 3 FLOODS ..................................................................................................... 2.3-1 3.1 Flood History .............................................................................................. 2.3-1 3.2 Flood Design Considerations...................................................................... 2.3-1 3.3 Effect of Local Intense Precipitation .......................................................... 2.3-1 4 PROBABLE MAXIMUM FLOOD (PMF) ON STREAMS AND RIVERS............................................................................................. 2.3-2 5 POTENTIAL DAM FAILURE, SEISMICALLY INDUCED................... 2.3-2 6 PROBABLE MAXIMUM SURGE AND SEICHE FLOODING.............. 2.3-2

tion Title Page 6.1 Probable Maximum Winds and Associated Meteorological Parameters................................................................................................... 2.3-2 6.2 Surge and Seiche Water Levels .................................................................. 2.3-3 6.3 Wave Action ............................................................................................... 2.3-3 6.4 Resonance ................................................................................................... 2.3-3 6.5 Protective Structures ................................................................................... 2.3-4 6.6 Probable Maximum Tsunami Flooding ..................................................... 2.3-4 7 ICE EFFECTS ............................................................................................ 2.3-4 8 COOLING WATER CANALS AND RESERVOIRS ............................... 2.3-4 9 CHANNEL DIVERSIONS......................................................................... 2.3-4 10 FLOODING PROTECTION REQUIREMENTS ...................................... 2.3-4 11 LOW WATER CONSIDERATIONS ........................................................ 2.3-4 11.1 Low Flow in Rivers and Streams ............................................................... 2.3-4 11.2 Low Water Resulting from Surges, Seiches, or Tsunamis ......................... 2.3-4 12 DISPERSION, DILUTION, AND TRAVEL TIMES OF ACCIDENTAL RELEASES OF LIQUID EFFLUENTS SURFACE WATERS. ................................................................................ 2.3-4 13 GROUNDWATER ..................................................................................... 2.3-4 14 TECHNICAL SPECIFICATION AND EMERGENCY OPERATION REQUIREMENTS .............................................................. 2.3-5 15 REFERENCES ........................................................................................... 2.3-5 GEOLOGY, SEISMOLOGY, AND GEOTECHNICAL ENGINEERING ......................................................................................... 2.4-1 1 BASIC GEOLOGIC AND SEISMIC INFORMATION............................ 2.4-1 2 VIBRATORY GROUND MOTION .......................................................... 2.4-1 2.1 Safe Fuel Storage Earthquake..................................................................... 2.4-1 3 SURFACE FAULTING ............................................................................. 2.4-1 3.1 Geologic conditions of the Site................................................................... 2.4-1 3.2 Evidence of Fault Offset ............................................................................. 2.4-1

tion Title Page 3.3 Earthquakes Associated with Capable Faults ............................................. 2.4-1 3.4 Investigation of Capable Faults .................................................................. 2.4-1 3.5 Correlation of Epicenters with Capable Faults ........................................... 2.4-2 3.6 Description of Capable Faults..................................................................... 2.4-2 3.7 Zone Requiring Detailed Faulting Investigation ........................................ 2.4-2 3.8 Results of Faulting Investigation ................................................................ 2.4-2 4 STABILITY OF SUBSURFACE MATERIALS AND FOUNDATIONS ........................................................................................ 2.4-2 5 STABILITY OF SLOPES .......................................................................... 2.4-2 6 EMBANKMENTS AND DAMS ............................................................... 2.4-2 7 REFERENCES ........................................................................................... 2.4-2 CHAPTER 3- FACILITY DESIGN AND OPERATION DESIGN CRITERIA .................................................................................. 3.1-1 1 CONFORMANCE WITH 10 CFR 50 APPENDIX A GENERAL DESIGN CRITERIA............................................................... 3.1-1 1.1 Summary Discussion .................................................................................. 3.1-1 1.2 Systematic Evaluation Program and Three Mile Island Evaluations of General Design Criteria ...................................................... 3.1-1 2 CLASSIFICATION OF STRUCTURES, SYSTEMS, AND COMPONENTS ........................................................................................ 3.1-1 2.1 Seismic Classification................................................................................. 3.1-1 2.2 Safety Related Classification ...................................................................... 3.1-3 2.3 Non-Safety Related Plant Functions Maintained in the Defueled Condition..................................................................................... 3.1-4 2.4 SSCs Important to the Defueled Condition ................................................ 3.1-4 3 WIND AND TORNADO LOADINGS ...................................................... 3.1-8 4 WATER LEVEL DESIGN ......................................................................... 3.1-8 5 MISSILE PROTECTION ........................................................................... 3.1-8

tion Title Page 5.1 Internally Generated Missiles ..................................................................... 3.1-8 5.2 Missiles Generated by Natural Phenomena ................................................ 3.1-8 5.3 Missiles Generated by Events Near the Site ............................................... 3.1-9 5.4 Aircraft Hazards.......................................................................................... 3.1-9 6 SEISMIC DESIGN ..................................................................................... 3.1-9 6.1 Comparison of Measured and Predicted Responses ................................. 3.1-10 7 DESIGN OF CLASS I AND CLASS II STRUCTURES......................... 3.1-10 7.1 Design Criteria, Applicable Codes, Standards and Specifications............................................................................................ 3.1-10 7.2 Loads and Loading Combinations ............................................................ 3.1-10 7.3 Structural Criteria for Class II Structures ................................................. 3.1-12 7.4 Seismic Class I and II Structures .............................................................. 3.1-12 8 SEISMIC QUALIFICATION OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT ................. 3.1-15 9 ENVIRONMENTAL DESIGN OF ELECTRICAL EQUIPMENT ........ 3.1-16 10 REFERENCES ......................................................................................... 3.1-16 SYSTEMS .................................................................................................. 3.2-1 1 FUEL STORAGE AND HANDLING ....................................................... 3.2-1 1.1 New Fuel Storage........................................................................................ 3.2-1 1.2 Spent Fuel Storage ...................................................................................... 3.2-1 1.3 Spent Fuel Pool Cooling System ................................................................ 3.2-3 1.4 Fuel Handling System ................................................................................ 3.2-5 2 MONITORING AND CONTROL FUNCTIONS ..................................... 3.2-6 3 DECAY HEAT REMOVAL (DHR) SYSTEM ......................................... 3.2-6 3.1 Design Bases .............................................................................................. 3.2-6 3.2 System Description ..................................................................................... 3.2-7 3.3 Safety Evaluation ........................................................................................ 3.2-7 3.4 Testing and Inspection ............................................................................... 3.2-7

tion Title Page 3.5 Instrumentation .......................................................................................... 3.2-7 4 MAKEUP WATER SYSTEM.................................................................... 3.2-7 4.1 Demineralized Water ................................................................................. 3.2-7 5 INTENTIONALLY DELETED ................................................................. 3.2-8 6 PROCESS SAMPLING SYSTEM ............................................................. 3.2-8 6.1 Design Bases............................................................................................... 3.2-8 6.2 System Description ..................................................................................... 3.2-8 6.3 Safety Evaluation ........................................................................................ 3.2-8 6.4 Testing and Inspection ............................................................................... 3.2-9 7 ELECTRICAL SYSTEMS ......................................................................... 3.2-9 7.1 Introduction................................................................................................. 3.2-9 7.2 Off Site Source............................................................................................ 3.2-9 7.3 Intentionally Deleted................................................................................... 3.2-9 7.4 On Site Electric System .............................................................................. 3.2-9 8 AIR CONDITIONING, HEATING, COOLING AND VENTILATION SYSTEMS..................................................................... 3.2-11 8.1 Reactor Building and SFPI Heating and Ventilation System ................... 3.2-11 8.2 Radwaste Building Ventilation System .................................................... 3.2-13 8.3 Intentionally Deleted................................................................................. 3.2-14 8.4 Turbine Building Heating and Ventilation ............................................... 3.2-14 9 FIRE PROTECTION SYSTEMS ............................................................. 3.2-15 9.1 Design Bases ............................................................................................ 3.2-15 9.2 System Description .................................................................................. 3.2-15 9.3 Safety Evaluation and Fire Hazards Analysis........................................... 3.2-19 9.4 Inspection and Testing .............................................................................. 3.2-20 9.5 Personnel Qualification and Testing......................................................... 3.2-21 10 REFERENCES ........................................................................................ 3.2-22

tion Title Page CHAPTER 4- RADIOACTIVE WASTE MANAGEMENT SOURCE TERMS ...................................................................................... 4.1-1 RADIATION PROTECTION DESIGN FEATURES ............................... 4.2-1 1 FACILITY DESIGN FEATURES ............................................................. 4.2-1 1.1 Design Basis ............................................................................................... 4.2-1 1.2 Ventilation .................................................................................................. 4.2-1 2 RADIATION PROTECTION PROGRAM................................................ 4.2-1 2.1 Organization................................................................................................ 4.2-1 ALARA PROGRAM .................................................................................. 4.3-1 1 POLICY CONSIDERATIONS ................................................................. 4.3-1 1.1 Design Considerations ................................................................................ 4.3-1 1.2 Operational Considerations......................................................................... 4.3-1 LIQUID WASTE MANAGEMENT SYSTEMS ....................................... 4.4-1 SOLID WASTE MANAGEMENT ............................................................ 4.5-1 1 DESIGN BASES ....................................................................................... 4.5-1 2 SYSTEM DESCRIPTION.......................................................................... 4.5-1 3 REFERENCES ........................................................................................... 4.5-2 EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING ....... 4.6-1 1 DESIGN ..................................................................................................... 4.6-1 1.1 Design Basis ............................................................................................... 4.6-1 1.2 System Design Description......................................................................... 4.6-1 2 AREA RADIATION MONITORING INSTRUMENTATION ................ 4.6-2 2.1 Design Bases............................................................................................... 4.6-2 2.2 System Description ..................................................................................... 4.6-2

tion Title Page 3 REFERENCE.............................................................................................. 4.6-2 CHAPTER 5- ACCIDENT ANALYSIS INTRODUCTION ...................................................................................... 5.1-1 1 ACCIDENT EVENT EVALUATION ....................................................... 5.1-1 1.1 Unacceptable Results for Design Basis Accidents (DBAs)........................ 5.1-1 1.2 Fuel Handling Accident Assumptions ....................................................... 5.1-1 1.3 Results......................................................................................................... 5.1-1 1.4 Radiological Consequences ........................................................................ 5.1-1 2 REFERENCES ........................................................................................... 5.1-2 FUEL HANDLING ACCIDENT ............................................................... 5.2-1 1 FUEL HANDLING ACCIDENT SCENARIOS IN THE SPENT FUEL POOL.................................................................................. 5.2-1 2 RADIOLOGICAL CONSEQUENCES...................................................... 5.2-2 3 REFERENCES ........................................................................................... 5.2-3 CHAPTER 6- CONDUCT OF OPERATIONS ORGANIZATIONAL STRUCTURE ....................................................... 6.1-1 1 MANAGEMENT AND TECHNICAL SUPPORT ORGANIZATION ...................................................................................... 6.1-1 1.1 Technical Support for Operations............................................................... 6.1-1 1.2 Organizational Arrangement....................................................................... 6.1-1 2 OPERATING ORGANIZATION ............................................................. 6.1-1 2.1 Plant Organization ..................................................................................... 6.1-1 2.2 Plant Personnel Responsibilities and Authorities ....................................... 6.1-1 2.3 Operating Shift Crews ................................................................................ 6.1-1

tion Title Page 3 QUALIFICATIONS OF NUCLEAR PLANT PERSONNEL .................. 6.1-2 3.1 Qualification Requirements ........................................................................ 6.1-2 4 REFERENCES ........................................................................................... 6.1-2 TECHNICAL SPECIFICATIONS ............................................................ 6.2-1 PROGRAMS ............................................................................................. 6.3-1 1 TRAINING ................................................................................................. 6.3-1 2 EMERGENCY PLAN ................................................................................ 6.3-1 3 PHYSICAL SECURITY PLANS............................................................... 6.3-1 4 QUALITY ASSURANCE PROGRAM DESCRIPTION (QAPD)

TOPICAL REPORT ................................................................................... 6.3-1 5 REFERENCES ........................................................................................... 6.3-2 PROCEDURES ......................................................................................... 6.4-1 REVIEW AND AUDIT.............................................................................. 6.5-1 1 ONSITE REVIEW...................................................................................... 6.5-1 2 INDEPENDENT REVIEW ........................................................................ 6.5-1 3 AUDITS ..................................................................................................... 6.5-1 CHAPTER 7- DECOMMISSIONING

SUMMARY

OF ACTIVITIES .................................................................. 7.1-1 1 DECOMMISSIONING APPROACH ....................................................... 7.1-2 1.1 Planning ..................................................................................................... 7.1-2 1.2 Site Characterization................................................................................... 7.1-3 1.3 Decontamination ......................................................................................... 7.1-3 1.4 Major Decommissioning Activities ............................................................ 7.1-4 1.5 Other Decommissioning Activities............................................................. 7.1-4

tion Title Page 1.6 Final Site Survey and Termination of License ........................................... 7.1-5 1.7 Site Restoration........................................................................................... 7.1-6 2 STORAGE OF RADIOACTIVE WASTE................................................. 7.1-6 2.1 High Level Waste ....................................................................................... 7.1-6 2.2 Low Level Waste ........................................................................................ 7.1-7 2.3 Waste Management..................................................................................... 7.1-7 3 RADIATION EXPOSURE MONITORING.............................................. 7.1-7 4 REFERENCES .......................................................................................... 7.1-7 ESTIMATE OF RADIATION EXPOSURE.............................................. 7.2-1 1 NUCLEAR WORKER .............................................................................. 7.2-1 2 GENERAL PUBLIC .................................................................................. 7.2-1 3 NORMAL TRANSPORTATION .............................................................. 7.2-2 CONTROL OF RADIATION RELEASES ASSOCIATED WITH DECOMMISSIONING EVENTS .................................................. 7.3-1 1 IN PLANT EVENTS ................................................................................. 7.3-1 2 TRANSPORTATION ACCIDENTS ......................................................... 7.3-1 NON-RADIOLOGICAL ENVIRONMENTAL IMPACTS ..................... 7.4-1 1 ADDITIONAL CONSIDERATIONS ........................................................ 7.4-1

mber Title BLE 1.1-1 Millstone Unit No.1 Licensing Milestones BLE 2.1-1 This Table has been Intentionally Deleted BLE 2.1-2 1990 Population and Population Densities - Cities and Towns within 10 miles of Millstone BLE 2.1-3 Population Growth 1960 - 1990 BLE 2.1-4 Population Distribution within 10 miles of Millstone - 1990 Census BLE 2.1-5 Population Distribution Within 10 Miles of Millstone 2000 Projected BLE 2.1-6 Population Distribution Within 10 Miles of Millstone 2010 Projected BLE 2.1-7 Population Distribution Within 10 Miles of Millstone 2020 Projected BLE 2.1-8 Population Distribution Within 10 Miles of Millstone 2030 Projected BLE 2.1-9 Population Distribution Within 50 Miles of Millstone - 1990 Census BLE 2.1-10 Population Distribution Within 50 Miles of Millstone - 2000 Projected BLE 2.1-11 Population Distribution Within 50 Miles of Millstone - 2010 Projected BLE 2.1-12 Population Distribution Within 50 Miles of Millstone - 2020 Projected BLE 2.1-13 Population Distribution Within 50 Miles of Millstone - 2030 Projected BLE 2.1-14 Transient Population Within 10 Miles of Millstone 1991-1992 School Enrollment BLE 2.1-15 Transient Population Within 10 Miles of Millstone - Employment BLE 2.1-16 Population Distribution Within 50 Miles of Millstone - 2030 Projected BLE 2.1-17 Low Population Zone Permanent Population Distributions BLE 2.1-18 Low Population Zone School Enrollment and Employment BLE 2.1-19 Metropolitan areas Within 50 Miles of Millstone 1990 Census Population BLE 2.1-20 Population Centers within 50 Miles of Millstone BLE 2.1-21 Population Density Within 10 Miles of Millstone 1990 (People per Square Mile)

BLE 2.1-22 Population Density Within 10 Miles of Millstone 2030 (People per Square Mile)

BLE 2.1-23 Population Density Within 50 Miles of Millstone 1990 (People per Square Mile)

mber Title BLE 2.1-24 Population Density Within 50 Miles of Millstone 2030 (People per Square Mile)

BLE 2.1-25 Cumulative Population Density Within 50 Miles of Millstone 1990 (People per Square Mile)

BLE 2.1-26 Cumulative Population Density Within 50 Miles of Millstone 2030 (People per Square Mile)

BLE 2.1-27 Description of Facilities BLE 2.1-28 List of Hazardous Materials Potentially Capable of Producing Significant Missiles BLE 3.1-1 Comparison with NRC General Design Criteria BLE 3.1-2 Allowable Stresses for Class I Structures BLE 4.6-1 Effluent Radiation Monitors BLE 4.6-2 Area Radiation Monitoring System Sensor and Converter Locations for Millstone Unit No. 1 BLE 5.2-1 Assumptions and Input Conditions for Fuel Handling Accident at Millstone Unit No. 1

mber Title URE 1.2-1 Plot Plan URE 1.2-2 General Arrangement Radwaste Buildings - Plans URE 1.2-3 General Arrangement Buildings Radwaste Buildings - Sections URE 2.1-1 General Site Location URE 2.1-2 General Vicinity URE 2.1-3 Site Layout URE 2.1-4 Site Plan URE 2.1-5 Towns Within 10 Miles URE 2.1-6 Population Sectors for 0 - 10 Miles URE 2.1-7 Population Sectors for 0 - 50 Miles URE 2.1-8 Roads and Facilities in the LPZ URE 2.1-9 LPZ Population Sectors Distribution URE 2.1-10 Instrument Landing Patterns at Trumbull Airport URE 2.1-11 Air Lanes Adjacent to Millstone Point URE 2.1-12 New London County - State Highways and Town Roads URE 2.3-1 Topography in the Vicinity of Millstone Point URE 3.1-1 Reactor Building Seismic Loads URE 3.1-2 Acceleration Diagram Under Seismic Loads 5 Percent Damping URE 3.1-3 Shear Diagram Under Seismic Loads URE 3.1-4 Moment Diagram Under Seismic Loads URE 3.1-5 Displacement Diagram Under Seismic Loads URE 3.1-6 Radwaste Building - Mathematical Model URE 3.2-1 P&ID: SFPI, Fuel Pool Cooling System URE 3.2-2 P&ID: SFPI, Fuel Pool Cooling System URE 3.2-3 P&ID: SFPI, Fuel Pool Cooling System (Refueling Bellows Leak Detection)

URE 3.2-4 P&ID: Reactor Building and HVAC Room SFPI Secondary Cooling (DHR)

System URE 3.2-5 P&ID: Reactor Building SFPI, Make-Up Water System

mber Title URE 3.2-6 P&ID SFPI HVAC System Composite URE 3.2-7 P&ID: HVAC B.O.P. System Composite URE 3.2-8 Intentionally Deleted URE 3.2-9 Intentionally Deleted URE 3.2-10 Intentionally Deleted URE 3.2-11 Intentionally Deleted URE 3.2-12 P&ID: HVAC Balance Of Plant System Composite URE 3.2-13 P&ID: HVAC System (Radwaste Storage Building)

URE 3.2-14 Fire Protection Composite

INTRODUCTION s Defueled Safety Analysis Report (DSAR) is submitted by the licensee in support of the ommissioning for Millstone Unit Number 1 at the Millstone Nuclear Power Station in erford, Connecticut. Dominion Nuclear Connecticut, Inc. owns and is responsible for the ommissioning of Millstone Unit Number 1.

DSAR is the principle licensing source document describing the pertinent equipment, ctures, systems, operational constraints and practices, accident analyses, and ommissioning activities associated with the existing defueled condition of Millstone Unit mber 1. As such, the DSAR is intended to serve in the same role as the Final Safety Analysis ort of Millstone Unit Number 1 during the periods of power operation between 1970 and

8. The DSAR is applicable throughout the decommissioning of Millstone Unit Number 1. The ommissioning process is dynamic. The issuance of the DSAR does not alleviate the licensee m continuing to follow all required surveillances, procedures, technical specifications or ilar documents, until those documents are officially modified using approved processes.

wings and figures of structures, systems, or components (SSCs) included or referenced in the AR, are included within the licensing basis of the facility only to the extent that they show s that are described in the text of the DSAR. Other contents of drawings and figures may not ect the current configuration of the facility and are not maintained.

struction of Millstone Unit Number 1 was authorized by a provisional construction permit R-20, on May 19,1966, in AEC Docket 50-245. Millstone Unit Number 1 was completed and y for fuel loading during October 1970. The plant went into commercial operation on ember 28, 1970. On July 21, 1998, pursuant to 10 CFR 50.82(a)(1)(i) and 10 CFR 2(a)(1)(ii), the licensee certified to the NRC that, as of July 17, 1998, Millstone Unit Number ad permanently ceased operations and that fuel had been permanently removed from the tor vessel. The issuance of this certification fundamentally changes the licensing basis of lstone Unit Number 1 in that the NRC-issued 10 CFR 50 license no longer authorizes ration of the reactor or emplacement or retention of fuel in the reactor vessel. Therefore, as of 21, 1998, only those conditions or activities associated with the safe storage of fuel and ological protection (including waste handling, storage and disposal) are applicable to the ueled Millstone Unit Number 1 plant.

lstone Unit Number 1 was a single cycle, boiling water reactor with a Mark I containment ch was designed, furnished and constructed by General Electric Company as prime contractor the licensee. The General Electric Company engaged Ebasco Services Incorporated as itect-engineer. Millstone Unit Number 1 had a reactor thermal output of 2011 megawatts and t electrical output of 652.1 megawatts. The Millstone site is located in the town of Waterford, w London County, Connecticut, on the north shore of Long Island Sound.

EVENT DATE nstruction Permit Issued May 19, 1966 AR Filed November 1, 1968 visional operating License Issued October 7, 1970 l-Term Operating License Issued October 31, 1986 l Power License October 7, 1970 tial Criticality October 26, 1970 nchronized to the Grid November 1970 0 Percent Power January 6, 1971 mmercial operation December 28, 1970 manently Ceased Operations July 21, 1998

1 PLANT SITE AND ENVIRONS 1.1 Location and Site site for the Millstone Nuclear Power Station consists of a tract of land of approximately 500 s located in the town of Waterford, Connecticut on the north shore of Long Island Sound and he east side of Niantic River Estuary. It is located 3.2 miles west-south-west of New London, 40 miles south-east of Hartford, Connecticut. The site is bounded on the west, south, and ions of the east sides by Long Island Sound. The nearest residential boundary is 855 meters h-east of the major structures of Millstone Unit Number 1. Chapter 2 contains more detailed rmation on the site and surrounding areas.

1.2 Site Ownership site is owned by Dominion Nuclear Connecticut, Inc.

thheld under 10 CFR 2.390 (d) (1) 1.4 Description of the Environs acent to the site to the north and west is cultivated land with residential dwellings. The village Niantic, consisting of a small commercial complex and attendant residential development, is miles north-west of the Reactor Building. Other residential areas adjoin the site at the end of plant access road and at distances of 1 to 3 miles.

w London, 3.2 miles ENE of the Reactor Building, is the nearest urban complex and includes ed residential, commercial, and industrial uses.

1.5 Geology site area is underlain by Monson gneiss and Westerly granite. The Westerly granite intrudes Monson gneiss, is more resistant to weathering and therefore forms ridges. Seismic surveys losed no unusual or extreme subsurface conditions. Chapter 2 contains more detailed rmation on geology and seismic qualities.

1.6 Seismology and Design Response Spectra Millstone Point site area is placed in Zone 2 (zone of moderate damage) on the seismic bability map of the 1964 Uniform Building Code.

preceding design criteria are for critical items only, that is, for Class I items. Class I items are ned in Chapter 3.

1.7 Hydrology plant site natural grade level is at an elevation of approximately 14 feet above mean sea level.

ause of the contours of the land and ground strata, and the distance of the reactor from water plies, no water accidentally released from the plant can reach industrial or drinking water plies.

pter 2 contains more detailed information on hydrology.

1.8 Meteorology meteorology of the site area is basically that of a sea-coast location with relatively favorable ospheric dilution conditions prevailing. The inland terrain in Connecticut is not pronounced ugh to produce any significant local modifications of synoptic conditions at the shoreline. The reline areas do, however, experience local modifications of synoptic patterns because of the perature differences between air over land and air over water.

site is located in an area occasionally traversed by hurricanes. The design basis hurricane for lstone has 124 mph maximum gradient winds and a 17 mph speed of translation. This is ificantly more intense than the worst on record (hurricane of 1938).

as been estimated that a tornado can be expected to strike a point on the Millstone site about ry 1,804 years. In spite of this low probability, the features of the plant important to the safe age of irradiated fuel have been designed to withstand 300 mph winds.

concluded that from the viewpoint of site meteorology, the site is suitable for the station as cribed. (Chapter 2 contains more detailed information concerning meteorology.)

1.9 Site Environmental Radioactivity Monitoring Program environmental radioactivity monitoring program was initiated and has been conducted at the since April 1967. Data are collected to measure radioactivity present in the environs. The gram is continuing in order to assure prompt detection and evaluation of any changes in oactivity.

3 SYSTEMS 3.1 Fuel Storage and Fuel Handling 3.1.1 Fuel Storage and Handling Equipment spent fuel storage pool holds fuel assemblies, control rods, and small vessel components. The l system contains provisions to maintain water cleanliness and instrumentation to monitor er level. Makeup water is available from the Unit 2 demineralized water system and the fire er system. The racks in which fuel assemblies are placed are designed and arranged to ensure criticality in the pool.

handling of spent fuel is performed within the Reactor Building. This employs a refueling form for underwater fuel transport, storage racks for fuel and control rods in a storage pool, erwater fuel preparation stations, and floor mounted jib cranes. Control rods can be stored in fuel pool racks or on hooks on the side of the pool.

ctural design of the fuel storage and equipment storage facilities meets all requirements for ss I structures. For additional information, refer to Chapter 3.

3.1.2 Fuel Pool Cooling System fuel pool cooling system provides cooling for the spent fuel pool water when required.

fuel pool cooling system consists of a circulating pump, heat exchanger, skimmer surge s, system piping, valves, and instrumentation and controls. Pool cleanup is provided by an in-l demineralizer and filter. For additional information, refer to Chapter 3.

3.2 Radioactive Waste Processing System radioactive waste processing system is designed to control the release of plant-produced oactive material to within the limits specified in 10 CFR 20 and Appendix I to 10 CFR 50.

s is done by collection, transfer, and evaporation.

uid waste drained or transferred to the Reactor Building sumps will be processed using the te Water Processing System (WWPS) or using an atmospheric evaporator. The Waste Water cessing System consists of four (4) 10,000 gallon Sample Tanks, recirculation pump, ineralizer, filters and associated piping. The A RBFD sump will pump to the WWPS ple tanks, where the water will be batch recirculated and sampled before subsequent harge. Radiological monitoring will be conducted using an in-line Liquid Effluent Monitor

-MG-110). Prior to discharge through DSN-001A (Emergency Service Water discharge ng to discharge canal), dilution flow requirements will be established by crediting Unit 2 ulating Water Flow to the common discharge canal. Alternatively, this system could be zed to pump the process liquids from the Reactor Building sumps to containers which would mit the process liquid to be processed on site or off site.

3.2.2 Solid Radwaste Handling d wastes originating from nuclear system equipment maybe stored in the spent fuel storage l and prepared for off site shipment in approved shipping containers.

d wastes are collected and appropriately prepared for off site shipment. Examples of these d wastes are filter residue, spent resins, paper, air filters, rags, and used clothing. For itional information, refer to Chapter 4.

3.3 Radiation Monitoring and Control 3.3.1 Radiation Monitoring and Sampling Spent Fuel Pool Island ventilation exhaust is monitored for gaseous radiation and iculates. A particulate sampling skid is provided for Unit 1 Balance of Plant (BOP) exhaust to mit sampling for any significant changes. For additional information, refer to Chapter 4.

3.3.2 Area Radiation Monitors iation monitors are provided to monitor for abnormal radiation at selected locations on the I. These monitors actuate alarms when abnormal radiation levels are detected.

3.3.3 Liquid Radwaste Processing System Control liquid radwaste system is designed to safely and economically collect, store, process, and ose of, or recycle, all radioactive or potentially radioactive liquid waste generated. The em operates on a batch basis.

3.3.4 Solid Radwaste Control d radwaste can be transferred to high integrity cask containers for shipment.

3.4.1 Decay Heat Removal (DHR) System DHR system provides cooling water to the spent fuel pool cooling system. The system sists of circulating pumps, air-water heat exchangers, an expansion tank, air separator and ciated piping valves and controls, and a portable ethylene glycol addition pump and tank.

3.4.2 Monitoring and Control Functions Millstone Unit 2 Control Room is continuously manned, and serves as the control room for lstone Unit 1. Millstone Unit 2 Operations personnel are responsible for the monitoring and trol of the Unit 1 spent fuel pool island (SFPI) and auxiliary systems via a computer console ted in the Millstone Unit 2 Control Room.

3.4.3 Fire Protection System protection and detection systems are provided at Millstone Unit Number 1 to protect ctures, systems, and components important to the defueled condition of the unit.

fire protection system includes a fire water supply system that consists of two fire water s, fire water pumps and a distribution system that delivers fire water to all parts of the plant.

water systems within the plant protect individual hazards and include sprinkler systems.

3.4.4 Electrical Power System 3.4.4.1 AC Power Supply electric power system includes the electrical equipment and connections required to supply er to station auxiliaries.

3.4.4.2 DC Power Supply SFPI 125 V DC system is provided via rectified AC at the point of use. In addition, a separate ommissioning 125V DC system powered by batteries and a battery charger provides a source C power to the decommissioning electrical system.

I 24V power is provided by power supplies within the SFPI Programmable Logic Controller C) panels.

3.5 Station Communication System plant communication system provides for reliable on site and off site communications both er normal and contingency conditions.

s system provides demineralized makeup water to Millstone Unit Number 1 for use in the nt fuel pool.

thheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-1 PLOT PLAN

thheld under 10 CFR 2.390 (d) (1)

FIGURE 1.2-2 GENERAL ARRANGEMENT RADWASTE BUILDINGS - PLANS

thheld under 10 CFR 2.390 (d) (1)

GURE 1.2-3 GENERAL ARRANGEMENT BUILDINGS RADWASTE BUILDINGS -

SECTIONS

1 APPLICANTS SUBSIDIARIES minion Nuclear Connecticut, Inc. is responsible for the decommissioning of Millstone Unit mber 1. Dominion Nuclear Connecticut, Inc. is a wholly owned subsidiary of Dominion rgy, which is wholly owned by Dominion Resources, Inc.

2 NUCLEAR STEAM SUPPLY SYSTEM SUPPLIER eral Electric Company was the nuclear steam system supplier for the plant.

3 ARCHITECT/ENGINEER sco Services Incorporated was the Architect/Engineer for Millstone Unit Number 1.

4 TURBINE-GENERATOR SUPPLIER turbine generator was manufactured by General Electric Company.

following is a list of material incorporated by reference in the Defueled Safety Analysis ort (1).

1. As identified in the List of Figures, the engineering controlled plant drawings that are, coincidentally, DSAR Figures.

2. The Quality Assurance Program Description (QAPD) Topical Report.

3. Specific sections of the Millstone Unit Number 2 and Number 3 FSARs as identified within the text of the DSAR.

Information incorporated by reference into the Defueled Safety Analysis Report is subject to update and reporting requirements of 10 CFR 50.71(e) and change controls of 10 CFR 50.59 ss separate NRC change control requirements apply (e.g., 10 CFR 50.54(a)).

1

SUMMARY

DISCUSSION AEC issued Appendix A General Design Criteria to 10 CFR 50 in July 1971. In November 0, Safety Guides, later to become Regulatory Guides, began to be published. These guides vided acceptable means for complying with specified general design criteria. They were not in ct at the time Millstone Unit Number 1 began operation with Provisional Operating License L) DPR-21, issued October 7, 1970.

lstone Unit Number 1 submitted summaries of compliance to these guides in the early 1970s upport of the application for a full-term operating license (Reference 1.5-1).

ore acting on this application, the NRC (formerly AEC) initiated the Systematic Evaluation gram (SEP) in 1977 to review the designs of older operating nuclear reactor plants in order to firm and document their safety. Millstone Unit Number 1 was identified as an SEP plant.

SEP objectives were:

• To establish documentation that shows how the criteria for each operating plant reviewed compare with current criteria on significant safety issues and to provide a rational for acceptable departures from these criteria.

• To provide the capability to make integrated and balanced decisions with respect to any required backfitting.

• To provide for early identification and resolution of any significant deficiencies.

• To assess the safety adequacy of the design and operation of currently licensed nuclear power plants.

• To use available resources efficiently to minimize requirements for additional resources by NRC or industry.

• To ensure that the safety assessments were adequate for conversion of provisional operating licenses to full-term operating licenses.

final version of the SEP program report included the status of all applicable generic activities I and USIs), including those that formed the basis for the Integrated Safety Analysis Program AP) being implemented by the Licensee. Based upon the acceptable conclusions reached in

, the NRC issued the full-term operating license for Millstone Unit Number 1 on October 31, 6.

2 REFERENCE 1 Millstone Nuclear Power Station Unit 1 Application for Full-Term Operating License, September 1, 1972.

LOCATION AND AREA Millstone site is located in the Town of Waterford, New London County, Connecticut, on the h shore of Long Island Sound. The 524 acre site occupies the tip of Millstone Point between ntic Bay on the west and Jordan Cove on the east and is situated 3.2 miles west-southwest of w London and 40 miles southeast of Hartford.

Millstone Unit Number 1 containment structure is located immediately south of Millstone 2

3. The geographical coordinates of the centerline of the reactor is as follows:

lstone Unit Number 1 Latitude and Longitude Northing and Easting N 41 18'32" N 173, 800 W 72 10'04" E 759, 965 site is owned by Dominion Nuclear Connecticut, Inc. Figures 2.1-1 through 2.1-4 identify site.

site protected area is considered the restricted area. The restricted area has been spicuously posted and administrative procedures, including periodic patrolling, have been osed to control access to the area. For the purpose of radiological dose assessment of dents, the exclusion area boundary (EAB) was considered the actual site boundary for rland sectors, except in the Fox Island / discharge channel area on the south end of the site. For water sectors, the nearest land site boundary distance was used.

significant normal releases are discharged to the atmosphere via the Unit Number 1 BOP aust point and the SFPI ventilation exhaust point. The distance from the Unit Number 1 BOP aust point and the SFPI ventilation exhaust point to the nearest residential property boundary he Millstone Point Colony development (Point A on Figure 2.1-3) is greater than 2,800 feet.

s development, adjacent to the eastern site boundary, consists of single family homes on 104 acre lots. One of the conditions of the sale of the site to the Hartford Electric Light Company the Connecticut Light and Power Company was that permanent dwellings would never be mitted in the beach area of the development. Because of this restriction, normal release doses calculated at Point A rather than at the nearest point on the site boundary.

thheld under 10 CFR 2.390 (d) (1) ensure the safety of people within the exclusion area during an emergency, an emergency plan the site has been prepared. The plan includes provisions for alarms both inside and outside

lusion area extending offshore through a written agreement between the licensee and the U.S.

st Guard at their station in New London, Connecticut.

owners have encouraged public use of portions of the site. Ownership rights have not, ever, been relinquished, and the owners can, and have provision to, fulfill their obligations h respect to 10 CFR 20, Standards for Protection Against Radiation.

ortion of the exclusion area is leased to the Town of Waterford for public recreation and is d primarily for soccer and baseball games. Figure 2.1-3 shows the general location of these vities. No attempt is made to restrict the number of persons using these facilities. Estimates of imum attendance indicate that about 2,000 visitors could be within the exclusion area at any time at the soccer and baseball fields. The licensee's Emergency Plan provides for removal of visitors from the site. The number and configuration of roads and highways assure ready ss from the areas described above (Figures 2.1-2, 2.1-3, and 2.1-4).

1 POPULATION total 1990 population within 10 miles of the station was estimated to be 120,443. This ulation is expected to increase to about 129,846 people by the year 2000 and to a total of roximately 142,277 people by the year 2030 (New York State Department of Economic elopment, 1989 (Reference 2.1-1); State of Connecticut Office of Policy and Management, 1 (Reference 2.1-2); US Department of Commerce, Bureau of the Census, 1990 Census of ulation (Reference 2.1-3)). The 10 mile area includes portions, or all of, New London and dlesex Counties in Connecticut and a small portion on Suffolk County of Fishers Island which art of the town of Southold, New York. Figure 2.1-5 shows counties and towns within the 10 e area. Town populations and population densities are provided in Table 2.1-2.

Town of Waterford, in which Millstone Unit Number 1 is located, contained a total ulation of 17,930 people in 1990 at an average density of 547 people per square mile (US artment of Commerce Bureau of the Census 1991) (Reference 2.1-3). The population growth Waterford was small with the 1990 total representing only a 0.5 percent increase over its 1980 ulation. Compared to towns immediately surrounding it, with the exception of New London, erford had the lowest increase in population between 1980 and 1990 (US Department of mmerce Bureau of the Census, 1991 (Reference 2.1-3)).

erford's growth has been consistently slowing down over the past 30 years, as shown in le 2.1-3. This slow growth is projected by state demographers to continue at a low rate ugh the year 2000, at which time the population is expected to reach 18,480. After that, it is ected to decrease in population. By the year 2010 (the last year of projections), the town's ulation is projected to be 18,080 (Connecticut Office of Policy and Management, Interim ulation Projections, 1991 (Reference 2.1-2)). Population distribution by sector for the area hin 10 miles of Millstone Unit Number 1 is shown for the years 1990, 2000, 2010, 2020 and 0 in Tables 2.1-4 through 2.1-8, that are keyed to the population sectors identified in ure 2.1-6.

r its land area, unless USGS 7.5 minute quadrangle maps indicated the population to be centrated in only on portion of the Block. The proportion of each Block area in each grid or was determined and applied to the Block total population, yielding the population in each sector. Population projections, by municipality, supplied by Connecticut's Office of Policy Management provided growth factors for projection of Projections, 1991 (Reference 2.1-2).

1.1 Population Distribution Within 50 Miles area within 50 miles of Millstone Unit Number 1 includes portions, or all, of eight counties in necticut, four counties in Rhode Island and one county in New York. Figure 2.1-7 shows nties and towns within the 50 mile area. In 1990, the 50 mile area contained approximately 5,159 people (U.S. Department of Commerce), 1990 Census of Population and Housing ference 2.1-4). This population is projected to increase to about 3,223,654 by the year 2030 nnecticut Office of Policy and Management, 1991 (Reference 2.1-2); New York State artment of Economic Development, 1989 (Reference 2.1-1); Rhode Island Department of ministration, 1989 (Reference 2.1-5); US Department of Commerce, 1990 Census of ulation and Housing, 1991 (Reference 2.1-4)). Population distribution by sector for the area hin 50 miles of Millstone Unit Number 1 is shown for the years 1990, 2000, 2010, 2020 and 0 in Tables 2.1-9 through 2.1-13, which are keyed to the population sectors identified in ure 2.1-9.

ulation distribution and projections within the 50 mile region surrounding Millstone Unit mber 1 were calculated based on population by municipalities and were assigned to sectors ed on land area allocation. Projections for the 50 mile area were based on country-wide ections.

1.2 Transient Population sonal population increases resulting from an influx of summer residents total approximately

00. However, many of the beaches and recreation facilities in the area are used by residents, therefore, do not represent any increase in population but instead a slight shift in population.

re are, however, a number of schools, industries, and recreation facilities which create daily seasonal variations in sector populations. Tables 2.1-14 through 2.1-16 show annular sector ulation variations resulting from school enrollments, industrial employment, and recreation lities (with documented attendance).

1.3 Low Population Zone low population zone (LPZ) surrounding Millstone Unit Number 1 encompasses an area hin a radial distance of about 2.4 miles. The distance was chosen based on the requirements of CFR 100.11. Figure 2.1-8 shows topographical features, transportation routes, facilities, and itutions within the LPZ.

LPZ contained approximately 9,846 people in 1990, with an average density of 545 people square mile. By the year 2030, the LPZ population is projected to increased to about 11,629,

ference 2.1-2); US Geological Survey (Reference 2.1-6)). The LPZ population distribution for 0 and 2030 is shown in Table 2.1-17. Table 2.1-18 shows the 1991-1992 school and loyment distribution within the LPZ. Both tables are keyed to Figure 2.1-9.

ly and seasonal variations due to transient population are minimal within the LPZ. Several ches are located within the area; however, they are predominantly used by local residents and erally have no facilities for parking or accommodation of large groups. Three schools, Great k Elementary and Southwest Elementary in Waterford, and Niantic Elementary in East Lyme, located within the LPZ. Major employment consists of the Camp Rell Military Reservation Hendel Petroleum. The New London Country Club is also located within the LPZ.

1.4 Population Center closest population center to Millstone Unit Number 1 (as defined by 10 CFR 100 to contain e than 25,000 residents) is the city of New London which contained a 1990 population of 540 people at an average population density of 5,189 people per square mile (US Department ommerce Bureau of the Census 1991). The distance between Millstone Unit Number 1 and city's closest corporate boundary is about 3.3 miles to the northeast, just beyond the minimum ance requirement set by 10 CFR 100.

region within 50 miles of Millstone Unit Number 1 includes portions, or all, of 11 ropolitan Statistical Area's. The populations of these areas are shown in Table 2.1-19.

re were 38 population centers within 50 miles of Millstone Unit Number 1, containing 25,000 ore people in 1990. They are listed in Table 2.1-20 with the populations indicated.

population of the area within 50 miles of Millstone was approximately 2,800,000 in 1990, h an average density of 361 people per square mile. This density is lower than the NRC parison figure of 500 people per square mile (NRC Regulatory Guide 1.70, Revision 3, erence 2.1-7). Within 30 miles of Millstone, the population density is considerably less, at an rage of 189 people per square mile. By 2030, the 50 mile population is projected to increase to 0,000 or an average population density of about 410 people per square mile, considerable er than the NRC comparison figure for end-year plant life of 1,000 people per square mile.

hin 30 miles, the average density will be 223 persons per square miles by the year 2030.

ulation densities by sector for 1990 and 2030 are shown for within 10 miles of Millstone in les 2.1-21 and 2.1-23 respectively, which are keyed to Figure 2.1-6, and for within 50 miles Millstone in Tables 2.1-23 and 2.1-24, respectively, which are keyed to Figure 2.1-7.

mulative population densities 1990 and 2030 are shown in Tables 2.1-25 and 2.1-26, ectively.

2 LAND USE area around the Millstone site contains three major industrial facilities (Dow Chemical poration, Pfizer Corporation, and Electric Boat division of General Dynamics Corporation);

transportation facilities (Groton/New London) Airport and the New London Transportation

re is also an interstate highway (Interstate 95), passenger and freight railroad lines, gas ribution lines, above ground gas and oil storage facilities and two major waterways (Long nd Sound, Thames River) in the vicinity of the Millstone site.

re are no major gas transmission lines, oil transmission or distribution lines, under ground gas age facilities, drilling or mining operations, or military firing, or bombing ranges near the site.

craft patterns and routes are shown of Figures 2.1-10 and 2.1-11. Figure 2.1-12 shows the d and highway system in the area of the Millstone site.

2.1 Description of Facilities ummary of the significant industrial, transportation, military, and industrial related facilities, products and materials used, is shown in Table 2.1-27 as listed below.

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.

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 and from Pfizer corporation by truck and/or railroad.

Electric Boat Division of General Dynamics of Eastern Point Road, Groton, Connecticut is located approximately 5 miles east-northeast of the site. Electric boat employs approximately 12,000 persons, and is a producer 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 substantial amounts of nuclear material which is licensed under the Naval Reactors Division. All material is moved by truck, railroad, and/

or barge to and from the company with the exception of completed ships which leave under their own power.

Groton / New London Airport, approximately 6 miles east-northeast of the site, handles regularly scheduled commercial passenger flights. Approximately 13 persons are employed at Groton/New London Airport on a full-time basis, excluding airline and car rental employees. The National Guard has an aircraft repair facility at the airport that has approximately 140 full time employees.

20 persons are employed there 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 entrance and exit for New London for civilian and commercial travel.

U. S. Navy Submarine Base, Groton, Connecticut is located on the east bank of the Thames River, approximately 7 miles northeast 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 Submarine 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 ship, to and from this government installation.

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 approximately 360 military and civilian personnel are employed here.

Camp Rell, located approximately 2 miles northwest of the site, 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 basis, 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 occupied by varying numbers of troop units for administrative training maneuvers, billeting, and supply functions 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 administrative 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.

In addition to Camp Rell, the Military Department of the State of Connecticut also maintains a field training facility known as Stone's Ranch Military Reservation, located approximately 7 miles northwest of the site. Fourteen persons are employed there full-time for two regional motor vehicle 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 Guard. During some weekend training sessions there may be up to 500 people at the facility.

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.

rotary-wing operations. Because of its distance 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.

Hess Oil Corporation of Eastern Point Road, Groton, Connecticut is located on the east bank of the Thames River, approximately 5 miles east-northeast of the site. It is located north of Pfizer Corporation, and south of General Dynamics-Electric 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 above ground tanks capable of storing heating oil, residual fuel oil, and kerosene. The fuel arrives by ships or barges and is distributed by trucks.

There is one medium-sized propane storage area in the proximity 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 facility for commercial and residential use. There are 5 above ground tanks (3-30,000 gallons and 2-16,000 gallons) which are capable of storing 126,000 gallons total of propane gas. The facility also stores 40,000 gallons of gasoline, and 40,000 gallons of Number 2 fuel oil. The propane for the facility arrives by train and truck, and is distributed by truck.

On the Millstone site, at the Fire Training Facility located approximately 2,800 feet to the north of the protected area are two 1,000 gallon propane cylinders. The two cylinders are used to supply propane to the fire simulator. The Fire Training Facility 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 Millstone site. Both cylinders are above ground domestic storage cylinders designed per ASME Code for Pressure Vessels,Section VIII Division 1-92.

Montville Station is a Fossil Fuel powered electric generating plant operated by Connecticut Light & Power Company in Montville, Connecticut. It is located on the west bank of the Thames River, approximately 9.5 miles north-northeast of the site.

Approximately 67 people are employed 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 barrels of fuel each; and two small above ground tanks, capable of storing approximately 250 barrels of fuel each. The fuel arrives by barge or trucks.

re are no major gas transmission lines within 5 miles of the site. There are two medium sure gas distribution lines in near proximity of the site. The nearest gas distribution line is roximately 2.9 miles from the site, located along Rope Ferry Road in Waterford. This 35 psi distribution line is a 6-inch plastic pipeline, buried approximately 3 feet deep. The control e for this line is located at the intersection of Clark Lane and Boston Post Road in Waterford.

second gas distribution line, ends at and serves the shopping center complex, near the rsection of I-95 and Parkway North, approximately 4 miles north of the site. This 35 psi gas ribution line is an 8 inch plastic pipeline buried approximately 3 feet deep. The control valve this line is located at the complex where it intersects with Parkway North.

re are no oil transmission or distribution lines within 5 miles of the Millstone site.

2.3 Waterways ps that pass by the site in the shipping channels of Long Island Sound are of two types: general o freighters, usually partially unloaded, with drafts of 20 to 25 feet, and deep draft tankers h drafts of 35 to 38 feet. Both of these classes of ships must remain at least 2 miles offshore to vent running aground on Bartlett Reef.

oil barges pass to the shore side of Bartlett Reef, and since there are no tank farms in Niantic

, no oil barges pass with 2 miles of the site.

ge traffic in the vicinity of the site has been diminishing over the past several years due to the rease in the amount of oil used by area facilities. Barge traffic is heaviest during the winter nths, and averages only 1 barge per day during these months. On the average of once a month, arge carrying 15,000 barrels of sulfuric acid is towed past the site outside of Bartlett Reef.

roximately 10 ships per day traverse the Reef in the vicinity, 6 miles of the site.

these reasons, it is concluded that shipping accidents would not adversely affect Millstone 3 ty related facilities.

2.4 Airports ton / New London Airport, approximately 6 miles east-northeast of the site, handles regularly eduled commercial passenger flights. It is served by U.S. Air Express. It has two runways:

3, which is 5,000 feet long; and 15-33, which is 4,000 feet long. Both runways are illuminated.

re is a control tower at Groton / New London, with ILS (Instrument Landing System) and R (Very High Frequency Omni Range) navigation aides located on the airfield. The ILS is ciated with runway 5. As shown on Figure 2.1-10, the landing patterns used do not direct fic near the Millstone site.

largest commercial aircraft to use Groton / New London Airport on a regularly scheduled s are Beechcraft 1900's which carry approximately 19 passengers. The only jets using the ort on a regular basis are two small chartered Cessna Citation which carry 10 passengers.

1995 there were approximately 4,490 military flights, approximately half of which were tary helicopters. Millstone Station is not in the flight path of these flights, and pilots are fed to avoid the site.

shown on Figure 2.1-11, the air lane nearest the site is V58 which is approximately 4 miles heast of the site. Other adjacent air lanes include V16, which is approximately 6 miles hwest of the site, and V308, which is approximately 8 miles east of the site. The nearest h-altitude jet route, J121-581, passes approximately 9 miles southeast of the site. A second jet e, J55, passes approximately 12 miles northwest of the site.

2.5 Highways area around the Millstone site is served by interstate, state and local roads. These are shown Figure 2.1-12. The nearest major highway which would be used for frequent transportation of ardous materials is U.S. Interstate 95, which is located 4 miles from the Millstone site. Other cipal highways which pass near the site include U.S. Highway 1 which is located 3 miles from site, and State Highway 156, located 1.5 miles from the site.

se separation distances exceed the minimum distance criteria given in Regulatory Guide 1.91, ision 1 and provide assurance that any transportation accidents resulting in explosions or toxic releases of truck size shipments of hazardous materials would not have a significant adverse ct on the safe operation or shutdown capability of the unit.

2.6 Railroads site is traversed from east to west by a Providence & Worcester (P&W)/Amtrak railroad t-of-way. The mainline tracks are more than 2,000 feet from the Millstone Unit Number 1 ctor Building structure.

motive force for the rail stock is both diesel and electric locomotives. Overhead electric lines er the former. These lines affect neither the site nor the overhead transmission lines leaving site and traversing the railroad right-of-way above the tracks.

Department of Transportation and P&W/Amtrak have been contacted for information cerning rail traffic on the mainline tracks. Approximately eighteen scheduled passenger trips day pass along the tracks near the Millstone site.

roximately one freight train per day passes by the site. Hazardous material shipped on the k include chlorine, anhydrous ammonia, carbon dioxide, propane, ethyl alcohol, rosin, monium nitrate, and hydrochloric acid. See Table 2.1-28 for a list of hazardous materials dled over this track which are potentially capable of producing significant missiles.

lstone site. See Section 2.1.3 for a more detailed evaluation of potential accidents.

railroad spur serves the Millstone Nuclear Power Station exclusively. The switch for that spur ormally set for through traffic. In order to reach any station facility, a train car must also pass ugh a second switch, which is normally set to direct traffic past the station to a dead end near Sound. Therefore, the possibility of unauthorized transport of hazardous materials on the spur ery remote.

re are no grade crossings on or adjacent to the site at which hazardous materials might be sported across the tracks.

2.7 Projections of Industrial Growth elines expansion of facilities is presently planned in the area for oil distribution within the theastern region of Connecticut. The gas distribution line along Rope Ferry Road ends at erford high School, approximately 2.9 miles from the Millstone site. The gas distribution line 95 and Parkway North ends at, and serves the shopping complex approximately 4 miles from Millstone site.

erways previously mentioned, ship and barge traffic in the area of Millstone site has decreased over past several years. No new ship or barge traffic is anticipated at this time in the Niantic Bay on Long Island Sound near location of the intake structures.

ports expansion of facilities at Groton / New London Airport is proposed although some rovements to the facility, such as expansion of the approach lights, and upgrading of the inal and runways in planned. Southeastern Connecticut Regional Planning Agency (SCRPA) mmends that a master plan be prepared for the airport before any major physical rovements are made. The agency has previously adopted the policy that Groton / New London port should remain a small feeder airport providing connection to larger airports and direct ice to a limited number of cities with a 500 mile radius.

3 DETERMINATION OF DESIGN BASIS EVENTS area around the Millstone site was investigated and found to contain no explosives, micals, airborne pollutants, flammable or dangerous gases, nor tanks or pipelines near enough he site to pose a danger if they were to explode or burn.

thheld under 10 CFR 2.390 (d) (1)

nearest major highway which would be used for frequent transportation of hazardous erials is U.S. Interstate 95, which is located at a distance of 4 miles from the Millstone site.

s separation distance exceeds the minimum distance criteria given in Regulatory Guide 1.91, ision 1; and therefore, provides assurance that any transportation accidents resulting in losions of truck size shipments of hazardous materials will not have an adverse effect on the operation of the plant.

ed upon the size of Groton / New London airport and the location of flight paths, the impact of irplane on Millstone Unit Number 1 is highly unlikely.

re are no major gas transmission lines within 5 miles of the site. The nearest low pressure gas ribution line is 2.9 miles from the site and is located near Waterford High School on Rope y Road.

closest oil transmission line is approximately 5 miles from the site in Groton Connecticut.

ause they are 5 miles or more away from the site, both the major gas and oil transmission lines stitute no threat to the safe conduct of activities associated with storage of irradiated fuel or ommissioning of Millstone Unit Number 1 or to the site in general.

4 EFFECTS OF DESIGN BASIS EVENTS thheld under 10 CFR 2.390 (d) (1)

1 New York State Department of Economic Development, Interim County, MSA and Region Projections, 1980 - 2010, 1989.

2 Connecticut Office of Policy Management, Interim Population Projections Series 91.1, 1991.

3 U.S. Department of Commerce, Bureau of the Census, 1990 Census of Population, P.L.94-171 Counts by Census Block, 1991.

4 U.S. Department of Commerce, Bureau of the Census, 1990 Census of Population and Housing - Connecticut, 1990 CPH-1-8, 1991.

5 Rhode Island Department of Administration, Projections by County, 1990 - 2020, 1989.

6 U.S. Geological Survey, 7.5-Minute Quadrangle Maps.

7 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.70, Revision 3.

8 AAR-RPI Number RA-01-2-7, 1972. Association of American Railroads and Railway Progress Institute Final Phase 01 Report on Summary of Ruptured Tank Cars Involved in Past Accidents, Revised July 1972. Chicago, IL.

9 AAR-RPI Number RA-02-2-18, 1972. Association of American Railroads and Railway Progress Institute Final Phase 02 Report on Accident Review, Chicago, IL.

10 Chemical Rubber Company, 1972. Handbook of Chemistry and Physics 44th and 53rd Editions.

11 ConRail 1980. Hazardous Materials Link Report between New Haven and New London, Connecticut from January 1978 through June 1979.

12 Giffen, 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.

13 Lotti, 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.

14 Kaiser, G. D. and Griffiths, R. F. 1982. The Accidental Release of Anhydrous Ammonia:

A Systematic Study of the Factors Influencing Cloud Density and Dispersion, Journal of the Air Pollution Control Association, Vol. 32, Number 1.

15 NASA Report 3023, 1978. Workbook for Estimating the Effects of Accidental Explosions in Propellant Ground Handling and Transport Systems.

17 NTSB-RAR-1, 1972.National Transportation Safety Board Accident Report for East St.

Louis, MO.

18 NTSB-RAR-75-7, 1974. National Transportation Safety Board Railroad Accident Report for Houston, TX.

19 NTSB-RAR-79-11, 1979. National Transportation Safety Board Railroad Accident Report for Crestview, FL.

20 NTSB-RAR-81-1, 1980. National Transportation Safety Board Railroad Accident Report for Muldraugh, KY.

21 NUREG-0800, 1981. Standard Review Plan: Evaluation of Potential Accidents (Section 2.23).

22 Perry & Chilton 1973. Chemical Engineers Handbook, 5th Edition McGraw-Hill, Inc.

23 Personal Communication between S.N. Bajpai and Robert Folden, Federal Railroad Administration, Office of Safety, February 17, 1982.

24 Deleted.

25 Research 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 Evacuations. Run Dated March 26, 1981., Covering Period December 22, 1970 to September 5, 1980.

26 Research and Special Programs Administration, U.S. Department of Transportation, Washington, D.C. 1981. Computer Printout of Incidents Involving Fire and Explosions by ConRail. Run dated 4/15/81 Covering Period June 6, 1973 through November 1, 1980.

27 Rhoads, R.E. et al 1978. An Assessment of Risk of Transporting Gasoline by Truck PNL-2133. Pacific Northwest Laboratory (Battelle Memorial Institute), Richland, Washington.

28 Siewert, R.D. 1972. Evacuation Areas for Transportation Accidents Involving Propellant Tank Pressure Bursts. NASA Technical Memorandum X68277.

29 Tilton, 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.

30 U.S. Department of Transportation. Incidents Involving LPG and Ammonia, Computer Runs Prepared for Stone & Webster, 1981.

32 US Department of Commerce, Bureau of the Census, State and Metropolitan Area Book 1991, a Statistical Abstract Supplement, 1991.

33 US Department of Commerce, Bureau of the Census, 1990 P.L.94-171 Counts by Municipality - New York, 1991.

34 US Department of Commerce, Bureau of the Census, 1990 Census P.L.94-171 Counts by Municipality - Rhode Island, 1991.

35 US Department of Commerce, Bureau of the Census, Number of Inhabitants:

Connecticut, PC(1)-A8, 1971; PC80-1-A8, 1981.

TABLE 2.1-1 THIS TABLE HAS BEEN INTENTIONALLY DELETED TOWNS WITHIN 10 MILES OF MILLSTONE 1990 Population 1990 Population Density 1980-1990 Municipality Total (People/Square Mile) Change (%)

st Lyme 15,340 451 10.6 oton 45,144 1,442 9.9 cluding City) dyard 14,913 391 8.6 me 1,949 61 7.0 ntville 16,673 397 1.3 w London 28,540 5,189 -1.0 d Lyme 6,535 283 6.1 d Saybrook 9,552 637 2.9 terford 17,930 547 0.5 uthold, New York 19,836 394 3.5 shers Island)

TES:

ed on 1990 US Census of Population and Housing.

udes total 1990 population of all municipalities totally or partially within 10 miles of the site.

Total Population  % Change Municipality 1960 1970 1980 1990 1960-1970 1970-1980 1980-199 East Lyme 6,782 11,399 13,870 15,340 68.1 21.7 10.6 Groton 29,937 38,523 41,062 45,144 28.7 6.6 9.9 Ledyard 5,395 14,558 13,735 14,913 169.8 -5.7 8.6 Lyme 1,183 1,484 1,822 1,949 25.4 22.8 7.0 Montville 7,759 15,662 16,455 16,673 101.9 5.1 1.3 New London 34,182 31,630 28,842 28,540 -7.5 -8.8 -1.0 Old Lyme 3,068 4,964 6,159 6,535 61.8 24.1 6.1 Old Saybrook 5,274 8,468 9,287 9,552 60.6 9.7 2.9 Waterford 15,391 17,227 17,843 17,930 11.9 3.6 0.5 SOURCES:

1980 Census of Population, Number of Inhabitants, Connecticut, PC80-1-A8, 12/81.

1970 Census of Population, Number of Inhabitants, Connecticut, PC10-A8, 4/71.

1980 Final Population and Housing Counts, Connecticut, PHC80-V-8, 3/81.

1990 Census of Population and Housing, Connecticut, CPH-1-8, 7/91.

Distance to Plant Sector 0-1 1-2 2-3 3-4 4-5 5-6 6-7 7-8 8-9 9-10 Tota N 16 722 866 784 116 213 542 209 536 1,717 5,721 NNE 13 359 1,146 1,978 1,861 1,622 2,242 2,242 2,192 3,142 16,221 NE 165 455 839 3,888 10,584 7,752 8,164 8,129 911 1,961 42,646 ENE 22 455 292 4,963 971 7,186 3,748 3,748 1,008 2,662 24,354 E 0 636 413 1,804 193 552 0 63 1,434 904 5,999 ESE 0 143 36 0 0 0 0 0 115 214 508 SE 0 0 0 0 0 0 0 0 0 0 0 SSE 0 0 0 0 0 0 0 0 0 0 0 S 0 0 0 0 0 0 0 0 0 0 0 SSW 0 0 0 0 0 0 0 0 0 0 SW 0 0 14 0 0 0 0 0 0 0 14 WSW 0 0 489 91 86 312 472 158 0 74 1,682 W 0 178 1,061 1,014 440 763 476 562 881 408 5,782 WNW 0 476 1,165 1,946 346 239 211 1,654 509 4-17 6,981 NW 0 634 873 1,192 1,140 644 599 101 209 81 5,473 NNW 148 314 892 522 646 918 221 429 456 314 4860 Total 354 4,372 8,086 18,200 16,383 20,201 16,098 16,594 8,251 11,894 120,44

Distance to Plant Sector 0-1 1-2 2-3 3-4 4-5 5-6 6-7 7-8 8-9 9-10 Tota N 18 778 932 845 126 230 582 225 578 1,852 6,166 NNE 14 387 1,234 2,131 2,006 1,749 1,796 2,415 2,366 3,389 17,487 NE 179 489 905 4,191 11,441 7,359 8,802 8,765 983 2,115 46,203 ENE 24 492 314 5,352 1,045 7,746 4,041 3,285 1,087 2,870 26,256 E 0 685 444 1,944 208 597 0 68 1,546 975 6,467 ESE 0 154 39 0 0 0 0 0 125 233 551 SE 0 0 0 0 0 0 0 0 0 0 0 SSE 0 0 0 0 0 0 0 0 0 0 0 S 0 0 0 0 0 0 0 0 0 0 0 SSW 0 0 0 0 0 0 0 0 0 0 0 SW 0 0 14 0 0 0 0 0 0 0 14 WSW 0 0 528 98 92 336 509 169 0 78 1,810 W 0 192 1,144 1,093 473 821 513 606 950 436 6,228 WNW 0 514 1,255 2,118 373 258 227 1,783 548 448 7,524 NW 0 684 940 1,285 1,229 695 646 108 226 88 5,901 NNW 158 304 961 564 696 990 238 462 491 339 5,239 Total 393 4,715 8,710 19,621 17,663 21,781 17,354 17,886 8,900 12,823 129,84

Distance to Plant Sector 0-1 1-2 2-3 3-4 4-5 5-6 6-7 7-8 8-9 9-10 Total N 18 803 961 871 129 237 600 230 595 1,908 6,352 NNE 14 399 1,272 2,197 2,068 1,804 1,853 2,492 2,437 3,495 18,301 NE 184 504 930 4,321 11,767 8,617 9,074 9,036 1,013 2,180 47,626 ENE 25 507 324 5,518 1,078 7,988 4,166 3,387 1,119 2,960 27,072 E 0 707 458 2,005 215 616 0 70 1,593 1,005 6,669 ESE 0 159 41 0 0 0 0 0 138 255 593 SE 0 0 0 0 0 0 0 0 0 0 0 SSE 0 0 0 0 0 0 0 0 0 0 0 S 0 0 0 0 0 0 0 0 0 0 0 SSW 0 0 0 0 0 0 0 0 0 0 0 SW 0 0 15 0 0 0 0 0 0 0 15 WSW 0 0 54 102 95 346 525 175 0 79 1,867 W 0 198 1,179 1,126 440 488 847 530 625 443 6,417 WNW 0 529 1,294 2,184 385 266 234 1,838 566 461 7,757 NW 0 705 969 1,325 1,267 716 666 111 232 90 6,081 NNW 163 350 992 582 718 1,021 245 476 506 350 5,403 Total 404 4,861 8,980 20,231 18,210 22,458 17,893 18,440 9,180 13,226 133,883

Distance to Plant Sector 0-1 1-2 2-3 3-4 4-5 5-6 6-7 7-8 8-9 9-10 Tota N 19 828 990 899 133 243 620 236 613 1,968 6,549 NNE 14 411 1,310 2,264 2,132 1,860 1,909 2,569 2,513 3,602 18,584 NE 188 519 960 4,455 12,134 8,885 9,355 9,318 1,044 2,247 49,105 ENE 25 523 333 5,689 1,220 8,236 4,296 3,492 1,151 3,052 27,907 E 0 728 472 2,067 222 635 0 72 1,642 1,036 6,874 ESE 0 162 41 0 0 0 0 0 144 268 615 SE 0 0 0 0 0 0 0 0 0 0 0 SSE 0 0 0 0 0 0 0 0 0 0 0 S 0 0 0 0 0 0 0 0 0 0 0 SSW 0 0 0 0 0 0 0 0 0 0 0 SW 0 0 15 0 0 0 0 0 0 0 15 WSW 0 0 562 105 98 356 541 180 0 80 1,922 W 0 205 1,216 1,161 504 874 546 644 1,011 450 6,611 WNW 0 544 1,226 2,252 398 274 242 1,895 583 476 8,000 NW 0 727 998 1,365 1,308 738 687 114 239 93 6,269 NNW 168 361 1,023 600 738 1,053 253 491 523 362 5,572 Total 414 5,008 9,256 20,857 18,777 23,154 18,449 19,011 9,463 13,634 138,02

Distance to Plant Sector 0-1 1-2 2-3 3-4 4-5 5-6 6-7 7-8 8-9 9-10 Tota N 19 855 1,021 927 136 250 638 242 631 2,027 6,746 NNE 14 425 1,351 2,334 2,196 1,916 1,968 2,650 2,590 3,712 19,156 NE 193 535 990 4,592 12,510 9,160 9,644 9,606 1,075 2,315 50,620 ENE 26 539 343 5,866 1,145 8,492 4,428 3,598 1,188 3,147 28,772 E 0 751 487 2,132 229 655 0 73 1,692 1,068 7,087 ESE 0 167 43 0 0 0 0 0 151 281 642 SE 0 0 0 0 0 0 0 0 0 0 0 SSE 0 0 0 0 0 0 0 0 0 0 0 S 0 0 0 0 0 0 0 0 0 0 0 SSW 0 0 0 0 0 0 0 0 0 0 0 SW 0 0 15 0 0 0 0 0 0 0 15 WSW 0 0 580 108 101 366 558 185 0 81 1,979 W 0 212 1,254 1,197 520 901 561 663 1,043 458 6,809 WNW 0 560 1,377 2,323 409 281 249 1,956 602 490 8,247 NW 0 748 1,029 1,407 1,349 761 708 116 246 95 6,459 NNW 174 371 1,055 618 761 1,085 261 507 539 374 5,745 Total 426 5,163 9,545 21,504 19,356 23,867 19,015 19,596 9,757 14,048 142,27

1990 CENSUS Distance to Plant Sector 0-10 10-20 20-30 30-40 40-50 Total 5,721 22,283 26,357 32,610 18,658 105,629 E 16,221 34,824 23,730 27,465 35,598 137,838 42,848 9,444 11,334 29,987 199,334 292,947 E 24,354 23,914 16,498 43,001 99,721 207,488 5,999 10,712 7,992 10,920 0 35,623 E 508 0 0 836 0 1,344 0 0 807 0 0 807 E 0 0 2,420 0 0 2,420 0 1,614 13,541 0 0 15,155 W 0 2,443 12,569 14,807 4,498 34,317 14 938 22,042 8,252 143,933 175,179 SW 1,682 2,471 0 0 20,389 24,542 5,782 27,956 34,384 184,723 267,465 520,310 NW 6,981 12,474 27,895 148,259 259,824 455,433 W 5,473 6,215 31,331 191,767 365,578 600,364 W 4,860 8,809 17,850 115,424 78,820 225,762 tal 120,443 164,097 248,750 808,051 1,493,818 2,835,159

2000 PROJECTED Distance to Plant Sector 0-10 10-20 20-30 30-40 40-50 Total 6,166 24,028 28,707 35,404 20,273 114,578 E 17,487 37,551 25,721 29,926 38,135 148,820 46,203 10,183 12,196 31,611 206,940 307,133 E 26,256 25,744 17,633 45,998 105,848 221,509 6,467 11,497 8,553 11,687 0 38,204 E 551 0 0 895 0 1,446 0 0 878 0 0 878 E 0 0 2,635 0 0 2,635 0 1,759 14,742 0 0 16,501 W 0 2,660 13,688 16,122 4,897 37,367 14 1,022 24,000 8,985 156,725 190,746 SW 1,810 2,641 0 0 22,201 26,652 6,228 29,887 36,343 195,006 281,709 549,173 NW 7,524 13,340 29,762 156,623 273,153 480,402 W 5,901 6,660 33,435 200,205 380,339 626,540 W 5,239 9,492 19,194 121,620 83,732 239,277 tal 129,846 176,464 267,517 854,082 1,573,952 3,001,861

2010 PROJECTED Distance to Plant Sector 0-10 10-20 20-30 30-40 40-50 Total 6,352 24,773 300,056 36,785 21,101 119,067 E 18,031 38,716 26,730 31,421 39,720 154,618 47,626 10,499 12,626 32,221 210,368 313,340 E 27,072 26,652 18,530 48,258 109,494 230,006 6.669 11,986 8,981 12,272 0 39,908 E 593 0 0 940 0 1,533 0 0 920 0 0 920 E 0 0 2,761 0 0 2,761 0 1,847 15,445 0 0 17,292 W 0 2,788 14,344 16,896 5,132 39,160 15 1,073 25,151 9,416 164,248 199,903 SW 1,867 2,689 0 0 23,267 27,823 6,417 30,426 37,096 199,100 286,889 559,928 NW 7,757 13,590 30,311 159,776 278,156 489,590 W 6,081 6,807 34,052 202,762 384,902 634,604 W 5,403 9,778 19,778 123,964 85,735 244,658 tal 133,883 181,624 276,781 873,811 1,609,012 3,075,111

2020 PROJECTED Distance to Plant Sector 0-10 10-20 20-30 30-40 40-50 Total 6,549 24,541 31,470 38,219 21,963 123,742 E 18,584 39,916 27,784 32,989 41,349 160,622 49,105 10,825 13,051 32,748 213,221 318,950 E 27,907 27,557 19,336 50,343 112,285 234,428 6,874 12,452 9,376 12,811 0 41,513 E 615 0 0 981 0 1,596 0 0 965 0 0 965 E 0 0 2,894 0 0 2,894 0 1,939 16,184 0 0 18,123 W 0 2,922 15,033 17,707 5,379 41,041 15 1,127 26,355 9,869 172,131 209,497 SW 1,922 2,737 0 0 24,383 29,042 6,611 30,974 37,863 203,283 292,190 570,921 NW 8,000 13,844 30,871 162,992 283,254 498,961 W 6,269 6,957 37,678 205,354 389,518 642,776 W 5,572 10,070 20,382 126,369 87,794 250,187 tal 138,023 186,861 286,242 893,665 1,643,467 3,148,258

2030 PROJECTED Distance to Plant Sector 0-10 10-20 20-30 30-40 40-50 Total 6,746 26,332 32,953 39,716 22,860 128,607 E 19,156 41,155 28,879 34,637 43,058 166,885 50,620 11,159 13,494 33,286 219,112 324,671 E 28,772 28,495 20,176 52,519 115,158 245,120 7,087 12,937 9,789 13,375 0 43,188 E 642 0 0 1,024 0 1,666 0 0 1,011 0 0 1,011 E 0 0 3,033 0 0 3,033 0 2,036 16,957 0 0 18,993 W 0 3,062 15,755 18,558 5,637 43,012 15 1,183 27,619 10,342 180,394 219,553 SW 1,979 2,787 0 0 25,554 30,320 6,809 31,532 38,647 207,551 297,607 582,146 NW 8,247 14,102 31,441 166,276 288,449 508,515 W 6,459 7,110 35,317 207,981 394,192 651,059 W 5,745 10,373 21,003 128,835 89,919 255,875 tal 142,277 192,263 296,074 914,100 1,678,940 3,223,654

Distance to Plant Sector 0-1 1-2 2-3 3-4 4-5 5-6 6-7 7-8 8-9 9-10 Tota N 0 310 0 0 0 0 0 74 0 413 797 NNE 0 0 0 374 897 2,073 174 0 0 444 3,962 NE 0 0 636 210 697 1,352 1,542 534 0 0 4,971 ENE 0 0 0 2,501 0 888 0 1,043 1,609 266 6,307 E 0 181 0 0 0 1,330 0 0 183 0 1,805 ESE 0 0 0 0 0 0 0 0 0 0 68 SE 0 0 0 0 0 0 0 0 0 0 0 SSE 0 0 0 0 0 0 0 0 0 0 0 S 0 0 0 0 0 0 0 0 0 0 0 SSW 0 0 0 0 0 0 0 0 0 0 0 SW 0 0 0 0 0 0 0 0 0 0 0 WSW 0 0 0 0 0 0 0 0 0 0 0 W 0 0 0 0 0 0 263 0 864 0 1,127 WNW 0 0 345 0 0 0 0 0 0 0 345 NW 0 0 0 843 0 0 0 0 0 0 843 NNW 0 0 0 298 1,250 0 0 0 0 0 1,548 TOTAL 0 602 981 4,226 2,844 5,643 1,979 1,651 2,656 1,191 21,773 Note:

Includes student enrollment only.

Sources:

Connecticut Department of Education listing of schools: Telephone survey conducted in March 1992.

Distance to Plant Sector 0-1 1-2 2-3 3-4 4-5 5-6 6-7 7-8 8-9 9-10 Tota N 0 0 0 300 0 0 0 0 0 200 500 NNE 0 0 0 0 0 0 375 375 109 277 1,134 NE 0 0 375 80 831 0 375 375 0 0 2,036 ENE 0 0 0 0 8,800 5,500 820 0 0 0 15,120 E 0 0 0 0 0 0 0 0 0 0 0 ESE 0 0 0 0 0 0 0 0 0 0 68 SE 0 0 0 0 0 0 0 0 0 0 0 SSE 0 0 0 0 0 0 0 0 0 0 0 S 0 0 0 0 0 0 0 0 0 0 0 SSW 0 0 0 0 0 0 0 0 256 0 256 SW 0 0 0 0 0 0 0 0 0 0 0 WSW 0 0 0 0 0 0 0 0 0 0 0 W 0 0 0 0 0 0 0 0 0 0 0 WNW 0 0 0 0 0 0 125 125 0 0 250 NW 0 500 0 843 0 0 125 125 0 0 750 NNW 0 0 0 0 0 0 0 0 0 0 0 TOTAL 0 500 375 380 9,631 5,500 1,820 1,000 363 477 20,046 Note: Firms with 50 employees or more. Excludes plant employee population.

Sources: Telephone survey conducted in March 1992.

2030 PROJECTED TOTAL ANNUAL SUMMER DAILY FACILITY LOCATION ATTENDANCE ATTENDANCE te Parks:

Bluff Point ENE/E 6-8 97,641 490

• Fort Griswold ENE 5-6 58,965 200

• Haley Farm ENE/E 7-9 11,675 60

• Harkness Memorial E 2-3 157,962 790

• Rocky Neck W 3-5 412,495 2,360 **

te Forests:

Nehantic WNW/NNW 7-10 81,146 400

• es:

aily summer attendance based on 90% of yearly attendance from April through September.

ncludes campers from April 15 to September 15.

rce:

e of Connecticut DEP - Office of Parks and Forests, 1990 Park Attendance.

DISTRIBUTIONS DIRECTION 1990 CENSUS 2030 PROJECTED N 1,298 1,536 NNE 903 1,065 NE 1,144 1,351 ENE 768 909 E 760 899 ESE 179 212 SE 0 0 SSE 0 0 S 0 0 SSW 0 0 SW 3 3 WSW 429 506 W 1,025 1,211 WNW 1,046 1,233 NW 1,167 1,377 NNW 1,124 1,327 TAL LPZ 9,846 11,629 rces:

0 Census of Population and Housing necticut Office of Policy and Management, Interim Population Projections Series 91.1, 4/91

EMPLOYMENT DIRECTION SCHOOL EMPLOYMENT N 310 0 NNE 0 0 NE 0 75 ENE 0 0 E 292 0 ESE 0 0 SE 0 0 SSE 0 0 S 0 0 SSW 0 0 SW 0 0 WSW 0 0 W 0 0 WNW 345 0 NW 0 500 NNW 0 0 TOTAL LPZ 947 0 es:

1-1992 Student Enrollment ms with 50 employees or more.

rces:

phone survey conducted in March 1992; Connecticut Department of Education school listing.

CENSUS POPULATION AREA 1990 POPULATION dgeport - Milford, CT PMSA 443,722 stol, CT PMSA 79,488 l River, MA-RI PMSA 157,272 rtford, CT PMSA 767,899 w Haven - Meriden, CT MSA 530,240 ssau - Suffolk, NY PMSA 2,609,212 w Britain, CT PMSA 148,188 w London - Norwich, CT-RI MSA 266,819 vidence, RI PMSA 654,869 terbury, CT MAS 221,629 ddletown, CT PMSA 90,320 es:

SA - Primary Metropolitan Statistical Area.

A - Metropolitan Statistical Area.

al population of metropolitan areas completely or only partially within 50 miles of the site.

STATE MUNICIPALITY 1990 POPULATION nnecticut Branford 27,603 Bristol 60,640 Cheshire 25,684 East Hartford 50,452 East Haven 26,144 Enfield 45,532 Glastonbury 27,901 Groton 45,144 Hamden 52,434 Hartford 139,739 Manchester 51,618 Meriden 59,479 Middletown 42,762 Milford 49,938 Naugatuck 30,625 New Britain 75,491 New Haven 130,474 New London 28,540 Newington 29,208 Norwich 37,371 Shelton 35,418 Southington 38,518 Stratford 49,389 Vernon 29,841 Wallingford 40,822 Waterbury 108,961 West Hartford 60,110 West Haven 54,021 Wethersfield 25,651 Windsor 27,817

STATE MUNICIPALITY 1990 POPULATION ode Island Coventry 31,083 Cranston 76,060 Johnston 26,542 Newport 28,227 Warwick 85,427 West Warwick 29,268 w York Brookhaven 407,779 Southampton 44,976 es: Municipalities with 25,000 people or more. Municipalities completely or only partially hin 50 miles.

rce: 1990 U.S. Census of Population and Housing.

Sector 0-1 1-2 2-3 3-4 4-5 5-6 6-7 7-8 8-9 9-10 Average N 82 1,226 883 571 66 99 212 71 161 460 292 NNE 66 610 1,168 1,440 1,054 751 653 762 657 843 827 NE 842 772 855 2,830 5,993 3,591 3,200 2,761 273 526 2,183 ENE 112 772 298 3,612 550 3,328 1,469 1,035 302 714 1,241 E 0 1,080 421 1,313 109 256 0 21 430 242 306 ESE 0 243 37 0 0 0 0 0 34 57 26 SE 0 0 0 0 0 0 0 0 0 0 0 SSE 0 0 0 0 0 0 0 0 0 0 0 S 0 0 0 0 0 0 0 0 0 0 0 SSW 0 0 0 0 0 0 0 0 0 0 0 SW 0 0 14 0 0 0 0 0 0 0 1 WSW 0 0 498 66 49 145 185 54 0 20 86 W 0 302 1,082 738 249 353 186 191 264 109 295 WNW 0 808 1,118 1,429 196 111 83 562 153 112 356 NW 0 1,076 890 868 646 298 235 34 63 22 279 NNW 755 533 909 380 366 425 87 146 137 84 248 Average 116 464 515 828 580 585 394 352 155 199 384 Source: 1990 Census of Population

Sector 0-1 1-2 2-3 3-4 4-5 5-6 6-7 7-8 8-9 9-10 Average N 97 1,452 1,041 675 77 116 250 82 189 544 344 NNE 71 722 1,377 1,700 1,243 887 771 900 776 995 976 NE 985 908 1,009 3,345 7,084 4,243 3,780 3,263 322 621 2,579 ENE 133 915 350 4,272 648 3,933 1,736 1,222 356 844 1,466 E 0 1,275 496 1,553 130 303 0 25 507 286 361 ESE 0 284 44 0 0 0 0 0 45 75 33 SE 0 0 0 0 0 0 0 0 0 0 0 SSE 0 0 0 0 0 0 0 0 0 0 0 S 0 0 0 0 0 0 0 0 0 0 0 SSW 0 0 0 0 0 0 0 0 0 0 0 SW 0 0 15 0 0 0 0 0 0 0 1 WSW 0 0 591 79 57 170 219 63 0 22 101 W 0 360 1,278 872 294 417 220 225 313 123 347 WNW 0 951 1,404 1,692 232 130 98 664 180 131 420 NW 0 1,270 1,049 1,025 764 352 278 39 74 25 329 NNW 888 630 1,075 450 431 503 102 172 162 100 293 Average 136 548 608 979 685 691 466 416 183 235 453 Source: CT Office of Policy and Management, Interim Population Projections Series 91.1, 4/91.

(PEOPLE PER SQUARE MILE)

Sector 0-10 10-20 20-30 30-40 40-50 Average N 292 378 269 237 106 215 NNE 827 591 242 200 202 281 NE 2,183 160 116 218 1,129 597 ENE 1,241 406 168 313 564 423 E 306 182 81 79 0 73 ESE 26 0 0 6 0 3 SE 0 0 8 0 0 2 SSE 0 0 25 0 0 5 S 0 27 138 0 0 31 SSW 0 41 128 108 25 70 SW 1 16 225 60 815 357 WSW 86 42 0 0 115 50 W 295 475 350 1,345 1,514 1,061 WNW 356 212 284 1,079 1,471 928 NW 279 106 319 1,396 2,070 1,224 NNW 248 150 182 840 446 460 Average 384 174 158 368 528 361 rce: 1990 Census of Population and Housing.

Sector 0-10 10-20 20-30 30-40 40-50 Average N 344 447 336 289 129 262 NNE 976 699 294 252 244 340 NE 2,579 190 138 242 1,224 662 ENE 1,466 484 206 382 652 499 E 361 220 100 97 0 88 ESE 33 0 0 7 0 3 SE 0 0 10 0 0 2 SSE 0 0 31 0 0 6 S 0 35 173 0 0 39 SSW 0 52 161 135 32 88 SW 1 20 81 75 1,021 447 WSW 101 47 0 0 145 62 W 347 536 394 1,511 1,685 1,187 WNW 420 240 320 1,210 1,633 1,036 NW 329 121 360 1,514 2,232 1,327 NNW 293 176 214 938 509 522 Average 453 204 189 416 594 410 Source: CT Office of Management, interim Population projections, Series 91.1, 4/91.

SQUARE MILE)

Sector 0-10 10-20 20-30 30-40 40-50 Average N 292 378 269 237 106 215 NNE 827 591 242 200 202 281 NE 2,183 160 116 218 1,129 597 ENE 1,241 406 168 313 564 423 E 306 182 81 79 0 73 ESE 26 0 0 6 0 3 SE 0 0 8 0 0 2 SSE 0 0 25 0 0 5 S 0 27 138 0 0 31 SSW 0 41 128 108 25 70 SW 1 16 225 60 815 357 WSW 86 42 0 0 115 50 W 295 475 350 1,345 1,514 1,061 WNW 356 212 284 1,079 1,471 928 NW 279 106 319 1,396 2,070 1,224 NNW 248 150 182 840 446 460 Average 384 174 158 368 528 361 Source: 1990 Census of Population and Housing.

SQUARE MILE)

Sector 0-10 0-20 0-30 0-40 0-50 N 344 421 374 337 262 NNE 976 768 505 394 340 NE 2,579 787 426 346 662 ENE 1,466 730 438 414 499 E 361 255 169 138 88 ESE 33 8 4 5 3 SE 0 0 6 3 2 SSE 0 0 17 10 6 S 0 26 108 60 39 SSW 0 39 107 119 88 SW 1 15 163 125 447 WSW 101 61 27 15 62 W 347 488 436 906 1,187 WNW 420 285 305 701 1,036 NW 329 173 277 818 1,327 NNW 293 205 210 529 522 Average 453 226 223 307 410

cility Location Approximate Approximate Sector Number Persons Distance From Employed or Site Miles Stationed ustrial Dow Chemical Corp Ledyard 115 10+ NNE Pfizer Corporation Groton 3,000 4.9 ENE Electric Boat (Division Groton 12,000 5 ENE of General Dynamics) nsportation Groton/New London Groton 153 6 ENE Airport (Trumbull)

New London New London 20 4 NE Transportation Center litary U.S. Navy Submarine Groton 10,300 7 NE Base U.S. Cost Guard New London 1,260 5.6 NE Academy Camp Rowland East Lyme 16 2 NW Stones Ranch Military East Lyme 14 7 NW Reservation ustrial Related Facilities Hess Oil Corporation Groton 14 5 ENE Hendel Petroleum Co. Waterford 75 2.5 NE Montville Station Montville 67 10 NNE Electric Generation Plant

PRODUCING SIGNIFICANT MISSILES Average Number of Cars per Train Containing Hazardous Approximate Number of Hazardous Material Materials Cars per Year Propane 2.20 44 Anhydrous Ammonia 0.266 5 Total 2.466 49

FIGURE 2.1-1 GENERAL SITE LOCATION FIGURE 2.1-2 GENERAL VICINITY Withheld under 10 CFR 2.390 (d) (1)

FIGURE 2.1-3 SITE LAYOUT

Withheld under 10 CFR 2.390 (d) (1)

FIGURE 2.1-4 SITE PLAN

FIGURE 2.1-5 TOWNS WITHIN 10 MILES FIGURE 2.1-6 POPULATION SECTORS FOR 0 - 10 MILES FIGURE 2.1-7 POPULATION SECTORS FOR 0 - 50 MILES FIGURE 2.1-8 ROADS AND FACILITIES IN THE LPZ FIGURE 2.1-9 LPZ POPULATION SECTORS DISTRIBUTION FIGURE 2.1-10 INSTRUMENT LANDING PATTERNS AT TRUMBULL AIRPORT FIGURE 2.1-11 AIR LANES ADJACENT TO MILLSTONE POINT FIGURE 2.1-12 NEW LONDON COUNTY - STATE HIGHWAYS AND TOWN ROADS rmation regarding meteorology is presented in Section 2.3 of the Millstone Unit 3 Final ety Analysis Report (Reference 2.2-1). With the exceptions given below, that information is rporated herein by reference.

1 REGIONAL CLIMATOLOGY e Section 2.3.1 of the Millstone 3 Final Safety Analysis Report of Reference 2.2-1).

2 LOCAL METEOROLOGY e Section 2.3.1 of the Millstone 3 Final Safety Analysis Report of Reference 2.2-1).

2.1 Potential Influence of the Plant and Its Facilities on Local Meteorology lstone Unit Number 1 used a once-through cooling water system, discharging its cooling water an existing quarry into which Units 2 and 3 also discharge and thence into Long Island nd. Thin wisps of steam fog occasionally form over the quarry and less frequently over the harge plume during the winter months, depending on tidal conditions and temperature erences between air and water. This fog dissipates rapidly as it moves away from the warm er area. Because the maximum discharge plume (defined by the 1.5F isotherm of temperature erential when all three Millstone units were at full power) is approximately an ellipse of 1500 er by 800 meters, the extent of the steam fog is negligible. With the permanent shutdown of lstone Unit Number 1, this maximum discharge plume size is further reduced.

2.2 Local Meteorological Conditions for Design and Operating Bases.

2.2.1 Design Basis Tornado specifications for the Millstone Unit Number 1 design basis tornado are:

Rotational velocity 300 mph Translational velocity 60 mph Total pressure drop 2.25 psi Rate of pressure drop 1.2 psi/sec 3 ON SITE METEOROLOGICAL MEASUREMENTS PROGRAM Millstone Site is served by a common meteorological tower, located south of Millstone Unit mber 1. The meteorological tower is capable of measuring wind speed, direction, and air perature at various heights. For details regarding the capability of the On Site Meteorological asurements program, see Section 2.3.3 of the Millstone 3 Final Safety Analyses Report erence 2.2-1, with the exception that Millstone Unit Number 1 no longer has the data rding systems and data recording capability to display parameters transmitted by modem/

ne line from the instrument shack at the base of the tower to the control room area for display he plant process computer described in Section 2.3.3.3.

4.1 Objective idents could result in short-term releases of radioactivity from several possible venting points.

ospheric diffusion factors (/Q) based on site meteorological data are calculated at the lusion area boundary (EAB) and low population zone (LPZ) for each downwind sector for h release point. The diffusion factors are calculated for different release time periods ending on the length of the release. These diffusion factors are used in the calculation of ological consequences of the releases.

4.2 Calculations 4.2.1 Venting Point and Receptor Locations LPZ is taken to be 3860 m in all sectors from any release point.

4.2.2 Models ident /Qs were calculated using the basic methods of Regulatory Guide 1.145. /Q values the Millstone Unit 2 Control Room due to ground level releases were calculated using the hods of Murphy and Campe. (Reference 2.2-2).

4.3 Results calculated /Qs used in design basis accident (DBA) radiological consequence calculations presented with the list of assumptions in Chapter 5.

5 LONG-TERM (ROUTINE) DIFFUSION ESTIMATES 5.1 Objective levels of radioactivity are routinely released on a continuous basis from the Unit Number 1 P exhaust point and the SFPI ventilation exhaust point. Atmospheric diffusion factors (/Q) ed on site meteorological data are calculated for various downwind receptor locations of rest. The meteorological data is used to calculate the dose consequences to the public from ine airborne effluents. The calculated doses are submitted periodically to the Nuclear ulatory Commission (NRC).

5.2 Calculations 5.2.1 Venting Point and Receptor Locations tine releases of radioactivity in gaseous effluents are vented from the Unit Number 1 BOP aust point and the SFPI ventilation exhaust point.

culations are performed on a periodic basis using the actual meteorology for this period.

5.2.3 Models values are ground level dispersion factors, and releases are modeled using a conventional ssian plume model.

6 REFERENCES 1 Millstone Unit 3, Final Safety Analysis Report, Section 2.3-Meteorology.

2 Murphy, 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.

rmation regarding hydrologic engineering is presented in Section 2.4 of the Millstone 3 Final ety Analysis Report (Reference 2.3-1). With the exceptions given below, that information is rporated herein by reference.

1 HYDROLOGIC DESCRIPTION e Section 2.4.1 of the Millstone 3 Final Safety Analysis Report, Reference 2.3-1.)

2 SITE AND FACILITIES lstone Point is located on the north shore of Long Island Sound. To the west of the site is ntic Bay and to the east is Jordon Cove. Figure 2.3-1 shows the general topography of the lstone area. The site grade elevation for Millstone Unit Number 1 varies from 14 feet to above eet mean sea level (MSL). Section 2.3.3.2 discusses the probable maximum hurricane used to ulate maximum water levels.

3 FLOODS s section reviews the flood history in the vicinity of Millstone Point, flood design siderations, and the effects of local intense precipitation.

3.1 Flood History oding near the site has historically been caused by hurricanes. The maximum historical ding was the result of a hurricane on September 21, 1938, which produced a flood level of 9.7 MSL at New London, Connecticut.

only sources of flooding that could affect Millstone Unit Number 1 are direct rainfall and m surges.

3.2 Flood Design Considerations controlling event for flooding at the Millstone site is a storm surge resulting from the urrence of a probable maximum hurricane (PMH) (see Section 2.3.6). The maximum water level is +18.11 feet MSL, and the associated wave run up is +22.3 feet MSL.

pter 3 describes the flooding protective features at Millstone Unit Number 1.

3.3 Effect of Local Intense Precipitation iscussion on the development of the probable maximum precipitation (PMP) for the site may ound in Section 2.3.2 of Reference 2.3-1.

udy was performed to determine the impact of the PMP intensity on the plant roof structures.

radwaste disposal building, intake structure, radwaste/control building, and southwest corner

turbine building, reactor building, warehouse, and heating/ventilation area roofs credit ppers to assure that the loads due to a PMP will remain below the roof design live loads.

thheld under 10 CFR 2.390 (d) (1) ing a PMP scenario, in-leakage through door openings could occur once the flood depths eed door sill elevations. Secured external and internal doors will have a tendency to limit or trol the amount of in-leakage.

4 PROBABLE MAXIMUM FLOOD (PMF) ON STREAMS AND RIVERS e Section 2.4.3 of Reference 2.3-1, the Millstone Unit 3 Final Safety Analysis Report.)

5 POTENTIAL DAM FAILURE, SEISMICALLY INDUCED e Section 2.4.4 of Reference 2.3-1, the Millstone Unit 3 Final Safety Analysis Report.)

6 PROBABLE MAXIMUM SURGE AND SEICHE FLOODING 6.1 Probable Maximum Winds and Associated Meteorological Parameters meteorologic characteristics used to calculate the probable maximum storm surge at the lstone Point site are those associated with the PMH as reported by the U.S. National Oceanic Atmospheric Administration (NOAA) in their unpublished report HUR 7-97. HUR 7-97 cribed the PMH as a hypothetical hurricane having that combination of characteristics ch will make it the most severe that can probably occur in the particular region involved. The icane should approach the point under study along a critical path and at an optimum rate of vement. Actually, nine different PMH storm patterns can be constructed using wind speed, m size and forward speed parameters given in HUR 7-97 in various combinations. The storm, ch would cause the maximum surge buildup at the entrance to Long Island Sound is one with rge radius to maximum wind and a slow speed of translation. Pertinent parameters are lated below:

Central Pressure Index The minimum surface atmospheric pressure in the eye of the hurricanes.

(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 atmospheric pressure at the outer edge of the hurricane where the hurricane circulation ends.

hough other parametric combinations give a higher wind speed, this particular combination ds the highest surge.

6.2 Surge and Seiche Water Levels hough frontal storms and squall lines cause tidal flooding in the Millstone Point area, by far the t severe flooding has resulted from hurricanes. For this reason, the PMH as defined in tion 2.3.6.1 was used to compute the design storm surge level at the site. The calculated total e height or still water level considers the wind setup, the water level rise due to barometric sure drop, the astronomical tide and forerunner or initial rise.

maximum still water level is +18.11 feet, and the associated wave run up elevation is +22.3 MSL.

6.3 Wave Action ve characteristics are dependent upon wind speed and duration, fetch length, and water depth.

lstone Point is sheltered from the direct onslaught of open ocean waves by Long Island.

he time of the peak surge, the wind is from the southeast direction and the wave attack would long the large axis of the point. Thus the intake structure, and the southeast portions of the ctor and Turbine Generator Buildings are primarily involved.

6.4 Resonance e Section 2.4.5 of the Millstone Unit 3 Final Safety Analysis Report Reference 2.3-1.)

he time of the peak surge, the wind is from the southeast direction and the wave attack would long the large axis of the point. Thus, the southeast portions of the Reactor Building would be arily involved.

6.6 Probable Maximum Tsunami Flooding e Section 2.4.6 of the Millstone Unit 3 Final Safety Analysis Report Reference 2.3-1.)

7 ICE EFFECTS re is no available history of ice or ice jams in Niantic Bay.

8 COOLING WATER CANALS AND RESERVOIRS e Section 2.4.8 of the Millstone Unit 3 Final Safety Analysis Report Reference 2.3-1.)

9 CHANNEL DIVERSIONS e Section 2.4.9 of the Millstone Unit 3 Final Safety Analysis Report Reference 2.3-1.)

10 FLOODING PROTECTION REQUIREMENTS e.

11 LOW WATER CONSIDERATIONS 11.1 Low Flow in Rivers and Streams ce Millstone Unit Number 1 does not depend on either rivers or streams as a source of cooling er, this section is not applicable.

11.2 Low Water Resulting from Surges, Seiches, or Tsunamis effect at Millstone Unit 1.

12 DISPERSION, DILUTION, AND TRAVEL TIMES OF ACCIDENTAL RELEASES OF LIQUID EFFLUENTS SURFACE WATERS.

e Section 2.4.12 of the Millstone Unit 3 Final Safety Analysis Report Reference 2.3-1.)

13 GROUNDWATER e Section 2.4.13 of the Millstone Unit 3 Final Safety Analysis Report Reference 2.3-1.)

ion Procedures address necessary precautions and actions to take in the event of anticipated icane, tornado, or flood conditions.

15 REFERENCES 1 Millstone Unit 3, Final Safety Analysis Report, Section 2.4 - Hydrologic Engineering.

2 Letter from J. J. Shea to W. G. Counsil, Millstone Nuclear Power Station, 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.

FIGURE 2.3-1 TOPOGRAPHY IN THE VICINITY OF MILLSTONE POINT 1 BASIC GEOLOGIC AND SEISMIC INFORMATION rmation regarding the geologic and seismic qualities of the Millstone site is presented in tion 2.5.1 of the Millstone Unit Number 3 Final Safety Analysis Report (Reference 2.4-1).

t information is incorporated herein by reference.

2 VIBRATORY GROUND MOTION rmation regarding vibratory ground motion at the Millstone site is presented in Section 2.52 eference 2.4-1. With the exceptions given below, that information is incorporated herein by rence.

2.1 Safe Fuel Storage Earthquake design of the plant is such that spent fuel pool remain intact during a ground motion of 0.17 g.

3 SURFACE FAULTING 3.1 Geologic conditions of the Site tion 2.5.1.2 of Reference 2.4-1 discusses the stratigraphy, structural geology, and geologic ory of the site are in detail.

3.2 Evidence of Fault Offset published geologic maps which include the site area do not indicate the presence of faulting.

iscussion of faults discovered during excavation of the Millstone Unit Number 3 site can be nd in Section 2.5.3.2 of Reference 2.4-1. This discussion can be considered typical for the re Millstone site.

3.3 Earthquakes Associated with Capable Faults re is no evidence of capable faults within the five-mile radius of the site. The majority of the ificant seismic activity has been associated with the White Mountain Plutonic Province. Some vity has been associated with the Ramapo fault system (Reference 2.4-2); however, the fault is considered capable (Reference 2.4-3).

3.4 Investigation of Capable Faults re are no capable faults within the site area. The faults uncovered in the excavation are ussed in Section 2.5.3.2 of Reference 2.4-1.

re has been no spatial correlation between earthquakes and folds in the site region. Some elation has been suggested with the Ramapo fault in New York and New Jersey. However, the apo is not considered capable (Reference 2.4-3).

3.6 Description of Capable Faults re are no capable faults within five miles of the site.

3.7 Zone Requiring Detailed Faulting Investigation ven incapable fault zones were uncovered during excavation at the Millstone Unit Number 3

. These faults have been mapped in detail and are discussed in Section 2.5.3.2 of erence 2.4-1.

3.8 Results of Faulting Investigation re is no evidence of capable faulting within the five mile radius of the site. The faults at the are related to the rifting associated with the Triassic-Jurassic Period or older, with the last vity occurring approximately 142 million years ago.

4 STABILITY OF SUBSURFACE MATERIALS AND FOUNDATIONS information on the stability of subsurface materials and foundations is available from the avation for Millstone Unit Number 1. A discussion of this subject for the Millstone Unit mber 3 excavation can be found in Reference 2.4-1. This information can be considered typical the Millstone site.

5 STABILITY OF SLOPES stability of slopes at the Millstone site was evaluated in Reference 2.4-4, wherein it was cluded that there are no natural or man-made slopes at the site that could be or become table such as to affect safety related structures, systems or components.

6 EMBANKMENTS AND DAMS embankments or dams have been constructed at the Millstone site.

7 REFERENCES 1 Millstone Unit 3, Final Safety Analysis Report, Section 2.5, Geology, Seismology, and Geotechnical Engineering.

2 Aggarwal, Y.P. and Sykes, L.R. 1978. Earthquakes, Faults, and Nuclear Power Plants in Southern New York and Northern New Jersey. Science, Vol. 200, Number 430, pages 425-429.

50-3, 50-247, and 50-285) ALAB-436.

4 Nuclear 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.

DESIGN CRITERIA 1 CONFORMANCE WITH 10 CFR 50 APPENDIX A GENERAL DESIGN CRITERIA 1.1 Summary Discussion General Design Criteria (GDC) for Nuclear Power Plants as listed in Appendix A to 10 CFR were effective May 21, 1971 and subsequently amended July 7, 1971.

lstone Unit Number 1, was issued a provisional operating license (POL) on October 7, 1970, is not obligated to comply with the GDC (Reference 3.1-3). Therefore, Millstone Unit mber 1, is not required to seek exemptions for those areas where it does not comply with the C. An evaluation of the design bases of the Millstone Nuclear Unit Number 1, as compared to GDCs, was performed in support of the application for a full term operating license (FTOL),

Reference 3.1-1. It was concluded therein that Millstone Unit Number 1 satisfies and is in pliance with the intent of the GDCs. Nevertheless, it should be noted that this comparison and clusion was not a commitment to meet all of the current GDCs or even to meet the intent of current GDCs. Instead, the Reference 3.1-1 comparison determined the degree of compliance h the GDCs at that time. Also, compliance is demonstrated based upon those interpretations in ct at the time the specific licensing question, or issue, was being addressed.

1.2 Systematic Evaluation Program and Three Mile Island Evaluations of General Design Criteria ing the systematic evaluation program (SEP) initiated by the NRC in 1977, a large number of eric and plant specific safety concerns were addressed and resolved (Reference 3.1-2). Many hese SEP issues, and later issues which arose from the Three Mile Island (TMI) accident, olved a consideration of the NRC GDC affected by a specific issue and how the plant design pared to the criteria. A compilation of this more recent evaluation of specific safety concerns the affected GDC are listed in Table 3.1-1.

2 CLASSIFICATION OF STRUCTURES, SYSTEMS, AND COMPONENTS 2.1 Seismic Classification Code of Federal Regulations requires that structures, systems, and components important to ty shall be designed to withstand the effects of earthquakes without loss of capability to orm necessary safety functions. 10 CFR 100, Appendix A further defines a safe shutdown hquake (SSE) and the structures, systems and components required to remain functional, as e plant features necessary to ensure:

(1) The integrity of the reactor coolant pressure boundary, (2) The capability to shut down the reactor and maintain it in a safe shutdown condition, and

exposures of 10 CFR 100.

ulatory Guide 1.29, Revision 3, describes an acceptable method for identifying and sifying those plant features that should be designed to withstand the effects of an SSE.

plant structures and equipment, including their foundations and supports, are divided into two ctural safety categories:

mic Class I: structures and equipment whose failure could cause significant release of radioactivity or which are vital to the removal of decay heat.

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

y the Reactor Building at and below elevation 108 feet 6 inches, the fuel pool liner and the nt fuel racks remain Seismic Class I in the permanently defueled condition. The Reactor lding houses, protects and supports the spent fuel pool. It supports maintenance of the fuel figuration in the fuel pool, provides protection from external hazards and supports ntenance of water in the fuel pool to a depth necessary to ensure the irradiated fuel is always ersed. The fuel storage racks are designed to assure subcriticality in the fuel pool and are gned to withstand the anticipated earthquake loadings as Class I structures.

Reactor Building structure above elevation 108 feet 6 inches (enclosure) is classified as mic Class II in the permanently defueled plant condition. The Reactor Building above ation 108 feet 6 inches provides a weather enclosure for the spent fuel pool and supports the tor building overhead crane. However, it has no structural function in providing support for spent fuel pool. Since the enclosure is no longer credited to provide secondary containment AR Section 3.1.2.2) and since its failure during a seismic event could adversely affect the nt fuel pool and its contents or adjacent safety related SSCs, the seismic design of the losure is categorized as Seismic II/I and is further discussed in Section 3.1.6.

mantlement of Seismically Designed Structures, Systems and Components SSCs designated as Seismic Class I in the permanently defueled condition, the following eria apply prior to performing dismantlement operations:

(1) Downgrading seismic classification of components shall be performed in accordance with appropriate engineering and design procedures and processes.

(2) When downgrading seismic classification of an SSC, a 10 CFR 50.59 evaluation shall be performed if:

a. the seismic classification is described in the DSAR, or

incident or an accident with off site doses exceeding the doses from the design basis accident.

(3) When downgrading seismic classification of an SSC, a 10 CFR 50.54 evaluation shall be performed if the structure classification is described in the Quality Assurance Program (QAP).

(4) When downgrading seismic classification of an SSC, a 10 CFR 50.59 evaluation shall be performed if, during a seismic event, its failure has the potential to drain the fuel pool water level lower than 9 feet above the active fuel.

2.2 Safety Related Classification lear plant SSC have traditionally been classified as safety related in accordance with 10 R 50.2 and in 10 CFR 100, Appendix A, Section III, if they are relied upon to remain ctional during and following design basis events to assure:

The integrity of the reactor coolant pressure boundary, The capability to shut down the reactor and 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 comparable to the applicable guideline exposures set forth in 10 CFR 50.34(a)(1) or 10 CFR 100.11.

arly, the first two parts of the safety related definition (reactor coolant pressure boundary, and ability to achieve and maintain safe shutdown) do not apply to a permanently defueled plant, en the license restrictions of 10 CFR 50.82. The third part of the safety related definition ident consequences comparable to 10 CFR 100 guidelines) is dependent on the results of new gn basis accident analysis assumptions and results developed to address the existing defueled t condition. SSC that are required to protect workers and the public from the consequences of gn basis events may need to remain classified as safety related.

es and consequences of potential accidents were reanalyzed and it was concluded that the only aining accident is a fuel handling accident. This accident was analyzed assuming no ondary containment or standby gas treatment system in operation, with a puff ground level ase. The resulting off site radiological exposure was determined to be significantly less than guideline exposures set forth in 10 CFR 50.34(a)(1) and 10 CFR 100.11. Therefore, no SSC is uired to be safety related to prevent or mitigate the consequences of the only remaining dent, except to account for assumptions inherent in the analysis.

act damage, did not change and that the fuel pool water remained in place. This implies that e is no failure of either the fuel pool structure or the fuel racks. Since these are passive ctural items, assumption of failure is not required as long as the items are safety related and gned to withstand these loads. Therefore, the fuel pool and supporting structure, fuel pool r, and the fuel racks must be considered as safety related to support the assumptions made in accident analysis. No other components, systems or structures meet this criterion.

2.3 Non-Safety Related Plant Functions Maintained in the Defueled Condition ddition to the Safety Related criterion above, other non-safety related plant functions must be ntained in the defueled condition. The following criteria were used to determine which SSC e still required:

Criterion 1 Is the SSC associated with storage, control or maintenance of nuclear fuel in a safe condition; or handling of radioactive waste?

Criterion 2 Is the SSC program associated with radiological safety?

Criterion 3 Is the SSC associated with an outstanding commitment to the regulators which remains applicable to storage, control, or maintenance of nuclear fuel in a safe condition; or handling of radioactive waste or radiological safety?

Criterion 4 Does the SSC satisfy a requirement based on regulations governing management of nuclear fuel or radioactive materials, including any SSC which is independently required by the License or Technical Specifications?

se criteria were applied to all Millstone Unit Number 1 SSC. A positive response to any erion, including the Safety Related criterion, results in the group of those SSC which must ain functioning.

2.4 SSCs Important to the Defueled Condition SSC that must remain functional will be maintained in accordance with applicable Millstone t Number 1 procedures or quality processes. Commitments exist for augmented quality related Fire Protection (FPQA), and Radwaste (RWQA). These requirements would apply to the ropriate portions of the SSC which meet the criteria above.

-nuclear safety standards apply to other SSC that do not fall under existing quality processes.

wever, to provide added assurance of adequate reliability for non-safety related SSC that are ortant to safeguarding the heath and safety of the public and workers, another augmented lity classification is being developed. This classification, Important to the Defueled Condition DC), implements management expectations but does not satisfy any regulatory requirement, will apply to selected systems and components which perform the following functions:

Storage, control, maintenance or handling of radioactive waste, if not already RWQA Radiological safety tems and components were reviewed for these functions. Note that application of these ctions differs from the 4 criteria in that requirements apply only to the primary SSC and are extended to supporting systems, equipment or structures. The intent of the ITDC augmented lity is to increase reliable operation of the system(s) primarily responsible for performing each ction. Acceptable performance of the supporting SSC are demonstrated during routine ration and/or periodic testing of the ITDC SSC.

itionally, certain regulatory requirements to which the licensee made a licensing commitment go beyond the functional scope of an SSC (e.g., Emergency Plan, Security Plan, Quality urance Program, etc.). These commitments and legal requirements were also considered in the assification process.

horizations, Restrictions and Limitations on use of the SSC reclassification criteria SSC reclassification criteria is used as a basis to change various Millstone Unit Number 1 cedures and programs, provided that the change involves an SSC that is non-ITDC and, vided that plant procedures contain an acceptable method for approving the change. The owing kinds of soft changes associated with non-ITDC SSCs are allowed:

SSC classifications, drawings, calculations, procedures, nonconforming items and corrective actions, external industry operating experience reports, commitments, open work orders (in process at the time the decision was made to decommission the plant) the application of 10 CFR 50 Appendix B criteria provided it does not represent a reduction in commitment.

of these criteria does not authorize:

1 CFR 50.82)

The physical removal/disassembly of existing SSCs, or the installation of new SSCs.

However, it may provide the basis for initiating such a change.

Changes to Technical Specification requirements.

Changes to regulations, license conditions, rules, and permits until such time that relief is granted by the regulating authority. However, it may provide the basis for requesting relief from the regulations, license conditions, rules, and permits.

Changes to commitments. Application of the commitment change process is required to change commitments.

Changes to the QAP. However, it may provide the basis for initiating a change to the QAP.

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.

Changes to the Emergency Plan. However, it may provide the basis for initiating a change to the Emergency Plan.

Changes to the Security Plan. However, it may provide the basis for initiating a change to the Security Plan.

Changes to the Fire Protection Plan. However, it may provide the basis for initiating a change to the Fire Protection Plan.

Changes to the Radiation Protection Program. However, it may provide the basis for initiating a change to the Radiation Protection Program.

ndaries and Interfaces for ITDC SSCs s identified as ITDC that require availability shall be maintained in a state such that the essary functional capability is maintained.

ineered Requirements for ITDC SSCs igher level of engineered quality is maintained for ITDC SSCs to assure that the capability ts to reliably meet performance expectations and requirements. However, ITDC SSCs are not ty related and are not required to satisfy 10 CFR 50 Appendix B requirements. Although not uired by regulation, the following criteria is developed and applied, as specified by ineering, to ITDC SSCs to assure continued reliability:

drawings, procedures and instructions. These measures shall include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled. Design changes, including field changes will be subjected to engineered design control measures commensurate with the importance of the SSC.

Procurement Document Control: Measures will be invoked to assure that applicable regulatory requirements, design basis, and other requirements which are necessary to assure adequate quality are suitably included or referenced in the documents for procurement of material, equipment, services.

Instruction, Procedures, and Drawings: Activities affecting SSCs will be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and will be accomplished in accordance with these instructions, procedures, and drawings. Instructions procedures, and drawings will include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

Control of Purchased Material, Equipment, and Services: Measures will be invoked to assure that material, equipment, and services conform to the procurement documents.

These measures shall include provisions, as appropriate, for source evaluation and selection, objective evidence of quality furnished, inspection at the source, and examination upon delivery.

Inspection: Inspection of activities affecting quality will be invoked and executed to verify conformance with the documented instructions, procedures, and drawings for accomplishing the activity.

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.

Test Control: Surveillance testing will be established for SSCs to ensure that the SSCs perform satisfactorily commensurate with the importance of their intended function.

Measuring and Test Equipment: Appropriate controls will be invoked to assure that measuring and test devices used on SSCs are properly controlled, calibrated and adjusted at specified periods to maintain accuracy within necessary limits.

Corrective Action: Measures will be invoked to assure that conditions adverse to quality are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures will assure that the cause of the condition is determined and corrective action is taken to preclude repetition.

CFR 50 (General Design Criteria 2), as implemented by Standard Review Plan (SRP) Sections 1 and 3.3.2 and Regulatory Guides (RG) 1.76 and 1.117, requires that the plant be designed to hstand the effects of natural phenomena such as wind and tornadoes.

Millstone Unit Number 1 capability to withstand wind and tornado loadings was evaluated in Systematic Evaluation Program (SEP) (Reference 3.1-4) as Topic III-2. Several submittals e made to the U.S. Nuclear Regulatory Commission (NRC) to address issues raised under that c (References 3.1-6, 3.1-7, 3.1-8, and 3.1-9). In an evaluation dated November 25, 1985 ference 3.1-5), the NRC concluded that the proposal will provide adequate protection against ado events.

4 WATER LEVEL DESIGN original design basis water level at Millstone Unit Number 1 is the probable maximum flood F) level of 19.0 feet above mean sea level (MSL). In the defueled condition, flooding of Unit ructures is acceptable. The spent fuel is stored in the upper elevations of the Reactor Building, as such is adequately protected from the PMF. The intake structure itself which was originally gned as seismic Class 1, is designed to withstand a water level of elevation 32.4 feet MSL.

s level accounts for an assumed 13.4 feet MSL still water level and for non-breaking waves ve this level as they strike the structure.

5 MISSILE PROTECTION tems and components have been examined to identify and classify potential missiles.

5.1 Internally Generated Missiles o broad categories of systems and components are reviewed to determine the potential for erating missiles; pressurized components and high speed rotating machinery. Only designs re a single failure could lead to a missile ejection were considered.

as determined there are no highly pressurized components or high speed rotating machines able of generating significant missile hazards in the permanently defueled condition.

refore, no internally generated missiles are postulated.

5.2 Missiles Generated by Natural Phenomena effects of missiles generated by natural phenomena has been evaluated References 3.1-13, 14, 3.1-15, 3.1-16, and 3.1-17). On the basis of those evaluations, Millstone Unit Number 1 is quately protected against such missiles inhibiting the ability to maintain safe storage of new irradiated fuel.

objective of this assessment (Reference 3.1-19) is to assure that the integrity of the safety ted structures, systems, and components will not be impaired and that they will perform their ty functions in the event of a site proximity missile.

potential for hazardous activities in the vicinity of the Millstone site are addressed in pter 2. The licensee concludes that the generation of missiles at these facilities does not pose a ible threat to the Millstone site. Therefore, no specific protection is required other than that cribed for tornado-generated missiles.

refore, Millstone Unit Number 1 does not present an undue risk to the health and safety of the lic as a result of proximity missile hazards.

5.4 Aircraft Hazards re is presently one small commercial airport approximately 6 miles east-northeast of the site.

ton/New London Airport handles regularly scheduled commercial passenger flights but is equate for handling large jets. The licensee has determined that the probability of an aircraft king safety related structures of Millstone Unit Number 1 is sufficiently low that it does not stitute a significant hazard.

6 SEISMIC DESIGN Millstone Unit Number 1 plant was designed for an earthquake (equivalent to the operating s earthquake or OBE) with a horizontal peak ground acceleration (HPGA) of 0.07g and ewed for an earthquake (equivalent to the safe shutdown earthquake or SSE) with a PGA of

g. A smoothed design response spectrum recommended by John Blume and Associates and north 69 west component of the 1952 Taft earthquake record normalized to the specified GAs were used as seismic input for the analyses and design. The vertical component of ground ion was assumed to be two-thirds of the horizontal components. For the dynamic analyses of mic Class I structures, the buildings (or structures) were modeled as lumped mass-spring ems with fixed base to simulate the rock founded foundations.

dynamic responses of the Reactor Building and Radwaste Building/Control were analyzed by e-history approach.

o methods were used for the analysis of safety related equipment:

(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.

pter 4 of the NRC NUREG/CR-2024 report, Seismic Review of the Millstone 1 Nuclear er Plant prepared for the NRC as part of the Systematic Evaluation Program ference 3.1-21), summarizes the details of the original analysis and design.

ty related SSCs (Seismic II/I criterion), this portion of the structure is analyzed for the istic, median-centered in-structure accelerations developed by Vectra Technologies ference 3.1-32) for use in the USI A-46 (SQUG) program evaluations of equipment in the egory I portion of the Reactor Building. These floor accelerations and spectra are considered e realistic since they incorporate the variabilities of the input motion at a rock site and the ctural parameters (mass and stiffness). The SSE floor accelerations at elevation 82 feet 9 es (highest elevation evaluated in the Vectra report) are approximately 80% of the esponding floor accelerations obtained from the EDS Report (Reference 3.1-23). Therefore floor accelerations at the operating floor and at the roof level are conservatively taken as 80%

he corresponding accelerations from Reference 3.1-23.

6.1 Comparison of Measured and Predicted Responses nt procedures have been developed for abnormal operational events such as earthquakes. If und motion is detected, plant walkdowns are initiated to determine plant capability.

7 DESIGN OF CLASS I AND CLASS II STRUCTURES 7.1 Design Criteria, Applicable Codes, Standards and Specifications design of all structures and facilities (Class I and II) conformed to the applicable general es or specifications in effect at time of design.

7.2 Loads and Loading Combinations eral requirements for the design of all structures and equipment include provisions for sting the stresses resulting from dead loads, live loads and wind or seismic loads with impact s considered as part of the live load. The treatment of equipment stresses are generally limited hose produced by non-operating loads such as the effect of building motion due to earthquake he anchorage or support for a piece of equipment. However, the loads resulting from operating sures or temperatures on equipment are considered where they would increase the stresses.

rmal gradients in the foundation were not considered in the design.

ection of materials to resist the expected loads is based on standard practice in the power plant

d. The use of these materials is governed by local building codes and the experience and wledge of the designers and builders.

loadings of concern are the following:

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.

Design earthquake load.

Wind load.

criteria which have been followed for all Class I structures with respect to stress levels and combinations for the postulated events are noted below:

ss I portions of Reactor Building and Radwaste Building D + E Normal allowable code stresses are used (AISC for structural steel, ACI for reinforced concrete). The customary increase in design stresses, when earthquake loads are considered, is not permitted.

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.

D + W Normal allowable code stresses (AISC for structural steel, ACI for reinforced concrete) with the customary increases in stresses when wind loads are considered.

maximum allowable stresses used for various loading conditions are given for Class I ctures in Table 3.1-2.

or live loads were established based upon equipment and operating loads and applied to the c building code, which is recommended to the boroughs by the State of Connecticut. Roof live s are a minimum of 60 psf for Class I buildings and 40 psf for Class II buildings.

Class I structures will withstand the maximum potential loadings resulting from a wind city of 115 miles per hour with gusts up to 140 miles per hour. Although some damage to e structures could occur, this damage would under no circumstances impair the functions for capability of safe storage of irradiated fuel.

idental torsion on the structures was not considered in the analyses. The Reactor Building is a

-like structure with heavy columns and thick walls which give it a high torsional rigidity.

idental eccentricity would therefore produce negligible stresses and has been ignored.

hough a lack of symmetry applies to the arrangement of Class I structures, it is felt that ause the buildings are not generally structurally connected, torsional effects are likely to be of e consequence.

he analysis of concrete structures, the design modulus of 3x106 psi is in accordance with the I building code requirements for reinforced concrete (ACI 318-63), Section 1102, which is dard design practice. However, it is recognized that the modulus of elasticity of concrete

(1) Curing temperature (2) Initial temperature (3) Variations in mixes (4) Amount of hydration elastic modulus is not directly proportional to the strength of concrete: nevertheless, the ct of increasing the strength causes an increase in the modulus. However, the increase in the dulus due to age is not believed to be significant in the light of all the uncertainties affecting modulus of concrete.

atever the small change in the modulus may be, this effect is partially accounted for by cracks he concrete structure due to shrinkage and temperature. Such cracks tend to make the structure e flexible, which tends to compensate for the increased modulus. Also, the percent change in modulus is small compared to other inputs in the analysis such as dimensions, areas, cross ions, mass grouping, etc. Hence, the effect of a small modulus change on the validity of the amic analysis is considered to be negligible.

7.3 Structural Criteria for Class II Structures ss II structures and equipment are designed following the normal practice for the design of er plants in the State of Connecticut, but as a minimum, this was not less than given in the iform Building Code for Zone 2. The usual practice of determining the stress due to hquakes by applying a static load based on a specified seismic coefficient was followed. The gn of the Class II portion of the Reactor Building is addressed in Section 3.1.6.

owable stresses for building materials in Class II structures are as specified in the Basic lding Code, which is recommended to the boroughs by the State of Connecticut. A one-third ease is allowed for combinations including seismic or wind loads.

7.4 Seismic Class I and II Structures 7.4.1 Reactor Building ction functions of the Reactor Building are to enclose the spent fuel pool and associated equipment protect it from the weather. It supports maintenance of the fuel configuration in the fuel pool, vides protection from external hazards and supports maintenance of water in the fuel pool to a th necessary to ensure the irradiated fuel is always immersed.

sed to seismic Class II with the requirement that the enclosure is capable of sustaining an SSE hout collapse (Seismic II/I criterion).

cription thheld under 10 CFR 2.390 (d) (1) new fuel storage vault and the spent fuel storage pool are located in the Reactor Building. The tor service and refueling area is serviced by an overhead bridge crane. A refueling service form with necessary handling and grappling fixtures services the spent fuel storage pool.

l storage pool is a reinforced concrete structure, completely lined with seam-welded stainless l plate which is welded to reinforcing members embedded in concrete.

thheld under 10 CFR 2.390 (d) (1) liner was designed considering thermal stress, and the welds were dye penetrant inspected to ure leak tight integrity. Construction materials used in the construction of the spent fuel age facility includes 4000 psi, 28 day strength concrete, 40 ksi deformed bar reinforcing steel, ASTM, A-167, Type 304 stainless steel.

ctor Building Seismic Design and Analysis ed on the recommended earthquake design criteria established for the station, envelopes of imum acceleration, displacement, shear and overturning moment versus height have been eloped and are presented for the two assumed earthquake directions. See Figures 3.1-1 ugh 3.1-5. Based on the data developed by John A. Blume and Associates, engineers, the gn criteria have been established as follows for computerized analysis: the mathematical del was subjected to an excursion through the north 69 west component of the 1952 Taft hquake with an applied factor of 7/17. The resulting maximum shears, moments and lacements were used for design.

maximum envelopes of building design shears, moments and displacements are presented phically in Figures 3.1-3 through 3.1-5, respectively. These curves have been used in the mic design of the Reactor Building. Loads and shears from reactor pressure vessel and

pressive loads to the mat. Shears are transferred to the mat by friction and bearing.

Reactor Building was designed to resist the seismic shears and moments presented herein hout the usual increase in stress for short-term loadings. In addition, the structure was ewed to assure that it can resist 2.4 times the postulated seismic shears and moments without sing injury to the structure. In addition to the horizontal accelerations, a vertical building (and ipment) acceleration was used for design.

Reactor Building enclosure structure (above elevation 108 feet 6 inches) is analyzed for a istic median-centered SSE, as described in Section 3.1.6, and is shown to resist the resulting tial loads from the accelerations with no loss of structural integrity.

7.4.2 Control Room and Radwaste Treatment Building cription waste treatment facility is north of and adjacent to the reactor building. The building includes ipment and tankage space below grade with the plant control room above grade. The area w grade is of reinforced concrete construction with shielded compartments provided for the ous pieces of radwaste equipment. The control room above grade is of reinforced concrete ls with a two foot thick reinforced concrete roof. The control room and radwaste facility are sidered seismic Class II. The analytical model used in the seismic analysis of the control room radwaste building is shown in Figure 3.1-6 and is similar to those for the Reactor Building.

Radwaste Building is seismically analyzed consistent with Regulatory Guide 1.143.

eral Structural Features building substructure is founded on rock. The maximum bearing pressure on the rock is 10 per square foot. The exterior walls are of cast-in-place concrete and designed for an earth sure per square foot at any depth equal to the depth in feet times 90 pounds. The exterior walls the base slab were originally designed to resist hydrostatic pressure and uplift due to exterior ding to elevation 19 feet 0 inches. In the defueled condition, the below grade elevations are wed to flood such that the uplift force is reduced.

interior walls of the substructure are of cast-in-place concrete and those for the superstructure either cast in place or made of concrete masonry units. With minor exceptions, all floors are red-in-place concrete slabs.

east half grade floor at elevation 14 feet 6 inches, including the concrete shielding plugs ch close hatchways over equipment in the substructure, is designed for a uniform live load of psf.

tanks are made of ductile metal and all sump pits are lined so that these containers can be jected to substantial distortion without rupture.

ve basement level.

7.4.3 Intake Structure intake structure is a reinforced concrete frame supported on a reinforced concrete structure which is founded on rock. The building has a flat roof consisting of 10 gauge steel h concrete slab covered with insulation and a tar and felt roofing membrane. Hatches are vided in the roof for removal of major pieces of equipment. The front wall of the intake cture is designed to resist the standing wave. Seismic stress levels were calculated using fficients of 0.07 g at grade and 0.12 g at the roof level for design earthquake and 2.4 times e values for the maximum earthquake. The structure is capable of withstanding 300 mph wind not the tornado internal pressure of 2.5 psi. However, the large number of hatches in the roof release this pressure. Although originally design as seismic Class I, the intake is considered mic Class II in the permanently defueled condition.

intake structure is located west of the main plant and has five 11 foot 2 inch wide bays. Each is provided with manually raked trash racks and stop log guides.

vision for service and cooling water strainers is made in a separate covered pit adjacent to the ke.

7.4.4 Turbine Building Turbine Building is a Class II structure. The Turbine Building foundation consists of a forced concrete mat supported on rolled structural steel H section bearing piles. All piles were en to rock or to refusal in the dense strata immediately above rock. Reinforced concrete shield ls are provided up to the operating deck at elevation 54 feet 6 inches.

remaining portions of the building have steel framing and metal siding. The Turbine Building und floor consists of a reinforced concrete slab supported on sand fill over the foundation

s. The turbine generator pedestal is a massive reinforced concrete pedestal designed to support turbine generator. It is supported on a six foot thick mat which forms an integral part of the aining building mat foundations. The roof is covered with metal decking, insulation and ing material flashed at the parapet walls. An overhead rolling door at the west end of the ding provides rail car access into the building.

8 SEISMIC QUALIFICATION OF SEISMIC CATEGORY I INSTRUMENTATION AND ELECTRICAL EQUIPMENT e of the plant process instrumentation provides safety related functions in conjunction with storage and handling of irradiated fuel or radioactive waste, or is credited with any function in safety evaluations performed to ensure that no undue risk to the health and safety of the public ts. No plant instrumentation or electrical systems are required for mitigation of the design s fuel handling accident. Seismic qualification of plant instrumentation and electrical ipment is not required.

s section is related to qualification of the electrical portion of the engineered safety features to orm their intended functions in the combined normal, accident and post accident ironments. There are no non-structural engineered safety features related to the safe storage handling of the irradiated fuel or radioactive waste, or credited in the safety evaluations ormed to ensure that no undue risk to the health and safety of the public exists. No non-ctural engineered safety features are credited in accident analysis to prevent or mitigate the sequences of the current design basis fuel handling accident.

10 REFERENCES 1 Millstone Nuclear Power Station Unit Number 1 Application for Full Term Operating License, September 1, 1972.

2 NUREG-0824, Integrated Plant Safety Assessment, Systematic Evaluation Program, Millstone Nuclear Power Station, Unit Number 1, February 1983.

3 Samual J. Chilk (Nuclear Regulatory Commission) memo to J. M. Taylor (Nuclear Regulator Commission), SECY-92-233 Resolution of Deviations Identified during the Systematic Evaluation Program dated September 18, 1992.

4 Integrated Plant Safety Assessment, Systematic Evaluation Program, Millstone Nuclear Power Station, Unit Number 1, NUREG-0834, Supplement Number 1, November 1985.

5 Letter, C.I. Grimes (NRC) to J.F. Opeka, subject: IPSAR Sections 4.4 Wind and Tornado Loadings and 4.7 Tornado Missiles.

6 Letter, 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.

7 Letter, 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 Settlement 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.

8 Letter, October 7, 1983, from W. G. Counsil to D. M. Crutchfield (NRC),

Subject:

Millstone Nuclear Power Station Unit Number 1, SEP Topic III-2 Wind and Tornado Loadings.

9 Letter, December 3, 1982, from W. G. Counsil to D. M. Crutchfield (NRC),

Subject:

Millstone Nuclear Power Station Unit Number 1, SEP Topic III-2 Wind and Tornado Loadings.

11 Letter, W. G. Counsil to D. M. Crutchfield (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 Topic III-6, Seismic Design Considerations.

12 10 CFR 50, Appendix A, General Design Criterion 4.

13 Letter 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.

14 Letter of March 9, 1982, W.G. Counsil to D. M. Crutchfield,: Millstone Nuclear Power Station Unit Number 1, SEP Topic III-4.A, Tornado Missiles.

15 Letter 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.

16 Letter of August 31, 1981, W.G. Counsil to D. M. Crutchfield,: Millstone Nuclear Power Station Unit Number 1, SEP Topic III-4.A, Tornado Missiles.

17 Letter of October 16, 1985, J. F. Opeka to C. I. Grimes, Millstone Nuclear Power Station Unit Number 1, Integrated Safety Assessment Program.

18 Letter 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.

19 Letter 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.

20 Letter of September 17, 1981, W. G. Counsil to D. M. Crutchfield, Millstone Nuclear Power Station Unit Number 1, SEP Topic III-4.D, Site Proximity Missiles.

21 NRC NUREG/CR-2024 Report, Seismic Review of the Millstone-1 Nuclear Power Plant, July 1981.

22 SEP Safety Topics III-6, Seismic Design Considerations and III-II, Component Integrity

- Millstone Nuclear Power Station Unit Number 1, SAR dated 6/30/82.

23 EDS Report Number 02-0240-1094, Generation of In-Structure Seismic Response Spectra Millstone Unit Number 1, dated June 1982.

24 NRC letter, Site Specific Ground Response Spectra for SEP Plants Located in the Eastern United States, June 17, 1981.

26 Letter J. F. Opeka to C.I. Grimes, Millstone Nuclear Station, Unit Number 1 ISAP Topic 1-19, Integrated Structural Analysis, dated January 6,1986.

27 Letter from D. G. Eisenhut, NRC, to W. G. Counsil, dated January 1, 1980.

28 Letter from D. M. Crutchfield, NRC, to W. G. Counsil, dated July 28, 1980.

29 Letter from W. G. Counsil to D. M. Crutchfield, NRC, dated October 16, 1985.

30 Millstone Unit 2 Final Safety Analysis Report Section 5.8.6.

31 Millstone Unit 3 Final Safety Analysis Report Section 3.7.4.2.

32 Vectra Technologies Report Number 0024-00099-RB-1, Rev. 1, Reactor Building A-46 Spectra, dated June 10, 1996.

SEP AND TMI SAFETY ISSUES WHICH LISTED THE SPECIFIED GDC AS PART GENERAL DESIGN CRITERIA OF CONCERN AFFECTED REGULATORY GUID I OVERALL REQUIREMENTS 1 QUALITY STANDARDS AND SEP II-3.A, II-3.B, II-3.C, III-3.A AND III-7.B 1.27, 1.59 RECORDS 2 DESIGN BASES FOR SEP II-2.A, II-3.A, II-3.B, II-3.C, II-4.E, II-4.F, 1,27, 1.,32, 1.59, 1.60, 1.61, 1.68, 1.75 PROTECTION AGAINST II-4.3, III-19 III-2, III-3.A, 1.76, 1.92, 1.102, 1.117, 1.120, 122, NATURAL PHENOMENA III-3.B, III-3.C, III-6, III-7.B, III-8.C, III-11, 1.127, 1.129, 1.132 VIII-3.A, VIII-3.B, TMI II.B.1 3 FIRE PROTECTION (SEE DSAR Section 3.2.9) 4 ENVIRONMENTAL AND SEP II-1.C, II-3.A, II-3.B, II-3.C, III-1, III-4.B, 1.3, 1.4, 1.7, 1.20, 1.27, 1.29, 1.32, 1.3 MISSILE DESIGN BASES III-5.A, III-5.B, III-7,B, III-11, 1.45, 1.46, 1,59, 1.68, 1.75, 1.115, 1.12 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 5 SHARING OF STRUCTURES, SEP III-1, VIII-3.A AND VIII-3.B 1.32, 1.75, 1.129 SYSTEMS AND COMPONENTS II PROTECTION BY MULTIPLE FISSION PRODUCT BARRIERS 10 REACTOR DESIGN NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION

SEP AND TMI SAFETY ISSUES WHICH LISTED THE SPECIFIED GDC AS PART GENERAL DESIGN CRITERIA OF CONCERN AFFECTED REGULATORY GUID 11 REACTOR INHERENT NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE PROTECTION DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 12 SUPPRESSION OF REACTOR NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE POWER OSCILLATIONS DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 13 INSTRUMENTATION AND NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE CONTROL DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 14 REACTOR COOLANT NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE PRESSURE BOUNDRY DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 15 REACTOR COOLANT SYSTEM NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE DESIGN DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 16 CONTAINMENT DESIGN NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 17 ELECTRIC POWER SYSTEMS SEP III-1, VII-7, VIII-2, VIII-3.A VIII-3.B, TMI 1.6, 1.9, 1.32, 1.75, 1.129 II.E.3.1, II.G.1 18 INSPECTION AND TESTING NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE OF ELECTRIC POWER DEFUELED CONDITION PERMANENTLY DEFUELED SYSTEMS CONDITION

SEP AND TMI SAFETY ISSUES WHICH LISTED THE SPECIFIED GDC AS PART GENERAL DESIGN CRITERIA OF CONCERN AFFECTED REGULATORY GUID 19 CONTROL ROOM NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION III PROTECTION AND REACTIVITY CONTROL SYSTEMS 20 PROTECTION SYSTEM NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE FUNCTIONS DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 21 PROTECTION SYSTEM NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE RELIABILITY AND DEFUELED CONDITION PERMANENTLY DEFUELED TESTABILITY CONDITION 22 PROTECTION SYSTEM NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE INDEPENDENCE DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 23 PROTECTION SYSTEM NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE FAILURE MODES DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 24 SEPARATION OF NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE PROTECTION AND CONTROL DEFUELED CONDITION PERMANENTLY DEFUELED SYSTEMS CONDITION 25 PROTECTION SYSTEM NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE REQUIREMENTS FOR DEFUELED CONDITION PERMANENTLY DEFUELED REACTIVITY CONTROL CONDITION MALFUNCTIONS

SEP AND TMI SAFETY ISSUES WHICH LISTED THE SPECIFIED GDC AS PART GENERAL DESIGN CRITERIA OF CONCERN AFFECTED REGULATORY GUID 26 REACTIVITY CONTROL NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE SYSTEM REDUNDANCY DEFUELED CONDITION PERMANENTLY DEFUELED AND CAPABILITY CONDITION 27 COMBINED REACTIVITY NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE CONTROL SYSTEMS DEFUELED CONDITION PERMANENTLY DEFUELED CAPABILITY CONDITION 28 REACTIVITY LIMITS NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 29 PROTECTION AGAINST NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE ANTICIPATED OPERATIONAL DEFUELED CONDITION PERMANENTLY DEFUELED OCCURRENCES CONDITION IV FLUID SYSTEMS 30 QUALITY OF REACTOR NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE COOLANT PRESSURE DEFUELED CONDITION PERMANENTLY DEFUELED BOUNDARY CONDITION 31 FRACTURE PREVENTION OF NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE REACTOR COOLANT DEFUELED CONDITION PERMANENTLY DEFUELED PRESSURE BOUNDARY CONDITION 32 INSPECTION OF REACTOR NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE COOLANT PRESSURE DEFUELED CONDITION PERMANENTLY DEFUELED BOUNDARY CONDITION

SEP AND TMI SAFETY ISSUES WHICH LISTED THE SPECIFIED GDC AS PART GENERAL DESIGN CRITERIA OF CONCERN AFFECTED REGULATORY GUID 33 REACTOR COOLANT NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE MAKEUP DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 34 RESIDUAL HEAT REMOVAL NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 35 EMERGENCY CORE COOLING NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 36 INSPECTION OF EMERGENCY NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE CORE COOLING SYSTEM DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 37 TESTING OF EMERGENCY NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE CORE COOLING SYSTEM DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 38 CONTAINMENT HEAT NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE REMOVAL DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 39 INSPECTION OF NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE CONTAINMENT HEAT DEFUELED CONDITION PERMANENTLY DEFUELED REMOVAL SYSTEM CONDITION

SEP AND TMI SAFETY ISSUES WHICH LISTED THE SPECIFIED GDC AS PART GENERAL DESIGN CRITERIA OF CONCERN AFFECTED REGULATORY GUID 40 TESTING OF CONTAINMENT NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE HEAT REMOVAL SYSTEM DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 41 CONTAINMENT NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE ATMOSPHERE CLEANUP DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 42 INSPECTION OF NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE CONTAINMENT DEFUELED CONDITION PERMANENTLY DEFUELED ATMOSPHERE CLEANUP CONDITION SYSTEMS 43 TESTING OF CONTAINMENT NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE ATMOSPHERE CLEANUP DEFUELED CONDITION PERMANENTLY DEFUELED SYSTEMS CONDITION 44 COOLING WATER NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 45 INSPECTION OF COOLING NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE WATER SYSTEM DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 46 TESTING OF COOLING NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE WATER SYSTEM DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION V REACTOR CONTAINMENT

SEP AND TMI SAFETY ISSUES WHICH LISTED THE SPECIFIED GDC AS PART GENERAL DESIGN CRITERIA OF CONCERN AFFECTED REGULATORY GUID 50 CONTAINMENT DESIGN NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE BASIS DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 51 FRACTURE PREVENTION OF NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE CONTAINMENT PRESSURE DEFUELED CONDITION PERMANENTLY DEFUELED BOUNDARY CONDITION 52 CAPABILITY FOR NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE CONTAINMENT LEAKAGE DEFUELED CONDITION PERMANENTLY DEFUELED RATE TESTING CONDITION 53 PROVISIONS FOR NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE CONTAINMENT INSPECTION DEFUELED CONDITION PERMANENTLY DEFUELED AND TESTING CONDITION 54 SYSTEMS PENETRATING NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE CONTAINMENT DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION 55 REACTOR COOLANT NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE PRESSURE BOUNDARY DEFUELED CONDITION PERMANENTLY DEFUELED PENETRATING CONDITION CONTAINMENT 56 PRIMARY CONTAINMENT NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE ISOLATION DEFUELEDCONDITION PERMANENTLY DEFUELED CONDITION

SEP AND TMI SAFETY ISSUES WHICH LISTED THE SPECIFIED GDC AS PART GENERAL DESIGN CRITERIA OF CONCERN AFFECTED REGULATORY GUID 57 CLOSED SYSTEMS NOT APPLICABLE TO THE PERMANENTLY NOT APPLICABLE TO THE ISOLATION VALVES DEFUELED CONDITION PERMANENTLY DEFUELED CONDITION VI FUEL AND RADIOACTIVITY CONTROL 60 CONTROL OF RELEASES OF SEP II.2.C, XI-1, XI-2, TMI II.B.2, II.B.3, 1.3, 1.4 RADIOACTIVE MATERIALS 2.1.6.A, 2.1.8.A TO THE ENVIRONMENT 61 FUEL STORAGE AND SEP XI-1, XI-2 HANDLING AND RADIOACTIVITY CONTROL 62 PREVENTION OF CRITICALITY IN FUEL STORAGE AND HANDLING 63 MONITORING FUEL AND SEP XI-1, XI-2 WASTE STORAGE 64 MONITORING SEP II-2.C, XI-1, XI-2; TMI II.B.2, II.B.3, RADIOACTIVITY RELEASES 2.1.6.A, 2.1.8.A

Structural Concrete Concrete Steel Structural Reinforcing Max. Max. Concrete Tension Structural Steel Steel Max. Allowable Allowable Max. On The Steel Shear Compression Stru Allowable Compression Shear Allowable Net On Gross On Gross S loading Conditions Stress Stress Stress Bearing Section Section Section Be (1) DEAD LOADS PLUS 0.5 Fy 0.45 f c 1.1 f c 0.25 f c 0.60 Fy 0.40 Fy VARIES 0.66 LIVE LOADS,* PLUS WITH TO OPERATING LOAD SLENDERN 0.60 PLUS SEISMIC LOADS ESS RATIO (0.07g)

(2) DEAD LOADS PLUS 0.667 Fy 0.60 f c 1.467 f c 0.333 f c 0.80 Fy 0.53 Fy VARIES 0.88 LOADS,

• PLUS WITH TO OPERATING LOADS SLENDERN 0.80 PLUS WIND LOADS ESS RATIO (3) DEAD LOADS PLUS GROSS STRUCTURAL LIVE LOADS,* PLUS INTEGRITY CAN BE MAIN-OPERATING LOADS, TAINED PLUS SEISMIC LOADS (SEE NOTE 1 BELOW) 0.17g 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 assure that gross structural integrity can be maintained during ground motion having 17/7 the intensity of the operating basis earthquake described in Section 3.1.6, even though stresses in some of the materials may exceed the yie point.

FIGURE 3.1-1 REACTOR BUILDING SEISMIC LOADS GURE 3.1-2 ACCELERATION DIAGRAM UNDER SEISMIC LOADS 5 PERCENT DAMPING

FIGURE 3.1-3 SHEAR DIAGRAM UNDER SEISMIC LOADS FIGURE 3.1-4 MOMENT DIAGRAM UNDER SEISMIC LOADS FIGURE 3.1-5 DISPLACEMENT DIAGRAM UNDER SEISMIC LOADS FIGURE 3.1-6 RADWASTE BUILDING - MATHEMATICAL MODEL 1 FUEL STORAGE AND HANDLING 1.1 New Fuel Storage ce Millstone Unit 1 is a de-commissioned unit, new fuel will no longer be received.

1.2 Spent Fuel Storage 1.2.1 Design Bases design bases for the storage of spent fuel are as follows:

A fuel storage pool for the underwater storage of 2959 fuel assemblies.

Maintain a keff of less than 0.95 at all times, including 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.

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.

Structural design of the fuel storage and equipment storage facilities meets all requirements for Class I structures.

The fuel storage racks for the fuel are designed to assure subcriticality in the fuel pool.

The storage racks are an interconnected honeycomb array of square stainless steel boxes forming individual cells for fuel storage. 1045 storage cells contain Boraflex sheets (not credited) on four sides, and 2184 storage cells contain B4C plates for neutron absorption.

Of the 1045 storage cells with Boraflex, only 775 cells are allowed to contain fuel.

Criticality Accident Requirements. Millstone 1 has chosen to comply with 10 CFR 50.68(b).

1.2.2 Facilities Description fuel pool contains water which is not borated. The fuel storage pool is a reinforced concrete cture, completely lined with seam-welded, stainless steel plate (11 gauge) which is welded to forcing members (channels, I-beams, etc.) embedded in concrete. The liner is reinforced by eased thickness and suitable insert strips in areas subject to heavy loading such as the cask dling area The concrete shielding walls are two or more feet thick and are designed as part of Class I portion of the Reactor Building structure.

rconnecting drainage paths are provided behind the liner welds to:

To control the loss of pool water and To provide liner leak detection and measurement capability.

drainage paths are suitably grouped to indicate the area of leakage. To avoid unintentional ning of the pool, there are no penetrations that would permit the pool to be drained below roximately nine feet above the top of the active fuel, and all lines extending below this level equipped with suitable valving to prevent backflow. The passage between the fuel storage l and the refueling cavity above the reactor vessel is provided with two gates. The refueling ity is maintained in a drained down state. The gate adjacent to the refueling cavity is welded to passage liner forming a permanent pressure boundary for the fuel storage pool. The double ed gate adjacent to the fuel storage pool is removable but normally maintained in the closed ition. A normally open drain line between the gates permits detection of leaks from the gate cent to the fuel storage pool. The drain line may be isolated and the volume between the gates ded to support removal of the gate for repairs in the event of such leakage.

esponse to the NRC I.E. Bulletin 84-03, augmented leak detection capability has been alled in the spent fuel pool to indicate high/low level in the pool.

water in the pool is cooled and filtered as required by the spent fuel pool cooling and in-pool nup system described in Section 3.2.1.3.

storage pool is designed to hold 20 fuel channels.

area of approximately seven feet by seven and one half feet is reserved for loading a spent fuel ping cask.

isters containing irradiated reactor vessel internals and other materials classified per 10 CFR s greater than class C (GTCC) waste are stored in the fuel storage pool adjacent to the fuel ping cask area.

1.2.3 Safety Evaluation spacing of fuel bundles in the spent fuel storage pool, the presence of neutron absorbing ons (where credited) in the fuel storage racks, not placing fuel in prohibited locations tified in the Technical Specifications, and the design of the fuel bundles maintains keff less or equal to 0.95. This is assured by limiting the fuel assemblies in the pool to those that have aximum K of 1.24 in the normal reactor configuration at cold conditions, and an average 35 enrichment of 3.8 weight percent or less. The criticality analysis confirms acceptable lts regardless of the spent fuel pool temperature.

diated fuel being moved in the fuel storage pool is covered by an eight foot minimum of water ve the top of active fuel, which is sufficient for radiation shielding. Radiation monitors in the storage pool work area monitor the radiation level and alarm upon excessive levels.

refueling platform equipment lock upon loss of power.

fuel storage racks are analyzed to withstand the impact of a dropped fuel assembly and dling tool with a combined dry weight of 1675 pounds from the maximum lift height of the eling platform telescoping mast. The analyses performed (References 3.2-9 to 3.2-12) onstrate that the spent fuel racks remain functional and that the spent fuel remains in a critical, submerged and coolable condition.

quid level transmitter, monitoring pool water level, is provided to detect loss of water from the

l. A level transmitter, monitoring the skimmer surge tank, is provided to permit water loss ction by initiating a low level alarm and provide level indication in the Millstone Unit 2 trol Room.

afety 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 8.

1.3 Spent Fuel Pool Cooling System Spent Fuel Pool Cooling System has been analyzed to remove the maximum heat load from spent fuel pool.

1.3.1 Design Bases Fuel Pool structure, pool liner, fuel racks, and external cooling system have been designed for mperature of approximately 150F. However, all of these structures and components have n demonstrated to be structurally adequate for abnormal temperature excursions to 212F.

h a complete loss of external cooling and a closed airspace above the pool, it would take roximately 10 days for the pool temperature to rise to 212F from an initial SFP bulk water perature of 100F, or approximately 7.5 days to rise to 212F if starting from the TRM upper perature limit of 140F. The spent fuel pool cooling system and secondary DHR cooling em have been qualified for satisfactory operation with pool temperatures as high as 170F.

s is greater than the maximum anticipated pool water temperature, following loss of cooling, vided that natural ventilation within the reactor building is established within approximately 5 s if starting from an initial SFP bulk water temperature of 100F, or 2.5 days if starting from TRM upper temperature limit of 140F.

ltiple methods are available to add water to the pool and adequate time is available to repair, ually reinstate or line up the system used for pool water cooling. Most significantly, if this em is not used to cool the pool water, no fuel damage would result and the potential off site osure would not approach the guidelines established in 10 CFR 50.34(a) or 10 CFR 100.11 vided makeup is initiated at a rate equal to or greater than the maximum evaporation rate at time prior to fuel uncovery. Water above the fuel provides shielding and heat sink functions.

the permanently defueled condition, the design bases for the fuel pool cooling system is:

To provide high clarity water to the fuel pool using the in-pool cleanup system.

To remove radioactivity released to the pool water using the in-pool cleanup system.

1.3.2 Spent Fuel Pool Heat Load lstone Unit Number 1 has permanently ceased power operation and all irradiated fuel has been manently removed from the reactor vessel. There are 2885 irradiated fuel assemblies in the nt fuel pool including one segmented bundle, consisting of 19 fuel rods. A decay heat load ulation was performed utilizing the computer program ORIGEN2, an industry standard for h analysis (Reference 3.2-13). The results show that total heat load in the pool was 1.781 tu/hr on 1/1/99. The spent fuel pool secondary cooling system (DHR) has been sized to ove the spent fuel decay heat load of approximately 1.5 Mbtu/hr, projected to exist on 6/1/00.

1.3.3 Loss of Fuel Pool Cooling h the spent fuel pool heat load established, a second calculation (Reference 3.2-14) was ormed to determine the transient and steady state spent fuel pool and reactor building peratures without active cooling to the spent fuel pool. Several cases were analyzed with erent ventilation configurations such as forced ventilation, natural ventilation and no tilation through the building. Steady state and transient calculations were performed to blish maximum pool and building temperatures and evaporation rates, as well as time frames potential operator actions. All analyses were performed using the GOTHIC computer gram.

limiting case evaluated was during summer conditions (92F, 50% Relative Humidity) owing the loss of active spent fuel pool cooling and without the reactor building HVAC system peration. In this case the time to reach 212F in the spent fuel pool is approximately 7.5 days tarting from the TRM upper temperature limit of 140F. This calculation also establishes a imum evaporative loss of 3.8 gpm under the above conditions. If natural ventilation is blished, by opening the reactor building truck bay doors, equipment hatch garage doors and tornado dampers on the reactor building roof, the maximum calculated pool temperature is F and the maximum evaporation rate is 3.0 gpm.

1.3.4 System Description spent fuel pool cooling system cools water in the fuel pool on an as needed basis to maintain er temperature. An in-pool demineralizer and filter maintain purity and water quality. Water is ulated by either one or two pumps which take suction from the skimmer surge tanks. The stable spent fuel pool weir gates maintain pool level and skim water from the surface of the pool. System lineups may vary due to decreasing heat removal needs. The flow diagram for spent fuel pool cooling system is shown in Figures 3.2-1 through 3.2-3.

mal water level of the fuel storage pool, and a local temperature indicator. The transmitter put is monitored in the Millstone Unit 2 Control Room via the Programmable Logic Controller C) which provides both indication of bulk temperature and notification of a high and low er temperature conditions within the fuel storage pool.

in-pool fuel pool demineralizer operates on an as needed basis to maintain pool water mistry. The in-pool filter operates on an as needed basis to maintain pool water clarity. The mmer surge tanks are shielded with concrete.

fuel pool cooling system is controlled and operated locally and from the Millstone Unit 2 trol Room. The system is provided with indicators and alarms for system flow, water level, temperature, skimmer surge tank level, and component operating status.

1.3.5 Safety Evaluation fuel pool water acts passively to transfer decay heat from the fuel and will protect the fuel m damage without human intervention as long as the fuel is completely immersed in water. If rnal cooling is stopped, the pool water temperature would gradually increase, resulting in no damage. In the most severe case of a closed airspace, with the current decay heat load in the lstone Unit Number 1 Fuel Pool and no external cooling, the pool temperature would only h equilibrium (stop rising) when the pool water boils, which is the natural limit of water perature in a space at atmospheric pressure. The fuel pool structure, pool liner, fuel racks, and rnal cooling system have been demonstrated to be adequate for abnormal temperature ursions to 212F. With a complete loss of external cooling and a closed airspace above the l, it would take approximately 10 days for the pool temperature to rise to 212F from an initial bulk water temperature of 100F, or approximately 7.5 days to rise to 212F if starting from TRM upper temperature limit of 140F. This is significantly longer than required to reinstate rnal cooling of the water. If natural ventilation is established, by opening the reactor building k bay doors, equipment hatch garage doors and the tornado dampers on the reactor building

, the maximum calculated pool temperature is 163F.

1.4 Fuel Handling System 1.4.1 Design Bases design bases for the fuel handling system are as follows:

No release of contamination or exposure of personnel to radiation will exceed the 10 CFR 20 limits.

Limited work on irradiated components will be possible at any time.

1.4.2 System Description fuel handling system handles irradiated fuel.

lding crane, which is equipped with a 110 ton main hoist and a seven-ton auxiliary hoist.

se hoists can reach any major equipment storage area on the operating floor.

1.4.3 Safety Evaluation refueling bridge and other fuel handling equipment are required for movement of fuel and er items stored in the fuel pool into storage/shipping containers. The reactor building crane is uired to move storage and shipping casks in the reactor building. These functions are required he permanently defueled condition, but are not safety related.

2 MONITORING AND CONTROL FUNCTIONS Millstone Unit 2 Control Room serves as the control room for Millstone Unit 1, and is tinuously manned. It is described in Section 7.6 of the Millstone Unit 2 Final Safety Analysis ort. Millstone Unit 2 Operations personnel are responsible for the monitoring and control of Unit 1 spent fuel pool island (SFPI) and auxiliary systems via a computer console located in Millstone Unit 2 Control Room. The computer console in the Millstone Unit 2 Control Room rfaces with a Programmable Logic Controller (PLC) for data acquisition and trending. The is located in the Millstone Unit 1 Central Monitoring Station (CMS). The CMS is located hin the Maintenance Shop.

Millstone Unit 1 CMS is not manned. It contains two computer consoles that may only be d as monitors, because they are normally in a locked supervisory mode.

re are no monitoring stations in the original Unit 1 Control Room. The original Unit 1 Control m no longer performs any Unit 1 function.

3 DECAY HEAT REMOVAL (DHR) SYSTEM 3.1 Design Bases DHR system is designed to provide cooling to the spent fuel pool cooling system. The system gn bases are:

ign Temperature: 170F ign Flow Rate (maximum) 625 gpm per pump ign Pressure: 200 psig DHR system is normally in service to supply spent fuel pool cooling system cooling loads as ded. System lineups vary during the permanently defueled condition due to reduced heat oval needs.

DHR system provides a supply of cooling water to the shell side of the spent fuel pool heat hangers. Water is circulated in a closed loop by the DHR pumps. Heat is removed from the em by the four DHR air-water heat exchangers located outside on the roof above the H&V

. System configuration may vary depending on heat load. The remainder of the system sists of a cooling water expansion tank, an air separator, piping and valves, and controls and rumentation. A demineralizer maintains system activity below established limits. The flow ram for the system is shown in Figure 3.2-4.

3.3 Safety Evaluation DHR system supplies cooling water to the fuel pool heat exchangers. Fuel pool cooling is a ction that is required for the permanently defueled condition, but is not safety related.

refore, this function of the DHR system is not safety related.

3.4 Testing and Inspection system components and instrumentation are tested periodically as necessary to ensure rational readiness.

3.5 Instrumentation R system instrumentation and controls are located locally and in the Millstone Unit 2 Control m.

4 MAKEUP WATER SYSTEM 4.1 Demineralized Water 4.1.1 System Description spent fuel pool makeup system will supply and store demineralized water to makeup for poration and leakage in the pool. The primary source will be from the Unit 2 Primary Makeup tem which is supplied from the onsite water treatment facility. A 5,000 gallon storage tank and sfer pump are installed in the reactor building to provide makeup water to the spent fuel pool ng period when the normal makeup from Unit 2 is unavailable. A connection to the pool eup line is also provided near the reactor building truck bay door to allow makeup to be vided by a tanker truck or fire water if necessary.

piping, tanks and other equipment of the spent fuel pool water storage and makeup system makeup system are of corrosion resistant metals which prevent contamination of the makeup er with foreign material.

flow diagram for the system is shown in Figure 3.2-5.

spent fuel pool makeup water system provides demineralized makeup water to the spent fuel l and spent fuel pool cooling system. This function supports fuel pool cooling, but is not safety ted.

4.1.3 Testing and Inspection ration of the makeup system is on demand at intermittent intervals to replenish water in the nt fuel pool makeup water storage tank and the skimmer surge tanks. The equipment is ally inspected periodically. Sampling of the makeup water storage tank is a standard nitoring procedure.

4.1.4 Instrumentation motor control switch for the makeup water transfer pump is located locally at the pump.

al makeup storage tank level indication is also provided.

5 INTENTIONALLY DELETED 6 PROCESS SAMPLING SYSTEM 6.1 Design Bases reason for sampling process gases is to provide representative samples for testing to obtain from which the performance of the plant equipment and systems are determined.

6.2 System Description Unit Number 1 BOP ventilation exhaust flow is continuously sampled for radioactive iculates. The sample is taken from the exhaust duct which runs along the north exterior wall of Reactor Building. A single point sample nozzle is positioned to obtain a representative sample he well mixed exhaust air. The sample passes through a particulate filter and is then expelled k into the exhaust duct.

SFPI ventilation exhaust flow is continuously monitored for gaseous radiation and iculates. The sample is taken from the exhaust duct near the reactor building exhaust plenum.

ngle point sample nozzle is positioned to obtain a representative sample of the turbulent and l mixed exhaust air. The sample passes through a particulate filter and a gas monitor and is expelled back into the exhaust duct.

b samples can be taken from the BOP and SFPI ventilation exhaust ducts and analyzed for oactive content.

6.3 Safety Evaluation BOP and SFPI ventilation systems are not safety related.

ctional tests were performed after installation. Routine use substitutes for subsequent periodic ing, with the exception of calibration and maintenance.

7 ELECTRICAL SYSTEMS 7.1 Introduction station electrical systems include the equipment and facilities which provide power to desired t equipment, instrumentation and controls. The system is designed to provide reliable power the permanently defueled condition. The power system is designed with a sufficient source, y protection, control, and necessary switching.

thheld under 10 CFR 2.390 (d) (1) 7.3 Intentionally Deleted.

7.4 On Site Electric System 7.4.1 Introduction ficient time is available to operators following a loss of offsite power to assure the continued storage of fuel without reliance on emergency sources of power.

power is provided through the Unit 1 service transformer 11YT-U1. The Unit 1 service sformer has adequate capacity to supply all normal auxiliaries required to support the manently defueled condition. Power for the SFPI and other decommissioning related activities om the 11YT-U1 via Bus 14H.

I and decommissioning related 125V DC power is obtained from rectified AC power at the nt of use, and a separate 125V DC source consisting of a 125V DC battery, a battery charger, onnect switch and distribution panel.

7.4.2 4160 Volt System Unit 1 service transformer (11YT-U1) steps down 23 kV to 4160 volts for the auxiliary buses.

he permanently shutdown condition, the plant will normally be operated with auxiliary trical loads supplied from the Unit 1 service transformer.

uit breaker is from the decommissioning 125 volt DC system.

major component of the 4160 volt power system is described below.

Unit 1 Service Transformer Unit 1 service transformer is an outdoor, 22,860-4160 volt three phase, 60 Hz., 125 kV BIL, 0 KVA transformer.

7.4.3 480 Volt System er from 4160 volt bus 14H is stepped down through transformers energizing the 480 volt es SFPI-B1 and FAC-B2.

and MCC supply breakers are opened and closed locally. All breakers will trip automatically n overload conditions exist.

7.4.4 120 Volt Systems SFPI electrical system utilizes its own dedicated 120V AC power derived from the SFPI AC er system.

SFPI instrument AC system is provided by the SFPI 120V AC distribution system and ked up by point of use UPS equipment. The SFPI PLC system has an integral 24V DC power ply.

7.4.5 AC Power System Design Criteria Interrupting Capacity - The switchgear, load centers, motor control centers, and distribution panels are sized for interrupting capacity based on maximum short circuit availability at their location. Low voltage metal enclosed breakers at load centers and molded case breakers at motor control centers are adequately sized for these maximum available short circuit currents.

Electrical System Protection - Electrical system protection is provided by protective devices or relays which monitor the electrical characteristics of the equipment and/or power system to assure operation consistent with design parameters, as follows:

(a) Initiate removal from service 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.

I related 125V DC utilizes rectified AC power. The rectifiers are located at the SFPI 480V AC tchgear bus. In addition, the decommissioning 125V DC system consists of a 125V DC ery, charger, disconnect switch and distribution panel.

7.4.7 Intentionally Deleted 7.4.8 Safety Evaluation he permanently defueled condition portions of the electrical systems are required for power

/or control of required non-safety related equipment in other systems. Since none of the ipment powered by these systems is safety related (Class 1E), all of the electrical systems are

-safety related. Although single failure criteria still apples to the unit, it need not be applied to ems and equipment that are non-safety related. Since none of the electrical systems or ipment is safety related or required for Regulatory Guide 1.97 (post accident monitoring) mitments, the EEQ program need not be applied. General Design Criteria Number 17 ctric Power Systems) includes certain requirements for availability of offsite power to support cal functions. Since the reactor cannot be made critical under allowed plant conditions in the manently defueled condition, no power source is required to be operable or available.

8 AIR CONDITIONING, HEATING, COOLING AND VENTILATION SYSTEMS 8.1 Reactor Building and SFPI Heating and Ventilation System 8.1.1 Design Bases Reactor Building and SFPI heating and ventilation systems are operated to maintain a perature above freezing within the areas of that building.

systems also maintain a slightly negative pressure when compared to the outside atmosphere.

s is performed to ensure that there will be no inadvertent unmonitored release to the site area m the reactor building.

er vapor from quiescent evaporation of liquid waste may be released into the ventilation em. The process allows only the distillate vapor into the ventilation system, assuring positive trol over the species and concentration of radionuclides released with Reactor Building aust air.

tilating air flow is routed to areas of progressively greater radioactive contamination prior to l exhaust. Back-draft dampers are provided to prevent reverse flow between areas of different tamination potential.

ering of supply air is provided to reduce the presence of dust particles.

ctor Building main supply and exhaust units consist of fan, motor, and their associated trols.

8.1.2 System Description Reactor Building and SFPI HVAC systems provide for the protection of personnel and ipment from airborne radioactive contaminants and excessive thermal conditions. Air flow is cted to areas of progressively greater radioactive contamination prior to exhaust.

Reactor Building is provided with supply and exhaust ventilation to ensure proper air flow ction and remove heat generated from equipment.

SFPI system includes variable speed supply and exhaust fans to maintain space temperature hin acceptable limits while also maintaining a negative pressure within the SFPI envelope tive to the outside and to Reactor Building areas outside the SFPI envelope.

low diagram of the Reactor Building HVAC system is given in Figure 3.2-12. The SFPI AC system is shown in Figure 3.2-6.

ctor Building HVAC supply segment of the system provides fresh air to all levels in the Reactor Building outside SFPI envelope. Outside air passes through fixed louvers, a damper, filters, and electric heating

s. One fan is available to deliver air flow. Electric unit heaters are provided inside the drywell freeze protection. Exhaust air flow combines in a common duct and continues on to the main aust fan plenum.

tem components, in addition to those mentioned above, include screens, filters, ductwork with pers, supply outlets, return and exhaust intakes, heating coils, and instrumentation and trols. Control actuation, indication, and alarm instrumentation are incorporated in a central AC master control panel.

I HVAC System supply segment of the system provides fresh air to the operating floor of the Reactor lding, portions of the 82 feet 9 inches elevation and the spent fuel pool pump area. Outside air ses through fixed louvers in the side of the reactor building wall, filters, and electric heating

s. A single variable speed 100% capacity fan is available to deliver air flow.

ngle variable speed exhaust fan discharges air from the SFPI envelope through a HEPA filter, fixed louver in the reactor building wall. The exhaust fan is operated in conjunction with the ply fan to maintain space temperatures within acceptable limits while also maintaining a slight ative pressure in the SFPI envelope relative to the outside and to Reactor Building areas ide the SFPI envelope.

ctric unit heaters are installed in all SFPI areas to maintain acceptable space temperatures.

cation, and alarm instrumentation are incorporated in a local control panel. Indication and m functions are provided in the Millstone Unit 2 Control Room.

ural ventilation cooling capability is also provided by opening the Reactor Building truck bay rs, equipment hatch garage doors and the tornado dampers located on the Reactor Building

. This path would be used following an extended loss of all spent fuel pool cooling capability.

8.1.3 Safety Evaluation Reactor Building and SFPI heating and ventilation systems maintain environmental ditions in building spaces (to support personnel comfort or operation of equipment located on e spaces), direct ventilation air from areas of low radioactive contamination to areas of gressively greater contamination (to minimize the spread of contamination), and vent entially contaminated exhaust air. Natural ventilation cooling capability is also provided for nt fuel pool cooling following an extended loss of all active pool cooling capability. The ctor Building and SFPI heating and ventilation systems are not safety related, but are required he permanently defueled condition because they house SSCs that are associated with the safe age and handling of irradiated fuel or radioactive waste.

8.2 Radwaste Building Ventilation System 8.2.1 Design Bases Radwaste Building ventilation system operates to supply filtered air to this building's areas.

ply air is filtered. The presence of dust particles potentially increases the spread of radioactive tamination.

s system also filters the exhaust air prior to its discharge, to limit the release of any radioactive taminants to the environment.

tilating air flow is routed to areas of progressively greater radioactive contamination potential r to final exhaust. Back-draft dampers are provided to prevent reverse flow between areas of erent contamination potential.

8.2.2 System Description ure 3.2-13 shows the ventilating flow through the Radwaste Building. The ventilating system esigned to provide a passive supply of filtered air and exhaust it after filtration. Air is drawn ugh the building by the main exhaust fan. An exhaust filter unit is provided.

side air is drawn into the system through two inlets above the roof of the building and ected by bird screening. The air is drawn through a filter designed to remove dust. A header veys fresh air to various areas of the building.

exhaust air is passed through the filtering system before discharge through the main exhaust 8.2.3 Safety Evaluation Radwaste Building ventilation directs ventilation air from areas of low radioactive tamination to areas of progressively greater contamination (to minimize the spread of tamination), and vents potentially contaminated exhaust air. The Radwaste Building tilation system is only required, in the permanently defueled condition, to support personnel ess to the space.

8.3 Intentionally Deleted 8.4 Turbine Building Heating and Ventilation 8.4.1 Design Bases Turbine Building ventilation system is operated to maintain a slight negative pressure in the ding to prevent any radioactive out-leakage, as well as, to provide fresh air to support onnel access.

8.4.2 System Description sh air is supplied to the Turbine Building through louvers in the walls and roof.

ventilation system is arranged with one supplementary transfer fan and connecting ductwork nduce flow to the north end of elevation 14 feet 6 inches.

Turbine Building exhaust system collects air from various areas into an exhaust air header discharges it into a plenum which also receives air from the Reactor Building and Liquid waste Building. One exhaust fan is furnished to handle the combined exhaust from these three dings. This fan discharges into a duct which runs along the north wall of the Reactor Building ore releasing the exhaust air to the environment. Potentially contaminated areas in the Turbine lding are maintained at a negative pressure by exhausting from these areas. The exhaust air is wn from adjacent spaces. This arrangement controls the air flow pattern and prevents out age.

Turbine Building ventilation air is normally discharged to the atmosphere without treatment.

ow diagram of the Turbine Building area ventilation system is shown in Figure 3.2-7.

Turbine Building ventilation system directs ventilation air from areas of low radioactive tamination to areas of progressively greater contamination (to minimize the spread of tamination), and vents potentially contaminated exhaust air. The Turbine Building ventilation em is only required, in the permanently defueled condition, to support personnel access to the ce.

9 FIRE PROTECTION SYSTEMS licensees Nuclear Plant Fire Protection Program has been developed to ensure that any le fire will not cause an unacceptable risk to public health and safety, and will not ificantly increase the risk of radioactive release to the environment.

ire Protection Program has been established at Millstone Unit Number 1. This program blishes the fire protection policy for the protection of structures, systems, and components ortant to the defueled condition of the unit and the procedures, equipment, and personnel uired to implement the program.

9.1 Design Bases achieve and maintain a high level of confidence for the Fire Protection Program, it has been anized and is administered using the defense-in-depth concept. The defense-in-depth concept res that if any level of fire protection fails, another level is available to provide the required nse. In fire protection terms, this defense-in-depth concept consists of the following levels; Preventing fires from starting, Early detection of fires that do start, and Controlling and/or extinguishing them quickly so as to limit their damage.

e of these levels can be perfect or complete, but strengthening any one level can compensate ome measure for weaknesses, known or unknown, in the others.

9.2 System Description 9.2.1 Site Water Supply System underground fire protection water supply system consists of a 12 inch cast and ductile iron, ent-lined pipe extending around Millstone Unit Number 1, 2, and 3 in a loop arrangement.

supply system services individually valved lines feeding fixed pipe water suppression ems (sprinklers, waterspray, and standpipes) throughout the plant and hydrants located around exterior of the plant.

e. The Millstone Unit Number 3 pumphouse contains one electric driven pump (M7-8), fed m Millstone Unit Number 3 power, and the diesel-driven fire pump (M7-7). The Millstone t 2 pumphouse contains one electric driven pump (P-82) fed from Unit 2 power. All three ps have individual connections to the underground supply system. Maximum system flow pressure requirements can be met with any one of the three pumps out of service.

tem operation is such that a 50 gpm electric jockey pump (M7-11) maintains system pressure utomatically starting when line pressure drops to 105 psig and will run until pressure reaches psig as indicated by a line pressure switch. A hydro-pneumatic tank is provided in the system revent short cycling of the jockey pump. At pressures below 105 psig, the MP2 P-82 electric p first starts at 98 psig to maintain system pressure and flow. The Millstone Unit Number 3

-8 electric pump then will start at 85 psig and it is fed 480 VAC from MCC-CD-6 (MCC ber 22A-2 Compartment number 1A). This pump is auto-started by a pressure switch set at 85 decreasing, while the M7-7 diesel-driven fire pump is auto-started by a separate pressure tch set at 75 psig decreasing. The diesel pump is started by its own self-contained battery em. A battery charger is provided for recharging. Both Millstone Unit Number 3 electric and el-driven fire pumps deliver 2000 gpm at 100 psi discharge pressure and remain in operation l they are manually shut down. Electrical interlocks stop the jockey pump when either of the Millstone Unit Number 3 fire pumps start.

fire pumps are supplied from two 250,000 gallon ground level tanks. The tanks are matically filled through a water line fed from city water.

major fire in any location of the MP-1 site should occur, the combined water tank and makeup er capacity would provide an adequate water supply for MP-1. The necessary pressure and would be maintained through the use of any two simultaneously operating 2,000 gpm rated ps.

9.2.2 Fixed Suppression Systems fire protection features for the Unit 2 Control Room are discussed in Section 9.10 of the lstone Unit 2 Final Safety Analysis Report.

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)

• Dry Pipe Manual Sprinkler Systems (Condenser Bay, Turbine Building Truck Unloading Area, and Reactor Building Rail Airlock Sprinkler Systems)

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 Buildings) 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 accessible to the plant operators or responding fire department members outside the fire areas being protected by the dry pipes.

The sprinkler systems 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.

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.

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 Unloading 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 have 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.

9.2.3 Portable Suppression Capabilities 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/CMS 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

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 shown in the FHA (Reference 3.2-19).

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 Underwriters Laboratories (UL) listed.

9.2.4 Fire Detection and Alarm Systems fire detection and alarm systems installed in the plant are designed in general compliance h NFPA Standard Number 72D, Standard for the Installation, Maintenance, and Use of prietary Protective Signaling Systems, and with NFPA Standard Number 72, National Fire rm Code.

detection systems are used for early warning detection and in some cases may have the ability to actuate fixed fire suppression systems.

ection devices consist of fixed temperature detectors and smoke detectors. Smoke detectors of the spot type, employing the ionization principle. Specific application of these detectors in h fire area is detailed in the FHA (Reference 3.2-19).

eneral, the installation of detector units is in accordance with the intent of the guidelines set h in NFPA Standard Number 72E, Standard on Automatic Fire Detectors.

/smoke detectors, as with waterflow indicators, and valve tamper devices are arranged to smit signals to local alarm panels and a fixed suppression system control panel, if applicable.

uation signals are also transmitted through the local alarm panels to control panels in the tral Monitoring Station (CMS). A Fire Alarm panel located in the CMS monitors those areas essary to support the Spent Fuel Pool Island. Trouble signals for these devices are transmitted similar manner. A general alarm is provided in the Unit 2 Control Room. Identification of the ct alarm or trouble signals must be performed locally in the Unit 1 CMS.

alarm system also monitors other miscellaneous fire protection system features.

9.2.5 Ventilation Systems and Smoke Removal oval of the products of combustion from any specific plant area requires the use of the mal plant ventilation system, which is designed to handle the expected normal environment hin a given area or the use of portable exhaust fans by the fire brigade. There are no cable nels, culverts, or other unventilated areas that pose any special venting problems. Removal of

ventilation and filtration systems of potential radiation release areas are discussed in detail the Reactor, Turbine, Radwaste, Radwaste storage, and Screenhouse Buildings in the FHA, erence 3.2-19.

9.3 Safety Evaluation and Fire Hazards Analysis 9.3.1 Evaluation Criteria evaluation of the overall Fire Protection Program as indicated by the FHA, ference 3.2-19), found that the program does provide reasonable assurance that a fire will not se an unacceptable risk to the public health and safety. The fire protection program omplishes this by assuring a fire will not significantly increase the risk of radioactive release he environment. Therefore, the Fire Protection Program meets the basic requirements of eral Design Criteria 3 and 5 as applicable to a permanently defueled facility. Branch hnical Position (BTP) APCSB 9.5.1, Guidelines for Fire Protection for Nuclear Power nts, provides the implementing criteria for GDC 3 and gives the general guidelines used to ew Millstone Unit Number 1. BTP APCSB 9.5.1 provides the guidelines acceptable to the C staff for implementing the following criteria:

General Design Criterion 3 (10 CFR 50, Appendix A) - Fire Protection.

Defense-in-Depth Criterion: For each fire hazard, a suitable combination of fire prevention, fire detection and suppression capability, and ability to withstand safely the effects of a fire is provided. Both equipment and procedural aspects of each are considered.

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).

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.

Backup Fire Suppression Capability: Total reliance for fire protection is not placed on a single automatic fire suppression system. Appropriate backup fire suppression capability is provided in the form of portable fire extinguishers or hose stations.

ddition to the specific guidance of the BTP, the evaluation considered the adequacy of the Fire tection Program on the effects of potential fire hazards throughout the plant based on sound protection engineering practices and judgments.

Protection was evaluated by conducting a fire hazard analysis of individual fire areas and fire es within the plant. The analysis methodology is described in the Fire Hazards Analysis ference 3.2-19).

9.3.3 Fire Hazard Analyses Results fire hazards analysis results for each fire area are contained in the FHA (Reference 3.2-19).

9.4 Inspection and Testing ministrative controls are provided through existing Plant Administrative Procedures, rating Procedures and the Quality Assurance Program to ensure that the Fire Protection gram and equipment is properly maintained. This includes QA audits of the program lementation, conduct of periodic test inspections, and remedial actions for systems and iers out of service.

technical requirements found in Millstone Unit Number 1 Technical Requirements Manual cribe the limiting condition for operation and surveillance requirements for the fire protection em. These technical requirements ensure the fire protection system is properly maintained and rated.

fire protection equipment and systems are subject to periodic inspections and tests in ordance with the intent of National Fire Codes and the Fire Protection Program.

following fire protection features 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) ipment out of service including fire suppression, detection, and barriers will be controlled ugh the administrative program and appropriate remedial actions taken. The program requires mpairments to fire protection systems to be identified and appropriate notification given to the Fire Marshal for evaluation.

ipment to service.

9.5 Personnel Qualification and Testing 9.5.1 Fire Protection Organization officer responsible for the Fire Protection Program at Millstone Unit Number 1 is defined in QAP. Formulation, and assessment of the effectiveness of the program are delegated as cated in Reference 3.2-20, the Fire Protection Program Manual.

9.5.2 Fire Brigade and Training Site Fire Brigade and Nuclear Training are a site (Units 1, 2, and 3) organizations. The lstone Site Fire Brigade consists of a minimum of a Shift Leader and four Fire Brigade onnel. MP-2 supplies an advisor, who is at a minimum a fully qualified Unit 1 Plant ipment Operator, to the Fire Brigade Shift Leader. The advisor will provide direction and port concerning plant operations and priorities.

mbers of the Fire Brigade are trained by the Nuclear Training Department.

Fire Brigade personnel are responsible for responding to all fires, fire alarms, and fire drills.

nsure availability, a minimum of a Shift Leader and four Fire Brigade personnel remain in the er controlled area and do not engage in any activity which would require a relief in order to ond to a fire (e.g., continuous fire watch).

ssistance is needed to fight a fire, additional equipment and manpower is supplied by the off local fire departments. Within a 5 mile radius of the plant there are numerous local volunteer companies. Letters of commitment to supply public fire department assistance have been ined from these fire companies.

Shift Leader coordinates the Site Fire Brigade activities, and ensures proper communications coordination of support for the local fire department chief or officer in charge once on site, other on site activities (e.g., Chemistry, Health Physics, and Security).

lear Training coordinates with the Site Fire Marshal and periodically familiarizes local fire artment personnel with the Stations layout and fire fighting equipment. The Site Fire Marshal rdinates with the Site Fire Brigade Personnel and all Unit Shift Managers, informing them of status of the site fire protection equipment, should equipment become inoperable or vailable.

Protection drills are planned and critiqued by Nuclear Training and members of the agement staff responsible for plant fire protection. Performance deficiencies of the Fire gade or of individual Fire Brigade personnel are remedied by scheduling additional training the Site Fire Brigade or individuals.

QA Program has been applied via the Fire Protection Program Manual to the FPSs which vide a function for the operating units.

10 REFERENCES 1 Docket Number 50-245, LS05-82-03-060, J. Shea to W.G. Counsil, 'SEP Topic IX-1, Fuel Storage (Millstone 1), March 9, 1982.

2 Docket 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 Docket 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.

4 Docket 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.

5 Docket 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.

6 Docket 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.

7 Reference deleted.

8 J.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.

9 Holtec Report Number H1-971914, Revision 1, Analysis Of 1675 Pound Fuel Assembly System Drop Onto The Irradiated Fuel Assembly.

10 Holtec Report Number AH1-971691, Revision 0, Criticality Safety Analysis Of The MP1 Racks With A Dropped Fuel Assembly.

11 Holtec Report Number H1-971698, Revision 0, Flow And Temperature Field Analysis Of Localized Cell Blockage In The Millstone Unit Number 1 Spent Fuel Pool.

12 Holtec Report Number H1-971675, Revision 1, Analysis Of Tetrabor And Boraflex Racks Under 1675 Pound Fuel Assembly System Impact.

14 Holtec Report Number H1-992125, Revision 0, Steady State Temperature of Millstone Unit 1 SFP and RB with No Active SFP Cooling.

15 Docket 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.

16 Docket Number 50-245, NUREG-0824, Integrated Plant Safety Assessment, Systematic Evaluation Program, Millstone Nuclear Power Station, Unit Number 1, February 1983, Topic IX-3, Station Service and Cooling Water Systems.

17 Docket Number 50-245, B10292, W.G. Counsil to D.M. Crutchfield Millstone Nuclear Power Station, Unit Number 1, SEP Topic IX-5, Ventilation Systems, November 19, 1981.

18 Docket 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.

19 Fire Hazard Analysis Millstone Unit Number 1, Revision 6, July 2000.

20 Millstone Nuclear Power Station Fire Protection Program Manual.

thheld under 10 CFR 2.390 (d) (1)

FIGURE 3.2-1 P&ID: SFPI, FUEL POOL COOLING SYSTEM

thheld under 10 CFR 2.390 (d) (1)

FIGURE 3.2-2 P&ID: SFPI, FUEL POOL COOLING SYSTEM

thheld under 10 CFR 2.390 (d) (1)

FIGURE 3.2-3 P&ID: SFPI, FUEL POOL COOLING SYSTEM (REFUELING BELLOWS LEAK DETECTION)

thheld under 10 CFR 2.390 (d) (1)

GURE 3.2-4 P&ID: REACTOR BUILDING AND HVAC ROOM SFPI SECONDARY COOLING (DHR) SYSTEM

thheld under 10 CFR 2.390 (d) (1)

FIGURE 3.2-5 P&ID: REACTOR BUILDING SFPI, MAKE-UP WATER SYSTEM

thheld under 10 CFR 2.390 (d) (1)

FIGURE 3.2-6 P&ID SFPI HVAC SYSTEM COMPOSITE

thheld under 10 CFR 2.390 (d) (1)

FIGURE 3.2-7 P&ID: HVAC B.O.P. SYSTEM COMPOSITE

FIGURE 3.2-8 INTENTIONALLY DELETED FIGURE 3.2-9 INTENTIONALLY DELETED FIGURE 3.2-10 INTENTIONALLY DELETED FIGURE 3.2-11 INTENTIONALLY DELETED thheld under 10 CFR 2.390 (d) (1)

FIGURE 3.2-12 P&ID: HVAC BALANCE OF PLANT SYSTEM COMPOSITE

thheld under 10 CFR 2.390 (d) (1)

FIGURE 3.2-13 P&ID: HVAC SYSTEM (RADWASTE STORAGE BUILDING)

thheld under 10 CFR 2.390 (d) (1)

FIGURE 3.2-14 FIRE PROTECTION COMPOSITE

SOURCE TERMS h the permanent defueled condition of Unit 1, fission, corrosion, and activation products from ration are no longer produced. The radioactive inventory that remains is primarily attributable activated reactor components and structural materials and residual radioactivity. The umulation of small amounts of solid waste as evaporator bottoms or contaminated materials easily be controlled. Planned liquid effluent releases will be evaluated prior to release, and ropriate controls (e.g., monitoring) will be established. The Radiological Effluent Monitoring Offsite Dose Calculation Manual ensures that Unit 1 complies with 10 CFR 50, Appendix I.

1 FACILITY DESIGN FEATURES iation shielding was provided to restrict radiation emanating from various sources throughout plant. The primary objective of radiation shielding is to minimize the radiation exposure of t personnel and the general public.

lstone Unit Number 1 is permanently shutdown and many installed components which are vided with shielding, are no longer required to safely store irradiated fuel. However, many of e installed components continue to contain radioactive material or remain radioactive mselves. Shielding that was originally designed to shield these components while they ported reactor operation, continues to provide shielding from the residual activity in the manently shutdown condition.

h the vessel in a drained down condition, a concrete shielding package is installed over the tor vessel head and reactor cavity floor to provide shielding from activated reactor vessel rnals.

1.1 Design Basis mal operating conditions determined the major portion of the original plant shielding design uirements. Two exceptions to this were the Control Room where shielding was determined by ation levels produced during the loss-of-coolant accident and the shutdown cooling system re shielding was determined by shutdown conditions. Although these conditions are no longer licable, these were the bases for the unit shielding.

1.2 Ventilation rmation on ventilation systems is contained in Chapter 3.

2 RADIATION PROTECTION PROGRAM 2.1 Organization radiation protection program is established to provide an effective means of radiation ection for permanent and temporary employees and for visitors at the station. The radiation ection program is developed and implemented through the applicable guidance of Regulatory des 8.2, Revision 0; 8.8, Revision 3; and 8.10 Revision 1.

radiation protection department and line function management implement and enforce the ation protection program.

officer responsible for implementing the radiation protection program is defined in the QAP.

qualifications specified in ANSI N18.1-1971.

1 POLICY CONSIDERATIONS the policy of the licensee to maintain individual and plant personnel total radiation exposure ARA. The licensees ALARA policy complies with 10 CFR 20 and 10 CFR 50.

1.1 Design Considerations basic objective of facility radiation shielding is to reduce external dose to plant personnel in junction with a program of radiologically controlled personnel access and occupancy in ation areas to levels which are both ALARA and within the regulations defined in 10 CFR 20.

h the reactor shutdown and all fuel stored in the spent fuel pool, the number and magnitude of ntial radiation sources have been reduced substantially from the original bases for the ation protection design features.

1.2 Operational Considerations iation surveys have been performed and will continue to be performed to ensure that plant s are correctly posted and barricaded.

uid waste from the Unit 1 Reactor Building Floor Drain (RBFD) System is collected in two (2) ve RBFD sumps. There are three (3) active sumps in the Unit 1 Radwaste Building that pump he A RBFD sump. In addition, water collected in the Unit 1 Turbine Building sumps, Unit 1 tilation Exhaust Duct (abandoned), Site Stack sump, and similar miscellaneous waste water ived from Units 2 and 3 are collected in the RBFD sumps.

primary method for disposing of this waste water will be using the Waste Water Processing tem (WWPS) located in the Unit 1 Reactor Building. The Waste Water Processing System sists of four (4) 10,000 gallon Sample Tanks, recirculation pump, demineralizer, filters and ciated piping. The A RBFD sump will pump to the WWPS sample tanks, where the water be batch recirculated and sampled before subsequent discharge. Radiological monitoring will onducted using an in-line Liquid Effluent Monitor (RE-MG-110). Prior to discharge through N-001A (Emergency Service Water discharge piping to discharging canal), dilution flow uirements will be established by crediting Unit 2 Circulating Water Flow to the common harge canal. In the future, the WWPS will be used to process, sample and discharge Unit 1 nt Fuel Pool water after all spent fuel assemblies are removed from the Spent Fuel Pool.

alternative method for processing waste water will be using an eight (8) gallon per hour ospheric evaporator. Waste water collected in the A RBFD sumps will be pumped to a ing tank. Collected liquids may be surveyed for activity and pumped to the evaporator. The illate vapor will be diluted in the Balance of Plant (BOP) Reactor Building Exhaust flow and ased as a ground level release. Radiological monitoring will be conducted by a particulate nitor in the BOP ventilation exhaust or by screening a grab sample of the process liquid.

centrates in the bottom of the atmospheric evaporator will be collected as required, and osed as Low Specific Activity (LSA) trash.

either of these liquid process methods are available, the RBFD sumps can be pumped to tainers which would permit the collected liquids to be processed at a later date, or sent offsite processing.

plant has no capability for processing concentrated waste solutions to a solidified product.

Activated Waste (DAW) is processed and stored in appropriate containers to allow for offsite ment.

rim on site storage facilities accept waste from Millstone Units 1, 2 and 3. Information rding facility design criteria is presented in Section 11.4 of the Millstone Unit 3 Final Safety lysis Report.

1 DESIGN BASES design basis objective of solid waste management is to provide for processing, packaging and dling solid dry wastes, and to allow for radioactive decay and/or temporary storage prior to ment off site and subsequent disposal.

d radwaste handling at Millstone Unit 1 ensures compliance with the following regulations Regulatory Guides:

10 CFR 20, Standards for Protection Against Radiation 10 CFR 50, Appendix I 10 CFR 61.55, Classification of Waste for Near Surface Disposal 10 CF 61.56, Waste Characteristics 10 CFR 71, Quality Assurance Criteria for Shipping Packages of Radioactive Material Regulatory Guide 1.143, Design Guidance for Radioactive Waste Management Systems, Structures and Components Regulatory Guide 8.8, ALARA Provisions 2 SYSTEM DESCRIPTION solid waste management process is designed to accommodate the following radioactive tes, which are typical for BWR power plants:

active wastes, which consist of contaminated clothing, tools and small pieces of equipment cannot be economically decontaminated; miscellaneous paper, rags, etc., from contaminated s; air filters from radioactive ventilation systems; used reactor equipment such as control rod es, temporary control curtains, fuel channels and in-core ion chambers - Radioactivity levels most DAW are low enough to permit handling by contact, it is processed and stored in ropriate containers to allow for off site shipment. Used radioactive equipment may be stored sufficient time to permit decay before removal for interim storage or off site shipment.

cific activity (LSA) dry active waste.

maries of solid waste shipments, types, volumes, and radionuclide composition are given in erence 4.5-1.

3 REFERENCES 1 Millstone Nuclear Power Station Unit Number 1, Docket Number 50-245, Annual Radioactive Effluents Report.

1 DESIGN 1.1 Design Basis effluent radiation monitoring system (RMS) provides nonsafety related functions. The em provides the means for compliance with Nuclear Regulatory Commission (NRC) ulations 10 CFR 20, 10 CFR 50 Appendix A General Design Criteria (GDC) 60, 63 and 64, CFR 50, Appendix I and Regulatory Guides (RG) 1.21, 4.15 and 8.8.

1.2 System Design Description I Ventilation Exhaust Monitor SFPI ventilation exhaust radiation monitor is designed with the capability to monitor, cate and record the discharge of gaseous radioactivity. Capability for sampling of particulate vity is provided. Annunciation in the Millstone Unit 2 Control Room occurs if setpoints are eeded.

hough the monitor cannot determine the individual activity level of the radionuclides in the uent gas, it provides the overall level and a basis for correlation with laboratory analyses of r and grab sample activities.

SFPI gas sample is taken from the exhaust duct near the reactor building exhaust plenum. A le point sample nozzle is positioned to obtain a representative sample of the turbulent and l mixed exhaust air. The monitor is located in a heated enclosure on the 65 foot elevation of Reactor Building directly below the exhaust duct. The sample passes through a particulate r and a shielded detection chamber (fixed volume) and is then expelled back into the exhaust

t. The particulate filters are periodically removed for detailed radiological quantitative lysis.

detector readout is sent to the PLC for display and recording. The range of indication is 10-6 ci/cc to 1 x 100 ci/cc (Kr-85).

P Ventilation Exhaust Monitor Unit Number 1 BOP ventilation exhaust flow is continuously sampled for radioactive iculates. The sample is taken from the exhaust duct which runs along the north exterior wall of Reactor Building. A single point sample nozzle is positioned to obtain a representative sample ell mixed exhaust air. The particulate sample skid is located in an insulated enclosure on the foot elevation, north wall, of the Reactor Building. The sample passes through a particulate r and is then expelled back into the exhaust duct. The particulate filter is periodically removed detailed radiological quantitative analysis.

2.1 Design Bases purpose of the ARM system is to warn of abnormal radiation levels in the SFPI where oactive material may be present, stored, handled, or inadvertently introduced. The system also vides information concerning radiation at selected locations in the SFPI.

2.2 System Description area radiation monitoring system detects, measures, and indicates ambient gamma radiation e rates at selected locations in the SFPI. It provides audible and visual alarms in the Millstone t 2 Control Room (locally at some locations) when radiation levels exceed pre-selected values when a monitor has operational failure. Table 4.6-2 lists the area radiation monitor locations ranges.

ueling Floor Area Radiation Monitor refueling floor ARM is a 3 channel digital unit. Each detector is a gama sensitive GM tube ted as described in Table 4.6-2. Each channel is provided with a failsafe High, Warn and ure alarm relay as well as an analog output. The alarms and analog output are sent to the PLC recording and alarm. Each unit has a built in check source and local audible and visual alarm cation.

3 REFERENCE 1 Letter from W.G. Counsil to D.G. Eisenhut dated July 1, 1981, Haddam Neck Plant, Millstone Nuclear Power Station, Unit Numbers 1 and 2, Post TMI Requirements -

Response to NUREG-0737, Docket Numbers 50-213, 50-245, 50-336.

Monitor Detector Range Trip Function PI ventilation exhaust (1) Beta Sinctillator 10-6 to 100 ci/cc None

CONVERTER LOCATIONS FOR MILLSTONE UNIT NO. 1 REACTOR BUILDING Station Number SENSOR AND CONVERTER LOCATION Range mR/hr M-SFPI-01 CH1 West Refuel Floor 0.01-102 M-SFPI-01 CH2 East Refuel Floor 0.01-102 M-SFPI-01-CH3 West Refuel Floor Hi Range 10.0-106

INTRODUCTION uly of 1998, the licensee certified to the NRC that Millstone Unit Number 1 had both manently ceased operations and that all fuel had been removed from the reactor vessel and ed in the spent fuel pool (Reference 5.1-1). Since Millstone Unit Number 1 will never again r any operational mode, reactor related accidents are no longer a possibility.

remaining analyzed accident that is in this chapter is the fuel handling accident. Conservatism quipment design, conformance to high standards of material and construction, the control of hanical and pressure loads, and strict administrative control over plant operations all serve to re the integrity of the fuel in the spent fuel pool.

w hazards, new initiators, and new accidents that may challenge offsite guideline exposures, be introduced as a result of certain decommissioning activities. These issues will be luated when the scope and type of the decommissioning activities are finalized.

1 ACCIDENT EVENT EVALUATION 1.1 Unacceptable Results for Design Basis Accidents (DBAs) following are considered to be unacceptable safety results for DBAs:

Radioactive material release that results in dose levels that exceed the guideline values of 10 CFR 100.

Nuclear system stresses in excess of those allowed for the accident classification by applicable industry codes.

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.

1.2 Fuel Handling Accident Assumptions l handling accident analysis assumptions are listed on Table 5.2-1.

1.3 Results results of the Fuel Handling Accident analytical evaluation are provided in Section 5.2.

1.4 Radiological Consequences sequences of radioactivity release during a fuel handling accident are presented in tion 5.2..

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.

the bounding accident analysis, an inadvertent release of radioactivity, as a result of a fuel dling accident in the spent fuel pool, was evaluated and is discussed below.

h the permanent cessation of operations of Millstone Unit Number 1, the prior limiting fuel dling accident, i.e., a fuel assembly drop onto the top of the core during fuel-handling rations, was no longer part of the plants design and licensing basis. Several fuel handling dent scenarios are still possible in the spent fuel pool. These scenarios are identified later in Section.

adiological consequences of a fuel handling accident in the spent fuel pool are described in section. For conservatism, a bounding analysis was made to calculate the radiological release m a failure of all fuel rods in four (4) fuel assemblies in the spent fuel pool. Other assumptions n into consideration are described later in this Section. The off site radiological consequences his release, i.e., from 4 failed fuel assemblies or, for example, 248 fuel rods for 8x8 fuel mblies, are substantially less than the 10 CFR Part 100 limits and are tabulated in this section.

1 FUEL HANDLING ACCIDENT SCENARIOS IN THE SPENT FUEL POOL consequences of the following postulated fuel handling drop events were evaluated:

Spent fuel pool gate (1200 lbs.) drop onto irradiated fuel and fuel storage racks in the spent fuel pool.

New fuel assembly drop (600 lbs.) onto irradiated fuel and fuel storage racks in the spent fuel pool.

Lifting of a Tri-Nuc Filter skid (965 lbs.) into the spent fuel 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.

se analyses utilized two sophisticated elasto-plastic finite-element models. The first represents fuel assembly components, the second represents the rack with its pedestals, liner and erlying reinforced concrete structure. The LS-DYNA3D computer code ference 5.2-1) was used. Conservative assumptions and restrictive inputs were utilized to lt in an upper bound estimate of the calculated damage for the postulated drop event.

following assumptions were utilized in the analysis:

arding the impactor movement and the target:

Both the impactor and the target are submerged.

The trajectory of the impactor is vertical.

The form drag force opposed to the impactor movement is proportional to its velocity squared.

The friction drag force is conservatively neglected.

arding 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 conservatively considered as a prefect rigid body, without deformability or energy absorption capacity.

arding 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 fuel rod and the bottom fitting) is not considered as a failure of the supported rod.

analysis of these additional accident scenarios has determined that the limiting event is the p of the spent fuel pool gate, which can result in extensive damage of the fuel assemblies, wing a total of 54 ruptured fuel rods. The drop of the new fuel assembly resulted in damage to targeted fuel assemblies, but no ruptured fuel rods were recorded for either the impactor or the et. Drop of an irradiated fuel assembly results in failure of all 64 guide ends, but no rupture of rods occurs. These results bounded all fuel types stored within the Millstone Unit Number 1 nt fuel pool for the analyses performed to date.

2 RADIOLOGICAL CONSEQUENCES ce the licensee has certified to the NRC that there is a permanent cessation of operations of lstone Unit Number 1 and that fuel has been permanently removed from the reactor vessel, a ulation evaluating the radiological consequences of a fuel handling accident in the spent fuel l was performed and eventually chosen as the new bounding accident (Reference 5.2-2).

ing into account the actual source term of the fuel in the spent fuel pool (i.e., appropriate ay time of fuel), the reanalysis assumed four fuel assemblies (e.g., 248 rods in an 8x8 mbly) failed in the spent fuel pool and resulted in an unfiltered, i.e., no Standby Gas atment System (SGTS) available and secondary containment not set, puff release. Additional mptions and input parameters are given in Table 5.2-1. This reanalysis was performed using guidelines of Standard Review Plan 15.7.4 Rev. 1 and Regulatory Guide 1.25. Doses were ulated using the TACT-III, ORIGEN-2, and ELISA computer codes.

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 exclusion area boundary 1.03E-03 REM Whole-body dose (calculated as TEDE) at the low-population zone 3.20E-05 REM se doses are well within the limits of 10 CFR 100, and are therefore acceptable.

es were also calculated to the Millstone Unit Number 2 Control Room. The results of this e assessment is as follows:

Thyroid dose to the Millstone Unit Number 2 Control Room 7.65E-02 REM Whole-body dose to (calculated 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 se doses are less than the limits specified in GDC 19. Doses were not calculated for the lstone Unit Number 3 control room since the atmospheric dispersion factor (/Q) is roximately 50 times less that the (/Q) to the Millstone Unit Number 2 control room.

refore, the dose to the Millstone Unit Number 3 control room would be approximately 50 es less than the Millstone Unit Number 2 control room dose.

3 REFERENCES 1 LS-DYNA3D, Version 932, Livermore Software Technology Corporation, May 1, 1995.

2 Calculation Package NUC-197, MP1 Defueled State - Radiological Analysis of a Fuel Handling Accident, Duke Engineering and Services, October 11, 1999.

NO. 1 Assumption Basis

1. Core Power Level During Irradiation = 2011 MWt Technical 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 burnup Regulatory Guide 1.25 See Ref. 5.2-3.

4. Pool Scrubbing Factor = 60 Extrapolation 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: 580 Technical Specifications

7. For radiological dose assessment: Number of fuel assemblies DSAR Section 5.2.2 assumed to fail = 4

8. Release fractions from fuel rods: Regulatory Guide 1.25 & conservative assumption

• 30 percent Noble Gases

• 12 percent Iodines

9. No credit taken for secondary containment Technical Specifications

• SGTS not in operation

• Puff release is an unfiltered ground release

10. Breathing rate = 3.47 x 10-4 m3/sec Regulatory Guide 1.25

11. Ground level dispersion factor (/Q): SEP Topic 11-2.c, Docket Number 50-245 EAB (0-2 hr.) = 6.10 x 10-4 sec/m3 LPZ (0-4 hr.) = 1.90 x 10-5 sec/m3

12. Decay Time for fuel = 3.8 years Based on the MP1 shutdown on November 4, 1995.

ORGANIZATIONAL STRUCTURE rmation regarding the organizational structure is presented in Section 1.0 of the Quality urance Program Description Topical Report (Reference 6.1-1). With the exception given w, that information is incorporated herein by reference.

owner, holding 100 percent of the Millstone Unit Number 1 nuclear plant, is Dominion lear Connecticut, Inc..

1 MANAGEMENT AND TECHNICAL SUPPORT ORGANIZATION rmation regarding the management and technical support organization is presented in tion 1.0 of Reference 6.1-1. That information is incorporated herein by reference.

1.1 Technical Support for Operations rmation regarding the technical support for operations is presented in Section 1.0 of erence 6.1-1. That information is incorporated herein by reference.

1.2 Organizational Arrangement rmation regarding the organizational arrangement is presented in Section 1.0 of erence 6.1-1. That information is incorporated herein by reference.

2 OPERATING ORGANIZATION 2.1 Plant Organization plant organization is as shown in Reference 6.1-1.

2.2 Plant Personnel Responsibilities and Authorities rmation regarding the plant personnel responsibilities and authorities is presented in Section of Reference 6.1-1. That information is incorporated herein by reference.

2.3 Operating Shift Crews minimum shift crew composition is contained in the Administrative Controls section of the lstone Unit Number 1 Technical Specifications.

3.1 Qualification Requirements lifications of plant managerial and supervisory personnel are established by the American ional Standards Institute (ANSI) N18.1 (Reference 6.1-2) except for the following:

a. The Operations Manager or at least one operations middle 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.

4 REFERENCES 1 Quality Assurance Program Description Topical Report.

2 American National Standards Institute, ANSI N 18.1-1971, Selection and Training of Nuclear Power Plant Personnel.

hnical Specifications set forth the limits, operating conditions and other requirements for the ection of the health and safety of the public. These specifications have been written in illment of 10 CFR 50.36 and are controlled pursuant to 10 CFR 50.90, 50.91, and 50.92.

hnical Specifications are maintained as Appendix A to the operating license.

Technical Requirements Manual (TRM) contains clarifications for certain technical cifications and a central location for other documents which place operating limits on the

t. Changes to the TRM are controlled pursuant to the 10 CFR 50.59 process.

1 TRAINING grams are credited to train plant personnel. Key technical operating personnel receive onsite sroom or guided self study and on-the-job training. Appropriate plant personnel receive ruction in emergency plan and radiation protection procedures. Specialized training in specific s conducted by the equipment manufacturers or other vendors is utilized as necessary.

ning on a continuing basis is used to maintain a high level of proficiency in the staff.

2 EMERGENCY PLAN staff approved Millstone Nuclear Power Station Emergency Plan (Reference 6.3-1) addresses criteria set forth in NUREG-0654, FEMA-REP-1, Criteria for Preparation and Evaluation of iological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, ision 1, November 1980 and NUREG-0737, Supplement 1. As such, the Emergency Plan vides for an acceptable state of emergency preparedness and meets the requirements of 10 R Part 50 and Appendix E thereto.

thheld under 10 CFR 2.390 (d) (1) 4 QUALITY ASSURANCE PROGRAM DESCRIPTION (QAPD) TOPICAL REPORT licensee has developed and implemented a comprehensive Quality Assurance Program P) to ensure conformance with established regulatory requirements as set forth by the lear Regulatory Commission, and accepted industry standards. The participants in the QAP re that the design, procurement, construction, testing, operation, maintenance, repair, and ommissioning of nuclear power plants are performed in a safe and effective manner.

QAPD Topical Report complies with the requirements set forth in Appendix B of 10 CFR 50, along with applicable sections of the Safety Analysis Report. The QAPD Topical Report corporated herein by reference.

mitted periodically to the NRC in accordance with 10 CFR 50.54(a).

5 REFERENCES 1 J. F. Opeka letter to U.S. Nuclear Regulatory 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].

2 J. F. Opeka letter to U.S. Nuclear Regulatory Commission, Millstone Nuclear Power Station, Unit Numbers 1, 2, and 3, Physical Security Plan, Revision 15, dated December 16, 1991 and subsequent revisions thereto.

tten procedures are required for maintenance, repair, or operational activities related to the ctures, systems and components which are safety related (Safety Class 1,2, or 3). Written cedures shall be established, implemented, and maintained in accordance with the Technical cifications.

rogram describing the review and audit of activities important to and affecting station safety, been established and complies with Regulatory Guide (RG) 1.33, Quality Assurance gram Requirements (Operation). The program provides a system to ensure that these vities are performed in accordance with company policy, rules, and approved procedures.

1 ONSITE REVIEW membership, duties, areas of review responsibility, and requirements of both the plant and operations review committees are described in the Quality Assurance Program Description PD) Topical Report (Reference 6.1-1).

2 INDEPENDENT REVIEW ependent review of activities affecting the unit's safety is performed by the Management ety Review Committee as described in the QAPD Topical Report (Reference 6.1-1).

3 AUDITS audit program for activities affecting safety related systems, structures, or components is as cribed in the QAPD Topical Report (Reference 6.1-1).

SUMMARY

OF ACTIVITIES lstone Unit Number 1 was shutdown for a normal refueling outage on November 4, 1995, and not operated since. On November 19, 1995, transfer of all fuel assemblies from the reactor sel into the spent fuel pool (SFP) for storage was completed. On July 17, 1998, the licensee ided to permanently cease further operation of the plant. Certification to the NRC of the manent cessation of operation and permanent removal of fuel from the reactor vessel, in ordance with 10 CFR 50.82 (a)(1)(i) & (ii), was filed on July 21, 1998 (Reference 7.1-1), at ch time the CFR 50 license no longer authorized operation of the reactor or placement of fuel in the reactor sel.

mission of the licensee is to decommission the plant safely and in a cost effective manner.

information contained in this section of the DSAR is based upon the best information ently available. The plans discussed herein may be modified as additional information omes available or conditions change.

cific conditions which are unique to the multi-unit Millstone Station require that certain lstone Unit Number 1 decommissioning activities be delayed and performed concurrently h the decommissioning of Millstone Unit Numbers 2 and 3. Other considerations may dictate y scheduling of certain decommissioning activities. Therefore, the approach to ommissioning Millstone Unit Number 1 can best be described as a modified SAFSTOR. In approach, decontamination and dismantlement activities may be undertaken early in the ommissioning wherever it makes sense from a safety or economic viewpoint. For instance, en the future uncertainty over access to a low level waste disposal site, early shipment of ain components will occur. The amount of decommissioning work completed prior to a FSTOR period depends upon a number of factors currently under evaluation.

h the DECON and the SAFSTOR options are approaches found acceptable to the NRC in its al Generic Environmental Impact Statement (GEIS) (Reference 7.1-2).

mpletion of the decommissioning schedule 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 decommissioning activities.

rently Millstone Unit Number 1 has access to Chem-Nuclear Systems Barnwell, S.C.

osal site and to the Envirocare disposal site in Tooele County, Utah. Escalation costs for the osal of waste have been incorporated into financial planning. Additionally, the licensee has sidered the possibility that during the decontamination and dismantlement phases, access to Barnwell low level waste disposal site could be denied or that the facility could be closed.

e resulted in a significant increase in the cost of decommissioning and, may require the use of ndependent spent fuel storage installation (ISFSI).

hough storage of the Millstone Unit Number 1 spent fuel in an ISFSI is presented in this AR as an option; an ISFSI has been contracted to ensure the continued operation of Millstone t Numbers. 2 and 3. Currently, after spent nuclear fuel is removed from the Unit 2 and Unit 3 tor core; it is safely stored in the existing SFPs. Capacity of these pools was designed with the mption the DOE high level waste repository would provide permanent storage. However, the selection, construction and licensing of such a repository have been delayed. As is the case h other nuclear facilities as the SFPs approach full capacity, spent fuel from Millstone Unit mbers 2 and 3 will be stored in the ISFSI. A description of the ISFSI is contained in the Unit mber 3 Final Safety Analysis Report.

er any eventuality such as unavailability of a LLW disposal site, temporary shortfall in ommissioning funding, or other unforeseen circumstances, 10 CFR 50.82 requires the licensee aintain the capability to suspend decontamination and dismantlement.

1 DECOMMISSIONING APPROACH licensee is planning on decommissioning Millstone Unit Number 1 using a modified FSTOR approach in which the decontamination and dismantlement of the systems, ponents, plant structures and facilities (i.e., DECON) are completed prior to and following a FSTOR period. In this plan, an ISFSI may be constructed and the transfer of spent fuel from spent fuel pool (SFP) could be completed during the SAFSTOR period. The SAFSTOR period s with decontamination and dismantlement of any remaining systems, structures, and ponents commence in coordination with Millstone Unit Number 2 and Millstone Unit mber 3 decommissioning.

nt fuel shipments from the ISFSI to DOE are scheduled, when practicable, following the ository commencing operations. Delays in the operation of the repository limits the transfer of and increases the cost of long term spent fuel storage.

following discussion provides an outline of the current decommissioning plan activities pleted to date and the remaining significant activities. The planning required for each ommissioning activity, including the selection of the process to perform the work, is pleted prior to the start of work for that activity.

1.1 Planning planning includes implementation of a site characterization plan, preparation of a detailed ommissioning plan, and the engineering development of task work packages. The detailed ineering required to support the decontamination and dismantlement of systems, structures, components are performed prior to the start of field activities.

nificant activities performed to date include:

Removal and disposal of legacy 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 segmentation, 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.

following activities remain:

Evaluate and choose a dry fuel storage system, if pursued. Investigate and prepare for the design and licensing of an ISFSI and prepare procurement specifications for a fuel canister system and ancillary equipment.

1.2 Site Characterization ing the initial portion of the planning period a detailed site characterization was undertaken ng which radiological, regulated and hazardous wastes were identified, categorized, and ntified. Surveys were conducted to establish the contamination and radiation levels throughout Millstone Unit Number 1 portion of the site. This information is used in developing cedures to ensure that hazardous, regulated or radiologically contaminated materials are oved and to ensure that worker exposure is maintained as low as reasonably achievable ARA). Selected surveys of the outdoor areas in the vicinity of Millstone Unit Number 1 may performed, although a detailed survey of the environs would likely be deferred pending ommissioning of Millstone Unit Numbers 2 and 3. It is worthwhile to note that site racterization is a process that continues throughout decommissioning. As decontamination and mantlement work proceed, surveys are conducted to maintain current characterization and that ommissioning activities are adjusted accordingly.

activation analysis of the reactor internals, the reactor vessel, and the biological shield wall undertaken as a part of the site characterization. Using the results of this analysis, these ponents were classified in accordance with 10 CFR 61 and form the basis for the detailed s for their packaging and disposal. The interior grid portion of the top guide structure was rmined to be greater than class C (GTCC) material, was segmented from the reactor vessel, is stored in the spent fuel pool in canisters sized to be compatible with ISFSI dry storage tainers.

1.3 Decontamination objectives of the decontamination effort are two fold. First, to reduce the radiation levels ughout the facility in order to minimize personnel exposure during dismantlement. Second, to n as much material as possible to unrestricted use levels, thereby permitting non radiological olition and minimizing the quantities of material that must be disposed of by burial as oactive waste.

ems may benefit from decontamination activities by reducing the radiation exposure to the kforce. Late dismantling may not require the components and systems to be decontaminated e the decay of the radiation sources reduces the radiation levels by significant amounts.

mical decontamination of the reactor recirculation system may provide value through reduced ker dose. An evaluation is performed to determine whether the expected reduction in the umulated workforce exposure is justified by the costs associated with the decontamination.

evaluation results are sensitive to the amount and type of work to be performed prior to a FSTOR period. Any decontamination method used employs established processes with well-erstood chemical interactions. The resulting waste is disposed of in accordance with plant cedures and applicable regulations.

second objective of the decontamination effort is achieved by decontaminating structural ponents including steel framing and concrete surfaces. The method used to accomplish this is hanical and requires the removal of the surface or surface coating. This process is used ularly in industrial and contaminated sites.

1.4 Major Decommissioning Activities defined in 10 CFR 50.2 a major decommissioning activity is any activity that results in manent removal of major radioactive components, permanently modify the structure of the tainment, or results in dismantling components for shipment containing GTCC waste in ordance with 10 CFR 61.55.

or decommissioning activities completed to date include the removal of the drywall head and oval of the reactor vessel internals by segmentation. The drywall head was sectioned and sent metal processor. The reactor vessel internals, classified as GTCC, are limited to the interior ion of the top guide structure, which has been segmented from the reactor vessel and is stored he spent fuel pool. The reactor cavity and reactor vessel have been drained. Without the GTCC rnals present, several options are available for later removal and disposal of the reactor vessel:

mentation, sectioning into pieces, or disposal as an intact package.

ed on an evaluation of activity levels, ease of execution, personnel exposure, schedule straints, disposal facility availability, and cost, segmentation of the internals may be postponed l after the fuel is removed from the SFP.

oval of the reactor vessel follows the removal of the reactor internals and may not occur until r a SAFSTOR period. It is likely that the vessel would be removed by sectioning or menting. Vessel sectioning or segmenting permits a substantial portion of the waste to be sent waste re-processor instead of a near surface disposal site. The dismantling of the drywell and pression chamber is undertaken as part of the reactor building demolition.

1.5 Other Decommissioning Activities er decommissioning activities include:

1. A license termination plan pursuant to 10 CFR 50.82

2. A spent fuel management program, pursuant to 10 CFR 50.54(bb) ddition to the major decommissioning activities listed above, the following decommissioning vities include:

Hazardous and regulated materials (e.g., asbestos, lead, mercury, PCBs, oil, chemicals) are identified during characterization and plans are developed for the removal of these materials.

Plant components removed from the Turbine Building include the Turbine Generator, Condenser, Feedwater Heaters, Moisture Separators and miscellaneous system and support equipment.

Miscellaneous solid waste removed 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 processed. When the spent fuel is removed, the SFP is drained and the water processed. Systems are then isolated and deactivated in a sequence compatible with the operations previously described. Spent fuel pool systems are isolated after removal of the spent fuel.

ioactively contaminated or activated materials are removed from the site as necessary to allow site to be released for unrestricted access. Low level waste is processed in accordance with t procedures and existing commercial options, and sent to licensed disposal facilities or waste cessors for further volume reduction. Wastes may be incinerated, compacted, or otherwise cessed by authorized and licensed contractors, as appropriate. Mixed wastes, if any, are aged according to all applicable federal and state regulations. Mixed wastes are transported y by authorized and licensed transporters and shipped only to authorized and licensed lities.

1.6 Final Site Survey and Termination of License ce Millstone Unit Number 1 and Millstone Unit Number 2 are contiguous and have common ctural boundaries, the plans for building demolition and for the license termination survey are lemented as a coordinated evolution for the two units. Consequently, the schedule for the lstone Unit Number 1 license termination is constrained by the need to terminate the Part 50 nse coincident with that of Millstone Unit Number 2. As a result of this delay in requesting

s surveyed in conjunction with completion of the Unit 2 decontamination and dismantlement.

licensee is required to prepare a License Termination Plan (LTP) for Millstone Unit Number The LTP defines the details of the final radiological survey to be performed once the ontamination activities are completed. The LTP conforms to the format defined in erence 7.1-5 and addresses the limits of 10 CFR 20 using the pathways analysis defined in erence 7.1-4. Use of this guidance ensures that survey design and implementation is ducted in a manner that provides a high degree of confidence that applicable NRC criteria are sfied. Once the survey is complete, the results are provided to the NRC in a format that can be fied.

1.7 Site Restoration restoration of the Millstone Unit Number 1 area of the Millstone site will be undertaken when 10 CFR Part 50 license for Millstone Unit Number 1 is terminated. This event may coincide h Millstone Unit Numbers 2 and 3 license terminations. Buildings, structures, and other lities which are not currently known to be radiologically contaminated, such as the Strainer Intake Structure, and the Discharge Structure are dismantled, as part of the building olition effort after the final license termination survey for Millstone Unit Number 1 is plete. These buildings can be removed late in the building demolition phase since there is no ommissioning operational need to remove them earlier. Site restoration requires that all dings be removed to an elevation 3 feet below grade or to an elevation consistent with the oval of the necessary amounts of contaminated material.

2 STORAGE OF RADIOACTIVE WASTE le 5.4-1 of the GEIS (Reference 7.1-2) provides an estimate for low-level waste disposal from ferenced boiling water reactor (BWR) of 18,975 cubic meters (669,817 cubic feet) for both the CON and SAFSTOR options. The licensee estimates the low-level waste burial volume for lstone Unit No. 1, will be at or below this value for the modified SAFSTOR alternative. The nsees estimate includes, by a reduction of approximately 40 percent (industry standard), the zation of present-day volume reduction techniques. For waste requiring deep geological al, i.e., GTCC waste, the licensee estimates that the volume for Millstone Unit Number 1 is at elow the 11.5 cubic meters anticipated for a reference BWR discussed in Section 5.4 of the S. These estimates support the conclusion that the previously issued environmental ements are bounding since the disposal of waste requires fewer resources, i.e., less waste osal facility area, than what was considered in the GEIS.

2.1 High Level Waste gress passed the Nuclear Waste Policy Act in 1982, assigning the responsibility for disposal pent nuclear fuel created by the commercial nuclear generating plants to DOE. This legislation created a Nuclear Waste Fund to cover the cost of the program, which is funded, in part, by sale of electricity from the Millstone Unit Number 1 plant. The current DOE estimate for tup of the federal waste management system is 2010. For planning purposes, the licensee has

ctly from the ISFSI.

spent fuel is currently stored in the SFP. The licensee may license a dry, ISFSI. Fuel will be sferred from the pool and stored temporarily on site using licensed canisters. For the period of e when the fuel will be stored in the SFP, the systems necessary for SFP operations will be solidated into an Island concept and configured for SFP clean-up and cooling.

2.2 Low Level Waste ioactively contaminated or activated materials are removed to allow the site to be released for estricted access. Low level waste is processed in accordance with federal and state regulations, t procedures and existing commercial options, and transported to license disposal facilities.

2.3 Waste Management ajor component of the total cost of decommissioning Millstone Unit Number 1 is the cost of kaging and disposing of systems, components and structures, contaminated soil, water and r plant process liquids. A waste management plan incorporates the most cost effective osal strategy consistent with regulatory requirements for each waste type. The waste agement plan will be based on the evaluation of available methods and strategies for cessing, packaging, and transporting radioactive waste in conjunction with the available osal facility options and associated waste acceptance criteria.

3 RADIATION EXPOSURE MONITORING sonnel radiation exposure is maintained ALARA and monitoring is conducted in accordance h the radiation protection program described in Chapter 4. Exposure specifically related to ommissioning activities is identified and tracked. Exposure monitoring is used to identify vities that are causing excessive exposure and to initiate corrective actions to reduce personnel osure.

4 REFERENCES 1 Letter B17388 from Bruce D. Kenyon 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.

2 U. S. Nuclear Regulatory Commission report NUREG-0586, Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities, dated August, 1988.

3 Letter B17790 from R. P. Necci to U. S. Nuclear Regulatory Commission, Post Shutdown Decommissioning Activities Report, dated June 14, 1999.

5 U. S. Nuclear Regulatory Commission report NUREG-1700, Standard Review Plan for Evaluating Nuclear Power Reactor License Termination Plans, (Currently in Draft form).

decommissioning of Millstone Unit No. 1 is accomplished with no significant adverse ironmental impacts, in that no Millstone Unit No. 1 site specific factors should alter the clusions of the GEIS (Reference 7.1-2) or the Millstone Environmental Statement. The ation dose to the public during decommissioning is typically minimal. Decommissioning kers receive a fraction of the dose which radiation workers receive in an operating plant. The

-level radioactive waste that is removed from the site occupies only a small portion of the al volume at approved waste disposal sites. The non-radiological environmental impacts are porary and not significant.

occupational dose exposure for decommissioning Millstone Unit No. 1 is less than described he GEIS because of two main reasons. First, the licensee initiated a zinc injection program for lstone Unit No. 1 in 1987 that significantly reduced the buildup of contaminated corrosion ducts during the remaining plant operation period. Second, with the plant shutdown since 5, natural decay of leading radionuclides have reduced overall plant general dose levels ificantly by the time decontamination and decommissioning activities occur.

activities identified in this chapter resemble the DECON option. Therefore, the modified FSTOR occupational and public dose exposure is compared to the DECON option dose in the IS. The occupational and public dose effects for a modified SAFSTOR alternative is bounded he DECON option. The exposure from decontamination and dismantlement activities and the osure during transportation of the low-level wastes is included in this dose estimate. NUREG-6 (Reference 7.1-2), Table 5.3-2, estimates a total occupational dose of 18.74 person-Sv (1874 on-rem) for the DECON alternative for the reference BWR plant. The values estimated by the nsee will be at or below this value.

1 NUCLEAR WORKER ailed estimates for external occupational radiation exposure that accumulate dose for ommissioning workers during the dismantlement program are developed based on a task by analysis of personnel hours and expected radiation dose rates associated with each task.

se estimates are based on the following:

1. ALARA principles are implemented.

2. Radiation exposure is monitored to identify jobs that are causing excessive exposure and corrective actions are taken to reduce the severity.

2 GENERAL PUBLIC iation dose to the public is maintained below comparable levels when the plant was operating ugh the continued application of radiation protection and contamination controls combined h the reduced source term available in the facility.

pments of spent fuel and radioactive wastes are performed by exclusive use vehicles.

pments will be in accordance with the Department of Transportation (DOT) regulations.

eric industry estimates of the doses from routing transportation of radioactive materials are ed on the following assumptions:

Two truck drivers during a 500 mile trip would probably spend no more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> inside the cab and 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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 exposed to radiation when a truck stops for fuel or for the drivers to eat. The onlooker dose is calculated on the basis that 10 people spend an average of 3 minutes each at a distance 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.

REG/CR-0672, Table 11.4-2, provided a generic estimate of the routing radiation doses from k transportation of radioactive wastes. The doses are based on the maximum allowable dose s for each shipment in exclusive use trucks and are conservatively high, on the number of k shipments, and on the shipping distances. The estimated external radiation dose for routing sportation operations is 110 man-rem to transportation workers and 10 man-rem to the general lic.

licensee estimates the volume of both high level and low level wastes to be less than the umes used in NUREG/CR-0672. The total number of shipments of radioactive wastes is less those used to determine the exposure in the NUREG/CR, and therefore the exposure to the sportation workers and the general public is less than those identified above.

ing the decommissioning, processes may concentrate source terms. Non-routine events may ur with the potential to disperse the source term. This section of the DSAR establishes controls requirements to maintain potential consequences of such event to below analyzed accidents.

1 IN PLANT EVENTS DBA for Millstone Unit Number 1 is the fuel handling accident and a detailed discussion can ound in DSAR Chapter 5. The acceptance criteria for all other potential events at the plant are trolled such that the potential consequences of any postulated event are less than 1 REM at the lusion area.

2 TRANSPORTATION ACCIDENTS nsportation accidents have a wide range of severities. Most accidents occur at low speeds and e relatively minor consequences. In general, as speed increase, accident severity also eases. However, accident severity is not a function of vehicle speed only. Other factors, such he type of accident, the equipment involved, and the location can have an important bearing on dent severity.

mage to a package in a transportation accident is not directly related to accident severity. In a es of accidents of the same severity, or in a single accident involving a number of packages, age to packages may vary from none to extensive. In relatively minor accidents, serious age to packages can occur from impacts on sharp objects or from being struck by other cargo.

versely, even in very severe accidents, damage to packages may be minimal.

probabilities of truck accidents used in the NUREG/CR-0672 study were based on accident supplied by the DOT. Accidents are classified into five categories as functions of speed and duration. The five categories in order of increasing severity are: minor, moderate, severe, a severe, and extreme. Table N.5-3 of NUREG/CR-0672 provides the probabilities of urrence for each classification.

mated accident frequencies, release amounts and radiation doses to the maximum exposed viduals for selected accidents for transportation of radioactive material are discussed in endix N.5.2.3 of NUREG/CR-0672. The frequencies are calculated by multiplying the total ance of transport with the total probability of accident per distance traveled for each accident erity class.

maximum exposed individual is assumed to be located 100 meters from the point of a sportation accident. The calculated dose values provided in Table N.5.6 of NUREG/CR-0672 the first year dose and the fifty year dose commitment to the bone, lung, thyroid and whole y.

REG/CR-0672.

non-radiological environmental impacts from the Millstone Unit Number 1 decommissioning temporary and not significant. The largest occupational risk associated with the ommissioning is the risk of industrial accidents. This risk is minimized by adherence to work trols during decommissioning similar to the procedures followed during power operation.

cedures controlling work related to asbestos, lead, and other non-radiological hazards remain lace during the decommissioning. The primary environmental effects of the decommissioning temporary and include small increases in noise levels and dust in the immediate vicinity of the

, and small increases in truck traffic to and from the site for hauling equipment and waste.

se effects are similar to those experienced during normal refueling outages and certainly less ere than those present during the original plant construction. No significant socioeconomic acts or impacts to local culture, terrestrial or aquatic resources have been identified.

1 ADDITIONAL CONSIDERATIONS ile not quantitative, the following considerations are also relevant to concluding that ommissioning activities do not result in significant environmental impacts not previously ewed:

The release of effluents continues to be controlled by plant license requirements and plant operating procedures throughout the decommissioning.

With respect to radiological releases, Millstone Unit No. 1 continues to operate in accordance with the Offsite Dose Calculation Manual during decommissioning.

Release of non-radiological effluents 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, clothing, and breathing apparatus are used as appropriate.

Sufficient decontamination and source term reduction prior to dismantlement are performed to ensure that occupational dose and public exposure do not exceed those estimated in the Final Generic Environmental Impact Statement (Reference 7.1-2.

Detailed site radiological surveys are performed prior to starting the waste campaigns to confirm the burial volume of low-level radioactive waste and highly activated components which require deep geological disposal.

Transport of radioactive waste is in accordance with plant procedure, applicable Federal regulations, and the requirements of the receiving facility.

Site access control during decommissioning ensures that residual contamination is minimized or eliminated as a radiation release pathway to the public.