ML052560278

License Termination Plan, Revision 3

ML052560278
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	09/06/2005
From:	Connecticut Yankee Atomic Power Co
To:	NRC/FSME
References
Download: ML052560278 (68)

Text

t I . w- I 4 Connecticut Yankee Atomic Power Company K.> Date of Distribution: 09-06-05 Notice of Receipt of License Termination Plan (LTP)

Change No.: 05-01 To:

NRC Headquarters Office Copy No.: Washington DC Copy # UNCONTROLLED Please revise your controlled copy per instructions below:

INSERT: Revision 3 of the Haddam Neck Plant (HNP) License Termination Plan (LTP), dated August 2005, according to the directions provided.

ATTACH:

REMOVE:

REPLACE This acknowledges receipt of the revisions listed above. In addition, all superseded pages have been removed and destroyed.

Signature: Date:

Please Return This Sheet to the Administrative Office, Connecticut Yankee Within Thirty (30) Days.

gtMD I

Haddan Neck Plant License Termination'Plan Distribution of Revision 3, August 2005 Insert Instruction for Revision 3 of the HNP LTP Memo RACY-05-056 Remove Page (s) Number Insert Page (S) Number Front Matter- Cover Sheet Rev 2 Front Matter- Cover Sheet Rev 3 Table of content Table of content ii through xiii ii through xiii List of Effective Pages List of Effective Pages LEP- I thru 4 LEP-1 thru 4 After Chapter 1 Tab After Chapter 1 Tab Page 1-6 Pages 1-6 None Page 1-6a Pages 1-7 thru 1-8 Pages 1-7 thru 1-8 After Chapter 2 Tab After Chapter 2 Tab Pages 2-2 Pages 2-2 After Chapter 5 Tab After Chapter 5 Tab Page 5-2 Page 5-2 None Page 5-2a Page 5-7 Page 5-7 Pages 5-13 thru 5-18 Pages 5-13 thru 5-18 Pages 5-23 thru 5-26 Pages 5-23 thru 5-26 None Page 5-26a None After Page 5-41 Insert Page 5-41 a Pages 5-48 thru 5-49 Pages 5-48 thru 5-49 Pages 5-51 thru 5-53 Pages 5-51 thru 5-53 None Page 5-53a Pages 5-59 thru 5-60 Pages 5-59 thru 5-60 After Chapter 6 Tab After Chapter 6 Tab Pages 6-1 thru 6-22 Pages 6-1 thru 6-22 Figure 6-6 None

17

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HADDAM NECK PLANT LICENSE TERMINATION PLAN THROUGH REVISION 3 AUGUST 2005

Haddam Neck Plant License Termination Plan (L; ' *s:-:

3.3 Completed and Ongoing Decommissioning Activities and Tasks 3-3 3.3.1 Overview 3-3 3.3.2 RCS Chemical Decontamination 3-4 3.3.3 Turbine Rotors 3-6 3.3.4 Removal of Spare Auxiliary Transformer 3-6 3.3.5 Removal of Steam Generator Steam Domes 3-6 3.3.6 Removal of Steam Generator Lower Assemblies 3-6 3.3.7 Removal of the Pressurizer 3-6 3.3.8 Removal of the Reactor Coolant Pumps (RCPs) 3-7 3.3.9 Dismantlement of Buildings 3-7 3.3.10 Removal and Disposal of the RPV 3-7 3.3.11 Spent Fuel and GTCC Wastes 3-8 3.3.12 Additional Activities 3-8 3.4 Future Decommissioning Activities and Tasks 3-8 3.4.1 Overview 3-8 3.4.1.1 Detailed Planning and Engineering Activities 3-9 3.4.1.2 General Decontamination and Dismantlement Considerations 3-10 3.4.1.3 Decontamination Methods 3-11 3.4.1.4 Contaminated System Dismantlement 3-12 3.4.1.5 Removal Sequence and Material Handling 3-12 3.4.1.6 System Isolation/De-energization 3-13 3.4.1.7 Temporary Systems Required to Support Decommissioning 3-13 3.4.1.8 Specific Decommissioning and Dismantlement Activities 3-14 3.4.1.9 Decontamination and Disposition of Site Buildings 3-15 3.4.2 General Description of and Remediation Consideration for Remaining Systems, 3-15 Structures, and Components as of May 2004 3.4.2.1 Chemical and Volume Control System (CVCS) 3-16 3.4.2.2 Component Cooling Water (CCW) System 3-16 3.4.2.3 Service Water (SW) System 3-16 3.4.2.4 Spent Fuel Pool and Fuel Handling Equipment 3-16 3.4.2.5 Spent Fuel Pool Purification 3-17 3.4.2.6 Spent Fuel Pool Transfer Tube 3-17 3.4.2.7 Makeup Water (MW) 3-17 3.4.2.8 Main Steam and Feedwater (MS &FW) Systems 3-17 3.4.2.9 Reactor Coolant System (RCS) 3-17 3.4.2.10 Residual Heat Removal (RHR) System 3-18 3.4.2.11 Safety Injection (SI) System 3-18 3.4.2.12 Service Air System 3-18 3.4.2.13 Control Air System 3-18 3.4.2.14 Primary Water (PW) System 3-18 3.4.2.15 Primary Ventilation System/ Fuel Building Ventilation System 3-19 3.4.2.16 Liquid Waste System 3-19 3.4.2.17 Gaseous Waste System 3-19 3.4.2.18 Turbine Building Waste Water Treatment System 3-19 3.4.2.19 Well Water and Water Treatment System 3-20 3.4.2.20 Circulating Water and Vacuum Priming Systems 3-20 3.4.2.21 Closed Cooling System 3-20 3.4.2.22 Turbine Lube Oil System 3-20 3.4.2.23 Boron Recovery'System 3-20 3.4.2.24 Containment Systems and Miscellaneous Systems 3-20 3.4.2.25 Site Electrical Distribution 3-20 August 2005 ii Rev. 3

Haddam Neck Plant License Termination Plan 3.4.2.26 Fire Protection System 3-21 3.4.2.27 Heating Steam and Condensate System 3-21 3.4.2.28 Floor, Roof and Equipment Drains 3-21 3.4.2.29 Buildings 3-21 3.5 Radiological Impacts of Decommissioning Activities 3-23 3.5.1 Occupational Exposure 3-24 3.5.2 Radioactive Waste Projections 3-26 3.6 References 3-27 4 SITE REMEDIATION PLANS 4-1 4.1 Introduction 4-1 4.2 Remediation Levels and ALARA Evaluations 4-1 4.2.1 Generic ALARA Screening Levels 4-2 4.2.2 Survey-Unit Specific ALARA Evaluation 4-2 4.2.3 Groundwater ALARA Evaluations 4-3 4.3 Remediation Actions 4-3 4.3.1 Structures 4-3 4.3.2 Soils 44 4.3.3 Nonstructural Systems 4-4 4.4 References 4-5 5 FINAL STATUS SURVEY PLAN 5-1 5.1 Introduction 5-1 5.2 Scope 5-1 5.3 Summary of the Final Status Survey Process 5-1 5.4 Survey Planning 5-4 5.4.1 Data Quality Objectives 5-4 5.4.2 Classification of Survey Areas and Units 5-7 5.4.3 Survey Units 5-7 5.4.4 Reference Coordinate Systems 5-8 5.4.5 Reference Areas and Materials 5-9 5.4.6 Area Preparation: Isolation and Control 5-10 5.4.6.1 Structures 5-10 5.4.6.2 Open Land Areas 5-12 5.4.6.3 Excavation Land Areas Resulting from Radiological Remediation 5-12 5.4.6.4 Bedrock 5-12 5.4.6.5 Excavations Resulting from the Removal of Piping Conduit 5-12 5.4.7 Selection of DCGLs 5-13 5.4.7.1 Operational DCGLs 5-13 5.4.7.2 Gross Activity DCGLs 5-18 5.4.7.3 Surrogate Ratio DCGLs 5-18 5.4.7.4 Elevated Measurement Comparison (EMC) DCGLs 5-20 5.4.7.5 Building Basements and Footings 5-23 5.4.7.6 Release Limits for Non-Structural Components and Systems 5-23 5.5 Final Status Survey Design Elements - Surface Soils and Structures 5-25 5.5.1 Selecting the Number of Fixed Measurements and Locations 5-27 5.5.1.1 Establishing Acceptable Decision Error Rates 5-27 5.5.1.2 Determining the Relative Shift 5-28 iii Rev.3 II 2005 August 2005 ...

Rev. 3 1

Haddam Neck Plant License Termination Plan 5.5.1.3 Selecting the Required Number of Measurements for the WRS Test 5-29 5.5.1.4 Selecting the Required Number of Measurements for the Sign Test 5-29 5.5.1.5 Assessing the Need for Additional Measurements in Class 1 Survey Units 5-30 5.5.1.6 Determining Measurement Locations 5-34 5.5.2 Judgmental Assessments 5-35 5.5.3 Data Investigations 5-35 5.5.3.1 Investigation Levels 5-35 5.5.3.2 Investigations 5-36 5.5.3.3 Remediation 5-37 5.5.3.4 Re-classification 5-37 5.5.3.5 Re-survey 5-37 5.6 Survey Protocol for Non-structural Systems and Components 5-38 5.7 Survey Implementation and Data Collection 5-39 5.7.1 Survey Methods 5-39 5.7.1.1 Scanning 5-39 5.7.1.2 Fixed Measurements 5-40 5.7.1.3 Advanced Technologies 5-40 5.7.1.4 Other Advanced Survey Technologies 5-41 5.7.1.5 Samples 5-41 5.7.1.6 Contaminated Concrete Basements 5-41a 5.7.2 Survey Instrumentation 5-42 5.7.2.1 Survey Instrument Data Quality Objectives 5-42 5.7.2.2 Instrument Selection 5-42 5.7.2.3 Calibration and Maintenance 5-43 5.7.2.4 Response Checks 5-43 5.7.2.5 MDC Calculations 5-44 5.7.2.6 Typical Instrumentation and MDCs 5-48 5.7.3 Survey Considerations 5-50 5.7.3.1 Survey Considerations for Buildings, Structures and Equipment 5-50 5.7.3.2 Survey Considerations for Outdoor Areas 5-54 5.7.3.3 Surveillance following Final status Surveys , 5-57 5.8 Survey Data Assessment 5-59 5.8.1 Wilcoxon Rank Sum Test 5-60 5.8.2 Sign Test 5-61 5.8.3 Elevated Measurement Comparison 5-62 5.8.4 Unity Rule 5-63 5.8.5 Data Assessment Conclusions 5-63 5.9 Final Status Survey Reports 5-64 5.9.1 FSS Survey Unit Release Records 5-64 5.9.2 FSS Final Reports 5-65 5.10 Quality Assurance and Quality Control Measures 5-66 5.11 References . 5-69 6 COMPLIANCE WITH THE RADIOLOGICAL CRITERIA FOR LICENSE 6-1 TERMINATION 6.1 Site Release Criteria 6-1 6.1.1 Radiological Criteria for Unrestricted Use 6-1 6.1.2 Conditions Satisfying the Site Release Criteria 6-1 August 2005 iv Rev.3

Haddam Neck Plant License Termination Plan 6.2 Site Characteristics 6-2 6.3 Dose Modeling Approach 6-4 6.3.1 Overview 64 6.3.2 Resident Farmer Scenario 64 6.3.3 Building Occupancy Scenario 6-5 6.4 RESidual RADioactivity (RESRAD) and RESRAD-Build Codes 6-6 6.5 Parameter Selection Process 6-7 6.5.1 Classification 6-7 6.5.2 Prioritization 6-7 6.5.3 Treatment 6-8 6.5.4 Sensitivity Analyses 6-8 6.5.5 Parameter Value Assignment 6-8 6.6 DCGLs for Soil 6-9 6.6.1 Dose Model 6-9 6.6.2 Conceptual Model 6-10 6.6.3 Sensitivity Analysis Results 6-10 6.6.4 DCGL Determination 6-10 6.7 DCGLs for Groundwater 6-11 6.7.1 Dose Model 6-11 6.7.2 Conceptual Model 6-12 6.7.3 Sensitivity Analysis Results 6-13 6.7.4 DCGL Determination 6-13 6.8 DCGLs for Concrete 6-15 6.8.1 DCGLs for Concrete: Buildings Standing 6-15 6.8.1.1 Dose Model 6-15 6.8.1.2 Conceptual Model 6-15 6.8.1.3 Sensitivity Analysis Results 6-16 6.8.1.4 DCGL Determination 6-16 6.8.2 . Future Groundwater Dose Subsurface Structures and Basement /Footings:

Basement Fill Model 6-16 6.8.2.1 General Dose Calculation Model 6-17 6.8.2.2 Future Groundwater Dose Calculation for Basement Concrete 6-18 6.8.2.3 Future Groundwater Dose Calculation for Surface Contamination on the Containment Liner 6-20 6.8.2.4 Future Groundwater Dose Calculation for Surface Contamination on Embedded Piping 6-20 6.8.2.5 Summary of all Future Groundwater Dose Calculations 6-21 6.9 Operational DCGLs 6-21 6.10 References 6-21 7 UPDATE OF SITE-SPECIFIC DECOMMISSIONING COSTS 7-1 7.1 Introduction 7-1 August 2005 v Rev.3

Haddam Neck Plant License Termination Plan 7.2 Decommissioning Cost Estimate 7-2 7.2.1 Cost Estimate Previously Docketed in Accordance with 10 CFR 50.82 and 10 CFR 50.75 Post Shutdown 7-2 7.2.2 Summary of the Site Specific Decommissioning Cost Estimate 7-2 7.2.3. Dismantlement and Decontamination 7-6 7.2.4 Radiological Waste Disposal 7-6 7.2.5 Long-Term Spent Fuel Storage 7-6 7.2.6 Site Restoration and License Termination 7-6 7.3 Decommissioning Funding 7-6 7.4 References. 7-7 8 SUPPLEMENT TO THE ENVIRONMENTAL REPORT 8-1 8.1 Introduction 8-1 8.1.1 Overview 8-1 8.1.2 Proposed Site Conditions at the Time of License Termination 8-4 8.1.3 Remaining Dismantlement and Decommissioning Activities 8-4 8.2 Analysis of Site-Specific Issues - 8-7 8.2.1 Onsite-Offsite Land Use 8-7 8.2.2 Water Use . 8-7 8.2.3 Water Quality 8-8 8.2.4 Air Quality 8-9 8.2.5 Aquatic Ecology 8-10 8.2.6 Terrestrial Ecology 8-11 8.2.7 Threatened and Endangered Species 8-11 8.2.8 Radiological 8-12 8.2.9 Radiological Accidents 8-13 8.2.10 Occupational Issues 8-14 8.2.11 Socioeconomic Impacts . 8-14 8.2.12 Environmental Justice 8-15 8.2.13 Cultural and Historic Resources Impact 8-15 8.2.14 Aesthetic . 8-16 8.2.15 Noise 8-16 8.2.16 Transportation 8-17 8.2.17 Irretrievable 8-17 8.3 References 8-18 APPENDICES Appendix A, Acronym List vi Rev. 3 2005 August 2005 vi Rev. 3

Haddam Neck Plant License Termination Plan Appendix B, ALARA Evaluations Appendix C, Financial Materials Appendix D, Input to Sensitivity Analysis (using RESRAD Version 6.1 and RESRAD-Build Version 3.1)

Appendix E, Results of the Sensitivity Analyses (using RESRAD Version 6.1 and RESRAD-Build Version 3.1)

Appendix F, Input to Calculate DCGLs (using RESRAD Version 5.91 and RESRAD-Build Version 2.37)

Appendix G, Calculation of DCGLs (using RESRAD Version 5.91 and RESRAD-Build Version 2.37)

Appendix H Table 2-10, MARSSIM Classifications (Updated as of November 2001) vii..i Rev.3 I August 2005 Rev. 3 l

Haddam Neck Plant License Termination Plan

, LIST OF TABLES 1 GENERAL INFORMATION No tables 2 SITE CHARACTERIZATION Table 2-1, Typical Instruments Used at HNP 24 Table 2-2, Examples of Unplanned Gaseous Release Events 2-8 Table 2-3, Examples of Unplanned Liquid Release Events 2-11 Table 2-4, Summary of Unrestricted Release Confirmatory Survey Program 2-13 Table 2-5, Radiological Status of HNP Systems 2-17 Table 2-6, Monitoring Well Soil Sample Data 2-30 Table 2-7, Well Water Results 2-31 Table 2-8, Groundwater Level Elevations 2-35 Table 2-9, Temporal Trends in Groundwater Radionuclide Concentrations 2-37 Table 2-10, MARSSIM Classifications (Updated as of May 2004) 2-45 Table 2-1 IA, Nominal Radiological Data Supporting Classifications for Structures 2-60 Table 2-1 IB, Nominal Radiological Data Supporting Classifications for Land Areas 2-85 Table 2-1 IC, Nominal Radiological Data Supporting Classifications for Subsurface Areas 2-90 Table 2-12, Radionuclides Potentially Present at HNP 2-92 Table 2-13, Summary of Radionuclide Analysis 2-94 3 IDENTIFICATION OF REMAINING SITE DISMANTLEMENT ACTIVITIES Table 3-1, Status of Major HNP Systems, Structures, and Components as of May 2004 2 Table 3-2, Activity Removed During the HNP Reactor Coolant System Chemical Decontamination 3-5 Table 3-3, Radiation Exposure Projections for Decommissioning and*Fuel Storage Activities 3-25 Table 3-4, Projected Waste Quantities 3-26 4 SITE REMEDIATION PLANS No tables.

5 FINAL STATUS SURVEY PLAN Table 5-1, HNP Survey Unit Surface Area Limits 5-8 Table 5-2, Typical Media-Specific Backgrounds 5-9 Table 5-3, Survey Areas Affected by Groundwater Contamination 5-16 Table 5-4, Operational DCGL. Example for Cs-137 5-17 Table 5-5, Area Factors for the Resident Farmer Scenario 5-21 Table 5-6, Area Factors for the Building Occupancy Scenario 5-22 Table 5-7, Release Limits for Buried Piping,dpm/I00 cm2 5-24 Table 5-8, Investigation Levels 5-36 Table 5-9, Traditional Scanning Coverage Requirements 5-40 Table 5-10, Volumetric Concrete Sample Requirements 5-41a Table 5-11, Available Instruments and Associated MDCs 549 August 2005 viii Rev.3

Haddam Neck Plant License Termination Plan BeT i Table 5-12, Initial Evaluation of Survey Results (Background Reference Area Used) 5-60 Table 5-13, Initial Evaluation of Survey Results (Background Reference Area Not Used) 5-60 6 COMPLIANCE WITH THE RADIOLOGICAL CRITERIA FOR LICENSE TERMINATION Table 6-1, Base Case DCGLs for Soil 6-10 Table 6-2, Base Case DCGLs for Groundwater 6-14 Table 6-3, Parameter Values for Equations 6-4 and 6-5 6-18 Table 6-4, Concrete Diffusion Coefficients Used in the Basements Fill Model 6-19 7 UPDATE OF SITE-SPECIFIC DECOMMISSIONING COSTS Table 7-1, Actual and Projected Decommissioning Expenditures 74 Table 7-2, Decommissioning/Spent Fuel Trust Analyses 7-5 8 SUPPLEMENT TO THE ENVIRONMENTAL REPORT Table 8-1, Summary of Environmental Impacts from Decommissioning 8-20 Table 8-2, Population Changes in the Vicinity of HNP 8-21 APPENDIX A, ACRONYM LIST No tables.

APPENDIX B, ALARA EVALUATIONS No tables APPENDIX C, FINANCIAL MATERIALS No tables.

APPENDIX D: INPUT TO SENSITIVITY ANALYSIS (USING RESRAD VERSION 6.1 AND RESRAD-BUILD VERSION 3.1)

Table D-1 Input Parameters for Sensitivity Analysis for Soil D-2 Table D-2 Input Parameters for Sensitivity Analysis for Groundwater D-12 Table D-3 Input Parameters for Sensitivity Analysis for Concrete: Buildings D-20 Standing Table D-4 Input Parameters for Sensitivity Analysis for Concrete: Buildings D-23 Demolished August 2005 ix 3 IRev.

Haddam Neck Plant License Termination Plan APPENDIX E: RESULTS OF THE SENSITIVITY ANALYSES (USING RESRAD VERSION 6.1 K>J AND RESRAD-BUILD VERSION 3.1)

Table E-1 Results of Sensitivity Analysis and Assignment of Conservative Values for Soil E-2 Table E-2 Results of Sensitivity Analysis and Assignment of Conservative Values for Groundwater E-5 Table E-3 Results of Sensitivity Analysis and Assignment of Conservative Values for Concrete: Buildings Standing E-8 Table E-4 Results of Sensitivity Analysis and Assignment of Conservative Values for Concrete: Buildings Demolished E-10 APPENDIX F: INPUT TO CALCULATE DCGLS (USING RESRAD VERSION 5.91 AND RESRAD-BUILD VERSION 2.37)

Table F-1 Input Parameters for Soil DCGLs F-2 Table F-2 Input Parameters for Groundwater DCGLs F-12 Table F-3 Input Parameters for DCGLs for Concrete: Buildings Standing F-20 Table F-4 Input Parameters for DCGLs for Concrete: Buildings Demolished F-25 APPENDIX G: CALCULATION OF DCGLS (USING RESRAD VERSION 5.91 AND RESRAD-BUILD VERSION 2.37)

Table G-1 DCGLs for Soil G-2 Q) Table G-2-1 Determination of Peak Dose Considering Dose Contributions from Progency G-3 Table G-2-2 DCGLs for Groundwater .G-6 Table G-3 DCGLs for Concrete: Buildings Standing G-8 Table G4-1 DCGLs for Concrete: Buildings Demolished G-9 Table G-4-2 Concentration of Residual Radioactive Material in the Contaminated Zone (pCilg) and the Well Water (RESRAD Groundwater) (pCi/l) and the Equilibrium Groundwater Contamination (pCi/g) G-10 APPENDIX H: TABLE 2-10, MARSSIM CLASSIFICATIONS (UPDATE as of NOVEMBER 2001)

Table 2-10, MARSSIM Classifications H-2 August 2005 x Rev. 3

i:

Haddam Neck Plant License Termination Plan LIST OF FIGURES (All Figures Are Located at the End of the Associated Section) 1 GENERAL INFORMATION No figures.

2 SITE CHARACTERIZATION Figure 2-1, Site Grounds Figure 2-2, Site Structures.-.Soils/Foundations Figure 2-3, Remaining Site Structure Figure 2-4, Open Land Areas Figure 2-5, Fuel Building, Elevation 13'-6" Figure 2-6, Containment Building, All Elevations Figure 2-7, Containment Building, Ground Floor Elevation 1'-6" Figure 2-8, Screenwell Building, Elevations 8'-O" Figure 2-9, Subsurface Areas 3 IDENTIFICATION OF REMAINING SITE DISMANTLEMENT ACTIVITIES No figures.

4 SITE REMEDIATION PLANS Figure 4-1, Survey Unit ALARA Evaluation Process 5 FINAL STATUS SURVEY PLAN Figure 5-1, Site Grounds Grid Map Figure 5-2, Site Area Grid Map Figure 5-3, Groundwater Plume Influence Boundary Figure 5-4, Final Status Survey Organization 6 COMPLIANCE WITH THE RADIOLOGICAL CRITERIA FOR LICENSE TERMINATION Figure 6-1, Site Layout Figure 6-2, Industrial and Peninsula Area Cross Section Figure 6-3, Exposure Pathways Considered in the Resident Farmer Scenario Figure 6-4, Exposure Pathways Considered in the Building Occupancy Scenario Figure 6-5, Parameter Selection Process 7 UPDATE OF SITE-SPECIFIC DECOMMISSIONING COSTS No figures.

August 2005 xi Rev. 3

Haddam Neck Plant License Termination Plan 8 SUPPLEMENT TO THE ENVIRONMENTAL REPORT No figures.

August 2005 ..i Rev. 3 l

Haddam Neck Plant License Termination Plan This page intentionally left blank.

xiii Rev. 3 August 2005 xiii Rev. 3

Haddam Neck Plant License Termination Plan LIST OF EFFECTIVE PAGES Front Matter Page Revision Date i through ii 2 August 2004 iii through xiii 3 August 2005 LEP-1 through LEP-4 3 August 2005 1 GENERAL INFORMATION Page Revision Date 1-1 through 1-5 2 August 2004 1-6 through 1-7 . 3 August 2005 1-8 through 1-12 2 August 2004 2 SITE CHARACTERIZATION Page Revision Date 2-1 1 August 2002 2-2 3 August 2005 2-3 2 August 2004 24 through 2-5 1 August 2002 2-6 0 July 2000 2-7 through 2-8 1 August 2002 2-9 through 2-10 2 August 2004 2-llthrough2-12 1 August 2002 2-13 2 August 2004 2-14 1 August 2002 2-15 through 2-16 2 August 2004 2-17 through 2-18 1 August 2002 2-19 through 2-98 2 August 2004 3 IDENTIFICATION OF REMAINING SITE DISMANTLEMENT ACTIVITIES Page Revision Date 3-1 2A January 2005 3-2 through 3-28 2 August 2004 August 2005 LEP-1 Rev. 3 l

Haddam Neck Plant License Termination Plan LIST OF EFFECTIVE PAGES 4 SITE RENIEDIATION PLANS Page Revision Date 4-1 2 August 2004 4-2 1 August 2002 4-3 through 44 2 August 2004 4-5 through 4-6 1 August 2002 5 FINAL STATUS SURVEY PLAN Page Revision Date 5-1 1 August 2002 5-2 through 5-2a 3 August 2005 5-3 through 5-6 2 August 2004 5-7 3 August 2005 5-8 through 5-12 2 August 2004 5-13 through 5-18 3 August 2005 5-19 through 5-22 2 August 2004 5-23 through 5-26a 3 August 2005 5-27 through 541 2 August 2004 541a 3 August 2005 542 through 548 2 August 2004 5-49 3 August 2005 5-50 2 August 2004 5-51 through 5-53a 3 August 2005 5-54 through 5-58 2 August 2004 5-59 through 5-60 3 August 2005 5-61 through 5-65 2 August 2004 5-66 2A January 2005 5-67 through 5-69 2 August 2004 5-70 2A January 2005 6 COMPLIANCE WITH THE RADIOLOGICAL CRITERIA FOR LICENSE TERMINATION Page Revision Date 6-1 through 6-22 3 August 2005 August 2005 LEP-2 Rev. 3

Haddam Neck Plant License Termination Plan LIST OF EFFECTIVE PAGES 7 UPDATE OF SITE-SPECIFIC DECOMMISSIONING COSTS Page Revision Date 7-1 thru 7-7 2 August 2004 8 SUPPLEMENT TO THE ENVIRONMENTAL REPORT Page Revision Date 8-1 thru 8-22 2 August 2004 Figures Figures 2-1 thru 2-9 2 August 2004 Figure 4-1 2 August 2004 Figures 5-1 and 5-2 0 July 2000 Figure 5-3 and Figure 5.3-1 2A January 2005

- - ~Figure 5-4 2 August 2004 Figure 6-1 0 July 2000 Figure 6-2 0 July 2000 Figures 6-3 thru 6-5 1 August 2002 Appendix A Page Revision Date A-1 thru A-2 0 July 2000 Appendix B Page Revision Date B-1 thru B-4 1 August 2002 Page Revision Date C-I thru C-36 0 July 2000 August 2005 LEP-3 Rev. 3 I

Haddam Neck Plant License Termination Plan LIST OF EFFECTIVE PAGES Appendix D Page* Revision Date D-1 thru D-31 I August 2002 Appendix E .

Page Revision Date E-1 thru E-12 I August 2002 Appendix F Page Revision Date FI thru F-32 1 August 2002 Appendix G Page Revision Date G&thru G-12 *1 August 2002 Appendix H Page Revision Date H-1 thru H 2 August 2004 LEP-44 Rev.3 I August 2005 LEP Rev. 3

A. .4 -

Haddam Neck Plant License Termination Plan 1.3.5 Final Status Survey Plan The primary objectives of the final status survey are to:

• select/verify survey unit classification,

• demonstrate that the level of residual radioactivity for each survey unit is below the release criterion, and

• demonstrate that the potential dose from small areas of elevated activity is below the release criterion for each survey unit.

The purpose of the Final Status Survey Plan is to describe the methods to be used in planning,.designing, conducting, and evaluating final status surveys at the HNP site to'demonstrate that the site meets the NRC's radiological criteria for unrestricted use. Chapter 5 of the LTP describes the Final Status Survey plan, which is consistent with the guidelines of MARSSIM. The HNP survey plan allows for the use of advanced technologies as long as the survey quality is equal to or better than traditional methods described in MARSSIM. Since MARSSIM is not readily applicable to complex nonstructural components within buildings, the current "no detectable" criteria will be applied to nonstructural components and systems at time of FSS (with the exception of those items discussed in Section 5.4.7.5).

The plan also describes methods and techniques used to implement isolation controls to prevent contaminating remediated areas (as discussed in additional detail in Section 5.4.6). The HNP Final Status Survey Plan incorporates measures to ensure that final survey activities are planned and communicated to regulatory agencies to allow the scheduling of inspection activities by these agencies if so desired.

K 1.3.6 Compliance wvith the Radiological Criteria for License Termination Chapter 6 together with Chapter 5, Final Status Survey Plan, describes the process to demonstrate compliance with the radiological criteria of 10CFR20.1402 (Reference 1-16) for unrestricted use for the HNP site. CYAPCO has selected the RESRAD computer code (Version 5.91) to model dose from soils, and ground water, and its counterpart, RESRAD-BUILD (Version 2.37), to model dose from structures.

For building basements to remain after unrestricted release of the site, the Basement Fill Model is used to calculate the future groundwater dose. The future groundwater dose is that which results from the leaching of radionuclides from buried concrete, the containment liner and embedded piping that is contained in basements to remain. This model is discussed in detail in Section 6.8.2. The characterization sampling to be performed to supply the input to the calculation of future groundwater dose using the Basement Fill Model is discussed in Chapter 5 (or with the discharge tunnel if they are assessed after the completion of containment liner FSS).

For building footings, an alternate criteria to the Concrete Debris DCGLs will be applied as part of the Basement Fill Model. For footings that are to remain and are volumetrically contaminated, the radioactivity inventory in the footing will be assessed and the total quantity will be conservatively included with the other sources to the containment basement (or with the discharge tunnel if they are assessed after the completion of containment liner FSS) in calculating future groundwater dose. This bounds the dose calculation as the calculation of the future groundwater concentration in containment includes the major radioactivity sources contained in subsurface structures to remain after license termination. Basements other than the containment and the fuel pit will be analyzed independently using the Basement Fill Model as they are not expected to contain significant levels of radioactivity and occur later in the decommissioning.

Aurimt 9005 4 qeazi 1-6 Ad a J

I%%;v.

Haddam Neck Plant License Termination Plan i_> Two primary scenarios have been selected as input to the RESRAD codes for calculating the radionuclide-specific Derived Concentration Guideline Levels (DCGLs). DCGLs are the concentration and surface radioactivity limits that will be the basis for performing the final status survey. These scenarios are the resident farmer scenario for site soils, and ground water and the building occupancy scenario for site buildings. Current decommissioning plans do not include the placement of concrete debris in facility basements, the concrete debris scenario, approved as part of the LTP approved in November 2002 is no longer applicable. If the decommissioning plans change, the option to use concrete debris as backfill (and the associated concrete debris DCGLs) is retained.

August 2005 1-6a Rev. 3

Haddam Neck Plant License Tcrmination Plan 1.3.7 Update of Site-Specific Decommissioning Costs In accordance with 10CFR50.82 (a)(9)(ii)(F), Chapter 7 provides an updated, site-specific estimate of the remaining decommissioning costs. It also includes a comparison of these estimated costs with the present funds set aside for decommissioning and a description of the means to ensure that there will be sufficient funds for completing decommissioning.

1.3.8 Supplement to the Environmental Report In accordance with IOCFR50.82 (a)(9)(ii)(G), Chapter 8 demonstrates that decommissioning activities will be accomplished with no significant adverse environmental impacts. Decommissioning and license termination activities remain bounded by the site-specific decommissioning activities described in:

• the PSDAR,

• the previously issued environmental assessment,

• the environmental impact statement,

• NUREG-0586, "Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities (FGEIS)" (Reference 1-17), and

• NUREG-1496, "Generic Environmental Impact Statement in Support Rulemaking for Radiological Criteria for License Termination of NRC-Licensed Nuclear Facilities."

(Reference 1-18).

The HNP PSDAR was submitted to the NRC in accordance with IOCFR50.82 (a)(4)(i). In the PSDAR, CYAPCO performed an environmental review to evaluate actual or potential environmental impacts associated with proposed decommissioning activities. This evaluation used NUREG-0586 and two previous site-specific environmental assessments as its basis. One site-specific assessment was performed from the conversion of the provisional operating license to a full-term operating license, and another was performed more recently from the recapture of the construction period time duration. The environmental review concluded that the impacts due to HNP decommissioning are bounded by the previously issued environmental impact statements.

As discussed in Chapter 6, the DCGLs for site buildings are calculated using the building occupancy scenario as the primary modeling scenario. Buildings to remain after release from the NRC License which are decontaminated at or below the DCGLs, could be allowed to remain standing after the final status survey. Consideration of the building occupancy scenario in determining the DCGL is compatible with the information in SECY 00-41 (Reference 1-19). SECY 00-41 concluded that the building occupancy and resident farmer scenarios, as well as assumptions used in the FGEIS to estimate public dose, are sufficiently conservative to bound such a condition. Chapter 8 also provides a summary description of the process CYAPCO will use to ensure that the non-radiological aspects of decommissioning meet state and federal requirements for release of the site.

1.4 Decommissioning Approach 1.4.1 Overview This section provides an overview of CYAPCO's approach to decommissioning the HNP site. References to the section in the LTP, where details concerning the particular step or stage of the decommissioning process are described, are given in parentheses.

A n ... -+A August 2005, 1-7 Rev. 3

Haddam Neck'Plant License Termination Plan Upon the decision to permanently cease power operations at the HNP site, CYAPCO began site characterization activities (Chapter 2). This characterization effort, which was performed to the guidelines of MARSSIM, included a Historical Site Assessment (HSA); a review of historical survey documentation; and measurements, samples, and analyses to further define the present radiological conditions of the site. The effort also addressed the status of the site relative to hazardous and state regulated non-radioactive materials.

The initial site characterization, together with geologic and hydrogeologic investigations of the site, provides the basis for the conceptualization of the site and the selection of the appropriate scenarios, models, and critical groups to address the possible future uses of the site. Conceptualization (creating the overall model for the site), which considers future use, characterization, geologic and hydrogeologic data, is also important in selecting the dose modeling code to be used to calculate the DCGLs and in the development of the Basement Fill Model for calculating "future groundwater" dose. These DCGLs correspond to a dose to the average member of the selected critical group that does not exceed 25 mrem/yr TEDE (Chapter 6).

Concurrent with site characterization and the conceptualization of the site, decommissioning activities are taking place. Activities performed during this period include the removal of contaminated components from the site for final disposition and demolition of some site buildings (Chapter 3).

Remediation of some site structures and soils will be performed, based upon the input of the initial site characterization and the DCGLs determined by dose modeling. In addition, remediation of groundwater may also be necessary to meet the dose criteria. Title 10 ofthe CFR, Section 20.1402 has dual criteria, namely 25 mrem/yr TEDE and ALARA. Accordingly additional remediation activities are evaluated to determine the cost/benefit of remediation beyond that which is necessary to meet the DCGLs along with the future groundwater dose calculated by the Basement Fill Model for the remaining portions of the SSCs. If the additional remediation activities are determined to be appropriate, they will be performed (Chapter 4). Once survey areas have been remediated to the required level, controls will be put into place to prevent re-contamination of the surveyed areas. (Section 5.4.6)

The Final Status Survey Plan (Chapter 5) describes the methodology by which land areas and buildings will be verified to be at or below the DCGLs (after accounting for the future groundwater dose), and thus meet the site release criteria for unrestricted use. Once final status surveys are performed for a specific land area or building, the data collected will be documented in a release record. Periodically, several release records will be compiled into a FSS Report and made available to the NRC as evidence of completion of activities and acceptability of the area for unrestricted release. CYAPCO plans to communicate the schedules for these final status surveys to the NRC so that independent confirmatory surveys can be scheduled and performed, as necessary.

CYAPCO may pursue backfill activities once the survey results for a survey area or group of survey areas are completed. For facility SSCs remaining onsite, the final status survey results will be compiled in a series of reports by survey area(s) and will be made available for NRC review and inspection. CYAPCO plans to demolish most structures to 4 feet below grade and selected basements and to dispose of the wastes generated at an LLW waste or other appropriate facility. Final status surveys will be performed to document the radiological condition of all remaining footings/basements and soil. The dose modeling approach, described in Chapter 6, evaluates potential exposures resulting from any remaining concrete structures, footings/basements to ensure that the doses are bounded by the conservative DCGLs (after accounting for future groundwater dose) specified in the plan. CYAPCO does not intend to use on-site burial, disposal or incineration of any low-level radioactive waste. Materials remaining onsite will meet the appropriate DCGLs, after accounting for "future groundwater" dose for unrestricted release, and thus are not low-level radioactive waste.

August 2005 1-8 Rev. 3 l

Haddam Neck Plant License Termination Plan Characterization efforts at the HNP decommissioning project are an iterative process spanning all aspects of the remediation activities. The information developed during the initialHNP characterization program represents a radiological and hazardous material assessment based on the knowledge and data available at the end of 1999. This information was sufficient to satisfy the objectives listed above. Additional measurements and samples obtained during the remediation process will continue to be assessed to ensure adequacy of area classification and effectiveness of the Final Status Survey to show compliancewith the established Derived Concentration Guideline Levels (DCGLs), in accordance with the guidelines of MARSSIM and to provide adequate data to allow confident calculation of the future groundwater dose from the basements to remain on site after release from the license The LTP provides the detailed information related to the decommissioning approach, dismantlement and bulk disposal, which will be used by CYAPCO to complete the decommissioning of the HNP. CYAPCO has replaced the future groundwater dose calculation method for basements contained in LTP Revision I a with the Basement Fill Model discussed in theLTP Revision 3. The decommissioning approach provided in Revision la of theLTP may be elected in the future for selected areas of the site. In the event that this approach is selected, the area classification approach in Revision Ia of the LTP will be implemented.

Appendix H contains historical information from Table 2-10.

2.2 Historical Site Assessment 2.2.1 Introduction The HSA for the HNP commenced in 1997, under the direction of the CYAPCO Radiation Protection Department staff. The process for conducting the HSA was established in accordance with MARSSIM guidelines. The HSA focused on historical events and routine operational processes that resulted in contamination of the plant systems, onsite buildings, exterior grounds and subsurface areas within the Radiologically Controlled Area (RCA); and grounds and subsurface areas outside of the RCA, but within the owner controlled area. The HSA, as part of the initial characterization program, was conducted to support the objectives detailed in Section 2.1.

In 1999, the HSA process became the task of Bechtel Power Corporation, as the Decommissioning Operations Contractor (DOC) at the time. The HSA was completed in the fall of 1999. The initial characterization report was issued in January of 2000 (Reference 2-2), and a Historical Site Assessment Supplement (Reference 2-3) was issued in August of2001. Section 2 of the License Termination Plan provides a summary of findings from the HSA and the information that is the basis for area classifications, input into the development of DCGLs, development of remediation plans, and design of the Final Status Survey. The scope of the HSA included potential contamination from radioactive materials, hazardous materials, and state-regulated materials. Ongoing characterization activities are being conducted as part of the CYAPCO's self-managing of the HNP decommissioning. The LTP includes a summary of the information contained in References 2-1 and 2-2. Additional characterization information and confirmation will continue throughout the decommissioning as part of the FSS process.

The LTP will generally not be updated to include this additional characterization.

2.2.2 Methodology The HSA was designed to evaluate input from two separate sources -plant records and personnel interviews. The review of plant records consisted of routine radioactive effluent release reports, non-routine reports submitted to the NRC under provisions of the technical specifications, 10CFR20, or IOCFR50; plant incident reports or condition reports; and findings documented in accordance with other assessment processes such as the Quality Assurance Program (QAP) and oversight activities. The K>iinformation obtained through this process forms the input data for the records that are maintained on site August 2005 2-2 Rev. 3

Haddam Neck Plant License Termination Plan

• demonstrate that the potential dose from small areas of elevated activity is below the release criterion for each survey unit.

The final status survey process consists of four principal elements:

• planning,

• design,

• implementation, and

• assessment.

The Data Quality Objective (DQO) and Data Quality Assessment (DQA) processes are applied to these four principal elements. DQOs allow for systematic planning and are specifically designed to address problems that require a decision to be made and provide alternate actions (as is the case in FSS). The Data Quality Assessment (DQA) process is an evaluation method used during the assessment phase of FSS to ensure the validity of survey results and demonstrate achievement of the sampling plan objectives (e.g., to demonstrate compliance with the release criteria in a survey unit).

Survey planning includes review of the Historical Site Assessment (HSA) and other pertinent characterization information to establish the radionuclides of concern and survey unit classifications.

Survey units are fundamental elements for which final status surveys are designed and executed. The classification of a survey unit determines how large it can be in terms of surface area. If any of the radionuclides of concern are present in background, the planning effort may include establishing appropriate reference areas to be used to establish baseline concentrations for these radionuclides and their variability. Reference materials are specified for establishing background instrument responses for cases where gross activity measurements'are to be made. A reference coordinate system is used for documenting locations where measurements were made and to allow replication of survey efforts if w necessary.

Before the survey process can proceed to the design phase, concentration levels that represent the maximum annual dose criterion of IOCFR20.1402 must be established. These concentrations are established for either surface contamination or volumetric contamination. They are used in the survey design process to establish the minimum sensitivities required for the available survey instruments and techniques, and in some cases, the spacing of fixed measurements or samples to be made within a survey unit. Surface or volumetric concentrations that correspond to the maximum annual dose criterion are referred to as Derived Concentration Guideline Levels, or DCGLs. Volumetric sample results will in some cases be used to calculate the "future groundwater" dose in building basements/footings using the Basement Fill Model rather than the application of DCGLs. The future groundwater dose is that which results from the leaching of radionuclides from buried concrete, the containment liner and embedded piping that is contained in basements to remain. A DCGL established for the average residual radioactivity in a survey unit is called a DCGLW. Values of the DCGLW may then be increased through the use of area factors to obtain a DCGL that represents the same dose to an individual for residual radioactivity over a smaller area within a survey unit. The scaled value is called the DCGLE0,C, where EMC stands for elevated measurement comparison.

After the DCGLw is established, a survey design is developed that selects the appropriate survey instruments and techniques to provide adequate coverage of the unit through a combination of scans, fixed measurements, and sampling. This process ensures that data of sufficient quantity and quality are obtained to make decisions regarding the suitability of the survey design assumptions and whether the unit meets the release criterion. Approved site procedures will direct this process to ensure consistent implementation and adherence to applicable requirements.

25 August 2005 5-2 Rev. 3 l

Haddam Neck Plant License Termination Plan Survey implementation is the process of carrying out the survey plan (package) for a given survey unit.

K> This consists of scan measurements, fixed measurements, and collection and analysis of samples. Data will be stored using a data management system.

August 2005 5-2a Rev. 3

Haddam Neck Plant License Termination Plan Optimize the Design for Obtaining Data The first six steps are the DQOs that develop the performance goals of the survey. This final step in the DQO process leads to the development of an adequate survey design.

5.4.2 Classification of Survey Areas and Units The adequacy of the final status survey process rests upon partitioning the site into properly classified survey units of appropriate physical area. Chapter 2 of this document discusses in detail the HSA for the HNP site and the classifications assigned to all of the site structures and grounds. Characterization is an ongoing effort throughout the decommissioning process, and survey unit classifications may be modified on the basis of new characterization information or impacts from decommissioning activities. The process described in Section 1.5 will be used to evaluate these changes. Survey areas have been determined as described in Section 2.3.3.2. The current approach is generally to remove the above-grade portions of site buildings and structures. Originally, the above-grade portions had been identified as survey areas and had been given MARSSIM classifications in Table 2-10. As final status survey activities are no longer planned for these areas, their survey area designations have been subsequently removed from Table 2-10. However, Appendix H contains the historical information from Table 2-10.

If it becomes necessary to final status survey these areas, the survey area designation and initial classification listed in Appendix H will be used. For the subsurface areas that will be evaluated using the Basement Fill Model, area classifications and survey unit area do not apply.

5.4.3 Survey Units A survey area may consist of one or more survey units. A survey unit is a physical area consisting of structures or land areas of a specified size and shape which will be subject to a final status survey.

Compliance with the applicable criteria will be demonstrated for each survey unit.

Survey units are limited in size based on classification, exposure pathway modeling assumptions, and site-specific conditions. The surface area limits, used in establishing the initial set of survey units for the HNP Final Status Survey Plan, are provided in Table 5-1 for structures and land areas. The area limits for structures refer to floor area, and not the total surface area, which would include the walls and ceiling.

This is consistent with the guidance of DG4006 (as incorporated in Section 2 of Appendix E to NUREG-1727) and MARSSIM. The floor area limits given in Table 5-1 were also used to establish survey unit sizes for structures such as roofs or exterior walls of buildings. The limits given in Table 5-1 will also be used should the need arise to establish any new survey units beyond the initial set given in this plan.

As indicated in Table 2-10 and 2-1 1A, B, and C, and Figures 2-1 through 2-9, areas of HNP that are classified as impacted have been divided into survey units to facilitate survey design. Each survey unit ha's been assigned an initial classification based on the site characterization process and the historical site assessment.

August 2005 5-7 Rev. 3

Haddam Neck Plant License Termination Plan 5.4.7 Selection of DCGLs ,  ;- .

Residual levels of radioactive material that correspond to allowable radiation dose standards are calculated by analysis of various pathways (direct radiation, inhalation, ingestion, etc.), media (concrete, soils, and groundwater) and scenarios through which exposures could occur. These derived levels, known as Derived Concentration Guideline Levels (DCGLs), are presented in terms of surface or mass activity concentrations. DCGLs usually refer to average levels of radiation or radioactivity above appropriate background levels. DCGLs applicable to building or other structural surfaces are expressed in units of activity per surface area (dpm/1 00 cm2). When applied to soil, sediments or structural materials where the radionuclides are distributed throughout, DCGLs are expressed in units of activity per unit of mass (pCi/g).

Chapter 6 of this plan describes in detail the modeling performed to develop the radionuclide-specific DCGLs for soil, groundwater and building surfaces. DCGLs are not needed for building footings (volumetrically contaminated), basements and activated concrete sources. Dose from these locations will be calculated using the Basement Fill Model. For situations where gross activity measurement methods are used to demonstrate compliance with the license termination criteria, the radionuclide specific DCGLs will be used to establish gross activity DCGLs. These gross activity DCGLs will be established based on a representative radionuclide mix established for each survey unit. In cases where measurable activity still exists, it is expected that the radionuclide mix will be established based on gamma-ray spectroscopy and alpha spectroscopy (where conditions warrant) or equivalent analyses on representative samples, with scaling factors used to establish the activity contribution for any hard-to-detect radionuclides that might be present. Scaling factors will be selected from available composite waste stream analyses or similar assays. Such analyses are performed periodically and documented in support of waste characterization needs.

For cases of survey units for which there is no measurable activity distinguishable from background, a representative radionuclide mix will be selected based upon historical characterization information for the survey unit of interest or for units with similar history and physical characteristics (e.g., information from adjacent areas).

To show compliance with 25 mrem/yr and ALARA, the unity rule will be applied in those areas in which the dose can be a result of soil, existing groundwater and future groundwater residual radioactivity. Use of the unity rule, as discussed in Section 5.4.7.1, will result in the development of operational DCGLs on a radionuclide-specific basis.

5.4.7.1 Operational DCGLs The DCGLs are developed in Chapter 6 for exposures due to three potential media. These exposures include that from residual radioactivity in soil, existing groundwater (GW) radioactivity, and additional future groundwater radioactivity from the building basements and footings. The areas of the site where these exposures could occur concurrently are where subsurface structures and concrete are buried and August 2005 5-13 Rev. 3

Haddam Neck Plant License Termination Plan existing groundwater contamination may be present. This area represents approximately 15,600 mr and includes the industrial area of the site. For this area, the total dose from these sources, HTota can be expressed as:

HTOtal = HSoiU + HFxunnivgG + HFulreGw (Equation 5-1)

For soil and existing groundwater, the dose from the residual radioactivity from radionuclide i is:

H' - 25* CGL' (Equation 5-2)

For the future groundwater pathway, the dose is determined from the Basement Fill Model.

Since the limit for the total annual dose is 25 mrem from all media (and all pathways), a reduction to the soil and existing groundwater DCGLs in Chapter 6 is needed, since these are based on an annual dose of 25 mrem from each media. The DCGLs in Chapter 6 are therefore considered "Base-Case (Base)"

values. The reduced DCGLs, or "Operational DCGLs" (DCGLop), can be related to the base case DCGLs using the principal relationship from:

H' =25* DCGL'oaps (Equation 5-3)

In the case of existing groundwater, the contamination concentration to be used for calculating dose is the highest measured at any point within the survey area or within the plume area boundary distance (largest capture zone radius as determined by the capture zone analysis described below) from the subject survey area at the time of notification of the NRC of intent to release the subject survey area from the license.

The following considerations may be included in determining if the results trend is sufficient to utilize the groundwater well sample results in the dose calculation for an affected survey unit:

• Fate and transport simulations will identify the projected area of highest groundwater concentration on site.

• The locations of existing wells will be examined in relation to the simulation results and additional wells constructed to ensure adequate monitoring of the area(s) of anticipated highest groundwater radionuclide Substances Of Concern (SOC) concentrations.

• Monitoring wells from which the sample results are to be used for the dose calculation for a survey unit will have been sampled quarterly for at least 18 months including two springtime high water table periods. In the case of areas where remediation (e.g., removal of contaminated soil below the average water table) has been conducted using groundwater depression, the 18 month monitoring period will begin. when use of the groundwater depression systems has ended. Prior to turning off the depression system, remediation will have been completed and excavation backfilled.

• Monitoring well results show groundwater contaminant concentrations to be below closure criteria as discussed in this section, and exhibit steady or decreasing trends.

The 18 month monitoring period is sufficient for the following reasons:

A5 August 2005 5-14 Rev. 3 l

Haddam Neck Plant License Termination Plan

• Historical releases at HNP and subsequent migration of groundwater contaminants appear to have resulted in dispersion of SOCs in groundwater. Actions completed to date have removed primary contaminant sources (e.g., contaminated process solutions) and processes (e.g., bulk waste water processing with leaking tanks) that historically contributed to observed groundwater contamination. As a result, only secondary contaminant sources (which could include residual subsurface soil contamination, grossly-contaminated groundwater and contaminated subsurface structures) remain at the site. The highest concentrations generally remain near historical source areas in wells that are completed within the unconsolidated soil formation that is slated for remediation.

• For all areas where groundwater contamination has been detected, this duration (when two springtime periods are included) ensures that the effect of the high water table season is included twice. Seasonal high water table levels impacting contaminated soils above the average water table level is one of the factors that can cause a seasonal increase in groundwater radionuclide concentrations.

• For areas where remediation has been conducted below the normal water table for the purpose of removing media suspected of contributing to groundwater contamination, the 18 month period (after the area has been backfilled and returned to normal groundwater levels) is expected to provide sufficient time for groundwater to leach through the remediated and backfilled area and for sampling of nearby monitoring wells to ensure the effectiveness of the remediation. As stated above, this will be confirmed by ensuring that the groundwater activity levels are steady or decreasing during this 18 month monitoring period.

The future groundwater component of equation 5-1 can be further stated as follows:

HFutu"reGW- FutureGroundwaterDose (Equation 5-4)

For building basements and footings to remain, the future groundwater dose will be calculated by the Basement Fill Model. The dose calculation method future groundwater is discussed in Section 6.8.

The HE,&tjngGW term, from Equation 5-1, will be applied to survey areas in which the presence of groundwater contamination has been detected and survey areas that are within the capture zone, the influence boundary distance of detectable ground contamination. "Detected groundwater contamination" is defined as the presence of:

• Plant-related radionuclides, which are also present in background, at a concentration greater than two standard deviations over background, or

• Plant-related radionuclides, not present in background, at a concentration greater than the Minimum Detectable Concentration and greater than two times the standard deviation in the net concentration.

Table 5-3 provides the survey areas to which the HExLstingsGw term would currently be applied. Table 5-3 is based upon as additional groundwater characterization and completion of the capture zone analysis. The August 2005 5-15 Rev. 3

Haddam Neck Plant License Termination Plan S..e} .. ,, .

capture zone analysis determined a maximum zone of influence of 100 meters around a groundwater monitoring well, (see Figure 5-3 and 5-3.1) and reference 5-13, Estimated Zone of Influence/Capture Zone for Hypothetical Water Supply Wells in Post-Closure Dose Modeling, CH2MHILL< Technical Memorandum, dated January 11, 2005. These figures depict the capture zone around the wells which have shown detectable contamination at the perimeter of the industrial area and around peninsula wells.

It is noted, however, that characterization efforts for groundwater contamination are still ongoing and the survey areas to which the HEistinsGW term are applieidmai change. Those changes will be communicated to the NRC. This change may be caused by changes in the location of the plume, detection of groundwater contamination at locations outside the plume or changes to the capture zone or detection of groundwater contamination at locations outside the zone. The Phase 2 Hydrogeologic Work Plan, as described in Section 2.3.3.1.6, will provide additional characterization of groundwater that will be used to better define the groundwater contamination plume.

Prior to the request to release any portion of the site from the license, CYAPCO will prepare and make available for inspection a capture zone analysis (provided to the NRC in January 2005), based on data collected as part of the Phase 2 Hydogeological Work Plan, to better define the capture zone distance, Reference 5-13. The "capture zone" is the area surrounding a hypothetical well to be used by the resident farmer, from which existing groundwater contamination could be drawn into the resident farmer's well.

The analysis used to determine this area used the hydrogeological conditions and parameters assumed in the Resident Farmer Scenario as described in Chapter 6 of the LTP Table 5-3 Survey Areas Affected by. Groundwater Contamination K> __Survev Aren 1000 9306 9522 2000 9308 9527 3000 9310 9528 (Units 0,2&3) 4000 9312 9530(Units 1,2,3,&4) 5000 9313 9801 6000 9502 9802 9102 9512 .9803 0106 9514 9804 9226 9518 9805 9302 9520 9304 9521 The compliance formulation for these resident farmer exposure scenarios is re-written as:

(Equation 5-5) 1 ()F DCGLOp psoil + DCGLOP-EtstingGW HFtueGTV D BCGLase Soil DCGLase-EsfingGlW 25eW August 2005 5-16 Rev. 3

Haddam Neck Plant License Termination Plan i.; 1 - 9, s' For simplicity Equation 5-5 may be re-written as:

1 2 (i) k5o + ftinG + fAlhreGWI] Equation 5-6 where, for a given radionuclide i, f'Soul is the fraction of the total dose from soil..IL,,jgow is the fraction of the total dose from existing contamination in groundwater, andFwureGWis the fraction of the 25 mrem dose that is calculated by the Basement Fill Model.

The use of this equation requires that only one variable be unknown. Therefore, values for Future Groundwater Dose andfajisngGwwill need to be selected in order to calculatefo 01 i. As the building surface operational DCGL are independent of soil and groundwater dose contribution, they will be set based on an ALARA evaluation and/or an administrative dose level at or below 25 mrem/yr.

The following example is provided to illustrate the use of the operational DCGLs for land areas that have the potential for existing and future groundwater dose, with the following assumptions:

• foil= 0.3

• f.btigGw = 0.2 o Therefore, Future Groundwater Dose (fraction of 25 mrem/yr) = 0.5 The above determination will be made prior to the performance of any final status surveys of soils or building surveys in areas where existing groundwater contamination will impact the potential dose. This determination will be provided in the FSS Report or a technical support document and will be applied to the affected survey areas.

The following table provides as example of the building surface operational DCGL for Cs-137 using the fractional values from above.

Table 5-4 Operational DCGL Example for Cs-137 Using Fractional Values from Above

• Base Case Operational DCGL/Dose DCGL/Dose Soil (pCi/g) 7.91E+00 2.37E+00 Existing Groundwater (pCi/l) 4.3 1E+02 8.62E+01 Future Groundwater Dose (mrem/yr) 25 12.5 August 2005 5-17 Rev. 3 1

Haddam Neck Plant License Termination Plan 5.4.7.2 Gross Activity DCGLs For alpha or beta surface activity measurements, field measurements will typically consist of gross activity assessments rather than radionuclide-specific techniques. Gross activity DCGLs will be established, based on the representative radionuclide mix, as follows:

DCGLGA =, (Equation 5-7) s DCGL where:

f= fraction of the total activity contributed by radionuclide i i = the number of radionuclides DCGL, = DCGL for radionuclide i Gross activity DCGLs can be developed for gross beta measurements, or a gross beta DCGL can be scaled so that it acts as a surrogate for gross alpha (see Section 5.4.7.3). Equation 5-7 will be applied for radionuclides that are present in a survey unit in concentrations greater than 5% of their respective DCGL. The aggregate of all radionuclides not included in the gross activity DCGL, based on the percentage of their respective DCGL, will not exceed 10%. This practice is conservative relative to the process presented in 10CFR20 in which radionuclides that contribute less than 10% to dose, provided the aggregate does not exceed 30%, and are not required to be included in the dose assessment.

K~ 5.4.7.3 Surrogate Ratio DCGLs It is acceptable industry practice to assay a Hard-To-Detect (HTD) radionuclide by using a surrogate relationship to an Easy-To-Detect (ETD) radionuclide. A common example would be to use a beta measurement to assay an alpha emitting radionuclide. Another example would be to relate a specific radionuclide, such as cesium-137, to one or more radionuclides of similar characteristics. In such cases, to demonstrate compliance with the release criteria for the survey unit the DCGL for the surrogate radionuclide or mix of radionuclides must be scaled to account for the fact that it is being used as an indicator for an additional radionuclide or mix of radionuclides. The result is referred to as the surrogate DCGL.

.The following process will be applied to assess the need to use surrogate ratios for final status surveys (FSS).

• Determine whether HTD radionuclides (e.g., TRU, Sr-90, H-3) are likely to be present in the survey unit based on process knowledge, historical data or characterization.

• When HTD radionuclides are likely to be present establish a relationship using a representative number of samples (typically six or more). The samples may come from another survey unit if the source of the contamination and expected concentrations are reasonably the same. These August 2005 5-18 Rev. 3

Haddam Neck Plant License Termination Plan 5.4.7.5 Building Basement and Footings After completion of final status survey activities of the remaining portions of structures, some subsurface concrete may remain in the form of building footings.

As these structures are solid concrete or steel structures, and will not be left in a condition to allow them to realistically be occupied, the only applicable dose pathway is from groundwater contamination from the leaching of radionuclides from these structures. The dose model for the calculation of this "future groundwater" dose is call the Basement Fill Model and is discussed in Section 6.8. The sampling to be performed to determine the radioactivity inventory to be used in calculating future groundwater dose is discussed in Section 5.7.1.6.

5.4.7.6 Release Limits for Non-Structural Components and Systems In general, non-structural components and systems will be surveyed to site unconditional release limits, i.e., no detectable radioactive (licensed) material. These surveys will be performed in accordance with health physics procedures and are consistent with the requirements of NRC Information Notice 85-92, "Surveys of Wastes Before Disposal From Nuclear Reactor Facilities," and IE Circular 81-07, "Control of Radioactively Contaminated Material." Separate limits will be applied at the time of Final Status Survey to the buried piping located in the saturated subsurface areas of the site and to embedded piping and penetrations. These limits are discussed in the following paragraphs.

For buried piping in contact with the saturated zone, an analysis has been performed to determine surface activity limits for the remaining piping that will result in no more than a 1mremn/yr dose (Reference 5-8).

This piping will be grouted with concrete (after any required remediation and surveying), as agreed to with the State of Connecticut DPUC. To simplif' the analysis, the piping material is assumed to be K> eroded away, leaving the slug of grout with the contamination from the interior surface of the piping.

Consistent with these simplified assumptions, the DCGLs calculated in Chapter 6 of the LTP approved in November 2002 for concrete debris are used in developing the surface contamination limits for this piping.

In order to calculate the release limits for the piping (corresponding to 1 mrem/yr), first, for each radionuclide, the DCGL representing 25 mrem/yr from all pathways for concrete debris and the fraction of dose from the water dependent pathways were used to determine the volumetric limits from water dependent pathways only (as the buried piping is well below the soil surface, thus eliminating external dose contribution, and is in contact with the groundwater). These limits are then normalized to represent a volumetric limit that would result in l mrem/yr. Finally, the volumetric contamination is converted to surface contamination for various piping diameter sizes, (bounding value for the pipe diameters in question, because the larger the diameter, and subsequently the radius, the larger the surface activity limits can be). The release limits to be applied to this piping are given in Table 5-7. The surface contamination levels for the various piping sizes when converted to the volumetric contamination based on the grouting of the piping does not change the effective volumetric concentrations being left. This is a result of using the volumetric limit that results in 1 mrem/yr dose in pCi/gm to scale the surface contamination limits for various piping sizes..

August 2005 5 Rev. 3

( Haddam Neck Plant LicE Termination Plan C ITable 5-7 Relense Limits ror Buried Piping, dpm/IlOOcm2 1" 2" 2 1/2 3" 4" 8" 10l 12" 14" 16" 18" 20" 24" 26" Rodionucilde 11-3 1.301+03 2.616+03 3.262+03 3.91E+03 5.21E+03 1.04E+04 1.301404 1.564+04 1.8212+04 2.082+04 2.342+04 2.61E+04 3.131+04 8.47E+43 C-14 1.94E.404 3.89E+04 4.86E404 5.83E+04 7.77E+04 1.55E+05 1.94E+05 2.33E405 2.72E+05 3.11213+05 3.50F.+05 3.89E+05 4.66E+05 1.26E+05 Mn-54 1.332+04 2.66r.+04 3.32e+04 3.98sr404 5.31 E+04 1.062+05 1.33E+05 1.59E+05 1.86E+05 2.12E+05 2.39e+05 2.66E+05 3.1912+05 8.63E404 Fe-55 1.54E+04 3.09E+04 3.86E+04 4.632+04 6.171+04 1.23E405 1.542405 1.852+05 2.16E+05 2.472+05 2.781405 3.092405 3.702+05 1.001+05 Co-60 8.032+04 1.61 E+05 2.0 1E+05 2.41 2+05 3.21 e+05 6.422+05 8.032+05 9.632+05 1.122+06 1.282+06 I A44e+06 1.6112406 1.932+06 5.22E+0)5 Ni-63 3.80E+04 7.60e+04 9.50E+04 1.1412405 1.521+05 3.042+05 3.802+05 4.562+05 5.32E+05 6.08E+05 6.842+05 7.60E+05 9.12E+05 2.47e+05 Sr-90 4.68E+01 9.351+01 1.172+02 1.401+02 1.87E+02 3.742+02 4.682+02 5.61 12+02 6.552+02 7.482+02 8.4A2+02 9.3523+02 1.121+03 3.04 E+02 Nb-94 3.432+04 6.852404 8.56E+04 L.03E+05 1.37F+05 2.74E+05 3.43E+05 4.11E+05 4.802+05 5.482+05 6.171+05 6.85E+05 8.222+05 2.23E+05 Tc-99 6.10E+03 1.22E+04 1.53E+04 1.832+04 2.442+04 4.882+04 6. 101+04 7.322+04 8.542+04 9.76e+04 1.1I0+05 1.22+s05 1.46F.+05 3.97r+04 .I-Ag-l0nsm 3.432+05 6.85E+05 8.56E+05 1.03E+06 1.372+06 2.74E+06 3.43E+06 4.11C+06 4.80E+06 5.48E+06 6.17E+06 6.852+06 8.222+06 2.23E+06 Cs-134 2.092404 4.1 8E+04 5.222+04 6.262+04 8.352+04 1.67e+05 2.092+05 2.512E+05. 2.922+05 3.34E+05 3.76+05 4.1U8E405 5.01I +05 1.36fF405 Cs-137 2.42E+04 4.832+04 6.04E+04 7.252+04 9.662+04 1.93E+05 2.422+05 2.902+05 3.38E+05 3.86E+05 4.35E+05 4.83E405 5.803405 1.571o405 Eu-152 6.702+04 1.342+05 1.682+05 2.01E+05 2.682+05 5.362+05 6.70E+05 8.042+05 9.382+05 1.07e+06 1.21 E+06 1.3423+06 1.61 E+06 4.36E405 E2-154 4.68E+04 9.35E+04 1.17E+05 1.40n+05 1.872+05 3.74e+05 4.682+05 5.61 e+05 6.55E405 7.48E+05 8.42E+05 9.35s+s05 1.12E+06 3.042405 Eu-155 3.002+05 6.006+05 7.502+05 9.00-+05 1.20E+06 2.402+06 3.00E+06 3.60e+06 4.202+06 4.80E+06 5.402+06 6.003+06 7.20E+06 1.95E+06 Pu-238 1.882+02 3.752+02 4.692+02 5.632+02 7.502+02 1.50E+03 1.88E+03 2.25s+03 2.632+03 3.002+03 3.38e+03 3.752403 4.50E+03 1.22E+03 Pu-239 1.71 E+02 3.4 12+02 4.262+02 5.122+02 6.822+02 1.361+03 1.7 1+03 2.052+03 2.392+03 2.732+03 3.072+03 3.4 1E403 4.092+03 1.1 2E+03 Pu-241 2.852+03 5.7012+03 7.133403 8.552403 1.142404 2.28E+04 2.852+04 3A2E+04 3.99E+04 4.56E+04 5.132+04 5.70E+04 6.842+04 1.85E+04 Am-241 8.33E+0I 1.67E+02 2.082+02 2.502402 3.332+02 6.66E402 8.332402 9.992+02 1.172+03 1.3323403 1.502403 1.67E+03 2.002+03 5.41 E402 Cm-243 1.151s+02 2.3 1E+02 2.882+02 3.462+02 4.6 1E+02 9.222+02 I. I 1E+03 1.38E+03 1.6I2+03 1.842+03 2.072+03 2.3 I+03 2.772+03 7.49E+02 August 2005 5-24 Rev. 3

Haddam Neck Plant License Termination Plan Embedded pipe represents medium- io large-bore penetrations (up to 42-inch) or small-bore piping (4-inch to 12-inch) that was built into concrete walls and run through structures including walls, ceilings and floors. The length of the piping for each segment is short, approximately the length of the thickness of the structure that the pipe penetrates, and in most cases it is expected to communicate perpendicular to the surface penetrated. The total length of this type of pipe has been estimated to be less than 1000 feet, segregated into a substantial number of individual segments.

Where the gross activity beta-to-alpha ratio at the time of FSS is 15:1 or greater, the piping will be left in place, and the building surface DCGLs will be applied during FSS. The basis and rationale for applying these DCGLs to embedded pipe are provided below:

• It is unlikely that access to piping 24 inches or less in diameter could occur.

• The majority of piping and penetration lengths greater than 24 inches in diameter are either run vertically (i.e., run through floor or ceiling) or are located six feet or more above the floor elevation. Thus it is unlikely that access to these pipes and penetrations would occur.:.

• An evaluation of the doses associated with accessing the piping and penetration was performed using a conservative radionuclide mixture where the gross activity beta-to-alpha ratio is 15:1 (Reference 5-9). Based upon the information contained in HNP waste stream characterization data, this mixture is expected to bound those conditions found at the site. This mixture corresponds to a composite sanmple'of contamination from the Waste Disposal Building, where the beta emitting radionuclides corresponding to the gross beta activity include: Mn-54, Co-60, Sr-90, Nb-94, Tc-99, Ag-108m, Cs-134, and Cs-137; and the gross alpha radionuclides include:

Pu-238, Pu-239/240, Cm-233/234, and Am-241. This evaluation calculated doses for a variety of pipe diameters (12-, 24-, 36-, and 42-inch), conservatively assuming the same duration of occupancy used in the building occupancy scenario (2340 hours0.0271 days <br />0.65 hours <br />0.00387 weeks <br />8.9037e-4 months <br /> per year) and applied a dose due to inhalationi and ingestion that is tvice those calculated in the building occupancy scenario. The results of the evaluation showed that the doses calculated using these conservative assumptions were only slightly higher than those associated with the building occupancy scenario and were thus acceptable.

As the evaluation is valid for situations in which the gross beta-to-alpha ratio for an embedded pipe is 15:1 or greater (at the time of FSS), if this condition is not met at the time of FSS, the piping will be removed, grouted, or capped to prevent access.

When present in a survey unit, embedded pipe and penetrations will be evaluated using the data quality objective process during survey planning and either removed or incorporated into the survey sample design, using the building surface DCGL as the applicable release criteria (under the conditions stated above). The decision to remove these pipes will be done as part of an ALARA evaluation for the subject survey unit.

5.5 Final Status Survey Design Elements-Surface Soils and Structures and Basements Sampling and surveys required to support the Basement Fill Model calculation for future groundwater dose will be taken of the basement concrete structures to determine the volumetric concentrations for subsurface structures within the water table. The number of samples to be taken for each basement and major subsurface feature to remain is given in Section 5.7.1.6. An assessment will be performed to determine the inventory of radioactivity in any footings that exhibit measurable radioactivity (i.e. >2 August 2005 5-25 Rev. 3

Haddam Neck Plant License Termination Plan sigma error of the analysis MinimfuimtDetectable Activity) in concrete samples or in the surrounding soil.

These inventories for footings will be included in the Basement Fill Model calculation as described in K.-" Section 6.8. The final status survey design elements and requirements for all other media and materials is further discussed below.

The general approach prescribed by MARSSIM for final status surveys requires that at least some minimum number of measurements or samples be taken within a survey unit, so that the non-parametric statistical tests used for data assessment can be applied with adequate confidence. Decisions regarding whether a given survey unit meets the applicable release criterion are made based on the results of these tests. Scanning measurements are used to check the design basis for the survey by evaluating if any small areas of elevated activity exist that would require reclassification, tighter grid spacing for the fixed measurements, or both. However, MARSSIM also recognizes that alternatives to this general approach for final status surveys exist. Specifically, MARSSIM states that if the equipment and methodology used for scanning are capable of providing data of the same quality as fixed measurements (e.g., detection limit, location of measurements, ability to record and document results), then scanning may be used in place of fixed measurements, provided that results are documented for at least the number of locations that would have been necessary had fixed measurements been used.

Final status surveys for the HNP surface soils and structures will be designed, following MARSSIM guidance, using combinations of fixed measurements, traditional scanning surveys, and other advanced survey methods, as appropriate, to evaluate survey units relative to their applicable release criteria. As MARSSIM does not directly address final status survey for subsurface soils, the principles of MARSSIM will guide the design of these surveys. Subsurface survey considerations can be found in Section 5.7.3.2.2.

Under MARSSIM, the level of survey effort required for a given survey unit is determined by the K.- potential for contamination as indicated by its classification. Class 3 survey units receive judgmental scanning and randomly located measurements or samples. Class 2 survey units receive scanning over a portion of the survey unit based on the potential for contamination, combined with fixed measurements or sampling performed on a systematic grid. Class 1 survey units receive scanning over 100% of the survey unit combined with fixed measurements or sampling performed on a systematic grid. Depending on the sensitivity of the scanning method, the grid spacing may need to be adjusted to ensure that small areas of elevated activity are detected.

For combinations of fixed measurements and traditional scanning, MARSSIM methodology is to select a requisite number of measurement locations to satisfy the decision error rates for the non-parametric statistical test to be used for data evaluation and to account for sample losses or data anomalies. The purpose of scans is to confirm that the area was properly classified and that any small'areas of elevated activity are within acceptable levels (i.e., are less than the applicable DCGLEmc). Depending on the sensitivity of the scanning method used, the number of fixed measurement locations may need to be increased so the spacing between measurements is reduced. Details on selecting the number and location of fixed measurements are the subject of Section 5.5.1 and subsequent subsections of this plan. The coverage requirements that will be applied for scans performed in support of final status surveys for the HNP site are:

• ForClass 1 survey units, 100%ofthesurfacewillbescanned; Rev. 3 5-26 August 2005 August 2005 5-26 Rev. 3

Haddam Neck Plant License Termination Plan

• For Class 2 survey units, between 10% and 100% of the surface will be scanned in a combination of systematic and judgmental measurements for outdoor units and for floor and lower walls of structures; and 10% to 50% of the surface will be covered for upper walls and ceilings;

• Scanning will be done on a judgmental basis for Class 3 survey units.

Though the emphasis of the document is on conducting final status surveys through a combination of fixed measurements and scans, MARSSIM also allows for use of advanced survey technologies as long as these techniques meet the applicable requirements for data quality and quantity. "Advanced technologies" in this context refers to survey techniques where the instrument is capable of recording data as an area is surveyed and the measurement sensitivity is an acceptable fraction of the applicable DCGLw (see Section 5.7.1.3). Such methods are desirable for final status surveys since they allow survey units to be assessed with a single measurement rather than separate fixed measurements and scans.

Advanced survey techniques may be used alone or in combination with fixed measurements and scans to assess a survey unit. For Class 1 and Class 2 units, two conditions must be met for advanced technologies to be employed as the only survey technique: an acceptable fraction of the survey unit surface area must be scanned; and the minimum detectable concentration (MDC) for the measurements must be an acceptable fraction of the DCGLw. For Class 1 units, 100% of the area must be covered. For Class 2 units, the coverage requirements for advanced technologies to be used alone are from 50% to 100% of the area for outdoor survey units or for floors and lower walls; and from 10% to 50% of the area for upper walls and ceilings. In cases where these coverage requirements cannot be achieved by an advanced survey technology or where the MDC is too large relative to the applicable DCGLW (see Section 5.5.1.5),

the survey will be augmented with fixed measurements and traditional scans as necessary in accordance with Section 5.5.1 and subsequent subsections of this plan. Advanced technologies may be used for judgmental assessments in Class 3 areas as long as the following MDC requirements are met.

For fixed measurements, MARSSIM states that MDCs should be as far below the DCGLW as possible, with values less than 10% of the DCGLw being preferred, and up to 50% of the DCGLW being acceptable.

August 2005 5-26a Rev. 3

Haddam Neck Plant License Termination Plan 5.7.1.6 Contaminated Concrete Basements The Basement Fill Model treats contaminated concrete as a volumetric source of radioactivity. It is therefore appropriate to utilize volumetric concrete sample results to determine the data to be used in the calculation. Table 5-10 shows the number of samples that have been taken to date and the number of additional samples that will be taken to provide enough characterization to allow the confident calculation of the future groundwater dose. The samples taken will be analyzed so that the profile with the depth of the concrete can be confidently shown. Except for the in core sump, the sampling will include analysis of concrete from the inside and outside surfaces and for areas inside the wall with at least 15% of the wallfloor thickness characterized. For the In Core Sump, the total depth of the sample will be analyzed to determine the radioactivity profile for this area.

Table 5-10 Volumetric Concrete Sample Requirements Basement Area to Remain Concrete Samples Additional Concrete Minimum (Below Elevation 17'6") Collected to Date Samples to be Collected Total Number Of (10/30/2004) Samples To be Used for the Inventory Calculation Containment Mat 8 6 14 Containment Walls 4 6 10 In Core Sump 1 8 9 Spent Fuel Pool 0 12 12 Cable Vault 7 6 13 "B" Switchgear Building 0 8 8 Discharge 0 10 10 Tunnels/Structure Intake Structure 2 6 8 The mechanism that has caused volumetric contamination in concrete is in many cases specific to certain radionuclide. Radionuclides such as H-3 and Sr-90 have been detected in concrete in contact with contaminated groundwater. Areas that have been subject to substantial neutron flux typically display H-3, Fe-55, Co-60, Eu-152 among others. To adequately assess the volumetric contamination of concrete,.a wafer from at least 20% of the locations listed in table 5-10 will be analyzed for all 20 radionuclides listed in Table 2-12. For radionuclides expected in certain areas of concrete, a sufficient number of wafers from all locations will be analyzed for the expected radionuclides to allow determination of a profile in the concrete at that location.

August 2005 5-4al Rev. 3 l

Haddam Neck Plant License Termination Plan The scan MDCs wvill be documented prior to performing the final status survey.

5.7.2.6 Typical instrumentation and MDCs Table 5-11 provides nominal data for the types of field instrumentation anticipated for use in the final survey efforts for the Haddam Neck Plant. The efficiencies listed in Table 5-10 are the total efficiencies in counts/disintegration, and the background count-rates shown are nominal values for generic materials.

This table is provided to show the relative sensitivity of some of the types of instruments that will be used during the final status surveys and allow the readers to compare the sensitivities to the DCGLs in Chapter6 of the LTP. The instrument efficiency (el) and source efficiency (£5) will be evaluated for instruments used for final status survey measurements and documented as part of the calibration records.

This evaluation will include the effects of surface to detector distances, surface coatings and the depth of contamination in material (e.g., concrete) on instrument performance. Instrument calibration sources will be chosen that are appropriate for use for the radionuclides expected to be present post remediation.

Instrument readings will be converted to activity by selecting conservative efficiency factors based upon the building surface conditions (including the depth of contamination in concrete).

August 2005 5-48 Rev. 3

Haddam Neck Plant License Termination Plan.

Table 5-11 I

Available Instruments and Associate&MDCs Instrument Application Nominal Nominal Nominal MIDC Nominal Scan Efficiency Background (fixed MIDC (Not Media measurement)

Specific) pancake GM beta-gamma 17% (Tc-99) 50 cpm 1,050 dpm/100 cm' 3140 dpm/100 probe (20 cm2) scans or fixed (I minute count) cm2 measurements for structure surfaces gas proportional alpha or beta P plateau: 16% 350 cpm (P 560 dpm/100 cm" 1770 dpm/100 counter (100 cm2 ) scans or fixed (Tc-99); plateau); (P plateau) cm2 (P plateau);

measurements a plateau: 15 cpm (a 90 dprn/100 cm2 (a 400 dprn/100 for structure 23% (Am-241) plateau) plateau); I minute cm2 (a plateau) surfaces counts plastic scintillator beta-gamma 30% (Co-60) 600 cpm 390 dpm/100 cm' 1230 dpm/100 (100 cm2) scans or fixed (I minute count) cm2 measurements for structure surfaces dual-phosphor scans or fixed 20% (Co-60) 300 cpm (P 420 dpm/100 cm1 1300 dpm/100 scintillator measurements; a 18% (Am-241) mode); (1 mode); cm2 (,B mode);

(100 cm 2 ) and1, 6 cpm (a 80 dpm/100 cm2 (a 400 dpm/100 independently or mode) mode) cm7 (a mode) simultaneously _

ZnS scintillator alpha scans or 19% (Pu-239) 2 cpm 50 dpmIlOO cm' (I 400 dpmr/100 (100 cm2) fixed . minute count time) cm2 measurements on structure surfaces 1.25-inch by gamma scans for Varies with Varies with N/A 6 pCi/g Co-60 1.5-inch NaI soil energy energy 11 pCi/g Cs-137 2-inch by 2-inch gamma scans for Varies with Varies with N/A 1.5 pCi/g Co-60 NaI soil energy energy 6 pCi/g Cs-137 3-inch by 3-inch in-situ gamma Varies with Varies with 0.1 pCi/g Co-60 N/A NaI spectroscopy- energy and energy and 0.2 pCi/g Cs-137 soil geometry geometry (10 minute counts)

HPGe in-situ gamma spectroscopy -

Varies with energy and Varies with energy and 0.05 pCi/g Co-60 0.05 pCi/g Cs-137 N/A II soil geometry geometry (10 minute counts) position-sensitive scan-and-record Co-60 (,B): 18% 350 cpm/100 Typical values are 1,925 dprn/100 cm1 proportional surveys Am-241 (a): cm2 beta f and 200 dpm/100 cm2 a counter 23% 15 cpm/100

_ I cm2 alpha' I__

August 2005 549 Rev. 3 I

Haddam Neck Plant License Termination Plan survey locations. In some cases, it may be necessary to core, drill, or use other methods as necessary to y> gain access to areas for sampling.

5.7.3.1.1 Activity Beneath Surfaces X Floors, walls, and ceilings of structures may have surface irregularities such as cracks and crevices that require special consideration in the survey process. Such considerations may consist of fixed measurements, longer count times, adjustments to counting efficiencies, sampling of material, or any combinations of these approaches.

Plant areas where residual radioactive material beneath a painted surface is known or suspected to be present will also require special consideration. Sampling will be performed, as appropriate, to confirm or deny the presence of residual activity. If activity is found, the samples should be used to determine both the radionuclides that are present and the density-thickness of the paint layer(s) in order to assess the need for correction factors for counting efficiencies. Such corrections, if required, will be determined following the guidance given in Section 5 of NUREG-1507. The effect of any such corrections on instrument MDCs will be assessed to ensure that measurements can still be performed with the required sensitivity relative to the applicable DCGLs.

5.7.3.1.2 Below-Grade Building Foundations 5.7.3.1.2.1 Basements The interior surfaces of the containment liner located below grade will be surveyed and decontaminated to meet the Building Occupancy DCGLs discussed in Chapter 6. The contamination levels used in the design of this survey will be used as the basis of the inventory of radioactivity released from the containment liner and embedded piping in the Basement Fill Model. Exterior surfaces of below-grade basements will be evaluated using the historical site assessment and other pertinent records to determine the potential for sub-surface contamination on these surfaces of below-grade basements. One method available to evaluate the exterior surfaces is the use of core bores through foundation or walls and the taking of soil samples at locations having a high potential for the accumulation and migration of radioactive contamination to sub-surface soils. These biased locations for soil and concrete assessment could include stress cracks, floor and wall interfaces, penetrations through walls and floors for piping, run-off from exterior walls, and leaks or spills in adjacent outside areas, etc. If the soil is found to be free of residual radioactivity at the biased locations, it will be assumed that the exterior surface of the foundation is also free of residual activity. Otherwise, additional sampling may be necessary to determine the extent of decontamination and remediation efforts. Another method available for evaluating the exterior surfaces of below-grade foundations is gamma well logging. Soil in biased locations next to the exterior of the buildings may be evaluated using this technique. This technique can provide for rapid isotopic analysis of soils without sampling.

For basements that are to remain after removal of all except ISFSI areas from the license, an FSS will be performed on the internal surfaces. The results of characterization of the external surfaces will be used in the design and DQOs of the subsurface FSS to be performed except the containment liner in the area.

Basement concrete that is to remain (except the containment liner as discussed above) will be characterized to determine the extent of any volumetric contamination of the concrete in accordance with Table 5-10. The results will be used in the calculation of the Basement Fill Model.

nR Auimust9Af005 5-51 Rev. A

• i 0l;' rW '*

Haddam Neck Plant License Termination Plan 5.7.3.1.2.2 Footings After completion 'of final status survey activities of the remaining portions of structures, some concrete may remain in the form of building'footings.

There are several building foundations that are to remain. The current approach includes the demolition of buildings to four feet below grade. This will remove ground-level floors and portions of footings and foundation supports. Surfaces of these below-grade tstructuies will be evaluated using the historical site assessment and other pertinent records to determine the potential for sub-surface contamination on the surfaces of the foundations. Soil samples will be taken in the vicinity of the footings/foundation. If the soil is found to be free of residual radioactivity at the biased locations, it will be assumed that the exterior surface of the foundation is also free of residual activity. If soil samples contain residual radioactivity, the exterior surfaces will be sampled or assessed as follows:

These footings will be sampled using concrete sampling (at least 3 samples per footing or group of footings) or assessed using the results of nearby soil samples, the following methodology will be used:

• The average of the soil samples result (for plant-related radionuclides detected) will be calculated for each footing (at least 3 samples per footing or group of footings).

• For Tritium:

o Using the soil distribution coefficient data (Kds) shown in Table F-1 and equation 6-1 of the LTP, the groundwater concentration in equilibrium with the soil concentrations in the last bullet will be detennined.

o Using the concrete distribution coefficients shown in Table F-4 ofLTP and equation 6-1 the concrete concentrations in equilibrium with the groundwater concentrations calculated in the last bullet wvill be determined.

o For radionuclides other than H-3:

o At least six pairs of concrete samples with adjacent soil samples *vill be collected at locations affected by plant leakage and/or groundwater contamination.

o The average and %CV (coefficient of variation) of the ratios of concrete concentrations to soil concentrations will be calculated.

o IF the %CV of the data is less than 25%, the average ratio will be used to determine concrete concentrations from adjacent soil'samples.

o If the %CV. is 25% or greater, more samples will be taken until a satisfactory'variance is calculated or the worst case ratio will be used.

The results of this sampling or assessment will then be used as input to the calculation of future groundwater dose using the basement fill model as discussed in Chapter 6.

The results of sampling and/or assessment of the external surfaces will be used in the design and DQOs of the subsurface FSS to be performed in the area. Following any required remediation approximately 4 feet of backfill soil would be placed over the footings to restore the area to grade elevation.

August'2005 5-52 Rev. 3

Haddam Neck Plant License Termination Plan The need for a final status survey of the areas with below grade structures will be determined on a graded approach.

For footings and other structures to remain that have a very low or no potential for contamination such as:

• Buildings outside the RCA

• Building shown by characterization sampling to be free of residual radioactivity The final status survey will consist of a surface and subsurface FSS of the area including the subsurface structures.

5.7.3.1.3 Sewer Systems, Plumbing and Floor Drains Residual radioactivity in sanitary piping or floor drains will be evaluated in the same manner as for non-structural plant systems or components, discussed in Sections 5.4.7.5 and 5.6. Assessment of residual activity levels in piping or floor drains will be via sampling of sediments, fixed measurements, scanning, or a combination of these methods, as appropriate.

All non-RCA sanitary systems at the Haddam Neck Plant drain to on-site leach fields. These systems are independent of other plant systems and all surface water or storm drains. If any residual radioactivity is suspected in portions of the sanitary plumbing systems, evaluations for both the leach fields and the associated system piping may be required. Radiological assessments of piping will be made as described in Section 5.6 of this plan, i.e., by full length surveys of interior surfaces. Evaluations required for any affected leach fields will be made as described in Section 5.7.3.2.2 of this plan, for sub-surface activity.

All operable RCA-located systems currently drain to the aerated drains system and are part of the normal plant effluent. Thus, there is no leach fields associated with these systems. During the plant lifetime, toilet facilities, showers and sinks, contained within the RCA, drained to the plant sanitary system and associated leach field. Any piping associated with the systems, which is proposed to remain following decommissioning will be evaluated as described above.

5.7.3.1.4 Ventilation Ducts - Interiors Radiological assessments of ventilation systems will be made by taking measurements at appropriate access points where activity levels should be representative of those on the interior surfaces. Assessments may also be made using in-situ gamma-spectroscopy provided adequate instrument efficiencies and detection limits can be achieved. Exterior surfaces of such systems will be evaluated as part of the building or structure in cases where the system is attached to it or is otherwise an integral component.

5.7.3.1.5 Piping and Embedded Piping The construction of the Haddam Neck Plant was such that there is not expected to be a significant amount of embedded piping to consider in the final survey effort. Most of the radiologically affected piping is in pipe trenches, and thus can be accessed and removed as necessary. Currently approximately 1000 feet of embedded piping is forecasted to remain after Final Status Survey. Any affected embedded piping remaining at the time of Final Status Survey is expected to be in wall penetrations between areas. Sections of such piping are not expected to be very long (no longer than the wall thickness) and thus should be able to be sampled or surveyed as appropriate to evaluate residual activity levels against the applicable release criteria. The Final Status Survey design of areas containing embedded piping will address this media during the DQO process. Expected outputs of the DQO process include defining the appropriate type of data to collect; survey measurement processes and survey instrument sensitivity; potential contaminants and appropriate DCGL for the assumed exposure pathway.

August 2005 5-53 Rev. 3

Haddam Neck Plant License Termination Plan 5.7.3.1.6 Activated Concrete s.,t K> Although concrete cores have been obtained in Containment, they were not obtained in areas subject to the highest levels of neutron activation. Areas subject to the highest neutron activation are currently inaccessible, and, therefore, specific characterization data is not yet available in all areas. However, neutron activation data from Maine Yankee, Trojan and Yankee Nuclear Power Station indicate that H-3 and Fe-55 are present in the highest concentrations. Other radionuclides such as C-14, Co-60, Eu-152 and Ni-63 are also present. Based upon these data, the activation products Eu-152 and Eu-154 were included in the list of radionuclides expected to be present at HNP (Table 2-12).

As the decommissioning progresses and high dose rate components are removed, additional characterization of structures within Containment, including activated concrete and structural components, will take place. These characterization samples will typically be analyzed by gamma spectroscopy with some samples being analyzed for "hard-to-detect" radionuclides. Therefore, a representative sample of characterization and final status survey samples will be screened for neutron activation.

In-situ gamma spectroscopy may be used to perform remediation surveys for activated concrete to determine the radionuclide concentration to be used in the "Basement Fill Model" calculation. If in-situ gamma-spectroscopy is selected for use, a technical support document will be developed which describes the technology to be used and how the technology meets the objectives of the survey. This document will be available for NRC inspection in support of final status survey activities.

Such surveys would be conducted so that 100% of the affected volume was covered in overlapping measurements. Embedded materials (such as rebar) and activated piping and activated portions of the liner will be treated as concrete for purposes of calculating future groundwater dose with the Basement

<> Fill Model. Assessments for any "hard-to-detect" radionuclides that might be present in activated concrete will be by either direct measurements (core-bores or equivalent) or by establishing surrogate concentrations for these radionuclides relative to some radionuclide easily measured via gamma-spectroscopy (Co-60, for example). Surrogate ratios will be established using pertinent characterization data for the survey unit of interest. Final status surveys of these areas will also include collection and analysis of concrete and rebar samples.

Basements that may contain activated concrete and are to remain will be characterized to determine the extent of any volumetric contamination of concrete in accordance with Table 5-10. The results will be used in the calculation of the Basement Fill Model.

5.7.3.1.7 -Systems and Equipment Interiors and Exteriors Surface activity assessments for non-structural systems and components will be made by making measurements at traps and other appropriate access points where activity levels should be representative of those on the interior surfaces. Assessments may also be made via in-situ gamma-spectroscopy, Aupust

_ 2005 5

_-SSw Rev. 3

Haddam Neck Plant License Termination Plan assessment are greater than 2 standard deviations from the mean, then the investigation will be increased to include a larger physical area than the initial investigation assessment. If the final results of the investigation assessment are statistically different than the radiological assessment results, then a full radiological assessment of the affected bedrock areas will be performed in accordance with Section 5.7.3.

The results of the re-assessment and investigation assessment will be documented and maintained in the bedrock assessment files for the affected bedrock areas.

5.8 Survey Data Assessment The Data Quality Assessment (DQA) process, being adopted at HNP, is an evaluation method used during the assessment phase of FSS to ensure the validity of FSS results and demonstrate achievement of the survey plan objectives. The level of effort expended during the DQA process will typically be consistent with the graded approach used during the DQO process. The DQA process will include a review of the DQOs and survey plan design, will include a review of preliminary data, will use appropriate statistical testing when applicable (statistical testing is not always required, e.g., when all sample or measurement results are less than the DCGLw ), will verify the assumptions of the statistical tests, and will draw conclusions from the data.

Prior to evaluating the data collected from a survey unit against the release criterion, the data are first confirmed to have been acquired in accordance with all applicable procedures and QA/QC requirements.

Any discrepancies between the data quality or the data collection process and the applicable requirements are resolved and documented prior to proceeding with data analysis. Data assessment will be performed, by trained personnel, using approved site procedures.

The first step in the data assessment process is to convert all of the survey results to DCGL units. Next, the individual measurements and sample concentrations will be compared to DCGL levels for evidence of small areas of elevated activity or results that are statistical outliers relative to the rest of the measurements (see Section 5.5.3.1). Graphical analyses of survey data that depict the spatial correlation of the measurements are especially useful for such assessments and will be used to the extent practical.

The results may indicate that additional data or additional remediation and resurvey may be necessary. If this is not the case, the survey results will then be evaluated using direct comparisons or statistical methods, as appropriate, to determine if they exceed the release criterion. If the release criterion has been exceeded or if results indicate the need for additional data points, appropriate further actions will then be determined.

Interpreting the results from a survey is most straightforward when all measurements are higher or lower than the DCGLW. In such cases,' the decision that a survey unit meets or exceeds the release criterion requires little in terms of data analysis. However, formal statistical tests provide a valuable tool when a survey unit's measurements are neither clearly above nor entirely below the DCGLw.

The first step in evaluating the data for a given survey unit is to draw simple comparisons between the measurement results and the release criterion. The result of these comparisons will be one of three conclusions: 1) the unit meets the release criterion; 2) the unit does not meet the release criterion; or 3) no conclusion can be drawn from simple comparisons and thus one of the non-parametric statistical tests must be applied. The initial comparisons made for the results for a given survey unit depend on whether or not the results are to be compared against a background reference area.

If the survey data are in the form of gross (non-radionuclide-specific) measurements or if the radionuclide of interest is present in background in a concentration that is a relevant fraction of the DCGLw, then the initial data evaluation will be as described in Table 5.12.

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Haddam Neck Plant License Termination Plan Table 5-12 I Initial Evaluation of Survey Results (Background Reference Area Used)

Evaluation Result Conclusion Difference between the maximum concentration Survey unit meets the release criterion measurement for the survey unit and the minimum reference area concentration is less than the DCGLw Difference between the average concentration Survey unit does not meet the release criterion measured for the survey unit and the average reference concentration is greater than the DCGLw Difference between any individual survey result and Conduct either the Wilcoxon Rank Sum test any individual reference area concentration is greater or the Sign test; and the EMC test than the DCGLw and the difference between the average concentration and the average for the reference area is less than the DCGLw If the survey data are in the form of radionuclide-specific measurements and the radionuclide(s) of interest is not present in background in a concentration that is a relevant fraction of the DCGLw, then the initial data evaluation will be as described in Table 5-13.

Table 5-13 Initial Evaluation of Survey Results (Background Reference Area Not Used)

Evaluation Result Conclusion All measured concentrations less than the Survey unit meets the release criterion DCGLw Average concentration exceeds the DCGLw Survey unit does not meet the release criterion Individual measurement result(s) exceeds the Conduct the Sign test and the EMC test DCGLW and the average concentration is less than the DCGLw __

5.8.1 Wilcoxon Rank Sum Test Gross activity measurements or measurements for which the radionuclide of interest exists in background in concentrations that are a relevant fraction of the DCGLW may be evaluated using the Wilcoxon Rank Sum (WRS) test. In the WRS test, comparisons are made between the survey results for a given survey unit and reference (background) data for comparable materials. However, for survey units which contain multiple materials having different backgrounds, it may be advantageous to background-subtract gross activity measurements (using paired observation) and apply the Sign test (see Section 5.8.2).

The WRS test tests the null hypothesis that the median concentration in the survey unit exceeds that in the reference area by more than the DCGLW. The null hypothesis is assumed to be true unless the statistical August 2005 5-60 Rev. 3 l

Haddam Neck Plant License Termination Plan 6 COMPLIANCE WITH TIE RADIOLOGICAL CRITERIA FOR LICENSE TERMINATION 6.1 Site Release Criteria 6.1.1 Radiological Criteria for Unrestricted Use The site release criteria for the Haddam Neck Plant (HNP) site will correspond to the radiological criteria for unrestricted use given in 10 CFR 20.1402 (Reference 6-1):

• Dose Criterion: The residual radioactivity that is distinguishable from background radiation results in a Total Effective Dose Equivalent (TEDE) to an average member of the critical group that does not exceed 25 mrem/year, including that from groundwater sources; and

• ALARA Criterion: The residual radioactivity has been reduced to levels that are As Low As Reasonably Achievable (ALARA).

6.1.2 Conditions Satisfying the Release Criteria Levels of residual radioactivity that correspond to the allowable radiation dose and ALARA levels described above are calculated by analysis of various scenarios and pathways (e.g., direct radiation,

<2~" inhalation, ingestion) through which exposures could be reasonably expected to occur. LTP Section 2.3.3.4 discusses the radionuclides for which Derived Concentration Guideline Levels (DCGLs) and the future groundwater dose must be calculated. These DCGLs and the future groundwater dose calculation methodology form the basis for the following conditions which, when met, satisfy the site release criteria as prescribed in 10 CFR 20.1402:

• The average residual radioactivity in soils, standing above grade buildings and existing groundwater above background is less than or equal to the applicable combined DCGLs.

• In the case of buried concrete, embedded piping and the below-grade containment liner, the "future groundwater" dose will be determined using the "Basement Fill" Model. This approach will ensure that the dose from all pathways will be less than the release criteria of 10CFR20.1402.

The details of this model are presented in Section 6.8.2

• Individual measurements, representing small areas of residual radioactivity, which exceed the DCGL, do not exceed the elevated measurement comparison DCGLEMc. The use of the DCGLEMc is described in Section 5.4.7.4.

• Where one or more individual measurements exceed the DCGL, the average residual radioactivity passes the Sign or Wilcoxon Rank Sum (WRS) statistical test. (See Section 5.8 for a detailed discussion application of statistical tests).

• Remediation is performed where it is ALARA to reduce the levels of residual radioactivity to

~

V mbelow those concentrations necessary to meet the DCGL or in the case of the Basement Fill August 2005 6-1 Rev. 3

Haddam Neck Plant License Termination Plan Model, below the future groufidwater dose calculated by that model; (See Section 4 and Appendix B for detailed discussions of ALARA considerations).

The methods in MARSSIM (Reference 6-2) and the DCGLs may not be appropriate for complex non-structural components. For those non-structural components and systems to which MARSSIM does not apply (with the exception of those cases discussed in Section 5.4.7.5), site unconditional release limits apply (i.e., no detectable radioactive material). These surveys will be performed in accordance with health physics procedures and are consistent with the requirements of NRC Information Notice 85-92, "Surveys of Wastes before Disposal from Nuclear Reactor Facilities", and 1E Circular 81-07, "Control of Radioactively Contaminated Material."

As MARSSIM does not define a protocol for performing volumetric contamination sampling, to be used to calculate the future groundwater dose, protocols for this sampling is defined in Section 5.7.1.5.

CYAPCO will not use concrete debris from the demolition of buildings to backfill basements that remain after release of the buildings from the license. As described in Chapter 1, demolition debris will be shipped off-site to an appropriate disposal facility. The remaining basements will be backfilled with soil from off-site locations. This backfill soil has been demonstrated to be free from plant related radioactivity over background. Due to this change in the decommissioning strategy, the Concrete Debris Scenario is no longer applicable to building basements. In order to calculate future groundwater dose that was previously calculated using the Concrete Debris DCGLs for basements, the Basement Fill model described in Section 6.8.2 will be used. Should the decommissioning plans at CYAPCO change to include the use of concrete debris for basement fill; the Concrete Debris DCGLs developed and approved as part of the LPT approved in November 2002 will be used to demonstrate compliance.

6.2 Site Characteristics The following is a description of the physical, geologic, and hydrogeologic characteristics of the area and the relationship of these characteristics to contaminant source areas and potential pathways.

Physical Characteristics The industrial area of the HNP site is located on the east bank of the Connecticut River on a level, 600 ft wide terrace at an elevation of 21 ft mean sea level (msl). A parking lot occupies the area to the north of the industrial area. The area north of the parking lot is occupied by a pond. To the south, a 5,500 foot-long cooling water discharge canal leads to the river from the southern edge of the industrial area. It is separated from the Connecticut River by a 200 to 1,000 ft wide peninsula flood plain that ranges in elevation from about 5 to 15 ft msl. A steep wooded hill slope rises immediately east of the industrial area to elevations over 300 ft msl. The lowermost 30 to 40 ft of the hillside adjacent to the plant consists of nearly vertical rock cut.

Geologic and Hydrogeologic Characteristics The geology and hydrogeology of the industrial area is documented in the "Groundwater Monitoring Report" Malcolm Pirnie (1999) (Reference 6-3). Drawings depicting geologic and hydrogeologic characteristics are given in Figures 6-1 and 6-2. A brief discussion of the site characteristics is provided below. Note: As discussed in Chapter 2, the following information was current as of August of 2002.

This information has been and will continue to be updated in correspondence with the CT DEP concerning the Phase 2 Hydrogeologic Investigation Work Plan. As the NRC receives copies of all of this correspondence, the information in the LTP will not generally be updated.

August 2005 6-2 Rev. 3

Haddam Neck Plant License Termination Plan The topography of this area originally consisted of a north-south trending promontory approximately 400 ft wide that connected the steep hillside north of this area to a floodplain terrace along the river's edge. The steep hill slope extended southward to the northeastern most third of the Containment Building.

The southern part of the promontory consisted of large bedrock outcroppings in the area of the turbine building. Wetlands extended for 1,000 ft or more to the northwest and southeast of the promontory.

During construction of the HNP, the steep hill slope to the north and the higher portions of the promontory were cut and the adjacent wetlands were filled. The discharge canal was excavated through the wetland, terrace, and floodplain to the southeast. The subsurface portions of the Containment Building, Primary Auxiliary Building (PAB), Turbine Building, Discharge Tunnel, and Spent Fuel Pool were also excavated down to or below the original bedrock surface.

On either side of the bedrock promontory and on the peninsula are seven layers of unconsolidated sediments: artificial fill, wetland silt and organic matter, gray silt and fine-grained sand (alluvium),

gravelly sand, red fine-grained sand, brown sand, and glacial till or cobble gravel. The sediment thickness below the industrial area averages less than 20 ft but increases southeastward to over 100 ft beneath the peninsula.

Bedrock fractures are visible on the hill slope and potentially project into the industrial area. These fractures may be preferential pathways for groundwater migration within the bedrock. The bedrock itself consists of a suite of recrystallized volcanic rocks mapped regionally as the Monson Gneiss and Middletown Formation. These rocks are made of various silicate minerals (quartz, plagioclase, biotite, homblende, pyroxene, etc.) with essentially no porosity other than fractures.

The shallow groundwater flow beneath the industrial area occurs within the unconsolidated sediments and bedrock. The depth to the water table averages about 10 ft below ground surface (bgs) in this area.

<_y Groundwater generally flows southwest and downward near the hill slope, and upwards near the discharge canal and the Connecticut River. Locally, the Containment Building and mat drain sump are important hydrogeologic features. The groundwater flow pattern around the Containment Building was distorted with a component of flow toward the drainage system under the Containment Building. The mat drain sump, located on the southern side of the Containment Building, when operated, removed groundwater and depressed the water table around it. The pumps were shut off for several months but have been restarted. The cooling water discharge tunnels divert the shallow groundwater flowing around the southwestern side of the Containment Building farther to the south. Southwest of tunnels, the shallow groundwater appears to flow southwesterly and directly toward the river.

Contaminant Characteristics Soil within the industrial area contains residual radioactivity from licensed operations by unplanned liquid releases or long-term accumulation of material in the soil via effluent releases. The impacted soil includes that in current open areas as well as that which will be exposed in the future following demolition of overlying buildings and structures. The areas wherein soil could potentially contain residual radioactivity are identified and described in Chapter 2. Based on the documented release mechanisms and the results of site characterization surveys, the residual radioactivity is generally confined to the surface soil layer, although some subsurface residual radioactivity exists. The surface soils in the industrial area are composed of a silty sand that was imported as artificial fill. Site survey results indicate that there may be localized areas where the soil contamination is deeper, but still restricted to the unsaturated zone.

Site surveys have identified radionuclides that may be present in measurable quantities in site soils and that are likely associated with licensed plant operations. Table 2-12 summarizes these radionuclides and their half-lives.

August 2005 6-3 Rev. 3

Haddam Neck Plant License Termination Plan 6.3 Dose Modeling Approach 6.3.1 Overview To calculate DCGLs, dose models were developed, which translate levels of residual radioactivity into potential radiation doses to the public. Dose models, appropriate to the HNP site, are based on the guidance found in DG-4006 (Reference 6-4), NUREG-1549 (Reference 6-5), and NUREG/CR-5512, Volume 1 (Reference 6-6). A conceptual model was based on the site conditions expected at the time of unrestricted release. Conditions at the HNP site (e.g., pre-existing residual radioactivity in groundwater) required site-specific dose modeling be performed. The approach taken to dose modeling for the HNP site is consistent with the information provided in Chapter 5 and Appendix C of NUREG-1727 (Reference 6-7) for-site specific modeling, including the information regarding source term abstraction and scenarios, pathways, and critical groups.

In addition to calculating DCGLs, a "Basement Fill Model" will be used to determine the future groundwater dose from building basements and other subsurface materials on future uses of the site. This method uses actual characterization data (to determine the radionuclide inventory) and the calculated release rate of the radionuclides from the material to calculate the equilibrium maximum groundwater concentration that will result between back-fill soil and groundwater in the building basements. The future groundwater dose is calculated from the groundwater concentration and the groundwater DCGLs for these sources. This model is explained further in Section 6.8.2.

The dose model is defined by the three factors: 1) the scenario, 2) the critical group and 3) the exposure pathways. The scenarios described in NUREG/CR-5512, Volume 1, address the major exposure pathways of direct exposure to penetrating radiation and inhalation and ingestion of radioactive materials. The K> scenarios also identify the critical group. The critical group is the group of individuals reasonably expected to receive the greatest exposure to residual radioactivity within the assumptions of the particular scenario. The scenarios and their modeling are specifically designed to be reasonably conservative by generally overestimating rather than underestimating potential dose.

The approach outlined above was used to develop dose models to calculate DCGLs for the following media:

• Soil

• Groundwater, and

• Concrete

-Buildings Standing It should be noted that the scenarios described in NUREG/CR-5512, Volume 1, were developed to estimate potential radiation doses from radioactive material in standing buildings and soil. These scenarios were adapted to estimate potential radiation doses from groundwater.

6.3.2 Resident Farmer Scenario Scenario Definition:

The resident farmer scenario, described as the "Residential Scenario" in NUREG/CR-5512, Volume 1, was selected to estimate human radiation exposures resulting from residual radioactivity in the soil and groundwater to detennine corresponding DCGLs.

August 2005 64 Rev. 3

Haddam Neck Plant License Termination Plan Critical Group:

K> Given regional demographic and economic data (References 6-8 and 6-9) the average member of the critical group was determined to be the resident farmer who lives on the plant site following decommissioning, grows all or a portion of his/her diet on site, and uses the water from a groundwater source on the site for drinking water and irrigation. The dose from residual radioactivity in the soil, and groundwater is evaluated for the average member of the critical group as required by 10 CFR Part 20, Subpart E, and described in NUREG-1727, Appendix C and NUREG-1549.

It is unlikely that any other set of plausible human activities could occur onsite that would result in a dose exceeding that calculated for the hypothetical resident farmer. It is more likely that the behavior of future occupants would result in a lower dose. For example it is more likely that the HNP site (currently zoned "industrial") will be reused for a fossil-fired plant, making use of the current infrastructure, or for land conservation. The hypothetical dose from the soil to individual in these settings would be less than for a resident farmer, since the industrial worker would not ingest food derived from onsite. Therefore, the use of the resident farmer as the average member of the critical group is both conservative and bounding for the calculation of soil DCGLs.

Exposure Pathways:

The potential exposure pathways that apply to the resident farmer are listed below and depicted in Figure 6-3. These exposure pathways are based upon those in NUREG/CR-5512, Volume 1:

e Direct exposure to external radiation from the residual radioactivity; v Internal dose from inhalation of airborne radionuclides; and

• Internal dose from ingestion of

- Plant foods grown in media containing residual radioactivity and irrigated with water containing residual radioactivity,

- Meat and milk from livestock fed with fodder grown in soil containing residual radioactivity and water containing residual radioactivity,

- Drinking water (containing residual radioactivity) from a well,

- Fish from a pond containing residual radioactivity, and

- Media containing residual radioactivity.

6.3.3 Building Occupancy Scenario Scenario Definition:

The building occupancy scenario, based upon NUREG/CR-5512, Volume 1, was selected to estimate human radiation exposure resulting from residual radioactivity in concrete from standing buildings and building foundations/basements that could reasonable be occupied and to determine corresponding DCGLs. CYAPCO will not leave any basements in place that can be reasonably occupied. These DCGLs are also used to bound residual contamination levels on metal surfaces such as the containment liner and embedded piping to be subsequently used in the calculation of the "future groundwater" dose due to contamination on these metal surfaces.

Rev. 3 August 2005 August 2005 6-5 Rev. 3

Haddam Neck Plant License Termination Plan Critical Group:

K<_i Given the fact that the buildings associated with the HNP site are commercial, the average member of the critical group is an adult engaging in light industrial work within the buildings following decommissioning of the site. He/she occupies a commercial facility in a normal manner without deliberately disturbing sources of residual radioactivity. The dose from residual radioactivity in the concrete from the standing building is evaluated for the average member of the critical group as required by 10 CFR Part 20, Subpart E, and described in NUREG -1727, Appendix C.

Exposure Pathways:

The potential exposure pathways, described in NUREG/CR-55 12, Volume 1, are depicted on Figure 6-4 and listed below:

• Direct exposure to external radiation from

- Source

- Material deposited on the floor

- Submersion in airborne dust

• Internal dose from inhalation of airborne radionuclides

• Internal dose from inadvertent ingestion of radionuclides from the source 6.4 RESidual RADioactivity (RESRAD) and RESRAD-BUILD Codes The RESRAD family of computer codes is pathway analysis models developed at Argonne National Laboratory (ANL). This family of computer codes includes RESRAD, used to analyze pathways K>J associated with soil, and RESRAD-BUILD, used to analyze pathways associated with buildings.

The RESRAD computer code (Version 5.91) was used in this analysis to consider three major exposure pathways to a resident farmer from residual radioactivity in soil and groundwater.

• Direct exposure to external radiation from media containing residual radioactivity;

• Internal exposure from inhalation of airborne radionuclides; and

• Internal exposure from ingestion of radionuclides.

A newer version of the code released by ANL is RESRAD Version 6.1. This version of the code includes probabilistic modules to examine the sensitivity of input parameters on the resulting dose. A sensitivity analysis has been performed using the probabilistic modules in RESRAD 6.1. Information obtained from that analysis (identification of sensitive parameters and their correlation to dose, either positive or negative) is then used to select conservative values for the sensitive input parameters for the deterministic runs using RESRAD Version 5.91.

The RESRAD-BUILD computer code (Version 2.37) is used in this analysis to consider five exposure pathways to occupants of a building from residual radioactivity for above-grade building surfaces:

• External exposure directly from the sources;

• External exposure to material deposited on the floor;

• External exposure due to air submersion; August 2005 6-6 Rev. 3

Haddam Neck Plant License Termination Plan

• Inhalation of airborne radioactive particulates; and

• Inadvertent ingestion of radioactive material directly from the sources.

ANL has released a newer version of the code, RESRAD-BUILD Version 3.1. This version of the code includes probabilistic modules to examine the sensitivity of input parameters on the resulting dose. A sensitivity analysis has been performed using the probabilistic modules in RESRAD-BUILD Version 3.1.

Information obtained from that analysis (identification of sensitive parameters and their correlation to dose, either positive or negative) was then used to select conservative values for the input parameters for the deterministic runs using RESRAD-BUILD Version 2.37.

For subsurface structures, an inventory-based method for assessing total radioactivity in the subsurface environment is provided. This total inventory is converted into a future groundwater dose using the "Basement Fill Model" and then converted to a future groundwater dose using the groundwater DCGLs as described above. Therefore, the process of parameter selection and sensitivity analysis does not apply to this model.

Information on the use of the codes and their applications are outlined in NUREG/CRs-6676, -6692,

-6697 (References 6-10, 6-11, and 6-12), the "Users Manual for RESRAD, Version 6.0" (Reference 6-13), the "Manual for implementing Residual Radioactive Material Guidelines Using RESRAD, Version 5.0" (Reference 6-14) and for RESRAD-BUILD "A Computer Model for Analyzing the Radiological Doses Resulting from the Remediation and Occupancy of Buildings Contaminated with Radioactive Material" (Reference 6-15).

6.5 Parameter Selection Process The conceptual model underlying the dose model was developed based on site characteristics expected at the time of release of the site. The conceptual model is quantified by a set of parameters. The parameter selection process is outlined schematically in Figure 6-5. The process was developed in accordance with the guidelines presented in NUREG/CR-6755 (Reference 6-16), -6676, -6692 and -6697 and ensures that conservative values are selected. Components of the selection process are discussed in the following sub-sections.

6.5.1 Classification The parameters were classified as behavioral, metabolic or physical consistent with NUREG/CR-6697, Attachment A. Behavioral parameters depend on the behavior of the receptor and the scenario definition.

Metabolic parameters represent the metabolic characteristics of the receptor and are independent of the scenario definition. Physical parameters are the parameters that would not change if a different group of receptors were considered.

6.5.2 Prioritization The parameters were prioritized in order of importance consistent with NUREG/CR-6697, Attachment B.

Prioritization was based on 1) the relevance of the parameter in dose calculations, 2) the variability of the dose as a result of changes in the parameter value, 3) the parameter type and 4) the availability of parameter-specific data. Priority 1parameters are considered to be high priority; priority 2 parameters are considered to be medium priority; and priority 3 parameters are considered to be low priority.

August 2005 6-7 Rev. 3

Haddam Neck Plant License Termination Plan 6.5.3 Treatment KJ.} The parameters were treated as "deterministic" or "stochastic" depending on parameter type, priority, and availability of site-specific data and the relevance of the parameter in dose calculations. "Deterministic" modules of the code use single values for input parameters and generate a single value for dose.

"Probabilistic" versions of the code use probability distributions for input parameters and generate a range of doses. "Stochastic" parameters are parameters that are defined by a probabilistic distribution.

The behavioral and metabolic parameters were treated as deterministic. The physical parameters for which site-specific data were available were also treated as deterministic. The remaining physical parameters for which no site-specific data were available to quantify were classified as either priority 1,2, or 3. Priority 1 and 2 parameters were treated as stochastic. The priority 3 physical parameters were treated as deterministic.

6.5.4 Sensitivity Analyses The purpose of the sensitivity analysis was to determine which of the stochastic parameters have the greatest influence on the resultant dose and associated DCGLs. The analysis was performed using the probabilistic modules of RESRAD, Version 6.1, and RESRAD-BUILD, Version 3.1.

The stochastic parameters were generally assigned distribution types and corresponding distribution statistical parameters from NUREG/CR-6697, Attachment C. Sensitivity analyses were performed on the stochastic parameters using the assigned distributions. To perform the sensitivity analysis the following information was required:

Sample Specifications: The analyses were run using 300 observations and 1 repetition. The Latin Hypercube Sampling (LHS) technique was used to sample the probability distributions for each of the stochastic input parameters. The correlated or uncorrelated grouping option was used to preserve the prescribed correlations Input Rank Correlations: Correlation coefficients were assigned between correlated parameters.

Output Specifications: All of the output options were specified.

Sensitivity analyses were performed for each of the radionuclides. The Partial Rank Correlation Coefficient (PRCC) for the peak of the mean dose was used as a measure of the sensitivity of each parameter to the peak of the mean dose.

For the resident farmer scenario, a parameter was identified as sensitive if the absolute value of its PRCC (IPRCCI) was greater than or equal to 0.25 and non-sensitive if the IPRCCI value was less than 0.25. For the building occupancy scenario, a parameter was identified as sensitive if the IPRCCI value was greater than or equal to 0.10 and non-sensitive if the IPRCCI value was less than 0.10. These thresholds were selected based on the guidance included in NUREG/CR-6676.

6.5.5 Parameter Value Assignment The behavioral and metabolic parameters were assigned values from NUREG/CR-5512, Volume 3, NUREG/CR-6697, or the RESRAD default library.

Physical parameters were assigned values as follows:

August 2005 6-8 Rev. 3

Haddam Neck Plant License Termination Plan

• Physical parameters for which site-specific data were available were assigned site-specific values.

• Priority 1 and 2 physical parameters shown to be sensitive (IPRCCI > 0.25) were assigned conservative values. Depending on whether the parameter was positively or negatively correlated with dose, the 75% or 25% quantile value of the distribution was used, respectively. The mean value of the distribution was also calculated for those parameters positively correlated with dose.

If the mean value was greater than the 75% quantile value (positively skewed distribution), the parameter was assigned the mean value.

• Priority 1 and 2 physical parameters shown to be non-sensitive (IPRCCI < 0.25) were assigned median values from NUREG/CR-6697, Attachment C.

• Priority 1 and 2 physical parameters shown to be non-sensitive (IPRCCI < 0.25) but correlated with a physical parameter shown to be sensitive (see Section 6.5.4) were assigned values based on the conservative value assigned to the sensitive parameter.

• Priority 3 physical parameters were assigned values from NUREG/CR-55 12, Volume 3, or from the RESRAD default library.

6.6 DCGLs for Soil Residual radioactive material is considered to exist in soil underlying portions of the HNP site. The residual radioactivity is considered to be from licensed operations by unplanned liquid releases or long-term accumulation of material in the soil via effluent releases. The affected areas are generally confined to the industrial area of the site and include areas that are currently open and areas that may be open following the demolition of buildings and structures.

6.6.1 Dose Model The DCGLs for soil were calculated using the resident farmer scenario. The residual radioactive materials were assumed to be contained in a soil layer (surface and subsurface) on the property that can be used for residential and light farming activities. The average member of the critical group is the resident farmer that lives on the plant site, grows all or a portion of his/her diet onsite, drinks water friom a groundwater source onsite.

The potential pathways used to estimate human radiation exposure resulting from residual radioactivity in the soil include the following:

• Direct exposure to external radiation from soil containing residual radioactivity;

• Internal dose from inhalation of airborne radionuclides; and

• Internal dose from ingestion of:

- Plant foods grown in the soil material containing residual radioactivity;

- Meat and milk from livestock fed with fodder grown in soil containing residual radioactivity and water containing residual radioactivity;

- Drinking water containing residual radioactivity from a well,

- Fish from a pond containing residual radioactivity, and

- Soil containing residual radioactivity.

August 2005 6-9 Rev. 3

` Haddam Neck Plant License Termination Plan 6.6.2 Conceptual Model  ;,

K> The conceptual model underlying the dose model includes a contaminated zone, an unsaturated zone, and a saturated zone. The contaminated zone is exposed at the ground surface (no cover). Residual radioactivity is confined to the soils in the contaminated zone. The thickness of the contaminated zone is conservatively set at 3 meters. For the purpose of calculating soil DCGLs, the groundwater is assumed to be initially uncontaminated.

The parameters used to quantify the conceptual model are listed in Appendix D, Table D-1. The values!

distributions assigned to each of the parameters and the basis for assigning such values/distributions are shown on the table.

6.6.3 Sensitivity Analysis Results Parameter distributions assigned in the probabilistic RESRAD, Version 6.1, model is presented in Appendix D, Table D-1. An initial radionuclide concentration of I pCi/g was used for the soil comprising the contaminated zone.

The stochastic parameters identified as sensitive (IPRCCI 2 0.25) to the peak of the mean dose for each of the radionuclides are presented in Appendix E, Table E-1. For each radionuclide, the sensitive parameters are listed in order of decreasing sensitivity. Included in the table are the conservative values assigned to each of the sensitive parameters.

6.6.4 DCGL Determination Parameter values assigned in the deterministic RESRAD Version 5.91 model are presented in Tables E-1 K> (conservative values assigned to parameters shown to be sensitive) and Appendix F, Table F-i. The soil DCGLs were determined for a radiation dose limit of 25 mnrem/yr.

The soil DCGLs calculated for each of the radionuclides are presented in Appendix G, Table G-1. The time to the peak of the mean dose is also included on the table together with the percent contribution to dose from the exposure pathways (water independent and water dependent).

The soil DCGLs are summarized in Table 6-i:

August 2005 6-10 Rev. 3 l

"a.

Haddam Neck Plant Lien'se Termination Plan Table 6-1 Base Case DCGLs for Soil Radionuclide Soil DCGL (pCi/g)

H-3 4.12E+02 I C-14 5.66E+OO Mn-54 1.74E+O1 Fe-55 2.74E+04 Co-60 3.81E+OO Ni-63 7.23E+02 Sr-90 1.55E+OO Nb-94 7.12E+OO Tc-99 1.26E+Ol Ag-108m 7.14E+OO Cs-134 4.67E+OO Cs-137 7.91E+OO Eu-152 l.O1E+O1 Eu-154 9.29E+OO Eu-155 3.92E+02 Pu-238 2.96E+Ol Pu-239 2.67E+O1 Pu-241 8.70E+02 Am-241 2.58E+Ol Cm-243 2.90E+O1 6.7 DCGLs for Groundwater Residual radioactivity presently exists in groundwater underlying portions of the HNP site. The affected areas are generally confined to the industrial area of the site, as investigated by Malcolm Pirnie (Reference 6-3).

6.7.1 Dose Model The resident farmer scenario was selected to estimate human radiation exposures resulting from residual radioactivity in the groundwater and to determine corresponding DCGLs. The residual radioactive materials are assumed to be contained in the groundwater on the property, which is withdrawn via a groundwater source (well) and used for irrigation and drinking water. The average member of the critical group is the resident farmer who lives on the plant site, grows all or a portion of his/her diet onsite, and I drinks water from the groundwater source onsite.

The potential pathways used to estimate human radiation exposure resulting from residual radioactivity in the groundwater include the following ingestion pathways:

Plant foods irrigated with water containing residual radioactivity; Meat and milk from livestock fed with water containing residual radioactivity; and 0 Drinking water containing residual radioactivity from a well.

August 2005 6-11 Rev. 3

Haddam Neck Plant License Termination Plan Groundwater flow directions determined by Malcolm Pimie (1999) are such that the existing plumes k<,, migrate toward the Connecticut River. The flow rate of groundwater, potentially containing residual radioactivity, relative to the flow rate of the Connecticut River is likely very small. Therefore, the aquatic foods ingestion pathway is not considered applicable in this calculation.

6.7.2 Conceptual Model The conceptual model underlying the dose model was developed based on the site characteristics expected at the time of release of the site. The model assumes that the groundwater contains residual radioactivity at the time of site release and that all sources that contributed to this contamination have since been removed. It is further assumed that the residential farmer installs a well that supplies water for drinking, crop irrigation, and livestock, and that this well is drilled and completed within a portion of the groundwater system that contains residual radioactivity.

The parameters used to quantify the model are presented in Appendix D, Table D-2. The values /

distributions assigned to each of the parameters, the basis for assigning such values / distributions and the relevance of the parameters to the dose calculations are included in the table.

The RESRAD code is typically used to calculate radiation doses (and DCGLs) for a source above the water table. To develop a dose model consistent with the conceptual model, it was necessary to establish the parameters below as follows:

- Time since placement of material = 1 year

- Time for calculations = 1 year

- Model for water transport parameters = Mass Balance (MB) model

< - Distribution coefficient in the saturated zone =0 cm3 /g By doing so, the groundwater (well water) concentrations calculated by RESRAD were found to be greater than or equal to the groundwater concentrations in equilibrium with the contaminated zone, under saturated conditions, and the time to the peak of the mean dose was 0 years.

The equilibrium groundwater concentration associated with the contaminated zone was calculated using the principals of linear sorption theory described in Appendix H of the "Users Manual for RESRAD Version 6.0," from which the following equation was derived:

1000 SoPb C= (Equation 6-1)

[l+(Kd pb/n)] n where, C = Equilibrium groundwater concentration (pCi/l)

SO = Initial principal radionuclide concentration in contaminated zone (pCi/g)

Pb = Bulk density of contaminated zone (g/cm 3 )

Kd = Distribution Coefficient of contaminated zone (cm 3/g) n = Total porosity of contaminated zone (Fraction)

August 2005 6-12 Rev. 3

Haddam Neck Plant License Termination Plan 6.7.3 Sensitivity Analysis Results Parameter distributions assigned in the probabilistic RESRAD Version 6.1 model are presented in Appendix D, Table D-2. An initial raidionuclide concentration of 1pCi/g was used for the soil comprising the contaminated zone.

The stochastic parameters identified as sensitive (lPRCCI 2 0.25) to the peak of the mean dose for each of the radionuclides are presented in Appendix E, Table E-2. For each radionuclide, the sensitive parameters are listed in order of decreasing sensitivity. Included in Table E-2 are the conservative values assigned to each of the sensitive parameters.

6.7.4 DCGL Determination Parameter values assigned in the deterministic RESRAD Version 5.91 model are presented in Table E-2, (conservative values assigned to parameters shown to be sensitive) and Appendix F, Table F-2. The groundwater DCGLs were determined for a radiation dose limit of 25 mrem/yr. The groundwater DCGLs were calculated by scaling the groundwater (well water) concentrations calculated by RESRAD against the peak dose to determine the concentration that would give a radiation dose of 25 mrem/yr, as shown in the following equation:

Concww DCGLGW =

• 25 (Equation 6-2)

DosePEAK where, DCGLW DCGL for groundwater (pCi/1)

Concww = Groundwater (well water) (pCi/l)

DosePEAK = Peak Dose (mrem/yr) 25 = Radiation dose limit of 25 mrem/yr The above derivation of the groundwater DCGLs is applicable to the radionuclides that do not have progeny, as the peak dose occurs at the time of release of the site. For the radionuclides that have progeny, the above derivation is not applicable, as the peak dose may occur subsequent to release of the site, due to the in-growth of progeny, and therefore contributions to dose, with time. For these' radionuclides,(Eu-152, Pu-238, Pu-239, Pu-241, Am-241 and Cm-243), the groundwater DCGLs were calculated by modeling the decay of a unit source over 1000 yrs in RESRAD, Version 5.91.

In a "new file" in RESRAD, Version 5.91, the parameters from the RESRAD default library were established, together with the lpCi/g (the units lpCi/g or lpCi/l are arbitrary since the only interest is the decay of a unit source) and 1000-year calculation time. A couple of other parameters were established (the precipitation was set to zero and the water-dependent pathways were toggled off) to ensure the model operated as a "closed" system. The resulting concentrations of the parent radionuclides and progeny, as a function of decay time, are presented in Appendix G, Table G-2-1.

The concentrations of the parent radionuclides and progeny, calculated by RESRAD using an initial concentration of lpCi/g, were converted into dose, outside of RESRAD, by multiplying by effective Dose Conversion Factors (DCFeff's). The DCFff's were calculated outside of RESRAD using the peak dose and groundwater (well water) concentrations for the parent radionuclides. For the progeny, the dose and groundwater concentrations were obtained by re-running the progeny as parent radionuclides in the groundwater model in RESRAD, Version 5.91. The DCFeff's were calculated as shown in the August 2005 6-13 Rev. 3

-HaddamNeck Plant License Termination Plan Equation 6-4 below. Calculation of the DCFeff's for earch of the radion'uclides is shown in Table G-2-1, Footnote "C".

DOScPEAK DCF.ff: =

(Equation 6-3)

Concww where, DCFff = Effective Dose Conversion Factor (mrem/yr/pCi/I)

DOSePEAK = Peak Dose at time = O yrs (mrem/yr)

Concww = Groundwater (well water) (pCi/l)

Following the calculation of dose (using the DCF~fr's), the peak dose was calculated outside of RESRAD by summing the individual doses for the parent radionuclides and progeny ("total" dose) and identifying the highest dose, as shown (in bold) on Table G-2-1. The groundwater DCGLs were calculated outside of RESRAD by scaling the peak dose (based on a radionuclide concentration of lpCi/l) to obtain a concentration based on a radiation dose limit of 25 mrem/yr. These groundwater DCGL's are presented in Appendix G, Table G-2-1.

The groundwater DCGLs for each of the radionuclides are presented in Appendix G,Table G-2-2, with the percent contribution to dose from the exposure pathways (water dependent). Included in Table G-2-2 are the equilibrium groundwater concentrations associated with the contaminated zone and the groundwater (well water) concentrations for a known concentration of radioactive material in the contaminated zone for each of the radionuclides.

The groundwater DCGLs are summarized in Table 6-2.

Table 6-2 Base Case DCGLs for Groundwater Radionuclide Groundwater DCGL

. (pCi/)

H-3 6.52E+05 C-14 9.01E+03 Mn-54 2.42E+04 Fe-55 6.54E+04 Co-60 1.14E+03 Ni-63 3.15E+04 Sr-90 2.51E+02 Nb-94 6.75E+03 Tc99. 2.64E+04 Ag-108m 4.24E+03 Cs-134 3.42E+02 Cs-137 4.3 lE+02 Eu-152 7.33E+03 Eu-154 5.05E+03 Eu-155 3.25E+04 Pu-238 1.51E+01 Pu-239 1.36E+01 Pu-241 4.60E+02 Am-241 1.32E+01 Cm-243 1.94E+01 Rev. 3 6-14 August 2005 August 2005 6-14 Rev. 3

Haddam Neck Plant License Termination Plan 6.8 DCGLs for Concrete A few of the building basements and footings, at HNP will remain in pilake and be surveyed or assessed for residual radioactivity. Presently CYAPCO's plan is to backfill these partial structures with clean material, once the final status survey or assessment of the structure has been completed. The site dose contribution from these basements will be calculated using the Basement Fill Model as discussed in Section 6.8.2. For buildings to remain after release from the license, the Building Occupancy DCGLs will be used.

6.8.1 DCGLs for Concrete: Buildings Standing 6.8.1.1 Dose Model The DCGLs for building surfaces were calculated using the building occupancy scenario. These DCGLs will be applied during the final status survey of:

• any buildings to remain standing following release from the license of that portion of the site that contains the building, and

• Any structure for which the surface contamination is not accounted for in the volumetric sampling of the structure. Examples of this are the steel liner attached to the inside of the containment building and embedded piping to remain after the release of the building from the license. The likelihood of these structures being occupied is very small and conducting the surveys to the Building Occupancy DCGLs allows for the confident use of the contamination levels corresponding to those DCGLs as bounding values in the calculation of "future groundwater" dose using the Basement Fill Model.

The residual radioactivity was assumed to be uniformly distributed over all surfaces of a room, including the floor, ceiling, and four walls. The average member of the critical group is an adult working in the building, engaged in light industrial activities.

The potential pathways used to estimate human radiation exposure resulting from residual radioactivity on the building surfaces include the following:

• External exposure directly from the source;

• External exposure to material deposited on the floor;

• External exposure due to air submersion;

• Inhalation of airborne radioactive particulates and tritium; and

• Inadvertent ingestion of radioactive material directly from the sources.

6.8.1.2 Conceptual Model The conceptual model underlying the dose model consisted of a room of fixed area (10 m by 10 m by 2.5 m high), uniform concentrations of residual radioactivity on all room surfaces, and the receptor located at the, center of the room at a height of 1m. Two cases were considered for the source type: area (surface) sources and volume sources. Area sources consisted of a thin-layer of residual radioactivity on the surface, consistent with NUREG/CR-55 12, Volume 1. Volumetric sources consisted of 0.305 m (12 inches) of concrete to account for the possibility of volumetrically contaminated sources, either by migration of radioactive material into the depth of the source or by neutron activation.

August 2005 6-15 Rev. 3

Haddam Neck Plant License'Termination Plan The parameters used to quantify the conceptual model are listed in Appendix D, Table D-3. The values /

distributions assigned to each of the parameters and the basis for assigning such values / distributions are

<-' also shown on the table.

6.8.1.3 Sensitivity Analysis Results Parameter distributions assigned in the stochastic model are presented in Appendix D, Table D-3. The stochastic parameters identified as sensitive (IPRCCI 2 0.1O) to the peak of the mean dose for each of the radionuclides are presented in Appendix E, Table E-3. For each radionuclide, the sensitive parameters are listed in order of decreasing sensitivity. Included in Table E-3 are the conservative values assigned to each of the sensitive parameters.

6.8.1.4 DCGL Determination Using the results of the sensitivity analysis, which identified which input parameters were sensitive to dose, conservative input values were selected (see Table E-3). Parameter values assigned in the deterministic model for area sources are presented in Appendix F, Table F-3.

For volume sources, 0.305 m (12 inches) of concrete was assumed for each of the six sources, which modeled an infinite thickness for the radionuclides of interest. In RESRAD-BUILD, the airborne concentration is determined by the parameter erosion rate, instead of the parameters removable fraction and time for source removal. A conservative value (75% quantile) for the erosion rate of 2.8E-7 cm/day based on NUREG/CR-6697, Attachment C, was used for those radionuclides which exhibited sensitivity for that parameter.

Building occupancy DCGLs were calculated using RESRAD-BUILD 2.37. The'DCGLs are presented in

< Table G-3, Appendix G. DCGLs for area sources have units of disintegrations per minute per 100 cm2 (dpm1 00 cm ). DCGLs for volume sources have units of pCi/g. The building occupancy DCGLs for each of the radionuclides are summarized in Table 6-3.

6.8.2 Future Groundwater Dose Subsurface'Structures and Basements/Footings:

Basement Fill Model Equation 5-6 will to be used to demonstrate compliance with license termination criteria for land areas potentially affected by groundwater contamination. This equation has three dose components that must total at most the unrestricted release criteria. These dose components are:

• Dose due to residual radioactivity in soil

• Dose due to existing groundwater contamination

• Dose due to "future groundwater" from the burial of concrete structures. This component is from radioactivity being released to the groundwater contained in the building basements The first two dose components are determined using the Soil and Groundwater DCGLs as described in other portions of this LTP. The future groundwater dose is determined using the Basement Fill Model and includes radioactivity from building basement/footing concrete, the containment liner and embedded piping that will remain on-site after release of site buildings from the NRC license.

August 2005 6-16 Rev. 3'

Haddam Neck Plant License Termination Plan The calculation of buildup factors as provided in reference 6-19 accounts for the groundwater flow velocity (from the site hydrogeologic parameters), the retardation values'for each radionuclide, and radioactive decay for each radionuclide over time.

Table 6-3 summarizes the values of Kid, R, and Bi as used in this model and in equations 6-4 and 6-5 for each structure/component.

Table 6-3: Parameter Values for Equations 6-4 and 6-5 Radionuclide Kj K R B H-3 0.06 1.26 1.91 Fe-55 1200 5350 .1.62 Co-60 22 99 1.65 Sr-90 10 45.6 2.84 Cs-137 45 202 2.86 Eu-152 825 3678 1.93 Reference 6-19 provides an initial assessment of the concrete structures that will remain following license termination. In this assessment, a concentration of 1 pCi/g for each radionuclide and for each structure was used as an initial starting point. In the implementation of this model, the actual average concrete concentrations will be used to obtain the value of CW,, as described in section 5.7.1.6.

Using the above value for Cw,,, the future groundwater dose, Hfulum Gw, due to concrete in the containment and spent fuel pool basements is determined from the groundwater DCGLs, DCGLbwI as follows:

HfutureGW =25 C" (eq. 6-6)

DCGLGW.i It should be noted that the above methodology assumes that concrete, the metal liner and the rebar, piping and other metal items embedded in the concrete in areas of activation have the same radionuclide concentrations. As discussed in section 5.7.1.6, this factor will be confirmed during characterization sampling. Should the rebar or metal have higher concentrations then the concrete, the higher concentrations will be used in the future groundwater dose calculation.

As discussed in the Reference 6-19 this method is very conservative due to following simplifying assumptions:

• The literature diffusion rates used are the highest values from the range given in the literature

• The use of the Buildup Factor assigns the highest available radioactivity inventory to the first year even though, for several radionuclides, this maximum occurs in different years.

• The radioactivity is assumed to diffuse from both the inside and the outside of all concrete masses into the same dilution volume (i.e., containment basement). This conservative assumption does not account for additional dilution that will occur from groundwater flow around the subsurface basements.

6.8.2.2 Future Groundwater Dose Calculation for Basement Concrete From equation 64, the cumulative fractional release, CFR5,;, is needed for all structures/components, s, and radionuclides, i, contained in the saturated zone (i.e. water table). Reference 6-19 provides a physical inventory of these concrete structures. This inventory includes all dimensions along with the estimated August 2005 6-18 Rev. 3

Haddam Neck Plant-License Termination Plan volumes and masses. For each major structure, the CFR values are calculated from these dimensions and from a conservative selection of diffusion coefficients using:

2jSA, (D t /,z) 05 (eq. 6 7)'

CFRSJ v e.67 Where:

CFR = cumulative fractional release of the material.

f= conversion factor = 0.01 m/cm SA = surface area (i 2)

V = volume of concrete (m3)

D = diffusion coefficient (crni/s), and t = time (s) = 1 year or 3.17E7 sec Based on a literature review of available data and on a review of available experimental data from CYAPCO concrete, Reference 6-19 selects conservative values of Di as provided in Table 6-4.

Table 6-4: Concrete Diffusion Coefficients Used in the Basement Fill Model Selected Diffusion Radionuclide Coefficient, Di

.- (cm 2 /s)

H-3 5.5x 107 Fe-55 5.0x 10-Co-60 4.0 x 10'-

Sr 5.2x.

XO1' 0 Cs-137 3.0 x l0u9 Eu-152 I 1.0 x lo-,,

Using the parameter values above with equations 6-4 and 6-6, the calculated groundwater concentrations and dose are conservative as a result of the following simplifying assumptions.

• The concrete surfaces are assumed to be represented by a semi-infinite geometry.

• No credit is taken for the barrier to diffusion that the containment liner provides.

• The Brookhaven Study assumes that 2.5 feet of grout will be placed above the activated concrete region of the In-Core Sump. The depth grout placed above the activated region will actually be 5 feet, thereby providing additional resistance to transport.

The analysis provided by Reference 6-19 is specifically for the containment and spent fuel pool basements taken together. Other adjacent subsurface structures such as the cable vault portion of the containment and other footings will be inventoried in a similar manner and added to the containment dilution volume (footings assessed after completion of the containment liner FSS will be included with the discharge tunnel inventory). This approach will also be used for other basements that remain (i.e. "B" Switchgear and the discharge tunnels) although the released radioactivity from these additional basements will be assumed to migrate and be included with the discharge tunnel inventories.

Rev. 3 6-19 2005 August 2005 6-19 Rev. 3

-£ -

Haddam Neck Plant License Termination Plan 6.8.23 Future Groundwater Dose Calculation for Surface Contamination on the Containment Liner As mentioned previously, the containment liner below four feet belowgfide will remain in place after the containment building is released from the license. This section describes the method to be used to account for this added surface contamination source.

For the containment liner, a characterization and final status survey will be implemented using the building occupancy DCGLs adjusted based on the results of an ALARA evaluation or administrative level not to exceed 25 mrem/yr. These surveys will establish the appropriate radionuclide mix for this surface.

For purposes of the basement fill model, this radionuclide mix will be used to calculate a total inventory assuming that the average surface radioactivity concentration is equal to the building occupancy values.

Using this inventory, a CFR value of 1.0 will be used along with the total surface area, and, equations 64 and 6-6 will be used to calculate the future groundwater dose and concentrations. These values will be added to their respective values for the concrete case discussed in 6.8.2.2. This is conservative approach since it assumes

1. that the total inventory of radioactivity for each radionuclide is released to the groundwater/backfill soil system assumed in the containment basement in the first year after release of this area from the NRC license, and

2. that all surfaces are contaminated to a level equivalent to the building occupancy DCGL.

Using this calculational approach, the surveys conducted for this source will be performed using the MARSSIM guidance as described in Chapter 5 for surfaces. . The results of these surveys will be used to determine average activity concentrations and radionuclide distributions.

KJi The above methodology applies to the containment liner. For all other buildings that have basement surfaces, no liner will remain in place. The contamination on the surfaces of these basements are accounted for in the volumetric sampling and subsequently in the calculation of future groundwater dose due to volumetric contamination. Separate calculations of dose due to this surface contamination on exposed building surfaces are therefore not required.

6.8.2.4 Future Groundwater Dose Calculation for Surface Contamination on Embedded Piping The last source to be evaluated in determining the future groundwater dose from buried structures and components is that resulting for the surface contamination contained on embedded piping to remain after termination of the license.

Embedded piping is that which is present in the containment or spent fuel pool basement and will not be removed or grouted. This source will be included in the calculation of the containment interior groundwater concentration. As in the case of the containment liner, surface activity surveys will be performed in accordance with Section 5.4.7.4 using building occupancy DCGLs determined by an ALARA evaluation or administrative limit. Using the radionuclide mix from these and other characterization surveys, each pipe will be assumed to be contaminated to levels equivalent to a total contamination level corresponding to building occupancy DCGL being used. Using this inventory, a CFR value of 1.0 will be used along with the total surface area, and, equations 64 and 6-6 will be used to calculate the future groundwater dose and concentrations. These values will be added to their respective values for the concrete case discussed in 6.8.2.2. This is conservative approach since it assumes August 2005 6-20 Rev. 3

Haddam Neck Plant Lice'nse Termination Plan

1. that the total inventory of radioactivity for each radionuclide is released to the groundwater/backfill soil system assumed in the containment basement in the first year after K release of this area from the NRC license, and'

2. that all surfaces are contaminated to a level equivalent to the building occupancy DCGL.

6.8.2.5 Summary of All Future Groundwater Dose Calculations Sections 6.8.2.2 to 6.8.2.4 show the method to be used determine the future groundwater doses due to the individual sources, buried concrete, containment liner and embedded piping. The individual doses will next be summed to determine the total future groundwater dose from all sources. This calculated dose will be used to supply the "future groundwater" dose component of the compliance equation (5-1). In the case of the containment basement, the future groundwater dose will be calculated as given in the methodology above. This calculation will be performed concurrent with the survey of the containment and be available for NRC review prior to the backfilling of the containment basement.

This approach will also be used for other basements that remain (i.e. "B" Switchgear and the discharge tunnels) although the released radioactivity from these additional basements will be assumed to migrate and be included with the discharge tunnel inventories.

6.9 Operational DCGLs Since additional scenarios, beyond those described above, may be created by combining pathways from different scenarios (e.g., resident farmer with the future groundwater dose calculated using the Basement Fill Model), a method to assess doses from these combined pathways is necessary. Additionally, any K- initial residual radioactivity in groundwater that exists will also contribute to total dose. For example, a resident farmer may locate his residence and raise crops on soil containing residual radioactivity and use groundwater that is in contact with a buried basement, which may also contain residual radioactivity. Soil and groundwater DCGLs for these combined scenarios along with the future groundwater dose calculated using the Basement Fill Model will be determined on an operational basis, using the base case DCGLs for soil and groundwater and future groundwater dose, calculated in Sections 6.6, 6.7, and 6.8. Section 5.4.7.1 describes, in detail, the methodology to account for all of these contributions.

6.10 References 6-1 Code of Federal Regulations, Title 10, Section 20.1402, "Radiological Criteria for Unrestricted Use."

6-2 NUREG-1575, "Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM),"

dated December 1997.

6-3 "Groundwater Monitoring Report," Connecticut Yankee Atomic Power Company, Haddam Neck, Connecticut, Malcolm Pirnie, Inc., September 1999.

64 Draft Regulatory Guide 4006, "Demonstrating Compliance with the Radiological Criteria for License Termination," August 1998.

6-5 NUREG-1549, "Decision Methods for Dose Assessment to Comply with Radiological Criteria for License Termination," July 1998.

August 2005 6-21 Rev. 3

Haddam Neck Plant License Termination Plan X-6 NUREG/CR-5512, Volume 1, "Residual Radioactive Contamination from Decommissioning, Technical Basis for Translating Contamination Levels to Annual Total Effective Dose Equivalent," October 1992.,;

6-7 NUREG-1727, "NMSS Decommissioning Standard Review Plan," September 2000.

6-8 1997 Census of Agriculture, Volume I, Geographic Area Series, Table 1. County Summary Highlights: 1997.

6-9 Connecticut Town Profiles 1998-1999, Connecticut Department of Economics and Community Development, Research Section, Public and Government Relations Division.

6-10 NUREG/CR-6676, "Probabilistic Dose Analysis Using Parameter Distributions Developed for RESRAD and RESRAD-BUILD Codes," May 2000.

6-11 NUREG/CR-6692, "Probabilistic Modules for the RESRAD and RESRAD-BUILD Computer Codes", November 2000.

6-12 NUREG/CR-6697, "Development of Probabilistic RESRAD 6.0 and RESRAD-BUILD 3.0 Computer Codes", December 2000.

6-13 "Users Manual for RESRAD, Version 6.0," July 2001.

6-14 "Manual for Implementing Residual Radioactive Material Guidance using RESRAD, Version 5.0", September 1993.

6-15 Yu et al., "RESRAD-BUILD: A Computer Model for Analyzing the Radiological Doses Resulting from the Remediation and Occupancy of Buildings Contaminated with Radioactive Materials," ANL/EAD/LD-3, Argonne National Laboratory, November 1994.

6-16 NUREG/CR-6755, Technical Basis for Calculating Radiation Doses for the Building Occupancy Scenario Using the Probabilistic RESRAD-BUILD 3.0 Code, February 2002.

6-17 Batelle, Pacific Northwest Division (PNWD), "Radonuclide Desorption and Leaching Tests for Concrete Cores from Haddam Neck Nuclear Plant Facilities", March 2002 6-18 NotUsed.

6-19 Technical Support Document CY-HP-0 184,

Subject:

Estimates for Release of Radionuclides from Potentially Contaminated Concrete at the Haddam Neck Nuclear Plant.

6-20 Technical Support Document CY-HP-0 185,

Subject:

Kd Values of Backfill Material for Connecticut Yankee.

August 2005 6-22 Rev. 3