CY-04-115, Update of License Termination Plan - Revision 2

Update of License Termination Plan - Revision 2

ML042740450
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	08/30/2004
From:	Norton W Connecticut Yankee Atomic Power Co
To:	Document Control Desk, NRC/FSME
References
+sispmjr200505, -RFPFR, CY-04-115
Download: ML042740450 (345)

Text

o CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLANT 362 INJUN HOLLOW ROAD

• EAST HAMPTON, CT 06424-3099 AUG 30 2004 Docket No. 50-213 CY-04-115 Re: 10 CFR 50.71(e)

U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Haddam Neck Plant Update of License Termination Plan - Revision 2 In a letter dated July 7, 20001, as revised in August 2002 (Revision 1) and in October 2002 (Revision 1A), Connecticut Yankee Atomic Power Company (CYAPCO) submitted the Haddam Neck Plant (HNP) License Termination Plan (LTP) to the NRC as a supplement to the HNP Updated Safety Analysis Report (UFSAR). The LTP was subsequently approved by the NRC via License Amendment 1972. In accordance with the requirements of 10 CFR 50.71(e),

CYAPCO is providing the attached update (Revision 2) of the HNP LTP. This update includes a change in the decommissioning approach. Specifically, the current decommissioning plan has been modified to include building demolition, bulk material disposal at a regulated disposal facility, final survey and site restoration using clean soil as appropriate.

On July 1, 2004, CYAPCO filed an update3 to the decommissioning cost estimate to the Federal Energy Regulatory Commission (FERC). Chapter 7, "Update of Site-Specific Decommissioning Cost," of the HNP LTP has been updated to 1 K. Heider (CYAPCO) letter to U. S. NRC, "Haddam Neck Plant - License Termination Plan," dated July 7, 2000.

2 J. Donohew (USNRC) letter to K. Heider (CYAPCO), "Haddam Neck Plant -

Issuance of Amendment Re - Approval of License Termination Plan (LTP),"

dated November 25, 2002.

3 Letter CYAPCO to FERC, "Revision to CY Wholesale Power Contract," dated July 1, 2004.

,. tOpm

Document Control Desk CY-04-115 / Page 2 include this FERC update. In addition, Chapter 8, "Supplement to the Environmental Report" has been updated using the format of Supplement 1 to NUREG-0586, "Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities," dated November 2002.

CYAPCO reviewed the changes included in this revision against those criteria specified in 10 CFR 50.59, 10 CFR 50.82(a)(6) and 10 CFR 50.82(a)(7), and HNP License condition 2.C(7). CYAPCO concluded that the changes do not require NRC approval prior to being implemented. This update (Revision 2 to the HNP LTP) is included in the Attachment. An 'instruction sheet' is provided therein describing how to insert and/or replace the pages/tables and figures incorporating this revision.

Pursuant to 10 CFR 50.71 (e)(2)(i), I certify that this revision accurately presents changes completed since CYAPCO's last submittal (October 2002) that are necessary to reflect information and changes made pursuant to HNP License Condition 2.C(7).

There are no regulatory commitments contained in this letter.

If you should have any questions regarding this submittal, please contact Mr. G.

P. van Noordennen at (860)-267-3938.

Sincerely, 4 ne A~rt,/ Ddtei Subscribed and sworn before me

-this-, 4) day ofA r,2004

- a- _ Notary Public My Commission Expires 127 / 0 7

Document Control Desk CY-04-115 / Page 3 Attachment cc: S. J. Collins, NRC Region 1 Administrator T. B. Smith, NRC Project Manager E Wilds, Jr., Director, CT DEP, Radiation and Monitoring Division R. R. Bellamy, Chief, Decommissioning and Laboratory Branch, NRC Region 1 M. Rosenstein, US Environmental Protection Agency, Region 1

Docket No. 50-213 CY-04-1 15 Attachment Haddam Neck Plant Update of License Termination Plan Revision 2 August 2004

Haddam Neck Plant License termination Plan (LTP) Revision 2 Instruction for Removal/ Insertion of Pages into the Controlled Copv of the LTP Remove Page (s) Number Insert Page (S) Number Table of content Table of content i through xii i through xiii List of Effective Pages List of Effective Pages LEP-1 thru 4 LEP-I thru 4 After Chapter 1 Tab After Chapter 1 Tab Pages 1-1 thru 1-12 Pages 1-1 thru 1-12 After Chapter 2 Tab After Chapter 2 Tab Pages 2-1 thru 2-128 Pages 2-1 thru 2-98 Figures 2-1 thru 2-23 Figures 2-1 thru 2-9 After Chapter 3 Tab After Chapter 3 Tab Pages 3-1 thru 3-30 Pages 3-1 thru 3-28 After Chapter 4 Tab After Chapter 4 Tab Pages 4-1 thru 4-6 Pages 4-1 thru 4-6 Figure 4-1 Figure 4-1 After Chapter 5 Tab After Chapter 5 Tab Pages 5-1 thru 5-66 Pages 5-1 thru 5-70 Figure 5-4 Figure 5-4 After Chapter 6 Tab After Chapter 6 Tab Pages 6-1 thru 6-22 Pages 6-1 thni 6-22 Figure 6-6 Figure 6-6 After Chapter 7 Tab After Chapter 7 Tab Pages 7-1 thru 7-10 Pages 7-1 thru 7-7 After Chapter 8 Tab After Chapter 8 Tab Pages 8-1 thru 8-6 Pages 8-21 None After Appendix G, Page G-12, Add Appendix H Pages H-I thru H-47

Haddam Neck Plant License Termination Plan TABLE OF CONTENTS 1 GENERAL INFORMATION 1-1 1.1 Purpose 1-1 1.2 Historical Background 1-1 1.3 Plan Summary 1-2 1.3.1 General Information 1-2 1.3.2 Site Characterization 1-3 1.3.3 Identification of Remaining Site Dismantlement Activities 1-4 1.3.4 Site Remediation Plans 1-5 1.3.5 Final Status Survey Plan 1-6 1.3.6 Compliance with the Radiological Criteria for License Termination 1-6 1.3.7 Update of Site-Specific Decommissioning Costs 1-7 1.3.8 Supplement to the Environmental Report 1-7 1.4 Decommissioning Approach 1-7 1.4.1 Overview 1-7 1.4.2 Phased Release Approach 1-9 1.5 License Termination Plan Change Process 1-10 1.6 References 1-10 2 SITE CHARACTERIZATION 2-1 2.1 Introduction 2-1 2.2 Historical Site Assessment 2-2 2.2.1 Introduction 2-2 2.2.2 Methodology 2-2 2.2.3 Instrumentation Selection, Use and Minimum Detectable Concentration 2-3 2.2.4 Results 2-7 2.2.4.1 Routine Releases 2-7 2.2.4.2 Operational Events 2-7 2.3 Initial Site Characterization 2-14 2.3.1 Introduction 2-14 2.3.2 Methodology 2-14 2.3.3 Site Characterization/HSA Results 2-15 2.3.3.1 Radiological Status 2-15 2.3.3.2 Initial Area Classification 2-41 2.3.3.3 Non-Impacted Area Assessment 2-90 2.3.3.4 Radionuclide Suite Selection 2-91 2.3.3.5 Hazardous Material Status 2-96 2.4 References 2-96 3 IDENTIFICATION OF REMAINING SITE DISMANTLEMENT ACTIVITIES 3-1 3.1 Introduction 3-1 3.2 Spent Fuel Pool Island Activities 3-3 August 2004 i Rev. 2

Haddam Neck Plant License Termination Plan 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 3.4.2.26 Fire Protection System 3-21 August 2004 Rev. 2 ii

Haddam Neck Plant License Termination Plan 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 4-4 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 Summarv 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 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 5.5.1.3 Selecting the Required Number of Measurements for the WRS Test 5-29 August 2004 Rev. 2

Haddam Neck Plant License Termination Plan 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 I 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 5.7.2 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 543 5.7.2.4 Response Checks 5-43 5.7.2.5 MDC Calculations 544 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 6.2 Site Characteristics 6-2 August 2004 iv Rev. 2

Haddam Neck Plant License Termination Plan 6.3 Dose Modeling Approach 6-3 6.3.1 Overview 6-3 6.3.2 Resident Farmer Scenario 6-4 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 6-7 6.5.4 Sensitivity Analyses 6-7 6.5.5 Parameter Value Assignment 6-8 6.6 DCGLs for Soil 6-8 6.6.1 Dose Model 6-9 6.6.2 Conceptual Model 6-9 6.6.3 Sensitivity Analysis Results 6-9 6.6.4 DCGL Determination 6-10 6.7 DCGLs for Groundwater 6-10 6.7.1 Dose Model 6-10 6.7.2 Conceptual Model 6-11 6.7.3 Sensitivity Analysis Results 6-12 6.7.4 DCGL Determination 6-12 6.8 DCGLs for Concrete 6-14 6.8.1 DCGLs for Concrete: Buildings Standing 6-14 6.8.1.1 Dose Model 6-14 6.8.1.2 Conceptual Model 6-15 6.8.1.3 Sensitivity Analysis Results 6-15 6.8.1.4 DCGL Determination 6-15 6.8.2 DCGLs for Concrete: Building Demolished (Concrete Debris) 6-16 6.8.2.1 Dose Model 6-16 6.8.2.2 Conceptual Model 6-17 6.8.2.3 Sensitivity Analysis Results 6-18 6.8.2.4 DCGL Determination 6-18 6.8.3 Concrete DCGL Conversion 6-19 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 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 August 2004 v Rev. 2

Haddam Neck Plant License Termination Plan 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 84 8.1.3 Remaining Dismantlement and Decommissioning Activities 84 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 8-1 1 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 Appendix B, ALARA Evaluations Appendix C, Financial Materials vi Rev. 2 August 2004 vi Rev. 2

Haddam Neck Plant License Termination Plan 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 Rev. 2 August 2004 vii Rev. 2

Haddam Neck Plant License Termination Plan LIST OF TABLES I GENERAL INFORMATION No tables 2 SITE CHARACTERIZATION Table 2-1, Typical Instruments Used at HNP 2-4 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-I 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 3-2 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 Remaining Buried Piping 5-24 Table 5-8, Investigation Levels 5-36 Table 5-9, Traditional Scanning Coverage Requirements 5-41 Table 5-10, Available Instruments and Associated MDCs 549 August 2004 viii Rev. 2

Haddam Neck Plant License Termination Plan Table 5-1 1, Initial Evaluation of Survey Results (Background Reference Area Used) 5-60 Table 5-12, 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, Base Case Building Surface DCGLs (Building Occupancy Scenario) 6-16 Table 64, Base Case DCGLs for Building Demolished (Concrete Debris) 6-19 Table 6-5, Conversion of Base Case Concrete Debris DCGLs 6-21 7 UPDATE OF SITE-SPECIFIC DECOMMISSIONING COSTS Table 7-1, Actual and Projected Decommissioning Expenditures 7-4 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-l 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 lx Rev. 2 August 2004 ix Rev. 2

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

Table E-l 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-I 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-I DCGLs for Soil G-2 Table G-2-l 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-i DCGLs for Concrete: Buildings Demolished G-9 Table G4-2 Concentration of Residual Radioactive Material in (lie Contaminated Zone (pCi/g) 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 x Rev. 2 August 2004 x Rev. 2

Hladdam Neck Plant License Termination Plan LIST OF FIGURES (All Figures Are Located at the End of the Associated Section)

I 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 Figure 6-6, Process for Determining Building Surface DCGLs 7 UPDATE OF SITE-SPECIFIC DECOMMISSIONING COSTS No figures.

xi Rev. 2 August 2004 xi Rev. 2

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

August 2004 xii Rcv. 2

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

xiii Rcv. 2 August 2004 xiii Rev. 2

Haddam Neck Plant License Termination Plan LIST OF EFFECTIVE PAGES Front Matter Page Revision Date i thru xiii 2 August 2004 LEP-1 thru LEP-4 la August 2004 I GENERAL INFORMATION Page Revision Date 1-1 thru 1-12 2 August 2004 2 SITE CHARACTERIZATION Page Revision Date 2-1 1 August 2002 2-2 thru 2-3 2 August 2004 2-4 thru 2-5 1 August 2002 2-6 0 7/7/00 2-7 thru 2-8 1 August 2002 2-9thru2-10 2 August 2004 2-11 thru 2-12 1 August 2002 2-13 2 August 2004 2-14 1 August 2002 2-15 thru 2-16 2 August 2004 2-17 thru 2-18 1 August 2002 2-19 thru 2-98 2 August 2004 3 IDENTIFICATION OF REMAINING SITE DISMANTLEMENT ACTIVITIES Page Revision Date 3-1 thru 3-28 2 August 2004 4 SITE REMEDIATION PLANS Page Revision Date 4-1 2 August 2004 4-2 1 August 2002 LEP-1 Rev. 2 2004 August 2004 LEP- I Rev. 2

Haddam Neck Plant License Termination Plan LIST OF EFFECTIVE PAGES 4-3 thru 4-4 2 August 2004 4-5 thru 4-6 August 2002 5 FINAL STATUS SURVEY PLAN Page Revvision Date 5-1 1 August 2002 5-2 thru 5-70 2 August 2004 6 COMPLIANCE WITH THE RADIOLOGICAL CRITERIA FOR LICENSE TERMINATION Page Revision Date 6-1 thru 6-18 2 August 2004 6-19 1 August 2002 6-20 2 August 2004 6-21 thru 6-22 1 August 2002 7 UPDATE OF SITE-SPECIFIC DECOMMISSIONING COSTS Page Revision Date 7-1 thru 7-7 2 August 2004 8 SUPPLEMENT TO TIHE 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 7/7/00 Figure 5-3 August 2002 Figure 5-4 2 August 2004 Figure 6-1 0 7/7/00 Figure 6-2 0 7/7/00 LEP-2 Rex'. 2 August 2004 LEP-2 Rev. 2

lladdam Neck Plant License Termination Plan LIST OF EFFECTIVE PAGES Figures 6-3 thru 6-5 August 2002 Figure 6-6 2 August 2004 Appendix A Page Revision Date A- I thru A-2 0 7/7/00 Appendix B Page Revision Date B-i thru B4 August 2002 Page Revision Date C-l thru C-36 0 7/7/00 Appendix D Page Revision Date D- I thru D-31 August 2002 Appendix E Page Revision Date E-1 thru E-12 August 2002 Appendix F Page Revision Date F1 thru F-32 August 2002 Appendix G Page Revision Date GI thruG-12 August 2002 Appendix H Page Revision Date H-I thru H47- 2 August 2004 LEP-3 Rev. 2 2004 August 2004 LEP-3 Rev. 2

Haddam Neck Plant License Termination Plan LIST OF EFFECTIVE PAGES This page intentionally left blank.

August 2004 LEP4 Rev. 2

Haddani Neck Plant License Termination Plan I GENERAL INFORMATION 1.1 Purpose The objective for decommissioning the Haddam Neck Plant (HNP) site is to reduce residual radioactivity to levels that permit release of the site for unrestricted use and for termination of tile I OCFR50 license, in accordance with the Commission's site release criteria set forth in IOCFR20, Subpart E. The purposc of this HNP License Termination Plan (LTP) is to satisfy the requirements of IOCFR50.82, "Termination of License" (Reference 1-1) using the guidance provided in Regulatory Guide 1. 179, "Standard Format and Content of License Termination Plans for Nuclear Power Reactors" (Reference 1-2) and Draft Regulatory Guide-4006, "Demonstrating Compliance with the Radiological Criteria for License Termination" (Reference 1-3). In September of 2000, the NRC incorporated much of the guidance of DG-4006 into various sections of NUREG-1 727 (Reference 1-4). References to the corresponding sections of NUREG-1727 (in which the guidance of DG-4006 have been incorporated) has been given in specific sections of this LTP, as appropriate.

Tile LTP describes the decommissioning activities that will be performed, the process for performing the final status surveys, and the method for demonstrating that the site meets the criteria for release for unrestricted use. The LTP contains specific information oln:

• Historical Site Assessment and Site Characterization;

• Remaining Decommissioning Activities;

• Site Remediation Plans;

• Final Status Survey Design and Implementation Plan;

• Dose Modeling Scenarios;

• Update to the Site-Specific Decommissioning Cost Estimate; and

• Supplement to the Environmental Report.

Each section of the LTP is summarized in Section 1.3.

1.2 Historical Background The HNP is located on the east bank of the Connecticut River, approximately 21 miles south-southeast of Hartford, at 362 Injun Hollow Road, Haddam, Middlesex County, Connecticut. HNP is owned by Connecticut Yankee Atomic Power Company, CYAPCO (Reference 1-5). Figures depicting the site area and buildings are provided at the end of LTP Chapter 2.

HNP, Docket No. 50-213 (License No. DPR-61), began commercial operation in January 1968. The plant incorporated a 4-loop closed-cycle pressurized water type nuclear steam supply system (NSSS); a turbine generator and electrical systems; engineered safety features; radioactive waste systems; fuel handling systems; instrumentation and control systems; the necessary auxiliaries; and structures to house plant systems and other onsite facilities. HNP was designed to produce 1,825 MW of thermal power and 590 MW of gross electrical power (Reference 1-6).

On December 4, 1996, HNP permanently shut down after approximately 28 years of operation. On December 5, 1996, CYAPCO notified the Nuclear Regulatory Commission (NRC) of tile permanent cessation of operations at the HNP and the permanent removal of all fuel assemblies from the Reactor Pressure Vessel and their placement in the Spent Fuel Pool (Reference 1-7). Following the cessation of August 2004 l l Rev. 2

Haddam Neck Plant License Terminiation Plan operations, CYAPCO began to decommission the HNP. The Post Shutdown Decommissioning Activities Report (PSDAR) was submitted, in accordance with I OCFR50.82 (a)(4), on August 22, 1997 (Reference 1-8), and was accepted by the NRC (Reference 1-9). On January 26, 1998, CYAPCO transmitted an Updated Final Safety Analysis Report to reflect the plant's permanent shutdown status (Reference 1-10), and on June 30, 1998, the NRC amended the HINP Facility Operating License to reflect this plant condition (Reference 1-1 1). On October 19, 1999, the Operating License was amended to reflect the decommissioning status of the plant and long-term storage of the spent fuel in the spent fuel pool.

(Reference 1-12). Additional licensing basis documents were also revised and submitted to reflect long-term fuel storage in the spent fuel pool (Defueied Emergency Plan, Security Plan, QA program, and Operator Training Program).

In April of 1999, CYAPCO contracted Bechtel Power Corporation, as the Decommissioning Operations Contractor (DOC), to perform the decommissioning activities at HNP. CYAPCO continued to perform Spent Fuel Pool Island Operations and provide oversight of the activities performed by the DOC, until June 2003, when CYAPCO terminated the DOC contract. CYAPCO is now managing the decommissioning using staff augmentation and subcontractors for specialty work.

1.3 Plan Summary Termination of the NRC license and environmental closure of the HNP site are closely related activities, completion of which will allow the site to be released for future use. The License Termination Plan describes the processes to be used in meeting the requirements for terminating the NRC license. An integrated site closure plan is also being prepared to include the processes to be used for non-radiological cleanup and release of the site. This information will be submitted to the appropriate regulatory agencies.

An integrated approach to site release processes will be used to the extent practicable.

The decommissioning will be conducted by performing radiological and hazardous environmental surveys to allow for controlled demolition of structures, removal of the wastes generated from the site, performing the Final Status Survey (FSS) of the remaining foundations/basements and/or soils, and the use of appropriate backfill materials to restore the site to grade elevation. Soils identified to be contaminated above release limits will be removed, an FSS or assessment performed, the area restored to grade using an appropriate backfill material and an FSS satisfactorily completed if required.

Controlled demolition is the removal of a radiologically contaminated System, Structures, or Components (SSC) using industry accepted construction technologies with limited to continuous radiological controls, based on radiological surveys. These radiological surveys may include but are not limited to misting, air sampling periodic surveys, tenting, and personal protective equipment such as respirators.

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 site. Due to a change in the approach to waste disposal, some previously identified survey areas were removed from Table 2-10. Appendix H contains historical information from Table 2-10. If areas are later determined to require an FSS (due to a change in waste disposal approach), the classification information provided in Appendix H will be used.

1.3.1 General Information This LTP has been prepared for the HNP in accordance with the requirements of I OCFR50.82 (a)(9). The LTP is being maintained as a supplement to the HNP Updated Final Safety Analysis Report. Each of the August 2004 I1-2 Rev. 2

liaddani Neck Plant License Termination Plan sections required by IOCFR50.82 (a)(9) are outlined in the subsections below. Note that figures are located at the end of the corresponding section.

1.3.2 Site Characterization Chapter 2 discusses site characterization activities. The site characterization for HNP includes the results of surveys and evaluations conducted to determine the extent and nature of the contamination at the site.

The initial characterization, performed in accordance with the guidelines of the "Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM)," (Reference 1-13) began in 1997 and was completed in 1999. This initial characterization included a Historical Site Assessment (HSA), a review of historical documents, and measurements, samples, and analyses to further define the current conditions of the site.

The effort also evaluated hazardous and state-regulated non-radioactive materials at the site that may require remediation and disposal.

The HSA consisted of a review and compilation of the following information: historical records, plant and radiological incident files, operational survey records, and annual environmental reports to the NRC.

Personnel interviews were conducted with present and former plant employees and contractors to obtain additional information regarding operational events that caused contamination in areas or systems not designed to contain radioactive or hazardous materials.

Information from previous surveys was reviewed for historical information regarding radiological conditions throughout the site. The current HNP Radiation Protection Program requires that site radiological conditions be assessed and documented by performing operational surveys and evaluations throughout the decommissioning process. The radiological data collected during this process will supplement the initial characterization data and provide a basis for developing plans for remediation and final status surveys.

The information developed during the initial HNP characterization program represents a radiological and hazardous material assessment based on the knowledge and information available at the end of 1999. The objectives of this initial characterization program were:

I. To divide the HNP site into manageable sections or areas for survey and classification purposes;

2. To identify the potential and known sources of radioactive contamination in systems, on structures, in surface or subsurface soils, and in ground water;

3. To determine the initial classification of each survey area;

4. To develop the initial radiological and hazardous material informiation to support decommissioning planning including building decontamination, demolition, and waste disposal;

5. To develop the information to support Final Status Survey design including instrument performance standards and quality requirements; and

6. To identify any unique radiological or hazardous material health and safety issues associated with decommissioning.

Operational radiation surveys and additional characterization measurements and samples obtained during cleanup activities will be used to confirm the area classification and effectiveness of the cleanup activities before completing the Final Status Survey.

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Hladdam Neck Plant License Termination Plan The site characterization and historical site assessment efforts are summarized in two documents:

'Connecticut Yankee Haddam Neck Plant Characterization Report" (Reference I-14) and the "Haddam Neck Plant Historical Site Assessment Supplement" (Reference 1-15).

The LTP includes a summary of information contained in References 1-14 and 1-15. 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.

As a result of the HSA and site characterization, approximately 93 acres of the plant site have been initially identified as "non impacted" as defined in MARSSIM. Table 2-10 provides the area classifications for the various survey areas of the FINP site. The results of the surveys are being used to identify areas of the site that require decontamination, as well as to identify the cleanup methods and plan for their associated costs.

1.3.3 Identification of Remaining Site Dismantlemient Activities CYAPCO has begun decontamination and dismantlement activities at the HNP site consistent with activities discussed in the HNP PSDAR. Chapter 3 of the LTP describes those dismantlement and decontamination activities that remain at the HNP as of May 2004. Also included in this section are estimates of radiation dose to workers from decommissioning activities and projected volumes of radioactive waste.

CYAPCO's primary goals are to decommission the HNP safely and to maintain the safe storage of spent fuel and Greater Than Class C (GTCC) waste. To the extent practical, impacted facility materials and surfaces that remain will be decontaminated to allow for beneficial reuse. Materials that cannot be decontaminated will be sent to an offsite radioactive waste processor to recycle or to a low-level waste disposal site. Completion of decommissioning the HNP site depends on the availability of low-level waste disposal sites. Currently, HNP has access to low-level waste disposal facilities including those in Barnwell, South Carolina, and Clive, Utah.

Spent Fuel Pool Island One of the significant activities that CYAPCO has performed is the creation of the Spent Fuel Pool Island. This involved separating the systems and components required to support storage of spent fuel in the Spent Fuel Pool from systems that no longer support current and planned decommissioning activities.

This minimizes the effects that decommissioning activities have on safe spent fuel storage.

Future Decommissioning Activities and Tasks The remaining decontamination activities can be placed into several categories that may be performed concurrently. These include: contaminated system removal, clean system removal, decontamination of site buildings and cleanup of the site land areas.

Decontamination of plant structures can occur at the same time as equipment removal. Decontamination techniques may range from water washing to removal of a layer of building surface material.

Contaminated equipment and structural material may be packaged and either shipped to a processing facility, or shipped directly to a low-level radioactive waste disposal facility.

Decontamination and dismantlement activities will continue generally independent of those activities related to the operation of the Spent Fuel Pool Island. Once construction of the Independent Spent Fuel August 2004 l1-4 Rev. 2

Haddarn Neck Plant License Terniniatioin Plan Storage Facility (ISFSI) and the necessary modifications to the Spent Fuel Island were complete, transfer of spent fuel and GTCC waste from the Fuel Building to dry storage casks at the onsite ISFSI commenced.

One of two types of radiological surveys will be performed on systems, structures, or components that will be demolished and the wastes generated shipped to a LLW storage facility or a licensed clean material landfill, as appropriate:

1. If a Structure, System, or Component (SSC) is known to be contaminated and is to be demolished, a Contamination Verification Survey (CVS) will be performed to ensure that contamination levels are within the established levels to permit controlled demolition.

2. If the SCC is not suspected to be contaminated, a Unrestricted Release Survey (URS), in compliance with the HNP Radiation Protection Program, will be performed to document that the SSC meets the criteria for unrestricted release.

If a contaminated SSC is to remain onsite, it will be decontaminated to the required levels, and a final status survey will be performed and documented. This survey will confirm that the site meets the release criteria. The final status survey results for each survey area will be complied into a release record documenting the as-left radiological conditions demonstrating compliance with site remediation criteria.

Several release records will be compiled in a series of reports by area(s). These reports, each made up of several release records will be made available for NRC inspection. Following completion of the final status survey and in the absence of any NRC inspection finding the report deficient, surveyed areas may be released from NRC license control.

1.3.4 Site Rernediation Plans Chapter 4 of the LTP describes various methods that can be used during HNP decommissioning to reduce the levels of radioactivity to those which meet the NRC radiological release criteria, that is, does not exceed 25mrem/yr Total Effective Dose Equivalent (TEDE) and is As Low As Reasonably Achievable (ALARA). This section describes tile methodology that will be used to demonstrate that the residual radioactivity has been reduced to a level that is ALARA in compliance with the NRC requirements.

An ALARA analysis determines when cleanup, beyond that required to meet the 25 mrem/yr TEDE dose limit, is appropriate. Figure 4-1 shows the ALARA evaluation process. Generic ALARA screening values may be determined at the planning stage, prior to the start of cleanup, or after some or all of the characterization work is complete. Survey unit-specific ALARA evaluations may be performied later in the remediation and survey processes.

These ALARA evaluations establish remediation levels at which additional cleanup actions are to be taken to reduce residual radioactivity. These different types of cleanup actions may include, but are not limited to chemical decontamination, wiping, vacuuming, scabbling, or high-pressure washing. The methodology and equations to be used for calculating remediation levels are those provided in NRC's Draft Regulatory Guide DG-4006, "Demonstrating Compliance with the Radiological Criteria for License Termination," which was subsequently included in Appendix D to NUREG-1727.

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Haddam Neck Plant License Termination Plan 1.3.5 Final Stattis 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 tile release criterion, and

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

Thle 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, wvhich 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.

1.3.6 Compliance with 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 I OCFR20.1402 (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, concrete debris, concrete basements that may remain and ground water, and its counterpart, RESRAD-BUILD (Version 2.37), to model dose from structures.

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, concrete debris, concrete foundations/basements and ground water and the building occupancy scenario for site buildings. Since concrete buildings may be demolished after acceptance of the final status surveys, the future potential use of concrete debris has been evaluated to ensure that the reuse is adequately bounded by doses calculated in the LTP. This evaluation considered the use of concrete debris as backfill on site. This evaluation uses the resident farmer scenario to calculate impacts from the concrete including the conservative assumption that future drinking water originates in a well located in the buried debris. The results of this additional scenario have been analyzed to ensure the most limiting radionuclide-specific DCGLs are used to calculate operational DCGLs for building surface surveys. Although current decommissioning plans do not call for the placement of concrete debris in facility basements, the methodology outlined above will be conservatively applied for the remaining basement structures. The option to use concrete debris as backfill is retained.

I1-6 August 2004 Rev. 2 1

Haddamn Neck Plant License Termnination Plan 1.3.7 Update of Site-Specific Decommissioning Costs In accordance with I OCFR50.82 (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 wvill be sufficient funds for completing decommissioning.

1.3.8 Supplement to the Environmental Report In accordance with I OCFR50.82 (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 wvas 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. Adding to the conservatism, an additional modeling scenario has been considered as discussed in Section 6.8.2 (i.e., resident farmer for concrete debris). Building foundations/basements, wlhich are decontaminated at or below the DCGLs, could be allowed to remain standing after the final status survey. These buildings and building foundations/basements could then be demolished and the debris dispositioned in a number of different manners. Consideration of the building occupancy scenario (as well as other scenarios) 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 overviewv 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.

August 2004 l1-7 Rev. 2

liaddarn Neck Plant License Terinination Plan Upon the decision to permanently cease power operations at the HNP site, CYAPCO began site characterization activities (Chapter 2). This characterization effort, wvhich 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 tile 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 tile dose modeling code to be used to calculate the DCGLs. These DCGLs correspond to a dose to the average member of tile 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 of the 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 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 recontamination 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, 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 demolition 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 foundations/basements and soil. The dose modeling approach, described in Chapter 6, evaluates potential exposures resulting from any remaining concrete structures, debris, foundations/basements to ensure that the doses are bounded by the conservative DCGLs 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 for unrestricted release, and thus are not low-level radioactive waste.

August 2004 I -8 Rev. 2

Haddarn Neck Plant License Termination Plan CYAPCO may also choose to remove specific land areas (and any associated buildings) from the IOCFR50 license after they have been surveyed and the results documented and provided to the NRC for its review and concurrence. A more detailed discussion of the phased release approach is provided in the following subsection. Upon completion of remediation and/or demolition, final status surveys, and confirmation that land areas (and any associated buildings) on the HNP site meet the site release criteria, CYAPCO will have completed the decommissioning process.

1.4.2 Phased Release Approach CYAPCO may choose to remove specific areas from the license in a phased manner. The approach for phased release and removal from the license, after approval of the License Termination Plan, is as follows:

I. Following completion of decommissioning activities and final status survey of a survey unit, CYAPCO will compile a final status survey report to address the area or building, where decommissioning and remediation tasks are complete and the criteria of IOCFR20.1402 have been met. The results of these surveys will be documented in a report, which is provided to the NRC for its review. A report will contain a compilation of release records of the areas surveyed.

A release record documents the as-left radiological condition of a survey area or survey unit.

2. Prior to a request to release a survey area from the license, the licensee will perform a Capture Zone Analysis and will assure that the ground water dose contribution is included for all applicable survey areas per the process described in Section 5.4.7.1 of the LTP.

3. CYAPCO will review and assess the impacts on the following documents in preparation of removing a land area (and any associated buildings) from the license:

• Updated Final Safety Analysis Report and Technical Specifications;

• Environmental Monitoring Program;

• Offsite Dose Calculation Manual;

• Defueled Emergency Plan;

• Security Plan;

• Post Shutdown Decommissioning Activities Report;

• License Termination Plan;

• Ground Water Monitoring Program;

• I OCFR 100 Siting Criteria; and

• Environmental Report.

The reviews wvill include an assessment to ensure that the land area(s), and any associated building(s), to be released will have no adverse impact on the ability of the site in aggregate to meet the Part 20, Subpart E, criteria for unrestricted release. The reviews will also include the impacts on the discharge of effluents and the limits of IOCFR 20, as they pertain to the public.

4. A letter of intent to remove a portion of the property from the Part 50 license will be sent to the NRC, at least sixty (60) days before the anticipated date for release of the subject survey area(s).

This letter wvill contain a summary of the assessments performed, as described above, and, for areas designated as "impacted," will include the FSS report for the subject survey units(s) or area(s).

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Haddain Neck Plant license Termination Plan

5. Once the land area(s), and any associated building(s), have been verified ready for release, no additional surveys or decontamination of the subject building or area w'ill be required (beyond those outlined in Section 5.4.6 intended for isolation and controls) unless administrative controls to prevent re-contamination are known or suspected to have been compromised. Following completion of the final status survey and submittal of the associated report, the NRC wvill review the report and conduct the applicable NRC confirmatory inspections.

6. Once the area(s), and any associated building(s), have been released from the license, remaining material can be dispositioned in accordance with state and federal requirements.

7. Upon completion of the HNP Decommissioning Project, a final report wvill be prepared, summarizing the release of areas of the HNP site from the IOCFR50 license.

1.5 License Termination Plan Change Process CYAPCO submitted the License Termination Plan to the NRC as a supplement to the Updated Final Safety Analysis Report (Reference 1-20). The NRC subsequently approved the License Termination Plan via License Amendment No. 197 (Reference 1-21). License Amendment 197 also adds a license condition, which provides the criteria against which changes to the License Termination Plan are evaluated to determine if prior NRC approval is required in addition to the criteria specified ill 10 CFR 50.59 and 10 CFR 50.82(a)(6) and (a)(7). A change to the LTP requires NRC approval prior to being implemented, if the change:

(a) Increases the radionuclide-specific derived concentration guideline levels (as discussed in Section of the LTP) or area factors (as discussed in Section 5.4.7.4 of the LTP);

(b) Increases the probability of making a Type I decision error above the level stated in the LTP (discussed in Section 5.5. 1.1 of the LTP);

(c) Increases the investigation level thresholds for a given survey unit classification (as given in Table 5-8 of the LTP);

(d) Changes the classification of a survey unit from a more restrictive classification to a less restrictive classification (e.g. Class I to Class 2, or Class A to Class B). Definitions for the different classifications for structures and surface soils are provided in Section 2.3.3.2 of the LTP, and definitions for the different classifications for subsurface soils are provided in Section 2.3.3.1.5 of the LTP; (e) Reduces the coverage requirements for scan measurements (Table 5-9 of the LTP); or (f) Involves reliance upon statistical tests other than the WRS or Sign Test (as discussed in Section 5.8 of the LTP) for data evaluation.

1.6 References 1-1 Code of Federal Regulations, Title 10, Part 50.82, "Termination of License."

1-2 Regulatory Guide 1.179, "Standard Format and Content of License Termination Plans for Nuclear Power Reactors," January 1999.

August 2004 l -lo Rev. 2

1laddani Neck Plant License Termination IPlan 1-3 Draft Regulatory Guide-4006, "Demonstrating Compliance wvith the Radiological Criteria for License Termination," August 1998.

1-4 NUREG-1727, "NMSS Decommissioning Standard Review Plan," dated September 2000.

1-5 Haddam Neck Facility Operating License (DPR-61) issued December 27, 1974, as amended December 14, 1999.

1-6 Haddam Neck Updated Final Safety Analysis Report, dated August 8, 2000.

1-7 Letter B 16066 from CYAPCO to the USNRC, "Haddam Neck Plant Certifications of Permanent Cessation of Power Operation and that Fuel Has Been Permanently Removed from tile Reactor,"

dated December 5, 1996.

1-8 Letter CY-97-075 from CYAPCO to the USNRC, "Haddam Neck Plant Post Shutdown Decommissioning Activities Report," dated August 22, 1997.

1-9 USNRC Memorandum from Fairtile to Weiss dated January 28, 1998, regarding CYAPCO Post-Shutdown Decommissioning Activities Report.

1-10 Letter CY-98-005 from CYAPCO to the USNRC, "Decommissioning Updated Final Safety Analysis Report," dated January 26, 1998.

1-11 USNRC Safety Evaluation, related to Amendment No. 193 to Facility Operating License No.

DPR-6 1, Connecticut Yankee Atomic Power Company, Connecticut Yankee Atomic Powver Station, Docket 50-213, dated June 30, 1998.

1-12 USNRC Safety Evaluation, related to Amendment No. 195 to Facility Operating License No.

DPR-61, Connecticut Yankee Atomic Power Company, Connecticut Yankee Atomic Power Station, Docket 50-213, dated October 19, 1999.

1-13 NUREG-1575, "Multi-Agency Radiation Survey and Site Investigation Manual)," dated December 1997.

1-14 "Connecticut Yankee Haddam Neck Plant Characterization Report," dated January 6, 2000.

1-15 "Haddam Neck Plant Historical Site Assessment Supplement," dated August 14, 2001.

1-16 Code of Federal Regulations, Title 10, Part 20.1402, "Radiological Criteria for Unrestricted Use."

1-17 NUREG-0586, "Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities," dated August 1988, as supplemented on November 2002.

1-18 NUREG-1496, "Generic Environmental Impact Statement in Support Rulemaking for Radiological Criteria for License Termination of NRC-Licensed Nuclear Facilities," dated July 1997.

1-19 SECY 00-41, "Use of Rubblized Concrete Dismantlement to Address IOCFR Part 20, Subpart E, Radiological Criteria for License Termination," February 14, 2000.

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Hladdam Neck Plant License Terminiation Plan 1-20 Connecticut Yankee Atomic Power Company (CYAPCO) letter to USNRC dated July 7, 2000, and supplemental letters dated June 14, July 3 1, August 15, August 22, September 6, and September 7, 2001, and August 20 and October 10, 2002.

1-21 J. Donolew (USNRC) to K. J. Heider (CYAPCO), "Haddam Neck Plant - Issuance of Amendment RE - Approval of License Termination Plan (LTP) TAC No. MA979 I," dated November 25, 2002.

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Hladdam Neck Plant License Termination Plan 2 SITE CHARACTERIZATION 2.1 Introduction Initial site characterization of the Haddam Neck Plant (HNP) began following the permanent cessation of operations in the fall of 1997, and was completed in the fall of 1999. This initial characterization effort 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 characterization was performed to the guidelines of NUREG- 1575, "Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM)" (Reference 2-1). The l-ISA consisted of a review and compilation of site historical records, e.g., I OCFR50.75(g) records, radiological incident files, operational survey records, and annual environmental reports to the NRC. Personnel interviews were conducted with present and former plant employees and selected contractors to determine operational events that caused contamination in areas or systems not designed to contain radioactive or hazardous materials. Documentation from operational surveys, available through site document control facilities, was reviewed for historical information regarding radiological conditions throughout the site. The operational Radiation Protection Program provides continuing input regarding site radiological conditions. Measurements and samples beyond the scope of the operational survey program have been conducted in areas recognized as needing additional infomiation in order to assess the type, magnitude, and extent of contamination.

The site characterization program used the same QA practices as employed by the operational radiation safety program. These practices included the use of approved procedures for the calibration, testing and use of both laboratory and portable equipment. Trained and qualified personnel collected data. Samples were controlled by administrative procedures to ensure that sample integrity is maintained. When offsite laboratories were used, they were required to perform daily instrumentation checks. Other quality control measures for offsite laboratories included periodic method blanks, replicate (duplicate) samples and participation in an inter-laboratory comparison program (e.g., cross checks). Performance of these checks, by offsite laboratories, was verified periodically by QA auditors.

The objectives of a characterization program are to collect data adequate to:

I. Divide the HNP site into manageable sections or areas for survey and classification purposes;

2. Identify the potential and known sources of radioactive contamination in systems, on structures, in surface or subsurface soils, and in groundwater;

3. Determine the initial classification of each survey area or unit;

4. Develop the initial radiological and hazardous material informiation in support of facility dismantlement and remediation planning and radioactive waste disposal activities;

5. Develop the radiological information in support of Final Status Survey design including minimum instrument performance standards and Quality Assurance requirements; and

6. Identify any unique radiological or hazardous material health and safety issues associated with decommissioning.

August 2002 2-1 Rev. I

Iiaddarn Neck Plant License Terminationi Plan Characterization efforts at the HNP decommissioning project are an iterative process spanning all aspects of the remediation activities. The information developed during the initial HNP characterization program represents a radiological and hazardous material assessment based on the knowledge and data available at the end of 1999. This information wvas 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 compliance with the established Derived Concentration Guideline Levels (DCGLs), in accordance with the guidelines of MARSSIM.

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 not changed the dose modeling information from the LTP Revision I a. The decommissioning approach provided in Revision Ia of the LTP 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 of 2001. Section 2 of the License Termination Plan provides a summary of findings from thle 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 MNlethodology 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, IOCFR20, or I OCFR50; 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 information obtained through this process forms the input data for the records that are maintained on site August 2004 2-2 Rev. 2

Hladdain Neck Plant License Terminiation Plan to satisfy the requirements of IOCFR50.75(g)(1). The objective of the document reviews was to identify events that caused the contamination of systems, buildings, external surfaces, subsurface areas, or waterways, via atmospheric releases, liquid releases, or release of solid radioactive material. For each event, available supporting documentation regarding event description, facility and system design, radiological surveys and analysis, remediation efforts, and post remediation surveys was collected and reviewed. The CYAPCO nuclear records management system wvas the primary source of plant record information gathered during the HSA process.

To facilitate correlation of the impact of an event to physical locations on the plant site and to provide a means to correlate subsequent survey data, the owner-controlled area has been divided into areas with numeric designations. Figures 2-1 through 2-9 provide the area identification numbers for buildings and grounds within the owner controlled area. The area designations form the basis for survey units presented in Table 2-10.

In addition to the review of plant records, interviews wvith individuals involved in nuclear operations at HNP were conducted. Personnel interviewed included selected present and former employees and contractors involved in operations, maintenance, and radiation protection activities at the site. Information regarding unplanned releases or other events that could have resulted in site contamination was obtained through site staff and all-hands meetings, the daily plant newsletter, and Northeast Utilities system-wide publications. The effort was designed to ensure that historical events were identified that had an impact oln the radiological or hazardous material status of the site. Information gathered from the interviews was reviewed and included as appropriate in the IOCFR50.75(g) database. During the IISA, CY did not identify any time gaps in information for operational history.

In addition, CY has reviewved construction activities that resulted in redistribution of materials (soils).

Generally, materials that were removed from the Industrial Area or Radiological Controlled Area were placed on the Southwest Site Storage Area (Survey Area 9520), Central Peninsula Area (Survey Area 9530), Southeast Protected Area Grounds (Survey Area 9308) or the South East Landfill Area (Survey Area 9535). All these areas have received characterization surveys. Materials from construction activities that took place outside of the Industrial Area or Radiological Controlled Area generally have been released to offsite locations. These locations have been identified and surveyed under the Offsite Material Recovery Program (Section 2.2.4.2.4 contains further details).

2.2.3 Instruimentation Selection, Use, and M~inimium D)ctectable Concentration Radiological surveys performed in support of the initial site characterization were conducted by qualified Radiation Protection personnel. Surveys were performed by the station staff using instrumentation calibrated and maintained in accordance with station procedures utilizing National Institute of Standards and Technology (NIST) traceable calibration sources. The program consists of both periodic calibrations and response checks, when instruments are in use. The selection of instrumentation was based on the objectives of the surveys, expected radionuclide mix, and the ambient background in the area. Table 2-I identifies instrumentation typically used at the Haddarn Neck Plant. Where appropriate, typical efficiencies and Minimum Detectable Activities (MDA) at HNIP have been included.

Site characterization activities included land area surveys: within the protected area, of the landfill area, of selected portions of the Primary Auxiliary Building (PAB), of the containment and waste storage building, and surface contamination surveys of the paved areas inside the security fence but outside the Radiologically Controlled Area, and the Turbine Building floors at grade elevation and on the operating floor. The details of these surveys, including the instrumentation used and the minimum detectable concentrations (MDC) are included in the survey reports. Specific reports for these surveys are identified in the references section.

August 2004 2-3 Rev. 2 l

Hladdam Neck Plant License Termination Plan Table 2-1 Typical Instruiments Used at HNP Instrument Efficiencv MDA Probe I)etcction Units Characteristics (Nominal) (dpm) Type Eberline ASP-I NA NA GM (II P-270) 1y mr/hr Shipping instrument currently calibrated for gamma. This instrument can use various probes.

Eberline NA NA GM O3 mr/hr Dose rating E-130A laundry bags.

Eberline > 12% 700 GM Pancake Py Cpm1l Frisker, battery E- 140 operated.

Eberline NA NA G M ( 111-270) f3Y mr/hr Shipping E-520 instrument currently calibrated for

.gamma.

Eberline E-600 Scaler/count rate instrument with SHP-IOOCGS 15% 1000 Pa Cpmll various probes.

SHIP-300 NA NA PY piremr/hr SIIP-360 10% 60 P Cpnil SPA-3 15% 18000 Y cpm 239-1 F Floor Monitor 20% Varies Gas Flow A cpm Data Logger capability 30% with a cpm back-ground Eberline NA NA GM Tube Y mr/hr Shielded E-530N directional probe for high dose rates.

RSS-112 NA NA Pressurized Ion Sy tr/hr Low dose rate Chamber monitoring.

Eberline NA NA NA NA Pulses Used to calibrate MP-2 count rate instruments and AMS-3.

Eberline 12 - 17% 700 GM Pancake PY cpm Emergency Plan Ps- I instrument.

(Digital Scaler)

Used also to calibrate the Gamma- 10.

August 2002 2-4 Rev. I

Haddami Neck l'lant License Terminiation Plan Table 2-1 Typical Instriimenits Used at IINP Instrument Efficiencv 1 II)DA Probe Detection Units Characteristics (Nominal) (tlPmIl) Type Eberline 12- 17% 700 GM Pancake P cpIll Emergcncy Plan PS-2-2 instrument.

(Digital scaler)

Used also to calibrate the Gamma-10.

Eberline > 12% 700 GM Pancake p' cpnil Frisker, AC RM-14 powered.

Ludlum 19 NA NA Scintillation Y pr/hr Used for boundary sun cys, dose rates of trash, etc.

Ludlum 2200 17% P <1000P IIP-210 p Poa dpm Smcar counting.

I I%a < 100 a 43-2 a (Approx. 12 43-2 43-1 a probes and 4 43-1 probes.

Bicron Electra I B with > 20% p 1000 dual phosphor Pa cpm --100 cm sized DP6BD Probe scintillation probe

--Detects a & p simultaneously or

__ independently.

Bicron Electra I B with > 12% p 1000 dual phosphor P3a cpm Same as the DP6DD Probe 8% a scintillation DP6BD probe.

This probe has a double mylar window.

NE Technologies 10 - 30% <5000 Scintillation Y dpm Small article SAM-9 monitor.

Bicron Micro-Rem NA NA Scintillation y tr/hr Used for yard surveys and truck surveys. Pulscr instrument.

August 2002 2-5 Rev. I

Haddam Ncck Plant License Tcrminiation Plan Table 2-1 Typical Instruments Used at HNP Instrument Efliciencv Mi)A Probe I)etcction Units Characteristics (Nominal) (dpin) Type Exploranium Gammia NA NA Nal Y seiectable This instrument is Ray GR- 130 Minispec cpm to used only for In Situ Monitor mr/hr qualitative analysis and store up to 122 complete 256 channel spectra in memory. This instrument can be used in an operational mode as a survey meter and dose meter also. Data logger capability downloadable to a computer via serial link.

ISOCS - In Situ 40% variable Intrinsic Y variable Collimator.

Object Counting (intrinsic) germanium Analyzes spectra System data and generates reports.

CM-I 1 30-34% p 1000 DE-1 IA Gas Pa dpm Used mainly at 17-19% a 100 Flow control point for Proportional radon. 100 cm2 sized probe. One is on a portable cart. Pulser instrument.

APC 11 Counter 22%c a variable Gas Flow pa Counts low background 35% p _ Proportional scaler/counter XLB-I Counter 24% a variable Gas Flow P3a Counts low background 38% p _ Proportional scaler/counter Rev. 0 2-6 7/7/00 2-6 Rcv. O

Hladdami Neck Plant License Termination Plan 2.2.4 Results 2.2.4.1 Routine Releases Normal operations at l]NP resulted in releases of radioactive material through both liquid and gaseous pathways. Releases w'ere monitored in accordance with the requirements of the plant Technical Specifications. The routine gaseous release pathway was via the main stack, adjacent to the Containment Building. Routine liquid releases were made after sampling and analysis of the liquid in the test tanks.

Effluents were quantified and reported to the NRC in the Semiannual and later Annual Radioactive Effluent Release Reports, as required by the plant Technical Specifications. Analysis of the routine gaseous releases from the normal effluent pathwvays indicates typical short-lived radionuclides and inert gases typical of nuclear power plant operations. These gaseous releases did not result in site contamination and, therefore, do not impact the site relative to decommissioning activities.

2.2.4.2 Operational Events Information reviewed during the HSA identified several events that involved atmospheric releases, unplanned liquid releases, facility contamination and release of radioactive material. The following sections describe the major events in each of these categories and the site areas impacted by the events.

2.2.4.2.1 Atmospheric Releases During the initial years of operation at HNP, several events involving unplanned releases of airborne radioactivity occurred due to equipment failures or operator errors. During 1971 through 1979, these releases occurred in various areas of the plant, with the resulting radioactive discharges to the environment primarily involving inert gases and iodine. These releases were documented in station abnormal occurrence (AO) reports or plant incident reports (PIR), with the discharges included in the Semi-Annual Effluent Release Reports. As indicated by and discussed in the Site Characterization Report, these occurrences involved short-lived radioactive gases and iodines that did not result in contamination of site areas that would impact decommissioning activities. Examples of typical gaseous releases are listed in Table 2-2. Additional information is provided in the "Haddam Neck Plant Historical Site Assessment Supplement" and in the HNP 50.75(g)(1) files, as well as Survey Area files associated with areas impacted by various events.

August 2002 2-7 Rev. I

Haddami Neck Plant License Termination Plan Table 2-2 Examples of Unplanned Gaseous Release Events Date CYAPCO Location of Survcv Area Medium Event Description Reference Event Number Document (Figs. 2-1 No. through 2-19) 5/10/71 CYi-1-1686, Containment 9000 Gas Iodine release due to dated 5/10/71, 1-131 operator error with Letters dated penetration seals.

6/11/71, 6/18/71I, 6/22/71, 4/27/99 5/19/72 AO 72-2 Ion Exchange 9114 Gas Unplanned airborne Cubicles Xe-133 & release from Xe- 135 Dernineralizer due to operator error 4/26/73 PIR 73-134 Steam 3111 Gas Primary to Generator Secondary Leak, High RMS Alarm -

Air Ejector Monitor 6/21/73 AO 73-6 Ion Exchange 9114 Liquid/Gas 500 gallons leaked Cubicles to the Aerated Drains Tank due to failed valve. Gas release.

3/30/76 LER 76-8/3-L Waste Gas 3111 Gas Waste Gas Decay Decay Tank Tank rupture disk Cubicle failed 9/18/77 LER 77-21/IP "A" \Waste 2000 Gas "A" Waste Gas Gas Decay Decay Tank rupture Tank disk actuated 12/16/79 PIR 79-125 Main Stack 2402, 9120, Particulate, Degassifier rupture 9302, 9307, Stack Release disk release of 9310, 9312, primary coolant to 9514, 9527 main stack August 2002 2-8 Rev. I

Hladdam Neck Plant license Terminiation Plan Atmospheric releases that have impacted the site area include a series of events that occurred in 1979.

Stack discharges involving particulate activity occurred in February and December of 1979 as a result of the failure of a level controller in the letdown degassifier, flooding of the degassifier, and actuation of the degassifier rupture disk. The letdown liquid (primary coolant) then overflowed the degassifier. The discharge line from the rupture disk was routed directly to the main discharge stack. Efforts to clean the stack following the first incident may also have resulted in particulate releases. Surveys identified a number of localized areas of elevated activity within the Radiologically Controlled Area, within the fenced area of the plant site, and in the parking lot and hillside east of the plant. The results of the site surveys are documented in Reference 2-4. The majority of the radioactive particles were found on the roof areas of buildings close to the stack, and on the ground within the Radiologically Controlled Area.

The extent of area outside the Radiologically Controlled Area impacted by the releases includes the parking lot north of the industrial area, the hillside cast of the plant out to 200 meters, areas adjacent to tile discharge canal, the lower parking lot and Information Center. At that time, the isolated spots of contamination were remediated. The parking lot area has since been expanded (with the previous asphalt removed) and paved. Surveys of the parking lot conducted more recently (1997) have identified no contamination.

Atmospheric releases, following the particulate releases of 1979, included gaseous and iodine releases documented in the Annual Effluent Reports submitted in accordance with the plant Technical Specifications. The releases predominantly consisted of inert gases and radioiodines with short half-lives, radionuclides that do not impact the site relative to decommissionins.

2.2.4.2.2 Liquid Releases In addition to routine releases, spills through the storm drain system affected Survey Area 9522, Southeast Site Grounds, and ultimately reached the discharge canal, Survey Area 9106. Sampling of the discharge canal occurred in the fall of 1997. Sediment samples were obtained at seven (7) locations down to a depth of two (2) feet. In the northern canal, none of the samples had plant-related radioactivity levels greater than the corresponding DCGL. In the southern canal, no plant-related radioactivity was detected.

A review of the available Annual Radiological Environmental Operating Reports (AREOR) indicates that licensed material (e.g., Cs-134, Co-60) has been identified in the past, in the vicinity of discharge in the southern canal. All measurements indicated concentrations at a small fraction of the applicable soil DCGL (i.e., <5%) with the exception of a single measurement performed in September 1975, which indicated a concentration of 0.54+/-0.09 pCilg for Co-60. Howvever, decay correcting this concentration to the time of LTP submittal results in an expected concentration that is a small fraction of the soil DCGL.

The information available at the time of initial classification supported classification of the southern unit of the Discharge Canal (from the Canal Road to the river) as a Class 3 area. More recent information (i.e., the 1999 AREOR) reported a concentration of 0.50+1-0.10 pCi/g for Co-60, which is approximately 15% of the Co-60 "base case" soil DCGL. Given the higher potential for residual radioactivity, the entire discharge canal has been reclassified as a Class 2. This example demonstrates the CYAPCO Survey Area/Unit classification process, which includes a unit reclassification process that drives the most appropriate classification.

In addition to the spills identified above, the HSA identified a number of leaks and unplanned liquid releases that have occurred during the operational lifetime of HNP. The majority of the occurrences were confined to the Radiologically Controlled Area. The leaks and unplanned releases were associated with equipment failures and operational events associated with components within the Containment Building, Primary Auxiliary Building, outside storage tanks, and the radioactive waste processing systems. The most significant unplanned liquid release occurred in 1984, the result of a failure of the reactor cavity August 2004 2-9 Rev. 2

Haddani Neck Plant License Terminiation Plan seal. Although this event did not result in any release outside of the buildings it is considered the most significant because of the large volume of water involved. This water, released to the basement of containment, w'as pumped through the purification system to the Refueling Water Storage Tank (RWST).

Smears of the seal ring indicated 350 mR/hr gamma, 5.4 rad/hr beta with 120,000 dpm alpha. RWST radiation levels ranged from 65 to 200 mR/hr as the result of relocating water during this period. Surveys in the basement of Containment on August 23, 1984, indicated contamination levels to 295 mrad/hr beta and 6300 dpm/l 00 cm 2 alpha. Sampling of the yard drains concluded that no liquid or radioactivity was released to the environment.

Examples of typical events involving radioactive liquid leaks and unplanned releases are presented in Table 2-3. A record of unplanned liquid releases is maintained on site in accordance with the requirements of 10CFR50.75(g) and is identified in the affected area assessment in the Haddam Neck Plant characterization report (Reference 2-2) and in the Historical Site Assessment Supplement (Reference 2-3).

The principal impact of these events is to the ground within the Radiologically Controlled Area (RCA).

Migration of radioactivity to subsurface soils has occurred in the area of the Primary Auxiliary Building and the Refueling Water Storage Tank, adjacent to the Containment Building. Radioactivity has also been detected in groundwater wells on site. Assessments of radioactivity in soils and groundwater are further discussed in Sections 2.3.3.1.3 through 2.3.3.1.6. Radioactive materials from leaks have also impacted the area in the southeast corner of the protected area, a leachfield south of the protected area (but within the owner controlled area), and drain systems leading from the RCA. All of these areas were within the owner controlled area.

August 2004 2-10 Rev. 2

laddam Neck Plant License Termination Plan Table 2-3 Examples of Unplanned Liquid Release Events Date CYAPCO Location of Survey Area Medium Event Description Reference Event Number(s)

Document No. (Figs. 2-1 through 2-19) 11/1/73 AO 73-11 RWST 9110, 2110 Liquid Valve Leak, 270 liters of liquid released to storm drain 1/28/76 PIR 76-15 "A" Recycle 9108 Liquid 15 gal. of liquid leaked Test Tank from tank to diked area 5/22/76 LER 76-13/990 PAB Below 2228 Liquid Leakage from drain line Drumming below floor Room Floor 2/24/77 LER 77-2 "A" Recycle 9108 Liquid 1000 gal. of radioactive Test Tank water released to diked area around tank 2/23/79 PIR 79-27 Main Stack 9307 Liquid Manway leakage following SG blowdown rupture disk actuation. 20 gal. to yard area.

3/6/79 PIR 79-38 Main Stack 9120, 9307 Liquid Manway leakage following SG blowdown rupture disk actuation 3/28/83 PIR 83-37 Septic Tank 9520 Liquid 84 gal. of water from Chem Lab to Septic Tank 8/21/84 PIR 84-136 Containment 3002 Liquid Reactor cavity seal ring Building failure. 200,000 gals of water drained to lower levels of Containment Building 2/24/89 PIR 89-35 Leach Field, 9102, 9308 Liquid 50 gal. release from SFB 115 kV yard floor drain, line discharges to 115 kV yard 9/14/90 PIR 90-239 RWST 9110 Liquid 6 gallon per day leak from RWST identified from inventory monitoring 8/12/91 PIR 91-149 Pipe Trench 2110 Liquid 400 gal. release from open I _ I valve to pipe trench August 2002 2-1 1 Rev. I

liaddam Neck Plant license Terlliniationi Plan 2.2.4.2.3 Contamination of Buildings The HNP design includes several structures, engineered and constructed to contain radioactive material.

These structures include the Containment Building, the Primary Auxiliary Building, the Service Building, the Waste Storage Building, Ion Exchange Structure, Spent Resin Facility, and structures containing tanks for storage of radioactive liquids. Operations and maintenance activities in these buildings have resulted in surface contamination typical of nuclear power plants. Additionally, the HSA has identified a number of events that affected the radiological status of these structures. Following events resulting in internal structure contamination, decontamination activities were implemented based on ALARA considerations.

The decontamination efforts performed during normal operation were conducted to reduce occupational radiation exposure and were not undertaken to achieve site release conditions. The HSA did not identify any events that created an unexpected scope of contamination based on the design intent of these structures. Infornation gathered during the HSA did indicate that on two occasions-one in 1979 and one in 1989-the plant operated with failed fuel at a level that resulted in an increase in the level of alpha emitting radionuclides in the Reactor Coolant System. Events, as well as routine maintenance activities during these periods, increased the alpha emitting component of the source term.

The HSA identified primary-to-secondary leakage events resulting from steam generator tube leakage.

The events occurred during several operating cycles, with the first leakage identified in 1973 and the final events occurring in 1990. The leakage has resulted in measurable radioactivity in small areas of the secondary system piping, primarily in the high-pressure steam components within the Turbine Building.

2.2.4.2.4 Release of Radioactive Materials The Historical Site Assessment identified several events in which radioactive material was found outside the Radiologically Controlled Area. The primary locations of discovery were the southwest peninsula, Survey Area 9520, and the shooting range landfill area, Survey Areas 9535. These areas are currently controlled as "Radioactive Material Areas" with restricted access. During plant operations, the peninsula area has been used for storage of materials. The materials have been typically associated with maintenance activities performed during outages or plant modifications. Documentation indicates that the radioactive material was detected in 1980, 1985, and 1989. Surveys performed in 1998, indicated some previously stored materials contained detectable radioactivity. Since the shutdown of the plant, these stored materials removed from the peninsula area have been evaluated using the same process as material leaving the Radiologically Controlled Area. Surveys for free release have been required prior to that material leaving tile industrial area.

Other instances of minor levels of contamination being identified in areas outside the Radiologically Controlled Area but within the site boundary are documented in the Historical Site Assessment. The levels indicated in surveys are detectable, but typically are below the DCGLs. The areas affected were areas of high personnel traffic, such as the Administration Building or the Steam Generator Mock-up Building. Upon discovery, remediation occurred and more extensive surveys of the areas were performed. In 1997, an extensive survey was conducted of the material within those support buildings housing materials that may have, at one time, been in the plant. Any material or object with any indication of radioactive material was returned to the Radiologically Controlled Area for disposition.

Surveys of approximately 12,000 items located outside of the Industrial Area indicated only 23 items that had contamination that was greater than or equal to 1000 dpm as measured with a HP21o probe. The maximum contamination level found on an individual item was 8,000 dpm. None of these items exhibited detectable removable contamination. Approximately 175,000 items located outside of the RCA but inside of the Industrial Area were surveyed. Of these, only 105 items had detectable August 2002 2-12 Rev. I

Haddani Neck Plant License Termination Plan contamination greater than or equal to 1000 dpm by direct frisk. The maximum contamination level found on an individual item was 45,000 dpm. Three items with removable contamination were found in the operations tool cage of the Turbine Building. These were 1) on the inside of a small pipe reducer bushing - 2,600 dpm beta and 180 dpm alpha, 2) inside a l/2-inch pipe elbow - 200 dpm beta and 30 dpm alpha, and 3) inside a 1/4-inch pipe to ferruled adapter - 50 dpm beta and no alpha. During the course of the project, no other removable contamination was identified.

Table 2-4 Summarv of Unrestricted Release Confirimatorv Survey Program

1. of Items Items > or Maximum Area Surveyed =1000 Contamination Level dpm(3 and y in dpm)

Outside the Industrial Area Recycle Building 1667 I 2000 Warehouse 8716 21 8000 Miscellaneous 1381 1 1000 NNcithin the Industrial Area Sea Van Containers 23822 30 15000 Fabrication Shop 777 4 15000 Maintenance Shop 68051 23 20000 Electrical Shop 15270 3 4000 Butler Bldg -Clean 469 0 Control Room 75 1 3000 Turbine Bldg Areas 23126 15 10000 Gen Const Bldg 6330 1 6000 Weld Shop 8548 9 45000 I&C Shop 29137 9 15000 Records Vault 34 0 Additional events have been documented, indicating the release of potentially contaminated material from the site. Tile material in question has been primarily construction materials such as concrete blocks and excavated soils. These materials were addressed through the Offsite Material Recovery Program that was completed in 2003. Materials identified as radioactive and nonradioactive materials were appropriately dispositioned.

August 2004 2-13 Rev. 2

lladdani Neck Plant License Termiiiatiort Plan 2.3 Initial Site Characterization 2.3.1 Introduction Radiological characterization of the HNP site has been on-going since the plant began operation in 1968.

Radiological surveys and sampling for radionuclides have been conducted as part of a routine surveillance program in support of the plant radiological safety program, the environmental monitoring program, and in response to operational events. At the time of final shutdown and the cessation of nuclear operations in 1996, a substantial amount of radiological information existed. This information was evaluated to determine the need for additional data. The radiological informiation base was compared to the results of the HSA to determine the type and amount of newv or updated information that is necessary. Although radiological characterization wvill continue throughout the decommissioning process, an initial amount of characterization data is necessary to support preparation of the LTP.

The site characterization process focused on data for structures, systems and the site environs, considering radiological, hazardous and state-regulated materials. Groundwater is included in the assessment of the site environs. In addition, activation analyses have been performed on components and structures subjected to neutron flux, to support planning efforts.

The information provided in this section summarizes the characterization of the HNP site. The data is based on surveys, samples and analysis performed through the end of 1999 and is the planning basis for remediation activities, establishment of area classifications, and the development of the Final Status Survey Plan. The figures included in this section depict the plant site at the time of plant shutdown.

2.3.2 Methodology A Data Quality Objective (DQO) approach was applied to the characterization process. This approach focused the effort on gathering sufficient information to achieve the objectives identified in Section 2. 1.

The site characterization process began with the consolidation of information gathered during the HSA, radiological survey and sample analysis data maintained within the HNP document control program, and current site radiological data maintained by the Radiation Protection Department. This information has been augmented with the results from surveys performed in support of decommissioning operations contractor (DOC) proposal development. The HSA provides the basis for identifying suspect areas outside of those designed to contain radioactivity or those expected to be impacted by normal operations of a nuclear plant.

To facilitate the evaluation of information, available data wvere sorted by building or structure, and by land areas. A second sort of information wvas developed based on systems. A unique characterization report was developed for each building or structure with a cross reference to those systems contained within those areas. The characterization process was controlled by site procedures. These procedures established a consistent approach to the evaluation of each area for radiological, hazardous and state-regulated materials that are known to exist or are potential contaminants in structures, systems or soils within that area. The procedural process included the evaluation or determination of:

• the structure or area bounded by the evaluation;

• systems contained within the area, if any;

• HSA identification of events that may have impacted the area;

• area use, present and historic; August 2002 2-14 Rev. I

Haddani Neck Plant License Tcrmination Plan

• materials of construction;

• presence of radioactive or hazardous material storage areas;

• radiological survey information including extent of area containing radioactivity, radionuclides, exposure rates and contamination levels, both present and historic; and

• potential for migration of contaminants from contiguous areas.

The evaluation of each area included a walkdowvn of the area by a professional experienced in radiological hazards and a second professional experienced in hazardous and state-regulated materials.

Fuel cladding failures occurring during operations, primarily affected the Reactor Coolant System, as well as liquid systems that interfaced with the Reactor Coolant System. As the result of leaks during operation and refueling outage maintenance activities, it is acknowledged that Transuranics (TRUs) may have been released from these systems to surrounding rooms or cubicles. These rooms, cubicles and systems were monitored extensively during outages for personnel protection purposes, and the resulting data obtained from monitoring activities were used to classify them as Class I. The upper areas of the Containment and Containment Dome were initially classified as Class 2. These areas are currently planned for demolition and disposal as radioactive waste.

During the planning of characterization activities, Transuranics activity was considered in survey design by reviewing the operational history and operational surveys of the area. If TRU activity was suspected in an area, proper instruments were selected, smears counted for alpha activity and soil samples analyzed for Am-241, a predominant alpha emitter at the HNP.

Upon completion of the records review and the walkdown of each area, a characterization report for that area was completed. The characterization report for each area contains:

• a description of the area;

• survey units;

• summary of radiological, hazardous, and state-regulated material conditions within the area;

• classification of the area in accordance with the categories defined by MARSSIM;

• radiologically impacted systems within the area; and

• additional sampling and analysis necessary to support reniediation.

The area characterization reports have been compiled in the "Connecticut Yankee Haddam Neck Plant Characterization Report" dated January 6, 2000 (Reference 2-2). Additional information is provided in the "Haddam Neck Plant Historical Site Assessment Supplement" (Reference 2-3).

2.3.3 Site Cliaracterization/11SA Results 2.3.3.1 Radiological Status 2.3.3.1.1 Systems An extensive review of systems was conducted to determine those systems that contain radioactive materials or in which radioactive material was detected at sonic time during the operating history of the plant. Systems that are identified as "affected" require additional surveys to define the extent and magnitude of radioactivity. For those systems that may have been impacted due to steam generator tube leakage or other operational events in the past, but for which subsequent samples have not identified August 2004 2-15 Rev. 2

Haddam Neck Plant License Termination Plan radioactivity, the "affected" status is maintained. Table 2-5 provides a listing of plant systems and their status relative to the potential for radioactivity. The assessment considers the internal portions of the systems. Systems that might be assessed as "unaffected" and are located in contaminated areas may themselves be externally contaminated and may be considered for remediation or disposal as radioactive waste.

For those systems designed to contain radioactivity, such as the Reactor Coolant System and Radioactive Waste Processing Systems, the associated radiological conditions are continuously changing, with the most recent information necessary to support radiation protection activities maintained by the site Radiation Protection Department. These systems will be evaluated for remediation or disposal as radioactive waste based on economic evaluation of the alternatives.

Several components, such as the gland seal and turbine casing, have been identified as "affected" based on primary-to-secondary leakage identified in operating cycles as recent as 1990. These components contain low levels of radioactivity. The extent of the contamination, although appearing to be limited to small portions of the high pressure steam portion of the system, will be further defined as the systems are disassembled and the internal surfaces become accessible. These items are identified in the sections of the characterization report associated with the areas containing the systems.

August 2004 2-16 Rev. 2 1

Haddam Neck Plant License Termination Plan Table 2-5 Radiological Status of HNP Systems System Name System Contamination Survey Potential Comments Code Auxiliary Boiler 0001 A Identified radioactivity through sampling, 9/9/97 Auxiliary Feedwater 0002 B Blowdown 0003 B Boric Acid 0004 B Boron Recovery 0005 A Charging 0006 A Chemical Addition 0007 B System consists entirely of addition tank.

Circulating Water 0008 C Closed Cooling Water 0009 A Identified radioactivity through sampling, 9/3/97 Component Cooling 0010 A Identified radioactivity Water through sampling, 9/3/97 Condensate 0011 B Containment Air 0012 A Recirculation Containment Heating 0013 A Containment Iodine 0014 A Removal Containment Leak 0015 A Monitoring Containment Purge 0016 A Containment Rod Drive 0017 A Mechanism Cooling Control Air 0018 B Demineralized Water 0019 B Grouped system with Water Treatment for survey purposes as systems are closely aligned.

Diesel Generator 0020 C Domestic Water 0021 C A - System contains radioactive materials, known to be contaminated.

B - Radioactivity detected in portions of the system, surveys required to determine full scope of contamination C - System has no history of radioactive contamination, no samples that have indicated detectable contamination.

August 2002 2-17 Rev. 1

Haddam Neck Plant License Termination Plan Table 2-5 Radiological Status of HNP Systems System Name System Contamination Comments Survey Code Potential Feedwater 0022 B Feedwater Heater, 0023 B Extraction Steam, Drains, Vents Fire Protection 0024 C Floor/Equipment Drains 0025 A High Pressure Safety 0026 A Injection High Pressure Steam 0027 B Dump Isolated Phase Bus Duct 0028 C Cooling Letdown 0029 A Low Pressure Safety 0030 A Injection Low Pressure Steam 0031 B Dump Main Generator Seal Oil 0032 C Main Steam 0033 B Main Turbine & 0034 A Survey data indicates areas Auxiliaries of radioactivity Misc. Ventilation 0035 B Includes Turbine Bldg, Systems Cable Vault, New Diesel Gen Bldg, Aux Feed Pump Enclosure, Office Bldg and Service Bldg.

Nitrogen/Hydrogen/ 0036 B Carbon Dioxide PAB Ventilation 0037 A Post Accident Sampling 0038 A Primary Grade Water 0039 B Purification 0040 A Radiation Monitoring 0041 A RCP Seal Water 0042 A Injection Reactor Coolant 0043 A A - System contains radioactive materials, known to be contaminated.

B - Radioactivity detected in portions of the system, surveys required to determine full scope of contamination C - System has no history of radioactive contamination, no samples that have indicated detectable contamination.

August 2002 2-18 Rev. I

Haddam Neck Plant License Termination Plan Table 2-5 Radiological Status of HNP Systems System Name System Contamination Comments Survey Code Potential Residual Heat Removal 0044 A Roof Drains 0045 B Sampling 0046 A Septic 0047 B Service Air 0048 B Service Water 0049 B Fuel Building Exhaust 0050 A Spent Fuel Pool Cooling 0051 A Storm Drains 0052 B Turbine Lube/Control 0053 C Oil Vacuum Priming 0054 C Waste Disposal 0055 A Waste Gas 0056 A Waste Water Treatment 0057 B Water Treatment 0058 C A - System contains radioactive materials, known to be contaminated.

B- Radioactivity detected in portions of the system, surveys required to determine full scope of contamination C- System has no history of radioactive contamination, no samples that have indicated detectable contamination.

2.3.3.1.2 Buildings Radiological surveillances performed routinely at the HNP are designed to ensure compliance with IOCFR20 requirements regarding posting of areas, and to provide a basis for establishing controls for the safety of workers involved in those areas. The radiological information from the Contamination Verification Surveys (CVS) provides the basis for the demolition of Systems, Structures, or Components (SSCs) at the site. Once the initial remediation processes have adequately reduced the ambient radiation levels for controlled demolition, access and control of the SSC will be formally turned over to the demolition contractor. The current plan is for the demolition contractor to demolish the SSCs to 4 feet below grade elevation and leave any remaining foundations or less than 4 feet below grade if all of the structure and foundation are removed. In some of the structures all of the forms and foundation concrete will be removed within the 4 feet below grade elevation. In structures where portions of the foundations remain once demolition activities reach the 4 feet below grade elevation, the remaining concrete will be part of the final status survey of the land area before the area is backfilled to grade elevation with clean material. When basements remain below the demolition elevation, a final status survey will be performed on the portions that are to remain before they are backfilled with clean material.

The extent and nature of radioactive material in the primary structures on site, as identified in the "Connecticut Yankee Haddam Neck Plant Characterization Report" and Historical Site Assessment, are discussed in the following paragraphs. The planned scope for the disposition of each area is included.

Note, buildings that have already been removed from the plant site are not discussed. I August 2004 2-19 Rev. 2

Haddam Neck Plant License Termination Plan 2.3.3.1.2.1 Turbine Building Primary to secondary leakage has resulted in measurable radioactivity in small areas of the secondary system piping, primarily in the high-pressure steam components within the turbine building.

In the fall of 1997, systems in the Turbine Building were systematically sampled at several locations (e.g.,

sumps, filters, pumps, valves) to evaluate internal contamination levels. Scans and Total Surface Contamination measurements as well as smears were obtained at the access locations. Positive sample results ranged from 0.01 pCi/g up to several picocuries per gram of Cs-137 and Co-60.

Also in 1997, scoping surveys performed in the Turbine Building, covering more than 30,000 square feet of the surfaces of the operating floor and grade level identified only one small area of elevated activity.

That area, near the normal entrance/egress path to the Radiologically Controlled Area, was remediated at the time of the survey.

Surveys of accessible areas of the systems have shown fixed radioactive material in levels up to approximately 450,000 dpm/100 cm2 . Isotopic analysis has identified that Cs-137 is the principal radionuclide that has carried over in the steam following primary to secondary leakage. No alpha emitting radionuclides have been identified in any surveys for either fixed or removable radioactivity.

2.3.3.1.2.2 PrimnaryAuxiliary Building The Primary Auxiliary Building (PAB) is designed to house systems containing radioactive materials.

The building is designed to contain and control leakage occurring during routine operations as well as unusual conditions. The radiological status of the building is maintained by the radiation protection department staff through surveys performed in support of daily plant activities.

With the exceptions of the service water, primary de-ionized water, control air, fire protection, nitrogen gas and service air, all of the systems within the PAB are radiologically contaminated. Contamination levels in several of these systems are such that high radiation areas exist in their vicinity. Most of the cubicles that contain major systems are posted as contaminated areas identifying removable radioactive material.

The PAB contains pipe trench and pump pit areas, where conditions in these areas include high dose rates, possibility for high airborne contamination levels and alpha-emitting radionuclides. The lower level of the PAB under the boron recovery equipment is contaminated due to past releases of evaporator bottoms. The concrete surfaces making up these pit and trench areas will require remediation.

The PAB, fuel building and containment air handling systems all contain contaminated filter elements that will have to be removed and disposed of once these systems are declared abandoned.

The roof of the PAB has been radiologically impacted by historical plant events. Contamination of the PAB roof has occurred on multiple occasions due to emissions from the PAB roof exhaust and identified problems with the exhaust ducting. In 1980, an epoxy coating was used to fix contamination identified on the PAB roof.

Historically, leaks have been found at the junction between the steam generator blowdown line and the service water discharge line beneath the floor of the PAB drumming room. On at least one occasion, a leak has resulted in contamination of the soil beneath the drumming room floor.

August 2004 2-20 Rev. 2 1

Haddam Neck Plant License Termination Plan Surveys conducted indicated beta/gamma contamination levels in the PAB range from less than 1000 dpm/100 cm2 up to hundreds of thousands of dpm/100 cm2 . Alpha contamination levels range from less than 50 dpm/100 cm2 to several thousand dpm/100 cm2 Radiation levels in the PAB range from less than 5 mR/hr up to several thousand mR/hr. Based on the building design basis, events that have occurred within the building, historical survey information, and the present status of areas that are controlled as contaminated areas, much of the interior surfaces of the PAB are expected to contain radioactivity above the DCGL. Core bore information has identified that typical penetration depth is about 0.5 inch, however a depth of up to 2 inches has been identified in an area expected to contain contamination amongst the highest in the building. Most of the PAB is considered to be Class 1. The PAB is to be demolished in its entirety. The waste material will be sent offsite for disposal.

2.3.3.1.2.3 ReactorContainment Building The Reactor Containment Building houses numerous systems containing primary coolant as well as radioactively contaminated support systems. System leakage and maintenance activities over the operating life of the plant have resulted in radiological conditions similar to the Reactor Containment Buildings at other pressurized water reactors of similar vintage. As in the PAB, beta/gamma contamination levels in the Reactor Containment Building range from less than 1000 dpm/I00 cm 2 up to hundreds of thousands of dpm/100 cm2 . Alpha contamination levels range from less than 50 dpm/100 cm2to several thousand dpm/100 cm2 . Radiation levels in the Reactor Containment Building range from less than 5 mR/hr up to several thousand mR/hr. Some components, equipment, structural steel and concrete have become radioactive due to neutron activation. Addition characterization for activation will be conducted as the areas become accessible.

Concrete core bores were obtained in three locations in Survey Area 3104 at the I ft inch elevation.

The cores were cut into half-inch thick wafers and analyzed. The first core was in the floor. All of the radioactivity was detected in the top half inch of the core with a Co-60 concentration of 23.4 pCi/g, Cs-137 concentration of 279.0 pCi/g, and a Cs-134 concentration of 2.76 pCi/g.

The second core was taken on a vertical surface one (1) foot above the floor with essentially all the contamination found in the top half inch of the core. The Co-60 concentration was 1.68 pCi/g, Cs- 137 concentration was 13.66 pCi/g, and the Cs-134 concentration was 0.21 pCi/g.

The third core was taken on a vertical surface three (3) feet above the floor with essentially all the contamination found in the top half-inch of the core. The Co-60 concentration was 0.39 pCi/g and the Cs-137 concentration was 2.12 pCi/g. No Cs-134 was detected.

During 1998, a site characterization study was performed for polychlorinated biphenyls (PCBs), RCRA metals, and radioactivity in paints used on the structures on primary and secondary sides of the HNP.

Many paint-chip and concrete-chip samples from Containment were collected and analyzed by gamma spectral analysis. The purpose of the sampling program was to determine the extent of remediation required, and the waste management requirements due to potential PCB or RCRA metals in paint used in the Reactor Containment Building. Complete results of this study are provided in Reference 2-5. A summary of the results is as follows:

• The average concentrations of radioactivity in paint on the steel liner are about 1200 pCi/g on the charging floor level and about 300 pCi/g on the grade level. The primary radionuclides are Cs-137, Co-60 and Cs-134.

• The concentrations of radioactivity in paints on equipment vary markedly in both total activity and radionuclide distribution, depending on location and use. For example, the core barrel lift rig August 2004 2-21 Rev. 2 l

Haddam Neck Plant License Termination Plan contained approximately 25,000 pCi/g of Co-60 in paint while the polar crane contained 30 pCi/g of Co-60 in paint.

• The total radioactivity in floor paint averaged approximately 8200 pCi/g and is essentially the same from the charging floor, grade level and lower level.

• The total radioactivity in wall paint averaged approximately 490 pCi/g and is essentially the same from the charging floor, grade level and lower level.

• The radioactivity in the paint/concrete samples is greater than the radioactivity in the underlying concrete samples. The radioactivity in the paint/concrete and concrete only samples is greater on the floors than in the corresponding samples from the walls.

Based on the building design basis, and the operating history as well as the present status of areas that are controlled as contaminated areas, most areas of the Reactor Containment Building are considered to be Class I areas. The steel liner to 4 feet below grade and all interior walls of the Containment (previously designated as Class 2 areas) will be removed.The waste material will be sent offsite for disposal.

2.3.3.1.2.4 Radivaste Reduction Facility)

The Radwaste Reduction Facility is a structure used for staging and packaging various radioactive, and RCRA mixed waste streams. The Radwaste Reduction Facility contained radiologically contaminated items, both as radioactive waste and processing equipment. Therefore, the potential for residual contamination exists throughout the building. The Radwaste Reduction Facility historically contained a variety of equipment such as a waste shredder/compactor previously used for waste processing. The shredder/compactor is known to be internally contaminated. This equipment is typical of equipment used throughout the life of the facility since the primary purpose was the sorting and volume reduction of radioactive material. Additionally, there is a permitted lead work booth that is radiologically contaminated and may represent a mixed waste concern. The floors and the floor drains of the facility represent the primary concerns for residual contamination. The Radwaste Reduction Facility contains no plant-related process systems (such as service air, control air, etc.). The systems within the building are support systems such as electrical service and ventilation. Historical surveys of the building show contamination levels range from non-detectable up to 2000 dpm/l OOcm 2 beta/gamma.

The Radwaste Reduction Facility is being demolished to 4 feet below grade. The waste material will be sent offsite for disposal.

2.3.3.1.2.5 Fuel Building The Fuel Building is a structure designed for the storage of new and spent fuel. The spent fuel handling area encloses the Spent Fuel Pool and the equipment necessary for safe handling and storage of spent fuel.

The Spent Fuel Pool is a 36 ft long by 37 ft wide by 35 ft deep pool located in the northern half of the building. The pool is filled with borated water and contains storage racks for the spent fuel assemblies stored there. Highly irradiated reactor components and other debris are stored in the pool as well.

The Fuel Building contains radiologically contaminated items and process equipment. The potential for residual contamination exists throughout the building. Historical surveys of the building show that contamination levels range from non-detectable (ND) to greater than 100,000 dpm/O00cm 2 beta/gamma; alpha contamination levels range from non-detectable to greater than 500 dpm/lOOcm 2 . The highest alpha contamination levels were measured in the Spent Fuel Pool area.

August 2004 2-22 Rev. 2 1

Haddam Neck Plant License Termination Plan The Fuel Building will be demolished to 4 feet below grade and sent offsite for disposal. The Spent Fuel Pool liner will be removed and shipped for disposal. The remaining portions of the Fuel Building will be remediated and will undergo an FSS.

2.3.3.1.2.6 Waste DisposalBuilding The Waste Disposal Building contains handling and processing systems for liquid and gaseous waste streams. These systems include the Waste Gas Decay System, the Liquid Waste Evaporator Systems, the Floor and Equipment Drain System, the Degassifier System, and the Distillate System. All of these systems are contaminated, and many are currently posted as high-radiation areas. Some lines have hot spots where dose rates exceed 100 mR/hr. There is evidence of historical system leakage in many areas, and there are systems that are presently exhibiting leakage (for example, the Waste Evaporator Cooler Discharge). There is also evidence of boron in the Waste Gas System. Contaminated Floor Drains are located throughout the building. Highly Contaminated Waste Evaporator bottoms have leaked repeatedly onto the floor. Historical surveys of the building show that contamination levels range from non-detectable (ND) to greater than 100,000 dpm/lOOcm 2 beta/gamma and alpha contamination levels range from ND to greater than 500 dpm/IOOcm 2 .

The Waste Disposal Building will be demolished in its entirety. The waste materials will be sent offsite for disposal.

2.3.3.1.2.7 Miscellaneous Buildings Service Building Most of the Service Building areas are, and have always been, subject to routine health physics surveys.

Many years' worth of these survey records, covering the operating life of the plant, were reviewed to identify areas found to have had radionuclide contamination in the past. Such areas include: the sampling room located on the first floor of the Health Physics facilities building (contaminated via spills from the Post-Accident Sampling System), the clean side machine shop (Survey Area 5120 specifically),

the hot side machine shop, the decontamination room, the radiochemistry lab, the men's RCA shower area, and the myriad of miscellaneous cable trays and duct work that traverse the area above the radiochemistry lab. The area underneath the Health Physics facilities building is also suspect due to historical events (ruptured lines) that resulted in contamination of the soil beneath the floor of the drumming room located in the PAB.

It should be noted that the Service Building has been reconfigured on multiple occasions to either re-locate or add facilities, resulting in areas that were previously under radiological control becoming "clean" areas and areas that were previously "clean" coming under radiological control. An example of this is that the Health Physics Office Area was originally the Drum Storage Area and on occasion was contaminated. It is noted in interviews that long-time staff members assert that there are contaminated floor drains and drain lines under the floor of what is now the maintenance shop kitchen/break area.

The Service Building is being demolished to 4 feet below grade. The waste materials will be sent offsite for disposal.

August 2004 2-23 Rev. 2

Haddam Neck Plant License Termination Plan Penetration Building The Penetration Building includes the Terry Turbine Building. The building has three basic levels. The lower level room contains the steam-driven auxiliary feedwater pumps and feedwater piping. The middle level room contains atmospheric main steam dump valves encased by vented metal doors and walls. The upper level room contains main steam stop valves and an outside catwalk that extends to the Service Building located directly above main steam and feedwater transfer piping. The upper level also serves as the alternate access Health Physics control point to the Reactor Containment Building via an opening cut through the Reactor Containment Building wall. Reviews of routine surveys from 1998 and 1999 indicate no detectable contamination activity on smears taken from surfaces in the Penetration Building.

& The Penetration Building is being demolished to 4 feet below grade. The waste materials will be sent offsite for disposal.

Administration Building The Administration Building is located outside of the radiologically controlled area, contained no radioactive systems or materials. and has had an ongoing documented history of being a clean area. With the exception of the auxiliary boiler heating system, this building does not contain interrelated systems that provide the potential for being cross-contaminated with radioactive materials under normal operating conditions.

The Administration Building is being demolished to 4 ft below grade. The waste materials will be sent offsite for disposal.

Shutdown Auxiliary Feed Pump House The Shutdown Auxiliary Feed Pump House contains a pump, connected to the above-ground Demineralized Water Storage Tank. This building is an independent stu-cture with outer aluminum siding, inner sheetrock walls, slightly tapered sheet metal roof, asphalt floor, building exhaust ventilation, heat traced piping, electrical control panels/wiring and a concrete support base beneath the pump.

No significant radiological events have been identified that affect this area. The most probable mechanism for affecting this building with radioactive material is by transfer of loose surface contamination from personnel or equipment.

The Shutdown Auxiliary Feed Pump House is being demolished to 4 feet below grade. The waste materials will be sent offsite for disposal.

August 2004 2-24 Rev. 2

Haddam Neck Plant License Termination Plan Steam Generator Mlockup Building The Steam Generator Mockup Building, mobile trailer complex and industrial trash compactor are located in the clean area outside of the industrial area. Each of these locations has a history of being a radiologically clean area. However, during February 1997 it was determined that some camera equipment with fixed contamination (up to 1000 corrected counts per minute or ccpm) had been temporarily stored in the Steam Generator Mockup Building and was subsequently inadvertently released from the plant during January of 1997. These items were retrieved and returned to the plant RCA. Follow-up radiological surveillance of the Steam Generator Mockup Building established that all contaminated items had been removed and no other building areas were affected.

These buildings contain no systems that would provide the potential for being contaminated with radioactive material under normal operating conditions. The most probable mechanism for affecting this building with radioactive material has been by transfer of loose surface contamination.

The Steam Generator Mockup Building is being demolished to 4 feet below grade. The waste materials will be sent offsite for disposal.

Emergency Operations Facility The Emergency Operations Facility is located outside of the radiologically controlled area and contained no radioactive systems or materials. The decision to make the areas Class 3 is based on the proximity of this area to Class I or 2 impacted areas and the fact that personnel move freely between various areas of the site.

The present plan is for the Emergency Operations Facility to be demolished to 4 feet below grade elevation and removed from the site. If the structure remains, an FSS will be performed and documented.

North Warehouse The North Warehouse is located in a clean area outside of the protected area and has a history of being a radiologically clean area. This building had not been used for any radiological activity until 1999. Valve handles and bonnets removed from the Terry Turbine main steam system and air purification equipment have been stored in the north bay within a temporary RCA. All components have been placed in containers and subsequently placed onto spill-plex containment devices to prevent contact with the asphalt floor surface. Radiological results of items being stored in this area are as follows:

• 100-200 ccpm fixed beta/gamma contamination and < MDA beta/gamma/alpha loose contamination inside of #2 non-return valve operator support assembly (bonnet)

• 400 ccpm fixed beta/gamma contamination and < MDA beta/gamma/alpha loose contamination on the outside of #4 non-return valve operator support assembly (bonnet)

This building contains no systems that would provide the potential for being contaminated with radioactive material under normal operating conditions. The most probable mechanism for contaminating this building with radioactive material is by the transfer of loose surface contamination. Confirmatory surveys performed during 1998 on equipment and handtools within this area identified no contaminated items.

The North Warehouse is being demolished to 4 feet below grade. The waste materials will be sent offsite for disposal.

August 2004 2-25 Rev. 2 l

Haddam Neck Plant License Termination Plan South Warehouse The South Warehouse has a history of being a clean area, even though the southwest corner of this building has previously been controlled as a Radioactive Materials Area (RMA) for the storage of contaminated outage equipment. The RMA used approximately one-quarter of the building area and consisted of a lockable caged area with a painted epoxy asphalt floor surface. Historical data does not indicate the spread of loose surface contamination from items stored within this area to surrounding surfaces. The type of radioactive material typically stored within the RMA was fixed-contamination items and containerized loose-contamination items. The RMA was removed from this building and the area was free released during 1999. This building contains no systems that would provide the potential for being contaminated with radioactive material under normal operating conditions. The most probable mechanism for affecting this building with radioactive material is by transfer of loose surface contamination. Confirmatory surveys performed during 1998 on equipment and handtools within this area identified no contaminated items.

The fenced off area of pavement (approximately 40' by 80') adjoining the east side of the building has been used for storage of containerized radiologically contaminated soil. This area was established in support of the off-site remediation effort during 1998.

The South Warehouse is being demolished to 4 feet below grade. The waste materials will be sent offsite for disposal.

Training Stores Office Building No radiological-related activities have been known to occur within the Training Stores Office Building.

The building was initially used as a warehouse prior to being converted into classrooms and offices. The routine radiological surveys, that have been reviewed, indicate smear results to be less than 1000 dpm/100 cm2 for beta/gamma and less than 100 dpm/l 00 cm2 for alpha contamination. Annual random frisk surveys taken in 1998 and 1999 indicate no detectable beta/gamma activity. No major power plant systems are located in this building. None of the systems that are located within this building are anticipated to be contaminated.

The Training Stores Office Building is being removed to 4 feet below grade. The waste materials will be sent offsite for disposal.

Warehouse #1 Warehouse # I has a history of being a clean area. However, during February 1996, Containment Air Recirculation (CAR) fan charcoal filters having fixed and loose contamination were identified during a routine radiological survey of the area. The filters had been stored on a shelf location for a long period of time and believed to have been there since 1979. The filters were contained and returned to the plant RCA for proper storage.

Warehouse # 1 contains no interrelated systems that would provide potential for being cross-contaminated with radioactive material under normal operating conditions. The south end of this building contains a drive-through truck bay entrance into the industrial area. The most probable mechanism for affecting this building with radioactive material has been by transfer of loose surface contamination. Confirmatory surveys performed during 1998 on loose equipment and hand tools within this area did not identify contaminated items.

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___ _e_. ZUU4 Augs 4 o_

August :Z-:Z6 Kev. z

Haddam Neck Plant License Termination Plan Warehouse #1 is being demolished to 4 feet below grade. The waste materials will be sent offsite for disposal.

NN'arehouse #2 Warehouse # 2 has a history of being a clean area. It contains no interrelated systems that would provide potential for being cross-contaminated with radioactive material under normal operating conditions. The most probable mechanism for affecting this building with radioactive material has been by transfer of loose surface contamination. Confirmatory surveys performed during 1998 on loose equipment and hand tools within this area did not identify contaminated items. The soil under this structure may contain trace levels of radioactive material due to historical RCA runoff.

Warehouse #2 is being demolished to 4 feet below grade. The waste materials will be sent offsite for disposal.

Information Center The Information Center is located outside of the radiologically controlled area and contained no radioactive systems or materials. The Information Center is being demolished to 4 feet below grade. The waste materials will be sent offsite for disposal.

Screenwell Building The Screenwell Building is located outside of the radiologically controlled area, contained no radioactive systems or materials, and has had an ongoing history of being a clean area. With the exception of the auxiliary boiler heating system, this building does not contain interrelated systems that provide a potential for being cross-contaminated with radioactive materials under normal operating conditions. The major systems housed in the structure are the main circulating and service water pumps and their associated four Connecticut River water suction bays.

The screenwell house will be removed to 4 feet below the contour of the river bank and below the river bed. The waste material will be sent offsite for disposal.

2.3.3.1.3 Radiologically Controlled Area Grounds The Radiologically Controlled Area Grounds will have the asphalt removed. A final status survey will be performed of the areas.

The Radiologically Controlled Area (RCA) ground consists of paved areas around the Containment Building, Primary Auxiliary Building, Refueling Water Storage Tank and waste storage tanks, and the Fuel Building. The Historical Site Assessment identified events involving unplanned liquid releases that have radiologically impacted the area. Portions of the area have been posted as contaminated (removable contamination greater than 1000 dpm/1 00 cm2 ) due to system leakage. The contents of the tanks have caused radiation areas to exist. The paved areas served as the pathways for personnel movement between buildings and for vehicles moving materials, including radioactive waste. The area has also been used for temporary processing equipment in support of operations and maintenance activities during outages.

Radiological surveys performed during the plant operating years have identified areas of removable contamination. For example, a temporary cask washdown area was operated in the areas south of the Fuel Building. Contaminated pavement and soil resulting from its use have been identified and partially August 2004 2-27 Rev. 2

Haddam Neck Plant License Termination Plan remediated. Fixed contamination measurements have been limited due to the relatively high ambient radiation levels that currently exist in the RCA.

The area contains several drains for rainwater runoff. During several events involving leakage of radioactive materials in the area, samples of materials from the bottom of the drains identified detectable radioactivity. Details regarding those events impacting the grounds within the Radiologically Controlled Area are found in the Connecticut Yankee Haddam Neck Plant Characterization Report dated January 6, 2000. Historical surveys identified radioactive material in excess of the anticipated DCGLs.

Consequently, grounds within the Radiologically Controlled Area were classified as Class I areas.

2.3.3.1.4 Non-Radiologically Controlled Area Grounds The Historical Site Assessment identified that some material contaminated with radioactive material was placed in the shooting range landfill area along with construction debris. The assessment identified that between 1974 and 1996, construction materials from approximately 32 site projects had been placed on the landfill. Examples of materials identified are discharge canal dredging spoils, excavated soils, construction debris and sand. The materials originated from areas both within and outside the Radiologically Controlled Area boundary. A separate landfill area southeast of the borrow pit was permitted by the State of Connecticut for disposal of bulk wastes.

When certain construction projects were undertaken, bulk materials that were generated were transported and placed in piles in the shooting range area. As multiple projects were completed, the piles increased in height to the current estimated height of approximately 3 meters from the original elevation. Some bulk materials were deposited in the landfill area without a proper survey to evaluate concentrations or quantities of radioactive material. This issue was discussed in Inspection Report 50-213/97-08 and in CYAPCO's December 5, 1997, reply to the notice of violation contained in the inspection report.

LER 97-017 also discussed this issue and was submitted on November 18, 1997. Based upon the Historical Site Assessment, there are no known areas on the HNP site that were excavated for the purpose of burial of radioactive material.

The site characterization group performed radiological surveys of the landfill area with confirmatory surveys conducted by ORISE (Oak Ridge Institute for Science and Engineering). The initial characterization surveys were completed in 1997. Results of the radiological survey and ground penetrating radar survey established the size of the landfill area to be approximately 5000 square meters and an approximate thickness of 3 meters. This currently fenced area is Survey Area 9535 located about I mile southeast of the Containment Building on higher elevations between the Salmon River and the discharge canal on owner controlled property.

Radiological results indicated the presence of both Co-60 and Cs- 137, with maximum concentrations in soil of 5.0 pCi/g and 52.9 pCi/g respectively. A study was completed to determine the extent and magnitude of the plant-derived radionuclides or hazardous materials that might have accompanied the construction debris into the landfill area. This radiological characterization consisted of:

• Gamma scan surveys centered on 9 m2 areas;

• Exposure rate measurements at one (1) m above the surface; and

• Soil samples collected for analysis from the surface, and at depths within the ranges 0-1 m, 1-2 m, and 2-3 m.

August 2004 2-28 Rev. 2

Haddam Neck Plant License Termination Plan The results of the survey identified eight (8) samples with positive Co-60 results between the minimum detectable activity of 0.15 pCi/g (environmental LLD) and a maximum of 2.8 pCi/g. Additionally, 40 samples measured positive Cs-137 results between the minimum detectable activity of 0.18 pCi/g (environmental LLD) and a maximum of 31.4 pCi/g. Only 10 of the 40 positive Cs-137 results were greater than the local background level of 1.68 pCi/g due to historical fallout from weapons testing. The gamma scan results showed a variance of a factor of 4 or 5 times the lowest levels that are indicative of background. The exposure rate measurements ranged from 9.1 to 15.6 prem/hr. The scan and exposure rate measurements both indicate the presence of low levels of radioactive material but are not able to establish if the variance is due to plant derived radionuclides or represent variations in naturally occurring radionuclides.

A hazardous materials assessment was also performed in the survey area. Two composite samples were analyzed for the Envirocare Suite of analytes. A 10 CFR 61 series of analyses were also performed on the composite samples. The test results indicated that all the analyses were below the Connecticut Department of Environmental Protection Remediation Standards for soil except for the chemical thallium.

Additional investigations will be performed. The level of thallium is above the residential limit, but below the industrial limit. Co-60 and Cs-137 were the only plant-derived radionuclides present in the composite samples.

These areas have been remediated of landfill materials to support removal of the State of Connecticut permit. An FSS, incorporating additional characterization information, has been performed for each area.

In December 1997, surveys were performed of paved areas within the industrial area but outside the Radiologically Controlled Area (specifically Survey Areas 9302, 9304, 9306 and 9308). These surveys identified seven discrete areas of contamination ranging from 10,000 to 65,000 dpm, each limited to less than 100 cm2 . The seven discrete areas were found in Survey Areas 9302, 9306 and 9308. Information concerning these surveys is provided in Reference 2-6. The radioactivity identified was remediated prior to completion of the survey. During 1998 and 1999 over one hundred asphalt and subsurface soil samples, in some cases down to six feet in depth, were collected in the same areas in support of plant modifications and site characterization activities. None of the samples had plant related radioactivity levels greater than the corresponding DCGL. Their classifications are presented in Table 2-10. Because of the use of the Survey Area 9308 for storage of the steam generator lower assemblies and the pressurizer, it was classified conservatively as a Class I area.

Land areas adjacent to the industrial area had been surveyed in response to the events occurring in 1979, discussed in Section 2.2.4.2.1. Subsequent surveys were performed in the land areas adjacent to the industrial area fence. The criteria for surface area coverage was to survey to a distance from the plant twice that of the furthest point where localized areas of elevated activity were detected. These areas of localized elevated activity were detected and removed from the area close to the industrial area fence, with few areas found beyond 100 meters. The details of the surveys and the results are presented in Reference 2-4.

Additionally in 1989, Warehouse #2 and the Office Building #3/PAP were built and the primary parking lot was re-configured. The primary parking lot was enlarged, re-paved and storm drains installed. Only the parking lot asphalt, excluding the road way, and near surface soil was removed for the reconfiguration. The storm drain installation helped reduce a long-standing problem with drainage in the southeast corner of the lot. The storm drains were installed that routed run off from the parking lot and other nearby storm drains to the pond. Other storm drains that connect to this drain header include: storm drains located on the south side of the Office Building #3/PAP, a drain from the drainage ditch on the east side of the parking lot and roof drains from Warehouse #1, #2 and the Office Building #3/PAP. The August 2004 2-29 Rev. 2

Haddam Neck Plant License Termination Plan parking lot storm drain header in turn drains to the pond located north of the parking lot. An overflow in the pond discharges to the river.

Eight wells were installed, in support of a proposed repowering of the site, in August 1999. Six of the eight wells (MW-502, -503, -504, -505, -507 and -508) were installed in the parking lot. Samples obtained from the well installations were analyzed by gamma spectrometry. Table 2-6 summarizes sample data obtained during the well installation.

Water samples were obtained from several of the wells in August 1999. Water samples were analyzed by gamma spectrometry, gross alpha, gross beta and H-3. Table 2-7 summarizes the sample results. No water sample results are available for wells MW-501 and -506.

Table 2-6 Monitoring Well Soil Sample Data (Through August 1999)

I Sample Control Location Co-60 Cs-137 No. pCi/g pCi/g Soil, 0-6' 990806004 MW-501 <0.036 0.076 Soil, 6-37' 990806005 MW-501 <0.033 <0.029 Soil, 0-6' 990909012 MW-502 <0.035 <0.037 Soil, composite 990906013 MW-502 <0.074 <0.043 Soil, 40-70' 990810014 MW-502,-503 <0.074 <0.034 Soil, 0-6" 990728014 MW-503 <0.050 <0.029 Soil, 990723005 MW-504 <0.036 <0.034 Soil, 990723006 MW-504 <0.041 <0.037 Soil, 990726009 MW-504 <0.041 <0.031 Soil, 15-50' 990726008 MW-505 <0.058 <0.047 Soil, 0-6" 990727007 MW-505 <0.055 <0.032 Soil, 0-6" 990730005 MW-506 <0.064 <0.055 Subsurface soil 990730006 MW-506 <0.061 <0.058 Soil, 5-10' 990804001 MW-507 <0.082 <0.071 Asphalt 99080301 1A MW-507 <0.047 <0.040 Soil 990803012 MW-507 <0.033 <0.040 Asphalt 990803009 MW-507D <0.035 <0.027 Soil 990803010 MW-507D <0.031 <0.036 Asphalt 990813004 MW-508S <0.064 <0.034 Soil, 0-15' 990813006 MW-508S <0.057 <0.053 Soil, 0-6' 990812003 MW-508 <0.048 <0.056

'Soil 990812002 MW-508 <0.029 <0.053 Asphalt 990812001 MW-508S <0.056 <0.055 August 2004 2-30 Rev. 2 I

Haddam Neck Plant License Termination Plan Table 2-7 Well Water Results (Through August 1999) l Vell # Gamma Spec. H-3 Gross Gross Co-60 Cs-137 Control # pCi/t Beta Alpha pCi/I pCi/I l__

_pCi/i

__ _ _ pCi/ _

MW-502 G85642 <67 7.8 <3.2 <5.4 <6.1 MW-503 G85644 <67 <0.86 <1.5 <6.0 <5.8 MW-504 G85646 <67 144.8 ( <1.9 0) <3.3 <4.6 MW-505 G85649 <67 6.5 <2.9 <7.2 <6.2 MW-505 G85648 <67 6.63 <2.1 <7.9 <6.0 MW-507S G85640 <67 6.32 <1.9 <5.5 <6.0 MW-507D G85656 <67 8.35 (2) 15.2 (2) <5.2 <4.7 MW-508S G85650 <67 9.83 <2.9 <5.5 <5.2 MW-508D G85651 <67 10.9 3) <3.8 <3.6 (1) Sample MW-504 was later recounted and results were 8.5 pCi/l gross beta and <2.2 pCi/l gross alpha.

(2) Sample MW-507D was later recounted and results were 13.5 pCi/i gross beta and 13.0 pCi/l gross alpha.

(3) Sample MW-508D was later recounted and results were 19.8 pCi/l gross beta and 8.68 pCi/I gross alpha.

2.3.3.1.5 Subsurface Soils In this context, surface soil refers to outdoor areas where the soil is, for purposes of dose modeling, considered to be uniformly contaminated from the surface down to some specified depth. Subsurface radioactivity refers to residual radioactivity that is underneath structures such as building floors/foundations or material that is covered with clean soil or some other unaffected layer(s).

The historical site assessment was consulted to identify those survey areas where the potential exists for subsurface radioactivity. Such areas include, but are not limited to, areas under buildings, building floors/foundations, or components where leakage was known or suspected to have occurred in the past; on-site storage areas where radioactive materials have been identified; and areas containing spoils from past dredging of the discharge canal. Soil data from both the historical site assessment and any pertinent characterization data will be used to establish a bounding depth profile for any potential sub-surface radioactivity. However, the assessment of all subsurface soil contamination is not currently complete.

Soil in difficult to access areas such as under tanks will be deferred until later in the decommissioning process, when access will be more readily available.

During 1998 and 1999 over one hundred subsurface soil samples, in some cases down to six feet in depth, were collected in Survey Areas 9302, 9304, 9306 and 9308 in support of plant modifications and site characterization activities. None of the samples had plant-related radioactivity levels greater than the corresponding base case soil DCGL. During the same time period, over two hundred subsurface soil samples, in some cases down to approximately two meters in depth, were collected inside the RCA in Survey Areas 9307, 9310, 9312 and 9227 (Bus-10 Pad and ground underneath). Some isolated locations showed Co-60 and Cs-137 activity levels up to several hundred pCi/g each, under the Bus-10 pad (Survey Area 9227). Most of the sampling was performed in Survey Areas 9310 and 9227 in support of Spent August 2004 2-31 Rev. 2

Haddam Neck Plant License Termination Plan Fuel Pool isolation. The soil was removed until residual radioactivity was below the generic soil screening DCGLs, adjusted to 10 mrem/yr. Groundwater dose contribution was not included in this evaluation.

The subsurface soils at CY are divided into 3 classifications: Class A, Class B and Class C. Class A soils have had known contaminating events and have a high potential to be at or exceed the DCGL. Class B soils have had contaminating events or may have been impacted by events in Class A soils but are not expected to exceed the DCGL. Contamination levels in Class C soils are expected to be a small fraction of the DCGL.

The classification guides the number of measurements or samples to be taken. Class A soils will receive the highest number of samples while Class C will receive the lowest. There are no limitations on the size of the area. Subsurface soils in the Radiological Control Area, with the exception of the soil beneath Survey Area 9308, are considered as Class A (the size of this area is reflected in the dose modeling for soil). Due to large number of subsurface samples already collected and analyzed in Survey Area 9308, subsurface soils beneath Survey Area 9308 and in the remainder of the Industrial Area outside the RCA are considered Class B. Areas northwest and southeast of Class B areas are considered Class C.

Figure 2-9 provides the designation and location of affected subsurface soil in the vicinity of the plant power block.

Subsurface soils contaminated above the applicable DCGLs will be removed. Groundwater dewatering wells, approved by the State of Connecticut, are being installed at strategic locations to draw down groundwater as necessary to support contaminated soil and structure removal activities.

2.3.3.1.6 Groundwater Additional groundwater monitoring results, as well as updated hydrogeologic information and changes to workplan deliverables, has been and will continue to be submitted to the CT DEP as a part of the Phase 2 Work Plan. This information is also being provided to the NRC. Considering this, the following sections of the LTP will generally not be revised to include the information from these reports.

2.3.3.1.6.1 Initial Groztmd vaterMonitoringProgramand Cliaracterization As part of the site characterization efforts and to address issues related to leakage of radioactive liquids from the Refueling Water Storage Tank (RWST), a groundwater monitoring sampling and analysis plan was developed in March of 1998. An investigation was conducted through May of 1999, with the primary purpose of radiological characterization of groundwater in four geographically separate areas of interest. The initial areas of interest were the power plant area, the peninsula water supply area, the Emergency Operations Facility area, and the landfill area. Results of this investigation are provide in the Groundwater Monitoring Report, revised on September 1999 (Reference 2-7). Since that time, a Phase 2 Hydrogeologic Investigation Work Plan (Reference 2-8) has been developed that identifies radiological release areas based upon the Historical Site Assessment and other related studies, delineates background wells for comparison of groundwater quality (includes naturally-occurring radiological and geochemical) within the plumes, and has been expanded to include the parking lot and surrounding warehouses, and two offsite properties. Three identified episodic releases of water containing plant-related radionuclides and boron have resulted in groundwater plumes that have migrated through the bedrock towards the Connecticut River, particularly tritium and boron which are much more mobile than the metallic radionuclides. These releases have ceased, and the plumes have shown a marked attenuation since periodic monitoring began on March 1999.

August 2004 2-32 Rev. 2

Haddam Neck Plant License Termination Plan The Phase 2 Hydrogeologic Final Work Plan contains a complete discussion regarding well installation, sampling protocol, groundwater modeling, and reporting. As discussed in the Task I Phase 2 Hydrogeologic Investigation Work Plan, monitoring wells were installed to characterize the water table and shallow bedrock aquifers near potential contaminant source areas. Six rounds of sampling, March 1999 through December 2001, were analyzed for dissolved boron, tritium, gross alpha, gross beta and gamma spectroscopy. The December 2001 sampling results revealed that tritium was quantified above detection limits (Minimum Detectable Activities of 700 to 1000 pCi/1) at 9 of 40 monitoring well locations (II of 40 previously). The elevated tritium concentrations, detected only in those wells in the plant area, ranged from 1,800 pCi/I (located between the Primary Auxiliary Building and the Turbine Building), to 21,300 pCi/I (located at the entrance to the peninsula area and adjacent to the former leaching field area and in the groundwater underlying the Industrial Area). Cs-137 was previously detected in I of 40 monitoring well locations, again in the power plant area. However, the December 2001 sampling results did not indicate the presence of Cs-137 at detectable levels in any of the wells. No other gamma emitting radionuclides were above the minimum detectable levels. Beta-emitting Sr-90 and Tc-99 have been detected in 3 wells and I well, respectively, in the area of the Reactor Containment Building. Boron, a naturally present chemical element as well as a power plant chemical additive, was found at concentrations above background at 10 of the 40 well locations (maximum of 2,400 ug/l), and those values roughly correlated to where tritium was identified.

Initial sampling results, along with initial results of geologic and hydrogeologic characterization activities, are documented in Reference 2-7. This report has been submitted to the Connecticut DEP and has been provided to the Nuclear Regulatory Commission and the Environmental Protection Agency.

The primary conclusions of that report are as follows.

• The general flow of groundwater is from north to south, or from hillside to river across the plant site, primarily within the fractured bedrock.

• The groundwater migrates down under and around the deep foundations of the plant structures in complex patterns and then moves upward again to the river.

• Areas such as the Emergency Operations Facility, water supply well and landfill areas show no tritium above minimum detectable activity and no boron above background levels. No contamination in any of the water supply wells has been detected. Although there was a very low level detection of Sr-90 in the water supply well in January of 2002, this detection was not confirmed by subsequent sampling of that well.

• Groundwater beneath and immediately around the power plant buildings has been affected by boron, H-3 and Cs-137. Subsequent sampling also indicates the presence of Sr-90 and Tc-99.

• The likely source of plant-generated radionuclides in the groundwater is the RWST. This source was eliminated when the RWST was drained. However, the Chemistry and Laboratory drain system and Primary Auxiliary Building Liquid Waste Discharge line failures and leaks are also possible source for radionuclides in the groundwater and, accordingly, will be fully investigated.

• More groundwater data and monitoring wells are needed to identify the bottom and core of the plume.

A more detailed discussion of the extent of plant-generated radionuclides in the power plant area is provided in the following discussion.

Rev. 2 2-33 August 2004 August 2004 2-33 Rev. 2

Haddam Neck Plant License Termination Plan Tritium Tritium is a radionuclide created in the reactor water as a normal by-product. Because it behaves like water, its hydrologic behavior and subsurface migration are the same as groundwater - it is not subject to the physical/chemical factors affecting the movement of dissolved chemicals, such as volatilization, retardation, or chemical degradation.

Figure 2-1 1 shows the observed tritium concentrations and extent of tritium as reported in Reference 2-7.

Based on observed concentrations and their spatial distributions, there have been three possible sources of tritium:

• The RWST located north of the Containment Building. Tritium has been observed at both MW-103S and MW-103D. Tritium has not been observed at sampling points upgradient of the RWST (i.e.,

MW-101).

• A potential source near well cluster MW-105, located southwest of the Containment Building.

• A minor, potential source near well cluster MW-I 02 and MW-I 14S.

The tritium from the former RWST source appears to be flowing downward and towards the Containment Building, with much of it captured by the mat drain sump. However, due to the complexity of flow patterns around the Containment Building, this source may have contributed to other downgradient plumes.

From MW- 105, a portion of the tritium is migrating in both the shallow and deep groundwater horizons southeastward past MW-106 to MW-107. The direction of the tritium appears to be controlled by the presence of the discharge tunnel that blocks shallow groundwater movement to the west. A downward component has carried some of the tritium into the bedrock at well MW- 106D.

A second, larger portion of the tritium from the MW- 105 area migrates downward underneath the discharge canal. It continues southwestward and southward towards well clusters MW-109 and MW-I I0, which are adjacent to the Connecticut River. The presence of local downward gradients, discharge canal and foundation walls combine to divert the plume into the bedrock under the turbine building where it eventually is carried upward by gradient toward the river, where it eventually discharges.

Cesium- 137 Cs-137 has been observed only at well MW-103S. The presence of Cs-137 coincides with the tritium findings and apparently originated from the RWST. Cs-137 concentrations in soil have been observed to decrease with depth, indicating a residual surficial source. The Cs-137 is not as mobile as tritium and should be retained in soils near the release area, which is consistent with the observed distribution. Since cessation of the mat drain sump pump, Cs-137 has not been observed at MW-103S. This is likely due to the re-equalization and distribution of constituents.

Strontium-90 and Technetium-99 In order to confirm that no hard-to-detect radionuclides were present in the groundwater under the CY site, in the summer of 2001, a number of well samples were analyzed for radionuclides known to have been present in the reactor water at CY. Analysis was conducted for Iron-55, Nickel-63, Strontium-89/90, Technetium-99, Americium-24 1, Plutonium-23 8/239/240/241 and Curium-243/244. The August 2004 2-34 Rev. 2

Haddam Neck Plant License Termination Plan radionuclide that was present at any significant level was Strontium-90 at a concentration ranging from 120 to 143 pCi/liter in one well (MW-l05S) near the Reactor Containment Building and adjacent to the area where the Chemistry and Laboratory drain system pipe leaks and weir box overflows were known to have occurred. Technetium-99 was present at a concentration of 3.9 pCi/liter in wvell MW-103D near the RWST.(Note: A follow-up sampling did not confirm the presence of Tc-99). Both areas are being fully investigated as potential sources, and a remedial plan of action will be formulated accordingly, once additional sampling data (soil and groundwater) are available from the work described in the Phase 2 Hydrogeological Investigation Work Plan. See Figures 2-10 and 2-11.

Groundwater Level and Concentration Trends Tables 2-8, "Groundwater Level Elevations" and 2-9, "Temporal Trends in Groundwater Radionuclide Concentrations," summarize the groundwater level and radionuclide results of six rounds of sampling at key monitoring locations. As the data indicate, levels are generally decreasing with time. This trend is consistent with the removal of the source of contamination (i.e., draining of the RWST). Groundwater monitoring will continue on a periodic basis during decommissioning to confirm this trend.

Table 2-8 Groundwater Level Elevations (Through January 2002)

GROUNDWVATER LEVEL ELEV. (ft above mean sea

• 'eil or Grade Refrernce Ie cl)

Location Elevation Elevation es 03/02/99 1 04/06199 1 06101100 T 07/20/01 01/08/02 Plant and Northern Peninsula Area WVells MW-10IS 20.86 20.54 16.14 15.49 15.24 14.42 14.76 MW-1I0D 20.86 20.53 12.73 12.38 10.93 8.64 5.43 MW-102S 20.81 16.94 8.62 8.47 7.37 8.83 5.65 MW-102D 20.48 20.06 8.86 10.76 9.16 3.88 17.51 MW-103S 21.18 20.85 12.77 12.42 10.75 8.66 5.49 MW-103D 21.16 20.95 9.92 10.45 -1.17 7.57 14.86 MW-104S 20.31 20.16 14.58 13.45 no data 10.75 8.23 MW-105S 20.84 20.59 8.74 9.54 9.09 7.86 4.27 MW-105D 20.87 20.58 12.15 12.01 10.33 8.52 5.21 MW-106S 20.76 20.51 9.78 10.21 no data 7.64 4.58 MW-106D 20.85 20.63 8.93 9.90 6.65 6.98 4.29 MW-107S 20.73 20.35 7.56 8.40 7.25 6.10 3.22 MW-107D 20.73 20.48 8.10 8.91 7.18 5.71 3.43 MW-108S 10.30 12.23 4.86 6.31 no data 4.53 2.42 MAW-109S 20.80 20.57 3.15 4.69 2.89 2.48 2.01 MW-109D 20.81 20.50 4.28 5.46 4.25 3.82 2.59 MW-I lOS 20.08 22.49 2.75 4.64 1.65 1.79 1.38 MW-11OD 20.18 22.85 3.95 5.37 2.90 2.79 1.90 MW-1 I IS 15.39 18.11 2.49 4.32 1.43 0.91 1.38 MW-112S 12.28 14.33 2.40 4.23 no data 0.94 1.55 MW-113S 11.49 13.48 2.27 4.33 no data 1.18 1.68 August 2004 2-35 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-8 Groundwater Level Elevations (Through January 2002)

Well or Grade Reference GROUNDWATER LEVEL ELEV. (ft above mean sea Location Elevation Elevation Icvel) 03/02/99 04/06/99 06/01/00 07/20/01 01/08/02 MW-114S 21.04 20.73 11.01 11.23 9.36 5.15 4.58 MW-115S 21.17 20.73 10.80 11.20 9.23 8.08 4.39 AST-I 21.55 23.59 7.04 7.16 6.69 6.59 5.06 Mat Drain Sump 21 21.5 -21.5 -25.5 -24.2 -21.9 4.3 Parking Lot Wells MW-100S 16.67 16.45 13.91 14.97 no data 12.76 13.42 MW-100D 16.69 16.47 16.47 15.92 14.97 12.56 12.76 MW-502 18.45 18.18 no data no data no data no data 3.12 MW-503 15.77 15.43 no data no data no data no data 3.10 MW-504 17.14 16.81 no data no data no data no data 3.35 MW-505 15.57 15.17 no data no data no data no data 11.02 MW-507S 19.02 18.67 no data no data no data no data 11.48 MW-507D 19.02 18.59 no data no data no data no data 4.30 MW-508S 18.96 17.84 no data no data no data no data 10.66 MW-508D 18.22 17.81 no data no data no data no data 2.82 River at boat dock NA 8.00 2.00 3.10 no data no data 0.89 Pond NA 15.00 11.23 11.00 no data no data 10.90 EOF Area Wells XX E-OF-1 24.45 24.09 16.03 j14.79 no data jndaa no data EOF-2 24.52 24.12 15.97 14.72 no data j 13.92 j 10.71 W_'ater Supply WVell Area W\'ells MW-117S 13.2 15.7 6.1 5.8 no data 4.00 1.44 MW-13 (C) 17 20.5 3.7 3.8 no data 1.36 1.15 TW-I (D) 17 18 3.7 3.7 no data 1.69 1.29

_Landfill Area Wells MW-200 52.18 54.67 39.19 37.4 no data 38.84 dry MW-201 55.96 58.78 29.61 28.37 no data 28.34 22.88 MW-202 49.67 51.62 38.42 38.77 no data 38.14 33.37 MW-203 44.51 46.23 37.8 37.36 no data 37.41 33.25 MW-204 39.96 41.85 36.03 36.15 no data 35.90 32.71 MW-205 38.66 40.55 33.13 32.78 no data 33.13 30.39 MW-206 41.15 43.08 35.51 36.13 no data 35.34 32.82 MW-207 43.88 45.98 no data 34.3 no data 34.80 28.90 Italic date - measurements obtained after the October 2001 cessation of mat drain sump pumping.

Bold - Indicates lower hydraulic head at well clusters. Vertical gradient is toward the bold datum.

August 2004 2-36 Rev. 2 I

Haddam Neck Plant License Termination Plan Table 2-9 Temporal Trends in Groundwater Radionuclide Concentrations (Through December 2001) I Well Number Round March April September June July December 1 1999 I 1999 1999 I 2000 2001 2001 Cesium (in PCi/I)

MW-103S 76 33 29 72 35 <13 Tritium (in pCi/l)

MW-100D <700 <1000 NS <MDC <270 <210 MW-IOOS <700 <1000 NS NS <270 <200 MW-IOID <700 <1000 NS NS <260 <260 MW-101S <700 <1000 NS <MDC <260 <210 MW-102D 2,740 3,160 2,640 2,470 2,620 4,110 MW-102S <700 <1000 NS 5,540 7,250 20,600 MW-103D 22,180 17,550 19,660 20,900 20,800 8,100 MW-103S 2,580 9,260 2,980 1,230 1,120 5,350 MW-104S <700 <1000 NS NS <270 <210 MW-105D 4,590 2,450 3,030 2,150 1,360 2,110 MW-105S 138,700 67,400 23,480 15,900 12,200 1,800 MW-106D 3,320 1,590 5,830 1,810 1,450 14,200 MW-106S 24,290 <1000 NS NS 780 2,130 MW-107D <700 <1000 NS <MDC <270 <210 MW-107S <700 <1000 NS <MDC <270 <220 MW-108S <700 <1000 NS NS <270 <210 MW-109D 33,070 31,600 21,230 15,800 6,550 <210 MW-109S <700 <1000 NS <MDC <270 <240 MW-I IOD 27,630 23,280 27,230 18,300 18,700 21,300 MW-lbIS 3,090 <1000 2,470 2,360 1,890 <240 MW-IllS <700 <1000 NS <MDC <270 <210 MW-I 12S <700 <1000 NS NS <270 <240 MW-113S <700 <1000 NS NS <270 <240 MW-114S <700 <1000 NS <MDC <270 NS MW-115S <700 <1000 NS 5,550 4,500 NS MW-117S <700 <1000 NS NS <180 <240 AST-1 <700 <1000 NS NS <260 <210 Mat Sump 2,630 2,320 NS 2,890 NS NS EOF 2 <700 <1000 NS <MDC <270 <200 TW-I <700 <1000 NS <MDC <270 <250 MW-13 <700 <1000 NS <MDC <270 <240 MW-200 <700 <1000 NS <MDC <180 NS MW-201 <700 <1000 NS <MDC <180 NS MW-202 <700 <1000 NS <MDC <180 <210 MW-203 <700 <1000 NS <MDC <270 <250 MW-204 <700 <1000 NS <MDC <180 <210 August 2004 2-37 Rev. 2 I

Haddam Neck Plant License Termination Plan Table 2-9 Temporal Trends in Groundwater Radionuclide Concentrations (Through December 2001) I WVell Number I Round March April September June July December

_ 1999 I 1999 1999 2000 2001 2001 MW-205 <700 <1000 NS <MDC <180 <210 MW-206 <700 <1000 NS <MDC <180 <250 MW-207 <700 <1000 NS <MDC <180 <250 EOF Supply Well NS NS NS NS NS <210 MW-502 NS NS NS NS <67 NS MW-503 NS NS NS NS <67 NS MW-504 NS NS NS NS <67 NS MW-505 NS NS NS NS <67 NS MW-507S NS NS NS NS <67 NS MW-508S NS NS NS NS <67 NS MW-508D NS NS NS NS <67 NS Technicium-99 (in PCi/I)

MW-103D NS lNS [NS lNS 13.9 1<l0 Strontium-90 (in pCi/1)

MW-103S NS NS NS NS 2.55 1.82 MW-105S NS NS NS NS 129/120 69.7

__ _ __ _ _ _ __ __ __ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - 1 43 / 1 4 1 _ _ _ _ _ _ _ _

MW-106S NS NS NS NS 6.6 4.67 Mat Sump NS NS NS NS 5.91 NS NA - Not Analyzed NS - Not Sampled In prep. - In preparation MDC - Minimum Detectable Concentration 2.3.3.1.6.2 Ongoing GroundwaterMonitoringProgram Additional groundwater characterization and monitoring work will be implemented to support and supplement existing data, as described in the Phase 2 Hydrogeologic Investigation Work Plan (Reference 2-8). The program includes the followving elements:

1. Installation of additional groundwater monitoring wells (single and multi-level);

2. Conducting hydraulic testing of principal water bearing units;

3. Measuring groundwater levels, and any influence from on-site precipitation;

4. Evaluating interactions between groundwater and surface water, including possible tidal influence;

5. Sampling and analyzing groundwater for radionuclides (see Table 2-9); and

6. Developing groundwater flow and transport modeling.

August 2004 2-38 Rev. 2 I

Haddam Neck Plant License Termination Plan An initial round of analysis for boron, tritium, alpha and gamma spectroscopy, and hard to detect beta-emitting radionuclides will be performed at all locations. Based on these results, quarterly monitoring will continue at selected existing wells and all new well locations for those radionuclides detected during the initial round. The following provides a summary of the areas of additional or on-going groundwater monitoring:

Additional Monitoring Wells CYAPCO proposes drilling five borings, principally within bedrock, in the presumed pathways of the plumes along the Connecticut River. These boring locations were selected based upon their ability to provide spatial data and a maximum of bedrock information per hole, but still be close enough together to allow testing for hydraulic interconnectivity and good resolution on lateral radar imagery. Detailed information on the bedrock lithologies, fractures, hydraulics, and distributions of substances of concern in each borehole will be determined first. Then multiple sampling points in each location, either within the original borehole, or in separate, clustered boreholes, will be installed depending on the initial findings.

An offsite well cluster is also proposed to characterize groundwater flow gradients and quality across the river from the site. Four other conventional monitoring wells will also be installed onsite to characterize and delineate specific edges of or preferential migration pathways within the plumes.

Hydraulic and Geophysical Aquifer Testing CYAPCO will use rising or falling slug tests for overburden monitoring wells to characterize the hydraulic conductivity of the overburden aquifer materials. These data will be used to calculate groundwater flow and radionuclide migration rates in the overburden aquifer.

The hydraulic characteristics of the bedrock aquifer will be determined using packer tests and geophysical methods. The bedrock portion of the proposed five additional bedrock borings will be completed in twenty (20) foot increments (e.g., length of one drill rod). Following the completion of each 20-foot bedrock section, the borehole will be isolated with a packer assembly and groundwater will be collected and analyzed for tritium and gamma-emitting radionuclides (e.g., Cs-137 and Co-60). Selected samples will be analyzed for hard-to-detect radionuclides. During the sample purging process, water levels will be measured above the packer interval to monitor for potential packer leak(s) and failure of the packer. To later help define the hydraulic properties of each 20-foot zone tested, CYAPCO will measure and record the flow rate, water level data and the potential change in hydraulic head during purging. Following the completion of the packer test sampling, CYAPCO will perform geophysical testing at each borehole location. The methods that may be utilized to characterize the location, orientation, and the groundwater flow in fractures includes caliper, gamma, electromagnetic induction, borehole deviation, optical televiewer, acoustic televiewer, fluid resistivity, temperature, and electromagnetic flowmeter logs. These techniques will allow for a determination of potential fracture connectivity, plume migration pathways, and groundwater and radionuclide fluxes. Additionally, these tests will allow for an evaluation of monitoring well construction details that will provide for the collection of the most representative groundwater quality data. A complete overview of well installation and sampling procedures is contained within the Phase 2 Hydrogeologic Work Plan, Section 3.2.3.

Groundwater Monitoring Groundwater levels will be measured synoptically in monitoring wells in the Industrial Area on a quarterly basis for at least one year. An initial round of analysis for boron, tritium alpha and gamma spectroscopy, and hard to detect beta-emitting radionuclides will be performed at all locations. Based on these results, quarterly monitoring will continue at selected existing wells and all new well locations for those substances of concern detected by the initial round. These data will be used to develop August 2004 2-39 Rev. 2

Haddam Neck Plant License Termination Plan potentiometric surfaces for the overburden, shallow bedrock and deep bedrock units, and to more definitively establish hydraulic gradients and groundwater flow paths for each of these water-bearing units. The groundwater level data will also be used to assess the temporal variability in groundwater level and flow direction in response to seasonal changes in precipitation and recharge rates.

Sampling of the existing monitoring well locations, as well as new locations, or a selected subset, will be completed in accordance with the Phase 2 Hydrogeologic Investigation Work Plan. CYAPCO will initially analyze the monitoring wells (including all new wells, landfill, parking lot, well across Connecticut River) for the radionuclides of interest contained in Table 2-12.

Based on these results, quarterly monitoring will continue at selected existing wells and all new well locations for those substances of concern detected by the initial round. If any reactor-generated radionuclides are detected from the initial sampling round, then additional sampling and analyses for alpha and HTD radionuclides will be performed in subsequent sampling rounds.

It is anticipated that the data resulting from these sampling activities will be used to document concentration trends over space and time. With the installation of the additional monitoring wells, the groundwater-monitoring network should have sufficient density to allow characterization of the highest concentrations of the groundwater plume. Data from the additional monitoring wells should also allow for the bounding of the vertical and horizontal extent of the groundwater plume.

Groundwater-Surface Water Interaction Water levels in the Connecticut River and five selected overburden, shallow bedrock, and deep bedrock wells will be monitored simultaneously at 15-minute or shorter intervals for a period of three days during the period of maximum predicted tidal fluctuation. The resulting data will be used to evaluate the relationship between groundwater elevations in the Industrial Area and surface water elevation of the Connecticut River, including the degree of hydraulic communication between the groundwater and the surface water. These data will also be used to assess whether plant generated radionuclides discharge into the Connecticut River.

Flow and Transport Modeling Based on the results of the site characterization activities described above, a ground water flow and transport model will be developed to assess quantitatively groundwater flow velocities, groundwater flow paths, and advective radionuclide transport rates and residence times in overburden, shallow bedrock, and deep bedrock units. This model would represent the Containment Building foundation, the turbine foundations, and the discharge tunnels to assess the influence of these impermeable subsurface structures on groundwater flow. The depth to bedrock and the thickness of each of the modeled water bearing units will be developed from existing boring logs and the boring logs that will result from installing the additional wells. Results from the slug tests, andlor other aquifer evaluation methods, will be used to assign the hydraulic conductivity of each water bearing unit. The model will be calibrated using the observed groundwater elevations. Once the groundwater flow model is calibrated, a fate and transport model will be used to predict groundwater flow paths, radionuclide transport rates and concentrations, and how long radionuclides will persist in the water bearing units.

Dose Modeling The groundwater sampling and analysis will be used to characterize the radionuclide concentrations in groundwater. As described in Section 6 dose modeling, using the RESRAD computer code, has been August 2004 2-40 Rev. 2

Haddam Neck Plant License Termination Plan performed to produce DCGLs representing 25 mrem/yr. A method for ensuring compliance with the IOCFR20.1402, by accounting for the dose from all media (including groundwater), is provided in Section 5.4.7.1.

Reporting and Documentation Results of the characterization activities described above will be documented as described in the Phase 2 Hydrogeologic Investigation Work Plan. CYAPCO will prepare a monitoring report for each quarterly round of sampling and analysis within approximately 90 days from the completion of sampling for the quarter.

2.3.3.1.7 Surface Water Monitoring As indicated in Section 2.3.3.1.6 any surface water interaction with groundwater, including possible tidal influence, will be evaluated and incorporated into the conceptual groundwater model. Surface water releases were considered when developing the areas of concern in the Phase 2 Hydrogeologic Work Plan.

Surface water permits are in place to prevent either radiologically or non-radiologically pollution from uncontrolled discharges and releases. Permits include the National Pollutant Discharge Elimination Source (NPDES), Radiological Effluent Monitoring Program (REMP), and State of Connecticut Stormwater Restrictions. The permits impose the limits for flow, temperature, and effluent chemistry (radiological and nonradiological). To date, no surface water contamination attributable to the site has been observed.

2.3.3.2 Initial Area Classification The current classifications for the HNP site are presented in Table 2-10 for site grounds (surface and subsurface) and structures. Survey area definitions were initially established for decommissioning and were used and expanded upon during subsequent site characterization activities. Some area boundaries were redefined to make use of logical physical boundaries and site landmarks. Many areas are further subdivided into survey units. A survey unit is a physical area consisting of structures or land areas of specified size and shape for which a separate decision will be made as to whether or not residual contamination in that area exceeds the release criterion. Note that survey areas for subsurface soils include any sub-surface features that are present such as piping and drain systems. The survey areas corresponding to the current plan are depicted in the maps presented as Figures 2-1 through 2-9.

The current decommissioning approach is generally to remove the above-grade portions of site buildings and structures. The above-grade portions had been previously identified as survey areas and had been given MARSSIM classifications as a part of the original LTP. Their survey area designations have been subsequently removed as final status activities and are no longer planned.

Classification of a survey area has a minimum of two stages: (1) initial classification and (2) final classification. Initial classification is performed only once, at the time of identification of the survey unit using the information available. Final classification is performed and verified as an objective of the final status survey plan.

Although it is expected that the existing survey areas will require no modification with regard to boundaries or classification, the characterization process is iterative. When additional information is obtained during the decommissioning process through characterization surveys, remediation surveys (performed to track the effectiveness of decontamination techniques), or turnover surveys (discussed below), the data will be assessed using the DQO process to verify that the current classification of the August 2004 2-41 Rev. 2

Haddam Neck Plant License Termination Plan survey area is appropriate, to guide reclassification of the survey area, and/or to guide the design of subsequent surveys. Approved site procedures govern the process of classification and mandate appropriate documentation of the classification results.

Correct classification of a survey unit is crucial for appropriate final status survey design, and the potential for making decision errors increases when a survey unit is incorrectly classified. Thus, the initial assumption for classifying a survey unit is that the survey unit contains residual radioactivity levels greater than the applicable Derived Concentration Guideline Levels (DCGLs) and, thus is a Class I area.

Available information is subsequently used to support classification of an area or unit as Class 2, Class 3, or non-impacted. Approved site procedures guide this determination process and mandate appropriate documentation for the determinations made.

The definitions of Class 1,2, and 3 (per MARSSIM) are as follows:

• Class I Areas: Areas that have, or had prior to remediation, a potential for radioactive contamination (based on site operating history) or known contamination (based on previous radiation surveys) above the DCGLW. Examples of Class I areas include: 1) site areas previously subjected to remedial actions, 2) locations where leaks or spills are known to have occurred, 3) former burial or disposal sites, 4) waste storage sites, and 5) areas with contaminants in discrete solid pieces of material and high specific activity.

• Class 2 Areas: Areas that have, or had prior to remediation, a potential for radioactive contamination or known contamination, but are not expected to exceed the DCGLW .To justify changing the classification from Class I to Class 2, there should be measurement data that provides a high degree of confidence that no individual measurement would exceed the DCGLW. Other justifications for reclassifying an area as Class 2 may be appropriate based on site-specific considerations. Examples of areas that might be classified as Class 2 for the final status survey include: 1) locations where radioactive materials were present in an unsealed form, 2) potentially contaminated transport routes,

3) areas downwind from stack release points, 4) upper walls and ceilings of buildings or rooms subjected to airborne radioactivity, 5) areas handling low concentrations of radioactive materials, and

6) areas on the perimeter of former contamination control areas.

• Class 3 Areas: Any impacted areas that are not expected to contain any residual radioactivity, or are expected to contain levels of residual radioactivity at a small fraction of the DCGLw, based on site operating history and previous radiation surveys. Examples of areas that might be classified as Class 3 include buffer zones around Class I or Class 2 areas, and areas with very low potential for residual contamination but insufficient information to justify a non-impacted classification.

Class I areas have the greatest potential for contamination and therefore receive the highest degree of survey effort for the final status survey, followed by Class 2 areas, and then by Class 3 areas. Non-impacted areas do not require any level of survey coverage because they have no potential for residual contamination. As a survey progresses, reevaluation of classifications may be necessary based on newly acquired survey data. The final status survey plan includes a process by which measurements that approach investigation levels, defined as fractions of the DCGLs (and discussed further in Section 5.5.3.1), are reviewed to see if reclassification of an area(s) is necessary.

The classifications, provided in Table 2-10, were conservatively chosen based upon the review of a large volume of historical radiological survey data (routine and non-routine), collected over the plant's operating history; a review of the historical information maintained under 10 CFR 50.75(g); and a review August 2004 2-42 Rev. 2

Haddam Neck Plant License Termination Plan of the scoping survey information compiled for the decommissioning of the HNP site. The radiological survey information in particular represents a substantial volume of information.

A supplement to the original Historical Site Assessment (HSA) was developed to provide more detailed information on the radiological status of structures and land at the Haddam Neck Plant. This supplement provides the additional information that was used to establish the initial MARSSIM classifications of survey areas and survey units as shown in LTP Table 2-10.

The HSA Supplement (and LTP Tables 2-1 IA, -I lB, and -I IC) summarizes a comprehensive review of numerous plant records such as plant incident reports, condition reports, investigations, radiological surveys (1967 to 2000), annual effluent reports, interviews with past and present employees and site walk-downs.

In all cases, radiological survey data was obtained by trained technicians using calibrated instrumentation.

Since the data set spans a large period of time (33 yrs), the procedures used in the calibration and other quality control processes have likely changed. However, most of the data presented in Tables 2-1 IA,

-1 IB, and -I IC is a result of surveys conducted within the last 5 years where the calibration and quality control processes have been consistent and similar to those used today. Within the last 5 years, all instruments have been calibrated against NIST traceable standards and undergo daily operability checks and in the case of laboratory sample analysis participate in third-party quality control performance testing.

In all cases however, radiological surveys and laboratory sample results have been generated by trained technicians and reviewed and approved by supervision. These processes ensure the validity of the collected data. In general, the tables are summaries of numerous routine and decommissioning support surveys. As a result, some radiation data in the tables (e.g., <0.2 or <1 mR/hr) is limited due to the instrumentation used for the surveys. This is also the reason for the large standard deviations in contamination levels. Between 1997 and 1999 limited site characterization surveys were performed.

Results of samples from these surveys are listed in the Comments column of Table 2-I I A and 2-I IC and in the Concentration columns of Table 2-1 lB.

The survey of many inaccessible or not readily accessible areas or surfaces has been deferred. Examples of areas where surveys are deferred include soils under structures, contaminated sumps, pipe trenches, and the Containment dome. The decision to defer the survey of an area was based on one or more of the following conditions:

• ALARA considerations (e.g., the area is either a high radiation or high contamination area and additional data would likely not change the survey area or unit MARSSIM classification of the location or surrounding areas),

• safety considerations (e.g., difficulty of access to the upper reaches of the Containment dome due to height above the charging floor),

• historical data shows that the area could be classified without further characterization,

• access for characterization would require significant deconstruction of adjacent systems, structures or other obstacles the removal of which could result in an unsafe condition or interfere with continued operation of required components, or

• the ability to use engineering judgement in assigning the area a MARSSIM classification based on physical relationship to surrounding areas and the likelihood of the area to have radiological conditions represented by the conditions in these adjacent areas.

August 2004 2-43 Rev. 2 1

Haddam Neck Plant License Termination Plan As access is gained to areas that were previously inaccessible, additional characterization data will be collected, evaluated and stored with other radiological survey data in a survey history file for the survey unit. Sampling for this additional characterization data will be chosen to include several locations such as cracks, floors and walls to establish the variability and extent of the contamination. This data will be used to establish the radionuclides present and variability in the radionuclide mix for both easy-to detect (ETD) and hard-to-detect (HTD) radionuclides. In addition, as the decommissioning progresses, data from operational events caused by equipment failures or personnel errors, which may affect the radiological status of survey unit(s), will be captured by CYAPCO's Condition Reporting process. These events will be evaluated and, when appropriate, stored in the 50.75(g) database. This additional characterization data will be used in validating the initial classification and in planning for the final status survey of each survey unit.

As decommissioning proceeds, areas will, as necessary, be decontaminated to remove loose surface contamination (as well as fixed contamination) to levels that will meet the conditions for controlled demolition or unrestricted release conditions for demolition. When an SSC is ready for demolition, a documented formal turnover from CYAPCO to the demolition contractor is made for access and control of the area. Following the demolition and when an area is believed to be ready for final status survey, a "turnover assessment" will be performed. If the results of this assessment indicate that the Final Status Survey acceptance criteria will be met, then physical and administrative control of the area will be transferred to the Final Status Survey personnel for preparation, design, and performance of the FSS.

Othenvise, additional remediation may be required. This assessment may include a "turnover survey,"

primarily for Class I and 2 areas within the Industrial Area and in land areas outside of the industrial area, that are impacted by existing groundwater contamination. This "turnover survey" process will include a MARSSIM-type survey in which a combination of scanning and fixed measurements will be obtained and evaluated against the FSS criteria for the survey unit. If the results of this survey indicate that the FSS acceptance criteria will be met, then physical and administrative control of the area will be transferred to the Final Status Survey personnel for preparation, design, and performance of the FSS. Othenvise, additional remediation may be required. The "turnover survey" process together with any additional characterization and remediation survey performed, represent at least one, but possibly several, opportunities to collect additional survey data prior to conducting the FSS. For each survey type (characterization, remediation, turnover, and final status) a documented survey plan will be developed using the DQO process. The level of effort with which the DQO process is used as a planning tool is commensurate with the type of survey and the necessity of avoiding a decision error. This is the graded approach of defining data quality requirements. For example, scoping and characterization survey plans intended to collect data might only require a survey objective and the instrumentation and analyses specifications necessary to meet that survey objective. Remediation and final status plans which require decisions would need additional effort during the planning phase according to the level of risk of making a decision error and the potential consequences of making that error. These survey plans will contain the appropriate data assessment to ensure that several objectives are met. These objectives include:

• Appropriate instrument selection to ensure the proper sensitivity relative to the applicable DCGLs,

• Appropriate instrument quality control measures to ensure operability,

• Appropriate survey techniques, as described in NUREG-1507, to ensure that the field measurement techniques are consistent with the calibration methodologies,

• Appropriate sample collection and analysis to determine spatial variability and variability in radionuclide ratios,

• Data analysis criteria to identify follow-up actions such as remediation and the collection of additional samples, and

• Appropriate classification of survey area.

August 2004 2-44 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-10 MARSSIM Classifications (Updated as of May 2004)

Area (in 2 )

Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Figure Coe Code Description Code Code Description MARSS.M Area Area Area: No.

______________________________________Area) 1000 Fuel Building Soils/Foundation 1 N/A 630 N/A 2-2 1308 Fuel Building Spent Fuel Pool Pit 0001 Floor Area, Pool Sides and 1 75 125 1.7 2-5 Bottom - Section 1 0002 Floor Area, Pool Sides and 1 75 125 1.7 2-5 Bottom - Section 2 2000 Primary Auxiliary Building 0001 East Soil/Foundation I N/A 1800 N/A 2-2 0002 West Soil/Foundation I N/A 1690 N/A 2.2 3000 Reactor Containment Soils/Foundations I N/A 1600 N/A 2-2 3002 Reactor Containment Enclosure 0000 Floor Area and Walls 1 25 115 4.6 2-6 Under Reactor Vessel 3004 Reactor Containment Enclosure 0000 Floor Area and Walls including 1 20 100 5.0 2-6 Sump Area Under Reactor Vessel tube to in-core sump area August 2004 2-45 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-10 MARSSIM Classifications (Updated as of May 2004)

Area (m2 )

Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Figure Area Code Description lUnit Code DescrveyUniption llMARSSIM Area Area A CdCoeDsrpinCode Code Decito Class. Floor No 3101 Reactor Containment Enclosure #4 0001 Floor Area, Walls and Ceiling - 1 65 300 4.6 2-7 Outer Annulus Lower Level NE Section 1 0002 Floor Area, Walls and Ceiling - 1 65 300 4.6 2-7 Section 2 3102 ReactorContainment Enclosure #1 0001 Floor Area, Walls and Ceiling - 1 65 300 4.6 2-7 Outer Annulus Lower Level NW Section I 0002 Floor Area, Walls and Ceiling - 1 65 300 4.6 2-7 Section 2 3103 Reactor Containment Enclosure #2 0001 Floor Area, Walls and Ceiling - 1 65 300 4.6 2-7 Outer Annulus Lower Level SW Section 1 0002 Floor Area, Walls and Ceiling - 1 65 300 4.6 2-7 Section 2 3104 Reactor Containment Enclosure #3 0001 Floor Area, Walls and Ceiling - 1 65 300 4.6 2-7 Outer Annulus Lower Level SE Section 1 0002 Floor Area, Walls and Ceiling - 1 65 300 4.6 2-7 Section 2 Rev. 2 2-46 2004 August 2004 2-46 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-10 MARSSIM Classifications (Updated as of May 2004)

Area (m2l Ratio Survey Survey Area Survey Suve .Ui Initial Floor Total (Total Figure Area Area Unit Codeurrey Description l A Ar l Area: N Area) 3105 Reactor Containment Enclosure 0000 Floor Area, Walls and Ceiling 1 40 185 4.6 2-7 Reactor Containment Sump Area 3107 Reactor Containment Enclosure 0001 Floor Area, Walls and Ceiling - 1 70 290 4.1 2-7 Cable Vault Outside Reactor Section I Containment .

0002 Floor Area, Walls and Ceiling - 1 70 290 4.1 2-7 Section 2 3111 Reactor Containment Enclosure 0001 Floor Area, Walls and Ceiling - 1 60 225 3.8 2-7 Loop # I Inner Annulus Lower Section I Level NE 0002 Floor Area, Walls and Ceiling - 1 55 220 4.0 2-7 Section 2 3112 Reactor Containment Enclosure 0001 Floor Area, Walls and Ceiling - 1 75 255 3.4 2-7 Loop #2 Inner Annulus Lower Section I Level NW . _

0002 Floor Area, Walls and Ceiling - 1 75 255 3.4 2-7 Section 2 3113 Reactor Containment Enclosure 0001 Floor Area, Walls and Ceiling - 1 75 255 3.4 2-7 Loop #3 Inner Annulus Lower Section I Level SW .

Rev. 2 2-47 2004 August 2004 2-47 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-10 MARSSIM Classifications (Updated as of May 2004)

I Area(m2)

Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Figure 3114A CpCodeR eactor Codien D Escriptiosure Unit Flo o Code de D D rMARSSIM Class. Area Area AFloor: N No.

______ _____Area) 0002 Floor Area, Walls and Ceiling - 1 75 255 3.4 2-7 Section 2 3114 Reactor Containment Enclosure 0001 Floor Area, Walls and Ceiling - 1 75 285 3.8 2-7 Loop 114 Inner Annulus Lower Section 1

____Level SE 0002 Floor Area, Walls and Ceiling - 1 75 270 3.6 2-7 Section 2 4000 Turbine Building Soils/Foundations 2 N/A 2650 N/A 2-2 5000 Service Building 0001 North Soils/Foundations I N/A 1230 N/A 2-2 0002 South Soils/Foundations 1 N/A 1230 N/A 2-2 5502 Circulating Water System 0001 Unit I Discharge Tunnel 2 290 950 3.3 2-3 Discharge Structures 0002 Unit 2 Discharge Tunnel 2 470 2,110 4.5 2-3 August 2004 2-48 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-10 MARSSIM Classifications (Updated as of May 2004)

Are ()

Ratio SuvySre raSurvey Stirvey Unit Iial Floor Ttl (Total Fgr Corde Code Description Unit Code Description MARSSIM Area Area Floor No.

6000 Waste Disposal Building Soils/Foundations I 2-2 7002 Screenhouse Building CW Circ 0000 Floor Area, Walls and Ceiling 3 180 600 3.3 2-8 Pump A&B Head 9102 YD 115KV Switchyard Area 0001 Trench and Adjoining Land I N/A 255 N/A 2-1 Area 0002 Land Area 2 N/A 1,760 N/A 2-1 0003 Demin Water Tank Land Area I N/A 280 N/A 2-1 9106 Discharge Canal 0001 Bank Land Area and Canal 2 N/A 3,065 N/A 2-4 Sediment 0002 Bank Land Area and Canal 2 N/A 6,583 N/A 2-4 Sediment 0003 Bank Land Area and Canal 2 N/A 8,294 N/A 2-4 Sediment August 2004 2-49 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-10 MARSSIM Classifications (Updated as of May 2004)

Artam Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Figure Area Cod Description Unit Code Dscrition MARSSIM Ar l Area Area: No.

CdCoeDsrpinCode CodeDecito Class. Ara re Floor No 0004 Bank Land Area and Canal 2 N/A 9,890 N/A 24 Sediment 0005 Bank Land Area and Canal 2 N/A 9,620 N/A 2-4 Sediment 0006 Bank Land Area and Canal 2 N/A 9,703 N/A 2-4 Sediment 0007 Bank Land Area and Canal 2 N/A 9,849 N/A 2-4 Sediment 0008 Bank Land Area and Canal 2 N/A 9,750 N/A 2-4 Sediment 0009 Bank Land Area and Canal 2 N/A 9,919 N/A 2-4 Sediment 0010 Bank Land Area and Canal 2 N/A 9,514 N/A 2-4 Sediment 0011 Bank Land Area and Canal 2 N/A 6,386 N/A 2-4 Sediment August 2004 2-50 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-10 MARSSIM Classifications (Updated as of May 2004)

I.

Arca (m')

Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Figure Area CoeDsrpinUnit CoeDsrpinMARSSIM Ae Ara Area: No CCode Code Description Class. A FlooraN

______________ ________________________________Area) 9202 Switchgear Building "B" 0000 Lower Switchgear Structure 2 N/A 430 N/A 2-2 Including Floor, Walls and Soils 9226 Radwaste Reduction Facility 0001 Soils/Foundations 1 N/A 500 N/A 2-2 9302 Northwest Protected Area Grounds 0000 Land Area 3 N/A 6,410 N/A 2-1 9304 Southwest Protected Area Grounds 0000 Land Area 3 N/A 5,800 N/A 2-1 9306 South Central Protected Area 0000 Land Area 2 N/A 4,780 N/A 2-1 Grounds 9308 Southeast Protected Area Grounds 0001 Land Area I N/A 1,570 N/A 2-1 0002 Land Area I N/A 1,570 N/A 2-1 0003 Land Area I N/A 1,570 N/A 2-1 Rev. 2 2-51 August 2004 2-51 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-10 MARSSIM Classifications (Updated as of May 2004)

. Area (m)

Ratio Arce Survey Area Survey Survey Unit Initial Floor Total (otal Figure Code Code Description Code Code Description Class. Area Area Floor No.

. CArea)odeCass.Flo 9310 East Protected Area Grounds 0001 Land Area From the Fuel I N/A 1,750 N/A 2-1 Building to the RadWaste

.__._. _Reduction Facility 0002 Land Area From the RadWaste I N/A 1,560 N/A 2-1 Reduction Facility to the East RCA Boundary 9312 Northeast Protected Area Grounds 0001 Land Area From the North RCA I N/A 1,810 N/A 2-1 Gate to Security Fence 0002 Land Area From Security Fence I N/A 1,910 N/A 2-1 to Fuel Building 0003 PAB/Service Bldg. Alley Way I N/A 1,945 N/A 2-1 9313 Central Site Grounds 0000 Land Area 2 N/A 5,850 N/A 2-1 9402 Emergency Operations Facility 0000 Soils/Foundations 3 N/A 1,300 N/A 2-2 9502 Northeast Site Grounds (Non- 0000 Land Area 3 N/A 8,700 N/A 2-1 Protected Area)

August 2004 2-52 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-10 MARSSIM Classifications (Updated as of May 2004)

Area (m2)

Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Figure SreSuvyArea SurveyescSurvey Unit MARSSIM Ara re Area:

CCode ode Description Code Code Description Class. Area Floor No.

___A ____ rea) 9504 Bypass Road / Secondary Parking 0000 Land Area 3 N/A 2,800 N/A 2-1 Lot 9506 North Site Grounds (Non-Protected 0000 Land Area 3 N/A 3,800 N/A 2-1 Area) 9508 Pond 0000 Land Area and Pond Sediment 3 N/A 10,000 N/A 2-1 9510 Access Road 0000 Paved Road 3 N/A 2,300 N/A 2-1 9512 Northwest site Grounds (Non- 0000 Land Area 3 N/A 19,500 N/A 2-1 Protected Area) 9514 Primary Parking Lot 0000 Land Area 3 N/A 20,000 N/A 2-1 9518 Southwest Site Grounds (Non- 0000 Land Area 2 N/A 5,900 N/A 2-4 Protected Area) 9520 Southwest Site Storage Area 0001 Land Area From Security Fence 2 N/A 7,500 N/A 2-4 to Load Distribution Tower August 2004 2-53 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-10 MARSSIM Classifications (Updated as of May 2004) l Ara (m2)

Ratio SuvySre raSurvey Survey Ui Initial Floor Total (Total Fgr Area Sode Description Unt Code Description MARSSIM Area Area Area: Figuoe C eCoeDsrpinCode CodeDecito Class. Floor No

_______________________________________Area) 0002 Land Area From Load 2 N/A 7,000 N/A 2-4 Distribution Tower East to 150m 0003 Land Area 150m East of Load 2 N/A 7,000 N/A 2-4 Distribution Tower to Gate 3 9521 Southeast Pond 0000 Land Area and Pond Sediment 3 N/A 23,100 N/A 2-4 9522 Southeast Site Grounds (Non- 0001 Land Area 2 N/A 8,700 N/A 2-1 Protected Area) 0002 Land Area I N/A 1,800 N/A 2-1 0003 Land Area I N/A 1,900 N/A 2-1 0004 Land Area I N/A 1,200 N/A 2-1 9523 Southeast Wetland Area 0000 Land Area 3 N/A 106,000 N/A 2-4 August 2004 2-54 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-10 MARSSIM Classifications (Updated as of May 2004)

Area (m')

Ratio Survey Survey Area Suni Su Unit Initial Floor Total (Total Figure Area Coute Descript GUroni Sodre Ucnit MARSSIM Area Area Area: No.

Area) 9524 South Site Grounds (Non-Protected 0000 Land Area 3 N/A 110,000 N/A 2-4 Area) 9525 Southeast Site Road 0000 Land Area 3 N/A 28,000 N/A 2-4 9526 Northeast Mountain Side 0000 Land Area 3 N/A 444,700 N/A 2-4 0001 Land Area 2 N/A 6,536 N/A 2-4 0002 Land Area 2 N/A 6,069 N/A 2-4 9527 East Mountain Side 0001 Land Area 2 N/A 8,600 N/A 2-1 0002 Land Area 2 N/A 9,740 N/A 2-1 0003 Land Area 2 N/A 8,200 N/A 2-1 Rev. 2 2-55 August 2004 2004 2-55 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-10 MARSSIM Classifications (Updated as of May 2004)

Area(m=) _

Ratio Srveay Survey Area Survey Survey Unit Initial Floor Total (Total Figure Cre Code Description Code Code Description Class. Area Area Floor No.

_______ A re a _ _ _ _ _ _

0004 Land Area 2 N/A 5,400 N/A 2-1 0005 Land Area 2 N/A 4,000 N/A 2-1 9528 Southeast Mountain Side 0000 Land Area 3 N/A 575,488 N/A 2-4 0002 Land Area 2 N/A 9,752 N/A 2-4 0003 Land Area 2 N/A 9,447 N/A 2-4 0004 Land Area 2 N/A 3,058 N/A 9530 Central Peninsula Area 0001 Land Area Bounded by and 2 N/A 8,000 N/A 2-4 Immediately Adjacent to the

_____________________________ _____Road ___

0002 Western Half of Diked Area and 2 N/A 5,000 N/A 2-4 Immediate Surrounding Sides August 2004 2-56 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-10 MARSSIM Classifications (Updated as of May 2004)

. Area (__2)

Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Figure Code Code Description' Cnde Code Description MARSSIM Area Area Area: No.

________________________ _____Area) 0003 Eastern Half of Diked Area and 2 N/A 5,000 N/A 2-4 Immediate Surrounding Sides 0004 Open Land Areas 3 N/A 97,000 N/A 2-4 0005 Land Area South of Diked Areas 2 N/A 7,000 N/A 2-4 9531 South End of Peninsula 0000 Land Area 3 N/A 118,000 N/A 2-4 9532 East Site Grounds (Non-Protected N/A N/A Non-impacted N/A 375,600 N/A 2-4 Area) 9535 South East Landfill Area 0001 Land Area I N/A 1,900 N/A 2-4 0002 Land Area 2 N/A 3,600 N/A 2-4 9536 Construction Piles Near Rifle Range 0000 Land Area 2 N/A 2,200 N/A 2-4 Rv.

257 Augut 204 August 2004 2-57 Rev. 2

Haddam Ncck Plant License Termination Plan Table 2-10 MARSSIM Classifications (Updated as of May 2004)

Area (m)

Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Figure Crdae Code Description Unit Code Description MARSSIM Ara Area Aloeor No.

._ Area) 9537 Permitted Landfill Area 0000 Land Area 2 N/A 2,200 N/A 2-4 9538 Material Storage Area 0000 Land Area 2 N/A 4,200 N/A 2-4 9801 Subsurface soils in Radiologically 0000 Subsurface Soil A* N/A 15,500 N/A 2-9 Controlled Area (excluding 9308) 9802 Subsurface soils associated with 0000 Subsurface Soil B* N/A 27,065 N/A 2-9 surface soil survey area 9308 and subsurface soil areas within the industrial area but outside the radiologically controlled area 9803 Subsurface soils associated with 0000 Subsurface Soil C* N/A 35,150 N/A 2-9 parking lot, Warehouses I & 2, and Steam Generator Mockup Building 9804 Subsurface soils associated with 0000 Subsurface Soil C* N/A 11,950 N/A 2-9 South East Site Grounds 9805 Subsurface soils associated with 0000 Subsurface Soil C* N/A 137,000 N/A 2-9 Peninsula (excluding area 9531)

August 2004 2-58 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-10 MARSSIM Classifications (Updated as of May 2004) r -

Area (m=) Ratio SuvySre raSurvey Survey Unit Initial For Total Aretal Figure Code Code Description Unit Code Description lMASSIM Area Area Floor No.

_ _ _ _ _ _ _ _ _ _ _ _ _ _ __ __ _ _ _ _A r e a) 9806 Subsurface soils associated with 0000 Subsurface Soil A* N/A 4250 N/A 2-9 surface soil portions of survey area 9535 (South East Landfill)

• MARSSIM does not cover media such as subsurface soil, which is considered beyond its scope. LTP Section 5.7.3.2.1 discusses the criteria applied during the classification of subsurface soils.

August 2004 2-59 Rev. 2 l

Haddam Neck Plant License Termination Plan Table 2-11A Nominal Radiological Data Supporting Classifications for Structures Survey Survey Area Area Code Description Radiation Levels Contamination Levels Comment Code' Cd ecito (m Jhr) Beta-gamma (dpm/lOOcm2 ) Alpha (dpmIlOOcm2 )

Min Max Min Max Average Stdev. 1 M11n Max Average Stdev 1102* FuelBuildingLaydown 100 ND >100000 56500 53627 ND 350 180 115 Area (< 57.2) (<14.1) 1104* Fuel Building Fuel Cask <I 15 ND 78000 12122 24956 57 324 N/A 2 N/A Decon Area (< 57.2) 1106* Fuel Building Skimmer 100 ND >100000 57333 80258 ND 198 N/A' N/A Pump and Sump Area (< 57.2) (<14.1) 2 1202* Fuel Building New Fuel <1 10 ND >100000 51540 79943 ND N/A'N/A Storage Area (< 57.2) (<14.1) 1204* Fuel Building Exhaust <1 2 ND 2000 N/A2 ND N/A 2 N/A 2 Filters and Fan (<57.2) (<14.1) 2 1302* Fuel Building Patio Area <0.2 ND N/ N/A 2 ND N/A' N/A'

(<42.0) (<8.92) 1304* Fuel Building New Fuel <1 15 ND 27000 7500 13000 ND N/A2 N/A2 Storage Area (< 57.2) (<14.1)

Denotes historical information for previously identified survey unit. This survey unit has subsequently been eliminated or incorporated into another survey area. However, CYAPCO retains the option to use the classifications in Appendix H.

August 2004 2-60 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-11A Nominal Radiological Data Supporting Classifications for Structures roea Code Desuription Radiation Levels Contamination Levels Comment Code' Cd ecito (mRlhr) Beta-gamma (dpm/l00cm2) Alpha (dpm/lOOcm')

Mn Max Mn Max Average Stdev. Min Max Average Stdev 2 2 13061 Fuel Building Cask 100000 NA N/A ND N/A N/A2 Laydown Area (<57.2) (<14.1) 1308 Fuel Building Spent Fuel 100000 38200 59137 ND >500 N/A' N/A 2 Pool Pit (<57.2) (<14.1) 1404* Fuel Building Roof Area <0.0 0.2 ND 71 N/A 2 N/A 2 ND N/A 2 N/A 2

(<58.1) (<14.4) 2002* PrimaryAuxiliaryBuilding 12 80 8000 >100000 64000 54129 ND 480 N/A' N/A 2 Core Co-60 6.93 RIIR Pump Room A (<20) Bore Cs-137 5.38 (pCilg) 2004* Primary Auxiliary Building 12 40 4500 >100000 111200 121356 ND >500 N/A2 N/A 2 RiiR Pump Room B (<20) 2006 Primary Auxiliary Building 20 >100 4000 100000 N/A 2 N/A 2 ND >500 542 439 RHR Heat Exchangers (<20) 2008* Primary Auxiliary Building I0 40 16000 >100000 233333 102144 ND >500 1185 1208 Core Co-60 13.1 Primary Drain Tank Pump (<20) Bore Cs-134 0.63 Room (pCi/g) Cs-137 5.38 2010* Primary Auxiliary Building 100 >100 10000 > 100000 287500 119269 ND >500 N/A' N/A 2 Primary Drain Tank Room (<20)

Rev. 2 2-61 August 2004 August 2004 2-61 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-11A Nominal Radiological Data Supporting Classifications for Structures Su rvey Survey Area Area Cd D iti Radiation Levels Contamination Levels Comment Code' Cd ecito (mR/hr) Beta-gamma (dpm/100cmI) Alpha (dpm/100cmI)

Min Max Min Max Average Stdev. Min Max Average Stdev 2012* Primary Auxiliary Building 50 >100 10000 >100000 N/A2 N/A >500 N/A N/A Aerated Drain Tank Room 2104' Primary Auxiliary Building 100000 202225 154296 18 >500 101455 197722 Core Co-60 34.1 Pipe Chase Under Hallway (<1000) Bore Cs-134 5.18 (pCi/g) Cs-137 74.0 N/A 2 2

2106* Primary Auxiliary Building I >100 4000 >100000 N/A 6 >500 N/A Pipe Chase Under Valve Room 2108' Primary Auxiliary Building 4 30 ND >100000 N/A' N/A2 20 >500 1638 1605 Boric Acid Evaporator Area (< 57.2 )

TK EVI-IA, EV2-IA 2110* Primary Auxiliary Building <2 >100 ND >100000 173333 92916 ND >500 N/A2 N/A=

Pipe Chase East & West (< 57.2 ) (<20)

Outside 2202' Primary Auxiliary Building 0.5 4 ND 1000 <1000 Note' ND N/A2 N/A' Hallway (< 57.2 ) (<20) 22040 Primary Auxiliary Building 0.1 2 ND 1000 N/A ND N/A= N/A2 Component Cooling Area (< 57.2 ) (<20) 2206* Primary Auxiliary Building l 80 ND >100000 NiA 2 N/A 22 >500 N/A 2 N/A2 Boric Acid Evaporator Area (< 57.2 )

Rev. 2 2-62 August 2004 August 2004 2-62 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-11A Nominal Radiological Data Supporting Classifications for Structures Survey SuvyAe Area Sure Area Radiation Levels Contamination Levels Comment Code' oeDscito (mR/hr) Beta-gamma (dpm/100cm 2) Alpha (dpm/1OOcm2)

M=n Max MIln Max Average Stdev. Min Max Average Stdev 2208* Primary Auxiliary Building 0.5 600 ND >100000 81250 33260 22 >500 1588 1825 Boric Acid Mix Tank Area (< 57.2) 22101 Primary Auxiliary Building 1 >100 ND >100000 81000 53560 ND >500 N/A 2 N/A 2 "B" Charging Pump Area (< 57.2 (<20) 2212* Primary Auxiliary Building 100 ND 22000 9000 9201 ND >500 N/A' N/A 2

'A" Charging Pump Area (< 57.2) (<20) 2214* Primary Auxiliary Building 5 >100 ND >100000 91280 59156 383 N/A 2 N/A 2 Metering Pump Area (< 57.2) 2216* Primary Auxiliary Building 1I I0 ND 50000 13938 21273 ND N/A' N/A' Purification Pump Area (< 57.2) (<20) 2218* Primary Auxiliary Building 0.5 50 ND 8000 3400 3362 ND N/A2 N/A 2 Primary Water Transfer (< 57.2) (<20)

Pump Area 2220( Primary Auxiliary Building I 40 143 >100000 2 N/A2 ND 240 N/A 2 N/A' Sample Room (<8.89) 2222* Primary Auxiliary Building 0.5 15 ND 24000 - 1 N/A 2 ND N/A 2 N/A' Steam Generator Blowdown (< 57.2) (<20)

Room August 2004 2-63 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-11A Nominal Radiological Data Supporting Classifications for Structures Survey SuvyAe Area Code Description Radiation Levels Contamination Levels Comment Code Cd ecito (mRlhr) Beta-gamma (dpmIlOOcm 2) Alpha (dpm/lOOcm2)

_Lin Mlax Mlin l Aax Average Stdev. Mmn Mlax Average Stdev 2

2224* Primary Auxiliary Building I 12 ND 12000 10200 9406 ND 123 N/A 7NA IIPSI Cubicle Area (< 57.2 ) (<20) 2226' PrimaryAuxiliaryBuilding 2 >100 ND 8000 3112 2886 ND N/A' N/A' LPSI Cubicle Area (< 57.2 ) (<20) 2228* PrimaryAuxiliary Building 0.5 >100 ND >100000 67733 97325 ND >500 733 1097 Soil Am-241 0.29 Drumming Room (< 57.2 ) (<20) (pCi/g) Co-60 8.73 Cs-134 0.16 Cs-137 126.6 2302* Primary Auxiliary Building <0.1 0.3 ND 1000 N/A2 N/A. ND N/A 2 N/A 2 Component Cooling Area (< 57.2 ) (<20) 23040 Primary Auxiliary Building 0.2 24 ND 40000 16000 21000 ND >500 N/A2 N/A2 Boric Acid Evaporator Area (< 57.2 ) (<20) 2306* Primary Auxiliary Building I 35 ND >100000 N/A 2 N/A 2 ND 420 A 2 N/A Boric Acid Mix Tank Area (< 57.2 ) (<20) 2308^ Primary Auxiliary Building 2 >100 ND >100001 45500 70995 ND 300 N/A 2 N/A Volume Control Tank (< 57.2 ) (<20)

Room 2310* Primary Auxiliary Building I 15 <1000 <1000 Note Purge and Dilution Fans (<20)

Rcv. 2 2-64 August 2004 2004 2-64 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-11A Nominal Radiological Data Supporting Classifications for Structures SreCy ue Area Radiation Levels Contamination Levels Comment Code Cd ecito

= (mR/hr) Beta-gamma (dpm/I00cm2) Alpha (dpnI10Ocmz)

Min Maxy MMax in Average Stdev. Min Max Average Stdev 2312* Primary Auxiliary Building I 6 <1000 <1000 NDe N/A2 NA 2 Service Water Strainer Area (<20) 2314* PrimaryAuxiliary Building I >100 <62 5000 <1000 NotND N/A N/A2 HEPA Filter and Hall Area (<15.2) 2316* Primary Auxiliary Building 0.2 0.5 <1000 <1000 ND N/A 2 N/A' Boric Acid Storage Room (<20) 2402* Primary Auxiliary Building <0.2 40 ND 3000 1500 1414 ND N/A N/A Roof Area (< 48.9) (< 16.73) 3002 Reactor Containment 10 >100 5000 >100000 N/A N/A ND 400 N/A N/A Enclosure Under Reactor (<14.82)

Vessel 3004 Reactor Containment 10 >100 1000 >100000 N/A 2 N/A 2 ND >500 N/A 2 N/A 2 Enclosure Sump Area (<14.82)

Under Reactor Vessel 3101 Reactor Containment <1 80 ND 40000 9240 11746 ND ND <100 Note Enclosure #4 Outer Annulus (<1000) (<20) (<20)

Lower Level NE 3102 Reactor Containment 100 ND 9000 4567 4349 ND 295 165 113 Enclosure #1 Outer Annulus (<I000) (<14.1)

Lower Level NW August 2004 2-65 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-11A Nominal Radiological Data Supporting Classifications for Structures Area Code Description Radiation Levels Contamination Levels Comment Code Cd ecito (mR/hr) Beta-gamma (dpmt/OOcm2) Alpha (dpm/lOOcm)

Min Max Min Max Average Stdev. Min Max Average Stdev 3103 Reactor Containment <1 175 ND >100000 21078 38702 ND 400 200 173 Enclosure #2Outer Annulus (<I000) (<14.1)

Lower Level SW_

3104 Reactor Containment <1 >100 ND 35000 8890 10182 ND 124 108 14 Core Co-60 23.4 Enclosure #3 Outer Annulus (<I000) (<14.1) Bore Cs-1 34 2.76 Lower Level SE (pCi/g) Cs-137 279 3105 Reactor Containment 30 >100 10000 >I00000 N/A' N/A' ND N/A' N/A2 Enclosure Reactor (<14.1)

Containment Sump Area 3107 Reactor Containment <0. i ND 5000 2333 2309 ND N/A2 N/Az Enclosure Cable Vault (<1000) (<14.1)

Outside Reactor Containment 3111 Reactor Containment 45 >100 5000 >100000 195111 141934 ND >500 3568 6587 Enclosure Loop #1 Inner (<20)

Annulus Lower Level NE 3112 Reactor Containment 10 >100 12000 >100000 204500 167706 ND >500 676 983 Enclosure Loop #2 Inner (<20)

Annulus Lower Level NW 3113 Reactor Containment 10 >100 2000 > I00000 86125 64570 ND >500 359 205 Enclosure Loop #3Inner (<20)

Annulus Lower Level SW 3114 Reactor Containment I5 >100 880 > 100000 84625 78053 ND >500 194 154 Enclosure Loop #4Inner (<14.1)

Annulus Lower Level SE Rev. 2 2-66 August 2004 2004 2-66 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-11A Nominal Radiological Data Supporting Classifications for Structures Survey Survey Area Area Radiation Levels Contamination Levels Comment Code Cd ecito (mR/hr) Beta-gamma (dpm/IOOcm 2) Alpha (dpm/I00cm2)

M=n Max Min Max Average Stdev. Alin Max Average Stdev 3201d Reactor Containment I 15 ND 30000 15225 9208 ND 250 150 87 Enclosure #1 Outer Annulus (I 000) (<14.1)

Ground Level NE 3202 Reactor Containment I >100 ND 25000 12713 8316 ND 200 133 58 Enclosure #2 Outer Annulus (I1000) (<14.1)

Ground Level NW 32030 Reactor Containment I 10 ND 49000 24650 18033 ND 295 149 98 Enclosure #3 Outer Annulus (<1 000) (< 14.1)

Ground Level SW 3204* Reactor Containment 1 >100 ND >100000 38100 30329 ND 250 138 75 Enclosure #4 Outer Annulus (<1 000) (<14.1)

Ground Level SE 3205* Reactor Containment 100000 66760 81252 ND >500 N/A2 N/A2 Enclosure Reactor (I 000) (<14.1)

Containment Foyer Area Ground Level 3206* Reactor Containment <I 15 ND 96000 N/A 2 N/A ND 140 N/A 2 N/A Enclosure Reactor (< 1000) (<14.1)

Containment Hatch Area Ground Level 3211

• Reactor Containment 4 30 25000 >100000 N/A 2 N/A2 8 >500 N/A 2 N/A 2 Enclosure Loop #1 Inner Annulus Mid Ground NE 3212* Reactor Containment 5 42 ND 33000 N N/A ND 102 N/A N/A2 Enclosure Loop #2 Inner (<1000) (<14.1)

Annulus Mid Ground NW Rev. 2 2-67 2004 August 2004 2-67 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-11A Nominal Radiological Data Supporting Classifications for Structures Survey Survey Area Area Code Description Radiation Levels Contamination Levels Comment 2

(mR/hr) l Beta-gamma (dpm/I00cm ) Alpha (dpm/1OOcml)

Min Max Min Nax Average Stdev. Min Max Average Stdev 2

3213 Reactor Containment 15 >100 ND 12000 N/A NA ND /A N/A' Enclosure Loop #3 Inner (<1000) (<14.1)

Annulus Mid Ground SW 3214* ReactorContainment 5 >100 ND 14000 N/A2 N/AX ND 375 N/A Enclosure Loop #4 Inner (<1000) (<14.1)

Annulus Mid Ground SE 3301

• Reactor Containment <1 35 ND 42000 16438 13606 ND 385 171 143 Enclosure #1 Outside Crane (<1000) (<14.1)

Charging Floor 3302* Reactor Containment 100000 34850 67415 ND 800 268 N 304 Enclosure #2 Outside Crane (<1000) (<14.1)

Charging Floor 3303 Reactor Containment 500 196 168 Enclosure #3 Outside Crane (<1000) (<14.1)

Charging Floor 3304* Reactor Containment <I 20 ND 16720 8000 7206 ND 275 168 83 Enclosure #4Outside Crane (<1000) (<14.1)

Charging Floor 3311* Reactor Containment 100 ND 70400 14160 21047 ND 500 153 173 Enclosure #1 Inside Crane (<1000) (<14.1)

Charging Floor 3312* Reactor Containment 100000 43493 96785 ND >500 164 191 Enclosure #2 Inside Crane (<1000) (<14.1)

Charging Floor August 2004 2-68 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-11A Nominal Radiological Data Supporting Classifications for Structures Survey SuvyAe Area Deription Survey Radiation Levels Contamination Levels Comment Code' Cd ecito (mRlhr) Beta-gamma (dpm/lOOcm2) Alpha (dpmnlOOcm 2 )

Min Max in MMax Average Stdev. Min Max Average Stdev 3313* Reactor Containment I >100 ND > 100000 56700 97932 ND >500 1844 3001 Enclosure #3 Inside Crane (<1000) (<14.1)

Charging Floor 3314' Reactor Containment I 00000 25711 53478 ND 328 130 116 Enclosure #4 Inside Crane (<1000) (<14.1)

Charging Floor 3315' Reactor Containment 4 >100 6000 18000 N/A 2 N/A 2 <20 N/A2 N/A2 Enclosure Removable Grating for RX Head Staging 3320* Reactor Containment 2.5 >100 1000 >100000 N/A2 >500 N/A 2 NA 2 Enclosure CTMT Rx Refuel Canal to Spent Fuel Pit 3322' Reactor Containment I0 >100 1000 >100000 N/A2 N/A 2 44 >500 N/A2 N/A2 Enclosure CTMT Reactor Refueling Cavity 2

3324' ReactorContainment 4 >100 ND > I 00000 N NFA/A ND >500 25500 20506 Enclosure CTMT Reactor (<1000) (<14.1)

Vessel Area 3326' Reactor Containment >100 >100000 N N/A2 N/A2 N/A 2 Enclosure Upper Core Package Storage Area 2

34030 Reactor Containment <1 20 N/ N/A2 N/A2 Core Co-60 130 Enclosure Inside Surfaces Bore Eu.152 64 (pCi/g) Cs-137 360 August 2004 2-69 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-1lA Nominal Radiological Data Supporting Classifications for Structures Survey Survey Area Area Radiation Levels Contamination Levels Comment Code' Code Decription (mR/hr) Beta-gamma (dpm/I00cm 2) Alpha (dpm/I00cm 2)

Min Max Min Max Average Stdev. Mn Max Average Stdev 2 2 3502* Reactor Containment <1.0 1.5 ND N/A N/A= ND N/A N/A2 Enclosure Outside Surfaces (<1000) (<20) 41020 Turbine Building North 0.01 ND <1000 <1000 Note' ND N/A N/A Floor Area (< 57.2) (<20) 4104' Turbine Building Oil Room, 0.01 ND <1000 <1000 Note ND <100 Note Core Co-60 0.06 Heater Drains, Emergency (< 57.2) (<20) Bore Cs-137 0.04 Power (pCifg) 4106* Turbine Building Air 0.01 ND <1000 <1000 Note' ND <100 Note Compressor Area (< 57.2) (<20) 4108' Turbine Building Steam 0.01 ND <1000 <1000 Note ND <100 Note' Generator Feed Pump Area (< 57.2) (<20) 4110' Turbine Building 0.01 ND <1000 <1000 Note ND <100 Note Chemistry/Closed Cooling (< 57.2) (<20)

Water Area 4112' Turbine Building Water 0.01 ND <1000 <1000 ND <100 Note' Treatment Area (< 57.2) (<20) 4114' Turbine Building 0.01 ND <1000 <1000 NoteND <100 Note' Condenser Pump and South (< 57.2) (<20)

Floor Area August 2004 2-70 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-11A Nominal Radiological Data Supporting Classifications for Structures Aurvey Survey AreaComn roea Code Description Radiation Levels Contamination Levels Comment Code' Cd eeito (mRihr) Beta-gamma (dpm/lOOemI) Alpha (dpmIlOOcmz)

Min Max Min Mlax Average Stdev. Min Max Average Stdev 4116* Turbine Building 0.01 ND <1000 <1000 Note' ND <100 Note' floist/Equipment Laydown (< 57.2) (<20)

Area 4118" Turbine Building 0.01 ND <1000 <1000 Note ND N/A2 N/A' Condenser 'A" Water Box (< 57.2) (<20)

'A & B" Area 4120" Turbine Building 0.01 ND <1000 <1000 Note' ND N/A2 N/A 2 Condenser 1B3Water Box (< 57.2) (<20)

'C & D' Area 4121" Turbine Building Secondary 0.01 ND <1000 N/A2 N/A 2 ND N/A' N/A2 Chem Lab (< 57.2) (<20) 4202" Turbine Building North End 0.02 ND <1000 N/A 2 N/AM ND N/A2 N/A2 Open Area (Walls & (< 57.2) (<20) supports) 4204* Turbine Building Oil 0.02 ND <1000 N/A2 N/A2 ND N/AI N/A2 Reservoir Area (< 57.2) (<20) 4206* Turbine Building S/G 0.02 ND <1000 . N/A' N/A' ND N/A' N/A 2 Feedwater IHeater 2A and (< 57.2) (<20) 2B Area 4208" Turbine Building S/G 0.02 ND <1000 N/A2 N/A 2 ND N/A 2 N/A2 Feedwater Ileater IA and (< 57.2) (<20)

I B Area Rev. 2 2-71 August 20042004 2-71 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-llA Nominal Radiological Data Supporting Classifications for Structures Survey Survey Area Area Code Description Radiation Levels Contamination Levels Comment Code' Cd ecito (mRlhr) Beta-gamma (dpm/lOOcm 2) Alpha (dpm/IOOcmz)

Min Max Min Max Average Stdev. Min Max Average Stdev 4210* Turbine Building Steam 0.02 ND <1000 N/A N/A ND N/A' N/A2 Generator Feedwater (< 57.2) (<20)

Control Valve Area 4212* Turbine Building South 0.02 ND <1000 N/A N/A ND N/A' N/A' End/Turbine Hall (< 57.2) (<20) 4216* Turbine Building S/G 0.02 ND <1000 N/A' N/A' ND N/A' N/A2 Feedwater Heater 6B and (< 57.2) (<20) 5B Area 4218* Turbine Building S/G 0.02 ND <1000 N7A N/A' NDDXA2 N/A' Feedwater Hleater 6A and (< 57.2) (<20) 5A Area 4302* Turbine Building 30" Main 0.02 ND <1000 N/A' N/A' ND N/A' N/A2 Steam Line Area (< 57.2) (<20) 4304* Turbine Building 24" Main 0.02 ND <1000 N/A' N/A' ND N/A' N/A' Steam Line Area (< 57.2) (<20) 4306* Turbine Building MSRHR 0.02 ND <1000 N/A N/A ND N/A IA and I B Area Reheater (< 57.2) (<20) 4308* Turbine Building MSRHR 0.02 ND <1000 N/A N/A ND N/A N/A IC and ID Area Reheater (< 57.2) (<20)

Rev. 2 2-72 2004 August 2004 2-72 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-llA Nominal Radiological Data Supporting Classifications for Structures Survey SuvyAe Area Suorvey Area Radiation Levels Contamination Levels Comment Code' Cd ecito (mRthr) Beta-gamma (dpm/lOOcm') Alpha (dpm/lOOcm)

Min Max Min Max Average Stdev. Min Max Average Stdev 4402* Turbine Building Laydown 0.02 ND <1000 <1000 ND N/A N/A 2 Area North Floor (< 57.2) (<20) 4404* Turbine Building S/G 0.02 ND <1000 <1000 Note' ND N/A2 N/A2 Feedwater Heater 3A Area (< 57.2) (<20) 4406* Turbine Building S/G 0.02 ND <1000 <1000 Ni ND N/A' N/A Feedwater Heater 4A Area (< 57.2) (<20) 4408* Turbine Building S/G 0.02 ND <1000 <1000 Note' ND N/A Feedwater Heater 3 B Area (< 57.2) (<20) 4410* Turbine Building S/G 0.02 ND <1000 <1000 ND NM Feedwater Heater 4B Area (< 57.2) (<20) 4412* Turbine Building i .P. 0.02 ND <1000 <1000 N ND N/A2 Turbine Area (< 57.2) (<20) 44140 Turbine Building L.P. #H1 0.02 ND <1000 <1000 Note' ND N/A2 N/A Turbine Area (< 57.2) (<20) 4416* Turbine Building L.P. #2 0.02 ND <1000 <1000 ND N/A N/A' Turbine Area _ (< 57.2) (<20)

Rev. 2 2-73 2004 August 2004 2-73 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-11A Nominal Radiological Data Supporting Classifications for Structures Survey Survey Area Area Code Description Radiation Levels Contamination Levels Comment Code' Cd ecito (mR/hr) Beta-gamma (dpm/lOOcm 2) Alpha (dpm/100em 2)

Min Max Min Max Average Stdev. Min Max Average Stdev 4418 Turbine Building Generator 0.02 ND <1000 <1000 Note' ND N/A' N/A2 Area (< 57.2) (<20) 4420' Turbine Building Exciter 0.02 ND <1000 <1000 Note ND N/A N/A Area (< 57.2) (<20) 4422* Turbine Building Laydown 0.02 ND <1000 N/A N/A ND N/A N/A Area South Floor (< 57.2) (<20) 4424' Turbine Building Open 0.02 ND <1000 N/A2 N/A2 ND NA2 N/A Hoist Area (< 57.2) (<20) 4502* Turbine Building Overhead 0.02 ND <1000 N/A' N/A2 ND N/A2 N/A2 Crane Area (< 57.2) (<20) 4603* Turbine Building RoofArea 0.02 ND <1000 N/A2 N/AND N/A N/A

(<57.2) (<20) 5102* Service Building "A" Diesel <1 ND <1000 <1000 Note' ND N/A N Generator Area (< 57.2) (<20) 5104* Service Building "B" Diesel <I ND <1000 <1000 NoteN N/A' N/A' Generator Area (< 57.2) (<20)

Rev. 2 2-74 2004 August 2004 2-74 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-llA Nominal Radiological Data Supporting Classifications for Structures Survey SuvyAe Areay Survey Area Radiation Levels Contamination Levels Comment Code Description (mR/hr) Beta-gamma (dpm/lOOcm') Alpha (dpm/lOOcml)

Min Max Min M1ax Average Stdev. Min Max Average Stdev 2

5106* Service Building Clean <c ND <1000 N/A N/A ND NI N/A2 Locker Room Area (< 57.2) (<20) 2 51085 Service Building Hot ND <1000 N/A N/A2 ND N N/A' Locker Room Area (< 57.2) (<20) 2 5110* Service Building HP <1 ND <1000 N/A N/A' ND 7XA N' Control Point and Ofiice (< 57.2) (<20)

Areas 5112' Service Building Woman's < ND <1000 N/A 2 N/A 2 ND N N/A 2 Locker Room Area (< 57.2) (<20) 5114* Service Building Hot <1 <1 ND 30000 N/A2 N/A 2 ND N/Az N/A 2 Chemistry Area (<1000) (<20) 5118* Service Building <1 2 ND 4000 N/A' N/AA ND *717 N/A2 Maintenance Decon Area (<1000) (<20) 5120* Service Building Machine <I ND <1000 N/A' N/A' ND 7M7XT N/A' Shop Clean Area (< 57.2) (<20) 5122* Service Building Machine <1 ND 18000 N/A 2 N/A' ND 7XT N/A' Shop Hot Area (< 57.2) (<20)

August 2004 2-75 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-11A Nominal Radiological Data Supporting Classifications for Structures S vySurvey Ae Area Survey Area Radiation Levels Contamination Levels Comment Code Cd ecito (mR/hr) Beta-gamma (dpm/IOOcm') Alpha (dpm/lOOcml)

Min Max Min Max Average Stdev. Min aUX Average Stdev 2

5124* Service Building <c ND <1000 N7A N/A' ND N/A N/A2 Maintenance Clean Shop (< 57.2) (<20)

Area 5126* Service Building "A" <I ND <1000 1Ai N/A ND N/A N/A Auxiliary Boiler Area (< 57.2) (<20) 5128* Service Building "B" <I ND <1000 N/A N/A ND N/A2 N/A' Auxiliary Boiler Area (< 57.2) (<20) 2 5130* Service Building East <I ND <1000 N/A N/A N/A 2 Hallway (< 57.2) (<20) 5132* Service Building Health <1 ND 4000 2000 1732 ND N/A N Physics Facility I' Floor (< 57.2) (<20) 5134* Service Building Health <I ND <1000 <1000 Note'ND N/A Physics Facility 2nd Floor (< 57.2) (<20) 5202* Service Building Switch <I ND <1000 <1000 NoteN N/A N/A Gear Area (< 57.2) (<20)

N/A 2 2 5302* Service Building Control <I ND <1000 N/A 2 N/A' ND N/

Room Area (< 57.2) (<20)

Rev. 2 2-76 2004 August 2004 2-76 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-11A Nominal Radiological Data Supporting Classifications for Structures Survey SuvyAe Area Survey Area Radiation Levels Contamination Levels Comment Code' Cd ecito (mR/hr) Beta-gamma (dpm/lOOcmz) Alpha (dpm/lOOcm)

M= Max Min Max Average Stdev. Mi Max Average Stdev 5304* Service Building Computer, <I ND <1000 N/A' N/A' ND NMA' N/A' Operations, Security Area (< 57.2) (<20) 5306* Service Building Machine <c ND <1000 N/A ND N/A' N/A and Equipment Area (< 57.2) (<20) 5308* Service Building Instrument <I ND <1000 N/A' N ND N/ N/A and Controls Shop (< 57.2) (<20) 5402* Service Building Roof <0.1 <1 ND ok A'in

(< 57.2) 5502 Screenhouse Building CW <c ND N/A2N/A' N/A' N/A2 System Trench (< 57.2) 6002* Waste Disposal Building <2 60 ND 24000 13333 9452 ND N/A2 N/A Hall Area Lower Level (< 000) (<20) 6004* Waste Disposal Building 0.6 80 ND 80000 25231 27875 ND >500 1371 3003 Area Outside Reboiler (<1000) (<20)

Room August 2004 2-77 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-11A Nominal Radiological Data Supporting Classifications for Structures Survey SuvyAe rea Code Desription Radiation Levels Contamination Levels Comment Code' Cd ecito (mR/hr) Beta-gamma (dpm/lO0cm2 ) Alpha (dpm/lOOcm2)

M=n Max Min M ax Average Stdev. Min Max Average Stdev 6006* Waste Disposal Building 18 >100 3000 >100000 57638 29353 ND >500 1379 3001 Core Am-241 11.03 Bottoms Pump and Reboiler (<20) Bore Co-60 160.55 Area (pCi/g) Cs-134 1.3 Cs-137 264 Nb-94 0.28 Eu- 154 4.05 Eu-155 0.86 6008* Waste Disposal Building <2 20 ND 30000 11821 26451 ND <100 Note Sump Trench Area Lower (<1000) (<20)

Level 6010* Waste Disposal Building- <2 >100 ND 12000 4383 3807 ND 480 209 236 Waste Decay Tank A,B,C (<1000) (<20)

Area 6012* Waste Disposal Building <2 >100 ND 6000 2200 2168 ND N/A 2 N/A' Surge Tank Area Lower (<1000) (<20)

Level 6102* Waste Disposal Building <2 2 ND N/A 2 N/A ND N/A NA Hall Area (<1000) (<20) 6202* Waste Disposal Building 2 ND N/A2 N/A2 ND N/A 2 N/A2 Hallway Area (<1000) (<20) 6304' Waste Disposal Building <2 >100 ND 10000 3589 51057 ND <100 Note Evaporator Area (<1000) (<20)

Rev. 2 2-78 August 2004 2-78 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-11A Nominal Radiological Data Supporting Classifications for Structures Survey SuvyAe Area Sure Arei Radiation Levels Contamination Levels Commcnt Code Code Description (mR/hr) Beta-gamma (dpm/l1Ocm2 ) Alpha (dpmIlOOcm2)

Min Max Min Max Average Stdev. Min Max Average Stdev 6306* Waste Disposal Building 8 100 ND > 100000 26625 45325 ND >500 190 215 Radwaste Liquid (<1000) (<20)

Evaporator 6308* Waste Disposal Building I >100 ND 80000 13236 32868 ND >500 144 204 Degassifier Transfer Pump (<1000) (<20)

Area 6312* Waste Disposal Building 1 >100 ND 40000 7379 25544 ND >500 203 261 Degassifier and Associated (<1000) (<20)

Valves 6404* Waste Disposal Building I 22 ND 80000 27667 15984 ND 331 N/A2 N/A2 Evaporator Area (<1000) (<20) 6406* Waste Disposal Building 20 60 ND 20000 28667 10440 ND 144 N/A 2 N/A 2 Liquid Evaporator Area (<1000) (<20) 6408* Waste Disposal-Waste Gas I >100 ND 20000 8000 4413 ND 350 N/A2 N/A 2 Compressor A&B Area (<1000) (<20)

N/A2' 2 6412* Waste Disposal Building I >100 ND >100000 30500 59072 ND >500 NA Degassifier Area and (< 1000) (<20)

Associated Valves 6502* Waste Disposal Building Roof Area Rev. 2 2-79 August2004 August 2004 2-79 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-11A Nominal Radiological Data Supporting Classifications for Structures Survey SuvyAe Area Code Description Radiation Levels Contamination Levels Comment Code' Cd ecito (mR/hr) Beta-gamma (dpmIlOOcm 2 ) Alpha (dpmnlOOcmz)

Min Max Min Max Average Stdev. Min Max Average Stdev 7002 Screenhouse Building CW <0.1 ND <1000 ND Cire Pump A&B Head (<57.2) (<14.)

7004* Screenhouse Building CW <0.1 ND <1000 ND Circ Pump C&D Head (< 57.2) (<14.)

7102* Screenhouse Building CW <0.1 ND <1000 ND Circ Pump Motor A&B (< 57.2) (<14.)

71041 Screenhouse Building CW <0.1 ND <1000 ND Cire Pump Motor C&D (< 57.2) (<14.)

7106* Screenhouse Building CW <0.1 ND <1000 ND IHypochlorite Tank Area (< 57.2) (<14.)

7108* Screenhouse Building CW <0.1 ND <1000 ND Intake and Screen Area (< 57.2) (<14.)

7202* Screenhouse Building CW <0.1 ND <1000 ND Roof Area (< 57.2) (<14.)

8100 Penetration Building Upper <0.1 ND <1000 Level (< 57.2)

August 2004 2-80 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-llA Nominal Radiological Data Supporting Classifications for Structures Survey Survey Area Area Radiation Levels Contamination Levels Comment Code' Cd ecito

_ (mRlhr) Beta-gamma (dpm/l00cm2) Alpha (dpm/l1OOcm 2)

M=n Max Min Max Average Stdev. Min Max Average Stdev 8200* Penetration Building Mid <0.1 ND <1000 Level (< 57.2) 8300* Penetration Building Lower <0.1 ND <1000 Level (< 57.2) 9108* YD North Tank Farm Area 0.8 >100 ND 10000 N-A2 N/A' ND 60 N/A

(<1000) (<20) 9110* YD South Tank Farm Area 2 >100 ND 50000 19000 26851 ND 132 N/A'

(<1000) (<20) 9112* YD Boron Storage Tank 15 >100 ND >100000 78000 115986 ND >500 30545 59643 Area (<1000) (<20) 9114* YD lon Exchange Area 1 >100 ND >100000 7A N/A' ND >500 N/A' N/A'

(<1000) (<20) 9116* YD Resin Slurry Area <0.2 >100 ND >100000 N/A ND >500 N/A N/

(<1000) (<20) 91180 YD Fuel Oil Tank Area <0.5 N/A N/A' N/A August 2004 2-81 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-11A Nominal Radiological Data Supporting Classifications for Structures SArvey Survey Area Radiation Levels Contamination Levels Comment Code Cd ecito (mR/hr) Beta-gamma (dpm/lOOcm2 ) Alpha (dpm/lOOcm')

Min Max M1in Max Average Stdev. Min Max Average Stdev 9120* YD Primary Vent Stack <0.2 70 ND > 100000 97440 153667 ND >500 809 1037

(<1000) (<20) 9126* YD Large Yard Crane Area <0.2 20 ND 1500 1167 289 N/A' N/A

(<1000) 9128* YD Demin Water Storage <0.2 <0.5 ND N/A NM N/A N/A Tank Area (<1000) 9202 Switchgear Building 'B" <0.2 ND N/A 2 NiA' N/A

(<57.2) 9208 Administration Building <0.2 ND <1000 ND N/A 2 N/A

(<57.2) (<20) 9214* Shutdown Auxiliary Feed N/A' N/A2 N/A2 N/A Pump House 2

9226 Radwaste Reduction <0.5 150 ND 2000 1143 378 N/A N/A 2 Facility (<1000) 9228 Unconditional Release <0.1 ND 5000 N/A2 N/A' ND N/A2 N/A' Facility (<45.4) (< 8.92)

Rev. 2 2-82 2004 August 2004 2-82 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-llA Nominal Radiological Data Supporting Classifications for Structures Survey Sre~e Area Survey Area Radiation Levels Contamination Levels Comment Code Cd ecito 2

_ (mR/hr) Beta-gamma (dpm/lOOcm ) Alpha (dpm/lOOcm')

Mn Max Min Max Average Stdev. Min Max Average Stdev 2 2 2 9402 Emergency Operations <0.2 ND N/A N/A ND N/A N/A' Facility (< 57.2) (<20 )

9403* Emergency Operations N/A 2 N/A2 N/ 2 N/AN Roof Cs-137 <0.18 Center Roof (pClg) 9404 North Warehouse <0.2 ND N/A2 N/A' ND N/A' NIA'

(<57.2) (<20) 9406 South Warehouse <0.2 ND <1000 Note' ND N/A2 N/A'

(<57.2) (<20) 2 2 9408' Miscellaneous Trailer <0.2 ND N/A ND NA Complex (< 57.2) (<20) 9410' Steam Generator Mockup <0.2 ND N/A' N/A' ND N/A' N/A' Building (< 57.2) (<20) 9412 Training Stores Office <0.2 ND N/A2 N/A' ND N/A' N/A' Building (< 57.2) (<20) 9414* Warehouse #1 <0.2 ND N/A2 N/A' 132 N/A 2 N/A'

(<57.2)

August 2004 2-83 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-11A Nominal Radiological Data Supporting Classifications for Structures Survey Survey Area Area Rtadiation Levels Contamination Levels Comment Code' Cd ecito (mR/hr) Beta-gamma (dpm/lOOcm2) Alpha (dpm/l00em')

Mini Max Min Max Average Stdev. Mlii Max Average Stdev 9416 Warehouse #2 <0.2 ND N/A N/A= ND T7Xy N/A'

(<57.2) (<20) 9481 O Building #3 and PAP <0.2 ND < N /A 2 Soil Cs-137 <0.18

(<57.2) (<20) (pCi/g) 94200 Office Trailer <0.2 ND N/A2 N/A 2 ND N/A N/A 2

(<57.2) (<20) 9422 Information Center <0.2 ND <1000 Note ND N/A' N/A 2

(<57.2) (<20) 9423* Information Center Roof N/A 2 N/A 2 N/A2 N/A 2 94241 All Buildings Contained in <I ND N/A 2 NM/2 ND 7/A= N/A 2 the Southwest Site Storage (<1000) (<20)

Area I All measurements less than MDA, stdev. not valid.

2 Insufficient data.

August 2004 2-84 Rev. 2 1

Haddam Neck Plant License Termination Plan Table 2-11B Nominal Radiological Data Supporting Classifications for Land Areas Survey Survey Area Radiation Levels Sample analysis Results Area Code Description Code (mR/hr) Concentration (pCi/g) Comments Min l Max Medium Nuclide Max 9102 YD 115KV Switchyard Area 0.015 1 Soil Co-60 2.87 Cs-137 4.49 9104* YD Main Transformer Area <0.5 9106 Discharge Canal <0. 1< Sediment Co-60 0.5 Cs-134 0.024 Cs-137 0.722 9122* YD Primary Water Storage Tank <0.2 5 Soil Co-60 521 Maximum concentration identified Area Cs-134 153 and removed during remediation

. Cs-137 793 9124 YD Backup Primary Water Storage <0.2 5 Soil Co-60 84.8 Maximum concentration identified Tank Area Cs-134 20.2 and removed during remediation Cs-137 155.2 9227* Busl3 5 Soil Co-60 22 Maximum concentration identified Cs-137 735.2 and removed during remediation 9302 Northwest Protected Area Grounds Soil Co-60 <0.12 Asphalt Cs-137 <0.10 August 2004 2-85 Rev.2

Haddam Neck Plant License Termination Plan Table 2-11B Nominal Radiological Data Supporting Classifications for Land Areas Survey Survey Area Radiation Levels Sample analysis Results Area Code Description Code (mR/hr) Concentration (pCi/g) Comments Min Max Medium Nuclide Max 9304 Southwest Protected Area Grounds 0.007 0.013 Soil Co-60 <0.10 Asphalt Cs-137 0.04 9306 South Central Protected Area Soil Co-60 0.532 Max concentration based on asphalt Grounds Asphalt Cs-137 0.775 results 9307 PAB / Service Building Alleyway 0.016 10 Co-60 202.4 Max concentration based on asphalt Cs-137 97.14 results 9308 Southeast Protected Area Grounds 0.009 0.016 Soil Co-60 0.31 Cs-137 0.33 9310 East Protected Area Grounds Soil Co-60 213.7 Max Cs-137 concentration based on Asphalt Cs-137 3095 asphalt results 9312 Northeast Protected Area Grounds 0.032 >100 Asphalt Co-60 0.31 Cs-137 7.94 9313 Central Site Grounds Soil Co-60 0.174 Cs-137 1.592 9502 Northeast Site Grounds (Non- Soil Cs-137 0.407 Protected Area)

Rev. 2 2-86 2004 August 2004 2-86 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-liB Nominal Radiological Data Supporting Classifications for Land Areas Survey Survey Area Radiation Levels Sample analysis Results Area Code Description Code (mR/lhr) Concentration (pCi/g) Comments Min Max Medium Nuclide Max 9504 Bypass Road / Secondary Parking Soil Cs-137 <0.18 Lot 9506 North Site Grounds (Non-Protected Soil Cs-137 0.08 Area) 9508 Pond Sediment Cs-137 0.233 9510 Access Road 9512 Northwest site Grounds (Non- 0.008 0.013 Soil Cs-137 0.238 Protected Area) 9514 Primary Parking Lot 0.012 0.015 Soil Cs-137 0.076 9518 Southwest Site Grounds (Non-Protected Area) 9520 Southwest Site Storage Area 0.003 0.012 Soil Co-60 0.13 Cs-137 6.30 August 2004 2-87 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-l1B Nominal Radiological Data Supporting Classifications for Land Areas Survey Survey Area Radiation Levels Sample analysis Results Area Code Description Code .-

(mR/hr) Concentration (pCi/g) Comments Min Max Medium Nuclide Max 9521 Southeast Pond 0.009 0.013 Soil Cs-137 0.596 9522 Southeast Site Grounds (Non- 0.012 0.015 Soil Co-60 7.50 Maximum concentrations could not Protected Area) Cs-137 7.20 be verified during 1998 scoping 9523 Southeast Wetland Area 9524 South Site Grounds (Non-Protected Soil Cs-137 <0.18 Area) 9525 Southeast Site Road Soil Cs-137 <0.18 9526 Northeast Mountain Side Soil Cs-137 0.79 9527 East Mountain Side 0.061 0.132 Soil Co-60 0.453 Cs-137 1.69 9528 Southeast Mountain Side Soil Cs-137 1.898 Rev. 2 2-88 August 2004 2-88 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-11B Nominal Radiological Data Supporting Classifications for Land Areas Survey Survey Area Radiation Levels Sample analysis Results Area Code Description Code (mR/lhr) Concentration (pCi/g) Comments Min Max Medium Nuclide Max 9530 Central Peninsula Area 0.008 0.014 Soil Co-60 0.041 Maximum concentrations identified Cs-137 1.69 with Class 2 area 9531 South End of Peninsula 9532 East Site Grounds (Non-Protected Non-impacted area Area) 9535 South East Landfill Area 0.011 Soil Co-60 5 Cs-137 52.9 9536 Construction Piles Near Rifle 0.009 0.3 Soil Cs-137 0.149 Range 9537 Permitted Landfill Area 0.006 0.01 Soil Cs-137 0.039 9538 Material Storage Area 0.006 0.011 Soil Cs-137 0.119 August 2004 2-89 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-l1C Nominal Radiological Data Supporting Classifications for Subsurface Areas Survey Area Soil Sample Results (pCi/g)

Code Radionuclide Max Comments 9801 Cs-137 127 Data reviewed from over 100 samples collected Co-60 9 in 1998 and 1999 in support of plant Am-241 0.3 modifications and characterization studies.

Maximum sample from under PAB pipe trench floor.

9802 Cs-137 16.7 Data reviewed from approximately 150 samples Co-60 0.65 collected from 1997 to 1999 in support of plant modifications and characterization studies.

Maximum sample in canal storm drain discharge area.

9803 Cs-137 0.076 Data reviewed from approximately 20 samples Co-60 MDA collected in 1999 in support of plant modifications and to support Partial Site Release.

9804 Cs-137 N/A No recent subsurface data is available; however, Co-60 subsurface soil data obtained in 1998 from the adjacent, northerly subsurface area, the RCA portion of 9802, was reviewed. Forty-two subsurface samples were obtained from twelve locations. Several samples were obtained down to a depth of 2 meters. The only plant-related gamma emitting radionuclide detected was Cs-137 with a maximum concentration equal to 2%

of the proposed base-case soil DCGL.

9805 Cs-137 6.3 Data reviewed from approximately 40 samples Co-60 MDA collected in 1998 for characterization study.

9806 Cs-137 58.5 Date from remedial action survey perform-ned in Co-60 2.21 2003.

Sr-90 0.442 2.3.3.3 Non-Impacted Area Assessment Non-impacted areas are those areas having no reasonable potential for residual contamination. Non-impacted areas are typically identified during initial classification using historical data and past or current radiological surveillance. Non-impacted areas should have no history of using, storing, or burying radioactive materials. Records and surveillances, including those required by IOCFR50.75(g)(1), should August 2004 2-90 Rev. 2 I

HNP License Termination Plan show that unplanned liquid releases, discharges and other occurrences have not resulted in the spread of contamination in these areas.

The Connecticut Yankee Haddam Neck Characterization Report has classified the East Site Grounds (Survey Area 9532) as non-impacted. This area consists of approximately ninety-three (93) acres of uninhabited, undeveloped land located about a third of a mile (0.29 miles) from the RCA (Radiologically Controlled Area). The East Site Grounds are bounded by steep, wooded hillsides to the east, an open clearing for power distribution lines to the south and west, and an access road (Wood Road) from the substation to the discharge canal to the north. Access to the interior of the East Site Grounds area is limited to a gated road (Cove Road) to the east, abandoned or seldom used logging paths and trails, and the power transmission clearing to the west. A walk down and visual inspection of the East Site Grounds area indicates the land was not used to store materials from the Haddam Neck Plant. There were no identified soil disturbances that would indicate dumping or burial of materials.

Historical data do not indicate that plant operations had an impact on the East Site Grounds area.

Historical data and radiological surveys have identified contamination from plant operations on the east hillside from the RCA boundary out to a distance of 200 meters or roughly an eighth of a mile (0.12 miles). Following identification of the contamination further surveys were conducted to a distance of approximately 400 meters, with no additional plant related radioactivity identified. Given the topography of the eastern hillside in general, and the distance from the RCA to the nearest boundary of the East Site Grounds area (0.29 miles), past occurrences from plant operations would not have had any radiological impact on the East Site Grounds.

Radiological environmental monitoring and sampling is performed in the East Site Grounds area in accordance with the Radiological Effluent Monitoring and Off-site Dose Calculation Manual.

Radiological analyses are performed with gamma exposure measuring devices, on samples of air particulates and iodine, and on broad leaf vegetation. The Haddam Neck Station Annual Radiological Reports show no long-lived radionuclides other than Sr-90 and Cs-137 above the MDL (Minimum Detectable Level). These radionuclides (Sr-90 and Cs-137) are measured at levels consistent with those found throughout the central Connecticut area and are attributed to past atmospheric nuclear weapons testing. It is important to note that the Haddam Neck Station Annual Radiological Reports consider all data statistically valid, including negative values, zeros, numbers below the MDL and those values with reporting errors greater than two standard deviations. The Haddam Neck Station Annual Radiological Reports present all valid data for strictly counting statistics purposes and to indicate background biases.

The historical data, use and topography of the land and radiological environmental monitoring results support the classification of the East Site Grounds area as a non-impacted area. In a letter dated April 29, 2004, CYAPCO provided written notification to the NRC of the intent to release the East Site Grounds (Survey Area 9532) from its Part 50 license.

2.3.3.4 Radionuclide Suite Selection As a part of site characterization, the list of radionuclides expected to be encountered during decommissioning was established. The radionuclides listed in Table 2-12 were identified using available waste characterization data from the HNP site.

August 2004 2-91 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-12 Radionuclides Potentially Present at HNP Radionuclide Half-lifes (years)

H-3 12.33 C-14 5,730 Mn-54 0.8561 Fe-55 2.685 Co-60 5.271 Ni-63 100 Sr-90 28.8 Nb-94 2.Ox 104 Tc-99 2.14x10 5 Ag-108m 1.27 x 102 Cs-134 2.062 Cs-137 30.17 Eu-152 13.3 Eu- 154 8.5 Eu-155 4.96 Pu-238 87.74 TPu-239 '- 2.41x10 4 Pu-241 14.4 Am-241 432.2 Cm-243 28.5 One DCGL was established for the Pu-239/240 pair and the Cm-243/244 pair, as laboratory radiochemical analyses do not report concentrations of these radionuclides separately. Bold indicates those radionuclides that are considered to be hard to detect.

Available regulatory documents, addressing radionuclides existing in bio-shield wall concrete and rebar and surface contamination, were reviewed in order to determine if other dose-significant radionuclides existed and should be included. NUREG/CR-3474, "Long-Lived Activation Products in Reactor Material" (Reference 2-9), Tables 5.4 and 5.6, and NUREG/CR-0130, "Technology, Safety and Cost of Decommissioning a Reference Pressurized Water Reactor Power Station" (Reference 2-10), Tables 7.3-5, 7.3-1 1 and 7.3-14, were used in this determination. A number of the radionuclides listed in the regulatory guidance have been found to be not significant to HNP based upon their extremely low dose contribution.

The methodology used to reach that conclusion follows.

August 2004 2-92 Rev. 2 I

HNP License Termination Plan The radionuclide inventory values, from the NUREGs referenced above, were used to determine the activity ratio of a given nuclide to that of Co-60. Co-60 was used as a reference nuclide, since it is a predominant beta/gamma emitting nuclide at the HNP. For situations in which a radionuclide was included more than once in the references, the highest ratio was selected as being representative of the relative activity for that nuclide. This ratio (representing activity at the time of reactor shutdown) was adjusted to represent decay to the current period (i.e., April 1, 2002). Doses, relative to Co-60, were calculated for the following exposure pathways:

• Inhalation Exposure

• Ingestion Exposure

• External Exposure to a Plane Source

• External Exposure to an Infinitely Thick Soil Contamination Source These relative doses were calculated by multiplying the applicable dose conversion factors (from FGR-l I and FGR-12) by the relative decayed activity. Table 2-13 provides a summary of the relative doses from each of the pathways, the maximum relative dose for each nuclide, and dose fraction relative to the total dose for each nuclide (Reference 2-1 1). This table is sorted in descending order by the dose fraction for each radionuclide relative to the total dose.

August 2004 2-93 Rev. 2

Haddam Neck Plant License Termination Plan Table 2-13 Summary of Radionuclide Analysis Fraction of Co-60 Dose _Maximum Pathwav Max Dose Fraction Relative to of Total Radionuclide Inhalation Ingestion Plane Infinite Co-60 Dose Cs-137 2.72E+00 3.46E+01 2.26E-03 8.64E-04 3.46E+01 6.48E-OI Am-241 4.28E+00 2.85E-01 2.47E-05 5.69E-06 4.28E+00 8.03E-02 Pu-238 3.65E+O0 2.42E-01 7.26E-07 1.90E-08 3.65E+00 6.84E-02 Cm-243 2.60E+00 1.73E-01 9.85E-05 6.66E-05 2.60E+00 4.88E-02 Sr-90 2.21 E+OO 1.96E+00 4.49E-05 1.61 E-05 2.21 E+OO 4.14E-02 Eu-152 1.74E+00 4.14E-01 8.06E-01 7.44E-01 1.74E+00 3.26E-02 Cs-134 1.04E-01 1.34E+00 3.19E-01 2.88E-01 1.34E+00 2.52E-02 Co-60 1.25E-11 3.35E-11 1.02E-10 9.26E-1 I.OOE+OO 1.87E-02 Pu-239 7.60E-01 5.08E-02 6.05E-08 7.05E-09 7.60E-01 1.42E-02 Fe-55 2.69E-01 4.93E-OI O.OOE+OO O.OOE+OO 4.93E-01 9.25E-03 Eu-154 4.51 E-O I 1.22E-01 1.75E-01 1.63E-01 4.51 E-O I 8.45E-03 H-3 1.03E-02 8.33E-02 O.OOE+OO O.OOE+OO 8.33E-02 1.56E-03 Pu-241 6.09E-02 4.1 OE-03 1.33E-09 5.88E-10 6.09E-02 1.14E-03 Hf-178m 3.26E-02 2.26E-03 2.84E-03 2.32E-03 3.26E-02 6.1 1E-04 U-233 1.68E-02 2.91 E-04 8.26E-09 2.34E-09 1.68E-02 3.15E-04 Ni-63 1.23E-02 9.16E-03 O.OOE+OO O.OOE+OO 1.23E-02 2.31 E-04 Na-22 3.85E-04 4.68E-03 9.81 E-03 9.26E-03 9.81 E-03 1.84E-04 Ca-41 7.15E-04 5.49E-03 O.OOE+OO O.OOE+OO 5.49E-03 1.03E-04 Sm-151 1.78E-03 1.87E-04 2.78E-08 7.87E-10 1.78E-03 3.33E-05 Mn-54 1.56E-04 5.22E-04 1.75E-03 1.61 E-03 1.75E-03 3.29E-05 Ba-133 3.1 OE-04 1.IOE-03 1.47E-03 1.06E-03 1.47E-03 2.75E-05 Eu-155 1.40E-03 4.18E-04 1.85E-04 8.28E-05 1.40E-03 2.62E-05 C-14 1.57E-04 1.27E-03 1.13 E-07 1.36E-08 1.27E-03 2.39E-05 Ho-166m 1.13E-03 9.53E-05 2.30E-04 2.02E-04 1.13E-03 2.1 E-05 Zn-65 5.09E-05 2.92E-04 1.28E-04 1.24E-04 2.92E-04 5.48E-06 Sb-125 5.64E-05 1.05E-04 1.83E-04 1.53E-04 1.83E-04 3.43E-06 Ca-45 4.68E-05 1.81 E-04 3.03E-08 5.96E-09 1.81E-04 3.40E-06 Nb-94 1.38E-04 1.92E-05 4.73E-05 4.33E-05 1.38E-04 2.58E-06 CI-36 9.52E-05 1.07E-04 2.72E-07 1.40E-07 1.07E-04 2.00E-06 g-108m 5.98E-05 1.31 E-05 3.14E-05 2.74E-05 5.98E-05 1.12E-06 Ni-59 4.69E-05 2.95E-05 O.OOE+OO O.OOE+OO 4.69E-05 8.78E-07 Ce-144 3.81 E-05 1.74E-05 1.93E-07 9.86E-08 3.81 E-05 7.14E-07 Mo-93 2.76E-05 1.06E-05 4.83E-07 7.74E-09 2.76E-05 5.18E-07 Ag-I lOim 5.78E-06 6.311E-06 1.78E-05 1.67E-05 1.78E-05 3.33E-07 Nb-93m 8.79E-06 1.27E-06 2.63E-08 4.22E-10 8.79E-06 1.65E-07 Tc-99 5.97E-06 8.511E-06 5.21 E-09 1.21 E-09 8.511E-06 1.60E-07 Pm-145 6.35E-06 8.02E-07 6.33E-07 8.25E-08 6.35E-06 1.19E-07 August 2004 2-94 Rev. 2 I

HNP License Termination Plan Table 2-13 Summary of Radionuclide Analysis Fraction of Co-60 Dose Maximum Pathway Max Dose Fraction Relative to of Total Radionuclide Inhalation Ingestion Plane Infinite Co-60 Dose Tb-158 2.42E-06 3.39E-07 6.8112-07 6.04E-07 2.42E-06 4.55E-08 Co-57 6.15E-07 6.52E-07 7.26E-07 4.58E-07 7.26E-07 1.36E-08 Zr-93 4.26E-07 1.79E-08 O.OOE+00 O.OOE+00 4.26E-07 7.99E-09 Sn-121m 3.08E-07 3.37E-07 1.22E-08 7.08E-10 3.37E-07 6.31 E-09 Se-79 1.OOE-09 7.19E-09 1.96E- 13 2.55E-14 7.19E-09 1.35E-10 Cs-135 1.91E-10 2.41E-09 1.30E-13 2.17E-14 2.41E-09 4.52-11I Sm-146 1.42E-09 2.86E-1 I O.OOE+00 O.OOE+00 1.42E-09 2.67E-1 1 1-129 1.08E-10 1.39E-09 1.49E-12 1.08E-13 1.392-09 2.60E-11 Mn-53 6.15E-10 1.082-09 0.002+00 0.00+00 1.08E-09 2.03E-I I Pb-205 4.86E-11 1.64E-10 1.73E-13 1.18E-15 1.64E-10 3.08E-12 S-35 6.75E-11 1.62E-10 4.27E-14 5.48E-15 1.62E-10 3.04E-12 Co-58 1.25E-11 3.35E-1I 1.02E-10 9.26E-11 1.02E-10 I.91E-12 Sc46 2.322-12 4.06E-12 1.40E-11 1.34E-11 1.40E-11 2.63E-13 Zr-95 3.46E- 13 4.492-13 9.86E- 13 8.93E-13 9.86E-13 1.85E-14 Sb-124 1.07E-14 3.49E-14 6.76E-14 6.70E-14 6.76E-14 1.27E-15 Fe-59 1.38E-15 5.06E-15 9.70E-15 9.59E-15 9.70E-15 1.82E-16 Sr-89 1.97E-15 3.57E-15 1.002- 17 5.82E- 18 3.57E-15 6.70E- 17 Ru- 103 7.26E-21 2.01 E-20 3.492-20 3.OOE-20 3.492-20 6.552-22 Nb-95 6.72E-22 2.422-21 8.06E-21 7.32E-21 8.06E-21 1.512E-22 Te-129m 1.492-22 5.41 E-22 2.19E-23 1.56E-23 5.412E-22 1.01 E-23 Ce-141 1.092-25 2.87E-25 8.37E-26 5.222-26 2.87E-25 5.382-27 Cr-51 7.772-27 2.782-26 6.66E-26 5.47E-26 6.662-26 1.252-27 P-33 1.432-27 4.60E-27 2.56E-30 4.92E-31 4.602-27 8.62E-29 Cs-136 5.93E-52 7.39E-51 1.572-50 1.452-50 1.572-50 2.95E-52 Ba-140 2.57E-54 5.29E-53 1.15E-53 9.562-54 5.29E-53 9.911E-55 Y-90 1.60E-239 1.66E-238 9.41 E-241 6.13E-241 1.662-238 3.11 E-240 La-140 0.002+00 O.OOE+00 0.002+00 O.OOE+00 0.002+00 O.OOE+00 1-133 0.002+00 0.002+00 0.002+00 O.OOE+00 O.OOE+00 O.OOE+00 The data provided in Table 2-13 allows the selection of significant nuclides based on both the expected relative abundance and the dose potential for each nuclide. The criterion for considering a radionuclide as being dose-significant was set at a fraction of total dose significance of 0.1% or greater. Although they do not meet the criteria for selection based upon dose significance, C-14, Mn-54, Ni-63, Nb-94, Tc-99, Ag-108m and Eu-155, listed in Table 2-12, are considered important for inclusion for HNP, as these radionuclides have been reported in plant waste streams by easy-to-detect methods (i.e., gamma spectroscopy) or are anticipated as a result of activation.

August 2004 2-95 Rev. 2 I

Haddam Neck Plant License Termination Plan Based upon this information, no additional radionuclides were selected for inclusion, and the list of radionuclides in Table 2-12 is considered to be the list of radionuclides of concern for HNP.

2.3.3.5 Hazardous Material Status The characterization process included the identification of hazardous materials and state of Connecticut regulated materials. The characterization process coupled the radiological and hazardous material evaluations such that resultant characterization report for each area included an assessment of materials known to be present as well as those further analyses needed to fully define the existence and scope of materials present. The hazardous material characterization effort used the same site procedure, following the methodology described in Section 2.3.2. As indicated in that section, a critical element of the characterization effort included a walk-down of each area by a professional experienced in hazardous and state-regulated materials.

The review of historical records and the familiarity of personnel performing the characterization with plant operations identified that the major hazardous materials encountered at HNP are asbestos, lead, PCBs and mercury. These materials are typically contained in building materials, paints, light bulbs, light fixtures, switches, electrical components and high voltage cables. In addition to the above materials, temporary RCRA waste storage areas were maintained on site in compliance with federal requirements.

These storage areas are identified in the area characterization reports, with further evaluation required to determine the extent, if any, of hazardous material contamination in those areas. An example of a building containing a RCRA waste storage area (90 day storage) is the North Warehouse, Building No. 160.

Full details of hazardous and state regulated materials identified in each survey area, and the additional actions and evaluations necessary to ensure the appropriate definition of the extent of the hazardous materials is presented in the "Connecticut Yankee Haddam Neck Plant Characterization Report," dated January 6, 2000.

2.4 References 2-1. NUREG-1575, "Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM),"

dated December 1997.

2-2. "Connecticut Yankee Haddam Neck Plant Characterization Report," dated January 6, 2000.

2-3. "Haddam Neck Plant Historical Site Assessment Supplement," dated August 14, 2001.

2-4 Investigation of the Source of Radioactive Contamination Found on the Connecticut Yankee Site March 10-30, 1980, dated April 1980.

2-5 Results of Phase 2 PCB and Radiological Characterization Study, CY Letter HP-98-423, dated July 28, 1998.

2-6 Executive Summary of Radiation Surveys Performed at Connecticut Yankee Atomic Power Station, dated January 22, 1998, Millennium Services, Inc.

2-7 Groundwater Monitoring Report, Connecticut Yankee Atomic Power Station Haddam Neck, Connecticut, by Malcom Pimey, Inc., dated July 1999, revised September 1999.

s Aiicrimt 7A04 Bt>-i _vv g 7-96

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HNP License Termination Plan 2-8 "Haddam Neck Plant Phase 2 Hydrogeologic Investigation Work Plan," May 2002.

2-9 NUREG/CR-3474, "Long-Lived Activation Products in Reactor Materials, dated August 1984.

2-10 NUREG/CR-0130, "Technology, Safety, and Costs of Decommissioning a Reference Pressurized Water Reactor Power Station," June 1978.

2-11 Technical Support Document, BCY-HP-0023, Revision 1,"Radionuclide List for DCGL Calculation in Support of the LTP."

August 2004 2-97 Rev. 2

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

August 2004 2-98 Rev. 2 l

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4 Haddam Neck Plant License Termination Plan 3 IDENTIFICATION OF REMAINING SITE DISMANTLEMENT ACTIVITIES 3.1 Introduction In accordance with IOCFR50.82 (a)(9)(ii)(B) (Reference 3-1), the LTP must identify the major dismantlement and decontamination activities that remain. The information includes those areas and equipment that need further remediation and an estimate of the radiological conditions that may be encountered. Included are estimates of associated occupational radiation dose and projected volumes of radioactive waste. These activities are undertaken pursuant to the current I OCFR50 license, are consistent with the PSDAR, and do not depend upon LTP approval to proceed.

CYAPCO's primary goals are to decommission the HNP safely and to maintain the continued safe storage of spent fuel. CYAPCO will decontaminate and dismantle the HNP in accordance with the DECON alternative, as described in the NRC's Final Generic Environmental Impact Statement.

Completion of the DECON option is contingent upon continued access to one or more low level waste disposal sites. Currently, HNP has access to low-level waste disposal facilities including those in Barnwell, South Carolina.

CYAPCO is currently conducting active decontamination and dismantlement activities at the HNP site in accordance with the HNP PSDAR (Reference 3-2). Decommissioning activities are being coordinated with the appropriate Federal and State regulatory agencies in accordance with plant administrative procedures. In order to minimize the impact of ongoing decommissioning activities, a Spent Fuel Pool Island has been established to separate spent fuel storage functions from other plant functions and other decommissioning activities.

Decommissioning activities at Haddam Neck will be conducted in accordance with the Haddam Neck UFSAR, Technical Specifications, existing Part 50 License and the requirements of l OCFR50.82(a)(6) and (a)(7). If an activity requires prior NRC approval under 10CFR50.59(c)(2) or a change to the Haddam Neck Plant Technical Specifications or license, a submittal will be made to the NRC for review and approval before implementing the activity in question. Decommissioning activities are conducted under the scrutiny of the existing CYAPCO Radiation Protection Program, Industrial Safety Program, and Waste Management Program. Such activities will be conducted in accordance with these programs, which are well established and frequently inspected by the NRC. Activities conducted during decommissioning do not pose any greater radiological or safety risk than those conducted during operations, especially those during major maintenance and outage evolutions.

Decontamination and dismantlement activities continue to be performed, as described in Section 3.3, while taking into account the specific system considerations as discussed in Sections 3.4.1 and 3.4.2.

These sections provide an overview and describe the major remaining components of contaminated plant systems and, as appropriate, a description of specific equipment remediation considerations. Table 3-1 contains a list of major systems and components that have been or are to be removed.

August 2004 3-1 .,- Rev. 2

Haddam Neck Plant License Termination Plan Table 3-1 Status of Major HNP Systems, Structures, and Components as of May 2004 I

SSC Required for Wet SSC Status Fuel Storage?

Reactor Coolant System No Partially Removed Reactor Vessel Internals No Segmentation completed Reactor Vessel No Removed from site Steam Generators No Removed from site Reactor Coolant Pumps No Removed from site Pressurizer No Removed from site Chemical and Volume Control No Partially removed System Safety Injection System No Partially removed Il Residual Heat Removal No Partially removed System Containment Spray System No Partially removed Component Cooling Water No Partially removed System Service Water System No Partially removed Spent Fuel Pool Yes In place Fuel Handling Equipment No In place Spent Fuel Pool Cooling and Yes In place Demineralizer System Condensate System No Partially removed Feedwater System No Partially removed Steam Generator Blowdown No Partially removed System Primary Makeup Water No Partially removed System Refueling Water Storage Tank No Removed from site I Plant Effluent Monitoring No In place System Containment Ventilation No Partially removed System Fuel Building Ventilation Yes In place System PAB Ventilation System No Partially removed I Auxiliary Boiler No Removed from site Instrument and Service Air No Partially removed System Gaseous Radioactive Waste No Partially removed System Solid Radioactive Waste No Partially removed System August 2004 3-2 Rev. 2

Haddam Neck Plant License Termination Plan Table 3-1 Status of Major HNP Systems, Structures, and Components as of May 2004 I

SSC Required for Wet SSC Status Fuel Storage?

Liquid Radioactive Waste No Partially removed System Makeup Water System Yes In place Radioactive Monitoring Yes In place System Process Sampling System No Partially removed Fire Protection System Yes (portions) Partially removed Electrical Systems Yes (portions) Partially removed Containment Building No Some equipment removed.

Decontamination activities are in progress.

Primary Auxiliary Building No Majority of equipment removed.

Fuel Building Yes In place Turbine Building No Majority of equipment removed.

Service Building No Majority of equipment removed.

3.2 Spent Fuel Pool Island Activities Plant closure activities were initiated following the decision to permanently cease HNP power operations in December 1996. At that time, CYAPCO performed evaluations of major plant Systems, Structures, and Components (SSCs) to determine what function, if any, these SSCs would be expected to perform during I the evolution to a decommissioned plant with a Spent Fuel Pool Island. Each major plant SSC was evaluated to determine if the SSC, in its entirety or any portion thereof, was required to support maintaining the spent fuel in a safe condition or was needed to perform a function during the decontamination and dismantlement of the plant.

Modifications were designed and installed to develop a Spent Fuel Pool Island. These modifications provided physical isolation between decommissioning activities and the Spent Fuel Pool. Additionally, they provided systems, independent from normal plant systems, for cooling, ventilation, and independent power supplies for the Spent Fuel Pool and its associated equipment and structure. A backup diesel generator was also installed to maintain electrical capabilities during a loss of normal power supply.

Construction of the ISFSI was completed in the Spring of 2004. The first casks moved to the ISFSI were those containing the GTCC waste, followed by casks containing spent fuel. The spent fuel and GTCC waste movements are scheduled to be complete by mid-2005.

3.3 Completed and Ongoing Decommissioning Activities and Tasks 3.3.1 Overview The major accomplishments described in the following sections are included in the LTP because they are similar to, and indicative of, the complexity of future activities to be performed. The successful August 2004 3-3 Rev. 2 l

Haddam Neck Plant License Termination Plan completion of these activities demonstrates the project team's ability to safely and effectively decommission the HNP site. CYAPCO initiated decommissioning activities in 1997 with plant resources and subsequently decided that a Decommissioning Operations Contractor (DOC) would be used to complete the decommissioning, with CYAPCO personnel performing an oversight function. Bechtel began project DOC work in April, 1999, beginning with preparation and performance of Transition Period activities. A transition plan was developed to define infrastructure, programmatic, and procedural elements to be reviewed, modified as appropriate and approved for safe and effective transition to a DOC-performed project. During transition, the existing CYAPCO infrastructure was used as necessary to facilitate project performance. Transition was completed in October 1999.

In June of 2003 CYAPCO transitioned decommissioning activities from the decommissioning operations contractor to self-managing the remaining scope of activities at the HNP site. CYAPCO will supplement its staff with specialty contractors and staff augmentation as necessary to safely and efficiently complete the decommissioning. CYAPCO changed the general approach to decommissioning the site from decontaminate, FSS and demolish as clean materials where appropriate to decontaminate to permit demolition using appropriate controls, dispose of the majority of the SSCs at regulated disposal sites, and perform FSS on the remaining structures and backfill. Presently, the scope is to remove SSCs to 4 ft.

below grade level, perform a FSS or assessment, backfill and satisfactorily complete an FSS as required.

The only industrial area structures that may remain are the below-grade portions of the Containment Building, Spent Fuel Building, Screenwell House, Discharge Tunnels, and the B Switchgear Building.

Generally, these buildings will have the internal structures removed, and will be decontaminated as necessary. An FSS or assessment, as appropriate, will be performed on remaining foundations/basements and soil prior to backfill, following backfill, an FSS will be satisfactorily completed as required. The Screenwell House will be removed to 4 ft below the contour of the river bank and below the river bed.

3.3.2 RCS Chemical Decontamination One of the first major decommissioning activities at the HNP site was the chemical decontamination of the Reactor Coolant System, which was conducted as an ALARA initiative from July to August of 1998.

The flush included the steam generators and pressurizer, as well as portions of the appended systems of letdown and charging, residual heat removal, loop fill and drains, and selected dead-leg piping.

Approximately 131 Curies (gamma emitting radionuclides, including Am-241) were removed from the RCS during this process. Table 3-2 provides a breakdown of the contamination removed from the Reactor Coolant System and is based on either gamma spectroscopy and Part 50/61 analysis of samples taken during the HNP chemical decontamination, or a Part 50/61 analysis of a stainless steel artifact removed from the HNP letdown line in November 1997. The activity for all other radionuclides (including hard-to-detect radionuclides) was below the Lower Limits of Detection (LLD) required for a Part 50/61 analysis. Although Ag-I 10m (half life of 8.4 months) is listed below as being present, it should be noted that the chemical decontamination was performed in mid-1998 and that this radionuclide will have decayed by more then 5 half lives by the time the Final Status Survey will be conducted. As discussed in Section 2.3.3.4, Co-57 and Ag-l1 Om are not included in the list of radionuclides considered significant for the conduct of the Final Status Survey (LTP Table 2-1 1), as they are expected to be insignificant contributors to dose.

August 2004 Rev.2 1 3-4

Haddam Neck Plant License Termination Plan Table 3-2 Activity Removed During the HNP Reactor Coolant System Chemical Decontamination

_ Radionuclides Activity (Ci) Reference/Basis A. Gamma Co-60 128.7 Final Report "Reactor Coolant System Emitting Decontamination with the Siemens Radionuclides CORD D UV Process," dated 11/9/98.

Am-241 0.32 Mn-54 1.36 Ag-1 10m 0.24 Co-57 0.12 Liquid Sample #980727029 taken during the Chem Decon. Value is calculated by scaling from Co-60 activity in sample.

Subtotal 130.74 B. Alpha Pu-238 0.30 Liquid #x0723 (Part5O/61) taken Emitting during the Chem Decon. Value is Radionuclides calculated by scaling from Am-241 Pu-239 0.10 cc Cm-242 0.0016 4; Cm-243 0.12 ..

Am-241 See "A" above Subtotal 0.5216 (without Am-241)

C. Other Fe-55 42.97 Chem Decon Stainless Steel Artifact, Radionuclides sample #z10449 (Part 50/61) -

Including Analyzed May 15, 1998 Value is I

HTDs calculated by scaling from Co-60 activity in this sample.

Ni-63 9.50 ..

Sr-90 0.024 ..

Pu-241 5.01 ..

Subtotal 57.504 August 2004 3-5 Rev. 2 1

Haddam Neck Plant License Termination Plan 3.3.3 Turbine Rotors/Generator The Low Pressure Turbine Rotors were removed from the system and transported to the Palisades Nuclear Plant. These were the first large components removed from site. The Low and High Pressure Turbine Rotors have been removed from the site, as well as the main generator.

3.3.4 Removal of Spare Auxiliary Transformer/Power Transmission Line The spare auxiliary transformer has been removed from the HNP site. The transformer was stripped of the electrical components and cooling fans, for loading onto a multi-wheeled transporter and subsequent loading onto a barge for shipment on the Connecticut River. This activity proved to be a very valuable demonstration and test run for future activities involving onsite component handling and component removal by barge, such as those involving the steam generators, pressurizer, and the reactor pressure vessel. The main station transformer has also been removed from the site, via barge, in 2001.

In the spring of 2004 a site wide power outage was coordinated to remove portions of the 115 kV system that provided power to the site. The Spent Fuel Pool Island and associated critical equipment was powered using the on site diesel generator while one of the two 115 kV system feeder lines to the site wvas removed along with its associated line towers and switch equipment.

3.3.5 Removal of Steam Generator Steam Domes The next major component removal task was the removal of the four steam generator steam domes. As the steam generators could not be removed in one piece without structural modification of the Containment Building, the generators were cut in the transition area at the feedwater ring. The steam domes were lifted and placed onto the Charging Floor where a cover plate was welded onto the bottom of the steam domes. Plates were also welded onto the top of the steam generator lower assembly. The nozzles were cut and capped. The domes were then rigged off of the Charging Floor, through the Containment Equipment Hatch, and onto a multi-wheeled transporter.

After final radiological survey, the domes were trucked approximately 13 miles to a local rail-head in Portland, Connecticut. The domes were then transported by rail to a licensed radioactive material handler in Tennessee.

3.3.6 Removal of Steam Generator Lower Assemblies The Steam Generator Lower Assemblies (SGLAs) were removed with the original intent of being placed on barges and shipped for disposal. However, due to low water levels on the Savannah River, barge transportation was not immediately achievable, and a temporary laydown area was established, with appropriate shielding installed. In May and June of 2001, two shipments, each consisting of two SGLAs, were barged from the site to Port Royal, South Carolina, where they were loaded onto railcars for shipment to their final destination at Barnwell, S.C. (Reference 3-3).

3.3.7 Removal of the Pressurizer The pressurizer was also removed from the Containment Building. Submittals to the DOT and the Envirocare Facility in Utah were made. CYAPCO received a "Notice to Transport" from Envirocare in December of 2000 and received a DOT exemption, to allow the pressurizer to be its own shipping package, in June of 2001 (Reference 3-4). The pressurizer was trucked to a local rail-head in Portland, Connecticut, and then shipped via rail to the Envirocare Facility in Utah on August 3, 2001.

August 2004 3-6 Rev. 2 1

Haddam Neck Plant License Termination Plan 3.3.8 Removal of Reactor Coolant Pumps (RCPs)

The RCPs were removed from the Containment Building. Each of the four RCPs was trucked to a rail-site in New Jersey (the first shipment left the HNP site on December 13, 2000 and the final shipment left January 9, 2001). The set of four RCPs were then shipped via rail to the Envirocare Facility in Utah and arrived February 12, 2001.

3.3.9 Dismantlement of Buildings Several service, maintenance, and fabrication shops were demolished and removed from the site. Those buildings were demolished using standard construction equipment and techniques, which included contamination control techniques such as dust suppression (by wetting down the structures and materials during demolition) and by stringent access control measures for the work area. Confirmatory monitoring for airborne contamination was performed and demonstrated the effectiveness of this approach.

Demolition of the buildings occurred only after a radiological survey of the buildings was made. Those for which no radioactivity was detected were released from the site as clean material. Buildings for which radioactivity was detected were sent to an offsite licensed radioactive material handler. The removal of these buildings was to facilitate large component removal.

Security functions have been relocated, and the old Security and Fitness Center has been demolished and released, with no detectable activity, using the currently approved survey and release process. The Engineering and Instrument and Controls Buildings have also been demolished.

3.3.10 Removal and Disposal of the RPV The RPV was disconnected and isolated from the coolant loops and placed in the lower transport canister.

The RPV was filled with 30 weight grout. Seventy (70) weight grout was injected into the area between the lower transport canister and the RPV. The upper canister was then mated to the lower canister and welded together. Following the installation, inspection and acceptance of the girth weld, grout was injected into the area between the RPV and the canister assembly. The RPV, canister and grout (weighing approximately 720 tons) were then lifted and down ended to its transport position. The package was moved out of the containment where it was placed on a transporter and moved to its staging location for barging to the Barnwell, South Carolina Low Level Waste Disposal Facility. The RPV contains approximately 2 1,000 curies of radioactivity in its packaged configuration.

Concurrently with the RPV packaging, the discharge canal was dredged to allow easy movement of the RPV transport barge into the barge loading area, approximately I mile from the Connecticut River.

The RPV package was secured to the transport barge and was transported down the east coast of the United States to the DOE Savannah River Site where it was transported overland to its final burial location.

August 2004 3-7 Rev.2

Haddam Neck Plant License Termination Plan 3.3.11 Spent Fuel and GTCC Wastes Spent fuel and GTCC wastes movements from the Spent Fuel Building began in the Spring of 2004. The first casks moved to the ISFSI were those containing GTCC wastes, followed by casks containing spent fuel. The spent fuel and GTCC waste movements are scheduled to be complete by mid-2005.

3.3.12 Additional Activities Additional activities that have been completed or are ongoing include, but are not limited to the following:

• Continued assessment of the functional requirements for plant systems, structures, and components.

Plant systems, structures, and components needed to support safe storage of the spent fuel, support spent fuel pool cooling, and facilitate ongoing plant activities have been identified.

• Isolation/de-energization and removal of plant systems, structures, and components.

A comprehensive plant isolation/de-energization program was developed and implemented as described in Section 3.4.1.6. Systems, structures, and components not required to support decommissioning or spent fuel storage continue to be decontaminated and/or removed in accordance with the license and approved plant procedures. Modifications were performed to provide for temporary systems and for construction power to support decommissioning activities, as needed.

• Reactor Vessel Internals Segmentation Because of Chem Nuclear's limit on a package's radioactive content (50,000 Curies),

segmentation, removal, and onsite storage of the Greater Than Class C (GTCC) internal components was performed. This reduced the RPV inventory to less than 50,000 Curies for shipment and disposal. The GTCC material is being stored on site until an approved facility is available for disposal of the GTCC waste.

• Continuing Removal of Wastes and Demolition Materials Plant activities working concurrently are radiological and non-radiological waste removal. These include removal of contaminated soils, asbestos abatement program, and removal of other hazardous and non-hazardous wastes.

3.4 Future Decommissioning Activities and Tasks 3.4.1 Overview Demolition and bulk disposal activities will continue generally independent of those activities related to the operation of the Spent Fuel Pool Island and ISFSI. Following any planned demolition (generally to 4 ft below grade) and/or decontamination of contaminated SSCs, an assessment survey will be performed, the area backfilled, and a comprehensive final radiation survey (Final Status Survey) will be conducted.

August 2004 3-8 Rev. 2 l

Haddam Neck Plant License Termination Plan This survey will confirm that the area meets the release criteria. Isolation and control measures will be established to prevent the area(s) from becoming re-contaminated. The final status survey results will be compiled in a series of reports (release records) by area(s) and will be made available to the NRC for inspection.

Detailed schedules for significant decommissioning activities (including Final Status Survey activities and building demolition) are routinely communicated to NRC Region I personnel (for example, through periodic decommissioning status meetings) to allow for NRC observation and inspection of these activities. Communication of significant decommissioning activities will be based on activities listed in detailed project schedules (or their equivalents).

The remaining dismantlement and decontamination activities can be grouped into several classifications, the implementation of which may overlap. The first phase included major component removal. The current phases includes contaminated system removal, clean system removal, and decontamination of site buildings to allow for controlled demolition and removal to a radioactive or clean waste facility, as appropriate. The phases may be implemented on an area-by-area basis. Under this approach, often only a part of a system will be removed, while the remaining portions await removal in a subsequent phase. The remaining contaminated systems and components will be decontaminated or removed, packaged, and either shipped to an offsite processing facility, shipped directly to a low-level radioactive waste disposal facility, or handled by alternate methods in accordance with applicable regulations.

Decontamination of plant structures, to allow for controlled demolition, may be completed concurrently with equipment removal. Decontamination of structures may include a variety of techniques ranging from water washing to surface material removal (Chapter 4 also discusses remediation methods). Contaminated structural material may be packaged and either shipped to a processing facility, or shipped directly to a low level radioactive waste disposal facility. Alternative disposal methods, in accordance with applicable regulations, may also be used.

The following sections provide a general description of the remaining decommissioning activities for the HNP site. These activities involve the reduction of radioactivity to As Low As Reasonably Achievable (ALARA) levels, not to exceed 25 mrem/yr TEDE, allowing for release of the site for unrestricted use.

This information provides the basis for development of programs and procedures for ensuring safe decommissioning and a basis for detailed planning and preparation of decontamination and dismantlement activities.

After demolition, a comprehensive final radiation survey will be completed as described in Chapter 5 on remaining building foundations/basements and soil. This survey will verify that residual radioactivity has been reduced to sufficiently low levels, as stipulated in IOCFR20.1402 (Reference 3-7), to allow the release of the site for unrestricted use. Upon completion of the final survey for a plant area, CYAPCO will document the results of that survey and will make them available for NRC inspection.

3.4.1.1 Detailed Planning and Engineering Activities Detailed project plans will continue to be developed in accordance with design control procedures to support the decontamination and dismantlement activities. These plans are used to develop work packages, to assess safety and waste issues for each major task, to support ALARA reviews, to aid in estimating labor and resource requirements, and to track decommissioning costs and schedule. CYAPCO has not made a final determination of the specific set of remediation technologies for the remaining systems, structures and components, since the variety of situations and conditions encountered do not readily allow a one-size-fits-all approach. This flexibility allows for use of emerging technologies over August 2004 3-9 Rev. 2

Haddam Neck Plant License Termination Plan the lifetime of the project, as they arise. A cost/benefit analysis (including ALARA assessment and waste management assessment) would be a component of making such a decision.

Work packages are used to implement the detailed plans and provide instructions for actual field implementation. The work packages address discrete units of work and include appropriate hold and inspection points. Administrative procedures control work package format and content, as well as the review and approval process.

3.4.1.2 General Decontamination and Dismantlement Considerations As has been the current practice and in accordance with the HNP PSDAR, the following general decontamination and dismantlement considerations, as applicable, will continue to be incorporated into decommissioning work packages during the decontamination and dismantlement period. Section 3.3.9 describes the process that has been used for dismantling buildings. Demolition techniques will vary for each building and will depend on building construction and contamination control methods. Surveys will verify that: (1) residual radioactivity has been reduced to sufficiently low levels to permit controlled demolition, (2) residual radioactivity has been reduced to sufficiently low levels, as stipulated in 10CFR20.1402; and buildings, and associated areas, may be released for unrestricted use upon their removal from the Part 50 license; (3) residual radioactivity levels are appropriate for shipment to an offsite licensed radioactive materials handler for processing and/or disposal; or (4) there is no detectable radioactivity, and the buildings may be free released. Specific considerations are described herein.

Dismantlement activities are reviewed to ensure that they do not impact the safe storage of spent fuel in the Spent Fuel Pool. Configuration control packages are implemented in accordance with administrative controls and require evaluations in accordance with the requirements of IOCFR50.59.

Temporary shielding and other standard Health Physics/Radiation Protection practices will be implemented to address ALARA considerations during decommissioning activities. These are described in further detail in subsequent paragraphs. Some dismantlement activities may be performed under water for shielding purposes, as well as for contamination control purposes.

As currently practiced at HNP and in accordance with the HNP PSDAR, the capability to isolate or to mitigate the consequences of a radioactive release will continue to be maintained during decontamination and dismantlement activities. Isolation is the closure or control of penetrations and openings to restrict transport of radioactivity to the environment. However, this consideration does not preclude the removal of penetrations and attachments to Containment, provided that effluents will be controlled or monitored.

Airborne radioactivity will be controlled considering the following:

• Operation of the appropriate portions of the containment ventilation and purge system, or an approved alternate system, during decontamination and dismantlement activities in the Containment Building;

• Operation of the appropriate portions of the Primary Auxiliary Building, Service Building, and Fuel Building ventilation systems, or an approved alternate system, during decontamination and dismantlement activities in these buildings;

• Use of local High-Efficiency Particulate Air (HEPA) Filtration Systems for activities expected to generate airborne radioactive particulates (e.g., grinding, chemical decontamination, or thermal cutting of contaminated components).

August 2004 3-10 Rev. 2

Haddam Neck Plant License Termination Plan Work activities are planned to minimize the spread of contamination. Contaminated liquids are contained within existing or supplemental barriers and may be processed by a liquid waste processing system prior to release, if necessary. To minimize the potential for spread of contamination, the following considerations will continue to be evaluated for incorporation into the planning of decommissioning work activities:

• Covering of openings in internally contaminated components to confine internal contamination;

• Use of contamination control barriers as appropriate around activities that may result in airborne contamination during cutting and removal processes;

• Decontamination and dismantlement of contaminated systems, structures, and components by decontamination in place, removal and decontamination, or removal and disposal;

• Removal of contaminated supports in conjunction with equipment removal or decontamination of supports in conjunction with building decontamination;

• Removal of contaminated systems and components from areas and buildings prior to structural decontamination (block shield walls, or portions of other walls, ceilings, or floors may be removed to permit removal of systems and components.);

• Removal or decontamination of embedded contaminated piping, conduit, ducts, plates, channels, anchors, sumps, and sleeves during area and building structural decontamination activities;

• Use of local or centralized processing and cutting stations to facilitate packaging of components removed in large pieces; and

• Removal of small or compact plant components and parts intact, where feasible. (This includes most valves, smaller pumps, some small tanks, and heat exchangers. These components could then be decontaminated in whole or part, and disassembled or segmented in preparation for disposal or release.)

3.4.1.3 Decontamination Methods Contaminated systems and components are typically removed and sent to an offsite processing facility, sent to a low-level radioactive waste disposal facility, or decontaminated onsite and released. Other decontamination methods typically include wiping, washing, vacuuming, scabbling, spalling, and abrasive blasting. Selection of the preferred method is based on the specific situation. Other decontamination technologies may be considered and used as appropriate.

Hand wiping may be used to remove loose surface contamination. Airborne contamination control and waste processing systems are used as necessary to control and monitor releases. If structural surfaces are washed to reduce contamination, controls are implemented in accordance with approved procedures to ensure that wastewater is collected for processing by liquid waste processing systems.

Tanks and vessels are evaluated and, if ALARA, are flushed or cleaned to reduce contamination and to remove sediments prior to sectioning and/or removal. In cases where tanks, vessels, and piping are major contributors to dose rates, the ALARA principle will be applied to determine removal sequence.

Precautions are taken to ensure that liquid inadvertently discharged from the tank is captured for processing by a liquid waste processing system. Sediment removed from the tank may be stabilized prior August 2004 3-1 1 Rev. 2

Haddam Neck Plant License Termination Plan to shipment or may be shipped without stabilization in an approved condition. Wastewater is processed and/or sampled and analyzed in accordance with the National Pollutant Discharge Elimination System (NPDES) permit and Offsite Dose Calculation Manual (ODCM) before being discharged.

Concrete that has surface or near-surface contamination may be cleaned, if necessary, to meet limits to permit controlled demolition or FSS, as applicable. Activated concrete may be removed as necessary to meet DCGLs (and ALARA considerations) and sent to a Low-Level Waste (LLW) Disposal Facility, or handled by other methods in accordance with applicable regulations. The removal of concrete may be performed using methods that control the removal depth to minimize the waste volume produced.

Vacuum removal of the dust and debris with HEPA filtration of the effluent may be used to minimize the spread of contamination and reliance on respiratory protection measures. Chapter 4 also discusses remediation methods.

3.4.1.4 Contaminated System Dismantlement Dismantlement methods for contaminated systems can be divided into two basic types: disassembly and cutting. Disassembly generally means removing fasteners and components in an orderly non-destructive manner (i.e., the reverse of the original assembly). Cutting methods include flame cutting, abrasive cutting, and cold cutting. Abrasive water-jet cutting was used to segment the Reactor Pressure Vessel internals.

Flame cutting includes the use of oxyacetylene and other gas torches, carbon arc torches, air or oxy arc torches, plasma arc torches, cutting electrodes, or combinations of these. Most of the torches can either be handheld or operated remotely. Abrasive cutting includes the use of grinders, abrasive saw blades, most wire saws, water lasers, grit blast, and other techniques that wear away metal. Cold cutting includes the use of bandsaws, bladesaws, mechanical disintegration methods, drilling, machining, shears, and bolt/pipe/tubing cutters. Selection of the preferred method depends on the specific situation. Other dismantlement technologies may be considered and used if appropriate.

Dismantling of systems includes the removal of valves and piping for disposal. Most valves can be removed with the piping. Larger valves and valves with actuators may be removed separately for handling purposes. Systems and components will be surveyed to ensure that levels are below the criteria to allow the system or components to remain and be demolished with the structure in a controlled environment.

3.4.1.5 Removal Sequence and Material Handling Removal sequences may be dictated by access and material handling requirements or by personnel exposure considerations. In some cases, a top-down approach may be used. Using this approach, materials and structures at the highest elevations are removed first to allow access to components in lower levels. In other cases, different approaches may prove more efficient.

In some cases, the first items removed are those that are not contaminated, or are only slightly contaminated, to preclude contamination by other equipment removal activities. However, personnel exposure considerations may not always allow this option. The ALARA principle will be applied by removal of hot spot items (e.g., piping with high dose rates) prior to other work. Where non-contaminated equipment or piping is not removed first, covers or other protection methods may be used to prevent cross-contamination.

August 2004 3-12 Rev. 2

Haddam Neck Plant License Termination Plan Where rapid cutting techniques are available, pipes and equipment can be sectioned into pieces that are manageable using light rigging or by manual lifting. Where slow cutting techniques are used, the largest manageable pieces will typically be freed and further reduced in size at a more convenient location.

The plant is equipped with multiple cranes, hoists, and lifting and transport systems. These systems can be used to lift and transport components and equipment to support plant decommissioning activities.

Forklifts, mobile cranes, front-end loaders, and other lifting and transport devices can also be used for plant decommissioning activities. The major installed plant cranes, hoists, and lifting and transport devices that may be available to support decommissioning include the Containment Building Polar Crane and Yard Crane.

Inspection requirements for these cranes meet the specific requirements of plant procedures. Other rigging equipment will be inspected and verified in accordance with procedures. Smaller rigging equipment (chain hoists, etc.) will be inspected and verified to be in good working condition prior to use.

The Containment Building Polar Crane is capable of reaching most locations inside the Containment Building and handling large, heavy loads. This crane has already been used for the Steam Generator Domes, Steam Generator Lower Assemblies and other smaller loads. The Yard Crane has been used for material handling.

Installed cranes and hoists may be used in conjunction with temporary or mobile lifting and transport devices to support decommissioning. The installed plant cranes, hoists, and other lifting devices may be dismantled when they no longer are required to support decommissioning activities.

3.4.1.6 System Isolation/De-encrgization Systems or components will continue to be deactivated prior to decontamination and dismantlement. In general, isolation/de-energization is implemented by mechanical isolation of interfaces with operating plant systems, draining piping/components, and de-energizing electrical supplies. Combustible materials (e.g., charcoal from filters, lube oil) are removed from the deactivated components, where practical.

Chemicals used in, or resulting from, decommissioning activities are controlled in accordance with the plant chemical safety program. Plant critical drawings are updated to indicate deactivated portions of systems, and plant procedures are modified accordingly to reflect the changes.

Isolation/de-energization of plant systems is administratively controlled by approved procedures.

Isolation/de-energization plans are established to implement the desired system valve lineup changes and electrical isolations. The design change process is used to remove components, lift electrical leads, install electrical jumpers, cut and cap piping systems, or install blank flanges.

Plant procedures also provide controls over the operation of deactivated system boundary valves. As additional systems are deactivated, existing isolation boundaries are re-evaluated and changed, as necessary, to reflect the new plant condition. Boundary valves are tagged for identification.

3.4.1.7 Temporary Systems Required to Support Decommissioning Decontamination and dismantlement of systems, structures, and components often require the removal of interferences. Removal of some of these interferences may eliminate power, service air, and other services needed to support decommissioning. Also, use of installed plant systems for decommissioning support may become impractical, due to the risk of encountering energized systems or circuits.

Temporary services and systems are being provided to support decommissioning activities. Temporary modifications to plant structures, systems, and components are controlled by design control procedures.

August 2004 3-13 Rev. 2

Haddam Neck Plant License Termination Plan Portable load centers are powered from motor control centers, plant load centers, or the yard loop. These portable load centers can supply cutting and hoisting equipment, temporary lighting, or other power needs. Service air can be provided by portable air compressors using hoses or temporary air manifolds.

Demineralized water is available from portable demineralizer skids or portable tankers brought onsite.

Portable hydraulic power centers can be used to power hydraulic equipment.

Temporary liquid and solid waste processing systems are being used during decommissioning for processing plant waste. These systems include filters and/or demineralizers, and may be used at one or more locations in the waste-processing path.

Portable radiation monitors and air monitoring equipment provide local radiation monitoring. Localized temporary ventilation equipment and HEPA filtration is being used to supplement building ventilation and minimize the spread of radioactive particulate contamination.

3.4.1.8 Specific Decommissioning and Dismantlement Activities 3.4.1.8.1 Reactor Pressure Vessel Internal Segmentation As stated earlier, segmentation of the internals was required so that the RPV package could be disposed of at Barnwell. The RPV internals were segmented using abrasive water-jet technology and Mechanical Disintegration Methods (MDM). The GTCC was removed and most of the GTCC waste was transferred to the ISFSI. The remaining GTCC waste is scheduled to transfer to the ISFSI by mid-2005. The remaining vessel internals were loaded into the reactor vessel. The RPV was prepared in accordance with a DOT-approved packaging plan and transported to a disposal facility.

3.4.1.8.2 Reactor Coolant Piping/Pressurizer The Reactor Coolant Pump (RCP) motors were separated from the pumps at the yoke and rigged out of the containment using the Polar Crane. The motors were packaged for transportation and disposal. The primary coolant loop piping at the pump nozzles was cut. The pumps have been removed from their foundation, properly packaged, and shipped to Envirocare for disposal.

Reactor coolant pipes were cut at the pressurizer nozzles and covered. Due to contamination of the pressurizer, a separate DOT exemption request was developed and submitted. The pressurizer was rigged out using the Polar Crane and prepared for transportation and ultimate disposal. The pressurizer was also shipped to Envirocare for disposal.

The remaining Reactor Coolant piping spools, after being cut away from the major components, were segmented for ease of removal. This piping was rigged out of the containment using the Polar Crane and packaged for transportation and ultimate disposal.

3.4.1.8.3 Major Component Removal Major Component Removal is being conducted with safety, ALARA, transportation, and disposal considerations. The approach to work for each major component was evaluated from the point of removal through ultimate disposal. Many technologies and processes were evaluated considering safety, dose, and cost.

August 2004 3-14 Rev. 2

Haddam Neck Plant License Termination Plan After the major components from the steam generator cubicles were removed, removal of ancillary systems, structures, and components commenced. The reactor internals have been segmented, and GTCC portions have been transferred to the SFP for storage (subsequently, the GTCC waste has been transferred to the ISFSI). Clean-up of the refueling cavity was performed in preparation for RPV removal. The RPV was placed into a container, grouted and sealed. The RPV was then moved to the equipment hatch area for final transportation preparation. The Containment will be cleared of components and bulk commodities (conduit, small bore piping, pipe hangers, etc.), building decontamination and preparation for structure demolition will begin.

The neutron shield tank will be segmented into manageable pieces and packaged for transportation and ultimate disposal.

The Low and High Pressure Turbine internals were previously removed from the plant site. The associated turbine casings have been segmented to facilitate removal. Segments were rigged from the turbine pedestal using the Turbine Building Crane to the rail bay and packaged for transportation and disposal. The generator set (consisting of the rotor and stator) was disassembled and removed. The exciter module was also disassembled and removed.

The turbine casings, steam chest, main condensers, and associated piping and valves have been removed from the Turbine Building. The Turbine Building will be surveyed and released for controlled demolition. The building will be demolished to 4 ft below grade, and an assessment survey will be performed, the area will be backfilled, a FSS survey performed and controls instituted to prevent re-contamination of the area. The waste materials, associated with the demolition, will be shipped to an appropriate waste facility.

3.4.1.9 Decontamination and Disposition of Site Buildings CYAPCO has modified its approach to the decommissioning. System, Structures and Components (SSCs) will be surveyed to ensure contamination levels are below levels acceptable for controlled demolition.

This minimizes the removal of contaminated components and the need for extensive decontamination prior to demolition. The wastes generated will be disposed of as low level radioactive wastes at regulated disposal facilities. The waste materials will not be used on site as backfill. Engineering controls will be used to control the demolition. Examples of engineering controls are decontamination and resurvey, applying a fixative to the elevated contamination areas, scabbling, scarifying, or tenting for partial removal of elevated areas of contamination.

The current plan is for SSCs that survey clean or can be cost effectively decontaminated and surveyed for unrestricted release will be demolished. These are typically secondary side SSCs. The wastes generated will be disposed of at regulated clean waste facilities.

Generally, SSCs will be demolished to 4 ft below grade elevation. The areas will be controlled to preclude contamination. An FSS survey of any remaining foundations/basements and the soils will be performed. An opportunity to perform verification will be provided to regulators. Then the foundations/basements will be backfilled to grade elevation, as needed. Surveillances will be performed on areas, for which FSS activities have been completed, to verify the as left radiological configuration 3.4.2 General Description of and Remediation Considerations for Remaining Systems, Structures, and Components as of May 2004 This section presents a summary description of the remaining HNP systems, components, and structures that are known to be or are considered to be internally contaminated or that may be used to support August 2004 3-15 Rev. 2 l

Haddam Neck Plant License Termination Plan decommissioning activities. This discussion includes general activities and remediation considerations associated with decommissioning these systems, structures, and components.

Because external contamination is generally considered to exist on systems, structures and components located in the RCA, it is not specifically discussed in the following system discussions. Systems, components, and structures that are externally contaminated will be decontaminated, as necessary, and released or disposed of as radioactive waste.

3.4.2.1 Chemical and Volume Control System (CVCS)

The CVCS no longer supports purification of the RCS, and resin is being removed from plant ion exchangers. The CVCS has been isolated from the flooded refueling cavity inside the Containment.

Remaining isolated portions of the CVCS located in containment and the Primary Auxiliary Building, including letdown piping, regenerative and non-regenerative heat exchangers, charging pumps, metering pump, RCP seal water heat exchanger, chemical addition tank, boric acid mixing subsystem, volume control tank, and associated piping have been or will be drained and will be prepared for dismantlement and removal. The CVCS is internally contaminated. Portions of the system may be demolished in place along with their associated structure and disposed of at a licensed disposal facility.

3.4.2.2 Component Cooling Water (CCXN') System Operation of the CCW system is no longer required. The system has been drained and prepared for dismantlement and removal. Portions of the CCWV system are internally contaminated. Portions of the system may be demolished in place along with their associated structure and disposed of at a licensed disposal facility.

3.4.2.3 Service Water (SWN) System Alternate flow paths exist, as allowed by the National Pollutant discharge Elimination System (NPDES) permit, for liquid discharges to the environment. Thus, the service water system components can be removed and/or demolished and disposed of at a licensed disposal facility.

3.4.2.4 Spent Fuel Pool and Fuel Handling Equipment The spent fuel storage and handling SSCs are located in the Fuel Building. These SSCs consist of: the Spent Fuel Pool, the spent fuel storage racks, the cask loading pit, and the area housing components of the modular systems for cooling and purification of the Spent Fuel Pool. The Spent Fuel Pool provides for storage of irradiated fuel until it is transferred to an alternative, licensed storage facility. The Spent Fuel Pool also provides storage of GTCC radioactive waste transferred from the Containment Building, until that waste is transferred to an alternative, licensed storage facility.

The GTCC waste was transferred from the Containment Building to the Spent Fuel Pool via the transfer canal in the Containment Building and the transfer tube between the Containment Building and the Fuel Building. Following the transfer of the GTCC waste from Containment to the Spent Fuel Pool, the Spent Fuel Pool was isolated from the transfer tube via a blind flange that was installed in place of the existing sluice gate. The Spent Fuel Building was modified to support the parallel use of two transfer casks for moving irradiated fuel and GTCC wastes. This modification reduces the amount of time required to move all of the spent fuel and GTCC wastes out of the SFP to the ISFSI for long-term storage.

August 2004 3-16 Rev. 2

Haddam Neck Plant License Termination Plan Spent fuel and GTCC waste transfers from the Spent Fuel Pool to the ISFSI have begun. Once all spent fuel and GTCC waste have been transferred to the ISFSI, the potential still exists that the remaining storage and handling components have high levels of contamination. The spent fuel storage racks may be accessed using the Fuel Building crane, and it is anticipated that these racks can be removed intact for sectioning and packaging at a different location. The liner for the spent fuel pool and cask loading pit may be sectioned for removal. To facilitate removal, the fuel handling cranes, the fuel transfer cart, associated tracks, and upender frames may be sectioned into pieces. The spent fuel pool bridge crane may also be used to support the decommissioning of the Spent Fuel Pool, following transfer of spent fuel and GTCC waste.

3.4.2.5 Spent Fuel Pool Purification Spent fuel pool purification is performed using an independent demineralizer system located in the Fuel Building. This system takes a slipstream off the pool via the Spent Fuel Pool Skimmer system. This system is an operable system and will be maintained until the spent fuel has been removed from the pool and transferred to dry cask storage.

3.4.2.6 Spent Fuel Pool Transfer Tube The Spent Fuel Pool transfer tube connects the pool with the fuel transfer canal within the reactor refueling cavity. The transfer tube was maintained functional to allow for the transporting of the RPV GTCC materials from the cavity to the pool. Upon completion of the RPV internal segmentation, the transfer tube was sealed from the Spent Fuel Pool. Sealing off the transfer tube allows for the removal of the fuel transfer canal inside Containment.

3.4.2.7 Makeup Water (MW\')

The Fuel Building Makeup Water (MW) system provides the required makeup water needed for the operation of the Spent Fuel Pool (SFP), Intermediate Cooling (IC) system, and the Spray Cooling system.

These systems are required to support the heat removal from the SFP. The MW is stored in existing tanks (formerly the PWST and RPWST) that are now called "A & B" MWSTs. Piping and components supporting this system will be maintained until the spent fuel is removed from the pool and transferred to the ISFSI. This system is internally contaminated.

3.4.2.8 Main Steam and Feedivater (MS & FN\') Systems The Main Steam System is no longer required. Portions of the system have been removed to support steam generator removal and turbine rotor removal. The Feedwater System has been drained and is being dismantled and removed. Components include feedwater heaters, Condenser, Turbine Generator, associated piping and components. The feedwater pumps located in the Turbine Building have been removed and disposed. Portions of the MS and FW Systems may be internally contaminated.

3.4.2.9 Reactor Coolant System (RCS)

The cavity, RPV, and associated RCS loop piping are isolated from the rest of the system via closure of RCS loop stop isolation valves and sealing of other cavity-related piping.

Cavity isolation has allowed for the removal of the steam generators, Reactor Coolant Pumps, Pressurizer, and connecting piping.

Completion of the RPV internal segmentation allowed for the cavity to be drained, the reactor vessel to be packaged, and remaining piping and valves to be removed. A portion of the concrete cavity structure was August 2004 3-17 Rev. 2

Haddam Neck Plant License Termination Plan cut and removed to allow for the removal of the reactor vessel package through the containment equipment hatchway. The RCS is internally contaminated and was filled with grout to the vessel isolation valves, during grouting of the vessel in preparation for disposal.

The remaining system and components in the Containment will be removed to support internal and external demolition of the structure.

3.4.2.10 Residual Heat Removal (RHR) System The RHR System is no longer required to perform its intended function. The system has been isolated from the refueling cavity inside the Containment.

Remaining portions of the system have been isolated from operable systems, drained and are prepared for dismantlement and removal. Components include RHR pumps, heat exchangers, containment spray, containment charcoal filters, and associated piping. The RHR System is internally contaminated.

3.4.2.11 Safety Injection (SI) System The Safety Injection System is no longer required to perform an operating function. The system has been isolated from the reactor cavity.

The high-pressure Safety Injection System was used for a flow path to flush portions of the system containing spent resin deposited during chemical decontamination activities. With the completion of resin cleanup, the SI system upstream of the isolated cavity will be drained, dismantled, and removed. System components include high- and low-pressure SI pumps, RWST, and associated piping. The Safety Injection System is internally contaminated.

3.4.2.12 Service Air System The plant Service Air System is no longer required. The compressors have ceased operation and will be removed. Construction air required for decommissioning activities will be provided via local industrial compressors. Portions of the Service Air System are internally contaminated.

3.4.2.13 Control Air System Based on the significantly decreased control air requirements, the plant control air compressors have been permanently shut down. Temporary compressors are tied into portions of the old plant control air system to make smaller closed loop systems. These loops supply tank level indicators, flow control and air operated valves for various systems (Well Water, Waste Processing, Tank Farm, Service Water, and Ventilation). Temporary compressors may be removed, as systems become abandoned and they are no longer required. The retired Control Air System and associated compressors will be dismantled and removed. Portions of the Control Air System may be internally contaminated.

3.4.2.14 Primary WX'ater (PNN) System Demineralized water required for decommissioning activities will be provided from the existing Condensate Storage Tank (CST). The Demineralized Water Storage Tank (DWST) is being used as a holding tank for water associated with drilling and developing monitoring wells. The water collected in the DWST will be processed and discharged through the monitored discharge pathway.

Aiiugt 2004 3-1 R As_ . . 2 Rev.

Haddam Neck Plant License Termination Plan Current decommissioning activities include emptying resin from the ion exchangers of the PW system.

Following the completion of this activity, the remaining water in the PW system will be drained; and the PW system may be dismantled and removed. Portions of the PW System are internally contaminated.

3.4.2.15 Primary Ventilation System/Fuel Building Ventilation System The Primary Ventilation System is being replaced by portable ventilation systems-one for the Primary Auxiliary Building and one for the Containment Building. With the permanent absence of nuclear fuel from the Containment Building, monitoring and filtration for fission by-product radioactive gasses in the primary ventilation exhaust are no longer necessary. However, the alternate Containment Ventilation pathways are monitored for radioactive particulate releases.

The Fuel Building (FB) ventilation system shares no components with the alternate ventilation system.

The Fuel Building ventilation system includes filters and monitors for radioactive particulates and any release of radioactive gasses from the stored nuclear fuel. This system will be operated while fuel is stored in the spent fuel pool; and it will be operated to support activities for the decommissioning of the Spent Fuel Pool and the Fuel Building. The Fuel Building Ventilation System includes equipment and ducting that are internally contaminated.

3.4.2.16 Liquid Waste System Currently, liquid radioactive wastes from decommissioning activities and Spent Fuel Pool Island operation are processed by a combination of a temporary, mobile water treatment demineralizer skid and selected components of the pre-existing liquid radioactive waste processing system (piping and valves).

The use of a temporary water-processing skid results in the release of other liquid waste system components for isolation and removal.

During future decommissioning activities, including operation of the Spent Fuel Pool, alterations in the configuration of liquid waste processing components may be implemented. Following the processing of liquid wastes, controlled effluent releases are performed in compliance with existing NPDES permit requirements and applicable requirements of IOCFR20. These effluent discharges are monitored in accordance with approved procedures.

During future decommissioning of the Spent Fuel Pool and the Fuel Building, an appropriate, temporary waste water treatment system may be used for processing of spent fuel pool water and other decommissioning process waste water. Appropriate monitoring systems will be in place to ensure compliance with applicable NPDES permit requirements and applicable requirements of I OCFR20.

Existing and temporary liquid waste systems are internally contaminated and will be decontaminated and removed or disposed of at a licensed disposal facility.

3.4.2.17 Gaseous Waste System The Gaseous Waste System is no longer required to perform its original design function. The Gaseous Waste System is internally contaminated and will be removed and disposed of at a licensed disposal facility.

3.4.2.18 Turbine Building Waste Water Treatment System The Turbine Building Wastewater Treatment System remained in operation until the completion of the Turbine Building dismantlement. Portions of the system were contaminated. The contaminated portions were removed and disposed of at a licensed disposal facility.

August 2004 3-19 Rev. 2

Haddam Neck Plant License Termination Plan 3.4.2.19 Well Water and Water Treatment System The Well Water and Water Treatment System supply water for both plant and personnel activities other than personal consumption. The system uses only one or both of wells "A" and well "B." Support systems consist of construction power and local compressed air.

Prior to the dismantlement of the Turbine Building, the system was modified to continue to provide water throughout dismantlement and removal of the Turbine Building. The well water system is not internally contaminated.

3.4.2.20 Circulating Water and Vacuum Priming Systems Water discharges are being performed; in conformance with existing NPDES permit requirements and REMODCM criteria. Alternate discharge paths exist, which do not require the operation of the Circulating Water System. The Circulating Water System has been shutdown to allow for dismantlement of the Turbine Building. The Circulating Water System is not contaminated.

3.4.2.21 Closed Cooling System The Closed Cooling Water system is no longer required. The system will be dismantled and removed.

Portions of the system are internally contaminated. The contaminated portions of the systems will be removed and disposed of at a licensed facility.

3.4.2.22 Turbine Lube Oil System The Turbine Oil Tank Room was used for the storage and transport of waste oil. The Turbine Lube Oil System was not internally contaminated. The oil has been removed from the system, and the system has been removed from the site.

3.4.2.23 Boron Recovery System The Boron Recovery System has not been used since 1991. The system has been drained, and boron recovery evaporators, coolers, reboilers, and associated pumps and interconnecting piping have been isolated, dismantled, and removed. The Boron Recovery System is internally contaminated. The Borated Waste Storage Tanks (BWST) and the Recycle Test Tanks (RTT) were used as part of water storage and processing. The BWSTs and RTTs have been dewatered and desludged, demolished, and removed for disposal at a licensed disposal facility.

3.4.2.24 Containment Systems and Miscellaneous Systems The reactor core sumps and the in-core instrumentation sumps collect miscellaneous leakage, drainage, or unplanned releases in the Containment. These sumps, associated piping and equipment will remain in use to collect and transfer this leakage to the waste water process system until demolition of the interior of Containment is complete. These systems are internally contaminated and will be decontaminated to required levels and portions of these systems left in place.

3.4.2.25 Site Electrical Distribution The existing plant electrical distribution system has been modified to institute the site decommissioning power distribution system. This new system provides the Spent Fuel Pool Island power and power required to support plant decommissioning and demolition activities from 115 kV system through Station August 2004 3-20 Rev. 2

Haddam Neck Plant License Termination Plan Service Transformner 12R-22S (399) via 4160V Bus 12. This allows the existing plant power distribution system and associated equipment, not re-aligned to the new decommissioning power distribution system, to be de-energized and removed from service.

Equipment and cables used for the site decommissioning power are identified with unique indicators to help ensure worker safety.

3.4.2.26 Fire Protection System The Fire Protection System provides manual and automatic fire suppression and automatic detection capabilities for plant areas. The Fire Protection System includes the following: portable fire extinguishers, water supply and distribution systems, fire suppression system, emergency lighting, and the fire detection and alarm system.

The Fire Water Distribution System is maintained operable by the use of the electric and diesel fire pumps located in the Screenwell Building, as well as the associated test headers, underground distribution system, fire hydrants, sprinklers and hose stations.

The fire protection supply to the Turbine Building, Maintenance Hot and Clean Shop, and the Boiler Room have been isolated in order to support removal activities in these areas. Other portions of the system will be de-activated to support future building decommissioning and removal activities. Such changes may require updates to the Technical Requirements Manual (TRM). The Fire Protection Systems are not internally contaminated.

3.4.2.27 Heating Steam and Condensate System The Heating Steam and Condensate System have been drained, dismantled and removed. Future heating and freeze protection of operable systems and buildings may be through the use of local heaters and/or independent heating systems. Portions of this system are internally contaminated. The contaminated portions of the system will be removed and disposed of at a licensed facility.

3.4.2.28 Floor, Roof and Equipment Drains The Floor, Roof, and Equipment Drain System will be maintained in operation. These drains are required to provide building drainage throughout the decommissioning effort and will be one of the last systems in each building to be removed. Portions of these drain systems are internally contaminated. The contaminated portions of the system will be removed and disposed of at a licensed disposal facility.

3.4.2.29 Buildings 3.4.2.29.1 Fuel Building The Fuel Building consists of a concrete structure with a superstructure on steel framing above the new fuel storage area and the spent fuel pool area. The Spent Fuel Pool is constructed of reinforced 6-foot thick concrete walls lined with a l4-inch thick stainless steel liner and is 35 feet deep. The roof is an 8-inch thick reinforced concrete slab. The building also contains the spent fuel pool independent demineralizer cleanup system and the spent fuel pool independent cooling system and associated equipment.

The Spent Fuel Pool will remain in use until the spent fuel has been transferred to the onsite ISFSI. Upon completion of this effort, the Fuel Building will be decontaminated and the associated systems and August 2004 3-21 Rev. 2

Haddam Neck Plant License Termination Plan components demolished to 4 ft below grade. Contaminated concrete will be scabbled to remove surface contaminated areas. Portions of the Spent Fuel Cooling System are internally contaminated.

The Spent Fuel Building was modified to permit the parallel use of two fuel transfer casks to reduce the amount of time it takes to move the irradiated fuel and GTCC waste to the ISFSI. Transfer operations began in the Spring of 2004 and are expected be completed by mid-2005.

3.4.2.29.2 Containment Building The Containment Building consists of two structures on a common foundation. One is the containment itself, and the other is the internal biological shield structure. Supports for equipment, operating decks, access stairways, and platforms are included in the containment internals.

The inside of the containment concrete shell is steel lined. The steel liner is %-inch thick at the sidewall, I/2-inch thick at the spherical dome, and 1/4/4-inch thick at the bottom. Penetrations in the Containment Building include the equipment hatch. This hatch has been permanently dismantled and removed. Since decommissioning activities began, an alternate personnel access opening has been created in the East side of the Containment Building.

Portions of the concrete surfaces inside the Containment Building are contaminated. Commodities are being removed from the Containment. The current plan is to demolish the interior structures of the Containment and liner to 4 ft below grade. The associated debris will be shipped to a licensed disposal facility. The remaining portion of the structure and lining will be decontaminated, as necessary.

CYAPCO retains the option to use Containment concrete debris as backfill.

The refueling cavity liner and surrounding concrete is being removed to allow for radiological surveys and any required decontamination activities.

The Containment Polar Crane will be used throughout decommissioning for removal of large components and structures. The Polar Crane will be one of the last items to be removed from Containment prior to its controlled demolition.

3.4.2.29.3 Primary Auxiliary Building (PAB)

The two-story Primary Auxiliary Building is primarily a reinforced-concrete structure with a one-story, braced steel frame superstructure. This building housed many of the reactor auxiliary systems. Shielding is provided by a combination of below grade construction and concrete walls and slabs on the first floor.

The second floor consists of structural steel framing for the Containment and PAB Ventilation System.

The pipe trench is located at or below ground elevation and extends the length of the building with laterals connecting to individual equipment cubicles. The pipe trench is covered by removable concrete slabs.

Ventilation equipment in the PAB supported radiological controls during decommissioning activities in the PAB, the Containment Building, and other structures supported by PAB ventilation. An alternate ventilation system has been installed to provide ventilation and exhaust monitoring for the PAB and Containment Building.

Selected water processing equipment in the PAB supported the processing of wastewater from decommissioning activities until they were either replaced by appropriate, alternative systems or the 0 9004

.Auriut _-. 3-22 Rev. 2 1

Haddam Neck Plant License Termination Plan associated supported decommissioning activities were verified to be complete. The concrete walls, floors, and pipe trench slabs may be decontaminated by scabbling methods to remove contaminated layers.

Floor and equipment drains have been abandoned in place and will be removed with the demolition of the structure. Portions of the PAB will be demolished to bedrock, and the waste materials generated will be disposed of at a licensed disposal facility. An assessment of the remaining bedrock will be performed (as discussed in Chapter 5). The area will then be backfilled to grade.

3.4.2.29.4 Other Buildings The Spent Resin Building will be demolished to bedrock. The waste materials generated including the system and components will be sent to a licensed disposal facility. The resins were processed prior to disposal at a licensed disposal facility.

The Service Building maintained the radiation control point until it, and associated personnel, were relocated. Subsequently the building will be dismantled and disposed. Portions of the building may be contaminated. The building has been divided into clean areas and contaminated areas. The clean portions of the building will be demolished to 4 ft below grade. The clean portions of the building will be disposed of at a regulated clean landfill, and the contaminated portions will be sent to a licensed disposal facility. An assessment survey of the excavation will be performed (along with any independent verification survey) The area will be backfilled to grade and an FSS will be performed.

The Turbine Building housed the plant's secondary side components, including the turbine generator, condenser and feedwater equipment. The portions of the Turbine Building systems that were internally contaminated were removed and disposed of at a licensed disposed facility. Portions of the Turbine Building will be demolished and disposed of at a regulated landfill.

The Screenwell House is located on the bank of the Connecticut River. The structure and its systems and components are not likely to be contaminated. The Screenwell House will be demolished to the contour of the river, and wastes will be sent to a regulated landfill or may be used as backfill. An FSS of the area and any verifications will be performed. The areas will then be restored.

The Waste Disposal Building will be surveyed to ensure contamination levels meet the criteria for controlled demolition. The building will be demolished to 4 ft below grade, and the associated wastes will be sent to a licensed disposal facility. An assessment of the remaining materials will be performed.

The area will then be backfilled to grade elevation and an FSS will be performed.

3.5 Radiological Impacts of Decommissioning Activities Decommissioning activities at Haddam Neck will be conducted in accordance with the Haddam Neck UFSAR, Technical Specifications, existing 10 CFR Part 50 License and the requirements of I OCFR50.82(a)(6) and (a)(7). If an activity requires prior NRC approval under IOCFR50.59(c)(2) or a change to the Haddam Neck Plant Technical Specifications or license, a submittal will be made to the NRC for review and approval before implementing the activity in question.

The decommissioning activities described herein are conducted under the provisions of the CYAPCO Radiation Protection Program and Radioactive Waste Management Program. These programs continue to be implemented as described in the HNP UFSAR. The Radiation Protection Program implements the regulatory requirements of IOCFR20 (Reference 3-7) through approved plant procedures established to maintain radiation exposures ALARA. The Radioactive Waste Management Program controls generation, August 2004 3-23 Rev. 2

Haddam Neck Plant License Termination Plan characterization, processing, handling, shipping, and disposal of radioactive waste per the approved CYAPCO Radiation Protection Program, Process Control Program, and plant procedures.

The current radiation protection program (described in UFSAR Chapter 12), waste management program (described in UFSAR Chapter 11), and Radiological Effluent Monitoring and Offsite Dose Calculation Manual (described in Section 6.6.3 of the HNP Technical Specifications) will be used to protect the workers and the public during the various decontamination and decommissioning activities. These well established programs are routinely inspected by the NRC to ensure that workers, the public, and the environment are protected during facility decommissioning activities. It is also important to note that most decommissioning activities involve very similar radiation protection and waste management considerations as those encountered during plant operations. As described in the PSDAR, the HNP decommissioning will be accomplished with no significant adverse environmental impacts in that:

• No site-specific factors pertaining to the HNP would alter the conclusions of the Final Generic Environmental Impact Statement (FGEIS).

• Radiation dose to the public will be minimal.

• Decommissioning is not an imminent health or safety concern and will generally have a positive environmental impact.

3.5.1 Occupational Exposure Detailed exposure estimates and exposure controls for specific activities are developed during detailed planning per Radiation Protection Program procedures. Table 3-3 provides estimated personnel exposures for various decommissioning and fuel storage activities. The total radiation exposure impact for decommissioning and spent fuel management is estimated in Table 3-3 to total approximately 935 person-rem (+/- 10%), as given in the PSDAR.

August 2004 3-24 Rev. 2

Haddam Neck Plant License Termination Plan Table 3-3 Radiation Exposure Projections for Decommissioning and Fuel Storage Activities Activity Exposure (person-rem)

Dismantlement Activities RCS Decon. 32 Asbestos Abatement (includes scaffolding) 136 Steam Generators and Pressurizer 88 Reactor Vessel Internals 90 Reactor Vessel & Head Prep (CRDM/ICI Structure) 72 Main Coolant System 17 Plant Systems 248 Structures 74 Miscellaneous 72 Waste Processing (includes shipping and prep) 12 Subtotal Dismantlement Activities 841 Operational Activities 1996 Operations--post certifications 1 1997 Operations 17 Spent Fuel Pool Isolation Modifications 10 Spent Fuel Storage 24 Fuel Transfer/Cask Loading 42 Subtotal Operational Activities 94 Total for Decommissioning and Fuel Storage Activities 935 Transportation (occupational and to general public) 71 Rex'. 2 3-25 August 2004 August 2004 3-25 Rev. 2 I

Haddam Neck Plant License Termination Plan 3.5.2 Radioactive Waste Projections The Radioactive Waste Management Program is used to control the characterization, generation, processing, handling, shipping, and disposal of radioactive waste during decommissioning. Activated and contaminated systems, structures, and components represent the largest volume of low level radioactive waste expected to be generated during decommissioning. Other forms of waste generated during decommissioning include:

1. Contaminated water;

2. Used disposable protective clothing;

3. Expended abrasive and absorbent materials;

4. Expended resins and filters;

5. Contamination control materials (e.g., strippable coatings, plastic enclosures); and

6. Contaminated equipment used in the decommissioning process.

Table 3-4 provides projections of waste quantities for decommissioning. These waste quantities are those reflected in the PSDAR. The total volume of HNP low-level radioactive waste for disposal has been estimated at 1,158,000 cubic feet. Actual waste volumes and classifications may vary, but the total quantity is not expected to exceed 1,158, 000 cubic feet.

Decommissioning planning at CYAPCO incorporates the assumption that cost-effective waste volume reduction methods are limited. It also assumes some significantly contaminated or activated materials are sent directly to a disposal facility. However, alternative processing methods consistent with the approaches described herein may be evaluated and used during decommissioning.

Table 3-4 Projected Waste Quantities Radwaste Volume Commodities 41,292 Containment 426,506 Misc Packaged Radwaste Awaiting Shipment 33,333 Misc. R2 Structures Disposal 101,454 PAB 69,330 Service Building 55,828 Spent Fuel Building 59,962 Terry Turbine Building 316 Waste Disposal Building 22,096 Containment Liner (Mixed PCB/radwaste) 12,407 Contaminated Roofs 17,949 Dredge Spoils 83,333 Rad Soil 187,142 Rad Asphalt 46,568 Rad Subtotal 1,157,516 August 2004 3-26 Rev. 2 I

Haddam Neck Plant License Termination Plan 3.6 References 3-1 Code of Federal Regulations, Title 10, Part 50.82, "Termination of License."

3-2 Letter CY-97-075 from CYAPCO to the USNRC, "Haddam Neck Plant Post Shutdown Decommissioning Activities Report," dated August 22, 1997.

3-3 Letter CY-98-144 from CYAPCO to the Department of Transportation, "Request for Exempting the Shipment of the Haddam Neck Plant Steam Generator Subassemblies," dated November 5, 1998 as revised by CY-99-073, dated June 3, 1999.

3-4 Letter CY-00-015 from CYAPCO to the Department of Transportation, "Exemption Request from the Surface Contaminated Object Demonstration Requirements of 49CFR1 73.403 and the Packaging Requirements of 49CFR173.427(b)(1) for the Shipment of a Pressurizer from the Haddam Neck Nuclear Plant," dated January 18, 2000.

3-5 Not used.

3-6 Not used.

3-7 Code of Federal Regulations, Title 10 Part 20, "Standards for Radiation Protection."

3-8 Revision 2 to the PSDAR, dated April 28, 2004.

August 2004 3-27 Rev. 2

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

August 2004 3-28 Rev. 2 l

Haddam Neck Plant License Termination Plan 4 SITE REMEDIATION PLANS 4.1 Introduction In accordance with IOCFR50.82 (a)(9)(ii)(C) (Reference 4-1), the LTP must provide the "plans for site remediation." These plans must include the provisions to meet the criteria from Subpart E of IOCFR20 (Reference 4-2) before the site may be released for unrestricted use:

• Annual total effective dose equivalent to the average member of the critical group not to exceed 25 mrem, and

• The dose to the public must be "as low as reasonably achievable," or ALARA.

Decontamination and dismantlement activities will be conducted in accordance with the CY Radiation Protection, Safety and Waste Management Programs, which are well established and frequently inspected. Changes have been made to these programs for D&D activities, and any future changes that may be made will be documented and processed with existing plant administrative procedures using IOCFR50.59 and the guidance contained in Regulatory Guide 1.187.

This section describes the methodologies and criteria that will be used to perform remediation activities of residual radioactivity and to demonstrate compliance with the ALARA criteria, required by IOCFR20.

More specific detail regarding remediation activities may be found in Chapter 3.

4.2 Remediation Levels and ALARA Evaluations When dismantlement and decontamination actions are completed, residual radioactivity may remain on building surfaces and on site soils. Residual radioactivity must satisfy the provisions of IOCFR20, Subpart E. As depicted on Figure 4-1, the ALARA cleanup levels for the HNP decommissioning may be established at one of two levels:

(I) a predefined generic ALARA screening, or (2) a survey unit-specific ALARA evaluation.

In either case, the ALARA evaluation uses an action level, referred to as a remediation level. This remediation level corresponds to a residual radioactivity concentration at which the averted collective radiation dose converted into dollars is equal to the costs of remediation (e.g., risk of transportation accidents converted into dollars, worker and public doses associated with the remediation action converted into dollars, and the actual costs to perform the remediation activity).

If the value of further dose reduction from remediation is greater than the "costs" of the action, then the remediation action being evaluated is cost-effective and should be performed. Conversely, if the value of further dose reduction is less than the costs, the levels of residual radioactivity are considered ALARA and therefore further remediation action would not be required. The methodology and equations used for calculating remediation levels are consistent with those provided in Draft Regulatory Guide DG-4006, "Demonstrating Compliance with the Radiological Criteria for License Termination" (Reference 4-3), as incorporated in Appendix D of NUREG-1727 (Reference 4-4). These are provided in Appendix B of the LTP. Documentation of ALARA evaluations will be included in the final status survey documentation for each survey area.

August 2004 4-1 Rev. 2

Haddam Neck Plant License Termination Plan The selection of appropriate instrumentation for post-remediation surveys is important from a planning and financial risk management perspective. Post-remediation surveys serve as a check to see if the remediation target is achieved. As shown on Table 5-10 of this LTP, if small handheld beta-gamma detectors are used to determine if remedial actions have been successful, the nominal frisk MDC is estimated to be between 1200 to 3200 dpm/100 cm 2 . Based upon site characterization data, the predominant beta-gamma emitting radionuclides at HNP are Co-60 and Cs-137. The corresponding building surface DCGLs for Co-60 and Cs-137 as shown in Table 6-3 are 11,100 and 43,000 dpm/100 cm2 respectively. Therefore, the MDCs of the handheld detectors are approximately only 3 to 30 percent of the predominant beta-gamma emitting radionuclide DCGLs. In other cases, the actual final status survey instrumentation may be used to evaluate remedial actions.

4.2.1 Generic ALARA Screening Levels As discussed in DG-4006 (and Appendix D of NUREG-1727), soil remediation beyond the DCGLs is not likely to be cost-beneficial due to the high costs of waste disposal. This has been confirmed in a generic ALARA evaluation for soils. Thus soil will be at ALARA levels when it meets the site-specific DCGLs discussed in Chapters 5 and 6.

For building surfaces, a generic ALARA screening value will be calculated using conservative estimates for building remediation costs. This generic ALARA screening value will be calculated using the guidance of DG-4006, after additional characterization has been undertaken and remediation methods have been evaluated for their effectiveness. This value will represent the level, expressed as a percentage or fraction of the DCGL, for which the benefit of further remediation of structures is greater than the associated costs.

Upon completion of post-remediation surveys and satisfaction of the 25 mrem/yr TEDE criteria, the level of residual radioactivity in the survey area will be compared against the appropriate generic ALARA screening value. Where the level of residual radioactivity is lower than the generic ALARA screening value, the remediation is clearly ALARA, no further remediation is required, and final status surveys can proceed. Where the level residual radioactivity is greater than the generic ALARA screening value, a survey-unit ALARA evaluation is performed to determine the unit-specific ALARA remediation level for comparison.

4.2.2 Survey-Unit Specific ALARA Evaluation In cases where levels of residual radioactivity are above the generic ALARA screening levels described above, survey unit-specific ALARA evaluations will be performed using approved site procedures. These survey unit-specific ALARA evaluations will be performed using data from post-remediation surveys in accordance with DG-4006 and will take into account:

• Radiation doses and environmental impacts for the decommissioning process and from the residual radioactivity remaining onsite following the decommissioning, and

• Other costs and risks associated with the decontamination and decommissioning of the site.

Once the total cost, CostT, for a survey-unit specific remediation action has been calculated, a remediation level, expressed as a fraction of a DCGL, can be determined and the ALARA evaluation can be performed using the process described in DG4006.

August 2002 4-2 Rev. I

Haddam Neck Plant License Termination Plan The remediation levels represent the radioactivity concentrations at which a remediation action is cost beneficial and, therefore, do not represent maximum or "not-to-exceed" concentrations. The ALARA criteria are met by performing the remediation action and not necessarily by achieving results below a specific remediation level. An ALARA analysis ensures that the efforts to remove residual contamination are commensurate with the risk that exists with leaving the residual contamination in place, even if the target remediation levels are not achieved. However the residual contamination must be low enough to assure the annual dose to the average member of the critical group does not exceed 25 mrem/yr TEDE.

4.2.3 Groundwater ALARA Evaluation As discussed in Section 1.6 of Appendix D to NUREG-1727, if there is residual radioactivity from site operations in groundwater, it may be necessary to calculate the collective dose from consumption of the groundwater. Sampling of the groundwater at the HNP Site indicates that residual radioactivity does exist. Dose modeling, as discussed in Chapter 6, assumes that the aquifer does not supply a large population, but only the resident farming family. Currently there is no population deriving its drinking water from a downstream supply, and based upon current knowledge of the aquifer onsite, it is doubtful that this aquifer would be used as a drinking water source for a large population. Wells providing water of potable quality cannot supply high well pumping rates needed to support a large population, and wells that can supply higher pumping rates are located in the unconsolidated sediments and provide poor quality water. These conditions will be re-confirmed during the program of ongoing groundwater monitoring, as described in Section 2.3.3.1.6.2. However, if it is determined that drinking water for a large population could be supplied by groundwater onsite, the collective dose for that population will be included in the ALARA calculation as indicated in NUREG-1727, Appendix D.

4.3 Remediation Actions Remediation actions may be required to reduce the residual radioactivity levels below the applicable cleanup criteria as provided in Chapters 5 and 6. The specific remedial actions depend on the type of area under consideration. These area types are categorized as one of the following:

1. Structures (including building interiors and exteriors, major freestanding exterior structures, exterior surfaces of plant systems, and paved exterior ground surfaces);

2. Soils; and

3. Nonstructural plant systems (including interior surfaces of process piping and components).

4.3.1 Structures Using the demolition and bulk disposal approach to decommissioning, concrete from contaminated structures will be remediated to levels meeting the radiological criteria for controlled demolition. The debris materials will be disposed of at a licensed radioactive material disposal facility. Nonstructural materials will be assessed using the process in Section 5.6.

Methods for remediating structures may include a variety of techniques ranging from water washing to surface material removal. A number of factors determine the choice of the remediation method for a given area, including: the size of the contaminated area, the extent of contamination, surface material, depth of contamination, and accessibility.

August 2004 4-3 Rev. 2

Haddam Neck Plant License Termination Plan Remediation activities for an area may include wiping, vacuuming, and washing with low- or high-pressure applications. Surfaces may also be remediated using surface removal techniques such as scabbling or grinding.

For concrete surfaces, remediation methods may include core drilling, concrete sawing, or scabbling.

Scabbling removes the concrete surface by bush heads, rotopeen devices, flappers, or similar devices and is effective for removing contamination that resides close to the surface. Abrasive blasting may also be used as an effective technique for contamination removal from surfaces that are not necessarily smooth.

Also, chipping, jackhammering, and other similar aggressive methods may be needed for removal of concrete surfaces as deep as the first mat of reinforcing steel. Strippable coatings can be used to remove contaminants from surfaces where more aggressive methods may not be appropriate or when other techniques are not successful.

4.3.2 Soils During 1998 and 1999 greater than one hundred subsurface soil samples, in some cases down to six feet in depth, were collected outside of the RCA in Survey Areas 9302, 9304, 9306 and 9308 in support of plant modifications and site characterization activities. None of the samples had plant related radioactivity levels greater than the corresponding base-case soil DCGLs. During the same time period, over two hundred subsurface soil samples, in some cases down to four feet in depth, were collected inside the RCA in Survey Areas 9307, 9310, 9312 and 9227 (Bus-10 Pad and ground underneath). Some isolated spots showed Co-60 and Cs-137 activity levels up to several hundred pCi/g each. Most of the sampling was performed in Survey Areas 9310 and 9227 in support of the Spent Fuel Pool project. Soil under Bus 10 was removed until that soil which remained was at or below the NRC's generic soil screening DCGLs, adjusted to 10 mrem/yr. Continued characterization of site soils has resulted in identification of contaminated areas and depth of the contamination. This will continue throughout the decommissioning.

Soils not meeting the applicable DCGLs will be removed and disposed of as radioactive waste. Offsite fill will be used to replace the excavated materials. As discussed previously in Chapter 2, the ongoing site characterization process establishes the location, depth and extent of soil contamination. As needed, additional investigations will be performed to ensure that any soil contamination profiles that may change during the remediation actions are adequately identified and characterized. In cases where offsite fill will be used to replace the excavated materials, a direct radiation survey will be conducted of either each load of fill or of the site from which the material will be obtained. This will be done as a documented survey to ensure that the background radiation levels (from the presence of naturally occurring radioactive material) from this fill material will not be significantly higher than that from the onsite material. This survey may be performed using CYAPCO's vehicle monitoring system (Bicron ASM 6000), by a handheld scintillation-based survey meter (sensitive to changes in ambient background radiation levels),

sample collection or analysis, or in-situ gamma spectroscopy measurements. Based upon the results of this survey, either background radiation levels will be accounted for in subsequent final status surveys or the material will be rejected for use.

4.3.3 Nonstructural Systems Chapter 3 describes the systems to be decontaminated, demolished, and disposed of. Contaminated plant systems and components may be sent to an offsite processing facility or to a low-level radioactive waste disposal facility. Slightly contaminated systems may be decontaminated onsite and released.

Nonstructural systems and components will be surveyed and released using existing plant procedures and processes (i.e., "free release criteria"), with the exception of those cases discussed in Section 5.4.7.5.

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-- 4-4 Rev f w

Ask

Haddam Neck Plant License Termination Plan Remediation methods typically used for system decontamination include chemical decontamination, wiping, washing, vacuuming, or abrasive blasting. Selection of the preferred method is based on the specific situation. Other remediation technologies may be considered and used, as appropriate.

4.4 References 4-1 Code of Federal Regulations, Title 10, Part 50.82, "Termination of License."

4-2 Code of Federal Regulations, Title 10, Part 20.1402, "Radiological Criteria for Unrestricted Use."

4-3 Draft Regulatory Guide-4006, "Demonstrating Compliance with the Radiological Criteria for License Termination," August 1998.

4-4 NUREG-1727, "NMSS Decommissioning Standard Review Plan," September 2000.

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Haddam Neck Plant License Termination Plan Figure 4-1 Survey Unit ALARA Evaluation Process I Initiate Pre-Remediation Characterization Surveys l August 2004 Rev. 2 1

Haddam Neck Plant License Termination Plan 5 FINAL STATUS SURVEY PLAN 5.1 Introduction 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. These surveys serve as key elements to demonstrate that the dose from residual radioactivity is less than the maximum annual dose criterion for license termination for unrestricted use specified in IOCFR20.1402 (Reference 5-1). The additional requirement of IOCFR20.1402 that all residual radioactivity at the site be reduced to levels that are as low as reasonable achievable (ALARA) is addressed in Chapter 4. The Final Status Survey Plan was developed using the guidance of Draft Regulatory Guide DG-4006, "Demonstrating Compliance with the Radiological Criteria for License Termination" (Reference 5-2); NUREG-1575, "The Multi-Agency Radiological Site Survey and Investigation Manual (MARSSIM)" (Reference 5-3); and Regulatory Guide 1.179, "Standard Format and Content of License Termination Plans for Nuclear Power Reactors" (Reference 5-4). In September of 2000, the NRC incorporated much of the guidance of DG-4006 into various sections of NUREG-1727; "NMSS Decommissioning standard Review Plan" dated September 2000 (Reference 5-5). References to the corresponding sections of NUREG-1727 (in which the guidance of DG-4006 has been incorporated) have been given in specific sections of this LTP, as appropriate.

The final status survey process described in this plan adheres to the guidance of MARSSIM for the design of final status surveys. However, advanced survey technologies may be used to conduct radiological surveys that can effectively scan 100% of the surface and record the results. This survey plan allows for the use of these advanced technologies, where survey quality and efficiency can be increased, as long as certain criteria are met. These criteria ensure that the survey results are at least equivalent to those that would have been obtained using the non-parametric sampling methods of MARSSIM in terms of their statistical confidence. In cases where advanced survey technologies are to be used, a technical support document will be developed to describe the technology to be used and to demonstrate how the technology meets the objectives of the survey. These technical support documents will be referenced, as appropriate, in Final Status Survey Reports.

5.2 Scope The final status survey plan encompasses the radiological assessment of all affected structures, systems and land areas for the purpose of quantifying the concentration of any residual activity that exists following all decontamination activities. Concentration limits will be established to represent the maximum annual dose rate criterion for unrestricted release specified in I OCFR20.1402.

5.3 Summary of the Final Status Survey Process The final status survey provides data to demonstrate that all radiological parameters satisfy the established guideline values and conditions. The primary objectives of the final status survey are to:

• select/verify survey unit classification,

• demonstrate that the potential dose from residual radioactivity is below the release criterion for each survey unit, and August 2002 5-1 Rev. I

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 necessary.

Before the survey process can proceed to the design phase, concentration levels that represent the maximum annual dose criterion of I OCFR20.1402 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. 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 DCGLEMC, 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.

Survey implementation is the process of carrying out the survey plan (package) for a given survey unit.

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

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Haddam Neck Plant License Termination Plan The Data Quality Assessment (DQA) approach is applied to FSS results to ensure their validity and to demonstrate that the objectives of the FSS are met. Data assessment includes data Verification and Validation (V&V), review of survey design bases, and data analysis. For a given survey unit, the survey data are evaluated to determine if the residual activity levels in the unit meet the applicable release criterion and if any areas of elevated activity exist. In some cases, data evaluation will simply serve to show that all of the measurements made in a given survey unit were below the applicable DCGLw. If so, demonstrating compliance with the release criterion is a simple matter and requires little in the way of analysis. In other cases, residual radioactivity may exist where measurement results both above and below the DCGLw are observed. In these cases, statistical tests must be performed to make a decision as to whether the unit meets the release criterion. The statistical tests that might be required to make decisions regarding the residual activity levels in a survey unit relative to the applicable DCGLW must be considered in the survey design to ensure that a sufficient number of measurements are collected.

MARSSIM specifies two non-parametric statistical tests to be applied to final status survey data to evaluate whether a set of measurements demonstrates compliance with the release criterion for a given survey unit. These statistical tests are discussed in detail in Section 5.8.

Quality assurance and control measures are employed throughout the final status survey process to ensure that all decisions are made on the basis of data of acceptable quality. Quality assurance and control measures are applied to ensure:

• the plan is correctly implemented as prescribed,

• DQOs are properly defined and derived,

• all data and samples are collected by individuals with the proper training following approved procedures,

• all instruments are properly calibrated,

• all collected data are validated, recorded, and stored in accordance with approved procedures,

• all required documents are properly maintained, and,

• if necessary, corrective actions are prescribed, implemented and followed up.

These measures apply to any services provided in support of final status survey.

Survey results will be converted to appropriate units (i.e., either dpm/I00 cm2 or pCi/g) and compared to investigation levels to determine appropriate follow-up action. Measurements exceeding investigation levels will be verified and investigated and, following confirmatory measurement(s), the affected area may be remediated and/or reclassified and a re-survey performed consistent with the guidance in MARSSIM (Section 8.5.3, "If the Survey Unit Fails") and commensurate with the classification and extent of contamination.

Documentation of the final status survey will transpire in two types of reports. An FSS Survey Unit Release Record will be prepared to provide a complete record of the "as left" radiological status of an individual survey unit, relative to the specified release criteria. Survey Unit Release Records will be made available to the NRC for review upon request. An FSS Final Report, which is written report submitted to the NRC, will be prepared to provide a summary of the survey results and the overall conclusions which demonstrate that the Haddam Neck Plant site, or portions of the site, meets the radiological criteria for unrestricted use. These reports are discussed in detail in Section 5.9.

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Haddam Neck Plant License Termination Plan The documentation describing the final status survey for a given survey unit will include:

• a physical description of the survey area which encompasses the unit(s) (in many cases, the survey areas and survey units will be the same);

• the characterization data associated with the area, including any required investigations, re-classifications or subdivisions;

• the classification history of the unit;

• the remediation activities (if any) performed in the survey unit;

• results and discussion of any ALARA evaluations performed;

• a discussion of the survey design (combination of scans, fixed measurements, samples, number of measurements, grid spacing, etc.);

• tabular and graphical depictions of survey results;

• discussions of data assessments, including graphical depictions; and

• conclusion that survey unit meets all applicable criteria.

It is anticipated that final status survey results will be documented and made available to the NRC for survey areas rather than for individual survey units. Reports will be compiled after final status survey activities for all of the survey units for a given area are completed. This approach should minimize the submittal of redundant historical assessment information and provide for a logical approach to perform reviews and independent verification.

5.4 Survey Planning 5.4.1 Data Quality Objectives The DQO process is incorporated as an integral component of the data life cycle at HNP. The DQO process is used in the planning phase for scoping, characterization, remediation, and final status survey plan development using a graded approach. Survey plans that are complex or that have a higher level of risk associated with an incorrect decision (such as final status surveys) would require significantly more effort than a survey plan used to obtain data relative to the extent and variability of a contaminant. This process, described in MARSSIM, is a series of planning steps found to be effective in establishing criteria for data quality and developing survey plans. DQOs allow for systematic planning and are specifically designed to address problems that require a decision to be made and provide alternate actions.

Furthermore, the DQO process is flexible in that the level of effort associated with planning a survey is based on the complexity of the survey and nature of the hazards. Finally, the DQO process is iterative allowing the survey planning team to incorporate new knowledge and modify the output of previous steps to act as input to subsequent steps. The Final Status Survey Quality Assurance Project Plan (FSSQAPP) provides a general description of the application of the DQO process to the different elements of the final status survey.

The DQO process consists of performing the following seven steps:

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Haddam Neck Plant License Termination Plan

• State the Problem

• Identify the Decision

• Identify the Inputs to the Decision

• Define the Boundaries of the Decision

• Develop a Decision Rule

• Specify Tolerable Limits on Decision Errors

• Optimize the Design for Obtaining Data The actions taken to address these DQO process steps during the planning of a final status survey for a particular survey area are addressed below.

• State the Problem The first step of the planning process consists of defining the problem. This step provides a clear description of the problem, identification of planning team members (especially the decision-makers), a conceptual model of the hazard to be investigated and the estimated resources. The problem associated with FSS is to determine whether an area meets the radiological release criterion of IOCFR20.1402.

• Identify the Decision This step of the DQO process consists of developing a decision statement based on a principal study question (i.e., the stated problem) and determining alternative actions that may be taken based on the answer to the principal study question. Alternative actions identify those measures to resolve the problem. The decision statement combines the principal study question and alternative actions into an expression of choice among multiple actions. For FSS the principal study question is "Does residual radioactive contamination present in the survey unit exceed the release criteria?" The alternative actions may include no action, investigation, resurvey, remediation and reclassification.

• Identify Inputs to the Decision The information required depends on the type of media under consideration (e.g., soil, water, concrete) and whether existing data are sufficient or new data are needed to make the decision. If the decision can be based on existing data then the source(s) will be documented and evaluated to ensure reasonable confidence that the data are acceptable. If new data are needed, then the type of measurement (e.g., scan, direct measurement and sampling) will be determined.

Sampling methods, sample quantity, sample matrix, type(s) of analyses and analytic and measurement process performance criteria, including detection limits, are established to ensure adequate sensitivity relative to the action level and to minimize bias. Action levels provide the criterion for choosing among alternative actions (e.g., whether to take no action, perform confirmatory sampling). These action levels may be radioactivity concentration (pCi/g) or measurement device response (count rate corrected for background).

• Define the Boundaries of the Study.

This step of the DQO process includes identification of the target population of interest, the spatial and temporal features of the population pertinent to the decision, time frame for collecting August 2004 5-5 Rev. 2

Haddam Neck Plant License Termination Plan the data, practical constraints and the scale of decision making. In FSS, the target population is the set of samples or direct measurements that constitute an area of interest (i.e., the survey unit).

The medium of interest (e.g., soil, water, concrete, steel) is specified during the planning process.

The spatial boundaries include the entire area of interest including soil depth, area dimensions, contained water bodies and natural boundaries, as needed. Temporal boundaries include those activities impacted by time-related events including weather conditions, seasons (i.e., more daylight available in the summer), operation of equipment under different environmental conditions, resource loading and work schedule.

• Develop a Decision Rule This step of the DQO process develops the binary statement that defines a logical process for choosing among alternative actions. The decision rule is a clear statement using the "If...then..."

format and includes action level conditions and the statistical parameter of interest (e.g., mean of data). Decision statements can become complex depending on the objectives of the survey and the radiological character of the affected area.

• Specify Tolerable Limits on Decision Errors This step of the DQO process incorporates hypothesis testing and probabilistic sampling distributions to control decision errors during data analysis. Hypothesis testing is a process based on the scientific method that compares a baseline condition to an alternate condition. The baseline condition is technically known as the null hypothesis. Hypothesis testing rests on the premise that the null hypothesis is true and that sufficient evidence must be provided for rejection.

The primary consideration during FSS will be demonstrating compliance with the release criteria.

The following statement may be used as the null hypothesis at HNP: "The survey unit exceeds the release criteria."

Decision errors occur when the data set leads the decision-maker to make false rejections or false acceptances during hypothesis testing. The cc error (Type I error) is set at 0.05 (5%). The P error may be variable depending upon the objectives of the surveys. A nominal value of 0.05 (5%) has been established for the f error (Type II error). Another output of this step is assigning probability limits to points above and below the gray region where the consequences of decision errors is considered acceptable. The upper bound corresponds to the release criteria. The Lower Bound of the Gray Region (LBGR) is determined in this step of the DQO process. LBGR is influenced by a parameter known as the relative shift. The relative shift is set between (and including) I and 3. If the relative shift is not between (or including) I and 3, then the LBGR is adjusted.

Graphing the probability that a survey unit does not meet the release criteria may be used during FSS. This graph, known as a power curve, may be performed retrospectively (i.e., after FSS) using actual measurement data. This retrospective power curve may be important when the null hypothesis is not rejected (i.e., the survey unit does not meet the release criteria) to demonstrate that the DQOs have been met.

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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.

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 DG-4006 (as incorporated in Section 2 of Appendix E to NUREG-l 727) 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 IA, 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 has been assigned an initial classification based on the site characterization process and the historical site assessment.

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Haddam Neck Plant License Termination Plan Table 5-1 HNP Survey Unit Surface Area Limits Survey Unit Classification Surface Area Limit Class 1:

Structures (floor area) <100 m2 Land areas <2,000 m2 Class 2:

Structures (floor area) 100 m2 < area < 1,000 m2 Land areas 2,000 m2 < area < 10,000 m2 Class 3:

Structures (floor area) no limit Land areas no limit A survey unit can have only one classification. Thus, situations may arise where it is necessary to create new survey units by subdividing areas within an existing unit. For example, residual radioactivity may be found within a Class 3 survey unit, or residual radioactivity in excess of the DCGLw may be found in a Class 2 unit. In such cases, it may be appropriate to define a new survey unit within the original unit that has a lower (more restrictive) classification. Alternately, the classification of the entire unit can be made more restrictive.

Likewise, survey units may need to be added or sub-divided to account for attached fixtures (such as cable trays, piping, and pipe hangers) that remain in an area after decommissioning activities are completed. The decision to define a new survey unit to account for attached fixtures can be made out of necessity, for compliance with the area limits from Table 5-1, or out of convenience to allow for a consistent survey approach within a given unit. If situations arise where it is neither necessary nor convenient to define additional survey units for attached fixtures, the fixtures will be considered to be part of the unit they are attached to. Attached fixtures and their impact on survey unit definitions cannot be addressed a priori, since major decommissioning activities are still ongoing.

5.4.4 Reference Coordinate Systems The reference coordinate system depicted in Figures 5-1 and 5-2 will be used to provide a general reference for locations within a survey unit. This coordinate system will not be used to explicitly specify locations for fixed measurements or samples, but instead will serve as a convenience for documenting survey efforts and other information pertaining to a given survey unit. The coordinate system could also provide a means to specify general locations for measurements or samples performed for quality control or verification purposes.

At a minimum, each survey unit will have a benchmark defined that will serve as an origin for documenting survey efforts and results. This benchmark (origin) will be provided on the map or plot included in the final status survey package. Any coordinate systems used for surveys will typically take the form of a grid of intersecting, perpendicular lines; but other patterns (e.g., triangular and polar) may be used as convenient. Physical gridding of a survey unit will only be done in cases where it is beneficial and cost effective to do so. When physical gridding is used, benchmark locations will be designated by either marking a spot with surveyor's paint (or equivalent) for indoor areas or setting an iron pin (or August 2004 5-8 Rev. 2

Haddam Neck Plant License Termination Plan equivalent) for outdoor areas. If needed, grid lines or measurement locations will be marked (e.g., with chalk lines, paint, surveyor's flags), as appropriate. Global positioning systems will also be used as practical.

5.4.5 Reference Areas and Materials The DQO process will be used during the planning phase in the preparation of a final status survey plan to determine whether media specific backgrounds, ambient area background or no background will be applied to a survey area or unit. The approach used for a specific survey unit will be based on the survey unit classification and the applicable DCGLs.

If applied, media specific backgrounds will be determined via measurements made in one or more reference areas and on various materials selected to represent the baseline radiological conditions for the site. The determination of media specific background will be controlled with a documented survey plan, which will include the DQO process. These data will be evaluated in a technical support document and available for inspection by the NRC. This process will ensure that the collected data meet the needs of the final status survey. The collected data may be used as the reference area data set when using the Wilcoxon Rank Sum test, or, for survey units with multiple materials, background data may be subtracted from survey unit measurements (using paired observations) if the Sign Test is applied.

Table 5-2 gives typical ranges for backgrounds expected to be encountered at the HNP during final survey activities. Ranges are given for several detector types (gross counters) and encompass the variability expected for different materials. The data in Table 5-2 are derived from both NUREG-I 507 (Reference 5-6) and from experience at the HNP.

Table 5-2 Typical Media Specific Backgrounds Instrument Nominal Background Range gas proportional counter (100 cm2 )

a-only mode 1 cpm - 20 cpm for ceramic tile; 1 cpm - 10 cpm for other materials f3-only mode a+p mode 300 cpm - 1,250 cpm for all materials 280 cpm - 1,250 cpm for all materials pancake GM probe (20 cm2 ) 40 cpm - 125 cpm for all materials ZnS (100 cm2 ) I cpm - 10 cpm for ceramic tile; I cpm - 5 cpm for other materials plastic scintillator (100 cm2 ) 500 cpm - 1,500 cpm for all materials Nal I inch by I inch 2,000 cpm - 4,000 cpm for soil 1.25 inch by 1.5 inch 3,000 - 6,000 cpm for soil 2 inch by 2 inch 8,000 cpm - 16,000 cpm for soil Depending on the values of the applicable DCGLs, an alternative method to using material specific I backgrounds may be used during final status surveys. This alternative method will involve the determination of the ambient area background in the survey unit and will only be applicable to beta-gamma detecting instruments. This determination will be made prior to performing a final status survey August 2004 5-9 Rev. 2 l

Haddam Neck Plant License Termination Plan at a location within a survey area that is of sufficient distance (or attenuation) from the surfaces to eliminate beta particles originating from the surfaces from reaching the detector. At such a location, the ambient background radiation will be due only to ambient gamma radiation and will be a background component of all surface measurements. The average background determined at this location can be used as a conservative estimate since it is expected to be less than the material specific background for the material in the room because it does not fully account for the naturally occurring radioactivity in the materials. Using this lower ambient background will result in conservative calculated residual radioactivity levels. If the average background reading exceeds a predetermined value, the survey would be terminated and an investigation performed to determine and eliminate the reason for the elevated reading. Each of the survey unit readings would subtract this average background value and the Sign Test applied.

Whether or not they are radionuclide specific, all background measurements should account for both spatial variability over the area being assessed and the precision of the instrument or method being used to make the measurements. Thus, the same materials or areas may require more than one background assessment to provide the requisite background information for the various survey instruments or methods expected to be used for final status surveys. The result of these background assessments provides the basis for determining the mean and its associated standard deviation.

5.4.6 Area Preparation: Isolation and Control Before final status survey activities can begin in an area, a transition must occur where planned decommissioning activities are completed and the area is subsequently assessed to scope the required isolation and control measures. This includes establishing if the area is ready for final survey activities and identifying any work practice issues that must be addressed in survey planning and design.

Determination of readiness for final status survey will be based on characterization and/or remediation surveys indicating that the residual radioactive material is likely to comply with the final status survey criteria.

During and following this assessment, the remediation of an area for the purposes of removal of residual radioactivity, a Remediation Action Survey (RAS) will be performed. This RAS will include scanning and sampling of the areas, as necessary, using appropriate instrumentation to ensure the intended remediation has been accomplished. If no further remediation is required, a turnover survey may be performed as necessary. Data from the RAS may be used for the turnover survey and as input to the design and DQOs for the surface and subsurface FSS.

For areas where an excavation has resulted from the removal of materials not contaminated with residual radioactivity, and contamination has not been detected or suspected, a graded approach will be used to assess the excavation. Limited biased sampling (and scanning of suspect areas) will be conducted in the excavation. If contaminated material is detected, additional sampling will be conducted to determine the extent and magnitude of the contamination. Remediation and a Remedial Action Survey will be conducted as necessary. The characterization information from this assessment may be used for any turnover survey needed and as input to the design and DQOs for the surface and subsurface FSS.

5.4.6.1 Structures The structures that remain (including below-ground portions of structures below elevation 17flt-6in) will be decontaminated and prepared for final status survey. Following a readiness assessment for final status survey, isolation and control measures will be implemented to prevent the introduction of plant-related contamination into soils or structures in the area, prior to, during or after final survey activities. These August 2004 5-10 Rev. 2

Haddam Neck Plant License Termination Plan control measures will include posting (e.g., with a placard or sign) areas that have been turned over for final status survey. Isolation and control measures are implemented for areas such as an entire structure.

If additional remediation is required in an area following the implementation of isolation and control measures, local contamination control measures such as tents, HEPA filters, or vacuums will be employed as appropriate.

Prior to transitioning an area from decommissioning activities to isolation and control, a walkdown may be performed to identify access requirements and to specify the required isolation and control measures.

The physical condition of the area will also be assessed, with any conditions that could interfere with final survey activities identified and addressed. If any support equipment needed for final survey activities, such as ladders or scaffolding, are in place, it will be evaluated to ensure that it does not pose the potential for introducing radioactive material into the area. Industrial safety and work practice issues, such as access to elevated areas requiring fall protection, will also be identified during the pre-survey evaluation.

Operational health physics or decontamination support data, if available, will be reviewed to identify any potential areas where additional decontamination may be required prior to commencing final survey activities. In some instances, turnover surveys may be performed to verify that an area is ready for final survey.

The following criteria must be met for an area to be deemed ready for isolation and control:

• all planned decommissioning activities in the area are complete, including removal, as necessary, of items (e.g., equipment mounts, wall hangers, and exposed studs) that could interfere with final survey activities;

• all planned decommissioning activities in areas either adjacent to the area to be isolated or that could otherwise affect it are controlled using isolation and control techniques, are complete or are deemed not to have any reasonable potential to spread plant-related radioactive material to the area;

• all tools and equipment not needed for final survey activities are removed;

• any equipment to be used for final survey activities is evaluated to ensure it does not pose the potential for introducing plant-related radioactive material into the area; and

• where practical, transit paths to or through the area, except those required to support final survey activities, are eliminated or re-routed.

Once the area meets the isolation and control criteria, isolation and control will be achieved through:

• a combination of personnel training, physical barriers and postings, and site notices as appropriate, to prevent unauthorized access to an isolated area;

• implementation of provisions to prevent the introduction of plant-related radioactive material by persons authorized to enter the area; and

• measures to prevent the introduction of plant-related radioactive material through the air or through other paths, such as systems or piping.

August 2004 5-1 1 Rev. 2

Haddam Neck Plant License Termination Plan Measures to prevent against the introduction of plant-related radioactive material by persons entering an isolated area may include personnel frisking stations at the entry point, the use of "sticky pads", or other such routine methods. Isolation from airborne material may include sealing off openings, including doors and ventilation ducts. Although not likely to be encountered, if a potential for waterborne material is deemed to exist (e.g., floor drains or penetrations left by decommissioning activities), similar measures will be taken to ensure such sources are sealed off from the isolated area. In addition to these physical controls, access points to buildings will be posted with signs providing contact information for approval to conduct decommissioning and demolition activities in the area. An administrative process will be used to evaluate, approve (or deny), and document all plant related activities conducted in these areas during and following final status surveys.

Following the final status survey, and any regulatory confirmation, the excavations associated with the structures will be backfilled with bulk fill material. Any isolation and control measures needed at the restored surface will be implemented to protect the area from contamination.

5.4.6.2 Open Land Areas For open land areas, access roads and trails will be posted (as well as informational notices) with signs providing contact information for approval to conduct plant-related activities in the area. An administrative process will be used to evaluate, approve (or deny), and document all plant related activities conducted in these open land areas during and following final status surveys. Land areas will be inspected quarterly and any material that has been deposited since the last inspection will be investigated.

5.4.6.3 Excavation Land Areas Resulting from Radiological Remediation These are land areas where there has been excavation for the purpose of radiological remediation of the soil. These areas will be posted with signs providing contact information for approval to conduct decommissioning and demolition activities in the area. An administrative process will be used to evaluate, approve (or deny), and document all plant related activities conducted in these excavations during and following final status surveys.

5.4.6.4 Bedrock There are areas of the site where bedrock will be exposed as a result of building demolition and soil remediation. These areas include, but are not limited to, the Tank Farm area, the Spent Resin Facility and Ion Exchange Facility, and the RHR pit area of the Primary Auxiliary Building. Isolation and control of bedrock areas will be the same as for open land areas with added controls for deep excavation personnel safety requirements.

5.4.6.5 Excavations Resulting from the Removal of Piping Conduit Areas that are excavated for the purposes of removing piping, conduit or other subsurface construction will be controlled to ensure personnel safety and to reduce the potential for plant-related activities to contaminate the area. These areas will be posted with signs providing contact information for approval to conduct decommissioning and demolition activities in the area. An administrative process will be used to evaluate and approve (or deny), and document all plant-related activities conducted in these excavations.

Any isolation and control measures needed at the restored surface will be implemented to protect the area from contamination.

August 2004 5-12 Rev. 2

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/l 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, concrete debris, building surfaces, building foundations/basements and activated concrete. These values will be used to establish DCGLs for survey units in cases where measurements are made that are not radionuclide specific or when hard-to-detect radionuclides are present that necessitate the need for a surrogate radionuclide. In such cases, appropriate 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).

Chapter 6 of this plan establishes the basis for the DCGLs in soil and sediments to be the resident farmer scenario, and the DCGLs for structures to be the building occupancy scenario. For structures, an additional scenario was evaluated for buried concrete debris and foundations/basements. This is described in Section 6.8.

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 both surface radioactivity bounded by the resident farmer - concrete debris scenario and resident farmer - soil scenario. 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 structures and the potential burial of subsurface concrete structures and possibly concrete debris containing residual radioactivity. The areas of the site where these exposures could occur concurrently are where subsurface structures and concrete are buried and existing August 2004 5-13 Rev. 2

Haddam Neck Plant License Termination Plan groundwater contamination may be present. This area represents approximately 15,600 m2 and includes the industrial area of the site. For this area, the total dose from these sources, H~ota, can be expressed as:

HTotal = Hsoil+ H~singG;V +HFutureGIV (Equation 5-1)

For these individual media, the dose from the residual radioactivity from radionuclide i is:

C' H' = 25

• CL (Equation 5-2)

DCGL!

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:

Hi = 25* DCGLp (Equation 5-3)

DCGL' .

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:

August 2004 5-14 Rev. 2

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.

In the case of potential future groundwater contamination from buried site building subsurface foundations/basements or concrete debris containing residual radioactivity, the dose to the resident farmer is limited to the water dependent pathways from the buried concrete debris scenario as described in Chapter 6. Therefore for the concrete debris case, the dose component for each radionuclide i is modified by the fraction of the total dose delivered from the water dependent pathway, f, as calculated using the information in Appendix G, Table G-4. This is provided as:

HFutureGV = 25*fi DCGLOP-ConcreteDebrjs (Equation 5-4) 4 f

DCG Oase-ConretDebris The HExistingGW 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 plume 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 HEistingGw term would currently be applied. Table 5-3 is based upon the groundwater plume (locations where tritium concentration was measured/predicted as Augut 204 515 Rv.I August 2004 5-15 Rev. 2

Haddam Neck Plant License Termination Plan Table 5-3 provides the survey areas to which the HExistingGw term would currently be applied. Table 5-3 is based upon the groundwater plume (locations where tritium concentration was measured/predicted as 1000 pCi/liter or greater) as depicted in the 1999 Malcolm Pirnie Groundwater Monitoring Report, including a plume influence boundary distance of 100 meters (Figure 5-3).

It is noted, however, that characterization efforts for groundwater contamination are still ongoing and the survey areas to which the HEXistingGW term are applied may change. Those changes will be communicated to the NRC. This change may be caused by changes in the location of the plume, detection of groundwvater contamination at locations outside the plume or changes to the plume influence boundary distance. 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 based on data collected as part of the Phase 2 Hydogeological Work Plan, to better define the plume influence boundary distance. 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 will use the hydrogeological conditions and parameters assumed in the Resident Farmer Scenario as described in Chapter 6 of the LTP. If this capture zone analysis (using the well pumping rate assumed in the dose calculations presented in Chapter 6) determines that the maximum capture zone radius is greater than 100 meters, the NRC will be notified.

Table 5-3 Survey Areas Affected by Groundwater Contamination Survey Area 9102 9120 9312 9104 9126 9313 9106 9128 9502 9108 9302 9512 9110 9304 9514 9112 9306 9518 9114 9307 9520 9116 9308 9522 9118 9310 The compliance formulation for these resident farmer exposure scenarios is re-written as:

DCGOLP'Soil DCGEOP-ExistingG;V I D+L"a-Cncre.eDebris (Equation 5-5)

Base-Soil Base-ExistingG DCGLase-ConceteDebris For simplicity Equation 5-5 may be re-written as:

August 2004 5-16 Rev. 2

Haddam Neck Plant License Termination Plan 1 Ž f~,SN + fatsuingGIV + ff on^reeDbr (Equation 5-6) where, for a given radionuclide i, fslj is the fraction of the total dose from soil, fAAisngG1' is the fraction of the total dose from existing contamination in groundwater, f.fC1oncreteDcbris is the fraction of the total dose from the water dependent pathways from subsurface foundation/basement or concrete debris.

The use of this equation requires that only one variable be unknown. Therefore, values for fs5 j,and fExjitijgGw will need to be known or selected in order to calculate the building surface operational DCGL.

The final selection of the building surface operational DCGL will be the lower of the DCGLs determined from Equation 5-5 or the building occupancy DCGL from Chapter 6. For areas of surface contamination, this comparison will be performed using the surface contamination building occupancy DCGLs, listed in Table 6-3, and concrete debris DCGLs in units of dpm/100 cm2 given in Table 6-5. For areas that are volumetrically contaminated, this comparison will be performed using the volumetric contamination building occupancy DCGLs, as listed in Table 6-3, and concrete debris DCGLs in units of pCi/g, as listed in Table 6-4. Like the base-case DCGLs, the operational DCGL will be different for each nuclide.

The following example is provided to illustrate the use of the operational DCGLs for surficial contamination with the following assumptions:

• fso = 0.3

• fexistngGov= 0.2

• ThereforefwfconcelDcbh,= 0.5 Using these values, the operational DCGLs for the buried building debris are calculated as 0.5 of the base-case concrete debris values from Chapter 6 and compared to the building occupancy DCGLs also in Chapter 6. In this final comparison, the more restrictive DCGL will be used as the operational DCGL for building surfaces.

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 an example of the building surface operational DCGL for Cs-I 37 using the fractional values from above.

Table 5-4 Operational DCGL Example for Cs-137 Using Fractional Values from Above and Assumingf, = 0.354 Base Case DCGL Operational DCGL Soil (pCilg) 7.91E+00 2.37E+00 Existing Groundwater (pCi/I) 4.3 1E+02 8.62E+01 Concrete Debris' (dpm/100cm 2) 8.31E+06 1.17E+07' Building Occupancy (dpm/100cm 2) 4.30E+04 Selected Building Surface Operational DCGL L4. .. 4  :; 30E+04 August 2004 5-17 Rev. 2

Haddam Neck Plant License Termination Plan CG~~s 0.5 x8.3 1E +06 0C1 lDCGL~o2pncreteDbu =i

• 0.354 1.17E+O7dpm/lOOem 2 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 = 1 (Equation 5-7)

ZDCGL, 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 IOCFR20 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.

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 A ___._.^nhAff 10 no. -

August ZUU4 D- I Z Kev.2i

Haddam Neck Plant License Termination Plan samples will be analyzed for ETD and HTD radionuclides using gross alpha, alpha spectroscopy, gross beta analysis or gamma spectroscopy techniques.

• Screen HTD radionuclides using the 5% and 10% rule described in Section 5.4.7.2.

Radionuclides not screened out will require a surrogate DCGL. Surrogate relationships will be determined from the samples results using one of methods described below.

• Develop a surrogate relationship for each HTD radionuclide.

DCGLsurrogate=DCGLETDx DCGLTTD (filTD: ETD x DCGLETD) + DCGLtITD (Equation 5-8)

• Determine the average surrogate DCGL and the standard deviation from the surrogate relationships.

If the %CV (coefficient of variation) of the average surrogate DCGL is within 25% then the average surrogate DCGL will be applied to the survey area. The %CV is the percent ratio of the standard deviation to the average surrogate DCGL. If this criterion is not met, the following steps will be applied.

- Following a more detailed spatial analysis of the radionuclide mix distribution, the unit may be subdivided into separate survey units based on the spatial distribution.

- The lowest surrogate DCGL from the observed radionuclide mix may be applied to the entire survey unit.

- Additional samples may be collected and analyzed to allow for a detailed analysis and documented evaluation of the radionuclide distribution resulting in the use of a specific DCGL for the survey unit.

• The surrogate DCGL may be computed from a simple recurrence formula as:

CETD CErD Cl C2 Ci DCGLsurrogate DCGLETD DCGLi DCGL2 DCGLi Equation (5-9) or, for simplification CE CE Cl C2 Ci

-- +-+- +..+-

Dsurrogate DE Di D2 Di Equation (S-JO) where:

DE E the DCGL for the easy-to-detect radionuclide DI }the DCGL for the first hard-to-detect radionuclide August 2004 5-19 Rev. 2

Haddam Neck Plant License Termination Plan D2 - the DCGL for the second hard-to-detect radionuclide Di -the DCGL for the ith hard-to-detect radionuclide f, the activity ratio of the first hard-to-detect radionuclide to the easy-to-detect radionuclide f2 -the activity ratio of the second hard-to-detect radionuclide to the easy-to-detect radionuclide f --the activity ratio of the ith hard-to-detect radionuclide to the easy-to-detect radionuclide Consider the case of three HTD radionuclides from which a surrogate will be calculated.

DCGLsufoat- (DED 1DD3)

Surrogate = (D 1D2D3)+(fiDED2D3)+(f2DED ID3)+(fMDED 0D2)

Example 5-1 A general expression for the surrogate equation based on recursive relationships is provided by Equation 5.11 for n HTD radionuclides.

DCGLsurrogate = Equation 5.11

/ E + E fi ID; 11D qaio .

i=l 5.4.7.4 Elevated Measurement Comparison (EMC) DCGLs The DCGL established for the average residual contamination in a survey unit is DCGLW. Values of the DCGLW may be scaled through the use of area factors to obtain a DCGL that represents the same dose to an individual from residual contamination over a smaller area within a survey unit. Such a value is called DCGLEMC, where the subscript EMC stands for elevated measurement comparison. The DCGLEMC is defined as the product of the applicable DCGLW and a correction factor know as the area factor.

The area factor is equal to the ratio of the dose from the base-case contaminated area to the dose from a smaller contaminated area with the same radioactive source concentration. Area factors are required for both the resident farmer and the building occupancy scenarios. Area factors for both the resident farmer and building occupancy scenarios have been calculated (Reference 5-7) for the radionuclides of concern at the HNP site considering all applicable potential pathways of exposure.

For the resident farmer scenario, RESRAD (Version 5.91) was used to determine area factors. For the building occupancy scenario, RESRAD-BUILD (Version 2.37) was used to determine area factors. Area factors will not be computed for areas smaller than I m2 for either the resident farmer or the building occupancy scenarios.

Table 5-5 summarizes the outputs of the RESRAD code for the radionuclides of concern for the Resident Farmer scenario. Table 5-6 summarizes the outputs of the RESRAD-BUILD code for the Building Occupancy scenario.

Aiimict "E"a t)OA

-- vv J-WV A

2() Ie

-\e

Haddam Neck Plant License Termination Plan Table 5-6 Area Factors for the Building Occupancy Scenario Size of Elevated Area (mL2 100 75 50 25 10 8 6 4 2 1 H-3 1.0 1.3 2.0 4.0 10.0 12.5 16.6 25.0 50.1 100.0 C-14 1.0 1.3 2.0 4.0 10.0 12.5 16.7 25.0 50.1 100.0 Mn-54 1.0 1.1 1.2 1.6 2.4 2.8 3.3 4.2 7.1 12.6 Fe-55 1.0 1.3 2.0 4.0 10.0 12.5 16.7 25.1 49.9 100.0 Co-60 1.0 1. 1 1.2 1.6 2.5 2.8 3.3 4.3 7.1 12.7 Ni-63 1.0 1.3 2.0 4.0 10.0 12.5 16.7 25.0 50.2 100.0 Sr-90 1.0 1.3 1.9 3.3 7.0 8.4 10.6 14.8 27.0 50.5 Nb-94 1.0 1.1 1.2 1.6 2.4 2.8 3.3 4.3 7.1 12.6 Tc-99 1.0 1.3 2.0 4.0 10.0 12.5 16.7 25.1 50.2 100.0 Ag-108m 1.0 1.1 1.2 1.6 2.4 2.8 3.3 4.3 7.1 12.7 Cs-134 1.0 1.1 1.3 1.6 2.5 2.8 3.4 4.4 7.3 13.1 Cs-137 1.0 1.1 1.3 1.7 2.6 2.9 3.5 4.5 7.5 13.4 Eu- 152 1.0 1.1 1.2 1.6 2.5 2.8 3.3 4.3 7.2 12.7 Eu- 154 1.0 1.1 1.2 1.6 2.4 2.8 3.3 4.3 7.1 12.7 Eu-155 1.0 1.1 1.3 1.7 2.6 2.9 3.5 4.5 7.5 13.4 Pu-238 1.0 1.3 2.0 4.0 10.0 12.5 16.6 24.9 49.7 99.5 Pu-239 1.0 1.3 2.0 4.0 10.0 12.5 16.7 24.9 49.8 99.5 Pu-241 1.0 1.3 2.0 4.0 10.0 12.5 16.7 25.0 50.1 100.0 Am-241 1.0 1.3 2.0 3.9 9.7 12.1 16.0 23.8 47.2 93.7 Cm-243 1.0 1.3 1.9 3.8 9.0 11.0 14.5 21.2 40.8 79.6 August 2004 5-22 Rev. 2 I

Haddam Neck Plant License Termination Plan 5.4.7.5 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 I mrem/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 simplify the analysis, the piping material is assumed to be 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 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 I 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 I mrem/yr. Finally, the volumetric contamination is converted to surface contamination, assuming a 4-inch diameter (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.

August 2004 5-23 Rev. 2

Haddam Neck Plant License Termination Plan Table 5-7 Release Limits For Remaining Buried Piping Radionuclide Surface Limit Resulting in I mrem/yr Dose (dpm/lOOcm2 )

H-3 5.21 E+03 C-14 7.77E+04 Mn-54 5.31 E+04 Fe-55 6.17E+04 Co-60 3.21 E+05 Ni-63 1.52E+05 Sr-90 1.87E+02 Nb-94 1.37E+05 Tc-99 2.44E+04 Ag-108m 1.37E+06 Cs- 134 8.35E+04 Cs-137 9.66E+04 Eu-152 2.68E+05 Eu- 154 1.87E+05 Eu-155 1.20E+06 Pu-238 7.50E+02 Pu-239 6.82E+02 Pu-241 1.14E+04 Am-241 3.33E+02 Cm-243 4.61 E+02 Embedded pipe represents medium- to 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.

August 2004 5-24 Rev. 2 I

Haddam Neck Plant License Termination Plan

• 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 sample 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 inhalation and ingestion that is twice 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 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 potential for contamination as indicated by its classification. Class 3 survey units receive judgmental August 2004 5-25 Rev. 2 1

Haddam Neck Plant License Termination Plan 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 I 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:

• For Class I survey units, 100% of the surface will be scanned;

• 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 I 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 I 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 2004 5-26 Rev. 2

Haddam Neck Plant License Termination Plan These same criteria will be used when deciding if advanced survey techniques can be used in place of fixed measurements and traditional scans for a given survey unit. MDCs for advanced techniques will be computed using background count rates obtained using appropriate reference materials.

With respect to the survey methods and techniques discussed above, the survey design criteria that will be employed for final status surveys for the HNP site are summarized below. Note that "fixed measurements" is used interchangeably to refer to measurements or samples taken at specific locations.

• For Class I or Class 2 survey units, advanced survey technologies may be used exclusively only in survey units for which the above coverage requirements can be achieved and MDCs are no greater than 50% of the applicable DCGLw.

• For Class 1 or Class 2 survey units for which advanced technologies would have an acceptable MDC, but the above coverage requirements cannot be achieved, advanced technologies may be used over 100% of the accessible area with a combination of fixed measurements and traditional scans used over the remainder of the area as specified in Section 5.5.1 and subsequent subsections of this plan.

For any survey units for which advanced survey techniques are impractical, fixed measurements and traditional scans will be used exclusively in accordance with this plan.

5.5.1 Selecting the Number of Fixed Measurements and Locations The MARSSIM methodology for evaluating whether a survey unit meets its applicable release criterion using fixed measurements plus scans is based on using non-parametric statistical tests for data assessment. Specifically, the methods of MARSSIM are based on two non-parametric tests: the Wilcoxon Rank Sum (WRS) test and the Sign test, as discussed in Section 5.8 Selection of the required minimum number of data points depends on which statistical test is going to be used to evaluate the data, and thus depends on what type of measurements are to be made (gross measurement, net measurement or radionuclide specific) and if the radionuclide(s) of interest appear(s) in background.

5.5.1.1 Establishing Acceptable Decision Error Rates One input to the process of selecting the required number of data points for a given survey, which does not depend on the statistical test applied, is the selection of the acceptable decision error rates. Decision errors refer to making false decisions by either rejecting a null hypothesis when it is true (a Type I error) or accepting a null hypothesis when it is false (a Type II error). With respect to final status surveys, the null hypothesis is that the survey unit of interest contains residual contamination in excess of the applicable release criterion. Thus, a Type I error refers to concluding that an area meets the release criteria when in fact it does not. The probability of making a Type I error is referred to as alpha (a).

Likewise, a Type 1I error refers to concluding a unit does not meet the release criteria when it actually does. The probability of making a Type II error is denoted beta (f3). Selecting values of a or P that are too low will result in an excessive number of fixed measurements being required. Likewise, selecting a P value that is too large can result in excessive costs in that survey units that meet the release criterion could be subjected to superfluous remediation efforts. Under the current regulatory models, an a value that is too large equates to greater risk to the public in that there is a greater chance of releasing a survey unit that does not meet the release criterion.

August 2004 5-27 Rev. 2

Haddam Neck Plant License Termination Plan NRC draft regulatory guide DG-4006 (as incorporated in Section 7.2 of Appendix E to NUREG-1727) recommends that the a decision error rate be set to 0.05 (5%) and that "any value of f3 is acceptable to the NRC." Thus, decision error rates for final status surveys designed for the HNP site will be set as follows:

• the a value will always be set to 0.05 unless prior NRC approval is granted for using a less restrictive value;

• the P value is nominally set to 0.05, but may be changed if it is found that more fixed measurements than necessary are being made to demonstrate compliance with the release criterion.

5.5.1.2 Determining the Relative Shift Another input to the process of selecting the required number of measurements that is somewhat independent of the statistical test to be employed is the determination of what is called the relative shift.

The relative shift is a parameter that quantifies the concentrations to be measured in a survey unit relative to the variability in these measurements. The relative shift is a function of the DCGLW, a parameter called the "Lower Bound of the Gray Region" (LBGR), and either the expected standard deviation of the measurements to be made in the survey unit (<s.) or the standard deviation established for the corresponding reference area (6,). The choice of a, or A, depends on whether the survey data are to be evaluated against a reference area(s). Reference areas are used if the WRS test is applied or, where gross measurements are to be background subtracted, the Sign test may be used. If a reference area is required, the larger of the values of a, or a, is used. The a, values will be selected by:

• using existing characterization or remediation support survey data or

• making preliminary measurements.

Values of a, will be computed using data collected from measurements in reference areas or from reference materials, as appropriate.

Given that a, and a, values should reflect a combination of the spatial variability in the concentration and the precision in the method of measurement, these values will be selected based on existing survey data only when the existing measurements were made using techniques equivalent to those to be used during the final status survey.

The LBGR represents the concentration to which the survey unit must be cleaned (decontaminated) in order to have an acceptable probability of passing the statistical test. The difference between the DCGLw and the LBGR, known as the shift, can be thought of as a measure of the resolution of the measurements that will be made in a survey unit. The shift is denoted as A.

The relative shift (A/cl) is computed as the quotient of the shift and the appropriate standard deviation values. If no reference area data are needed to evaluate the survey results, the expected standard deviation of the measurements (a.) is used. If a reference area is required, the larger of the values of as or cr, is used.

To compute the relative shift, the appropriate sigma value and an initial LBGR are selected. The initial value for LBGR will be based upon site specific information, if available; otherwise, per MARSSIM and DG-4006 (as incorporated in Section 7.1 of Appendix E to NUREG-I 727) the initial value for the LBGR will be set to one-half of the DCGLw. If the resulting relative shift is not in the range of 1.0 and 3.0, the August 2004 5-28 Rev. 2

Haddam Neck Plant License Termination Plan LBGR is adjusted until it is. If the relative shift is too low, the LBGR is decreased; and if the relative shift is too high, the LBGR is increased.

5.5.1.3 Selecting the Required Number of Measurements for the CURS Test The minimum number of fixed measurements required when the WRS is computed by the following equation:

N! (Zia +ZI-)2 (Equation 5-12) 2 3(P,-o.5) 2 where: N the minimum number of measurements required for each survey area or reference area; Zi.a the percentile represented by the a decision error; ZI-p the percentile represented by the P decision error; and Pr the probability that a random measurement from the survey unit exceeds a random measurement from the reference area by less than the DCGLW when the survey unit median is equal to the LBGR concentration above background.

Values of Pr, Z1, and Z,.p will be taken from Tables 5.1 and 5.2 of MARSSIM. P. is a function of the relative shift, and Zt c, and ZI.p depend on the selected values for a and P.

The value of N computed for the WRS test applies for both the survey unit and the reference area (i.e., at least N measurements should be performed in both areas). To ensure against lost or unusable data, the value of N will be increased by at least a factor of 1.2 when assigning the number of measurements to be made.

5.5.1.4 Selecting the Required Number of Measurements for the Sign Test The minimum number of fixed measurements required when the Sign test is computed by the following equation:

N__ (Z1- +z,- 1,)2 (Equation 5-13) 4(Sign P _ O.5) 2 I

where: N the minimum number of measurements required; Zt-a the percentile represented by the ax decision error; ZN03 the percentile represented by the P3 decision error; and Sign p the probability that a random measurement from the survey unit will be less than the DCGLW when the survey unit median concentration is equal to the LBGR.

Values for Sign p will be taken from Table 5-4 of MARSSIM.

To ensure against lost or unusable data, the value of N will be increased by at least a factor of 1.2 when assigning the number of measurements to be made.

August 2004 5-29 Rev. 2

Haddam Neck Plant License Termination Plan 5.5.1.5 Assessing the Need for Additional Measurements in Class I Survey Units Given the potential for small areas of elevated activity in Class I survey units, evaluations must be performed to assess the potential for missing such areas while scanning in locations not covered by fixed measurements. This evaluation, referred to as the Elevated Measurement Comparison (EMC), is performed by comparing the MDC of the scanning technique to the DCGLEMC for the survey unit of interest. If the scanning MDC is larger than the DCGLENic, additional measurements may be required beyond the minimum number computed via Equation 5-9 or 5-10. The effect of these additional measurement points is to tighten the grid spacing for the fixed measurements, thus reducing the probability of missing a small area of elevated activity to an acceptable level.

The adequacy of the scanning technique will be evaluated by calculating a scanning MDC, expressed as a fraction of the DCGLENC as shown below.

As described in Section 5.4.7.4, the relationship between the DCGLENMC and the DCGLw using the area factor for nuclide i is:

DCGL'AC = AF'DCGL'v (Equation 5-14)

Where: AF' is the area factor for radionuclide i.

For soil, the relationship between a scanning minimum detectable count rate (MDCR) and the minimum detectable soil concentration is:

MDC'(pCi/g)= MDCR(cpin) (Equation 5-15)

E (cpm / pCi/g)

Where: El is the conversion factor (in cpmlpCi/g) for the radionuclide i (instrument efficiency for scanning)

The soil scanning MDC expressed as a fraction of the DCGLEMC is calculated by the following equation:

MDC(IDCGLEfC) = MDCRE f --MDCRI f (Equation 5-16)

ECE'DCGLEAC E'AF'DCGL!,v Where: fl is the decimal fraction of the radionuclide mix comprised by radionuclide i and is based upon characterization data, as a part of the Final Status Survey.

An example calculation to determine the soil scanning MDC expressed as a fraction of the DCGLENIC when multiple radionuclides are present is shown as follows:

August 2004 5-30 Rev. 2

Haddam Neck Plant License Termination Plan Assumptions:

Two radionuclides are present; Cs-137 and Co-60 Cs-137 fraction in mix (f) = 0.75 Co-60 fraction in mix (f) = 0.25 Cs-137 efficiency (E) = 228 cpm/pCi/g Co-60 efficiency (E) = 882 cpm/pCi/g Elevated area = 100 m2 Cs-137 area factor (AF) from Table 5-5 = 2.93 Co-60 area factor (AF) from Table 5-5 = 1.41 Cs-137 DCGLw from Table 6-1 = 7.91 pCi/g Co-60 DCGLW from Table 6-1= 3.81 pCi/g MDCR = 2,000 cpm MDC(JDCGLEAIfC) = 2,000F 0.75 + 0.25 10.4 L(2 2 8 )( 2 .9 3 )( 7 .9 1) (882)(1 .41)(3.81) ]

For scanning building surfaces, the following equation from MARSSIM provides the method to calculate the MDC for beta-gamma measurements. It has been repeated here below for clarity:

MDC(dpn/lOOccm) )= I.B A (Equation 5-17)

V _t100 where t is the time the detector spends over a source of radionuclide i which can be related to the travel velocity of the probe, V(cm/min), and the minimum dimension of the detector, L (cm), as:

t(min) = L(cm) (Equation 5-18)

I V(c / mmin)

Equation 5-14 can be rewritten as follows:

1.38 2 1.38 Rb 1.38FR, MDC (dpn/l10cm2 ) = LVb I (Equation 5-19)

-loo)(1P August 2004 5-31 Rev. 2 I

Haddam Neck Plant License Termination Plan Substituting equation 5-15 into 5-16 gives:

MDC'(dpm IIOOcm2 ) = 1.38JR (Equation 5-20)

(A°° )

In accordance with MARSSIM, the MDCR for an analog detector with an audible signal (for d'=1 .38) is:

I 38/i~ 1.384f 1.38+/f MDCR(cpm) = - - 3 8X (Equation 5-21)

TV I

Using this, equation 5-20 is re-written as:

MDC'(dpm/lOOcm2 )= MDCR (Equation 5-22)

To allow for multiple radionuclides, the scan MDC expressed as a fraction of the DCGLEMC is:

MDC(./DCGLEAIC) = -DRY P' (Equation 5-23) I NFPA ce'eDCGLVE!

By substituting DCGLA!aC = AF'DCGL!. into Equation 5-23 yields the building surface scanning MDC I equation expressed as a fraction of the DCGLEMC:

MDC(J/DCGLEAic) =MDR2 (Equation 5-24) I

'r-(

A c'AF'DCGLV An example calculation to determine the building surface scanning MDC expressed as a fraction of the DCGLEMC when multiple radionuclides are present is shown below:

August 2004 5-32 Rev. 2 I

Haddam Neck Plant License Termination Plan Assumptions:

Two radionuclides are present; Cs-137 and Co-60 Cs-137 fraction in mix (f) = 0.75 Co-60 fraction in mix (f) = 0.25 Probe width (L) = 10.2 cm (4 inches)

Scan rate (V) =305 cm/min (2 inches/sec)

Background count rate (Rb) = 200 cpm

£; = 0.3 for Co-60

£;=0.38 forCs-137 c, = 0.25 for Co-60 es = 0.5 for Cs-137 Surveyor Efficiency, p=0.5 Probe area (A) = 100 cm 2 MDCR = 107 cpm Elevated area =10 m2 Cs-137 area factor (AF) from Table 5-6 = 2.6 Co-60 area factor (AF) from Table 5-6 = 2.5 Cs-137 DCGLw from Table 6-3 = 4.30E+04 dpm/100 cm2 Co-60 DCGLw from Table 6-3= 1.11 E+04 dpm/1 00 cm2 MDC(IDCGLEC) = 107 E-( 100o

[ (0.38)(0.5)(2.6)(4.30E4) 0.75 + 0.25 (0.3)(0.25)(2.5)(1.1 1E4)-

1 = 0.02

• 100)

As shown in these two examples, the fraction of DCGLE1,C is less than one. Therefore no additional measurements are required.

If the value of MDC (fDCGLEMC) is greater than one, additional measurements may need to be taken in the survey unit as determined by taking the following steps.

Determine the size of the elevated area from Table 5-5 or Table 5-6 corresponding to the highest fDCGLEMC which is still less than one. That area is denoted as AENSC-The number of measurements (NEOWC) required to detect an area of elevated concentration equal to AEMC is then computed as NEAIC = A (Equation 5-25) where A is the total area of the survey unit. NEMC (computed via Equation 5-25) is then compared to N, the number of fixed measurement points computed via Equation 5-12 or 5-13. The larger of NEMC or N is then used as the requisite number of fixed measurement locations and to compute the grid spacing.

Aunust

• *t, 2004 - 5-331 Rev.

.\_ ..2

Haddam Neck Plant License Termination Plan 5.5.1.6 Determining Measurement Locations For Class I and Class 2 survey units, fixed measurements will be performed over a systematic measurement pattern consisting of a grid having either a triangular or a square pitch. The pitch (grid spacing) will be determined based on the number of measurements required and whether the desired grid is triangular or square.

Systematic grids will not be used for surveys involving fixed measurements for Class 3 units. Instead, fixed measurement locations will be selected at random throughout the survey unit area by generating pairs of random numbers between zero and one. One pair of random numbers will be generated for each fixed measurement to be made. The random number pairs, representing (x, y) coordinates, will be multiplied by the maximum length and width dimensions of the survey unit to yield the location for each fixed measurement. For odd-shaped survey units, a rectangular area encompassing the survey unit will be used to establish the maximum length and width. A new pair of random numbers will be generated if any of them give locations that are not actually within the survey unit boundaries. New pairs of numbers will also be generated in cases where a measurement cannot be made at a specific location because of an obstruction, inaccessibility, etc.

The spacing to be used in setting up the systematic grid used to establish fixed measurement locations for Class I and Class 2 areas will be computed as L= A6 for a triangular grid, or (Equation 5-26)

L= for a square grid (Equation 5-27) where L = grid spacing (dimension is square root of the area),

A = the total area of the survey unit, and N = the desired number of measurements.

In the case of Class I units, the value used for N in Equations 5-26 and 5-27 should be the larger of that from Equations 5-12 or 5-13 (if the scan MDC is sufficient to see small areas of elevated activity) or Equation 5-25. In all cases, the value of N should include additional measurements required to ensure against losses or unusable data.

Once the grid spacing is established, a random starting point will be established for the survey pattern using the same method as described above for selecting random locations for Class 3 units. Starting from this randomly-selected location, a row of points will then be established parallel to one of the survey unit axes at intervals of L. Additional rows will then be added parallel to the first row. For a triangular grid, additional rows will be added at a spacing of 0.866L from the first row, with points on alternate rows spaced mid-way between the points from the previous row. For a square grid, points and rows will be spaced at intervals of L. Section 5.5.2.5 of MARSSIM describes the process to be used for selecting fixed measurement locations and provides examples of how to establish both a systematic grid and random measurement locations.

Software tools that accomplish the necessary grid spacing, including random starting points and triangular or square pitch, may be employed during Final Status Survey. When available, this software will be used August 2004 5-34 Rev. 2

Haddam Neck Plant License Termination Plan with suitable mapping programs to determine coordinates for a global positioning system (GPS). The use of these tools will provide a reliable process for determining, locating and mapping measurement locations in open land areas separated by large distances and will be helpful during independent verification.

5.5.2 Judgmental Assessments For those Class 2 and Class 3 survey units for which 100% of the area is not surveyed, it is important to consider performing judgmental assessments to augment any regimented measurements made in accordance with the above guidance. Such assessments may consist of biased sampling or measurements performed in locations selected on the basis of site knowledge and professional judgment. Judgmental assessments serve to provide added assurance that residual contamination at the site has been adequately located and characterized.

In addition to any judgmental measurements deemed necessary to provide comprehensive survey coverage for a given survey unit, the survey process should include an isotopic mix evaluation in cases where measurable activity still exists. Doing so will allow an assessment of the adequacy of the DCGLw selected for the survey unit in question to be made during the subsequent data assessment phase. For gross count measurements (i.e., not radionuclide specific), radionuclide mix information will also allow for an evaluation of the suitability of the efficiencies applied in converting raw count data to activity.

The basis for judgmental assessments will be documented in the Final Status Survey Report and will receive a technical review in accordance with plant procedures.

5.5.3 Data Investigations 5.5.3.1 Investigation Levels An important aspect of the final status survey is the selection and implementation of investigation levels.

Investigation levels are levels of radioactivity used to indicate when additional investigations may be necessary. Investigation levels also serve as a quality control check to determine when a measurement process begins to deviate from expected norms. For example, a measurement that exceeds an investigation level may indicate a failing instrument or an improper measurement. However, in general, investigation levels are used to confirm that survey units have been properly classified.

When an investigation level is exceeded, the first step is to confirm that the initial measurement/sample actually exceeds the particular investigation level. Depending on the results of the investigation actions, the survey unit may subsequently require reclassification, remediation, and/or resurvey. Investigation levels are established for each class of survey unit. The investigation levels (criteria), to be employed for the HNP final status survey effort, are given in Table 5-8.

August 2004 5-35 Rev. 2

Haddam Neck Plant License Termination Plan Table 5-8 Investigation Levels Survey Unit For fixed measurements or samples, For scan measurements, perform Classification perform investigation if: investigation if:

Class I > DCGLE,%fC or > DCGLW and a > DCGLEMC statistical outlier.

Class 2 > DCGLw > DCGLw or > MDCSC..n if MDCSCfl is l__ greater than the DCGLw Class 3 > 0.5 x DCGLw Detectable over background.

For Class 1 survey units, measurements above the DCGLW are not necessarily unexpected. However, such a result may still indicate a need for further investigation if it is significantly different than the other measurements made within the same survey unit. Thus, some additional evaluation criterion is needed to assess if results from fixed measurements or samples in a Class 1 survey unit that exceed the DCGLw warrant further attention. Measurements in Class I survey units that exceed the DCGLW and differ from the mean of the remaining measurements by more than three standard deviations will therefore be investigated. Measurements in Class I units that exceed the DCGLw, but do not differ from the mean by as much may still be investigated on the basis of professional judgment, as may any measurements that differ significantly from the rest of the measurements made within a given survey unit.

In Class 2 or Class 3 areas, neither measurements above the DCGLW nor areas of elevated activity are expected. Thus, any fixed measurements or sampling results that exceed the DCGLW in these areas will be investigated. In the case of Class 3 areas, where any residual radioactivity would be unexpected, fixed measurement or sample results that are greater than 0.5 x DCGLw will be investigated. Because the survey design for Class 2 and Class 3 survey units is not driven by the elevated measurement comparison, any indication of residual radioactivity in excess of the DCGLW during the scan of a Class 2 unit will warrant further investigation. For Class 3 units, any scan measurement that shows a positive indication over background will be investigated.

In cases where an advanced survey method is used instead of fixed measurements or samples, the investigation levels given in Table 5-8 for fixed measurements or samples will be applied with the exception of the statistical outlier test for measurements in Class I survey units. In cases where advanced survey methods are used as a means of traditional scanning, the investigation levels for scan measurements in Table 5-8 will be used.

5.5.3.2 Investigations Locations where initial measurements give results that exceed an applicable investigation level will be identified for confirmatory measurements. If it is confirmed that residual activity exists in excess of the investigation level, additional measurements will be made to determine the extent of the area of elevated activity and to provide reasonable assurance that other areas of elevated activity do not exist. Potential sources of the elevated activity will be postulated and evaluated against the original classification of the survey unit and its associated characterization data. The possibility of the source of the elevated activity having affected other adjacent or nearby survey units will also be evaluated. Documentation will be compiled containing the results from the investigation surveys and showing any areas where residual activity was confirmed to be in excess of the investigation level. If residual activity in excess of the August 2004 5-36 Rev. 2

Haddam Neck Plant License Termination Plan applicable investigation level is confirmed, the documentation will also address the potential source(s) of the activity and the impact this has on the original classification assigned to the survey unit. A decision will then be made regarding re-classification of the unit in whole or in part.

5.5.3.3 Remediation "Remediation" in the context of the LTP is intended to mean activities performed to meet the criteria of IOCFR20, Subpart E. Activities to remove materials may be performed for other reasons, and thus are not considered to be "remediation." If during the time of Final Status Survey, any areas of residual activity found to be in excess of the DCGLEMC, they will be remediated to reduce the activity to acceptable levels. Areas of residual activity may also need to be remediated to meet the ALARA criterion.

Remediation actions are discussed in Chapter 4 and documented as described in Section 5.9.

5.5.3.4 Re-classification The decision to reclassify an area, or part of an area, is made following a review of the basis for the original classification, considering the evaluation process outlined in Section 5.5.3.2 (consistent with MARSSIM). This process includes sufficient additional measurements to confirm the residual contamination, determine the nature and extent of the contamination present, provide assurance that other areas of elevated activity do not exist within the survey unit, and evaluate the impact (if any) of the affected area on nearby survey units. The results of these measurements will be evaluated, and the area, or part of the area, will be reclassified and resurveyed per Section 5.5.3.5 in a manner that is consistent with the process described in MARSSIM. Additionally, if required remediation actions are taken in the area, it will be resurveyed per Section 5.5.3.5 in a manner that is consistent with the process described in MARSSIM. Re-classification of areas from a less to a more restrictive classification may be done without prior NRC approval; however, re-classification to a less restrictive classification would require prior NRC approval.

5.5.3.5 Re-survey If a survey unit is re-classified (in whole or in part), or if remediation is performed within a unit, then the areas affected are subject to re-survey. Any re-surveys will be designed and performed as specified in this plan based on the appropriate classification of the survey unit. That is, if a survey unit is re-classified or a new survey unit is created, the survey design will be based on the new classification.

For example, a Class 3 area that is subdivided due to the unexpected presence of radioactivity will be divided into at least two areas. One of these may remain as a Class 3 area while the other may be a Class 2 area. In order to maintain the survey design Type I and Type 11 decision error rates in the Class 3 area, additional measurements may be required to be performed at randomly selected locations until the required total number of measurements is met (see Section 5.5.1). The new sub-divided Class 2 survey area will then be surveyed using a new survey design. The Type I and II decision error rates used are documented in the final status survey documentation.

A Class 2 area that is subdivided due to the levels of radioactivity identified will be divided into at least two areas as well. In this case if the original survey design criteria has been satisfied, no additional action is required, otherwise the remaining Class 2 survey unit will be redesigned. The new sub-divided survey unit will be surveyed against a new survey design.

If remediation is required in only a small area of a Class I survey unit, any replacement measurements or samples required will be made within the remediated area at randomly selected locations following Awynt 9004 5-37 Rev.

Haddam Neck Plant License Termination Plan verification that the remediation activities did not affect the remainder of the unit. Re-survey will be required in any area of a survey unit affected by subsequent remediation activities.

5.6 Survey Protocol for Non-Structural Systems and Components The guidance provided in MARSSIM and DG-4006 for conducting final status surveys does not include guidance for conducting final status surveys for non-structural system or components. Per DG-4006, "non-structural systems and components" refers to anything not attached to or not an integral part of a building or structure. Given that the methods of the MARSSIM do not apply to non-structural systems and components, an alternative set of release criteria must be chosen to facilitate site remediation for license termination purposes.

The current site unconditional release limit of no detectable radioactive (licensed) material will be used to survey non-structural systems and components (excluding the cases discussed below). Non-structural systems and components meeting the criteria can be released, after survey. Those not meeting the release criteria will be disposed of as radioactive waste.

Buried pipe that is located within the saturated subsurface areas of the site (to remain on site) will be surveyed to the limits set forth in Table 5-7. Full-length surveys will be performed for this piping, typically using conventional methods and instrumentation. If advanced technology instrumentation, such as 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. Detection limits for surface activity assessments for this buried piping should be at least equivalent to the release limits given in Table 5-7 at the 95% confidence level. Detection limits will be computed using the methods described in Section 5.7.2.5 of this plan. If necessary, scaling factors may be applied to establish gross activity levels via radionuclide-specific measurements or other assessments, as appropriate.

As discussed in Section 5.4.7.5, embedded piping to remain on site will be surveyed to the building surface DCGLs, provided that the gross activity beta-to-alpha ratio associated with that piping (or penetration) at the time of FSS is greater than or equal to 15:1. If the gross activity beta-to-alpha ratio at the time of FSS is found to be less than 15: 1, the associated piping or penetration will be capped, grouted, or removed.

Evaluations as to whether material should be considered as a structure or a component will be via the guidance of Section B of DG-4006 and comparisons with the dose modeling scenarios used to develop the DCGLs that govern release of grounds and structures. Examples of parts of buildings or structures that are considered in the development of DCGLs include floors, walls, ceilings, doors, windows, sinks, hoods, lighting fixtures, built-in laboratory benches, and built-in furniture. Examples of non-structural systems and components include pumps, motors, heat exchangers, and piping between components.

August 2004 5-38 Rev. 2

Haddam Neck Plant License Termination Plan 5.7 Survey Implementation and Data Collection The requirements and objectives outlined in this plan and the project QA plan will be incorporated into Standard Operating Procedures (SOPs). Procedures will govern the survey design process, survey performance and data assessment (decision making). The final status survey design will be carried out in accordance with the SOPs and the QA plan, resulting in the generation of raw data. The product of the survey design process is a survey package, which addresses various elements of the survey, including, but not limited to:

• maps of the survey area showing the survey unit(s) and measurement/sample locations, as appropriate;

• applicable DCGLs

• instrumentation to be used;

• instrument calibration;

• types and quantities of measurements or samples to be made or collected;

• investigation criteria;

• QA/QC requirements (e.g., replicate measurements or samples);

• personnel training;

• applicable health and safety procedures;

• approved survey procedures; and

• applicable operating procedures.

An important element of the survey design process is establishing the DCGLs for the measurements to be made. The DCGLs will be determined as described in Section 5.4.7 based on characterization data for the survey unit(s) being considered. Isotopic mix, material backgrounds, and the variability of these will all be considered. The detection limit requirements dictated by the DCGLs affect the selection of both the instrumentation to be used for a given survey and the survey method(s) to be employed (advanced survey methods, fixed measurements, sampling; or combinations thereof).

5.7.1 Survey Methods The survey methods to be employed in the final status surveys will consist of combinations of advanced technologies, scanning, fixed measurements, sampling, and other methods as needed to meet the survey objectives. Additional methods may be used if such become available between the time this plan is adopted and the completion of final survey activities. However, any new technologies must still meet the applicable requirements of this plan. Note that in some cases, the same instrument may be used for more than one type of survey. For instance, a sodium-iodide (Nal) detector may be used in either a scanning mode or for fixed spectroscopic measurements.

5.7.1.1 Scanning Scanning is the process by which the operator uses portable radiation detection instruments to detect the presence of radionuclides on a specific surface (i.e., ground, wall, floor, equipment). The term scanning survey is used to describe the process of moving portable radiation detectors across a surface with the intent of locating residual radioactivity. Investigation levels for scanning surveys are determined during survey planning to identify areas of elevated activity. Scanning surveys are performed to locate radiation anomalies indicating residual gross activity that may require further investigation or action. These investigation levels may be based on the DCGLw or the DCGLEMC.

August 2004 5-39 Rev. 2

Haddam Neck Plant License Termination Plan No matter what survey approach is selected (combination of instrumentation and techniques), one of the most important elements of a survey is a prioriscanning to confirm that the unit is properly classified and to identify any areas where residual activity levels are elevated relative to the DCGLw. The purpose of scanning is to detect areas of residual activity that may not be detected by other measurement methods.

Thus, scanning should always be performed prior to any fixed measurement or sample collection in a survey unit. If the scanning indicates that the unit or some area within the unit has been improperly classified, then the survey design process must be evaluated to either assess the effect of re-classification on the survey unit as a whole (if the whole unit requires re-classification) or a new design must be established for the new unit(s) (in the case of sub-division). A new survey design will require a re-evaluation of the survey strategy to decide if it can meet the requirements of the revised survey design. If not, the survey strategy must be revised based on the available instrumentation and methods.

Table 5-9 gives the area coverage requirements when scanning is used with fixed measurements.

Table 5-9 Traditional Scanning Coverage Requirements Survey Unit Classification Required Scanning Coverage Fraction Class 1 100%

Class 2 Outdoor areas, floors, or lower walls of buildings:

10% to 100%

I Upper walls or ceilings: 10% to 50%

Class 3 Judgmental 5.7.1.2 Fixed Measurements Fixed measurements are taken by placing the instrument at the appropriate distance above the surface, taking a discrete measurement for a pre-determined time interval, and recording the reading. Fixed measurements may be collected at random locations in a survey unit or may be collected at systematic locations and supplement scanning surveys for the identification of small areas of elevated activity. Fixed measurements may also be collected at locations identified by scanning surveys as part of an investigation to determine the source of the elevated instrument response. Professional judgment may also be used to identify locations for fixed measurements to further define the areal extent of contamination. Locations for fixed measurements specified by a given survey design will be established as discussed in Section 5.5.

5.7.1.3 Advanced Technologies In the context of this Plan, advanced technologies refer to survey instruments or methods that create a spatially-correlated log of the measurements made as the detector is passed over an area. This logging of all of the measurements allows quantitative assessments of activity levels to be made, thus serving the same role as fixed measurements. Having all of the measurements logged allows statistical analyses to be made using a large number of samples, which provides for enhanced detection sensitivity relative to traditional scanning. The sensitivity achieved using advanced survey methods may, in some cases, be small enough relative to the DCGLW that the advanced method alone will allow a decision to be made as to whether a survey unit meets the release criterion without the need for additional fixed measurements.

The fact that the instrument records every measurement made over the entire area it covers inherently addresses the issue of small areas of elevated activity. Average and maximum residual activity August 2004 5-40 Rev. 2

Haddam Ncck Plant License Termination Plan concentrations can be quantified over any area desired, allowing one to assess compliance with the applicable criteria (DCGLw or DCGLEhC) by inspection.

If advanced technology instrumentation 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.

5.7.1.4 Other Advanced Survey Technologies Other instruments and methods that may be used for final status surveys include, but are not limited to, in situ gamma spectrometry, in situ object counting systems, and systems capable of traversing ducting or piping. Like the advanced technologies discussed above, these other methods may in some cases provide sufficient area coverage so that augmenting the measurement with scanning is not necessary.

In situ gamma spectrometry is an established technique for assaying the average radionuclide concentration in large volumes of material (for example, soil and activated concrete). It has the advantage of being able to assess large areas with a single measurement. If desired, the detector's field of view can be reduced through collimation to allow assay of smaller areas.

In situ object counting refers to gamma spectrometry systems that include software capable of modeling photon transport in complex geometries for the purpose of estimating detector efficiencies. This eliminates the need for a calibration geometry representing the object to be counted. Such systems are useful for assaying complex components such as heat exchangers. "Pipe crawler" systems may be employed to survey a length of piping or ducting.

If advanced technology instrumentation 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.

5.7.1.5 Samples Sampling is the process of collecting a portion of a medium as a representation of the locally remaining medium. The collected portion of the medium is then analyzed to determine the radionuclide concentration. Examples of materials that may be sampled include soil, sediments, concrete, paint, and groundwater.

Section 5.10, "Quality Assurance and Quality Control Measures" addresses QA requirements for final status survey activities that apply to onsite and offsite laboratories employed to analyze samples as a part of the final status survey process. Performance of laboratories will be verified periodically by QA auditors. This verification will include reviews of personnel training, procedures and equipment operation.

Trained and qualified individuals will collect and control samples. All sampling activities will be performed under approved procedures. CYAPCO will utilize a chain-of-custody (COC) process to ensure sample integrity.

August 2004 5-41 Rev. 2

Haddam Neck Plant License Termination Plan 5.7.2 Survey Instrumentation 5.7.2.1 Survey Instrument Data Quality Objectives The data quality objectives process includes the selection of instrumentation appropriate for the type of measurement to be performed (i.e., fixed measurement, scan or both), that are calibrated to respond to a radiation field under controlled circumstances; evaluated periodically for adequate performance to established quality standards; and sensitive enough to detect the radionuclide(s) of interest with a sufficient degree of confidence. The specific DQOs for instruments are established early in the planning phase for FSS activities, implemented by standard operating procedures and executed in the survey plan.

Further discussion of the DQOs for instruments is provided below.

5.7.2.2 Instrument Selection The selection and proper use of appropriate instruments for both fixed measurements and laboratory analyses is one of the most important factors in assuring that a survey accurately determines the radiological status of a survey unit and meets the survey objectives. The survey plan design must establish acceptable measurement techniques for scanning and direct measurements. The DQO process must include consideration as to the type of radiation, energy spectrum and spatial distribution of radioactivity as well as the characteristics of the medium to be surveyed (e.g., painted, scabbled, chemically decontaminated).

The particular capabilities of a radiation detector establish its potential for being used in conducting a specific type of survey based on the factors discussed above. Radiation survey parameters that will be needed for final survey purposes include surface activities and radionuclide concentrations in soil. To determine these parameters, both field measurements and laboratory analyses will be necessary. For certain radionuclides or radionuclide mixtures, both alpha and beta radiation may have to be measured. In addition to assessing average radiological conditions, the survey objectives must address identifying small areas of elevated activity.

Instruments must be stable and reliable under the environmental and physical conditions where they will be used, and their physical characteristics (size and weight) should be compatible with the intended application. This has been the case for typical radiation detection instrumentation used at HNP for operational surveys as well as scoping and characterization surveys.

The radiation detectors to be used for final survey activities at the Haddam Neck Plant can be divided into three general classes:

• gas-filled detectors,

• scintillation detectors, and

• solid-state detectors.

Gas-filled detectors include ionization chambers, proportional counters (both gas-flow and pressurized) and Geiger-Mueller (GM) detectors. Scintillation detectors include plastic scintillators, zinc-sulfide (ZnS) detectors and sodium-iodide (Nal) detectors. Solid-state detectors include both n-type and p-type intrinsic germanium detectors.

Finally, the DQO process must evaluate, depending on the type of radiation of interest, and on the application, the ability of instrumentation to measure levels that are less than the DCGL. In some cases instruments used for scanning may have detection limits that are greater than the DCGLw. This is August 2004 542 Rev. 2

Haddam Neck Plant License Termination Plan recognized by MARSSIM as an acceptable approach as long as the grid spacing (for Class I survey units) and investigation levels used are in accordance with Sections 5.5.1.5, 5.5.1.6 and 5.5.3.1, respectively, of this plan. The DQO process for instrument selection is performed in the planning phase for an FSS activity and is typically documented by a technical support document, which is referenced in the survey plan.

5.7.2.3 Calibration and Maintenance All instrumentation used for measurements to demonstrate compliance with the radiological criterion for license termination at the Haddam Neck Plant will be calibrated and maintained under approved plant procedures and the project QA plan or vendor QA plan that satisfies the requirement of the project QA plan. Instruments will be calibrated for normal use under typical field conditions at the frequency specified by vendor instructions or by approved plant procedures (at least annually). Calibration standards will be traceable to the National Institute of Standards and Technology (NIST). If external vendors are used for instrument calibration or maintenance, these services must be approved and conducted under the project QA plan. Calibration records will be maintained as required by plant procedures and the project QA plan.

Instruments used to measure gross beta surface activity will be calibrated using radionuclides such as Tc-99, Co-60, or Cs-137 so as to represent the beta energies for the beta-emitting radionuclides that will be encountered during final survey activities. Likewise, radionuclides such as Pu-239 or Th-230 may be used to calibrate instruments used to assess alpha surface activity so the alpha energies of the transuranic (TRU) radionuclides that may be encountered are adequately represented.

The DQO process must consider the field conditions the instrument will be used in to determine the affect and magnitude of variation from conditions established during calibration. These conditions might include source to detector geometry (including distance and solid angle), size and distribution of the source relative to the detector, and composition and condition of surface to be assessed. Most of these factors should have been determined during the instrument selection process. In some cases, instrument efficiencies may require modifications to account for surface conditions or coverings. Such modifications, if necessary, will be established using the information in Section 5 of NUREG-1507 and pertinent site characterization data. This will be performed during the planning process and documented by a technical support document and referenced in the survey plan. This technical support document will include the evaluation supporting instrument selection.

5.7.2.4 Response Checks The DQO process determines the frequency of response checks, typically before issue and after an instrument has been used (typically at the end of the work day but in some cases this may be performed during an established break in activity, e.g., lunch). This additional check will expedite the identification of a potential problem before continued use in the field. Instrumentation will be response checked in accordance with plant procedures. If the instrument response does not fall within the established range, the instrument will be removed from use until the reason for the deviation can be resolved and acceptable response again demonstrated. If the instrument fails a post-survey source check, all data collected during that time period with the instrument will be carefully reviewed and possibly adjusted or discarded, depending on the cause of the failure. In the event that data are discarded, the affected area will be resurveyed.

August 2004 5-43 Rev. 2

Haddam Neck Plant License Termination Plan 5.7.2.5 MDC calculations The DQO process evaluates the ability of the instrument to measure radioactivity at levels below the applicable DCGL. This evaluation will be performed and documented by a technical support document and referenced by the survey plan. This evaluation may also be included with the technical support document discussed in Section 5.7.2.2 above.

Instrument detection limits are typically quantified in terms of their Minimum Detectable Concentration, or MDC. The MDC is the concentration that a given instrument and measurement technique can be expected to detect 95% of the time under actual conditions of use.

Instruments and methods used for field measurements will be capable of meeting the investigation level in Table 5-8.

Before any measurements are performed, the instruments and techniques to be used must be shown to have sufficient detection capability relative to the applicable DCGLs. The detection capability of a given instrument and measurement technique is quantified by its MDC.

5.7.2.5.1 MDCs for Fixed Measurements Per NUREG-1507, MDCs for fixed measurements are computed as MDCfe + (Equation 5-28) where 3 and 4.65 =constants as described in NUREG-1507; B = background counts during the measurement time interval (t);

t = counting time; and K = a proportionality constant that relates the detector response to the activity level in the sample being measured.

The proportionality constant K typically encompasses the detector efficiency, self-absorption factors and probe area corrections, as required. The dimensions of the counting interval "t" are consistent with those for the MDC and the proportionality constant K. Thus, "t" would be in minutes to compute an MDC in dpm/100 cm2 .

An example calculation to determine the MDCfrxed for the detection of Co-60 with a 100 cm2 gas proportional detector is shown below.

Assumptions:

Background count rate = 200 cpm t = 1 minute B = 200 counts in the measurement time interval (t)

K = cis,(A/100), where A = area of the detector in cm 2 E; = 0.38 cpm/dpm Es = 0.25 (from ISO 7503-1) emissions per disintegration A = 100 cm2 August 2004 5-44 Rev. 2

Haddam Neck Plant License Termination Plan 3 + 4.65)(1i060 2 MD~ftCd =(0.38)(0.25)(100/ 100)(1) = 724dpm /1 00Cm Actual values for Es will be selected from ISO 7503-1 (Reference 5-12) or NUREG-1507 or empirically determined and documented prior to performing the final status survey.

5.7.2.5.2 MDCs for Beta-Gamma Scan Surveys for Structure Surfaces As recommended in Draft Guide-4006 (and as incorporated into Section 5.1 of Appendix E to NUREG-1727), MDCs for surface scans for structure surfaces for beta and gamma emitters will be computed via 1.38Vi dpmI1OOcm 2 MlDCstructurescan (Equation 5-29)

(A ° where 1.38 = sensitivity index, B = number of background counts in time interval t, p = surveyor efficiency,

£; = instrument efficiency for the emitted radiation (cpm per dpm),

= source efficiency (intensity) in emissions per disintegration, A = sensitive area of the detector (cm2 ),

t = time interval of the observation while the probe passes over the source (minutes).

The value of 1.38 used for the sensitivity index corresponds to a 95% confidence level for detection of a concentration at the scanning MDC with a false positive rate of 60%. The numerator in Equation 5-29 I represents the minimum detectable count rate that the observer would "see" at the performance level represented by the sensitivity index. The surveyor efficiency (p) will be taken to be 0.5, as recommended by DG-4006 (and incorporated into Section 5.1 of Appendix E to NUREG- 1727). The factor of 100 corrects for probe areas that are not 100 cm2 . In the case of a scan measurement, the counting interval is the time the probe is actually over the source of radioactivity. This time depends on scan speed, the size of the source, and the fraction of the detector's sensitive area that passes over the source; with the latter depending on the direction of probe travel. The source efficiency term (s) in Equation 5-29 may be I adjusted to account for effects such as self-absorption, as appropriate.

An example calculation to determine the MDCstructure, scan for the detection of Co-60 with a 100 cm2 gas proportional detector follows.

Assumptions:

Probe width = 4 inches Scan rate = 2 inches/sec Background count rate = 200 cpm t = 2 seconds = 0.033 minute B = 6.7 counts in the measurement time interval (t) p=0.5 i= 0.38 cpm/dpm August 2004 5-45 Rev. 2

Haddam Neck Plant License Termination Plan cs = 0.25 (from ISO 7503-1) emissions per disintegration A = 100 cm 2 MDC Structure4scaf = 1.38r.7 (10 = 161ldpm/lOOcm 16 2

,/.(0.38)(0.25)( o')(0.033)

Actual values for s&,will be selected from ISO 7503-1 or NUREG-1507 or empirically determined and documented prior to performing the final status survey.

5.7.2.5.3 MDCs for Alpha Scan Surveys for Structure Surfaces In cases where alpha scan surveys are required, MDCs must be quantified differently than those for beta-gamma surveys because the background count rate from a typical alpha survey instrument is nearly zero (I to 3 counts per minute typically). Since the time that an area of alpha activity is under the probe varies and the background count rates of alpha survey instruments is so low, it is not practical to determine a fixed MDC for scanning. Instead, it is more useful to determine the probability of detecting an area of contamination at a predetermined DCGL for given scan rates. In general, it is expected that separate alpha and beta surface activity measurements will not be necessary at the HNP and that surrogate measurements will instead be used for alpha surface activity assessments (see Section 5.4.7.3).

For alpha survey instrumentation with a background around one to three counts per minute, a single count will give a surveyor sufficient cause to stop and investigate further. Thus, the probability of detecting given levels of alpha emitting radionuclides can be calculated by use of Poisson summation statistics.

Doing so (see MARRSIM Section 6.7.2.2 and Appendix J for details), one finds that the probability of detecting an area of alpha activity of 300 dpm/IOOcm 2 at a scan rate of 3 cm per second (roughly I inch per second) is 90% if the probe dimension in the direction of the scan is 10 cm. If the probe dimension in the scan direction is halved to 5 cm, the detection probability is still 70%. Choosing appropriate values for surveyor efficiency, instrument and surface efficiencies will yield MDCs for alpha surveys for structure surfaces. If for some reason lower MDCs are desired, then scan speeds can be adjusted, within practical limits, via the methods of Section 6.7.2.2 and Appendix J of the MARSSIM.

5.7.2.5.4 MDCs for Gamma Scans of Land Areas As recommended in DG4006 (and Section 5.1 of Appendix E to NUREG-1727), the values given in Table 6.7 of MARSSIM may be adopted for gamma scans of land areas if Nal detectors of the dimensions considered in the table are used. If larger Nal detectors (e.g., 3 inch by 3 inch) or other detector types (e.g., plastic scintillator) are used, then the scan MDC will be computed using the methods of Section 6.7.2.1 of MARSSIM. This is the same method as was used to derive the values given in MARSSIM Table 6.7. As an alternative, a specific technical study may be performed and documented to establish efficiency to a soil standard consistent with MARSSIM guidance.

The radionuclides represented in MARSSIM Table 6.7 encompass those expected to be encountered in gamma scans for land areas at the HNP. If desired, the methods of Sections 5.4.7.2 and 5.4.7.3 of this plan may be used to establish scan MDCs based on radionuclide mix ratios. Alternatively, the most limiting value for the radionuclide mix may be used, with most limiting in this case meaning the August 2004 5-46 Rev. 2

Haddam Neck Plant License Termination Plan radionuclide for which the MDC is the largest fraction of its DCGLw for soil, while still meeting the criteria of 5.5.3 .1.

An example calculation to determine the MDCl,,nd, ,,, for the detection of Cs-137 with a 2"x2" Nal detector is shown below.

The Minimum Detectable Count Rate (MDCR) for a surveyor must be calculated prior to determining the scan MDC. The MDCR is dependent upon the background counts expected during time, t, at which the detector is located over the localized contamination. The MDCR for a surveyor is calculated using the following expression:

MDCRs o MCsui.vyor 1.384 N t- (Equation 5-30) where b = the background counts expected during time, t t= the time the detector is located above the localized contamination p = the surveyor efficiency Assumptions:

Scan speed = 0.5 meters/sec Localized contamination diameter = 56 cm Background count rate = 7000 cpm b = 130.67 counts in the measurement time interval (t) t = 0.0187 minute p=O. 5 1.38 IJ3i.67 MDCRMCRsurwyor = /iF 1.8

._ (0.01I9) 195cp lSp Next, the Minimum Detectable Exposure Rate (MDER) is calculated by dividing the MDCRsurveyorby the response to exposure rate factor for Cs-137 of 900 cpni/pR/h from MARSSIM Table 6.7 as follows:

MDER = 1195cpn = 1.331 d/I 900cpm/1,uR/I The MicroshieldTNl modeling code is used to calculate the exposure rate from the localized contamination.

Assuming a localized contamination depth of 15 cm, a density of 1.6 g/cm3 , a dose point of 10 cm above the surface and an initial concentration of 5 pCi/g of Cs-137, results in a calculated exposure rate equal to 1.268 pRIh. The scan MDC is calculated by dividing the MDER by the localized contamination exposure rate conversion factor as follows:

Cs-l37scanMDC=5pCi/g1.33/R / =-5.24pCi/g 1.268iR/Ih August 2004 5-47 Rev. 2

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

5.7.2.6 Typical instrumentation and MDCs Table 5-10 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 (Es) 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 2004 5-48 Rev. 2

Haddam Neck Plant License Termination Plan Table 5-10 Available Instruments and Associated 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/lOO cm 2 3140 dpm/I00 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 cm2 1770 dpm/100 counter (100 cm2) scans or fixed (Tc-99); plateau); (P plateau) cm2 (P plateau);

measurements a plateau: 15 cpm (a 90 dpm/l00 cm 2 (a 400 dpm/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 2 1230 dpm/100 (100 cm 2) scans or fixed (I minute count) cm2 measurements for structure surfaces dual-phosphor scans or fixed 20% (Co-60) 300 cpm (I 420 dpm/I00 Cm2 1300 dpm/I00 scintillator measurements; a 18% (Am-241) mode); (P mode); cm2 (O mode);

(100 Cm2) and, 6 cpm (a 80 dpm/100 cm 2 (a 400 dpm/100 independently or mode) mode) cm2 (a mode) simultaneously ZnS scintillator alpha scans or 19% (Pu-239) 2 cpm 50 dpm/100 cm2 (I 400 dpm/I00 (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 Nal 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 Nat 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 Nal spectroscopy - energy and energy and 0.2 pCi/g Cs-137 soil geometry geometry (10 minute counts)

HPGe in-situ gamma Varies with Varies with 0.05 pCi/g Co-60 N/A spectroscopy- energy and energy and 0.05 pCi/g Cs-137 soil geometry geometry (10 minute counts) position-sensitive scan-and-record Co-60 (0): 18% 350 cpm/100 Typical values are 1,925 dpm/l00 cm2 proportional surveys Am-241 (a): cm2 beta p and 200 dpm/I00 cm2 a counter 23% 15 cpm/100 cm2 alpha August 2004 5-49 Rev. 2

Haddam Neck Plant License Termination Plan 5.7.3 Survey Considerations The available complement of survey instrumentation and techniques will be evaluated to select an integrated approach that will effectively measure residual radioactivity for a given survey unit. The survey design must rely on both the historical site assessment and pertinent data from characterization or remediation support surveys to ensure a complete survey approach. Considerations that will be addressed in the selection of survey instrumentation and techniques include, but are not limited to:

• the types of measurements required;

• suitability for the expected physical and environmental conditions;

• MDCs for advanced survey methods, traditional scanning surveys, fixed measurements, and sampling relative to the DCGLw and the DCGLEMC;

• radionuclide mix, including hard-to-detect and alpha-emitting radionuclides;

• expected spatial variability of any suspected residual contamination;

• accessibility of areas (may impact coverage for scanning surveys); and

• the need for any judgmental assessments to address areas believed to have a higher potential for contamination or situations such as potential sub-surface contamination where prudence would dictate some additional sampling.

5.7.3.1 Survey Considerations for Buildings, Structures and Equipment The condition of surfaces following decontamination activities can affect the choice of survey instruments and techniques. Removing contamination that has penetrated a surface usually involves removing the surface material. As a result, the floors and walls of decontaminated facilities can be scarred or broken up and uneven. Such surfaces are more difficult to survey because it is not possible to maintain a fixed distance between the detector and the surface. In addition, scabbled or porous surfaces may attenuate radiation - particularly alpha and low-energy beta particles, and pose an increased risk of damage to detector probe faces. Surface irregularities may also cause difficulty in rolling or maneuvering detector systems on wheels.

Part of the planning for the final status survey of a particular survey unit will include an evaluation of the surfaces to be monitored. For conventional instrumentation, surface anomalies will be identified as part of this process and will be taken into account when selecting efficiencies to convert instrument readings to activity and in the calculation of the corresponding MDCs. Conservative values will be chosen based upon surface conditions. If the condition of the surface in the area changes in a more conservative direction (e.g. shorter detector to surface distance), the effect on the MDC will be assessed but may not be re-derived. If the condition of the surface changes in a non-conservative direction (e.g. different construction material which has higher natural radioactivity) the MDC will be assessed and re-derived.

Expansion joints, stress cracks, floor/wall interfaces, and penetrations into floors and xvalls for piping, conduit, anchor bolts, etc., are potential sites for accumulation of contamination and pathways for migration into sub-floor soil and hollow wall spaces. Roof surfaces and drainage points are also important August 2004 5-SO Rev. 2

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 gain access to areas for sampling.

5.7.3.1.1 Activity Beneath Surfaces 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 below-grade basements (i.e., those more than four feet below ground level) will be surveyed and decontaminated to meet the DCGLs in Chapter 6. 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 in the area.

5.7.3.1.2.2 Footings and Foundations 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 August 2004 5-51 Rev. 2

Haddam Neck Plant License Termination Plan foundation supports. Surfaces of these below-grade structures 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 further characterized. The results of characterization of the external surfaces will be used in the design and DQOs of the subsurface FSS to be performed in the area.

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.

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Haddam Neck Plant License Termination Plan 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.

5.7.3.1.6 Activated Concrete 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 demonstrate that it meets the applicable volumetric DCGLs. 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) will be treated as concrete for purposes of assigning DCGLs. 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 DCGLs 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.

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, August 2004 5-53 Rev. 2

Haddam Neck Plant License Termination Plan provided adequate instrument efficiencies and detection limits can be achieved. If necessary, scaling factors may be applied to establish gross activity levels via radionuclide-specific measurements or other assessments, as appropriate.

5.7.3.2 Survey Considerations for Outdoor Areas 5.7.3.2.1 Residual Radioactivity in Surface Soils In this context, surface soil refers to outdoor areas where the soil is, for purposes of dose modeling, considered to be uniformly contaminated from the surface down to some specified depth. These areas will be surveyed through combinations of sampling, scanning, and in-situ measurements, as appropriate.

5.7.3.2.2 Residual Radioactivity in Subsurface Soils Residual radioactivity in subsurface soils refers to residual radioactivity residing under the top 6 inches of soil or is underneath structures such as building floors/foundations. Such areas include, but are not limited to, areas under buildings, building floors/foundations, or components where leakage was known or suspected to have occurred in the past; on-site storage areas where radioactive materials have been identified; and areas containing spoils from past dredging of the discharge canal. However, the assessment of all subsurface soil contamination is not currently complete. Soil in difficult to access areas such as under buildings will be deferred until later in the decommissioning process. As a part of survey planning, borehole logs will be reviewed, when available.

The DQO process for subsurface areas will be similar to the DQO process used for other surveys at HNP (e.g., final status survey for surface soils). However, there may be differences in design input parameters as necessary to satisfy the objectives of the plan. Additional detail regarding subsurface input parameters and methodology are provided below. Surveys (i.e., scoping, characterization, remediation and final status survey) for subsurface areas will be performed under a documented survey plan developed using the DQO process. The level of effort with which the DQO process is used as a planning tool is commensurate with the type of survey and the necessity of avoiding a decision error. This is the graded approach of defining data quality requirements as described previously in the LTP. For example scoping and characterization survey plans intended to collect data might only require a survey objective and the instrumentation and analyses specifications necessary to meet that survey objective. Remediation and final status survey plans which require decisions would need additional effort during the planning phase according to the level of risk of making a decision error and the potential consequences of making that error.

The DQA process will be used to assess data and demonstrate achievement of the sampling 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, review of preliminary data, use of appropriate statistical testing when applicable (see discussion below), verify the assumptions of the statistical tests and draw conclusions from the data.

Evaluation of subsurface soil at HNP during final status survey will be a combination of systematic and biased measurements. Measurements may be either in-situ gamma spectroscopy by well logging or other advanced technology, provided the MDC meets the criteria discussed in Section 5.7.2, or by sampling. If advanced technology instrumentation 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.

August 2004 5-54 Rev. 2

Haddam Neck Plant License Termination Plan This document will be available for NRC inspection in support of final status survey activities. Sample locations will be selected randomly, for Class C areas, or by random-start systematic grid for Class A and B areas, supplemented with biased measurements. Biased measurements will be obtained at the locations of localized contamination. Where samples are taken, each 3-meter core (or other means of collection, such as test pits) will be homogenized and measured.

The horizontal extent of contamination will only be established for judgmental sampling and for samples within a systematic sampling area that exceed the DCGLEMc. For the case where the DCGLEMC comparison is made, the value used for the area factor will be determined from the area bounded by the adjacent samples or by the area bounded by additional samples at or below the DCGLW. This approach is consistent with the model used to calculate DCGLs in Chapter 6.

As discussed in Section 2.3.3.1.5, subsurface soils at HNP are divided into 3 classifications: Class A, Class B and Class C. This classification defines the measurement or sample density. There will be 31 measurement locations (approximately one per 500 m2 ) in Class A areas. In addition, biased measurements or samples will be obtained at the locations of localized remediation efforts where subsurface leaks are suspected to have created soil contamination. Random measurements or samples will be obtained in Class B and Class C areas. There will be 25 measurement locations in a Class B area.

There will be 15 measurement locations in Class C areas. In addition, biased measurements or samples will be obtained in Class B and Class C areas based upon characterization data and professional judgment. If a systematic or random sample location falls on a building foundation, a sample will be obtained at that location unless the building is in contact with bedrock. The range of the number of measurements in Class A, B, and C areas (31 measurements in the Class A area to 15 measurements in Class C areas) corresponds to the range of values for N (for Sign test, or N/2 for WRS), considering a = 0.05, , = 0.05, and I.O<A/cr <3.0. All samples will be evaluated against the soil DCGLs by using either the Sign or WRS test.

Investigation levels applicable to surface soils (given in Table 5-8) will be applied to subsurface soils.

Similarly the area factors for surface soils (given in Table 5-5) will be applied to subsurface soils. That is no sample can exceed the DCGLEMC without an investigation being performed. These investigations would be similar to those performed for surface soils.

Samples will be obtained to a depth of 3 meters or bedrock, whichever is reached first. These samples will be homogenized over the entire depth of the sample obtained. In cases where refusal is met because of bedrock, the sample will be used "as is." In cases where a non-bedrock refusal is met prior to the 3-meter depth, the available sample will be used to represent the 3-meter sample, if the viable sample is at least 1.5 meters in depth. If a non-bedrock refusal is met before the 1.5-meter depth, then a new sample will be obtained within a 3-meter radius from the original location. All samples will be analyzed by gamma spectrometry. Because the mobility of some of the radionuclides, believed to be present, is not well understood, some of the samples will undergo analysis for all hard-to-detect radionuclides. A minimum of 5% of the samples will be analyzed for hard-to-detect radionuclides. During specific investigations, such as the identification of the horizontal extent of contamination, analysis of a larger percentage of samples for hard-to-detect radionuclides will be performed.

5.7.3.2.3 Paved areas Paved areas that remain at the HNP following decommissioning activities may require surveys for residual radioactivity on the surface, beneath the surface, or both. As part of the survey design and planning process, historical information will be reviewed to determine whether radiological incidents or plant alterations have occurred in the survey unit. Where indications are that impacted soil could have been mixed by grade work prior to paving, this will be factored into final survey design to establish a August 2004 5-55 Rev. 2

Haddam Neck Plant License Termination Plan reasonable depth of disturbed soil for evaluation. If it is determined that the soil beneath pavement has been impacted, the final status survey will incorporate appropriate surveys and sampling.

If residual radioactivity is primarily on or near the surface of the paved area, for purposes of surveying, measurements will be taken as if the area were surface soil. If the residual radioactivity is primarily beneath the paving, it will be treated, for purposes of surveying, as subsurface residual radioactivity.

5.7.3.2.4 Groundwater Assessments Assessments of any residual activity in groundwater at the Haddam Neck Plant will be via groundwater monitoring wells. The monitoring wells installed at the site will monitor groundwater at both deep and shallow depths. Section 2.3.3.1.6 describes, in depth, the groundwater monitoring to be conducted.

5.7.3.2.5 Bedrock Assessments Exposed bedrock from the demolition of structures or the remediation of contaminated soils creates another area requiring a methodical radiological assessment.

Several areas of the site will be excavated to bedrock either through the demolition of buildings or the removal of contaminated soils. Initial excavation in the Tank Farm area to bedrock will allow for data to be assessed on the potential magnitude and distribution of contamination within the bedrock. This assessment will include the determination of:

• The degree of contamination on the bedrock surface,

• The degree of contamination migration into bedrock cracks and fissures, and

• The observation of surface conditions of the bedrock.

As remediation progresses, the bedrock surfaces will be cleaned of readily removable material using techniques such as vacuuming and air pressure removal (combined with vacuum collection of removed material). Following remediation, the radiological conditions will be assessed prior to backfilling the excavation. The backfill will ultimately consist of clean fill.

The dose pathway that would apply to such open bedrock excavations will be from potential future groundwater contamination since other pathways such as direct exposure, and plant uptake would not apply to this material (clean backfill provides substantial shielding to the surface and farming plants would not be grown in bedrock). Therefore, the post-remediation field assessment and dose assessment methods focus on the radioactivity inventory potentially available to future uncontaminated groundwater in contact with the remaining radioactivity in bedrock from each bedrock excavation.

The monitoring of the bedrock area will be through the installation and sampling of groundwater monitoring wells. Once the bedrock area is backfilled with clean fill material, the dewatering wells, pumps, used to suppress the groundwater will be turned off allowing the groundwater to return to an equilibrium condition in the unconsolidated backfill materials. Monitoring wells will be installed at locations to provide groundwater samples for monitoring. The installed monitoring wells will be sampled quarterly for at least 18 months to include two springtime periods when groundwater has its greatest impact.

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Haddam Neck Plant License Termination Plan 5.7.3.2.6 Surface Water and Sediments Sediments will be assessed by collecting samples within locations of surface water ingress or by collecting composite samples of bottom sediments, as appropriate. Such samples will be collected using approved procedures based on accepted methods for sampling of this nature. Sample locations will be established using the methods of Section 5.5.1 of this plan. Scanning in such areas is not applicable.

Note that per.the agreement with the Connecticut DPUC, radiological sampling of fish, water and sediment will be performed for the pond (Survey Area 9508).

Also note, per the agreement with the Connecticut DPUC, the sampling density will be adjusted in the area of the canal, from the outfall (the beginning of the canal closest to the plant) to 50 feet past the weir, to be twice the density that would otherwise be required of a Class 2 survey unit.

Sediment samples will be evaluated against the DCGLs for soil. This is considered appropriate given that the action that would result in the greatest radiological impact to future inhabitants of the site would be to dredge up the sediment and use it for farming. If the sediment is left in place, then use of the soil DCGLs is conservative since many of the pathways considered in developing the soil DCGLs (direct exposure, uptake by plants, etc.) would not apply.

Assessment of residual activity levels in surface water drainage systems will be via sampling of sediments, fixed measurements, or both, as appropriate, making measurements at traps and other appropriate access points where activity levels should be representative or bound those on the interior surfaces.

5.7.3.2.7 Storm Drains and Buried Piping Other Than Described in Section 5.7.3.1.5 Most buried piping (including storm drains) will be removed from the site; however, any buried piping that remains following decommissioning activities that has a potential to contain residual activity will be surveyed using the criteria given in Section 5.4.7.5.

5.7.3.3 Surveillance Following Final Status Surveys Isolation and control measures will be implemented through approved plant procedures and will remain in force throughout final survey activities and until there is no risk of contamination from decommissioning or the survey area has been released from the license. In the event that isolation and control measures established for a given survey unit are compromised, evaluations will be performed and documented to confirm that no radioactive material was introduced into the area that would affect the results of the Final Status Survey.

To provide additional assurance that land areas and the limited number of structures that have successfully undergone FSS remain unchanged until final site release, documented periodic evaluations of the FSS areas will be performed. The periodic evaluation will consist of:

• A walkdown of the areas to check for proper postings,

• Check for materials introduced into the area since the last evaluation,

• Any general disturbance that could change the FSS including the potential for contamination from adjacent decommissioning activities,

• A review of the 50.75(g) files.

Evaluations will be documented and controlled in accordance with site procedures August 2004 5-57 Rev. 2

Haddam Neck Plant License Termination Plan 5.7.3.3.1 Surveillance of Structures Routine surveys will be performed on structures that have completed their final status survey until the structure is backfilled. These routine operational health physics surveys will be used to verify that the as-left radiological conditions in the area have not changed. These routine surveys will typically include survey locations on the floor and on lower walls, and areas of ingress, egress, and storage. Locations will be selected on a judgmental basis, based on technician experience and conditions present in the survey area at the time of the evaluation, but primarily designed to detect the migration of loose surface contamination from decommissioning activities taking place in adjacent areas and other areas in close proximity that could cause a potential change in conditions.

If the area is suspect following the routine surveillance survey, then a full FSS survey of the affected unit(s) will be performed in accordance with the LTP. The results for the FSS re-survey and investigations surveys will be documented and maintained in the FSS files for the affected survey unit(s).

Additionally, for any area that has completed FSS activities, any soil, sediment, or equipment relocated to that area will require a demonstration that the material being introduced does not result in resident radioactivity that is statistically different that that identified in the FSS. Once a structure has been backfilled, the periodic surveillance will be similar to the surveillance employed for open land areas.

5.7.3.3.2 Surveillance of Open Land Areas Open land areas that have been final status surveyed will be evaluated periodically, not to exceed semi-annually.

If the area is suspect following the evaluation an investigation survey will be performed to confirm the FSS surveys validity. This investigation survey will involve judgmental sampling of the suspect areas. If the results of the investigation survey indicates that contamination is statistically different than the initial FSS results (>2 standard deviations from the mean), then the investigation survey will be increased to include a larger physical area than the initial investigation survey. If the final results of the investigation survey are statistically different than the FSS survey results, then a full FSS survey of the affected areas will be performed in accordance with the LTP. The results of the FSS re-survey and investigation surveys will be documented and maintained in the FSS files for the affected survey units. Additionally, for any area that has completed FSS activities, any soil or sediment relocated to that area will require demonstration that the material introduced does not result in residual radioactivity that is statistically different than that in the FSS.

5.7.3.3.3 Surveillance of Bedrock Areas Generally, bedrock areas will not remain exposed for a period of time such that surveillance would be necessary. Typically the bedrock area will be assessed for radiological conditions and then backfilled with clean fill material. Any necessary groundwater level adjustments will be made, which include stopping groundwater dewatering from the bedrock area to allow the "normal" groundwater levels to be restored and the installation of any monitoring wells needed to support ongoing radiological groundwater monitoring.

If the bedrock area is suspect, following the evaluation, an investigation assessment will be performed to confirm the radiological assessment's validity. This investigation assessment will involve judgmental sampling of boundary and/or potential access points to bedrock area. If the results of the investigation August 2004 5-58 Rev. 2

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 5-11.

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Haddam Neck Plant License Termination Plan Table 5-11 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 5-12.

Table 5-12 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 2004 5-60 Rev. 2

Haddam Neck Plant License Termination Plan test indicates that it should be rejected in favor of the alternative. The alternative hypothesis is that the median concentration in the survey unit exceeds that in the reference area by less than the DCGLw. Note that some or all of the survey unit measurements may be larger than some reference area measurements, while still meeting the release criterion. Indeed, some survey unit measurements may exceed some reference area measurements by more than the DCGLW. The result of the hypothesis test determines whether or not the survey unit as a whole is deemed to meet the release criterion. The EMC is used to screen individual measurements.

The WRS test is applied as described in the following steps:

1. Adjust the reference area measurements by adding the DCGLw to each one.

2. Pool the adjusted reference area measurements and the sample (survey unit) measurements and rank them in increasing order from I to the total number of data points (reference measurements plus sample measurements).

3. For any measurements that have the same value, the rank assigned to that set of measurements is the average of their ranks.

4. Sum the ranks of the adjusted reference area measurements.

5. Compare the sum of the adjusted reference area measurements (W.) with the critical value from Table 1.4 of the MARSSIM for the appropriate values of m (the number of reference measurements), n (the number of sample measurements), and a (the decision error rate).

If the value Wr determined from steps 1 through 5 above exceeds the critical value from Table 1.4 of the MARSSIM, then the null hypothesis is rejected and the alternate accepted. In other words, the results show that the survey unit meets the release criterion.

Note that the WRS test described in steps I through 5 above assumes that there are no "less than" results in the data set, i.e., that all of the data points have a quantitative value rather than "background" or "less than MDC." Though it is not anticipated that data of this nature would be among that collected for a final status survey, if it is encountered and must be used, the method described in Section 8.4.2 of the MARSSIM will be used to assign rank to these values. If more than 40% of the data collected for a final status survey are "less than" values, then the WRS test cannot be used.

5.8.2 Sign Test Radionuclide specific measurements for which the radionuclide(s) of interest either does not exist in background or is not present in a concentration that is a relevant fraction of the DCGLw will be evaluated using the Sign test. In addition, the Sign test may be used to evaluate gross activity measurements from survey units containing multiple materials by subtracting the appropriate background using paired measurements.

The null hypothesis tested by the Sign test is the same as that used for the WRS test. As with the WRS test, some individual survey unit measurements may exceed the DCGLw even when the survey unit as a whole meets the release criterion. In fact, a survey unit average that is close to the DCGLw might have almost half of its individual measurements greater than the DCGLw. Such a survey unit may still not exceed the release criterion. As with the WRS test, the EMC is used to screen individual measurements.

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Haddam Neck Plant License Termination Plan The Sign test is applied as described in the following steps:

1. List the survey measurements.

2. For each survey unit measurement, subtract the measurement from the DCGLw and record the differences.

3. Discard any difference that is exactly zero and reduce the total number of measurements (N) by the number of zero differences.

4. Count the number of positive differences. This value is the test statistic S+.

5. Compare the number of positive difference (S+) to the critical values from Table 1.3 of MARSSIM for the appropriate values of N (total measurements) and a (decision error rate). (A positive difference corresponds to a measurement below the DCGLW and contributes evidence that the survey unit meets the release criterion.)

If S+ is greater than the critical value in Table 1.3, then the null hypothesis is rejected and the alternate accepted.

Note that "measurements" in Step I above refers to the net result in cases where background-subtracted gross activity measurements (using the paired observation methodology) are being evaluated.

Though it is not anticipated, if any of the data collected from a final status survey are reported as "less than MDC" or as background, actual values will be assigned, even if negative, for purposes of applying the Sign test.

5.8.3 Elevated Measurement Comparison The Elevated Measurement Comparison (EMC) consists of comparing each measurement from the survey unit with the investigation levels discussed in Section 5.5.3. The EMC is performed for both measurements obtained on the systematic-sampling grid and for locations flagged by scanning measurements. Any measurement from the survey unit that is equal to or greater than an investigation level indicates an area of relatively high concentrations that should be investigated, regardless of the outcome of the nonparametric statistical tests. Thus, the use of the EMC against the investigation levels may be viewed as assurance that unusually large measurements will receive proper attention regardless of the outcome of those tests and that any area having the potential for significant dose contributions will be identified. The EMC is intended to flag potential failures in the remediation process. It should not be used as the primary means to identify whether or not a unit meets the release criterion.

If residual radioactivity exists in an isolated area of elevated activity in addition to residual radioactivity distributed relatively uniformly across a survey unit, the unity rule will be used to ensure that the total dose is within the release criterion, i.e.,

Celevated <1 (Equation 5-31)

DCGLw 1 (AreaFactor)xDCGLJ where: 8 = average concentration outside the elevated area, Celevted = average concentration in the elevated area.

A separate term will be used in Equation 5-31 for each elevated area identified in a survey unit.

August 2004 5-62 Rev. 2

Haddam Neck Plant License Termination Plan Note that EMC considerations generally apply only to Class I survey units, since areas of elevated activity should not exist in Class 2 or Class 3 survey units.

5.8.4 Unity Rule When radionuclide specific measurements are made in survey units having multiple radionuclides, compliance with the radiological release criterion will be assessed through use of the unity rule, also known as the sum of fractions. The unity rule, represented in the expression below, is satisfied when radionuclide mixtures yield a combined fractional concentration limit that is less than or equal to one, i.e.:

____ C2 C.__

+~ +~ ... + *~ 1 (Equation 5-32)

DCGLi DCGL2 DCGL(

where:

C, = Concentration of radionuclide n DCGLn= DCGL for radionuclide n 5.8.5 Data Assessment Conclusions The result of the data assessment is the decision to reject or not to reject the null hypothesis. Provided that the results of investigations triggered by the EMC were resolved, a rejection of the null hypothesis leads to the decision that the survey unit meets the release criterion. If the data assessment concludes that the null hypothesis cannot be rejected, this may be due to one of two things: 1) the average residual concentration in the survey unit exceeds the DCGLw; or 2) the analysis did not have adequate statistical power. "Power" in this context refers to the probability that the null hypothesis is rejected when it is indeed false. Quantitatively, the power is I - A,where P is the Type II error rate (the probability of accepting the null hypothesis when it is actually false). A retrospective power analysis can be used in the event that a survey unit is found not to meet the release criterion to determine if this is indeed due to excess residual activity or if it is due to an inadequate sample size.

Retrospective power analyses, if necessary, will be performed following the methods of MARSSIM Sections 1.9 and 1.10 for the Sign test and WRS test, respectively. If the retrospective power analysis indicates insufficient power, then an assessment will be performed to determine whether the observed median concentration and/or observed standard deviation are significantly different from the estimated values used during the DQO process. The assessment may identify and propose alternative actions to meet the objectives of the DQOs. These alternative actions may include failing the unit and starting the DQO process over, remediating some or all of the survey unit and starting the DQO process over and adjusting the LBGR to increase sample size. For example, the assessment determines that the median residual concentration in the survey unit exceeds the DCGLW or is higher than was estimated and planned for during the DQO process. A likely cause of action might be to fail the unit or remediate and resurvey using a new sample design. As another example, the assessment determines that additional samples are necessary to provide sufficient power. One course of action might be to determine the number of additional samples and collect them at random locations. Note, this method may increase the Type I error, therefore agreement with the regulator will be necessary prior to implementation. Another action would be to resample the survey unit with a new (and appropriate) number of samples and/or a new survey design.

There may be cases where the decision was made during the DQO process by the planning team to accept lower power. For instance, during the DQO process the calculated relative shift was found to be less August 2004 5-63 Rev. 2

Haddam Neck Plant License Termination Plan than 1. The planning team adjusts the LBGR, evaluates the impact on power and accepts the lower power. In this case, the DQA process would require the planning team to compare the prospective power analysis with the retrospective power analysis and determine whether the lower power is still justified and the DQOs satisfied.

5.9 Final Status Survey Reports Documentation of the final status survey will transpire in two types of reports and will be consistent with Section 14.5 of NUREG-1727 and Section 8.6 of NUREG-1575. An FSS Survey Unit Release Record will be prepared to provide a complete record of the as-left radiological status of an individual survey unit, relative to the specified release criteria. Survey Unit Release Records will be made available to the NRC for review upon request. An FSS Final Report, which is a written report that is provided to the NRC for its review, will be prepared to provide a summary of the survey results and the overall conclusions which demonstrate that the Haddam Neck Plant site, or portions of the site, meets the radiological criteria for unrestricted use.

5.9.1 FSS Survey Unit Release Records An FSS Survey Unit Release Record will be prepared upon completion of the final status survey for a specific survey unit. Sufficient data and information will be provided in the release record to enable an independent re-creation and evaluation at some future time. The format and content of the FSS Survey Unit Release Record is as follows:

• Survey Unit Description, including unit size, descriptive maps, plots or photographs, including reference coordinates and historic changes in description;

• ClassiflcationBasis, including significant historical site assessment and characterization data used to establish the final classification as well as a statement on the impact groundwater had on the final classification;

• Data Quality Objectives stating the primary objective of the survey, and a brief description of the DQO process;

• Survey Design describing the design process, including methods used to determine the number of samples or measurements required based on statistical design, number of biased or judgmental samples or measurements required, method of sample or measurement locating, and a table providing a synopsis of the survey design;

• Survey Implementation describing survey methods and instrumentation used, accessibility restrictions to sample or measurement location, number of actual samples or measurements taken, documentation activities, Quality Control samples or measurements, and scan data collected in tabular format;

• Survey Results including types of analyses performed, types of statistical tests performed, statement of pass or failure of the statistical test(s);

• Quality Control results to include discussion of split samples and/or QC replicate measurements; August 2004 5-64 Rev. 2

Haddam Neck Plant License Termination Plan

• Investigations and Results;

• Remediation activities, both historic and resulting from the final status survey;

• Changesfrom the FinalStatus Survey Plan including field changes;

• Data QualityAssessment;

• Anomalies occurring during the survey or in the sample results;

• Conclusion as to whether or not the survey unit satisfied the specified release criteria, a discussion of ALARA evaluations performed, and whether or not sufficient power was achieved;

• Attachments and enclosures to include supporting maps, diagrams, and sample statistical data.

5.9.2 FSS Final Reports The ultimate product of the Data Life Cycle is an FSS Final Report which will be, to the extent practical, a stand-alone document with minimal information incorporated by reference. To facilitate the data management process, as well as overall project management, FSS Final Reports will usually incorporate multiple FSS Survey Unit Release Records. To minimize the incorporation of redundant historical assessment and other FSS program information, and to facilitate potential partial site releases from the current license, FSS Final Reports will be prepared and submitted in a phased approach. The format and content of the FSS Final Report is as follows:

• Introduction, including a discussion on the phased approach for submittals;

• FSS Program Overview to include sub-sections on survey planning, survey design, survey implementation, survey data assessment, and Quality Assurance and Quality Control measures;

• Site Information to include sub-sections on site description, survey area/unit description (specific to current phase submittal), summary of historical radiological data, conditions at the time of survey, identification of potential contaminants, and radiological release criteria;

• Final Status Survey Protocol to include sub-sections on Data Quality Objectives, survey unit designation and classification, background determination, final status survey plans, survey design, instrumentation (detector efficiencies, detector sensitivities, instrument maintenance and control and instrument calibration), survey methodology, and quality control surveys;

• Survey Findings to include sub-sections on survey data conversion, survey data verification and validation, evaluation of number of sample/measurement locations, and comparison of findings with DCGLs

• Appendix A: Survey Unit Release Records (specific to each phased submittal);

• Appendix B: FSS Program and Implementing Procedures (initial phased submittal - subsequent submittals contain only revisions or additions to program and/or implementing procedures);

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Haddam Neck Plant License Termination Plan

• Appendix C: FSS Technical Basis Documents (initial phased submittal - subsequent submittals contain only revisions or additions to FSS technical basis documents).

5.10 Quality Assurance and Quality Control Measures Connecticut Yankee Atomic Power Company (CYAPCO) has developed and is implementing a comprehensive Quality Assurance Program to assure conformance with established regulatory requirements, set forth by the Nuclear Regulatory Commission (NRC), and accepted industry standards.

The participants in the Connecticut Yankee Quality Assurance Program (CYQAP) assure that the design, procurement, construction, testing, operation, maintenance, repair, and modification of nuclear power plants are performed in a safe and effective manner.

The CYQAP complies with the requirements set forth in Appendix B, of 10 CFR Part 50, along with applicable sections of the Updated Final Safety Analysis Report (UFSAR) for the license application, and is responsive to Regulatory Guide 1.70, which describes the information presented in the Quality Assurance Section of the UFSAR for nuclear power plants. References to specific industry standards for quality assurance and quality control measures governing final status survey activities are reflected in supporting procedures, plans, and instructions.

These Quality Control (QC) and Quality Assurance (QA) measures are integrated into all decommissioning activities, including the development of the LTP and implementation of the final status survey. The CYQAP concepts, as defined in implementing procedures, adequately encompass the risk-significant decommissioning activities. All final status survey activities essential to data quality will be implemented and performed under approved procedures. Effective implementation of administrative controls will be verified through audit activities, with corrective actions being prescribed, implemented and verified in the event any deficiencies are identified. These measures apply to the related services provided by off-site vendors, in addition to on-site sub-contractors.

With regard to the final status survey effort, QA/QC activities will serve to ensure that surveys are performed by trained individuals using approved written procedures and properly calibrated instruments that are sensitive to the suspected contaminant. In addition, QC measures will be taken to obtain quantitative information to demonstrate that measurement results have the required precision and are sufficiently free of errors to accurately represent the site being investigated. QC checks will be performed as prescribed by the implementing procedures required by the CYQAP for both field measurements and laboratory analysis (both on-site and third party). For field measurements, replicate measurements will be made for randomly chosen survey units by a different technician at the same locations as the original measurements. Additionally, the CYAPCO Oversight Organization will be involved in assessing the performance of final status survey activities.

The concepts described in the CYQAP will be applied to the Final Status Survey activities. These activities include the following, as applicable:

August 2004 5-66 Rev. 2

Haddam Neck Plant License Termination Plan Organization The Director of Nuclear Safety/Regulatory Affairs is responsible for ensuring the implementation of site QA programs and processes. The Nuclear Safety Manager directs and administers independent audits, surveillances, and inspections for the final status survey. Both CYAPCO Independent Oversight and Nuclear Safety have the authority to stop unsatisfactory final status survey activities. An organizational chart of the Final Status Survey Group is provided as Figure 5-4.

Ouality Assurance Program The site characterization program used the same QA practices as employed by the operational radiation protection program. These practices included the use of approved procedures for the calibration, testing and use of both laboratory and portable equipment. Trained and qualified personnel collected data. Samples were controlled by administrative procedures to ensure that sample integrity was maintained. When offsite laboratories were utilized, they were required to perform daily instrumentation checks as well as split samples, blank samples, and cross check samples. Performances of these checks by offsite laboratories were verified periodically by QA auditors.

To support future site characterization and the FSS, quality assurance project plans as well, as Data Quality Objectives, will be developed. Documented procedures will be utilized for implementing quality activities at HNP. Additionally, the assignment of documented responsibilities for the conduct of activities affecting quality is defined. Through implementation of these controls, confidence is established that the performance of the FSS will be accomplished in a manner consistent with CYAPCO Policies. It also establishes the commitment that quality activities are performed by trained qualified personnel and that these activities are verified through audits, surveillances, and inspections.

Design Control Design control requirements are established to assure that the applicable regulatory bases, codes, technical standards and quality standards are identified in the Final Status Survey. Design controls including independent verification, and design interface control have been implemented to determine the DCGLs, MDCs, area factors, and other DQO and FSS elements.

Procurement Document Control The procurement of materials, equipment, and services for the FSS are performed in a controlled manner which assures compliance with applicable regulatory requirements, procedures, quality assurance standards and regulations. Service requests will be reviewed for technical adequacy and verification of supplier's quality assurance program will be performed as needed, to assure confidence with services provided. Performance of off-site audits will be used as deemed necessary by administrative controls.

Procedures. Instructions and Drawings The performance of the FSS will require procedures for personnel training, survey implementation, data collection, chain of custody, instrument calibration and maintenance, verification and record storage. These procedures will be developed to ensure compliance with the License Termination Plan and will meet applicable quality requirements. These requirements August 2004 5-67 Rev. 2 l

Haddam Neck Plant License Termination Plan include that the procedures be developed utilizing the guidance of an approved procedure and will receive the appropriate review and approval.

Document Control As stated above, procedures will be written to control the FSS performance. Additionally, procedures will be provided describing the requirements for the control and storage of survey and sample data developed by implementation of the FSS Plan. The results of the FSS will be maintained as records for a minimum of 3 years as required by IOCFR20.2103(a).

Control of Purchased Material, Equipment, and Services Vendors may be used for the performance of the FSS and laboratory activities. Quality-related services, such as instrument-calibration and laboratory analysis, are procured from qualified vendors whose internal QA program is subject to approval in accordance with the CYAPCO Quality Assurance Program. Additionally, audits and surveillance of these contractors will be performed to provide an adequate level of assurance that the quality activities are being effectively performed.

Inspection Inspections and verification activities will be delineated in implementing procedures. These programs and procedures will be used to verify sampling and surveying protocols are appropriately utilized. Inspections will also be conducted on off-site laboratories performing sample analysis for the FSS.

Control of Measuring and Test Equipment Approved procedures will be developed for the use, calibration, and testing of the equipment utilized for the FSS. These procedures will be developed to assure confidence in the data obtained. If additional equipment is procured for the FSS, associated maintenance, calibration, and testing procedures will also be developed. This includes both laboratory equipment and field use equipment. Instrument calibrations will be done periodically in accordance with approved administrative procedures.

Handling, Storage, and Shipping Some of the material samples will be transported to off-site laboratories for analysis. The process for controlling this material will be sufficient to ensure that a chain of custody is maintained.

Additionally, protocols must be established to ensure that there is no cross-contamination between samples and sample packaging. Appropriate controls will be defined in administrative procedures to ensure that sample integrity is maintained.

Nonconforming Materials. Parts. Components, or Services During the performance of the FSS, non-conforming conditions may be identified with equipment or services. The associated data will be segregated until such time that they are accepted, rejected, or reworked in accordance with an appropriate procedure.

August 2004 5-68 Rev. 2

Haddam Neck Plant License Termination Plan Corrective Action The existing Corrective Action Program established under the CYQAP will be utilized for the FSS Program to identify conditions adverse to quality and to support the development of corrective actions.

Quality Assurance Records As stated previously, the FSS records will be maintained in accordance with current administrative controls and will be retained for a minimum of 3 years.

Controls of Special Processes Procedures will be developed to implement special processes in support of FSS implementation.

Validated special processes will be implemented by trained, qualified individuals using approved procedures.

Quality Assurance Audits Audits of FSS activities will be periodically performed to verify the implementation of quality activities.

Inspection, Test, and Operating Status Project controls and schedules will be developed and implemented which identify the status of FSS activities. Measures will ensure identification mechanisms are in place to enable accurate determination of FSS status.

5.11 References 5-1 Code of Federal Regulations, Title 10, Part 20.1402, "Radiological Criteria for Unrestricted Use."

5-2 Draft Regulatory Guide-4006, "Demonstrating Compliance with the Radiological Criteria for License Termination," August 1998.

5-3 NUREG-1575, "Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM),"

dated December 1997.

5-4 Regulatory Guide 1.179, "Standard Format and Content of License Termination Plans for Nuclear Power Reactors," dated January 1999.

5-5 NUREG-1727, "NMSS Decommissioning Standard Review Plan," dated September 2000.

5-6 NUREG-1507, "Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions," December 1997.

5-7 Bechtel Calculation 24265-000-MOC-9000-0007-003, "DCGL Area Factors," R. K. Carr.

5-8 Technical Support Document, BCY-HP-1 05, Rev. 1,"Dose Evaluation for Buried Piping."

August 2004 5-69 Rev. 2

Haddam Neck Plant License Termination Plan 5-9 Technical Support Document, BCY-HP-1 14, Rev. 0, "Dose Comparison of Imbedded Pipe to Building Structures."

5-10 Information Notice 85-92, "Surveys of Wastes Before Disposal from Nuclear Reactor Facilities."

5-11 IE Circular 81-07, "Control of Radioactively Contaminated Material."

5-12 ISO 7503-1 Auwust

- ... 0 .- 2004

_ 5-70 Rev. 2 l

Haddam Neck Plant License Termination Plan Figure 5-4 Final Status Survey Organization President Staus Srve Fina I Nuclear Safety Director of D&D Independent Oversight Group I Site Closure Manager I Site Release Technical lFinal Status Surveyl I RCRA/CAP : : Groundwater 1 Support l Project I Project __

I Support Final Status Survey I

Ire-Demolition Surveys (URS/CVS)I Pf I Site Radiological Characterization I

August 2004 Rev. 2 I

Haddam Neck Plant License Termination Plan 6 COMPLIANCE WITH THE 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, 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) must be calculated. These DCGLs 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 above background is less than or equal to the DCGL.

• Individual measurements, representing small areas of residual radioactivity, which exceed the DCGL, do not exceed the elevated measurement comparison DCGL. The elevated measurement comparison DCGL (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 below those concentrations necessary to meet the DCGLs. (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 August 2004 6-1 Rev. 2

Haddam Neck Plant License Termination Plan of Wastes before Disposal from Nuclear Reactor Facilities", and IE Circular 81-07, "Control of Radioactively Contaminated Material."

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 5500 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.

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 August 2004 6-2 Rev. 2

Haddam Neck Plant License Termination Plan Middletown Formation. These rocks are made of various silicate minerals (quartz, plagioclase, biotite, hornblende, 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.

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.

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 I (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.

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 scenarios also identify the critical group. The critical group is the group of individuals reasonably August 2004 6-3 Rev. 2

Haddam Neck Plant License Termination Plan 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

- Buildings Demolished

- Foundations/Basements It should be noted that the scenarios described in NUREG/CR-55 12, 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 and concrete from demolished buildings.

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, groundwater and concrete from demolished buildings and building foundations/basements and to determine corresponding DCGLs.

Critical Group:

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, groundwater and concrete from demolished buildings and foundations/basements 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.

August 2004 6-4 Rev. 2

Haddam Neck Plant License Termination Plan 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-55 12, Volume 1:

• Direct exposure to external radiation from the residual radioactivity;

• 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-55 12, Volume 1, was selected to estimate human radiation exposure resulting from residual radioactivity in concrete from standing buildings and building foundations/basements and to determine corresponding DCGLs.

Critical Group:

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 August 2004 6-5 Rev. 2

Haddam Neck Plant License Termination Plan 6.4 RESidual RADioactivity WRESRAD) 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 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:

• 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:

• External exposure directly from the sources;

• External exposure to material deposited on the floor;

• External exposure due to air submersion;

• 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.

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

-6697 (References 6-10, 6-1 1, 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 Rernediation and Occupancy of Buildings Contaminated with Radioactive Material" (Reference 6-15).

August 2004 6-6 Rev. 2

Haddam Neck Plant License Termination Plan 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 I parameters 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.

6.5.3 Treatment 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 wvere 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 I 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:

August 2004 6-7 Rev. 2

Haddam Neck Plant License Termination Plan Sample Specifications: The analyses were run using 300 observations and I 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 !PRCCI 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:

• 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 2 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 I 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-5512, 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-August 2004 6-8 Rev. 2

Haddam Neck Plant License Termination Plan 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 from 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.

6.6.2 Conceptual Model 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.

August 2004 6-9 Rev. 2

Haddam Neck Plant License Termination Plan 6.6.4 DCGL Determination Parameter values assigned in the deterministic RESRAD Version 5.91 model are presented in Tables E- I (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 mrem/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 6-1:

Table 6-1 Base Case DCGLs for Soil Radionuclide Soil DCGL (pCi/g)

H-3 4.12E+02 C-14 5.66E+00 Mn-54 1.74E+01 Fe-55 2.74E+04 Co-60 3.81E+00 Ni-63 7.23E+02 Sr-90 1.55E+00 Nb-94 7.12E+00 Tc-99 1.26E+01 Ag-108m 7.14E+00 Cs-134 4.67E+00 Cs-137 7.91 E+00 Eu-152 1.O1E+01 Eu-154 9.29E+00 Eu-155 3.92E+02 Pu-238 2.96E+01 Pu-239 2.67E+01 Pu-241 8.70E+02 Am-241 2.58E+01 Cm-243 2.90E+01 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 August 2004 6-10 Rev. 2

Haddam Neck Plant License Termination Plan 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 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

• Drinking water containing residual radioactivity from a well.

Groundwater flow directions determined by Malcolm Pirnie (1999) are such that the existing plumes 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 = I year

- Time for calculations = I 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:

I1000 S. Pb C 1= (Equation 6-1)

[I +(Kd pb / n)] n August 2004 6-1 1 Rev. 2

Haddam Neck Plant License Termination Plan

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 (cm3 /g) n = Total porosity of contaminated zone (%)

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 radionuclide concentration of I pCi/g was used for the soil comprising the contaminated zone.

The stochastic parameters identified as sensitive (IPRCCI > 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, DCGLGw = DCGL for groundwater (pCi/l)

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 IpCi/l are arbitrary since the only interest is the August 2004 6-12 Rev. 2

Haddam Neck Plant License Termination Plan decay of a unit source) and I 000-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 IpCi/g, were converted into dose, outside of RESRAD, by multiplying by effective Dose Conversion Factors (DCFeff's). The DCFeff'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 Equation 6-4 below. Calculation of the DCF&ff's for each of the radionuclides is shown in Table G-2-1, Footnote "C".

DosepEAK DCFeff= (Equation 6-4)

Concww where, DCFeff = Effective Dose Conversion Factor (mrem/yr/pCi/l)

DosePEAK = Peak Dose at time = 0 yrs (mrem/yr)

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

Following the calculation of dose (using the DCFeff'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 IpCi/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.

August 2004 6-13 Rev. 2

Haddam Neck Plant License Termination Plan Table 6-2 Base Case DCGLs for Groundwater Radionuclide Groundwater DCGL (pci/I)

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 2.51 E+02 Nb-94 6.75E+03 Tc-99 2.64E+04 Ag-108m 4.24E+03 Cs-134 3.42E+02 Cs-137 4.31 E+02 Eu-152 7.33E+03 Eu-154 5.05E+03 Eu-155 3.25E+04 Pu-238 1.511E+01 Pu-239 1.36E+01 Pu-241 4.60E+02 Am-241 1.32E+01 Cm-243 1.94E+01 6.8 DCGLs for Concrete A few of the building basements, at HNP will remain in place and be surveyed for residual radioactivity.

Presently CYAPCO's plan is to backfill these partial structures with clean material, once they are released To ensure that the building surface DCGLs are conservative, two cases were evaluated. Figure 6-6 illustrates the process of determining building surface DCGLs using two scenarios. In one case, the building occupancy scenario was evaluated; and for the other case, the resident farmer (below-grade concrete material) scenario was evaluated. The two scenarios were evaluated separately and the more restrictive DCGL for each radionuclide will be adopted at the time of final status survey, as discussed in Section 5.4.7.1. This ensures that the potential dose to the average member of critical group will be conservatively estimated whether the foundation/basement remains, or the building is demolished and the material will be disposed of at a licensed facility.

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. The residual radioactivity was assumed to be uniformly distributed over all surfaces of a room, including the floor, August 2004 6-14 Rev. 2

Haddam Neck Plant License Termination Plan 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 I m. 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-5512, 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.

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 0.10) 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 (dpm/100 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:

August 2004 6-15 Rev. 2

Haddam Neck Plant License Termination Plan Table 6-3 Base Case Building Surface DCGLs (Building Occupancy Scenario)

DCGL DCGL Radionuclide for Surface Sources for Volumetric Sources (dpmn1 00 cm2 ) (pCi/g)

H-3 3.15E+08 1.47E+03 C-14 1.03E+07 1.18E+08 Mn-54 3.21E+04 9.06E+00 Fe-55 3.49E+07 9.54E+07 Co-60 1.1 E+04 2.90E+ 0 Ni-63 3.60E+07 4.11 E+07 Sr-90 1.27E+05 2.38E+03 Nb-94 1.71 E+04 4.83E+00 Tc-99 1.45E+07 3.09E+07 Ag-108m 1.65E+04 4.84E+00 Cs-134 1.65E+04 4.93E+00 Cs-137 4.30E+04 1.37E+01 Eu-152 2.34E+04 6.70E+00 Eu-154 2.19E+04 6. 1E+00 Eu-155 4.37E+05 3.23E+02 Pu-238 4.87E+03 6.61 E+02 Pu-239 4.44E+03 6.02E+02 Pu-241 2.29E+05 3.12E+04 Am-241 4.27E+03 4.16E+02 Cm-243 6.07E+03 7.53E+01 6.8.2 DCGLs for Concrete: Buildings Demolished (Concrete Debris)

After completion of final status survey activities of the remaining portions of structures, some concrete debris may remain. However, the current plans are to use clean backfill material and not concrete debris to backfill basements and restore the site to grade elevation. Previously it was assumed that concrete debris generated from the demolition of the buildings and any additional decontaminated concrete structures would be placed in the basements of the buildings, to a depth of approximately 3 feet below the ground surface. And approximately 3 feet of clean backfill soil would be placed over the concrete debris.

This concrete debris was considered to contain residual radioactive material. All building basements were assumed to operate as open-systems, allowing in-flow and out-flow of groundwater.

Although current decommissioning plans do not call for the placement of concrete debris in facility basements, the methodology outlined below will be conservatively applied for remaining basement structures.

6.8.2.1 Dose Model The resident farmer scenario is selected to estimate human radiation exposures resulting from residual radioactivity in concrete debris and to determine corresponding DCGLs. The residual radioactive materials are assumed to be contained in a subsurface layer of concrete debris on property that can be August 2004 6-16 Rev. 2 I

Haddam Neck Plant License Termination Plan used for residential and light farming activities. 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 drinks water from a groundwater source onsite.

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

• Direct exposure to external radiation from the concrete debris containing residual radioactivity;

• Internal dose from inhalation of airborne radionuclides; and

• Internal dose from ingestion of

- Plant foods grown in the soil cover and irrigated with water containing residual radioactivity;

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

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

- Concrete debris containing residual radioactivity.

Groundwater flow directions determined by Malcolm Pirnie (1999) are such that any radionuclides present in the groundwater underlying the industrial area would migrate toward the Connecticut River.

The flow rate of groundwater that may potentially contain residual radioactivity, relative to the flow rate of the Connecticut River, should be very small. Therefore, the aquatic foods ingestion pathway is not considered applicable in this calculation.

6.8.2.2 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 and the initially planned disposition of concrete debris in the basements of the buildings. Key assumptions associated with that conceptual model are as follows:

(i) All concrete debris contains residual radioactivity and this concrete debris comprises the contaminated zone. The contaminated zone extends below the water table, based on an ambient water table elevation of approximately 10 ft msl (Malcolm Pirnie, 1999).

(ii) Approximately 3 ft of soil fill is initially placed on top of the concrete debris. This soil does not contain residual radioactivity.

(iii) The residential farmer constructs his/her home over a debris-filled basement.

(iv) The well that supplies water for drinking, crop irrigation, and livestock is drilled and completed within the debris-filled basement.

The conceptual model described above bounds site characteristics expected at the time of release under the current decommissioning approach.

The parameters used to quantify the conceptual model are presented in Appendix D, Table D4. The values / distributions assigned to each of the parameters, the basis for assigning such values / distributions and the relevance of the parameter to the dose calculations are presented in Table D-4.

August 2004 6-17 Rev. 2

Haddam Neck Plant License Termination Plan The distributions and defining statistical parameters assigned to the distribution coefficients (Kd) and plant transfer factors for the concrete debris were determined from site-specific data contained in a report produced by Batelle, Pacific Northwest Division (PNWD) (Reference 6-17). Batelle performed tests on concrete cores taken from the Containment Building and the Waste Disposal Building and used groundwater from the site for the tests.

The conceptual model underlying the dose model includes a contaminated zone above and below the water 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 following parameters:

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

- Number of unsaturated zone strata = 0 By establishing the above parameters, the groundwater (well water) concentrations calculated by RESRAD were found to be greater than or equal to the groundwater concentrations in equilibrium with the concrete debris under saturated conditions. The equilibrium groundwater concentration associated with concrete debris was calculated using Equation 6-1 in Section 6.7.2 for a contaminated zone comprising concrete debris.

6.8.2.3 Sensitivity Analysis Results Parameter distributions assigned in the probabilistic RESRAD Version 6.1 model are presented in Appendix D, Table D-4. An initial radionuclide concentration of I pCi/g was used for the concrete debris 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-4. For each radionuclide, the sensitive parameters are listed in order of decreasing sensitivity. Included in Table E4 are the conservative values assigned to each of the sensitive parameters.

6.8.2.4 DCGL Determination Parameter values assigned in the deterministic RESRAD Version 5.91 model are presented in Table E-4 (conservative values assigned to parameters shown to be sensitive) and Appendix F, Table F-4. The groundwater DCGLs were determined for a radiation dose limit of 25 mrem/yr.

The concrete debris DCGLs calculated for each of the radionuclides are presented in Appendix G, Table G-4-1. The time to the peak of the mean dose is included in the table together with the percent contribution to dose from the exposure pathways (water independent and water dependent). The equilibrium groundwater concentrations associated the concrete debris and the groundwater (well water) concentrations for a known concentration of radioactive material in the concrete debris are presented in Appendix G-4-2. The concrete debris DCGLs are summarized in Table 6-4.

August 2004 6-18 Rev. 2

Haddam Neck Plant License Termination Plan Table 6-4 Base Case DCGLs for Building Demolished (Concrete Debris)

Radionuclide Concrete Debris DCGLs (pCi/g)

H-3 9.05E+O l C-14 2.05E+O0 Mn-54 5.51E+OI Fe-55 8.96E+Ol Co-60 9.07E+OI Ni-63 1.29E+02 Sr-90 3.77E-Ol Nb-94 7.74E+OO Tc-99 2.85E+O1 Ag-108m 2.59E+O1 Cs-134 3.21E+02 Cs-137 6.45E+02 Eu-152 2.27E+02 Eu- 154 1.94E+02 Eu-155 9.53E+03 Pu-238 1.14E+O1 Pu-239 1.OOE+O1 Pu-241 1.49E+02 Am-241 4.42E+OO Cm-243 3.83 E+OO 6.8.3 Concrete DCGL Conversion Table 6-3 shows the DCGLs calculated for building surfaces using the building occupancy scenario, and Table 6-4 shows the DCGLs calculated for concrete debris using the resident farmer scenario. Note that the units for surface sources from the building occupancy scenario are in units of dpm/lOOcm 2 , whereas, for volumetric sources from the resident farmer concrete debris scenario the units are pCi/g. In order to determine operational DCGLs, as discussed in Section 5.4.7.1, a method of converting the volumetric DCGLs to surface DCGLs (and vice versa) is needed.

This conversion is performed by assuming that the entire quantity of radioactivity within the volume occupied by the available fill area up to 3 feet below grade, V, is distributed on the internal surface area of the building, Ab. The value of Ab is determined based on the surface area of concrete prior to demolition.

This conversion is performed as follows.

August 2002 6-19 Rev. I

Haddam Neck Plant License Termination Plan DCG4(dpnz/lOOcoI2) = DCG4V(pCig)

• A(m,3)

• 106 cmI

• p(g/crn3 ) * (1-In)

• l,,2

• 2.22dpm

• 100 A4b (r 2) m3 104Crn 2 POi

( Equation 6-4) where: p is the density for concrete, or 2.4 g/cm 3 , and n is the porosity for the buried concrete debris, or 0.3.

100 is used to convert from cm2 to 100 cm2 Substituting these values and combining terms provides the following relationship:

DCGL,(dpm /1 00cm2 ) = 37,296

• DCGLCOfC(pCi/g)* V(m113 ) (Equation 6-5)

Ab~n2 ) (qain65 The minimum value of V/Ab has been evaluated and documented in the Technical Support Document (Reference 6-18). The minimum value of V/Ab was determined to be for the Containment Building where V = 5790 m3 , and Ab= 16764 m2 .

The value of Ab represents the surface area of concrete within the containment building. The values presented in LTP Table 2-10 represent all structural surfaces within the containment, including steel surfaces that will be included in the final status survey.

Using minimum values therefore ensures that additional conservatism is applied to all plant structures subject to concrete demolition. Substituting these values yields the following relationship for the DCGLs;:

DCGL, (dpm / 1OOcnm2 ) = 12,881 *DCGL_(pCiI g) (Equation 6-6)

The volumetric concrete debris DCGL values are converted to surface activity DCGL values using the above relationship. These surface activity DCGL values are conservative since all of the concrete debris within the backfill volume is assumed to contain residual radioactivity at a level corresponding to the volumetric DCGL. This is unlikely since a large fraction of internal surfaces of the plant structures contain very low levels of residual radioactivity. This total activity is then distributed uniformly over all internal surfaces of the building. The resulting surface DCGL is conservative considering that no allowances are made for the large area occupied by the non-contaminated surface areas and continues to bound the site characteristics expected at site release considering the current decommissioning approach.

August 2004 6-20 Rev. 2

Haddam Neck Plant License Termination Plan Table 6-5 Conversion of Base Case Concrete Debris DCGLs Radio- Base Case DCGLs Base Case DCGLs nuclide (pCig) (dpm/100 cm2)

H-3 9.05E+01 1.17E+06 C-14 2.05E+01 2.64E+05 Mn-54 5.51 E+0 1 7.1 OE+05 Fe-55 8.96E+0I 1.15E+06 Co-60 9.07E+01 1.17E+06 Ni-63 1.29E+02 1.66E+06 Sr-90 3.77E-01 4.86E+03 Nb-94 7.74E+00 9.97E+04 Tc-99 2.85E+01 3.67E+05 Ag-108r_ 2.59E+01 3.34E+05 Cs-134 3.21E+02 4.13E+06 Cs-137 6.45E+02 8.31 E+06 Eu-152 2.27E+02 2.92E+06 Eu-154 1.94E+02 2.50E+06 Eu- 155 9.53E+03 1.23E+08 Pu-238 1.14E+01 1.47E+05 Pu-239 l.OOE+O I 1.29E+05 Pu-241 _ 1.49E+02 1.92E+06 Am-241 4.42E+00 5.69E+04 Cm-243 3.83E+00 4.93E+04 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 buried debris scenario), a method to assess doses from these combined pathways is necessary. Additionally, any 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 the dispositioned concrete debris, which may also contain residual radioactivity. Soil and building surface DCGLs for these combined scenarios will be determined on an operational basis, using the base case DCGLs for soil, groundwater, and building surfaces, 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.

August 2002 6-21 Rev. I

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

6-4 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.

6-6 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 Technical Support Document, BCY-HP-0018, Building Concrete Surface Area and Volume Determination.

August 2002 6-22 Rev. I

Haddam Neck Plant License Termination Plan Figure 6-6 Process for Determining Building Surface DCGLs Historical Site Assessment &

Site Characterization Sensitivity Analysis and Parameter Selection I

Unit Concentration of Each Individual Radionuclide Plant Structures I Left Standing Evaluate/Select Pathways for Plant Structures Radionuclides Demolished I

Resident Farmer I Building Occupancy Scenario I (Concrete Debris) Scenario RESRAD 5.91 Using Site Specific and Conservative Input Data RESRAD-BUILD 2.37 Using (Appendix E)

Site Specific and Conservative Input Data Determine Base Case Building (Appendix E)

Demolished (Concrete Debris)

DCGLs that Result in 25 mrem/yr (As required)

Determine Base Case (Table 6-4)

Building Standing DCGLs that Result in 25 mrem/yr Convert Volumetric DCGLs to Pre-(Table 6-3) Demolition Surface DCGLs (Table 6-5) (As Required)

Determine Operational DCGL for Buildings Using Equation 5-5 August 2004 Rev. 2

Haddam Neck Plant License Termination Plan 7 UPDATE OF SITE-SPECIFIC DECOMMISSIONING COSTS 7.1 Introduction In accordance with 10 CFR 50.82(a)(9)(ii)(F) and Regulatory Guide 1.179, the site specific cost estimates and funding plans are provided. Regulatory Guide 1.179 discusses the details of the information to be presented.

The License Termination Plan (LTP) must:

Provide an estimate of the remaining decommissioning costs, and compare the estimated costs with the present funds set aside for decommissioning. The financial assurance instrument required by IOCFR50.75 (Reference 7-1) must be funded to the amount of the cost estimate. If there is a deficit in the present funding, the LTP must indicate the means for ensuring adequate funds to complete the decommissioning.

The decommissioning cost estimate should include an evaluation of the following cost elements:

• Cost assumptions used, including contingency

• Major decommissioning activities and tasks

• Unit cost factors

• Estimated decontamination and equipment and structure removal

• Estimated cost of radioactive waste disposal including disposal surcharges

• Estimated final survey costs

• Estimated total costs The cost estimate should focus on the remaining work, detailed activity by activity, including costs of labor, materials, equipment, energy, and services.

During plant operations, CYAPCO sold the entire electrical output of the Haddam Neck Plant (HNP) to wholesale power purchase contracts (i.e., Power Contracts) with the ten New England utilities that collectively own 100% of the common equity of CYAPCO (the "Customers"). Over the HNP's operating life, CYAPCO recovered, and since the shutdown continues to recover, its costs of providing service (including the estimated costs of decommissioning HNP) through a formula rate set forth in its Power Contracts. Collections for decommissioning have been placed in a trust established under Connecticut law, with two funds-the Qualified Fund and the Non-Qualified Fund (the "Decommissioning Trust").

On December 26, 1996, CYAPCO submitted amendments to its Power Contracts (the "Amendatory Agreements") for filing (Reference 7-2). These Amendatory Agreements were executed to implement the decision to cease operations permanently. Also in that filing, CYAPCO proposed to increase its annual decommissioning collections in order to allow CYAPCO to recover the estimated costs to decommission the HNP.

This cost estimate was prepared by Northeast Utilities Service Company (NUSCo) using the TLG Services, Inc., estimating model in accordance with 10 CFR 50.82(a)(8)(iii). The assumed method of decommissioning anticipated a prompt decommissioning technique commonly referred to as DECON.

August 2004 7-1 Rev. 2

Haddam Neck Plant License Termination Plan 7.2 Decommissioning Cost Estimate 7.2.1 Cost Estimate Previously Docketed in Accordance with 10 CFR 50.82 and 10 CFR 50.75 Post Shutdown A letter from CYAPCO to the USNRC dated August 22, 1997 (Reference 7-3) was docketed detailing the PSDAR. This submittal contained a site-specific decommissioning cost estimate. In accordance with 10 FR 50.82(a)(8)(iii), CYAPCO submitted the site-specific decommissioning cost estimate to the USNRC in a letter dated August 25, 1998 (Reference 74). The estimate included with the PSDAR had been filed with the FERC prior to docketing. This estimate was the basis for which CYAPCO collected its rates and its anticipated decommissioning expenses. This cost estimate was prepared by NUSCo in 1996 using the TLG model for estimating decommissioning costs. It was prepared in sufficient detail to identify an activity by activity work breakdown complete with costs for radioactive waste, utility labor, contractor labor, energy, materials and equipment.

In August of 1998, the presiding Administrative Law Judge (ALJ) in the FERC case issued an opinion to the FERC Commissioners that the estimate was "unreliable for rate-making" and recommendations were made identifying appropriate remedies based on the factual data submitted from CYAPCO and intervening parties. CYAPCO prepared a revised cost estimate and engaged in settlement discussion with the certain intervening parties to the rate case. A settlement agreement incorporating the revised decommissioning funding plan was reached in April 2000 and submitted to the FERC for their review and approval. The settlement was approved by the FERC in April 2000.

7.2.2 Summary of the Site Specific Decommissioning Cost Estimate The FERC approved decommissioning cost estimate is based on the April 7, 2000 Offer of Settlement between CYAPCO and the Connecticut Department of Public Utility Control and the Connecticut Department of Consumer Counsel (Reference 7-5). A copy of the FERC-approved Settlement Offer is provided in Appendix C. This FERC-approved decommissioning cost estimate completely replaces and supersedes the 1996 cost estimate prepared by Northeast Utilities Service Company using the TLG Services estimating model.

As stated in the FERC settlement, CYAPCO executed a contract with Bechtel Power Corporation on April 3, 1999 to perform the decommissioning and dismantlement operations on a fixed price, turnkey basis (the contract also provided for implementation of an ISFSI). CYAPCO commissioned a detailed independent decommissioning cost analysis by Nuclear Energy Services Inc. (NES) as part of its evaluation of the contractor's proposals. Special attention was paid to the LLW disposal estimates from the bidders and their initial estimates were evaluated for soil remediation and disposal. In the aggregate, the Decommissioning Operations Contractor (DOC) candidate bids were in reasonable agreement with the independent cost estimate.

The scope of the independent cost estimate was equivalent to the base work scope proposed for the DOC.

The contractor work scope assumes CY continues to operate and maintain the spent fuel island, hold the NRC license and provide oversight to the DOC. Site-specific costs, such as insurance costs, property taxes, and decommissioning oversight costs, are also the responsibility of CY. The DOC was responsible for all other costs associated with decommissioning activities. A Critical Path Method analysis was performed, incorporating the activity schedule hours to determine the decommissioning schedule. The schedule was bound by an April 2004 completion date. From this schedule, the CY and DOC staff requirements and costs were determined. The schedules were also used to determine all other period dependent costs such as small tool allowances, health physics supplies, energy requirements and security.

August 2004 7-2 Rev. 2

Haddam Neck Plant License Termination Plan The activity dependent and period dependent costs were added together and a contingency applied to arrive at the total decommissioning cost estimate.

In July of 2003, CYAPCO terminated the DOC contract after a 30 day cure period. Following termination CYAPCO restarted decommissioning activities by subcontracting various elements of the work and performing project management and engineering activities. CYAPCO decommissioning plans are described in Chapter 3.

In December of 2003, CYAPCO updated its Decommissioning Cost Estimate to incorporate the most recent assumptions with respect to the remaining decommissioning activities and related costs. CYAPCO filed an update to the decommissioning cost estimate together with updated decommissioning charges to its wholesale purchasers, with the FERC on July 1, 2004 (Reference 7-8). The updated revenue requirements are reflected in Table 7-2 and are proposed to be effective, subject to refund, January 1, 2005.

Table 7-1 identifies, as of January 1,2003, that the remaining cost to complete NRC required decommissioning activities, based on the updated decommissioning cost estimates, is $170.7 million.

The $170.7 million is comprised of $106.2 million for dismantlement and decontamination and $64.5 million for radioactive waste.

Table 7-1, also identifies: the long term spent fuel costs through 2023 ($317.9 million), and site restoration costs ($99.8 million).

August 2004 7-3 Rev. 2

Haddam Neck Plant License Termination Plan Table 7-1 Actual and Project Decommissioning Expenditures (SMillions)

Cost Categories Total 1997 - 2002 Dismnantlement and Decontamination S 326.9 2003 "To-go " Decommissioning Cost Estimate: 2003 - 2023 Cost Elements [Al:

(1) Dismantlement and Decontamination S 106.2 (2) Radioactive Waste Costs $ 64.5 Subtotal Dismantlement and Decontamination and Waste (1+2) $ 170.7 (3) Long Term SNF Storage $ 317.9 (4) Site Restoration $ 99.8 (5) Final Status Survey (FSS) $ 14.9 Subtotal SNF Storage, Site Restoration and FSS (3+4+5) $ 432.6 Total 2003 "To-go" Decommissioning Cost Estimate 2003 - 2023 S 603.3 Total Decommissioning Cost Including Incurred and Estimated Cost JBJ $ 930.2

[A] 2003 "To-go" cost estimate is stated in year 2003 dollars.

[B] Cost total excludes the FERC approved consolidation of the pre-1983 spent fuel fee obligation of $152.7M (principal & interest as of December, 2003) which was deposited into the decommissioning trust fund. The liability increases according to the spent fuel obligation interest rate requirements and the spent fuel obligation asset accretion assumptions are within the decommissioning funding model.

August 2004 7-4 Rev. 2

Haddam Neck Plant License Termination Plan Table 7-2 DECOMMISSIONING/SPENT FUEL TRUST ANALYSES (in Thousands of Dollars)

(Column 1) (Column 2) (Column 3) (Column 4) (Column 5)

After Total Decom Decommissioning Tax/Eamings Total 'To Go' Period Ending Contributions Expenses and Adjust Decom. Trust Cost Estimate 31-Dec-02 $222,271 $730,291 31-Dec-03 16,742 (99,326) 12,193 151,880 630,965 31-Dec-04 16,742 (146,453) 39,965 62,134 484,512 31-Dec-05 93,002 (119,117) 22,264 58,282 365,395 31-Dec-06 93,002 (89,047) (3,909) 58,329 276,348 1

31-Dec-07 93,002 (23,616) (9,805) 117,910 252,732 31-Dec-08 93,002 (13,982) (12,282) 184,648 238,750 31-Dec-09 93,002 (10,311) (9,824) 257,516 228,439 31-Dec-10 93,002 (16,837) (3,217) 330,464 211,602 31 -Dec-1I (13,347) 23,316 340,433 198,256 31-Dec-12 (14,288) 28,595 354,740 183,968 31-Dec-13 (14,703) 30,183 370,220 169,265 31-Dec-14 (15,016) 31,561 386,765 154,249 31-Dec-15 (15,859) 32,475 403,381 138,390 31-Dec-16 (16,216) 33,583 420,748 122,174 31-Dec-17 (17,176) 32,141 435,714 104,998 31-Dec-18 (17,909) 25,877 443,682 87,089 31-Dec-19 (18,269) 27,118 452,531 68,820 31-Dec-20 (19,363) 25,398 458,567 49,458 2

31-Dec-21 (434,675) 13,953 37,844 38,913 31-Dec-22 (25,519) 848 13,174 13,394 31-Dec-23 (13,393) 219 0 0 Totals $591,497 ($1,154,420) $340,652 TOTAL FUNDS AVAILABLE so

' Decommissioning completed except for OTHER NOTES:

ISFSI. - Column 2 Includes contingency and escalation.

2 The amounts in column 2 for the 2021 to 2023 time frame assume that - The cost estimate does not include the DOE the payment obligation to the DOE for the pre-1983 spent fuel fee would be paid pre-1983 spent fuel fee obligation which is in 2021 with fuel removed completed during payable when DOE first begins removing fuel 2021, followed by ISFSI decommissioning in 2022 and complete dissolution in 2023. as described in note 2.

August 2004 7-5 Rev. 2

Haddam Neck Plant License Termination Plan 7.2.3 Dismantlement and Decontamination The decommissioning method being used provides for the demolition of all site structures to 4 feet below grade level elevation. Foundations or building basements that remain will be decontaminated, surveyed and left in place. This is limited to the containment building and portions of the spent fuel building, the main circulating water tunnel structure, and miscellaneous foundations and footings. The waste materials generated during the demolition process will be disposed of at regulated land fills or licensed radioactive disposal facilities as appropriate. Following the demolition and disposal the remainder of the site will be surveyed to insure compliance with the LTP. The HNP license will be terminated except for the dry fuel storage facility.

7.2.4 Radiological Waste Disposal Radiological waste disposal includes: preparation, packaging, transportation and disposal of all forms of low-level radioactive wastes. The large components from the nuclear steam supply system include: the four steam generators, pressurizer, reactor coolant pumps and piping, and the reactor vessel. Table 3-1 contains a list of major systems and components that have been or are to be removed. Table 3-4 provides projections of waste quantities for decommissioning.

The large majority of the waste is Class A waste which is either sent to a licensed waste processing facility or a suitable disposal facility. The rates for these facilities are comparable to or lower than the published rates for the Barnwell facility. The portion of the waste that has gone or will go to Barnwell disposal facility consists mainly of Class B and C waste (e.g., resin liners) and large components (e.g.,

reactor vessel, steam generator lower assemblies).

7.2.5 Long-Term Spent Fuel Storage In parallel with the final phase of decommissioning, an Independent Spent Fuel Storage Installation (ISFSI) has been constructed and is operational for the long-term storage of the spent nuclear fuel. As of January 1,2003, the remaining cost for long term fuel storage through 2023 is $ 317.9 million. The cost estimate is in July 2003 dollars.

7.2.6 Site Restoration and License Termination The current estimate for site restoration includes the costs to perform the final site survey, restoration of the site, and terminate the Part 50 license. This cost includes the costs associated with non-radiological remediation required by Federal and State agencies for such items as RCRA and TSCA closure, asbestos disposal, etc.

7.3 Decommissioning Funding On July 18, 1990, CYAPCO submitted to the NRC a report as required by I OCFR50.75, indicating how reasonable assurance will be provided for funds to decommission the facility (Reference 7-6). The report described how CYAPCO has established an external sinking fund in 1984 to accumulate decommissioning funds. CYAPCO certified that each owner agreed to be financially responsible for its share of the decommissioning costs pursuant to the terms of the Power Contracts and Amendatory Agreements in accordance with the FERC regulations. These contracts have been filed with and approved by FERC. The Power Contracts and Amendatory Agreements were attached to the report.

August 2004 7-6 Rev. 2

Haddam Neck Plant License Termination Plan On March 31, 2000, CYAPCO provided the most recent status report on the decommissioning fund to the NRC in accordance with I0CFR50.75 (Reference 7-7). This report restated the obligation that each wholesale power purchaser is responsible for its share of the facility decommissioning costs pursuant to the Power Contracts regardless of when the costs occur.

Under the April 2000, Offer of Settlement agreement, CYAPCO was required to file with the FERC no later than July 1, 2004, for the purpose of examining any further rate adjustments and is not limited to the future cost of spent fuel storage. CYAPCO filed an update to the decommissioning cost estimate (summarized in Table 7-1) with the FERC on July 1, 2004. The updated revenue requirements for this estimate are reflected in Table 7-2 and are assumed to be effective January 1, 2005.

All decommissioning physical activities are scheduled for completion by year end 2006. All spent fuel is expected to be transferred to the ISFSI by early 2005. The long term spent fuel storage costs after 2005 consist of the operation, and maintenance of the ISFSI, as well as decommissioning of the ISFSI. As shown in Table 7-2, sufficient funding wvill exist, based on CYAPCO's assumption that the DOE will complete spent fuel storage and removal by 2023.

Finally, as demonstrated in Table 7-2, CYAPCO shows that sufficient funds will be collected and will be available to complete decommissioning upon approval of the FERC rate filing filed with the FERC on July 1,2004. Reliance upon the availability of the trust funds is not the sole basis for assurance that decommissioning costs will be paid when incurred. Assurance comes primarily from the power contract obligations of the owners of CYAPCO. Pursuant to 10CFR50.75 and 10CFR50.82 regulations, CYAPCO has demonstrated a financial plan which includes adequate reserves for the entire decommissioning and ISFSI-related costs, which therefore meet the requirements for decommissioning costs associated with decommissioning and dismantlement as defined by these regulations.

7.4 References 7-1 Code of Federal Regulations, Title 10, part 50.75, "Reporting and Recordkeeping for Decommissioning Planning."

7-2 Letter CYAPCO to FERC, "Amendatory Agreements," dated December 26, 1996.

7-3 Letter CYAPCO to USNRC, "Post Shutdown Decommissioning Activities Report," dated August 22, 1997, October 22, 2002.

7-4 Letter CYAPCO to USNRC, "Site-Specific Decommissioning Cost Estimate," dated August 25, 1998.

7-5 Letter CYAPCO to FERC, "Offer of Settlement," dated April 7, 2000, as supplemented by letter dated April 27, 2000.

7-6 Letter CYAPCO to USNRC, "Decommissioning Financial Assurance Certification Report,"

dated July 18, 1990.

7-7 Letter CYAPCO to USNRC, "Decommissioning Funding Assurance," dated March 31, 2000.

7-8 Letter CYAPCO to FERC, "Revision to CY Wholesale Power contract," dated July 1, 2004.

August 2004

__ _. 7-7

. . Rev... _2

Haddam Neck Plant License Termination Plan 8 SUPPLEMENT TO THE ENVIRONMENTAL REPORT 8.1 Introduction 8.1.1 Overview The Connecticut Yankee Power Company Decommissioning Environmental Review (Reference 8-1), dated 1997, was prepared and submitted in conjunction with the HNP Post-Shutdown Decommissioning Activities Report (Reference 8-2). The Environmental Review was previously provided to Federal and State agencies. The report concluded that the environmental impacts of decommissioning activities are bounded by previously issued environmental impact statements-NUREG-0586, "Final Generic Environment Impact Statement (FGEIS) on Decommissioning of Nuclear Facilities," (Reference 8-3); Final Environmental Statement, Haddam Neck Nuclear Power Plant, Docket No. 50-213, October, 1973 (Reference 8-4); and "Environmental Assessment for Proposed License Extension," dated November 23, 1987 (Reference 8-5).

The purpose of this section of the LTP is to describe any new information on significant environmental impacts associated with site-specific license termination activities and to determine if these impacts are within the scope of the environmental impacts previously evaluated either generically or on a site-specific basis by:

1. the environmental impact statement developed in support of the original facility,

2. the environmental impacts described in conjunction with the Decommissioning Plan (and PSDAR) related to decommissioning activities, or

3. The Final Generic Environmental Impact Statement addressing decommissioning (NUREG-0586).

The NRC has issued guidance associated with the impacts of decommissioning, including Supplement I to NUREG-0586 (Reference 8-6). Supplement 1 to NUREG-0586 focuses on the impacts of decommissioning nuclear power reactors licensed by the NRC, unlike the 1988 FGEIS, which took a broad look at decommissioning of a variety of sites and activities.

Supplement 1 to NUREG-0586 is intended to consider, in a comprehensive manner, all aspects related to the radiological decommissioning of nuclear reactor facilities. Supplement l uses an approach that defines a measure of significance and severity of potential environmental impacts and an applicability of these impacts to a variety of facilities. The significance of an impact is described as being SMALL, MODERATE, or LARGE. The applicability of impacts is described as being generic or site-specific. These terms are clearly defined in Section 4 of Supplement I to NUREG-0586.

Table H-1, located in Appendix H to Supplement I of NUREG-0586, provides a listing of activities for which the NRC has generically determined that no environmental impacts exist.

Because these activities have already been determined not to result in environmental impacts, no further review is required in connection with the LTP.

8-I Rev. 2 August 2004 2004 8-1 Rev. 2

Haddam Neck Plant License Termination Plan Table H-2 provides a summary of the decommissioning activities and associated environmental issues that have been determined to have potential impacts. As stated in Section 4.3 of Supplement 1 to the NUREG-0586, if these plant-specific impacts fall within the scope of the environmental impacts previously identified and evaluated by the NRC staff, these activities can be performed without further evaluation. The issues identified in Table H-2 to be evaluated for plant-specific impacts are:

• Onsite/offsite land use

• Water use

• Water quality

• Air quality

• Aquatic ecology

• Terrestrial ecology

• Threatened and endangered species

• Radiological

• Radiological accidents

• Occupational

. Socioeconomics

• Environmental justice

• Cultural impacts

• Aesthetics

. Noise

• Transportation

• Irretrievable resources.

According to Supplement 1 to NUREG-0586, the NRC assessed the impacts of each of these issues using data from previous studies and environmental reviews in addition to information obtained during site visits and provided by plants undergoing decommissioning. The NRC then examined the cumulative impacts of decommissioning activities and other past, present, and reasonably foreseeable future activities at the sites. After analyzing the issues, the NRC determined the impact of each and assigned a significance level (SMALL, MODERATE, or LARGE).

The NRC also determined whether the analysis of the environmental issues could be applied to all plants. Each environmental issue identified was assigned one of the following two categories:

generic or site-specific.

8-2 Rev. 2 August 2004 8-2 Rev. 2

Haddam Neck Plant License Termination Plan Generic issues met the following three criteria:

1. The environmental impacts associated with the issue have been determined to apply to all plants, or, for some, issues to a group of plants of a specific size, specific locations, or having a specific type of cooling system or site characteristic.

2. A single significance criterion (SMALL, MODERATE, or LARGE) has been assigned to describe the impacts.

3. Mitigation of adverse impacts associated with the issue has been considered in the analysis, and it has been determined that additional plant-specific mitigation measures are likely not to be sufficiently beneficial to warrant implementation.

If one or more of the above criteria cannot be met, the issue is considered to be "site-specific" and a site-specific evaluation of the issue is required. Table 8-1 summarizes the NRC's findings with respect to applicability and impact of the identified environmental issues pertinent to decommissioning.

Decommissioning and license termination activities at HNP fall within the range of activities evaluated for the FGEIS and NUREG-0586, Supplement 1. For those issues identified as "generic" in Table 8-1, the NRC's prior conclusions bound environmental impacts at HNP from decommissioning and license termination.

The LTP addresses the issues identified in Table 8-1 as "site-specific." In addition, consistent with regulatory guidance, the review focuses on any new information or significant environmental change associated with site-specific termination issues. Impacts associated with site-specific termination activities have been compared to previously analyzed decommissioning and termination activities, in this LTP and its references. The proposed termination activities related to the end use of the site do not result in significant environmental changes that are not bounded by the site-specific decommissioning activities described in the Decommissioning Plan, PSDAR, the FGEIS, or NUREG-0586.

Note that the review and conclusion in this Section relate only to activities and impacts associated with termination of the NRC license. CYAPCO is conducting other site characterization at HNP for non-radiological remediation and site restoration, which are not part of the license termination activities and are outside of the scope of NRC regulation. The non-radiological remediation activities are addressed in the RCRA Corrective Action Program (CAP) that is being reviewed by the EPA. The Property Transfer Program regulations from the Connecticut Department of Environmental Protection will be used to achieve site closure. This program will incorporate the remediation activities that occurred under the LTP and RCRA CAP.

Other agencies, such as the Connecticut Department of Public Health, the Army Corps of Engineers and the Haddam Wetlands Commission are also routinely involved in aspects of non-radiological site remediation.

8-3 Rev. 2 2004 August 2004 8-3 Rev. 2

Haddam Neck Plant License Termination Plan 8.1.2 Proposed Site Conditions at the Time of License Termination The HNP site is intended to be released for unrestricted use, under the radiological release criteria of 10CFR20.1402 (Reference 8-7) upon termination of its NRC license. Sections 3 and 4 of this LTP discuss in greater detail the activities that have been completed, those ongoing and remaining, and the proposed final state of the site.

8.1.3 Remaining Dismantlement and Decommissioning Activities 8.1.3.1 General Description of Remaining Dismantlement and Decommissioning Activities CYAPCO continues to implement the DECON alternative as the most appropriate alternative for decommissioning the HNP site. Evaluation of the environmental effects of the DECON alternative is contained in NUREG-0586 and its supplement.

8.1.3.2 Conclusions Regarding Environmental Impact Included in the PSDAR The PSDAR includes a discussion of environmental impacts from decommissioning. This information was based upon NUREG-0586; "The Connecticut Yankee Power Company Decommissioning Environmental Review," dated 1997; the "Final Environmental Statement, Haddam Neck Nuclear Power Plant, Docket No. 50-213," dated October 1973; and the "Environmental Assessment for Proposed License Extension," dated November 23, 1998.

The PSDAR concluded that the impacts due to decommissioning would be bounded by the previously issued environmental impacts statements. This was principally due to the following reasons:

• The postulated impacts associated with the method chosen, DECON, have already been considered in Supplement I to NUREG-0586.

• There are no unique aspects of the plant or decommissioning techniques to be utilized that would invalidate the conclusions reached in Supplement 1 to NUREG-0586.

• The methods to be employed to dismantle and decontaminate the site are standard construction-based techniques fully considered in Supplement 1 to NUREG-0586.

• The site-specific person-rem estimate for all decommissioning activities has been conservatively calculated using methods similar to those used in Supplement I to NUREG-0586.

Specifically, the review concluded that the HNP decommissioning will result in generally positive environmental effects, in that:

Radiological sources that create the potential for radiation exposure to site workers and the public will be minimized.

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Haddam Neck Plant License Termination Plan

• The site will be returned to a condition that will be acceptable for unrestricted use.

• The thermal impact on the Connecticut River from facility operations will be eliminated.

• Noise levels in the vicinity of the facility will be reduced.

• Hazardous material and chemicals will be removed.

• Local traffic will be reduced (fewer employees, contractors and materials shipments than required to support an operating nuclear power plant).

Furthermore, the HNP decommissioning will be accomplished with no significant adverse environmental impacts in that:

• No site specific factors pertaining to HNP will alter the conclusions of Supplement 1 to NUREG-0586

• Radiation dose to the public will be minimal.

• Radiation dose to decommissioning workers will be a fraction of the operating exposure.

• Decommissioning is not an imminent health or safety problem and will generally have a positive environmental impact.

Revisions 0 through 2 of the PSDAR estimated the total occupational exposure (excluding public and transportation dose) for the proposed decommissioning activities to be approximately 935 person-rem. Since the original estimate was made, a significant amount of the decommissioning tasks have been completed. A current estimate of the total occupational exposure as of March 2004 is within 10% of the original estimate (Reference 8-8). The current estimate uses the actual occupational exposure associated with completed decommissioning tasks and provides estimates for those tasks yet to be completed. The maximum estimate of the total occupational exposure is still bounded by the 1,115 person-rem exposure estimate in the FGEIS for the reference pressurized reactor.

Radiation exposure due to transportation of radioactive waste has been conservatively estimated to be approximately 71 person-rem. This value is bounded by the estimate in Supplement I to NUREG-0586 for occupational exposure for transport of radioactive material.

Radiation exposure to offsite individuals for expected conditions, or from postulated accidents is bounded by the Environmental Protection Agency's Protective Action Guidelines and NRC regulations. The public exposure due to radiological effluents will continue to remain well below the IOCFRPart 20 limits and the ALARA dose objectives of IOCFR50, Appendix I. This conclusion is supported by the HNP Annual Effluent Release Reports in which individual doses to members of the public are calculated for station liquid and gaseous effluents.

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Haddam Neck Plant License Termination Plan The total volume of HNP low-level waste (LLW) was estimated to be 283,117 cubic feet in Revision 0 and 1 of the PSDAR. The scenario associated with this estimate involved license termination with many site buildings remaining onsite. Since that time, the decommissioning approach has been modified, based upon lessons learned at other decommissioning facilities.

The modified approach assumes that these buildings will be demolished and disposed of as radwaste, prior to license termination. The change in decommissioning approach increases the estimated volume of LLW waste to approximately 1,158,000 cubic feet. This estimate exceeds the FGEIS LLW volume for the reference PWR by approximately 79% and the estimate in Supplement I to NUREG-0586 by 228%. Thus, the bases for the conclusions in the FGEIS were investigated to determine if the newly estimated volume at HNP would invalidate the conclusions in the FGEIS.

The change in the waste volume is a result of the decision to demolish the buildings and ship a large portion of them as radioactive waste, prior to license termination. The additional source term would be that which would have met the 25 mrem/yr criteria and would have remained onsite after license termination. Appendix K to Supplement 1 to NUREG-0586 classifies this type of waste as "Very Low Activity Waste" and states that "the activity estimates for very low level activity waste are sufficiently small that the activity may be neglected in the evaluation of radiological impacts of transportation of LLW." Approximately 1,086,000 cubic feet of waste is considered to be very low level activity waste, and thus can be ignored in the evaluation of radiological impacts of transportation of this waste. Thus, the increase in the waste volume due to the change in decommissioning approach does not increase the occupational, public, or "on-looker" dose for the decommissioning.

The FGEIS evaluates the impact of LLW waste from decommissioning in the context of the commitment of radwaste disposal space and dose to the public. The commitment of radwaste disposal space is related to the amount of LLW generated and requiring disposal. The FGEIS (Section 4.4) estimates the commitment of LLW disposal space based upon a volume of 18,340 cubic meters (647,600 cubic feet) of LLW, assuming shallow-land burial in standard trenches. The FGEIS concluded that 2 acres of radioactive waste disposal space would be required for the disposal. The FGEIS concluded that the environmental impact of the disposal would not be significant because the amount of radioactive waste disposal is small when compared to the amount of acreage associated with use of the plant site. The basis and conclusions remain valid for HNP. Based upon the amount of LLW estimated to require disposal (1,158,000 cubic feet), approximately 3.5 acres of LLW disposal space would be required. This commitment of acreage still represents a small fraction, when compared to the 525 acres associated with the HNP site.

No significant environmental impacts are anticipated in the event that LLW is required to be temporarily stored onsite because adequate storage space exists and LLW storage will be in accordance with all applicable federal and state regulations.

The non-radiological environmental impacts from decommissioning are temporary and are not significant. The largest occupational risk associated with decommissioning HNP is related to the risk of industrial accidents. The primary environmental effects are short term: small increases in noise levels and fugitive dust in the immediate vicinity of the site, as well as truck traffic to and from the site for hauling equipment and waste. No socioeconomic impacts, other than those Rv August 2004 August 2004 8-68-6 Rev. 2

Haddam Neck Plant License Termination Plan associated with the cessation of operations (loss of jobs and taxes) have been identified. Also, no significant impacts to local culture, terrestrial or aquatic resources, such as the Connecticut River have been identified.

8.2 Analysis of Site-Specific Issues 8.2.1 Onsite-Offsite Land Uses 8.2.1.1 Onsite Land Uses The environmental impacts associated with onsite land uses have been determined by the NRC to be generically applicable with a SMALL impact. The NRC's analysis of the environmental impacts of onsite land uses is documented in Section 4.3.1 of Supplement I to NUREG-0586.

HNP is located on approximately 525 acres. A small fraction of that area had been developed for plant use. Decommissioning activities utilize the same areas used during initial construction and operations. The use of a small fraction of the total site area land impacted by decommissioning and the re-use of areas used during initial construction are consistent with the NRC's assumptions in Supplement I to NUREG-0586, and thus there are no significant environmental impacts associated with HNP decommissioning.

CYAPCO has identified no new information or significant environmental change associated with the site-specific termination activities related to the end use of the site.

8.2.1.2 Offsite Land Uses Only areas within the existing site boundary will be used to support decommissioning and license termination activities (such as temporary storage areas and staging areas). As discussed previously in this section, and in detail in Chapter 5, isolation and control measures will be instituted to prevent the spread of contamination. These measures will also be monitored to ensure their effectiveness. Thus, no environmental impacts associated with the use of offsite lands are anticipated from HNP decommissioning and license termination activities.

8.2.2 Water Use The environmental impacts associated with water use, during decommissioning, have been determined by the NRC to be generically applicable with a SMALL impact. The NRC's analysis of the environmental impacts of water use is documented in Section 4.3.2 of Supplement I to NUREG-0586.

During plant operation, approximately 372,000 gpm was diverted from the Connecticut River and used to cool plants systems in a "once-through" condenser cooling system (Reference 8-9).

Environmental reviews associated with operation of the plant identified no significant environmental impacts associated with water use for the operating facility.

Since decommissioning began in 1997, a number of systems that contributed to water usage have been removed from operation. Chapter 3 of this LTP describes those water-containing systems that have been removed from service or drained and identifies the systems remaining in operation. Only a few systems remain to support cooling and make-up water demands associated with the Spent Fuel Pool or to support batch releases of liquid radiological waste.

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Haddam Neck Plant License Termination Plan These demands will diminish, and eventually be eliminated, by the removal of spent fuel and GTCC wastes from the SFP and their placement on the ISFSI pad.

Use of water for decontamination of systems such as the Reactor Coolant System and the Spent Fuel Pit are addressed in the FGEIS. Other water usage, such as for dust abatement, are similar to those that occurred during construction of the plant. Potable water for decommissioning contractor staff is being provided via bottled water, and sanitary services are provided by water from on-site wells for showers and toilets as well as portable toilet facilities thus minimizing the impacts on the on-site water supply.

In summary, the conditions for HNP decommissioning are consistent with the assumptions of Supplement 1 to the NUREG-0586, and thus there are no significant environmental impacts associated with water use during the decommissioning of the HNP. CYAPCO has not identified any new information or significant environmental change associated with the site-specific termination activities related to the end use of the site.

8.2.3 Water Quality The environmental impacts associated with surface water quality have been determined by the NRC to be generically applicable with a SMALL impact. The NRC's analysis of the environmental impacts of surface water quality is documented in Section 4.3.3 of Supplement 1 to NUREG-0586.

All discharges are controlled under the National Pollutant Discharge Elimination System (NPDES) permit (Reference 8-10) or stormwater permits. These permits are issued by the U.S.

Environmental Protection Agency (EPA). The Offsite Dose Calculation Manual (Reference 8-11) also addresses limitations on radiological doses to members of the public from liquid effluent and requires that they be maintained below the limits in:

• 10CFR50, Appendix 1;

• 10CFR20, Appendix B, Table II, Column 2 (pre-1994 version); and

. 40CFR190.

Radiological impacts are being assessed and monitored by use of on- and offsite groundwater monitoring wells for aquifers that discharge to the Connecticut River. A detailed discussion about future groundwater assessments and historical data are provided in Chapter 2 of this LTP.

As previously discussed, site buildings are being removed to four feet below grade, and any remaining basements are being remediated to meet the appropriate Derived Concentration Guideline Levels (DCGLs). Contaminated concrete debris from demolition of the buildings will be removed from the site and disposed of at an appropriate facility. This contaminated debris will not be used as backfill at the site, and thus does not have the potential to affect ground or surface water quality.

The conditions for HNP decommissioning are consistent with the assumptions of Supplement I to the NUREG-0586, and thus there are no significant environmental impacts associated with surface water quality during the decommissioning of HNP. CYAPCO has not identified any newv Augus 200 8-8 ev.

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Haddam Neck Plant License Termination Plan information or significant environmental change associated with the site-specific termination activities related to the end use of the site.

8.2.4 Air Quality The environmental impacts of decommissioning associated with air quality have been determined by the NRC to be generically applicable with a SMALL impact. The NRC's analysis of the environmental impacts of air quality is documented in Section 4.3.4 of Supplement I to the NUREG-0586.

Supplement 1 to the NUREG-0586 identifies the following decommissioning activities as having the potential for non-radiological impacts on air quality:

• Worker transportation to and from the site,

• Dismantling of systems and removal of equipment,

• Movement and open storage of materials onsite,

• Demolition of buildings and structures, and

• Shipment of material and debris to offsite locations.

Worker transportation: Consistent with the assumptions in the FGEIS, the work force at HNP has decreased from the time the plant ceased operation. The work force will further decrease as decommissioning nears completion. There will and have been occasional increases during specific decontamination and decommissioning activities. The work force during decommissioning is smaller than that associated with plant construction and refueling at HNP.

Accordingly, the adverse changes in air quality, associated with changes in worker transportation, will not be detectable and are not destabilizing.

Dismantling systems and removal of equipment: Generation of particulate matter associated with the physical activities of dismantlement and by the release of gases from systems during removal are potential sources that could impact air quality. Methods and provisions are available to minimize fugitive dust (e.g., wet suppression) and to minimize airborne contamination in buildings (e.g., isolation of areas and HEPA filtration). Local filtration systems may also be used when activities are located in areas that are not ventilated to the plant stack, and are likely to generate airborne radioactivity. Thus, it is highly unlikely that particulate matter generated during decommissioning and released to the environment will be detectable offsite. Any refrigerants will be disposed of in accordance with the applicable state and federal regulations.

Movement and open storage of materials onsite: Movement of equipment and open storage of materials during decommissioning may result in fugitive dust. Provisions as discussed in Chapter 3 and identified above can mitigate these effects. Thus, it is highly unlikely that particulate matter generated as a result of movement or storage of material onsite will be detectable offsite.

Demolition of buildings or structures: As discussed in the FGEIS, demolition of structures and buildings on the HNP site may result in a temporary increase in fugitive dust. The controlled August 2004 8-9 Rev. 2

Haddam Neck Plant License Termination Plan dismantlement and packaging of site components and structures will minimize the potential for fugitive dust from becoming an ambient air quality concern during decommissioning. Fugitive dust from demolition of buildings and structures generally involves large particles that settle quickly. Dust and smaller particles will be controlled using mitigation methods such as wet suppression. Thus, it is highly unlikely that particulate matter generated as a result of building or structure demolition will be detectable offsite.

Shipments of material to an offsite location: Material, debris, and equipment will be removed from the site during decommissioning. Although the remaining number of shipments to be sent during decommissioning is relatively large, these shipments are taking place over a number of years, and thus the average number of shipments per day is relatively small. As stated in the FGEIS, it is unlikely that the emissions associated with the small number of daily shipments would be detectable offsite.

Air effluent released from the site is monitored in accordance with the Radiological Effluent Monitoring Manual (REMM) which sets limits on doses caused by effluents, based upon the ALARA (as low as reasonably achievable) objectives of 10CFR50.34a, 10CFR50.36a, and Section IV.A of Appendix I to 10CFR50. Effluents are reported annually to the NRC.

Based upon the above considerations, it has been determined that the conclusions of the FGEIS are applicable to HNP, and decommissioning of HNP will not noticeably affect offsite air quality. CYAPCO has not identified any new information or significant environmental change associated with the site-specific termination activities related to the end use of the site.

8.2.5 Aquatic Ecology 8.2.5.1 Activities within the Operational Area The environmental impacts associated with aquatic ecology for decommissioning activities within the operational area have been determined by the NRC to be generically applicable with a SMALL impact. The NRC's analysis of the environmental impacts of aquatic ecology for activities within the operational area is documented in Section 4.3.5 of Supplement I to NUREG-0586.

8.2.5.2 Activities Outside of the Operational Area The FGEIS identifies generation of runoff due to ground disturbances and surface erosion as having the potential to impact aquatic resources. Provisions will be made to reduce surface erosion and runoff by appropriate environmental controls and implementation of the site stormwater pollution prevention program.

It is understood that decommissioning of shoreline and in-water structures has the potential to impact aquatic habitats and biota. CYAPCO will consult with regulatory and resource agencies to obtain permits and plan activities to minimize the duration and extent of these impacts.

Regardless, impacts would be limited to those areas previously disturbed during construction and operation, and these areas would be expected to re-colonize as they did following initial construction. Thus, even considering the removal of shoreline and in-water structures, the impacts of decommissioning on aquatic ecology are minimal.

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Haddam Neck Plant License Termination Plan CYAPCO has not identified any new information or significant environmental change associated with the site-specific termination activities related to the end use of the site.

8.2.6 Terrestrial Ecology 8.2.6.1 Activities Within the Operational Area The environmental impacts of decommissioning associated with terrestrial ecology for activities within the operational area have been determined by the NRC to be generically applicable with a SMALL impact. The NRC's analysis of the environmental impacts of terrestrial ecology for activities within the operational area is documented in Section 4.3.6 of Supplement I to the FGEIS.

8.2.6.2 Activities Outside the Operational Area Only areas within the existing site boundary that were used during plant operations are being used to support decommissioning and license termination activities (such as temporary storage areas and staging areas). These areas are within those areas that were disturbed during initial construction and operation. The FGEIS states that terrestrial habitats disturbed during the construction of the site often continue to be of low habitat quality during operation and decommissioning.

As discussed previously in this section, and in detail in Chapter 5, isolation and control measures will be instituted to prevent the spread of contamination, and these measures will be monitored to ensure their effectiveness. Because the HNP site has been in active decommissioning since the decision to permanently close the facility was made, it is reasonable to conclude that areas disturbed during the construction and operation of the plant have not become new sensitive areas with respect to terrestrial biota. Thus, no environmental impacts associated with the use of offsite lands are anticipated from HNP decommissioning and license termination activities related to the end use of the site.

8.2.7 Threatened and Endangered Species CYAPCO has reviewed the Natural Diversity Data Base (Reference 8-12) compiled by the CT DEP. In addition, a report has been prepared for CYAPCO's contractor, CLF Ventures, Inc., by Ickthyological Associates, Inc., to assess the aquatic resources on the Haddam Neck Plant site (Reference 8-13).

The report by Ickthyological Associates identified twelve species as being either endangered, threatened, or special species of concern in Connecticut. These were two fish species (the shortnose sturgeon and the Atlantic sturgeon); the bald eagle; six species of macroinvertebrates (the bronze copper, the midland clubtail, little bluet, the tidewater mucket, the eastern pond mussel, and the wooland pond snail and possibly the yellow lampmussel); and three plants (the swamp cottonwood, the smooth hedge-nettle, and the arrow leaf). The species present at the HNP site are: the eastern pondmussel, the tidewater mucket, and the swamp cottonwood.

A survey performed by certified biologists was conducted on April 2, 2004. The survey found two relic shells of the tidewater mucket and one of the eastern pondmussel in the vicinity of the 8-Il Rev. 2 August 2004 2004 8-11 Rev. 2

Haddam Neck Plant License Termination Plan Intake Structure. One live pondmussel was found approximately 50 feet upstream of the Intake Structure and approximately 20 feet from the shore. The live pondmussel was photographed and then returned to the river bottom close to its original location. It was noted the survey also identified the present of 100 to 200 of each of three common native species of fresh water mussel in the Intake Structure.

Based upon the survey results and on discussions with wildlife biologists, it has been concluded that there is no significant viable population of the tidewater mucket or the eastern pondmussel in the vicinity of the Intake Structure.

Dr. Priscillia Baillie, botanist and ecologist, later identified in a July 10, 2003, report to Tighe and Bond (Reference 8-14) that the species of arrow leaf located on the HNP site was the more common Sagittariamontevidentsis ssp. Sponiosus rather than the state-listed plant Sagittaris subulata.

With respect to other listed plant species, it was noted in Dr. Baillie's report that proposed decommissioning plans to deepen the canal would remove non-native invasive plant species from the canal and would remove the soft substrate, which is rich in organics. The effect of the removal of these harmful invasive species would more than offset the loss a relatively small number of native plants in the area. Removal of the sediments would curtail fill-in of the canal and would prevent the consumption of much of the available oxygen in the water by decomposition of the organic materials. The Swamp Cottonwood are located away from the areas affected by active decommissioning (including the ISFSI and ISFSI haul road). Thus, there would be no significant environmental impact on the identified listed plant species.

Field walk downs are performed in areas of the site undergoing decommissioning or remediation to verify that additional endangered or threatened species are not present.

Thus, decommissioning and license termination activities at the HNP site does not adversely impact threatened or endangered species.

8.2.8 Radiological 8.2.8.1 Activities Resulting in Occupational Doses to WVorkers The environmental impacts associated with radiological activities resulting in occupational doses to worker have been determined by the NRC to be generically applicable with a SMALL impact, because of the existence of guidance regulating doses to workers (I OCFR20) which remain applicable to the HNP. The NRC's analysis of the environmental impacts of radiological activities resulting in occupational doses to workers is documented in Section 4.3.8 of Supplement 1 to NUREG-0586.

8.2.8.2 Activities Resulting in Doses to the Public The environmental impacts associated with radiological activities resulting in doses to the public have been determined by the NRC to be generically applicable with a SMALL impact, because of the existence of guidance regulating and documenting doses to members of the public (I OCFR20). The NRC's analysis of the environmental impacts of radiological activities resulting in doses to the public is documented in Section 4.3.8 of Supplement 1 to NUREG-0586.

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Haddam Neck Plant License Termination Plan CYAPCO has not identified any new information or significant environmental change associated with the site-specific termination activities related to the end use of the site.

Potential doses to the public following license termination are not covered by the Supplement to the FGEIS but were evaluated during promulgation of rulemaking for the radiological criteria for license termination (IOCFR20.1402). The basis for public health and safety considerations associated with the license termination rule is discussed in NUREG-1496.

8.2.9 Radiological Accidents The environmental impacts associated with radiological accidents have been determined by the NRC to be generically applicable with a SMALL impact. The NRC's analysis of the environmental impacts of radiological accidents is documented in Section 4.3.9 of Supplement I to NUREG-0586. CYAPCO has not identified any new information or significant environmental change associated with the site-specific termination activities related to the end use of the site.

The NRC concluded that radiological impacts, due to accidents, are considered to be undetectable and non-destabilizing, in the National Environmental Policy Act (NEPA) sense, if the doses remain within regulatory limits. The HNP UFSAR provides a summary of the evaluation of plant transients that have a potential impact on both occupational and public safety and health. The risk of accidents resulting in a significant radiological release during decommissioning activities is considerably less than during plant operations.

The analysis of decommissioning events includes all phases of decommissioning activities:

decontamination, dismantlement, packaging, storage, radioactive materials handling, and license termination activities (including final status surveys). The following radiological events are identified in the UFSAR (February 2004 revision) as having the potential to affect public health and safety:

• Radioactive Waste System Failure

• Fuel Handling Accidents CYAPCO requested and received an exemption from the emergency preparedness requirements of IOCFR50.47 (Reference 8-13); however, approval of the exemption request was predicated on the absence of any accidents where the offsite dose consequences could exceed the EPA protective action guidelines (PAGs). Releases resulting from accidents postulated in the decommissioning accident analysis were evaluated using the EPA PAGs as an upper limit and found to be bounded by this criterion. Use of the EPA PAGs as an administrative limit also ensures that postulated accident offsite doses are significantly less than the 10CFRIOO reference values.

Thus, because the dose consequences resulting from radiological events, identified as having the potential to affect public health and safety, are below the EPA PAGs and the criteria of IOCFRIOO, the associated impacts on the environment are minimal.

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Haddam Neck Plant License Termination Plan 8.2.10 Occupational Issues The environmental impacts of occupational issues have been determined by the NRC to be generically applicable with a SMALL impact. The NRC's analysis of the environmental impacts of occupational issues is documented in Section 4.3.10 of Supplement I to NUREG-0586.

CYAPCO has not identified any new information or significant environmental change associated with the site-specific termination activities related to the end use of the site.

As Supplement 1 to the NUREG-0586 indicates, the Occupational Safety and Health Act of 1970 was enacted to protect the health of workers, and applicable regulations are administered by the Occupational Safety and Heath Administration (OSHA). HNP is subject to 29 CFR 1910 and 1926 for worker health and safety protection under OSHA regulations. These requirements are implemented under existing plant programs and procedures.

8.2.11 Socioeconomic Impacts The environmental impacts of socioeconomic impacts have been determined by the NRC to be generically applicable with a SMALL impact. The NRC's analysis of the environmental impacts of socioeconomic impacts is documented in Section 4.3.12 of Supplement I to NUREG-0586.

The impacts that are observed by the community are primarily those resulting from plant closure rather than from decommissioning, although some decommissioning activities began very shortly after closure. These impacts occur through changes in employment levels and local demands for housing and infrastructure or through decline of the local tax base and the ability of local government entities to provide public services. Supplement 1 to NUREG-0586 states that decommissioning, itself, has no impact on the tax base and no detectable impact on the demand for public services.

Additionally Supplement 1 to NUREG-0586 concludes that the effects of employment changes on population growth are:

1. not detectable if population changes (reductions or increases) are less than 3% per year,

2. detectable but not destabilizing if the population change is between 3% and 5%, and

3. de-stabilizing if the population change is greater than 5% per year.

Table 8-2 shows the change in population over the last two decades. For the decade 1990 to 2000, which includes the period of shutdown and partial decommissioning, the population in the vicinity of the site increased less than an average of 3% per year during this ten-year period (Reference 8-14). It is notable that the population continues to increase, and in most cases does so consistent with the previous decade. As can be seen, the average annual population change, based upon the data from 1990 and 2000, does not exceed the NRC's threshold of 3%, and thus signifies that the changes are neither detectable nor destabilizing. Thus no significant socioeconomic impacts are associated with HNP decommissioning and license termination activities related to the end use of the site.

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Haddam Neck Plant License Termination Plan 8.2.12 Environmental Justice Environmental Justice was addressed in the HNP Decommissioning Plan (Reference 8-1). An evaluation of the demographic data regarding low income and minority populations along the LLW transportation routes was performed to address Environmental Justice concerns. The results of the analysis support the conclusions that minority populations and low income populations are not disproportionately impacted by LLW waste shipments from the HNP site.

These conclusions remain valid. The types of decommissioning and license termination activities, conducted or planned at HNP, are not significantly different than those described in the Decommissioning Plan and the assumptions related to affected populations remain valid, considering the information from the 2000 Census, presented above. Thus, there are no environmental justice impacts introduced by decommissioning or license termination.

8.2.13 Cultural and Historic Resources Impacts 8.2.13.1 Activities Within the Operational Area The environmental impacts associated with cultural and historic resource impacts from activities within the operational area have been determined by the NRC to be generically applicable with a SMALL impact. The NRC's analysis of the environmental impacts of cultural and historic resource impacts from activities within the operational area is documented in Section 4.3.14 of Supplement 1 to NUREG-0586. CYAPCO has not identified any new information or significant environmental change associated with the site-specific termination activities related to the end use of the site.

8.2.13.2 Activities Outside the Operational Area At CYAPCO's request, a review was performed by American Cultural Specialists, LLC, to identify and evaluate the archaeological resources that might exist on the plant site, particularly in the area of the ISFSI (Reference 8-18). The conclusion reached was that the site activities overall, and specifically those planned in the area of the ISFSI, would present no impact on the area's cultural resources.

Archaeological research continues onsite outside the operational area. At the direction of the State Historic Preservation Office, CYAPCO will characterize and investigate, as appropriate, the archaeologically sensitive areas on site. Results of any investigations will be reported to the State Historic Preservation Office.

The State [of Connecticut] Historic Preservation Office reviewed the documents prepared by American Cultural Specialists, as well as reviewing the Venture Smith Archaeological Site, Haddam. Connecticut, National Register of Historic Places Nomination: Report on Background Research dated May 15, 2002, prepared by the Public Archaeology Survey Team, Inc. In addition, a Staff Archaeologist from the State Historical Preservation Office conducted onsite inspections of the area and associated areas, including the haul road; monitoring station, bulky waste burial pit, and the burial pit access road.

August 2004 8-15 Rev. 2

Haddam Neck Plant License Termination Plan The onsite inspection identified a diffuse scatter of prehistoric and historic archaeological artifacts. However, the State Historical Preservation Office believe that the artifacts lack stratigraphic context and scientific integrity and represent isolated finds that do not provide substantial information on the prehistory or history of the area. Thus, the Office agrees with the conclusions of the American Cultural Specialists that no further archaeological investigations are required in the ISFSI area and that the proposed activities will have no effect upon the state's archaeological heritage. (Reference 8-19) 8.2.14 Aesthetics The environmental impacts associated with aesthetics have been determined by the NRC to be generically applicable with a SMALL impact. The NRC's analysis of the environmental impacts of aesthetics is documented in Section 4.3.15 of Supplement I to NUREG-0586.

Aesthetic resources include natural and man-made landscapes and the way the two are integrated. As a part of construction and operation of the facility, the landscape was previously altered. Decommissioning activities will be conducted onsite, both inside and outside of existing buildings (in the case of dismantlement or shipping activities). The NRC has concluded that any visual intrusion resulting from decommissioning will be temporary and would serve to reduce the aesthetic impacts of the facility. CYAPCO will use best management practices to control many of the potentially adverse impacts of decommissioning on aesthetics (such as dust and noise), as discussed in other sections.

CYAPCO has not identified any new information or significant environmental change associated with the site-specific termination activities related to the end use of the site.

8.2.15 Noise The environmental impacts associated with noise have been determined by the NRC to be generically applicable with a SMALL impact. The NRC's analysis of the environmental impacts of noise is documented in Section 4.3.16 of Supplement 1 to NUREG-0586.

The plant is quite remote from well-traveled roads. To the northeast, just beyond the yard perimeter fence, the terrain rises steeply forming hills that parallel the Connecticut River. The heavily wooded terrain rises abruptly to hill crests of approximately 200 to 300 feet above mean sea level.

During operations, only at locations directly across the Connecticut River and in the immediate plant area could noise level, attributable to plant operations, be detected. With decommissioning, these noises have been eliminated. Decommissioning activities will, in general, be intermittent and temporary, and limited to a relatively small portion of the entire HNP site. Noise is attenuated by the mature forests surrounding the plant. During fall and winter, absence of foliage will allow some additional transmission of noise to the areas north and west of the plant. The presence of the Connecticut River will allow some transmission of noise over the water before attenuation by forest. Because decommissioning activities are expected to add minimally to ambient noise beyond the perimeter security fence, noise will have a negligible effect on the environment.

August 2004 8-16 Rev. 2

Haddam Ncck Plant License Termination Plan 8.2.16 Transportation The environmental issue of transportation has been determined by the NRC to be generically applicable with a SMALL impact. The NRC's analysis of the environmental impacts of transportation is documented in Section 4.3.17 of Supplement I to NUREG-0586.

The number of shipments and the volume of waste shipped are greater during decommissioning than during operations. In Supplement I to the NUREG-0586, the public health and safety impacts of transportation of radioactive wastes are evaluated on the basis of compliance with regulation. The NRC has concluded that compliance with regulation is adequate to protect the public against unreasonable risk from the transportation of radioactive materials. The supplement to the FGEIS notes that the evaluation leading to that conclusion was based, in part, on information in NUREG-0170 and that recent re-evaluation of transportation risks, using updated information and assessment tools, found that risks are lower than those estimated in NUREG-0170. Because HNP will comply with all applicable regulations when shipping radioactive wastes from decommissioning, the effects of transportation of that radioactive waste on public health and safety are considered to be neither detectable nor destabilizing.

Non-radiological impacts of transportation include increased traffic and wear and tear on roadways. Because the average number of daily shipments from the site will be relatively small, there will be no significant effect on traffic flow or road wear. Additionally, because of the industry's emphasis on training and adherence to established procedures, truck accident rates for activities at nuclear facilities has been lower than the national average for similar activities. The NRC has concluded that impacts of transportation accidents would neither be detectable nor destabilizing.

Thus, transportation of wastes associated with the HNP decommissioning and license termination activities do not present significant adverse impacts.

8.2.17 Irretrievable Resources The environmental issue of irretrievable resources has been determined by the NRC to be generically applicable with a SMALL impact. The NRC's analysis of the environmental impacts of irretrievable resources is documented in Section 4.3.18 of Supplement I to NUREG-0586.

Supplement I to the NUREG-0586 indicates that land associated with a site released for unrestricted use is available for other uses, regardless of whether or not the decommissioning process returned the land to an open space or to an industrial complex. Thus the land resource would not be considered "irretrievable." The Supplement to the NUREG-0586 evaluated other irretrievable resources such as the materials/equipment used to decontaminate the facilities and the fuel used for construction machinery and for transporting wastes and concluded these resources are minor.

CYAPCO plans to release the land for unrestricted use. Thus, the impact of decommissioning and license termination on irretrievable resources is neither detectable nor destabilizing.

8-17 Rev. 2 August 2004 2004 8-17 Rev. 2

Haddam Neck Plant License Termination Plan 8.3 References 8-1 Connecticut Yankee Atomic Power Company, Decommissioning Environmental Report.,

dated August 1997.

8-2 Letter CY-97-075 from CYAPCO to USNRC, "Haddam Neck Plant Post-Shutdowvn Decommissioning Activities Report," dated August 22, 1997, as revised.

8-3 NUREG-0586, "Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities," dated August 1988.

8-4 USNRC, Final Environmental Statement, Haddam Neck (Connecticut Yankee) Nuclear Power Plant, Docket No. 50-213, October 1973.

8-5 Letter, USNRC to CYAPCO, "Environmental Assessment for Proposed License Extension," dated November 23, 1987.

8-6 Supplement 1 to NUREG-0586, "Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities," dated November 2002.

8-7 Title 10 to the Code of Federal Regulations, Subpart E to Part 20.

8-8 Revision 2 to the PSDAR, dated April 28, 2004 8-9 Haddam Neck Plant UFSAR, revisions through February 2004.

8-10 State of Connecticut Department of Environmental Protection, NPDES Permit issued to Connecticut Yankee Atomic Power Company, expiration date September 29, 2005.

8-11 Radiological Effluent Monitoring Manual for the Haddam Neck Plant, Revision 17.

8-12 CTDEP to CYAPCO letter, "Natural Diversity Data Base," dated April 24, 2000.

8-18 Rev. 2 August 2004 8-18 Rev. 2

Haddam Neck Plant License Termination Plan 8-13 "Assessment of and Management Recommendations for the Aquatic Resources of the Connecticut Yankee Property, Haddam Neck, Connecticut," final report, prepared by Ichthyological Associates, Inc., dated June 4, 2003.

8-14 Letter from Priscilla W. Baillie to Kurt Prochorena, Tighe & Bond, Inc., "Re:

Connecticut Yankee Decommissioning Project, Reference Number 1260891," dated July 10, 2003.

8-15 Letter CY-04-113, from Gerard van Noordennen, CYAPCO, to Mike Grzywvinski, DEP, "Haddam Neck Plant Structures, Dredging & Fill Permit Application, Intake Structure Demolition, Additional Information," dated May 25, 2004, 8-16 Fredrichs, USNRC, to Mellor, CYAPCO, "Exemption from a Portion of 10 CFR Section 50.54(q) and Approval of Defueled Emergency Plan at Haddam Neck Plant (RAC No.

99015) dated August 28, 1998.

8-17 US Census Bureau, "Population Housing Units, Area, and Density: 2000, Connecticut Place and County Subdivision."

8-18 Phase IB Archaeology Reconnaissance Survey of the Proposed ISFSI Location on the Connecticut Yankee Property in Haddam Neck, CT," prepared by American Cultural Specialists, LLC, dated May 16, 2002.

8-19 Letter from John W. Shannahan, Director and State Historic Preservation Officer, Connecticut Historical Commission to K.J. Heider, CYAPCO, dated May 20, 2002.

August 2004 8-19 Rev. 2

Haddam Neck Plant License Termination Plan Table 8-1 Summary of Environmental Impacts from Decommissioning Issue Generic Impact LTP Section Onsite-Offsite Land Uses 8.2.1

• Onsite Land Uses Yes Small 8.2.1.1

• Offsite Land Uses No Site-Specific 8.2.1.2 Water Use Yes Small 8.2.2 Water Quality Yes Small 8.2.3 Air Quality Yes Small 8.2.4 Aquatic Ecology 8.2.5

• Activities within the operational area Yes Small 8.2.5.1

• Activities outside the operational area No Site-Specific 8.2.5.2 Terrestrial Ecology 8.2.6

• Within the operational area Yes Small 8.2.6.1

• Outside the operational area No Site-Specific 8.2.6.2 Threatened and Endangered Species No Site-Specific 8.2.7 Radiological 8.2.8

• Activities resulting in occupational doses Yes Small 8.2.8.1 to workers

• Activities resulting in doses to the public Yes Small 8.2.8.2 Radiological accidents Yes Small 8.2.9 Occupational issues Yes Small 8.2.10 Cost N/A N/At 7 Socioeconomic Yes Small 8.2.11 Environmental Justice No Site-Specific 8.2.12 Cultural and Historic Resource Impacts 8.2.13

• Activities within the operational area Yes Small 8.2.13.1

• Activities outside the operational area No Site-Specific 8.2.13.2 Aesthetics Yes Small 8.2.14 Noise Yes Small 8.2.15 Transportation Yes Small 8.2.16 Irretrievable Resources Yes Small 8.2.17

• The operational area is defined as the portion of the plant site where most or all of the site activities occur, such as reactor operation, materials and equipment storage, parking, substation operation, facility service, and maintenance. This includes areas within the protected area fences, the intake, discharge, cooling, and associated structures as well as surrounding paved, graveled, maintained landscape, or other maintained areas.

t A decommissioning cost assessment is not a specific National Environmental Policy Act (NEPA) requirement.

August 2004 8-20 Rev. 2

Haddam Neck Plant License Termination Plan Table 8-2 Population Changes in the Vicinity of HNP Municipality 1990  % change from 2000  % change (Ref 8-1) prior decade (Ref 8-17) in decade including shutdown Chester 3417 11.4 3742 9.5 Colchester 10980 41.5 14551 32.5 Deep River 4332 8.5 4610 6.4 Durham 5732 11.5 6627 15.6 East Haddam 6676 18.8 8333 24.8 East Hampton 10428 21.7 13352 28.0 Essex 5904 16.3 6505 9.2 Haddam 6769 6.0 7157 5.7 Hebron 7079 29.8 8610 21.6 Killingworth 4814 21.1 6018 25.0 Lyme 1949 7.0 2016 3.4 Madison 15485 10.4 17858 15.3 Marlborough 5535 16.6 5709 3.1 Middlefield 3925 3.4 4203 7.1 Middletown 42762 9.5 43167 1.0 Portland 8418 0.4 8732 3.8 Salem 3310 41.8 3858 8.3 Westbrook 5414 3.0 6292 16.2 8-21 Ret'. 2 August 2004 8-21 Rev. 2 I

Haddam Neck Plant License Termination Plan Appendix H Table 2-10 MARSSIM Classifications (updated as of November 2001)

August 2004 H-1 Rev. 2

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (m')

Ratio Area Survey Area Surnvity Survey Unit Initial Floor Total otal Figure MARSSIM Area:

CCode od ecito Code oeDsrpinclass. Floor No 1000 Fuel Building 2-2 1102 Fuel Building Laydown Area 0001 Floor Area, Walls and Ceiling - 1 80 250 3.1 2-5 Section 1 0002 Floor Area, Walls and Ceiling - 1 70 275 3.9 2-5 Section 2 0003 Floor Area, Walls and Ceiling - 1 65 315 4.8 2-5 Section 3 1104 Fuel Building Fuel Cask Decon 0000 Floor Area, Walls and Ceiling 1 40 245 6.1 2-5 Area 1106 Fuel Building Skimmer Pump and 0000 Floor Area, Walls and Ceiling 1 30 160 5.3 2-5 Sump Area 1202 Fuel Building New Fuel Storage 0000 Floor Area, Walls and Ceiling 1 90 340 3.8 2-5 Area 1204 Fuel Building Exhaust Filters and 0000 Floor Area, Walls and Ceiling 1 20 110 5.5 2-5 Fan August 2004 H-2 Rev. 2

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (m 2)l Ratio Survey Survey Area Survey Survey Unit Initial Floor Ttl (Total Fgr Are Code Description CUnit Code Description CARSS. Area Area Floor No.

I_ _ __ _ A rea) _ _ _ _ _ _

1302 Fuel Building Patio Area 0001 Roof Area I N/A 65 N/A 2-5 0002 Roof Area I N/A 65 N/A 2-5 1304 Fuel Building New Fuel Storage 0000 Floor Area, Walls and Ceiling 1 80 320 4.0 2-5 Area 1306 Fuel Building Cask Laydown Area 0000 Floor Area, Walls and Ceiling 1 80 320 4.0 2-5 1308 Fuel Building Spent Fuel Pool Pit 0001 Floor Area, Pool Sides and I 100 655 6.6 2-5 Bottom, Walls and Ceiling -

Section 1 0002 Floor Area, Pool Sides and 1 100 655 6.6 2-5 Bottom, Walls and Ceiling -

__ Section 2 1404 Fuel Building Roof Area 0001 Roof Area - Section I 1 N/A 90 N/A 2-3 0002 Roof Area - Section 2 1 N/A 90 N/A 2-3 H-3 Rev. 2 August 2004 H-3 Rev. 2

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Ar (m 2)

Ratio Survey Survey Area Survey Survey Unit Flooral(Total Area CoeDsrpinUnit M SInitialr Total Area: Figure Code Code Description Code Code Description Class. Area Area Floor No.

_________________________________ ________Area) 0003 Roof Area - Section 3 1 N/A 90 N/A 2-3 0004 Roof Area - Section 4 1 N/A 90 N/A 2-3 0005 East & West Exterior Walls 2 N/A 1,000 N/A 2-5 0006 North & South Exterior Walls 2 N/A 460 N/A 2-5 2000 Primary Auxiliary Building 2-2 2002 Primary Auxiliary Building RHR 0000 Floor Area, Walls and Ceiling 1 25 275 11.0 2-6 Pump Room A 2004 Primary Auxiliary Building RHR 0000 Floor Area, Walls and Ceiling 1 25 275 11.0 2-6 Pump Room B 2006 Primary Auxiliary Building RHR 0000 Floor Area, Walls and Ceiling . 30 390 13.0 2-6 Heat Exchangers H-4 Rev. 2 August 2004 H-4 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Figure MARSSIM Area Area A l C t Code CdDescription C Floor No.

______Area) 2008 Primary Auxiliary Building Primary 0000 FloorArea, Walls and Ceiling 1 35 380 10.9 2-6 Drain Tank Pump Room 2010 Primary Auxiliary Building Primary 0000 Floor Area, Walls and Ceiling 1 30 320 10.7 2-6 Drain Tank Room 2012 Primary Auxiliary Building Aerated 0000 Floor Area, Walls and Ceiling 1 25 285 11.4 2-6 Drain Tank Room 2104 Primary Auxiliary Building Pipe 0001 Floor Area, Walls and Ceiling - 1 85 300 3.5 -

Chase Under Hallway Section 1 0002 Floor Area, Walls and Ceiling - 1 70 215 3.1 -

Section 2 2106 Primary Auxiliary Building Pipe 0001 Floor Area, Walls and Ceiling - 1 60 155 2.6 -

Chase Under Valve Room Section 1 0002 Floor Area, Walls and Ceiling - 1 60 155 2.6 -

Section 2 2108 Primary Auxiliary Building Boric 0000 Floor Area, Walls and Ceiling 1 80 265 3.3 2-6 Acid Evaporator Area TK EV I-IA, EV2-1A August 2004 H-5 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (ml2 Ratio SreSuvyAreaSuvyure Unit Initial Floor Total (Total Figure Arve vode D esierption Unit Code De MARSSIM Area Area Area: No.

CdCoeDsrtonCode CodeDecito Class. Floor No 2110 Primary Auxiliary Building Pipe 0000 Floor Area, Walls and Ceiling 1 75 250 3.3 -

Chase East & West Outside 2202 Primary Auxiliary Building 0001 Floor Area, Walls and Ceiling - 1 80 270 3.4 2-6 Hallway Section 1 0002 Floor Area, Walls and Ceiling - 1 80 270 3.4 2-6 Section 2 2204 Primary Auxiliary Building 0000 Floor Area, Walls and Ceiling 1 100 320 3.2 2-6 Component Cooling Area 2206 Primary Auxiliary Building Boric 0000 Floor Area, Walls and Ceiling 1 80 300 3.8 2-6 Acid Evaporator Area 2208 Primary Auxiliary Building Boric 0000 Floor Area, Walls and Ceiling I 100 320 3.2 2-6 Acid Mix Tank Area 2210 Primary Auxiliary Building "B" 0000 Floor Area, Walls and Ceiling 1 35 220 6.3 2-6 Charging Pump Area 2212 Primary Auxiliary Building "A" 0000 Floor Area, Walls and Ceiling 1 35 220 6.3 2-6 Charging Pump Area August 2004 H-6 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Arrea (m2)

Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Figure Code Code Description Code Code Description Class. Area Area Floor No.

Area) . -

2214 Primary Auxiliary Building 0000 Floor Area, Walls and Ceiling 1 35 240 6.9 2-6 Metering Pump Area 2216 Primary Auxiliary Building 0000 Floor Area, Walls and Ceiling I 100 430 4.3 2-6 Purification Pump Area 2218 Primary Auxiliary Building Primary 0000 Floor Area, Walls and Ceiling 1 40 200 5.0 2-6 Water Transfer Pump Area 2220 Primary Auxiliary Building Sample 0000 Floor Area, Walls and Ceiling 1 20 150 7.5 2-6 Room 2222 Primary Auxiliary Building Steam 0000 Floor Area, Walls and Ceiling 1 45 240 5.3 2-6 Generator Blowdown Room 2224 Primary Auxiliary Building HPSI 0000 Floor Area, Walls and Ceiling 1 60 180 3.0 2-6 Cubicle Area 2226 Primary Auxiliary Building LPSI 0000 Floor Area, Walls and Ceiling 1 60 160 2.7 2-6 Cubicle Area 2228 Primary Auxiliary Building 0001 Floor, Walls and Ceiling - 1 70 460 6.6 2-6 Drumming Room Section I H-7 Rev. 2 August 2004 H-7 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Ratio Survey Survey Area Survey Survey Undi Initial Floor Total (Total Figure Aren Code Description AraArea Code Description Area Fr Code Code Class. No.o7 Area) 0002 Floor, Walls and Ceiling - I 100 540 5.4 2-6 Section 2 2302 Primary Auxiliary Building 0000 Floor Area, Walls and Ceiling 1 90 355 3.9 2-6 Component Cooling Area 2304 Primary Auxiliary Building Boric 0000 Floor Area, Walls and Ceiling 1 75 350 4.7 2-6 Acid Evaporator Area 2306 Primary Auxiliary Building Boric 0000 Floor Area, Walls and Ceiling 1 80 320 4.0 2-6 Acid Mix Tank Area 2308 Primary Auxiliary Building Volume 0000 Floor Area, Walls and Ceiling 1 40 220 5.5 2-6 Control Tank Room 2310 Primary Auxiliary Building Purge 0000 Floor Area, Walls and Ceiling 1 90 320 3.6 2-6 and Dilution Fans 2312 Primary Auxiliary Building Service 0000 Floor Area, Walls and Ceiling 1 25 170 6.8 2-6 Water Strainer Area 2314 Primary Auxiliary Building HEPA 0001 Floor Area, Walls and Ceiling - 1 85 310 3.6 2-6 Filter and Hall Area Section 1 H-8 Rev. 2 August 2004 H-8 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (m')

Ratio Survey Survey Area ut Srvey Unit Initial Floor Total (Total Figure Code Code Description Code Code Description Class. Area Area Floor No.

__________________________ _________Area)______

0002 Floor Area, Walls and Ceiling - 1 80 205 2.6 2-6 Section 2 0003 Floor Area, Walls and Ceiling - 1 70 180 2.6 2-6 Section 3 0004 Floor Area, Walls and Ceiling - 1 70 195 2.8 2-6 Section 4 0005 Floor Area, Walls and Ceiling - 1 95 250 2.6 2-6 Section 5 0006 Floor Area, Walls and Ceiling - 1 75 200 2.7 2-6 Section 6 0007 Floor Area, Walls and Ceiling - 1 65 190 2.9 2-6 Section 7 0008 Floor Area, Walls and Ceiling - 1 35 70 2.0 2-6 Section 8 2316 Primary Auxiliary Building Boric 0001 Floor Area, Walls and Ceiling - 1 60 155 2.6 2-6 Acid Storage Room Section I H-9 Rev. 2 2004 August 2004 H-9 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (m')

Ratio Survey Survey Area Survey Unit Initial Floor Total (Total Figure Code Code Description Code Code Description Class. Area Area Floor No.

____ ___ ___ ____ ___ ____ ___ ____ ___A rea) 0002 Floor Area, Walls and Ceiling - 1 60 215 3.6 2-6 Section 2 2402 Primary Auxiliary Building Roof 0001 Roof Area by Ventilation Duct- I N/A 75 N/A 2-3 Area Section 1 0002 Roof Area by Ventilation Duct- I N/A 75 N/A 2-3 Section 2 0003 Roof Area by Ventilation Duct- I N/A 75 N/A 2-3 Section 3 0004 Remaining Roof Area - Section 2 N/A 455 N/A 2-3 0005 Remaining Roof Area - Section 2 N/A 775 N/A 2-3 2

0006 Exterior Walls 3 N/A 2,260 N/A 2-3 3000 Reactor Containment 2-2 H-10 Rev. 2 2004 August 2004 H-10 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Ratio SreSuvyAreaSuvyure Unit Initial Floor Total (TotaI Figure Survea SCr ipy ion l l CodeDescriptin UnitU MARSSIM Area Area Area: No.

CdCoeDsrpinCode CodeDecito Class. Floor No Area) 3002 Reactor Containment Enclosure 0000 Floor Area and Walls I 25 115 4.6 2-7 Under Reactor Vessel 3004 Reactor Containment Enclosure 0000 Floor Area and Walls 1 20 100 5.0 2-7 Sump Area Under Reactor Vessel 3101 Reactor Containment Enclosure #4 0001 Floor Area, Walls and Ceiling - 1 65 300 4.6 2-8 Outer Annulus Lower Level NE Section 1 0002 Floor Area, Walls and Ceiling - 1 65 300 4.6 2-8 Section 2 3102 Reactor Containment Enclosure #1 0001 Floor Area, Walls and Ceiling - 1 65 300 4.6 2-8 Outer Annulus Lowver Level NW Section 1 0002 Floor Area, Walls and Ceiling - 1 65 300 4.6 2-8 Section 2 3103 Reactor Containment Enclosure #2 0001 Floor Area, Walls and Ceiling - 1 65 300 4.6 2-8 Outer Annulus Lower Level SW Section 1 0002 Floor Area, Walls and Ceiling - 1 65 300 4.6 2-8 Section 2 H-I I Rev. 2 August 2004 Atlgist H-1 I Rcv. 2

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (m2 )

Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Figure Coea Code Description Cndet Code Description ass Ara Floor l No.

______Area) 3104 Reactor Containment Enclosure #3 0001 Floor Area, Walls and Ceiling - 1 65 300 4.6 2-8 Outer Annulus Lower Level SE Section 1 0002 Floor Area, Walls and Ceiling - 1 65 300 4.6 2-8 Section 2 3105 Reactor Containment Enclosure 0000 Floor Area, Walls and Ceiling 1 40 185 4.6 2-8 Reactor Containment Sump Area 3107 Reactor Containment Enclosure 0001 Floor Area, Walls and Ceiling - 1 70 290 4.1 2-8 Cable Vault Outside Reactor Section I Containment 0002 Floor Area, Walls and Ceiling - 1 70 290 4.1 2-8 Section 2 3111 Reactor Containment Enclosure 0001 Floor Area, Walls and Ceiling - 1 60 225 3.8 2-8 Loop #1 Inner Annulus Lower Section I Level NE 0002 Floor Area, Walls and Ceiling - 1 55 220 4.0 2-8 Section 2 3112 Reactor Containment Enclosure 0001 Floor Area, Walls and Ceiling - 1 75 255 3.4 2-8 Loop #2 Inner Annulus Lower Section I Level NW .

H- 12 Rev.2 I 2004 August 2004 H-12 Rev. 2 l

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

_______A ra A L ____2_

Ratio Survey Survey Area Survey Survey Unit Initial Floor Total a otal Figure Code Code Description Code Code Description Class. Area Area Floor No.

._ Area) 0002 Floor Area, Walls and Ceiling - 1 75 255 3.4 2-8 Section 2 3113 Reactor Containment Enclosure 0001 Floor Area, Walls and Ceiling - 1 75 255 3.4 2-8 Loop #3 Inner Annulus Lower Section 1 Level SW 0002 Floor Area, Walls and Ceiling - 1 75 255 3.4 2-8 Section 2 3114 Reactor Containment Enclosure 0001 Floor Area, Walls and Ceiling - 1 75 285 3.8 2-8 Loop #4 Inner Annulus Lower Section I Level SE 0002 Floor Area, Walls and Ceiling - 1 75 270 3.6 2-8 Section 2 3201 Reactor Containment Enclosure #1 0000 Floor Area, Walls and Ceiling I 100 500 5.0 2-9 Outer Annulus Ground Level NE 3202 Reactor Containment Enclosure #2 0001 Floor Area, Walls and Ceiling - 1 65 360 5.5 2-9 Outer Annulus Ground Level NW Section 1 0002 Floor Area, Walls and Ceiling - 1 65 395 6.1 2-9 Section 2 H- 13 Rev.2 I August 2004 August 2004 H-13 Rev. 2 l

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (m')

Ratio Sv Area Survey Sunity Surve Unit Initial Floor Total (Total Figure SreSueyArea SudeveyscSurvey Unit Cd ecito MARSSIM Area Area Area: No CdCoeDsrpinCode CodeDecito Class. Floor No

_____ Area) 3203 Reactor Containment Enclosure #3 0001 Floor Area, Walls and Ceiling - 1 65 395 6.1 2-9 Outer Annulus Ground Level SW Section 1 0002 Floor Area, Walls and Ceiling - 1 65 360 5.5 2-9 Section 2 3204 Reactor Containment Enclosure #4 0001 Floor Area, Walls and Ceiling - 1 60 380 6.3 2-9 Outer Annulus Ground Level SE Section I 0002 Floor Area, Walls and Ceiling - 1 65 360 5.5 2-9 Section 2 3205 Reactor Containment Enclosure 0000 Floor Area and Walls 1 80 235 2.9 2-9 Reactor Containment Foyer Area Ground Level 3206 Reactor Containment Enclosure 0000 Floor Area and Walls 1 95 195 2.1 2-9 Reactor Containment Hatch Area Ground Level 3211 Reactor Containment Enclosure 0001 Floor Area, Walls and Ceiling - 1 55 225 4.1 2-9 Loop #1 Inner Annulus Mid Ground Section I NE 0002 Floor Area, Walls and Ceiling - 1 55 225 4.1 2-9 Section 2 H-14 Rev. 2 2004 August 2004 H-14 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001) 2 Area (i )

Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Figure Area Code Description Code Code Description CARSS.M Area Area Flo: No.

________________________ __________Area) 3212 Reactor Containment Enclosure 0001 Floor Area, Walls and Ceiling - 1 60 205 3.4 2-9 Loop #2 Inner Annulus Mid Ground Section I NW 0002 Floor Area, Walls and Ceiling - 1 60 300 5.0 2-9 Section 2 3213 Reactor Containment Enclosure 0001 Floor Area, Walls and Ceiling - 1 60 260 4.3 2-9 Loop #3 Inner Annulus Mid Ground Section I SW 0002 Floor Area, Walls and Ceiling - 1 60 205 3.4 2-9 Section 2 3214 Reactor Containment Enclosure 0001 Floor Area, Walls and Ceiling - 1 60 280 4.7 2-9 Loop #4 Inner Annulus Mid Ground Section I SE 0002 Floor Area, Walls and Ceiling - 1 60 200 3.3 2-9 Section 2 3301 Reactor Containment Enclosure #1 0001 Floor Area and Containment 1 75 110 1.5 2-10 Outside Crane Charging Floor Enclosure Wall up to el. 56' 6" -

Section 1 0002 Floor Area and Containment 1 75 110 1.5 2-10 Enclosure Wall up to el. 56' 6" -

H-IS Rev. 2 August 2004 H-15 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (M_)

Ratio Survey Area Survey Survey Unit Survey Initial Floor Total (Total Figure Code Code Description Code Code Description Class. Area Arla Aorea No.

_______ ___________________________Area)______

Section 2 3302 Reactor Containment Enclosure #2 0001 Floor Area and Containment 1 75 110 1.5 2-10 Outside Crane Charging Floor Enclosure Wall up to el. 56' 6" -

Section 1 0002 Floor Area and Containment 1 75 110 1.5 2-10 Enclosure Wall up to el. 56' 6" -

Section 2 3303 Reactor Containment Enclosure #3 0001 Floor Area and Containment 1 75 110 1.5 2-10 Outside Crane Charging Floor Enclosure Wall up to el. 56' 6" -

Section 1 _

0002 Floor Area and Containment 1 75 110 1.5 2-10 Enclosure Wall up to el. 56' 6" -

Section 2 3304 Reactor Containment Enclosure #4 0001 Floor Area and Containment 1 75 110 1.5 2-10 Outside Crane Charging Floor Enclosure Wall up to el. 56' 6" -

Section 1 0002 Floor Area and Containment 1 75 110 1.5 2-10 Enclosure Wall up to el. 56' 6" -

Section 2 H- 16 Rev. 2 August 2004 H-16 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Ar a (m 2)

Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Figure Code Code Description Code Code Description Class. Area Area Floor No.

________________________ ________Area) 3311 Reactor Containment Enclosure #1 0001 Floor Area 1 55 55 1.0 2-10 Inside Crane Charging Floor 0002 Floor Area 1 55 55 1.0 2-10 3312 Reactor Containment Enclosure #2 0001 Floor Area 1 60 60 1.0 2-10 Inside Crane Charging Floor 0002 Floor Area 1 60 60 1.0 2-10 3313 Reactor Containment Enclosure #3 0001 Floor Area 1 60 60 1.0 2-10 Inside Crane Charging Floor 0002 Floor Area 1 60 60 1.0 2-10 3314 Reactor Containment Enclosure #4 0001 Floor Area 1 60 60 1.0 2-10 Inside Crane Charging Floor 0002 Floor Area 1 60 60 1.0 2-10 H-17 Rev. 2 August 2004 H-17 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (m2 )

Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Figure Code Code Description Code Code Description MARSSIM Area Area Area: No.

__________ ________Area) ______

3315 Reactor Containment Enclosure 0000 Floor Area 1 40 40 1.0 2-10 Removable Grating for RX Head Staging .

3320 Reactor Containment Enclosure 0000 Floor and Walls 1 35 500 14.3 2-10 CTMT Rx Refuel Canal to Spent Fuel Pit .

3322 Reactor Containment Enclosure 0000 Floor and Walls I 100 460 4.6 2-10 CTMT Reactor Refueling Cavity 3324 Reactor Containment Enclosure 0000 Wall Area and Supports 1 40 210 5.3 2-10 CTMT Reactor Vessel Area 3326 Reactor Containment Enclosure 0000 Floor Area 1 10 10 1.0 2-10 Upper Core Package Storage Area 3403 Reactor Containment Enclosure 0001 Dome - Quadrant 1 2 N/A 675 N/A 2-10 Inside Surfaces 0002 Dome - Quadrant 2 2 N/A 675 N/A 2-10 0003 Dome - Quadrant 3 2 N/A 675 N/A 2-10 H-18 Rev. 2 August 2004 H-18 Rcv. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (m_ _

Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Figure Code Code Description Code Code Description Class. Area Area Floor No.

________Area) 0004 Dome - Quadrant 4 2 N/A 675 N/A 2-10 0005 Shell (el. 56'6" and up) - 2 N/A 840 N/A 2-10 Section 1 0006 Shell (el. 56'6" and up) - 2 N/A 840 N/A 2-10 Section 2 0007 Shell (el. 56'6" and up) - 2 N/A 840 N/A 2-10 Section 3 3502 Reactor Containment Enclosure 0001 Dome - Quadrant 1 2 N/A 725 N/A 2-3 Outside Surfaces 0002 Dome - Quadrant 2 2 N/A 725 N/A 2-3 0003 Dome - Quadrant 3 2 N/A 725 N/A 2-3 0004 Dome - Quadrant 4 2 N/A 725 N/A 2-3 H- 19 Rev.2 I 2004 August 2004 H-19 Rev. 2

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Arc a (m 2)

Ratio Area Survey Area Suviety Survey Unit InitialI Floor Total (Total Figure Code Code Description Code Code Description Class. Area Area Floor No.

________ ________ _________ A rea) _ _ _ _ _ _

0005 Shell - Quadrant 1, East Section 2 N/A 525 N/A 2-3 0006 Shell - Quadrant 1, North 2 N/A 525 N/A 2-3 Section 0007 Shell - Quadrant 2 3 N/A 1,050 N/A 2-3 0008 Shell - Quadrant 3 3 N/A 1,050 N/A 2-3 0009 Shell - Quadrant 4 3 N/A 1,050 N/A 2-3 4000 Turbine Building 2-2 4102 Turbine Building North Floor Area l 0000 Floor Area and Walls 2 450 1,450 3.2 2-11 4104 Turbine Building Oil Room, Heater 0000 Floor Area, Walls and Ceiling 2 810 1,650 2.0 2-1 1 Drains, Emergency Power August 2004 H-20 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Are a (m2)

Ratio urve Survey Area Sunvety Survey Unit Initial Floor Total (Total Figure Are CdeDecrptonUni CdeDecrptonMARSSIM Ae Ara Area: No CdCoeDsrpinCode CodeDecito Class. Floor No

______________________________ _____Area) 4106 Turbine Building Air Compressor 0000 Floor Area, Walls and Ceiling 2 170 460 2.7 2-11 Area 4108 Turbine Building Steam Generator 0000 Floor Area, Walls and Ceiling 2 325 800 2.5 2-11 Feed Pump Area 4110 Turbine Building Chemistry/Closed 0000 Floor Area, Walls and Ceiling 2 290 750 2.6 2-11 Cooling Water Area 4112 Turbine Building Water Treatment 0000 Floor Area, Walls and Ceiling 2 365 980 2.7 2-11 Area 4114 Turbine Building Condenser Pump 0000 Floor Area, Walls and Ceiling 2 755 1,660 2.2 2-11 and South Floor Area 4116 Turbine Building Hoist/Equipment 0000 Floor Area and Walls 2 165 330 2.0 2-11 Laydown Area 4118 Turbine Building Condenser "A" 0000 Floor Area, Walls and Ceiling 2 480 1,040 2.2 2-11 Water Box "A & B" Area 4120 Turbine Building Condenser "B" 0000 Floor Area, Walls and Ceiling 2 580 1,260 2.2 2-11 Water Box "C & D" Area August 2004 H-21 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (M2)

Ratio SreSuvyAreaSuvyure Unit Initial Floor Total (Total Figure S e SurveyArea ldSuDrveny l CodeSDscription MARSSTM Area Area Area: No.

CdCoeDsrpinCode CodeDecito Class. Floor No

______________________________________ _______ Area) 4121 Turbine Building Secondary Chem 0000 Floor Area, Walls and Ceiling 2 85 270 3.2 2-11 Lab 4202 Turbine Building North End Open 0000 Wall Area and Supports 2 N/A 320 N/A 2-12 Area 4204 Turbine Building Oil Reservoir 0000 Structure Area and Walls 2 N/A 250 N/A 2-12 Area 4206 Turbine Building S/G Feedwater 0000 Floor Area, Walls and Ceiling 2 425 1,410 3.3 2-12 Heater 2A and 2B Area 4208 Turbine Building S/G Feedwater 0000 Floor Area, Walls and Ceiling 2 270 840 3.1 2-12 Heater IA and IB Area 4210 Turbine Building Steam Generator 0000 Floor Area, Walls and Ceiling 2 165 600 3.6 2-12 Feedwater Control Valve Area 4212 Turbine Building South 0000 Wall Area and Supports 2 N/A 640 N/A 2-12 End/Turbine Hall 4216 Turbine Building S/G Feedwater 0000 Floor Area, Walls and Ceiling 2 570 1,290 2.3 2-12 Heater 6B and 5B Area August 2004 H-22 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (mr)

Ratio ArCea Survey Area Sunvety Survey Unit Initial Floor Total (Total Figure Are CdeDecrptonUni CdeDecrptonMARSSIM Ae Ara Area: No CdCoeDsrpinCode CodeDecito Class. Floor No

_ _ ___ __ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ ___ _ _ _ _ _A rea ) _ _ _ _ _ _

4218 Turbine Building S/G Feedwater 0000 Floor Area, Walls and Ceiling 2 800 1,810 2.3 2-12 Heater 6A and SA Area 4302 Turbine Building 30" Main Steam 0000 Floor Area, Walls and Ceiling 2 1,000 2,350 2.4 2-13 Line Area 4304 Turbine Building 24" Main Steam 0000 Floor Area, Walls and Ceiling 2 715 1,820 2.5 2-13 Line Area 4306 Turbine Building MSRHR IA and 0000 Floor Area, Walls and Ceiling 2 980 2,040 2.1 2-13 1B Area Reheater 4308 Turbine Building MSRHR IC and 0000 Floor Area, Walls and Ceiling 2 900 2,080 2.3 2-13 ID Area Reheater 4402 Turbine Building Laydown Area 0000 Floor Area and Walls 2 900 3,100 3.4 2-14 North Floor 4404 Turbine Building Steam Generator 0000 Floor Area and Walls 2 190 720 3.8 2-14 Fcedwater Heater 3A Area 4406 Turbine Building Steam Generator 0000 Floor Area and Walls 2 200 885 4.4 2-14 Feedwater Heater 4A Area H-23 Rev. 2 August 2004 H-23 Rev. 2

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

- r Area (m')

Ratio Survey Surve Area Survey Survey Unit Initial Floor Total (Total Figure Ara UitMARSSIM Area Are Code Code Description Code Code Description Class. Area Area Floor No.

____ _______ ___ ___ _ __ ___ ___ A rea) 4408 Turbine Building Steam Generator 0000 Floor Area and Walls 2 170 650 3.8 2-14 Feedwater Heater 3B Area 4410 Turbine Building Steam Generator 0000 Floor Area and Walls 2 180 690 3.8 2-14 Feedwater Heater 4B Area 4412 Turbine Building H.P. Turbine Area 0000 Floor Area 2 290 290 1.0 2-14 4414 Turbine Building L.P. #1 Turbine 0000 Floor Area 2 290 290 1.0 2-14 Area 4416 Turbine Building L.P. #2 Turbine 0000 Floor Area 2 315 315 1.0 2-14 Area Area 4418 Turbine Building Generator Area 0000 Floor Area 2 330 330 1.0 2-14 4420 Turbine Building Exciter Area 0000 Floor Area 2 60 60 1.0 2-14 4422 Turbine Building Laydown Area 0000 Floor Area and Walls 2 320 780 2.4 2-14 South Floor H-24 Rev. 2 August 2004 H-24 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (m2 )

Ratio SuvySurvey Area Survey Survey Unit Initial Floor Total (Total Figure CAorde Code Description Unit Code Description MARSSIM Area Area Flore No.

______________ __________________________Area) 4424 Turbine Building Open Hoist Area 0000 Wall Area 2 N/A 460 N/A 2-14 4502 Turbine Building Overhead Crane 0001 Ceiling Area - Section 1 2 N/A 810 N/A 2-14 Area 0002 Ceiling Area - Section 2 2 N/A 810 N/A 2-14 0003 Ceiling Area - Section 3 2 N/A 810 N/A 2-14 0004 Ceiling Area - Section 4 2 N/A 810 N/A 2-14 4603 Turbine Building Roof Area 0001 Roof Area - Section 1 2 N/A 810 N/A 2-3 0002 Roof Area - Section 2 2 N/A 810 N/A 2-3 0003 Roof Area - Section 3 2 N/A 810 N/A 2-3 August 2004 H-25 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (m2)

Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Figure Area CoeDsrpinUnit CoeDsrpinMARSSIM Ae Ara Area: No CCode od ecito Code oeDsrpinClas Ae Ara Floor No 0004 Roof Area - Section 4 2 N/A 810 N/A 2-3 0005 Exterior Walls 3 N/A 8,760 N/A 2-3 5000 Service Building 2-2 5102 Service Building "A" Diesel 0000 Floor Area, Walls and Ceiling 3 160 550 3.4 2-16 Generator Area 5104 Service Building "B" Diesel 0000 Floor Area, Walls and Ceiling 3 160 550 3.4 2-16 Generator Area 5106 Service Building Clean Locker 0000 Floor Area, Walls, and Ceiling 2 220 750 3.4 2-15 Room Area 5108 Service Building Hot Locker Room 0000 Floor Area, Walls, and Ceiling 2 145 490 3.4 2-15 Area 5110 Service Building HP Control Point 0000 Floor Area, Walls, and Ceiling 2 160 695 4.3 2-15 and Office Areas August 2004 H-26 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (m2 )

Ratio Area Survey Area Survey Survey UInitial Floor Total (Total Figure Code Code Description Code Code Description Class. Area rea Floor No.

5112 Service Building Woman's Locker 0000 Floor Area, Walls, and Ceiling 2 90 280 3.1 2-15 Room Area 5114 Service Building Hot Chemistry 0001 Floor Area, Walls, and Ceiling - 1 70 205 2.9 2-15 Area Section 1 0002 Floor Area, Walls, and Ceiling - 1 70 205 2.9 2-15 Section 2 5118 Service Building Maintenance 0000 Floor Area, Walls, and Ceiling 1 70 340 4.9 2-15 Decon Area 5120 Service Building Machine Shop 0000 Floor Area, Walls, and Ceiling 2 120 560 4.7 2-15 Clean Area 5122 Service Building Machine Shop Hot 0001 Floor Area, Walls, and Ceiling - 1 95 420 4.4 2-15 Area Section 1 0002 Floor Area, Walls, and Ceiling- 1 85 325 3.8 2-15 Section 2 5124 Service Building Maintenance 0000 Floor Area, Walls, and Ceiling 2 585 2,100 3.6 2-16 Clean Shop Area H-27 Rev. 2 August 2004 H-27 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (mL)

Ratio Survey Survey Area Sunit uveUi Initial Flor Tta Total Fgr A Code Deseription Unit Code Description MARSSIM a a rea l :

Areaa 5126 Service Building "A" Auxiliary 0000 Floor Area, Walls, and Ceiling 2130 530 4.1 2-16 Boiler Area 5128 Service Building "B" Auxiliary 0000 Floor Area, Walls, and Ceiling 2 130 530 4.1 2-16 Boiler Area 5128 Service Building E Hallway ast 0000 Floor Area, Walls, and Ceiling 2 130 830 8.3 2-15 5132 Service Building Health Physics 0000 Floor Area, Walls, and Ceiling 2 145 770 5.3 2-15 Facility Ist Floor 5134 Service Building Health Physics 0000 Floor Area, Walls, and Ceiling 2 145 770 5.3 2-15 Facility 2nd Floor 5202 Service Building Switch Gear Area 0000 Floor Area, Walls, and Ceiling 3 900 2,490 2.8 2-17 5302 Service Building Control Room 0000 Floor Area, Walls, and Ceiling 3 505 1,250 2.5 2-17 Area 5304 Service Building Computer, 0000 Floor Area, Walls, and Ceiling 3 195 600 3.1 2-17 Operations, Security Area H-28 Rev. 2 August 2004 H-28 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (m2 )

Ratio Area Survey Area Unitr l Survey Unit Initial Floor Total (Total Figure Are CdeDecrptonUni CdeDecrptonMARSSTM Ae Ara Area: No CdCoeDsrpinCode CodeDecito Class. Floor No

_ _ __ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ __ __ A re a) _ _ _ _ _ _

5306 Service Building Machine and 0000 Floor Area, Walls, and Ceiling 2 210 575 2.7 2-17 Equipment Area 5308 Service Building Instrument & 0000 Floor Area, Walls, and Ceiling 2 225 610 2.7 2-17 Controls Shop 5402 Service Building Roof 0001 Roof Area - Section 1 2 N/A 1,000 N/A 2-3 0002 Roof Area - Section 2 2 N/A 565 N/A 2-3 0003 Roof Area - Section 3 2 N/A 565 N/A 2-3 0004 Roof Area - Section 4 2 N/A 590 N/A 2-3 0005 Exterior Walls 3 N/A 2,100 N/A 2-3 5502 Screenhouse Building CW System 0001 Unit I Discharge Tunnel Surface 2 290 950 3.3 2-19 Trench Area H-29 Rev. 2 August2004 August 2004 H-29 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (m2 )

Ratio Survey Survey Area Survey Survey Unit Initial For Total Areta: Figure Area l Unit l Code Description ore (t Area: l CdCoeDsrpinCode CodeDecito Class. Ara re Floor No 0002 Unit 2 Discharge Tunnel Surface 2 470 2,110 4.5 2-19 Area 6000 Waste Disposal Building 2-2 6002 Waste Disposal Building Hall Area 0000 Floor Area, Walls and Ceiling 1 25 230 9.2 2-18 Lower Level 6004 Waste Disposal Building Area 0000 Floor Area, Walls and Ceiling 1 25 190 7.6 2-18 Outside Reboiler Room 6006 Waste Disposal Building Bottoms 0000 Floor Area, Walls and Ceiling I 15 105 7.0 2-18 Pump and Reboiler Area 6008 Waste Disposal Building Sump 0000 FloorArea, Walls and Ceiling 1 20 145 7.3 2-18 Trench Area Lower Level 6010 Waste Disposal Building-Waste 0000 Floor Area, Walls and Ceiling 1 20 135 6.8 2-18 Decay Tank A,B,C Area 6012 Waste Disposal Building Surge 0000 FloorArea, Walls and Ceiling I 10 100 10.0 2-18 Tank Area Lower Level H-30 Rev. 2 August 2004 H-30 Rev. 2

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (m2 )

Ratio Arve Survet Initial Flo(otl Total Fgr Srve Survey Area S iutey Survey Unit MARSSIM Aa Totalr Area: Figure Code Code Description Code Code Description Class. Ae Ara Floor No.

_______ ___________________________Area) ______

6102 Waste Disposal Building Hall Area 0000 Floor Area, Walls and Ceiling 1 10 40 4.0 2-18 6202 Waste Disposal Building Hallway 0000 Floor Area, Walls and Ceiling 1 30 195 6.5 2-18 Area 6304 Waste Disposal Building 0000 Floor Area, Walls and Ceiling 1 25 175 7.0 2-18 Evaporator Area 6306 Waste Disposal Building Radwaste 0000 Floor Area, Walls and Ceiling 1 15 100 6.7 2-18 Liquid Evaporator 6308 Waste Disposal Building 0000 Floor Area, Walls and Ceiling 1 30 185 6.2 2-18 Degassifier Transfer Pump Area 6312 Waste Disposal Building 0000 Floor Area, Walls and Ceiling 1 20 155 7.8 2-18 Degassifier and Associated Valves 6404 Waste Disposal Building 0000 Floor Area, Walls and Ceiling 1 25 175 7.0 2-18 Evaporator Area 6406 Waste Disposal Building Liquid 0000 Floor Area, Walls and Ceiling 1 15 100 6.7 2-18 Evaporator Area Rev. 2 August 2004 H-31I H-3 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (m2)

Ratio Vey Survey Unit Initial Floor Total (Total Figure lit CoeDsrpinMARSSIM Ae Ara Area: No

,e de eDsrpinClass. Ae Ara Floor No 00 Floor Area, Walls and Ceiling 1 30 185 6.2 2-18 00 Floor Area, Walls and Ceiling 1 20 155 7.8 2-18 01 Roof Area - Section I 1 N/A 60 N/A 2-18 02 RoofArea -Section 2 1 N/A 60 N/A 2-18 03 Exterior Walls 3 N/A 870 N/A 2-3 2-2 00 Floor Area, Walls and Ceiling 3 90 300 3.3 2-19 00 FloorArea,WallsandCeiling 3 90 300 3.3 2-19 August 2004 H-32 Rev. 2

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

I I Area (m')

Ratio Survey Survey Area Survey Survey Unit InitiaM Floor Total (Total Figure Coea Code Description Unit Code Description CaRss. Area Area Floor No.

____ __ _ ____ ____ _ __ ___A rea) 7102 Screenhouse Building CW Circ 0000 Floor Area, Walls and Ceiling 3 125 395 3.2 2-19 Pump Motor A&B 7104 Screenhouse Building CW Circ 0000 Floor Area, Walls and Ceiling 3 125 395 3.2 2-19 Pump Motor C&D 7106 Screenhouse Building CW 0000 Floor Area, Walls and Ceiling 3 55 250 4.5 2-19 Hypochloride Tank Area 7108 Screenhouse Building CW Intake 0000 Floor Area 3 180 180 1.0 2-19 and Screen Area 7202 Screenhouse Building CW Roof 0000 Roof Area and Exterior Walls 3 N/A 595 N/A 2-19 Area 8000 Penetration Building 2-2 8100 Penetration Building Upper Level 0000 Floor Area, Walls and Ceiling 1 90 330 3.7 -

8200 Penetration Building Mid Level 0000 Floor Area, Walls and Ceiling 1 90 330 3.7 =

H-33 Rev. 2 August 2004 H-33 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (m 2)

Ratio Survey Survey Area Survey Survey Unit Initial Ar Total Tora: Figure Area CoeDsrpinUnit CoeDsrpinMARSSIM Florea Ae rao CdCoeDsrpinCode CodeDecito Class. Ara re Floor N.

8300 Penetration Building Lower Level 0000 Floor Area, Walls and Ceiling 1 90 210 2.3 -

9000 Miscellaneous Buildings and Land 2-1 Areas 9102 YD 115KV Switchyard Area 0001 Trench and Adjoining Land I N/A 120 N/A 2-1 Area 0002 Land Area 2 N/A 1,220 N/A 2-1 9104 YD Main Transformer Area 0000 Land Area 3 N/A 930 N/A 2-1 9106 Discharge Canal 0001 Bank Land Area and Canal 2 N/A 9,300 N/A 2-1 Sediment 0002 Bank Land Area and Canal 2 N/A 9,300 N/A 2-1 Sediment 0003 Bank Land Area and Canal 2 N/A 9,300 N/A 2-1 Sediment H-34 Rev. 2 August 2004 H-34 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Ar a (m_2)

Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Figure Code Code Description Code Code Description Class. Area Area Floor No.

Area) 0004 Bank Land Area and Canal 2 N/A 9,300 N/A 2-1 Sediment 0005 Bank Land Area and Canal 2 N/A 9,300 N/A 2-1 Sediment 0006 Bank Land Area and Canal 2 N/A 9,300 N/A 2-1 Sediment 0007 Bank Land Area and Canal 2 N/A 9,300 N/A 2-1 Sediment 0008 Bank Land Area and Canal 2 N/A 9,300 N/A 2-1 Sediment 9108 YD North Tank Farm Area 0000 Land Area I N/A 200 N/A 2-2 9110 YD South Tank Farm Area 0000 Land Area I N/A 150 N/A 2-2 9112 YD Boron Storage Tank Area 0000 Land Area I N/A 60 N/A 2-2 H-35 Rev. 2 August 2004 H-35 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (in)

Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Fgr Area Code Description Unit Code Description MARSSIM Area Area Area: Fgr Code Code Class. Floor No 9114 YD Ion Exchange Area 0000 Land Area 1 N/A 300 N/A 2-2 9116 YD Resin Slurry Area 0000 Land Area I N/A 60 N/A 2-2 9118 YD Fuel Oil Tank Area 0000 Standing Structure and Land 3 N/A 200 N/A 2-1 Area 9120 YD Primary Vent Stack 0000 Structure and Land Area I N/A 10 N/A 2-1 9122 YD Primary Water Storage Tank 0000 Land Area I N/A 300 N/A 2-2 Area 9124 YD Backup Primary Water Storage 0000 Land Area 1 N/A 300 N/A 2-2 Tank Area 9126 YD Large Yard Crane Area 0000 Land Area I N/A 1,400 N/A 2-3 9128 YD Demin Water Storage Tank 0000 Land Area I N/A 340 N/A 2-2 Area H-36 Rev. 2 August 2004 H-36 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (m2 i Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Figure Area Code Description Code Code Description Class. Area Area Floor No.

9202 Switchgear Building "B" 0001 Upper Switchgear Structure 3 185 1,600 8.6 2-2 Including Floor, Walls, Ceiling,

.-. Roof and Exterior Surfaces _

0002 Lower Switchgear Structure 2 185 430 2.3 2-2 Including Floor, Walls and Ceiling . -

9208 Administration Building 0000 Structure Including Floor, 3 495 4,300 8.7 2-2 Walls, Ceiling, Roof and Exterior Surfaces . . . -

9214 Shutdown Auxiliary Feed Pump 0000 Structure Including Floor, 2 25 220 8.8 2-2 House Walls, Ceiling, Roof and Exterior Surfaces 9226 Radwaste Reduction Facility 0001 Floor, Ceiling and Walls From I 100 460 4.6 2-20 the Eastern Entrance to a Point

._ 15 Foot West 0002 Floor, Ceiling and Walls From 1 100 405 4.1 2-20 Point 15 Foot West to a Point 30 Foot West . -

0003 Floor, Ceiling and Walls From 1 100 230 2.3 2-3 Point 30 Foot West to a Point 45 Foot West H-37 Rev. 2 August 2004 H-37 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Areq (m')

Ratio Area Survey Area Survey Survey Unit Initial Floor Total (Total Figure Code Code Description Code Code Description Class. Area Area Flrea No.

0004 Floor, Ceiling and Walls From 1 100 305 3.1 2-20 Point 45 Foot West to West

_ _ _ _ _ _ _ W all 0005 Floor, Ceiling and Walls From 1 75 445 5.9 2-20 the Southwestern Rollup Door to the Opposite Shield Wall 0006 Roof 2 N/A 465 N/A 2-20 0007 Exterior Walls 2 N/A 665 N/A 2-3 9227 Busl3 0001 Pad and Ground Underneath I N/A 100 N/A -

[Former Bus 10 Pad]

0002 Bus 13 Structure 3 100 410 N/A -

9228 Unconditional Release Facility 0000 Slab 1 45 45 1.0 2-2 9302 Northwest Protected Area Grounds 0000 Land Area 3 N/A 4,200 N/A 2-i H-38 Rev. 2 August 2004 H-38 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (m')

Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Figure Code Code Description Code Code Description Class. Area Area Floor No.

Area) 9304 Southwest Protected Area Grounds 0000 Land Area 3 N/A 3,000 N/A 2-1 9306 South Central Protected Area 0000 Land Area 2 N/A 2,450 N/A 2-1 Grounds 9307 PAB / Service Building Alleyway 0000 Land Area 1 N/A 1,000 N/A 2-1 9308 Southeast Protected Area Grounds 0001 Land Area I N/A 1,470 N/A 2-1 0002 Land Area 1 N/A 1,470 N/A 2-1 0003 Land Area I N/A 1,470 N/A 2-1 9310 East Protected Area Grounds 0001 Land Area From the Fuel I N/A 1,120 N/A 2-1 Building to the RadWaste Reduction Facility 0002 Land Area From the RadWaste I N/A 1,350 N/A 2-1 Reduction Facility to the East RCA Boundary H-39 Rev. 2 August 2004 H-39 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Figure Code Code Description CodU Code Description Class. Area Floor No.

____ ___ ___ ____ ___ ____ ___ ____ ___A rea) 9312 Northeast Protected Area Grounds 0001 Land Area From the North RCA 1 N/A 1,190 N/A 2-1 Gate to Security Fence 0002 Land Area From Security Fence 1 N/A 1,230 N/A 2-1 to Fuel Building 9313 Central Site Grounds 0000 Land Area 2 N/A 600 N/A 2-1 9402 Emergency Operations Facility 0000 Structure Including Floor, 3 1,300 5,300 4.1 2-2 Walls, Ceiling and Exterior Walls 9403 Emergency Operations Center Roof 0000 Structure Roof 3 N/A 1,300 N/A 2-3 9404 North Warehouse 0000 Structure Including Floor, 3 400 2,900 7.3 2-2 Walls, Ceiling, Roof and Exterior Surfaces _

9406 South Warehouse 0000 Structure Including Floor, 3 400 2,900 7.3 2-2 Walls, Ceiling, Roof and Exterior Surfaces 9408 Miscellaneous Trailer Complex 0000 Structure Including Floor, 3 N/A 1,600 N/A 2-2 Walls, Ceiling, Roof and Exterior Surfaces H-40 Rev. 2 August 2004 H-40 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

_ _ _ _ _ __ __ _ _ _ _ __ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _A re (m 2 )

Ratio Survey Survey Area Survey Survey UUit Initial Floor Total (Total Figure Code Code Description Code Code Description Class. Area Area Floor No.

Area) 9410 Steam Generator Mockup Building 0000 Structure Including Floor, 3 N/A 460 N/A 2-2 Walls, Ceiling, Roof and Exterior Surfaces 9412 Training Stores Office Building 0000 Structure Including Floor, 3 480 2,620 5.5 2-2 Walls, Ceiling, Roof and Exterior Surfaces .

9414 Warehouse #1 0001 Structure Including Floor, 3 2,000 6,400 3.2 2-2 Walls, Ceiling 0002 Roof and Exterior Surfaces 3 N/A 4,400 N/A 2-3 9416 Warehouse #2 0000 Structure Including Floor, 3 1,500 8,500 5.7 2-2 Walls, Ceiling, Roof and Exterior Surfaces 9418 Office Building #3 and PAP 0000 Structure Including Floor, 3 900 7,500 8.3 2-2 Walls, Ceiling, Roof and Exterior Surfaces .

9420 Office Trailer 0000 Structure Including Floor, Walls 3 220 650 3.0 2-2 and Ceiling 9422 Information Center 0000 Structure Including Floor, 3 1,200 3,800 3.2 2-2 Walls, Ceiling and Exterior H-41 Rev. 2 2004 August 2004 H-41 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

_ __ _ _ __ _ __ __ _ _ __ _ __ _ _ _ _ __A rea (m )_ _ _ _

Ratio Survey Survey Area Survey Survey Unit Initial Floor Total (Total Figure Are Code Description Unit Code Description Cass. Area Area Area: No.

Code Code Class. Floor No Walls_ Area)

Walls 9423 Information Center Roof 0000 Structure Roof 3 N/A 1,200 N/A 2-3 9424 All Buildings Contained in the 0000 Structure Including Floor, 3 130 470 3.6 2-2 Southwest Site Storage Area Walls, Ceiling, Roof and Exterior Surfaces 9502 Northeast Site Grounds (Non- 0000 Land Area 3 N/A 8,700 N/A 2-1 Protected Area) 9504 Bypass Road / Secondary Parking 0000 Land Area 3 N/A 2,800 N/A 2-1 Lot 9506 North Site Grounds (Non-Protected 0000 Land Area 3 N/A 3,800 N/A 2-1 Area) 9508 Pond 0000 Land Area and Pond Sediment 3 N/A 10,000 N/A 2-1 9510 Access Road 0000 Paved Road 3 N/A 2,300 N/A 2-1 H-42 Rev. 2 August 2004 H-42 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (in)

Ratio Survey Survey Area Survey Survey Unit MAIniia Floor (Total Figure Code Code Description Code Code Description Class. Area Area Floor No.

9512 Northwest site Grounds (Non- 0000 Land Area 3 N/A 19,500 N/A 2-1 Protected Area) 9514 Primary Parking Lot 0000 Paved Lot 3 N/A 20,000 N/A 2-1 9518 Southwest Site Grounds (Non- 0000 Land Area 2 N/A 5,900 N/A 2-4 Protected Area) 9520 Southwest Site Storage Area 0001 Land Area From Security Fence 2 N/A 7,500 N/A 2-4 to Load Distribution Tower 0002 Land Area From Load 2 N/A 7,000 N/A 2-4 Distribution Tower East to 150m 0003 Land Area 150m East of Load 2 N/A 7,000 N/A 2-4 Distribution Tower to Gate 3 9521 Southeast Pond 0000 Land Area and Pond Sediment 3 N/A 23,100 N/A 2-4 9522 Southeast Site Grounds (Non- 0001 Land Area 2 N/A 8,700 N/A 2-1 Protected Area)

H-43 Rev. 2 August 2004 H-43 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (M2)

Ratio Area Sre Area Survey Survey Unit Initial Floor Total (Total Figure Code Code Description Code Code Description Class. Area Area Floor No.

__ _ _ _ _ _ _Area) _ _ _ _ _ _

0002 Land Area I N/A 1,800 N/A 2-1 0003 Land Area 1 N/A 1,900 N/A 2-1 0004 Land Area 1 N/A 1,200 N/A 2-1 9523 Southeast Wetland Area 0000 Land Area 3 N/A 106,000 N/A 2-4 9524 South Site Grounds (Non-Protected 0000 Land Area 3 N/A 110,000 N/A 2-4 Area) 9525 Southeast Site Road 0000 Paved Road 3 N/A 28,000 N/A 2-4 9526 Northeast Mountain Side 0000 Land Area 3 N/A 444,700 N/A 2-4 9527 East Mountain Side 0001 Land Area From Contour Line 2 N/A 9,800 N/A 2-1 120' to Upper Fence and Eastern

._ Ridge Edge H-44 Rev. 2 August 2004 H-44 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

. Area (m'2)

Ratio Survey Survey Area Survey Survey Unit Initial For Ttl (Total Fgr Area Desriptio Cde UnitMARSSIM Arear Areal Ae: Fgr Code iCode Code Description Class. Area FoArea or No.

____ ___ ____ __ _ ____ ___ A rea) 0002 Land Area From Contour Line 2 N/A 9,600 N/A 2-1 120' to Lower Fence and Eastern Ridge Edge 0003 Land Area From Ridge Edge to 2 N/A 9,300 N/A 2-1 9522A Boundary 9528 Southeast Mountain Side 0000 Land Area 3 N/A 553,000 N/A 2-4 9530 Central Peninsula Area 0001 Land Area Bounded by and 2 N/A 9,700 N/A 2-4 Immediately Adjacent to the Road 0002 Western Half of Diked Area and 2 N/A 7,000 N/A 2-4 Immediate Surrounding Sides 0003 Eastern Half of Diked Area and 2 N/A 7,000 N/A 2-4 Immediate Surrounding Sides 0004 Remaining Land Area 3 N/A 70,800 N/A 2-4 9531 South End of Peninsula 0000 Land Area 3 N/A 118,000 N/A 2-4 H-45 Rev. 2 August 2004 H-45 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (my Ratio Survey Survey Area Survey Survey UInitial Floor Total (Total Figure Code Code Description Code Code Description Class. Area Area Floor No.

______A r ea) _ _ _ _ _ _

9532 East Site Grounds (Non-Protected N/A N/A Non-impacted N/A 375,600 N/A 2-4 Area) 9535 South East Landfill Area 0001 Land Area 1 N/A 1,900 N/A 2-4 0002 Land Area 2 N/A 3,600 N/A 2-4 9536 Construction Piles Near Rifle Range 0000 Land Area 2 N/A 2,200 N/A 2-4 9537 Permitted Landfill Area 0000 Land Area 2 N/A 2,200 N/A 2-4 9538 Material Storage Area 0000 Land Area 2 N/A 4,200 N/A 2-4 9801 Subsurface soils in Radiologically 0000 Subsurface Soil A* N/A 15,500 N/A 2-21 Controlled Area (excluding 9308) 9802 Subsurface soils associated with 0000 Subsurface Soil B*N/A 27,065 N/A 2-21 surface soil survey area 9308 and subsurface soil areas within the August 2004 H-46 H-46 Rev. 2 Rev. 2 I

Table 2-10 MARSSIM Classifications (Updated as of November 2001)

Area (m=)

Ratio Survey Survey Area Survey SuvyUnit Initial Floor Total (Total Fgr SurveyArea Code Descr nUnit Ctde Description MARSSIM Area Ar No.

CdCoeDsrpinCode CodeDecito Class. Ara re Floor No

_______ __________________________Area) industrial area but outside the radiologically controlled area 9803 Subsurface soils associated with 0000 Subsurface Soil C* N/A 35,150 N/A 2-21 parking lot, Warehouses 1 & 2, and Steam Generator Mockup Building 9804 Subsurface soils associated with 0000 Subsurface Soil C* N/A 11,950 N/A 2-21 South East Site Grounds 9805 Subsurface soils associated with 0000 Subsurface Soil C* N/A 137,000 N/A 2-21 Peninsula (excluding area 9531) 9806 Subsurface soils associated with 0000 Subsurface Soil B* N/A 4250 N/A 2-21 surface soil portions of survey area 9535 (South East Landfill)

• MARSSIM does not cover media such as subsurface soil, which is considered beyond its scope. LTP Section 5.7.3.2.1 discusses the criteria applied during the classification of subsurface soils.

August 2004 H-47 Rev. 2 I

/