Text

Yankee Nuclear Power Station - Final Safety Analysis Report Biennial Update
	ML052150047
Person / Time
Site:	Yankee Rowe
Issue date:	06/30/2005
From:	Jesse Rollins; Yankee Atomic Electric Co
To:	; Document Control Desk, NRC/FSME
References
	BYR 2005-057
	Download: ML052150047 (158)
	v • d • e

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YANKEE ATOMIC ELECTRIC COMPANY Telephone (413) 424-5261 YANKEE 49 Yankee Road, Rowe, Massachusetts 01367 June 30. 2005 BYR 2005-057 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washinuton. D.C. 20555

References:

Subject:

1. License No. DPR-3 (Docket No. 50-29)

2. Letter. G.A. Maret (YAEC) to USNRC. Updated Final Safety Analysis Report. BYR 2003-049. June 27. 2003 Final Safetv Analysis Report Biennial Update Yankee Atomic Electric Company (YAEC) herewith encloses one paper cop) and one CD-ROM copy of the updated Yankee Nuclear Powver Station (YNPS) Final Safety Analysis Report (FSAR). A summary listing of the changes for each FSAR section is also enclosed. This biennial update of the YNPS FSAR is being submitted in accordance with the requirements of 10 CFR 50.7 l(e)(4). and replaces in its entirety the last update that was submitted in June 2003 (Reference 2). Unless otherwise noted, the update reflects changes made through June 2005.

Should there be any questions regarding this submittal. please contact the undersigned at (413) 424-2300.

Very truhl vours.

YANKEE ATOMIC ELECTRIC COMPANY ack D. Rolihs Licensin-and Reoulatorv Affairs Manager I declare under penalty of perjury that the foregoing is true and correct to the best of my knowledge.

Executed on (a

0 Subscribed and sworn to before me This 30 dav of June 2005

&tAFL 'I6 Date Commission Expires:

g f I X (D C3 Enclosures

<7 n

Jr)SYJ

USNRC Document Control Desk BYR 2005-057, Page 2 cc:

Ms. M. Miller. USNRC. Region I Branch Chief Mr. J. Kottan. USNRC. Region I Inspector Mr. J. Hickman. Project Manager. NMSS

SUMMARY

LISTING OF JUNE 2005 CHANGES TO YNPS FSAR Section Number Description of Change i thru xvii Revised Table of Contents to reflect permanent removal, and current status and configuration of systems, structures and components.

PSDAR Updated information relating to the License Termination Plan, dismantlement methods, status of systems, structures, and components, disposal site locations and decommissioning cost estimate.

Deleted schedule of activities already completed.

Updated the waste volume shipped, actual exposure (through 12/31/04), and the estimated "to go" exposure.

100 No change.

200 Deleted Section 200.4 that is no longer applicable.

211, 215, 220-221, 226-227 Deleted these sections since they are no longer applicable.

229 Revised text to update the description of the electrical system by noting that it is powered by one switchboard via one transformer.

230-231 Deleted these sections since they are no longer applicable.

232 Revised text to describe the Fire Detection System coverage for the ISFSI.

236-239, 241-248, 250-254 Deleted these sections since they are no longer applicable.

255 Revised text to eliminate tanks that no longer exist, and update the status of remaining tanks.

256 No change.

257 Revised text to reflect "current" as opposed to "future" tense for ISFSI component usage. Revised Figure 257-1 to reflect the current site configuration.

300 Revised Figure 300-1 to delete the Yankee's Visitor Center 301-303 No change to these sections.

304 Revised text to clarify current water supply.

305 Revised text to clarify the effluent release path during the final stages of decommissioning.

306 Eliminated reference to the License Termination Plan, which is discussed in the PSDAR.

307 No change.

400 Revised the General Overview to indicate that certain events have been retained until decommissioning activities are complete. Changed the Defueled Emergency Plan to the ISFSI Defueled Emergency Plan. Updated the bounding set of dose conversion factors in Table 400.1 (and associated notes, text and I of 3

references) to make them applicable until 1/1/2008.

Updated Table 400.3 to make it consistent with Section 408 with respect to the fuel storage location and governing events included in the NAC-MPC FSAR. Revised the Note to Table 400.4 to include the Vapor Container, and added a reference to it in the table.

401 Changed Defueled Emergency Plan to ISFSI Defueled Emergency Plan.

402 No change.

403 No change.

404 No change.

405 No change.

406 Changed the number and location of propane tanks with respect to the ISFSI; and indicated that the existing NAC analysis for overpressure bounds the explosion of the revised propane tank configuration.

407 Replaced statement that the physical consequences of an on-site transportation event does not adversely affect the Spent Fuel Pit structural integrity with a statement that it does not adversely affect the dry cask storage system at the ISFSI.

408 No change.

500 No change.

501 No change.

502 Revised statement to indicate that the Yankee Decommissioning Quality Assurance Program (YDQAP) requires indoctrination and training of staff personnel involved with YDQAP activities.

503 No change.

504 Text was added that addressed the submittal of the License Termination Plan (LTP) together with revision and supplement, and how the LTP demonstrates that the remaining decommissioning activities will be performed in accordance with Federal Regulations.

505 No change.

506 No change.

507 Changed the job title of various YNPS managers to reflect organizational changes and related responsibilities. Revised the frequency with which radiation workers will receive whole body counting.

Deleted statement that installed process and effluent monitors are used in accordance with the Off-Site Dose Calculation Manual. Deleted one reference.

508 On-site waste packaging areas that have been demolished were removed from the text. The text was also revised to indicate that radioactive waste management related programs will be maintained in compliance with the YDQAP, as opposed to 2of3

Technical Specifications. Reference to specific locations for solid radioactive waste storage awaiting shipment was eliminated. Reference to a specific direct off-site dose from storage of low-level radioactive waste in the certain site locations was eliminated. Reference to processing of contaminated liquids in the liquid waste evaporator was eliminated.

The text on Airborne Radioactive Waste Processing was revised to eliminate reference to systems and structures that no longer exist.

509 "Plant tests" was changed to "tests" to reflect the current site status.

510 The section on Health and Safety Organization and Functions was revised to reflect the current organization.

511 The section on Above-Ground Fuel Oil Storage Tanks was revised to reflect the removal of tanks and related contaminants.

512 The text was revised to indicate that the protected area is the ISFSI.

513 A Note was added to indicate that the remaining requirements of the Fire Protection Technical Requirements Manual were incorporated into site administration procedures and the manual cancelled.

The text was revised to indicate that the implementation of the Fire Protection Program has changed in that the active fire protection systems have been removed from service and is replaced by the onsite incipient fire brigade.

514 No change.

515 Text was added explaining that a revision to the Emergency Plan was issued which eliminated all facets of the Plan not related to the ISFSI.

Appendix A Updated several plant drawings.

3of3

I YNPS FSAR REV. 6/05 TABLE OF CONTENTS Page POST-SHUTDOWN DECOMMISSIONING ACTIVITIES REPORT SECTION 100 100 SPENT FUEL MANAGEMENT PLAN...................................

100-1 100.1 Background...................................

100-1 100.2 Special Nuclear Material Control...................................

100-1 REFERENCES...................................

100-2 SECTION 200 200 DECOMMISSIONING ACTIVITIES AND PLANNING.200-1 200.1 Introduction.200-1 200.2 Decommissioning Plan Approval.200-1 200.3 Decontamination and Dismantlement: General Information.200-2 200.3.1 Overview.200-2 200.3.2 Detailed Planning and Engineering Activities.200-3 200.3.3 General Decontamination and Dismantlement Considerations.200-3 200.3.4 Decontamination and Dismantlement Process.200-5 200.4 Materials Cutting Station.200-6 200.5 Decontamination and Dismantlement Plan: Systems, Structures and Components.200-6 200.5.1 Overview.200-6 201 REACTOR VESSEL (Deleted - No longer applicable) 202 STEAM GENERATORS (Deleted - No longer applicable) 203 MAIN COOLANT SYSTEM (Deleted - No longer applicable) 204 PRESSURE CONTROL AND RELIEF SYSTEM (Deleted -No longer applicable) 205 CHARGING AND VOLUME CONTROL SYSTEM (Deleted - No longer applicable) i

YNPS FSAR REV. 6/05 TABLE OF CONTENTS (Continued)

Page 206 CHEMICAL SHUTDOWN SYSTEM (Deleted - No longer applicable) 207 PURIFICATION SYSTEM (Deleted - No longer applicable) 208 COMPONENT COOLING WATER SYSTEM (Deleted - No longer applicable) 209 PRIMARY PLANT CORROSION CONTROL SYSTEM (Deleted - No longer applicable) 210 PRIMARY PLANT SAMPLE SYSTEM (Deleted -No longer applicable) 211 TEMPORARY WASTE WATER PROCESSING ISLAND SYSTEM (Deleted - No longer applicable) 212 SHUTDOWN COOLING SYSTEM (Deleted -No longer applicable) 213 PRIMARY PLANT VENT AND DRAIN SYSTEM (Deleted - No longer applicable) 214 EMERGENCY CORE COOLING SYSTEM (Deleted -No longer applicable) 215 RADIATION MONITORING SYSTEM (Deleted - No longer applicable) 216 VC VENTILATION AND PURGE SYSTEM (Deleted -No longer applicable) 217 VC HEATING AND COOLING SYSTEM (Deleted -No longer applicable) 218 POST-ACCIDENT HYDROGEN CONTROL SYSTEM (Deleted -No longer applicable) 219 CONTAINMENT ISOLATION SYSTEM (Deleted -No longer applicable) 220 FUEL HANDLING EQUIPMENT SYSTEM (Deleted -No longer applicable) 221 SFP COOLING AND PURIFICATION SYSTEM (Deleted -No longer applicable) 222 MAIN STEAM SYSTEM (Deleted - No longer applicable) 223 FEEDWATER SYSTEM (Deleted - No longer applicable) 224 STEAM GENERATOR BLOWDOWN SYSTEM (Deleted - No longer applicable) 225 EMERGENCY FEEDWATER SYSTEM (Deleted - No longer applicable) ii

YNPS FSAR REV. 6/05 TABLE OF CONTENTS (Continued)

Page l 226 AUXILIARY SERVICE WATER SYSTEM (Deleted - No longer applicable) l 227 DEMINERALIZED WATER SYSTEM (Deleted - No longer applicable) 228 COMPRESSED AIR SYSTEM (Deleted -No longer applicable) 229 ELECTRICAL SYSTEM............................................

229-1 229.1 Description............................................

229-1 229.2 Decontamination and Dismantlement Considerations..................................... 229-1 230 HEATING SYSTEM (Deleted - No longer applicable) l 231 VENTILATION SYSTEM (Deleted - No longer applicable) l 232 FIRE DETECTION SYSTEM............................................

232-1 232.1 Description.............................................

232-1 232.2 Decontamination and Dismantlement Considerations..................................... 232-1 233 PRIMARY PUMP SEAL WATER SYSTEM (Deleted - No longer applicable) 234 SAFE SHUTDOWN SYSTEM (Deleted -No longer applicable) 235 WATER CLEANUP SYSTEM (Deleted - No longer applicable) 236 VAPOR CONTAINER (Deleted - No longer applicable) 237 REACTOR SUPPORT STRUCTURE (Deleted - No longer applicable) 238 VAPOR CONTAINER POLAR CRANE (Deleted -No longer applicable) 239 RADIATION SHIELDING (Deleted - No longer applicable) 240 NEUTRON SHIELD TANK (Deleted -No longer applicable) 241 PIPE CHASES (Deleted -No longer applicable) l 242 FUEL TRANSFER CHUTE (Deleted - No longer applicable) l 243 YARD AREA CRANE AND SUPPORT STRUCTURE (Deleted - No longer applicable) l 244 ION EXCHANGE PIT (Deleted - No longer applicable) iii

YNPS FSAR REV. 6/05 TABLE OF CONTENTS (Continued)

Page l 245 PRIMARY VENT STACK (Deleted - No longer applicable) l 246 SPENT FUEL PIT AND SPENT FUEL PIT BUILDING (Deleted - No longer applicable) l 247 NEW FUEL VAULT (Deleted - No longer applicable) l 248 PRIMARY AUXILIARY BUILDING (Deleted - No longer applicable) 249 DIESEL GENERATOR BUILDING (Deleted - No longer applicable) l 250 WASTE DISPOSAL BUILDING (Deleted - No longer applicable) l 251 SAFE SHUTDOWN SYSTEM BUILDING (Deleted - No longer applicable) 252 POTENTIALLY CONTAMINATED AREA STORAGE BUILDINGS 1 AND 2 l

AND WAREHOUSE (Deleted - No longer applicable) l 253 COMPACTOR BUILDING (Deleted -No longer applicable) l 254 SERVICE BUILDING (Deleted - No longer applicable) 255 MISCELLANEOUS TANKS.................................................

255-1 255.1 Description.................................................

255-1 256 METEOROLOGICAL TOWER.................................................

256-1 256.1 Description.................................................

256-1 257 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)......................... 257-1 257.1 Description.................................................

257-1 257.2 Status..................................................

257-2 REFERENCES.................................................

257-2 SECTION 300 300 ENVIRONMENTAL-SITE CHARACTERISTICS..

300-1 300.1 Demography..

300-1 300.2 Geography and Land Use..

300-1 iv

YNPS FSAR REV. 6/05 TABLE OF CONTENTS (Continued)

Page 301 METEOROLOGY (GENERAL CLIMATE, SEVERE WEATHER).............................. 301-1 302 HYDROLOGY.................................................

302-1 303 GEOLOGY AND SEISMOLOGY.................................................

303-1 304 PLANT WATER SUPPLY.................................................

304-1 305 PLANT EFFLUENT.................................................

305-1 306 ENVIRONMENTAL RADIOLOGICAL STATUS.................................................

306-1 306.1 Facility Operating History.................................................

306-1 306.2 Radiological Scoping Survey.................................................

306-2 306.3 Radiological Environmental Monitoring Program.......................................... 306-2 REFERENCES.................................................

306-3 307 FINAL RADIOLOGICAL SURVEY - SITE RELEASE CRITERIA 307-1 SECTION 400 400 TRANSIENTS.................................................

400-1 400.1 General Overview 400-1 400.1.1 Radionuclide Release Limits.400-1 400.1.2 Assumptions.400-2 400.2 Event Identification Process.400-3 400.3 Summary..........................................................................................................400-3 REFERENCES.................................................

400-5 401 EVENTS AFFECTING OCCUPATIONAL HEALTH AND SAFETY.......................... 401-1 401.1 Radiological Events.................................................

401-1 401.2 Nonradiological Events.................................................

401-1 402 EVENTS AFFECTING PUBLIC HEALTH AND SAFETY........................................... 402-1 402.1 Off-Site Radiological Events.................................................

402-1 402.2 Nonradiological Events.................................................

402-1 v

YNPS FSAR REV. 6/05 TABLE OF CONTENTS (Continued)

Page 402.3 Radiological Events.402-1 402.4 Radiological Analysis Basis..

402-1 403 DECOMMISSIONING ACTIVITY EVENTS

.403-1 403.1 Decontamination Events..

403-1 403.2 Dismantlement Events..

403-2 403.3 Packaging Events..

403-3 403.4 Storage Events..

403-3 403.5 Materials Handling Events..

403-3 404 LOSS OF SUPPORT SYSTEM EVENTS.404-1 404.1 Loss of Off-Site Power.404-1 404.2 Loss of Cooling Water.404-1 404.3 Loss of Compressed Air.404-2 405 FIRE EVENTS.405-1 406 EXPLOSION EVENTS.406-1 406.1 Explosion Events: Vapor Container.406-1 406.2 Explosion Events: Potentially Contaminated Area Warehouse.406-2 406.3 Explosion Events: ISFSI.406-3 407 EXTERNAL EVENTS.407-1 407.1 Aircraft Impact..

407-1 407.2 Earthquake..

407-1 407.3 External Flooding..

407-2 407.4 Tornadoes and Extreme Winds..

407-2 407.5 Forest Fire..

407-3 407.6 Lightning Event...........................

407-3 407.7 On-Site Transportation Accidents..

407-3 408 SPENT FUEL STORAGE EVENTS.408-1 408.1 Events.408-1 vi

YNPS FSAR REV. 6/05 TABLE OF CONTENTS (Continued)

Page SECTION 500 500 ADMINISTRATION OF THE DECOMMISSIONING PLAN....................................... 500-1 500.1 Regulatory Basis for Administration of the Decommissioning Plan............... 500-1 REFERENCES......................................................

500-2 501 DECOMMISSIONING COST ESTIMATE AND FUNDING PLAN............................. 501-1 501.1 Decommissioning Cost Estimate.......................................................

501-1 501.2 Decommissioning Funding......................................................

501-1 502 DECOMMISSIONING ORGANIZATION AND RESPONSIBILITIES........................ 502-1 502.1 YAEC Commitment......................................................

502-1 502.1.1 Goals......................................................

502-1 502.1.2 Organizational Strategy......................................................

502-1 502.2 YAEC Organization and Functions......................................................

502-2 503 TRAINING PROGRAM......................................................

503-1 503.1 General Employee Training........................

503-1 503.2 Radiation Worker Training........................

503-1 503.3 Certified Fuel Handler........................

503-1 503.4 Specific Job Training........................

503-2 503.5 Nonradiation Worker Indoctrination........................

503-2 503.6 Training Records........................

503-2 REFERENCES.503-3 504 DEFUELED TECHNICAL SPECIFICATIONS.504-1 504.1 Description.504-1 REFERENCES 5304-3 505 PROCEDURES.505-1 505.1 Normal Defueled Operating Procedures.505-1 505.2 Emergency Procedures.505-1 vii

YNPS FSAR REV. 6/05 TABLE OF CONTENTS (Continued)

Page 506 MAINTENANCE PROGRAM.............................

506-1 507 RADIATION PROTECTION PROGRAM.............................

507-1 507.1 Introduction.............................................

507-1 507.2 Management Policies............................................

507-1 507.2.1 Management Policy Statement...........................

507-1 507.2.2 Administrative Policy...........................

507-2 507.2.3 ALARA Policy...........................

507-2 507.2.4 Regulatory Compliance Policy...........................

507-2 507.2.5 Waste Minimization and Disposal Policy...........................

507-3 507.2.6 Respiratory Protection Policy...........................

507-3 507.3 Decommissioning Exposure Projections............................................

507-3 507.4 ALARA Program............................................

507-3 507.4.1 General Program Description............................................

507-3 507.4.2 ALARA Program Organization and Responsibilities.................... 507-4 507.4.3 ALARA Training and Instruction............................................

507-5 507.5 Administrative Dose Control............................................

507-5 507.6 Radiation Work Permits............................................

507-5 507.7 Area Definitions and Postings............................................

507-6 507.8 External Dosimetry............................................

507-6 507.8.1 General Considerations............................................

507-6 507.8.2 Monitoring Whole Body Dose............................................

507-6 507.8.3 Dosimetry Quality Control............................................

507-7 507.9 Internal Dosimetry Control and Monitoring............................................

507-7 507.9.1 General Considerations............................................

507-7 507.9.2 Bioassay Program............................................

507-7 507.10 Respiratory Protection Program............................................

507-8 507.11 Radioactive Material Controls............................................

507-8 507.12 Surveillance............................................

507-9 507.13 Instrumentation............................................

507-9 507.14 Review and Audit............................................

507-9 507.15 Radiation Protection Program Performance Analysis................................... 507-10 viii

YNPS FSAR REV. 6/05 TABLE OF CONTENTS (Continued)

Page REFERENCES..................................

507-11 508 RADIOACTIVE WASTE MANAGEMENT 508-1 508.1 Solid Radioactive Waste Processing............................................

508-1 508.1.1 Solid Radioactive Waste Packaging............................................

508-1 508.1.2 Solid Radioactive Waste Storage Awaiting Shipment................... 508-2 508.1.3 Solid Radioactive Waste Shipment............................................

508-3 508.2 Liquid Radioactive Waste Processing............................................

508-3 508.3 Airborne Radioactive Waste Processing.............................................

508-4 508.4 Mixed Waste............................................

508-5 508.5 Radioactive Waste Minimization............................................

508-5 508.5.1 Radioactive Waste Reduction Program......................................... 508-5 508.5.2 On-Site Decontamination Methods............................................

508-6 508.5.3 Off-Site Radioactive Materials Processing.................................... 508-7 508.6 Decommissioning Radioactive Waste Projections.......................................... 508-8 REFERENCES..............................

508-9 509 TESTS.............................

509-1 510 OCCUPATIONAL SAFETY PROGRAM............................

510-1 510.1 Introduction......................................

510-1 510.2 Management Policy Statement......................................

510-1 510.3 Health and Safety Organization and Functions (Deleted)............................... 510-1 510.4 YAEC Occupational Safety Program......................................

510-1 510.5 Safety Training and Meetings......................................

510-2 511 NONRADIOACTIVE WASTE MANAGEMENT..................................

511-1 511.1 Introduction........................................

511-1 511.2 Management Policy Statement........................................

511-1 511.3 Hazardous Material Management........................................

511-1 511.4 Hazardous Waste Management........................................

511-1 511.4.1 Above-Ground Fuel Oil Storage Tanks........................................ 511-1 511.4.2 Underground Fuel Oil Storage Tanks (Deleted)............................ 511-2 511.4.3 Waste Oil........................................

511-2 511.4.4 PCB Contaminated Transformer Oil (Deleted)............................. 511-2 ix

YNPS FSAR REV. 6/05 TABLE OF CONTENTS (Continued)

Page 511.4.5 Mercury (Deleted)..........................

511-2 511.4.6 Asbestos Containing Materials..........................

511-2 511.4.7 Lead Based and PCB-Containing Paints..........................

511-2 511.5 Sampling and Remedial Actions..........................

511-3 511.6 Industrial Waste Management..........................

511-3 511.7 Training..........................

511-3 512 SECURITY PLAN..

512-1 512.1 General..

512-1 512.2 Site Access Control..

512-1 REFERENCES..

512-3 513 FIRE PROTECION..

513-1 513.1 Program Description..

513-1 513.2 Fire Protection Program Implementation..

513-2 REFERENCES..

513-3 514 DECOMMISSIONING QUALITY ASSURANCE PROGRAM..

514-1 514.1 Description..

514-1 REFERENCES..

514-2 515 EMERGENCY PLAN..

515-1 APPENDIX A - Plant Drawings x

YNPS FSARV REV. 6105 Table No.

200.2 200.3 200.4 306.1 400.1 400.2 400.3 400.4 504.1 LIST OF TABLES Title Decontamination Methods Mechanical Cutting/Removal Methods Thermal Cutting/Removal Methods Summary of Significant Radiological Contamination Events Exclusion Area Boundary Dose per Curie Conversion Factors (BASIS: 1/94)

Summary of Materials Released in Postulated Events Summary of Safety Analysis Requirements Summary of Decommissioning Planning Considerations from the Safety Analysis Technical Specification Changes for Transition to Permanently Defueled Condition Page 200-7 200-8 200-9 306-4 400-7 400-8 400-9 400-11 I

504-4 xi

YNPS FSAR REV. 6/05 LIST OF FIGURES Figure No.

257-1 300-1 Title 2003 Site Design Site Boundary and Plant Exclusion Area Page 257-3 300-2 xii

YNPS FSAR REV. 6/05 LIST OF ABBREVIATIONS, TERMS AND UNITS AC ALARA ANSI AOR ASME BOP CFH CFR Ci CRP DAC DC DECON DOE DPM EDM ENTOMB EPA FERC FSAR FTE GET GM HEPA ICRP INPO ISFSI ISR kV LLD LLW LPG M&TE fici MCS MDA MDC MDM mR MgR MWe Alternating Current As Low As is Reasonably Achievable American National Standards Institute Abnormal Occurrence Report American Society of Mechanical Engineers Balance of Plant Certified Fuel Handler Code of Federal Regulations Curie; Unit of Radioactivity = 3.7E-10 Disintegrations Per Second Component Removal Project Derived Air Concentration Direct Current Immediate Decontamination and Dismantlement Option Department of Energy Disintegrations per Minute Electrical Discharge Machining Encasement in Concrete with Future Dismantlement Option Environmental Protection Agency Federal Energy Regulatory Commission Final Safety Analysis Report Fuel Transfer Enclosure General Employee Training Geiger-Mueller High Efficiency Particulate Air (Filter)

International Commission on Radiological Protection Institute of Nuclear Power Operators Independent Spent Fuel Storage Installation Independent Safety Review Kilovolt Lower Limit of Detection Low Level Waste Liquid Propane Gas Measuring and Test Equipment 0.000001 Ci Main Coolant System Minimum Detectable Activity Minimum Detectable Concentration Metal Disintegration Machining 0.001 R 0.000001 R Megawatts Electric

xiii

YNPS FSAR REV. 6/05 MWt NCRP NIST NPDES NRC NSARC NSSS NST ODCM OSHA PAB PAG PASS PCA PCB pCi Person-rem PIC PIR POL PORC PSDAR R

RCA RCRA Rem REMP RETS RWP SAFSTOR SFP TEDE TLD TLG TS VC WD YDQAP YAEC YNPS Megawatts Thermal National Council on Radiation Protection and Measurements National Institute of Standards and Technology National Pollutant Discharge Elimination System Nuclear Regulatory Commission Nuclear Safety Audit and Review Committee Nuclear Steam Supply System Neutron Shield Tank Off-Site Dose Calculation Manual Occupational Safety and Health Administration Primary Auxilary Building Protective Action Guides Post-Accident Sample System Potentially Contaminated Area Polychlorinated Biphenyls 0.000000000001 Ci Collective Radiation Dose to a Population Pressurized Ion Chamber Plant Information Report Possession Only License Plant Operation Review Committee Post-Shutdown Decommissioning Activities Report Roentgen; Unit of Radiation Exposure Radiation Control Area Resource Conservation and Recovery Act Unit of Dose Equivalent Radiological Environmental Monitoring Program Radiological Effluent Technical Specifications Radiation Work Permit Delayed Decontamination and Dismantlement Option Spent Fuel Pit Total Effective Dose Equivalent Thermoluminescent Dosimeter TLG Engineering, Inc.

Technical Specifications Vapor Container Waste Disposal Yankee Decommissioning Quality Assurance Program Yankee Atomic Electric Company Yankee Nuclear Power Station xiv

YANKEE NUCLEAR POWER STATION (YNPS)

POST-SHUTDOWN DECOMMISSIONING ACTIVITIES REPORT (PSDAR)

(As of June 2005)

YNPS PSDAR REV. 6/05 TABLE OF CONTENTS Page Introduction and Overview of the PSDAR....................................................... 1 Historical Background......................................................

lI Decommissioning Alternative.......................................................

2 Description of Decommissioning Activities.................................................................................

. 2 Other Decommissioning Considerations.......................................................

4 Schedule of Decommissioning Activities.......................................................

8 Decommissioning Cost Estimate.......................................................

8 Environmental Impacts.......................................................

9 References............................................................................................................................................. 12 Table-i: Status of Systems. Structures and Components Described in FSAR................................... 13 ii

I YNPS PSDAR RE V. 6105 INTRODUCTION AND OVERVIEW OF THE PSDAR Pursuant to 10CFR50.82(a)(4)(i) and Regulatory Guide 1.185, this appendix contains information concerning post-shutdown activities remaining at the Yankee Nuclear Power Station (YNPS).

Section 50.82(a)(4)(i) requires that licensees develop a post-shutdown decommissioning activities report (PSDAR). Licensees with an approved decommissioning plan, however, may "replace their decommissioning plans with a PSDAR update that uses the format and content specified in this document" (Reference 13). The YNPS Decommissioning Plan (Reference 1), which was approved on February 14, 1995 and later became part of the Final Safety Analysis Report (FSAR), describes all remaining decommissioning activities, but in considerably more detail than that required in the PSDAR. Yankee Atomic Electric Company (YAEC) has elected to relocate pertinent information to a PSDAR, which conforms to the guidance in RG 1.185.

HISTORICAL BACKGROUND YNPS achieved initial criticality in 1960 and began commercial operations in 1961. The Nuclear Steam Supply System is a four loop pressurized water reactor designed by Westinghouse Electric Corporation. The original thermal power design limit of 485 MWt was upgraded to 600 MWt in 1963. The Turbine Generator, also designed by Westinghouse, was rated to produce 185 MWe.

Commercial operation ceased in 1992 after about 31 years of operation. During its operation, YNPS achieved an average capacity factor of about 74%.

YNPS shut down on October 1, 1991 in response to regulatory uncertainties associated with the integrity of the Reactor Vessel. During the outage and before February 26, 1992 all fuel assemblies, control rods, and neutron sources were removed from the Reactor Vessel and stored in the Spent Fuel Pit. A total of 533 fuel assemblies are stored in the Spent Fuel Pit. Plant systems required to support spent fuel storage and to support permanently defueled operations were in service until 2003.

On February 26, 1992, the YAEC Board of Directors decided to cease power operations permanently at YNPS. By letter, dated February 27, 1992 (Reference 2), YAEC notified the Nuclear Regulatory Commission (NRC) of the Company's decision to permanently cease power operations at the YNPS.

After notifying the NRC, YAEC initiated decommissioning planning and other plant closure activities to safely reduce radioactivity at the site to residual levels, allowing release of the site for unrestricted access.

On August 5, 1992, the NRC amended the YNPS Facility Operating License (DPR-3) to possession only status (Reference 3). This, combined with other amendments and program changes, formed the basis of the Decommissioning Plan. The plan (Reference 1) was submitted by YAEC in accordance with the requirements of 1 OCFR50.82(a) [pre-1996], that required submittal of a proposed Decommissioning Plan within two years of the permanent cessation of operations. The Decommissioning Plan was subsequently approved on February 14, 1995 (Reference 4). A commitment from the approval process required that the Decommissioning Plan be incorporated into the FSAR.

Plant closure activities, which were commenced following the decision to cease power operations, will continue in accordance with applicable regulatory requirements and YAEC's commitment to I

I YNPS PSDAR REV. 6/05 maintain the facility in a safe and economical manner. These activities have included site security modifications, control rod disposal, decontamination, disposal of radioactive components, lay-up of plant equipment, and facility modifications to improve plant operations.

After deciding to close YNPS permanently, YAEC reviewed the plant licensing basis to evaluate the applicability of existing Technical Specifications and NRC regulations to a permanently defueled condition. Beginning in June 2002, spent fuel and GTCC waste were transferred to dry casks and placed on the onsite Independent Spent Fuel Storage Installation (ISFSI). The transfer activities were concluded in June 2003. The remaining Technical Specifications that are applicable have been relocated to the Yankee Decommissioning Quality Assurance Program (Reference 14).

DECOMMISSIONING ALTERNATIVE Following an evaluation of the three decommissioning alternatives, YAEC selected DECON as the most appropriate alternative for decommissioning YNPS. YAEC's choice of a decommissioning alternative is consistent with NUREG-0586, "Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities,"(Reference 6). In that document, the NRC concluded that DECON and SAFSTOR alternatives are reasonable for decommissioning a pressurized water reactor. Implementation of the DECON alternative is within the 60 year limit (after cessation of operation) in 1 OCFR50.82(a)(3).

DESCRIPTION OF DECOMMISSIONING ACTIVITIES Since 1993 YAEC has removed and disposed of the steam generators, pressurizer, reactor vessel and reactor vessel internals. The reactor vessel internals components, which are Greater-Than-Class-C (GTCC) waste, remain onsite and are stored within a dry cask along with the 15 dry casks of spent fuel located at the onsite ISFSI. YAEC has sought accelerated acceptance of its spent fuel by the Department of Energy (DOE) in accordance with the current fuel disposal contract. The DOE's position is that they have not yet determined whether priority will be accorded shutdown reactors, or if priority is granted, under what specific circumstances it might be granted.

As of June 2005, all plant systems and components not required to support the storage of spent fuel have been dismantled and disposed of in accordance with the YNPS Decommissioning Plan and Final Safety Analysis Report (References 1 and 12).

The current status of the systems, structures, and components are described in Table-1. The purpose of this section is to identify the dismantlement and decontamination activities associated with the remaining systems, structures, and components at YNPS.

The current phase of decommissioning involves the dismantlement and decontamination of the SFP and its supporting systems, structures, and components. In the final phase of decommissioning, the possession only license will be terminated.

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I YNPS PSDAR RE:V. 6105 A.

Remaining Systems, Structures, And Components Not Necessary For Spent Fuel Storage (Phase 1)

After removing systems and components from an area or building, contaminated concrete, steel, and other building materials were decontaminated or removed. The structures listed below are not required to support operation of the ISFSI and have been demolished as of June 2005. Disposal of debris from the demolition activities is ongoing and expected to be complete by the end of 2005. Certain debris meeting appropriate acceptance criteria may remain onsite as approved fill in the Beneficial Use Determination permit.

Vapor Container

Reactor Support Structure

Upper and Lower Pipe Chases

Fuel Transfer Chute

South, East, and West Walls of Ion Exchange Pit

Primary Auxiliary Building

Waste Disposal Building

Safe Shutdown Building

Potentially Contaminated Area (PCA) Buildings 1 and 2

Compactor Building

Turbine Building

Service Building

Storage Warehouse Final status surveys are being completed to verify that structures and open land areas meet the release criteria. Independent verification of the results by the NRC will allow for the release of the individual surveyed structures, foundations and open land areas as non-radiologically controlled material. In order to facilitate remediation, the superstructures were demolished prior to remediating substructure and soils work beneath the structures.

Measures have been implemented to prevent recontamination of surveyed areas prior to final license termination. Partial release of the site may also be considered to reduce the site to what is needed for the dry cask storage facility.

B.

Systems, Structures, And Components Associated With Storage Of Spent Fuel In The Spent Fuel Pool (Phase 2)

With all the spent fuel and GTCC waste from the SFP having been transferred to the ISFSI, the remaining components of the systems listed below have been dismantled and disposed of:

Radiation Monitoring System

Fuel Handling System

SFP Cooling and Purification System 3

I YNPS PSDAR REV. 6105

Temporary Liquid Waste Processing System

Auxiliary Service Water System

Demineralized Water System

Electrical System

Ventilation System

Fire Protection and Detection System Systems and components from an area or building associated with the SFP Island, contaminated structural concrete, steel and other building materials have been decontaminated or removed. The structures listed below were decontaminated during decommissioning of the SFP Island.

Yard Area Crane and Support Structure

Ion Exchange Pit

Primary Vent Stack

Spent Fuel Pool and SFP Building New Fuel Vault

Potentially Contaminated Area Warehouse

Service Building C.

License Tcrmination (Phase 3)

The final phase of decommissioning will take place after all spent fuel and GTCC waste is taken off-site in order to complete the dismantlement and decontamination of the ISFSI.

Decommissioning of the ISFSI would consist primarily of the disposal of the Vertical Concrete Casks (VCCs) and the concrete storage pad. The VCC design minimizes neutron activation, thereby generating minimal radioactive waste. This waste should qualify for disposal at a low-level radioactive waste disposal site.

The I OCFRPart 50 license may be terminated after the last stage of final status surveys and independent NRC verification. Site restoration activities will then be conducted.

OTHER DECOMMISSIONING CONSIDERATIONS The dismantlement and, if necessary, decontamination of contaminated systems, structures and l components was accomplished using the following three approaches: decontamination in place, dismantlement and decontamination, or dismantlement and disposal. Furthermore, a combination of l these methods was used to reduce contamination levels, worker radiation exposures, and project costs. General considerations applicable to these activities are described in detail in Section 200 of l the FSAR and are summarized below. An approved work control process is used to perform demolition, decommissioning, and or decontamination work activities in accordance with an approved Independent Safety Review.

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I YNPS PSDAR REV. 6105 A.

General Decommissioning Activities Relating to Removal of Radiological Components

& Structures Components were safely and efficiently removed using techniques and appropriate methods for the particular circumstances and consistent with Decommissioning Work Packages.

Openings in components were typically covered and sealed to minimize the spread of contamination. The components were moved to a processing area for volume reduction and packaging into containers for shipment to a processing facility for decontamination or a low-level radioactive waste disposal facility. Buried contaminated components (e.g., piping, drains, etc.) will be decontaminated in place or excavated.

B.

Decontamination Methods Contaminated systems and components were removed and sent to an off-site processing facility or to a low level radioactive waste disposal facility. On-site decontamination of systems and components were generally be limited to activities needed to maintain personnel exposure as low as is reasonably achievable, to expedite equipment removal, and to control the spread of contamination.

Application of coatings and hand wiping were the preferred methods for stabilizing or removing loose surface contamination. If other methods were utilized (e.g., grit blasting, high pressure water), airborne contamination control and waste processing systems were used as necessary to control and monitor any releases of contamination.

Contaminated and activated concrete as well as other contaminated materials was removed and sent to a low-level radioactive waste disposal facility. Concrete removal methods, such as scabbling and scarifying, controlled the removal depth to minimize the waste volume produced. Vacuuming the dust and debris effluent with HEPA filtration minimized the need for additional respiratory protection control measures. While these methods represent the most practicable and widely used decontamination methods available at this time, YAEC also considered new decontamination technologies as appropriate.

C.

Dismantlement Methods YAEC used two basic dismantlement methods:

Mechanical Methods - Mechanical methods machine the surfaces of the material that is being cut. Typically, these methods are capable of cutting remotely without generating significant amounts of airborne contamination. This attribute made these methods attractive for removing most of the contaminated piping, equipment, and components at YNPS. The outside diameter machining method, in particular, was best suited for cutting large bore contaminated piping.

5

I YNPS PSDAR REV. 6105 Thermal Methods - Thermal methods melt or vaporize the surfaces of material. The cutting debris is transported from the cut region with a gas jet or water spray. Although thermal methods are significantly quicker than mechanical methods, they have high power requirements and generate airborne contamination when used on contaminated systems in air. Generation of airborne contamination was easily controlled when the method was used underwater. Thermal methods are suitable for segmenting large vessels in areas that can easily be sealed, filtered, or maintained underwater. The method is also suitable for use at a cutting station with air filtration. Thermal methods are appropriate for removing structural steel if it has been decontaminated or if a local contamination envelope with HEPA filtration is established. Appropriate lead and/or PCB paint removal controls must also be implemented.

D.

Special or Unusual Programs There are no special or unusual programs. All procedures and processes used at YNPS are consistent with those considered in the Final Generic Environmental Impact Statement (FGEIS).

E.

Removal of Low Level Radioactive Waste (LLWN) and Compaction or Incineration LLW is processed in accordance with plant procedures and sent to LLW disposal facilities.

While no incineration of low level Radwaste (to remove radionuclides) will occur onsite, YAEC has used an off-site licensed facility. YAEC has no intention, however, to use onsite compaction at this time.

F.

Soil Pavement and Concrete Remediation Soils and pavement are being surveyed and characterized in accordance with the site radiological characterization program. As necessary, soils, concrete and pavement is remediated (i.e., removed, processed and disposed of at a licensed facility) if determined to contain contamination levels above the derived concentration guideline levels (DCGLs) established in the License Termination Plan.

G.

Processing and Disposal Site Locations Currently, there are several facilities available for (1) the processing of waste materials to achieve volume reduction prior to disposal or (2) the disposal of low-level radioactive waste.

These locations include (but are not limited to) GTS Duratek - Barnwell, South Carolina; Envirocare - South Clive, Utah; GTS Duratek - Oak Ridge, Tennessee; RACE in Memphis Tennessee H.

Removal of Mixed Wastes Mixed wastes are managed according to all applicable federal and state regulations including NRC handling, storage, and transportation regulations. Mixed wastes from YNPS are 6

I YNPS PSDAR RE V. 6/05 transported only by authorized and licensed transporters and shipped only to authorized and licensed facilities. YAEC has used an appropriate approved process to render mixed waste non-hazardous onsite. A thermal desorption system was used to remove PCB materials from soils.

Storage/Removal of Spent Fuel and GTCC Waste In June 2003, YAEC completed transfer of spent fuel and GTCC waste to an onsite Independent Spent Fuel Storage Installation (ISFSI). YAEC cannot make a precise determination of when spent fuel and GTCC waste will be removed from the YNPS because the availability of a licensed DOE high level waste repository is uncertain. Currently, YAEC expects that the turnover of spent fuel and GTCC waste to the DOE will be completed in 2022.

J.

License Termination Plan, Final Radiological Survey and Site Release Criteria The objective for decommissioning the Yankee Nuclear Power Station (YNPS) is to reduce residual radioactivity to levels that permit release of the site for unrestricted use and for termination of the I OCFR50 license, in accordance with the Nuclear Regulatory Commission's (NRC's) site release criteria set forth in 1 OCFR20, Subpart E. The purpose of the YNPS License Termination Plan (LTP) is to satisfy the requirements of I OCFR50.82, "Termination of License" using the guidance provided in Regulatory Guide 1.179, "Standard Format and Content of License Termination Plans for Nuclear Reactors". NRC staff review guidance, in the form of NUREG-1700 and NUREG-1757 has also been considered.

In May 1997, YAEC submitted to the Commission for approval an LTP for YNPS, pursuant to IOCFR50.82, as amended by 62 Fed. Reg. 39091 (July 29,1996). YAEC's LTP employed a survey methodology based on the "Manual for Conducting Radiological Surveys in Support of License Termination (Reference 9)," known as the NUREG-5849 Methodology.

Subsequently, the Commission, jointly with a group of other federal agencies, approved an alternative survey methodology, known as NUREG-1575, "Multi-Agency Radiation Survey and Site Investigation Manual" (MARSSIM - Reference 10). In May 1999 (Reference 11),

YAEC advised the Commission that it intended to shift from the NUREG-5849 Methodology to MARSSIM and withdrew its previously submitted LTP application. YAEC submitted a new LTP in 2003 incorporating the more recent regulatory guidance and descriptive of the MARSSIM final status survey (FSS). FSS is being conducted to verify that structures and open land areas meet the release criteria. An independent NRC contractor is also conducting verification surveys, to confirm YNPS FSS results that the remaining structures and open land areas meet the unrestricted release criteria.After final status surveys and NRC verification, individual surveyed structures and open land areas will be released. YAEC will, nevertheless, maintain control over the site until termination of its IOCFRPart 50 license.

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I YNPS PSDAR REV. 6105 K.

Site Restoration Following termination of the YNPS possession only license by the NRC, YAEC will complete site restoration activities. Many of these activities have been initiated during the dismantlement period. Activities associated with the Vapor Container included removal of internal structures, disassembly of the Vapor Container shell, and demolition of the Reactor Support Structure and other concrete structures. The remaining plant structures are being demolished. All building foundations will be back filled with concrete and/or structural fill.

Site areas will be graded and landscaped as necessary.

SCHEDULE OF DECOMMISSIONING ACTIVITIES As stated above, decommissioning will be completed in three phases. The first phase consisted of the decontamination and dismantlement of remaining systems and components that did not support fuel storage. Next the spent fuel was removed from the SFP and the SFP was drained in 2003. The current phase of decommissioning involves the dismantlement and decontamination of the SFP and its supporting systems, structures and components. The final phase of decommissioning will involve the termination of the possession only license. License termination of the ISFSI will occur after all spent fuel has been taken off-site. All decommissioning activities accomplished to date have resulted in no significant adverse environmental impacts.

YAEC expects to complete the current phase of dismantlement and decontamination and final status surveys by late 2005. The design and construction of a dry cask storage facility was completed in 2001. Fuel transfer activities commenced in 2002 and were completed in 2003.

Following the transfer of spent fuel and GTCC waste from the SFP, decommissioning of the SFP l island was begun and should be completed over a period of approximately two years, including final status surveys. The dry cask storage facility is expected to be operated from 2002 to 2022, when the last fuel assembly is assumed to be taken off-site. Using this assumption, the YNPS license will be terminated after the dry cask storage facility is decommissioned.

Planning sequences and dates are based on current knowledge and could change in the future. Dry storage is addressed in the spent fuel management plans contained in the FSAR. Yankee will continue to inform the NRC of all major changes to the planned decommissioning activities in accordance with IOCFR50.82(a)(7).

DECOMMISSIONING COST ESTIMATE The YNPS Decommissioning Plan (Reference 1) was submitted to the NRC in December 1993 and included a cost study for operating the facility through a safe storage period, decommissioning the facility, restoring the site, and storing spent fuel until its transfer to the DOE. In October 1994, Yankee completed a revised cost study to assist the NRC in its review of the Decommissioning Plan and to fulfill a commitment to Federal Energy Regulatory Commission (FERC). This 1994 cost study was based on the assumption that dismantlement activities would not begin until a low-level radioactive waste disposal site became available to Yankee.

In June 1995, the State of South Carolina re-opened the low-level waste facility in Barnwell, South Carolina to radioactive waste generators throughout the United States. In response, Yankee updated 8

I YNPS PSDAR REV. 6/05 the cost estimate to reflect several significant changes in parameters affecting decommissioning costs. This study, called the 1995 Cost Study, was filed with FERC in August 1995. This study was a site-specific cost study that aidjusted the 1994 Cost Study for differences in decommissioning timing, waste disposal costs, and one year of escalation. The 1995 Cost Study estimate of "to-go" costs remaining as of January 1995 was $303.2 million. In addition, as part of the final December 1995 FERC settlement, Yankee was allowed to collect another $3.2 million in the decommissioning trust fund to adjust for adjudicatory delays during re-approval of the Decommissioning Plan, bringing the total January 1995 "to-go" cost to $306.4 million (1995 dollars).

As required by the FERC settlement, an updated cost estimate was filed in December 1999. The total decommissioning costs remaining (as of January 1, 1999) are $246 million in constant value 1999 dollars. The total cost of decommissioning (expressed in 1999 dollars) is summarized below:

Expended dollars (1993 - 1998)

$207,100,000

Dismantlement Activities 147,700,000

Spent Fuel Storage 98.300.000

Total Cost to Remove/Dismantle YNPS

$453,100,000 Yankee has collected its decommissioning funds through its Power Contracts. The collections were deposited in an independent and irrevocable trust at a commercial bank, with the principal and interest used to discharge decommissioning obligations as they are incurred. This trust is in compliance with 1 OCFR50.75(e)(1)(ii) and a copy of the trust document has been provided to the NRC. The Power Contracts obligate the purchasers for the full costs of decommissioning YNPS, including spent fuel. The FERC orders received by Yankee acknowledge the continuing obligation of the purchasers with respect to the full cost of decommissioning YNPS. The periodic reviews of decommissioning cost studies mandated by FERC provide the mechanism for updating the required payments under the Power Contracts to assure adequate funds for that purpose.

Yankee Atomic submitted a new decommissioning cost estimate to FERC in April 2003 to seek additional revenue requirements billed through 2010 for the period through December 2022. That estimate supercedes the 1999"to go" cost estimate. As of 1/1/2003 the remaining cost to complete NRC required decommissioning activities is $121.1 million stated in 2003 dollars. An update was provided to the NRC on March 31, 2005 providing the current status of the Decommissioning fund (Reference 15).

ENVIRONMENTAL IMPACTS YAEC prepared an Environmental Report [Reference 5] to evaluate all actual or potential environmental impacts associated with the proposed decommissioning activities. This evaluation used as its basis NUREG-0586, "Final Generic Environmental Impact Statement (FGEIS) on Decommissioning of Nuclear Facilities," [Reference 6] and the site-specific environmental assessment from the re-capture of the construction period time duration [Reference 7].

This Environmental Report concluded that the impacts due to decommissioning of the YNPS will be bounded by the previously issued environmental impact statements, specifically the FGEIS and previously issued environmental assessments. This is principally due to the following reasons:

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I YNPS PSDAR REV. 6/05

The postulated impacts associated with the method chosen, DECON, have already been considered in the FGEIS.

There are no unique aspects of the plant or decommissioning techniques to be utilized that would invalidate the conclusions reached in the FGEIS.

The methods to be employed to dismantle and decontaminate the site are standard construction based techniques fully considered in the FGEIS.

The site-specific person-rem estimate for all decommissioning activities has been conservatively calculated using methods similar to and consistent with the FGEIS.

Specifically, this review concludes that the YAEC 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 eliminated.

The site will be returned to a condition that will be acceptable for unrestricted use.

The thermal impact on the Deerfield River from facility operations will be eliminated.

Noise levels in the vicinity of the facility will be reduced.

Hazardous materials and chemicals will be removed.

Local traffic will be reduced (fewer employees, contractors and materials shipments than are required to support an operating nuclear power plant).

Furthermore, the YNPS decommissioning will be accomplished with no significant adverse environmental impacts in that:

No site specific factors pertaining to YNPS would alter the conclusions of the FGEIS.

Radiation dose to the public will be minimal.

Radiation dose to decommissioning workers will be a fraction of the operating experience.

Decommissioning is not an imminent health or safety problem and will generally have a positive environmental impact.

The total radiation exposure impact for decommissioning was estimated in the Decommissioning Plan (Reference 1) to be approximately 744 person-rem. This estimate was re-evaluated in 1996, resulting in a lower value of 580 person-rem (Reference 8). The estimate of 580 person-rem has been re-evaluated and the current estimate is 594 person-rem: The actual exposure, through 12/31/04, for decommissioning activities is 592.4 person-rem. The estimated radiation exposure for completion of the decommissioning activities at YNPS accounts for the one remaining dose 10

I YNPS PSDAR REV. 6/05 intensive task of decontamination and dismantlement of the spent fuel pit. This "to go" exposure has been estimated to be 1.6 person-rem (excludes public and transportation dose).

Radiation exposure due to transportation of radioactive waste has been conservatively estimated to be approximately 7 person-rem. This value is bounded by the FGEIS value of 100 person-rem for transportation occupational exposure.

Radiation exposure to off-site 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 10CFRPart 20 limits and the ALARA dose objectives of 10CFR50, Appendix I. This conclusion is supported by the YNPS Annual Effluent Release Reports in which individual doses to members of the public are calculated for station liquid and gaseous effluents.

No significant impacts are expected from the disposal of low-level radioactive waste (LLW). The total volume of YNPS low-level radioactive waste for disposal was estimated in the Decommissioning Plan to be approximately 132,000 ft3. A review of the annual effluent reports filed with the NRC has determined that, through the end of 2002, 144,184 ft3 of LLW from YNPS has been shipped off-site for burial. The waste volume that remains at YNPS is bounded by the FGEIS estimate of 647,670 ft3 for a reference PWR.

Since the approval of the Decommissioning Plan (Reference 1) and the issuance of the Decommissioning Environmental Report (Reference 5), YNPS has identified the presence of solid Polychlorinated Biphenyl's (PCB's) in some paint coatings. As in the case of radiologically contaminated lead paint, asbestos, and other hazardous materials, contaminated paint that contains PCB's will be managed according to all applicable federal and state regulations.

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 YNPS 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 significant socioeconomic impacts, other than those associated with cessation of operation (loss ofjobs and taxes), or impacts to local culture, terrestrial or aquatic resources such as the Sherman Reservoir and Deerfield River have been identified.

11

I YNPS PSDAR REV. 6/05 REFERENCES

1.

Letter, J. K. Thayer (YAEC) to T. E. Murley (USNRC), Decommissioning Plan for Yankee Nuclear Power Station (YNPS), December 20, 1993.

2.

Letter, A. C. Kadak (YAEC) to T. E. Murley (USNRC), Permanent Cessation of Power Operations at the Yankee Nuclear Power Station, BYR 92-024, February 27, 1992.

3.

Letter, M. B. Fairtile (USNRC) to J. M. Grant (YAEC), Issuance of Amendment #142 to Facility License DPR Yankee Nuclear Power Station, August 5, 1992.

4.

Letter, M. B. Fairtile to J. A. Kay (YAEC), Order Approving the Decommissioning Plan and Authorizing Decommissioning of the Yankee Nuclear Power Station, February 14,1995.

5.

Yankee Nuclear Power Station Decommissioning Environmental Report, December 1993.

6.

NUREG-0586, Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities, August 1988.

7.

Letter, M. B. Fairtile (USNRC) to G. Papanic (YAEC), dated June 2, 1988.

8.

USNRC Atomic Safety and Licensing Board Docket No. 50-029-DCOM, Supplemental Affidavit of Russell A. Mellor, September 3, 1996.

9.

NUREG-5849, "Manual for Conducting Radiological Surveys in Support of License Termination," Draft, June 1992.

10.

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

11.

Letter, D.K. Davis (YAEC) to USNRC, "Withdrawal of Proposed License Amendment to Approve Yankee Nuclear Power Station's License Termination Plan," May 25, 1999.

12.

Yankee Nuclear Power Station Final Safety Analysis Report.

13.

Regulatory Guide 1.185, Standard Format and Content for Post-Shutdown Decommissioning Activities Report, July 2000.

14.

Letter J.B. Hickman (USNRC) to J.A. Kay (YAEC), Issuance of Amendment #157 RE:

Deletion of Operational and Administrative Requirements Following Fuel Transfer to ISFSI, NYR 03-027, April 18, 2003.

15.

Letter Jack Rollins (YAEC) to DCD, Decommissioning Funding Assurance-(10CFS50.75(f), March 31, 2005.

12

YNPS PSDAR REV. 6/05 TABLE-1: STATUS OF SYSTEMS, STRUCTURES AND COMPONENTS DESCRIBED IN FSAR FSAR System. Structure, or Component Section Status Note I Reactor Vessel 201 Removed.

Steam Generators 202 Removed.

Main Coolant System 203 Removed.

Pressure Control and Relief System 204 Removed.

Charging and Volume Control System 205 Removed.

Chemical Shutdown System 206 Removed.

Purification System 207 Removed.

Component Cooling System 208 Removed.

Primary Plant Corrosion Control System 209 Removed.

Primary Plant Sample System 210 Removed.

Waste Disposal System 211 Removed.

Shutdown Cooling System 212 Removed.

Primarv Plant Vent and Drain System 213 Remox'ed.

Emergency Core Cooling System 214 Removed.

Radiation Monitoring System 215 Removed.

VC Ventilation and Purge System 216 Removed VC Hleating and Cooling System 217 Removed.

Post-Accident Hydrogen Control System 218 Removed.

Containment Isolation System 219 Removed.

Fuel Handling Equipment System 220 Removed.

SFP Cooling and Purification System 221 Removed.

Lain Steam System 222 Removed.

Feedwater System 223 Removed.

Steam Generator Blowdown SYstem z

24 Removed Emergency Feedwvater System 225 Removed.

Service Water System 226 Removed.

Demineralized Water System 227 Removed.

Compressed Air S stem 228 Removed.

Electrical System 229 Partially removed. in service. [Note 21 Hleating System 230 Removed.

Ventilation System 231 Removed.

Fire Protection and Detection System 232 Removed.

Primary Pump Seal Water System 233 Removed.

13

I YNPS PSDAR REV. 6/05 I

I I

I FSAR System, Structure, or Component Section Status Note I Safe Shutdown System 234 Removed.

Water Cleanup System 235 Removed.

Vapor Container 236 Removed Reactor Support Structure 237 Removed Vapor Container Polar Crane 238 Removed Radiation Shielding 249 Removed Neutron Shield Tank 240 Removed.

Pipe Chases 241 Removed Fuel Transfer Chute 242 Removed.

Yard Area Crane and Support Structure 243 Removed.

Ion Exchange Pit 244 Removed.

Primary Vent Stack 245 Removed.

Spent Fuel Pit and Spent Fuel Pit Building 246 Removed.

New Fuel Vault 247 Removed.

Primary Auxiliary Building 248 Removed.

Diesel Generator Building 249 Removed.

Waste Disposal Building 250 Removed.

Safe Shutdown System Building 251 Removed.

Potentially Contaminated Area Storage 252 Removed.

Buildings I and 2 and Warehouse Compactor Building 253 Removed.

Service Building 254 Removed.

Miscellaneous Tanks 255 Most removed; one tank remaining to be decontaminated.

Meteorological Tower 256 Function no longer required. A 10-meter tower located at the ISFSI provides wind speed and direction for onsite assessment purposes.

I I

Note 1: Certain foundations, duct banks, piping and slabs will remain onsite, once the structures have been determined to be acceptable from an environmental and radiological standpoint, as required by the MADEP Beneficial Use Determination. This material will also comply with the LTP.

Note 2: With all spent fuel and Greater-Than-Class-C waste having been removed from the Spent Fuel Pool, the remaining portions of this system or structure that were in service to support SFP operations have been decontaminated and dismantled. The portions of the system required to support the ISFSI will continue in service.

14

YNPS FSAR REV. 6/05 100 SPENT FUEL MANAGEMENT PLAN 100.1

Background

This section documents the YNPS spent fuel management plan in accordance with 10CFR50.54(bb)

(Reference 100-1). This regulation also requires a funding plan for spent fuel storage. The spent fuel storage cost estimate and funding plan are presented in the PSDAR.

YAEC contracted with NAC International to transfer the spent fuel from the spent fuel pit to a dry cask storage facility. The location of the new storage facility is at the south end of the YNPS site (see figure 257-1). Currently, there are 533 fuel assemblies stored in 15 NAC Multi-Purpose Canisters (MPCs) on the ISFSI pad. These fuel assemblies were discharged from the reactor between 1972 and 1992. There were also a number of failed fuel pins that were consolidated into a Reconfigured Fuel Assembly (RFA) which was then transferred to a dry cask. The Greater than Class C (GTCC) radioactive waste in the form of canisters containing reactor vessel internals (core baffle) and dross material and having the same external dimensions as fuel assemblies have also been transferred to a NAC-MPC. Several miscellaneous GTCC radioactive items also are stored in the NAC-MPC containing the GTCC (e.g. neutron sources, fuel assembly cage components, fixed incore detector wires, etc.).

YAEC is currently seeking accelerated acceptance of YNPS's spent fuel by the Department of Energy in accordance with the current fuel disposal contract. The Department of Energy's current position is that they have not yet determined whether priority will be accorded shutdown reactors, or if priority is granted, under what specific circumstances it might be granted. For planning purposes, fuel shipments are assumed to be completed in 2022. This projection is based on the Department of Energy's Acceptance Priority Ranking, Annual Capacity Report, and an extrapolation beyond the 10 year Department of Energy outlook. For planning purposes, YAEC's current decommissioning cost estimate assumes storage of fuel in the dry storage facility until 2022.

100.2 Special Nuclear Material Control Information regarding Special Nuclear Material Control is available for inspection by authorized personnel.

100-1

YNPS FSAR RE V. 6/05 REFERENCES 100-1 59-FR-10267, Notification of Spent Fuel Management and Funding Plans By Licensees of Prematurely Shut Down Power Reactors, March 4, 1994.

100-2 YRP 435/92, Spent Nuclear Fuel Storage Study Report and Recommendations, B. W. Holmgren, J. M. Buchheit, R. A. Mellor to J. K. Thayer, October 9, 1992.

100-3 YRP 303/93, Impact of Wet Spent Fuel Storage on Decommissioning, P. A. Rainey to R. A. Mellor, July 15, 1993.

100-2

YNPS FSAR REV. 6/05 200 DECOMMISSIONING ACTIVITIES AND PLANNING 200.1 Introduction This section presents a description of decommissioning activities and tasks for the implementation of the dismantlement activities at YNPS. The information presented in this section reflects initial planning of decommissioning activities. YAEC intends to complete more detailed planning prior to initiating each decommissioning activity. Detailed planning includes engineering design; ALARA planning; and cost, schedule, and resource refinement.

Decommissioning will be completed in three phases. The first phase which has been completed consisted of the decontamination and dismantlement of certain structures systems and components which do not support fuel storage. After the removal of all spent fuel from the Spent Fuel Pit (SFP),

the second phase of decommissioning will involve the dismantlement and decontamination of the SFP and the remaining systems, structures and components. The final phase of decommissioning will involve the termination of the possession only license. License termination will occur after all spent fuel has been removed off-site and, the dry cask storage facility is decommissioned. It is expected that decommissioning activities will be accomplished with no significant adverse environmental impacts.

200.2 Decommissioning Plan Approval A Decommissioning Plan (Reference 500-1) was prepared and submitted in response to the requirements of 1 OCFR50.82, "Application for Termination of License," using guidance from Draft Regulatory Guide DG-1005 (Reference 500-2). The Decommissioning Plan provided the framework for the decommissioning process by which the YNPS site will be returned to an unrestricted use condition.

The Decommissioning Plan was approved by NRC on February 14, 1995 (Reference 500-3).

Following NRC approval, the Decommissioning Plan was incorporated into the FSAR in accordance with Reference 500-4 and the FSAR was conformed to the Decommissioning Plan using the process described in 10CFR50.71(e). Essential features and functions that will be relied upon during decommissioning are now described in the FSAR.

In 2000, Yankee created a Post-Shutdown Decommissioning Activities Report (PSDAR) within the FSAR. NRC Draft Regulatory Guide DG-1071 recommends that licensees with an approved Decommissioning Plan (Dplan) "extract pertinent detail from the decommissioning plan and submit a PSDAR update in the format and content specified by [DG-1071]." Based on the NRC draft guidance, Yankee segregated, updated and condensed certain information concerning post-shutdown decommissioning activities in a manner that conforms to the standard format and content of a PSDAR.

200-1

YNPS FSAR REV. 6/05 200.3 Decontamination and Dismantlement: General Information 200.3.1 Overview This section presents a general description of the decontamination and dismantlement activities that are necessary to decommission YNPS. The information presented in this section provides sufficient detail to address the adequacy of the programs and procedures, ensuring safe and economic decommissioning of YNPS. The information in this section has been incorporated into the more detailed planning that will be completed prior to initiating each decommissioning activity. Detailed planning includes engineering design; ALARA planning; and cost, schedule, and resource refinement. Sections 201 through 256 present specific information for YNPS systems, structures and components.

The description of decontamination and dismantlement options includes the words "should" and "must" to describe alternatives. The word "should" implies that the alternative is preferred, however, alternatives may be available that are equally acceptable. The word "must" implies that the alternative is based on a programmatic, regulatory or safety analysis requirement. If another alternative is chosen, the corresponding requirements must be re-evaluated to ensure that the original intent is not adversely affected.

Before the start of the dismantlement activities, a decommissioning administrative and engineering organization will be mobilized. The following activities will occur:

Initiation of detailed project planning.

Preparation of engineering specifications and procedures.

Procurement of special equipment needed to support decommissioning.

Negotiation of service contracts required for decommissioning activities.

Reactivation and return to service of systems required for decommissioning.

The engineering and preparation phase is followed by the initiation of plant dismantlement activities.

The contaminated systems will be removed, packaged, and either shipped to an off-site processing facility or shipped directly to a low level radioactive waste disposal facility. Decontamination of plant structures will be completed concurrently with the equipment and system removal process.

Structure decontamination will include a variety of techniques ranging from high pressure water washing to removal of concrete to allow release of the structures. Contaminated structural material will be packaged and either shipped to a processing facility for decontamination or shipped directly to a low level radioactive waste disposal facility.

Following the removal of contaminated systems, structures and components, a comprehensive final radiation survey will be conducted. The survey will verify that radioactivity has been reduced to sufficiently low levels allowing unrestricted release of the site. Successful completion of the final survey will be demonstrated through a verification survey completed by an independent contractor selected by the NRC.

200-2

YNPS FSAR RE1,V. 6105 200.3.2 Detailed Planning and Engineering Activities YAEC is the prime entity with license responsibility for YNPS decommissioning. In this position, YAEC has direct control and oversight over all decommissioning activities. This role is similar to that taken by YAEC during the 31 year operation of YNPS. In that role, YAEC provided operational, technical, licensing, project management, and contractor oversight for all plant systems, structures and components.

Detailed project implementation plans will be developed to support decontamination and dismantlement activities before these activities are initiated. The plans will be used as a project management tool to support detailed engineering activities and ALARA Program implementation, estimate decommissioning labor requirements, and manage decommissioning cost and schedule.

Demolition and abatement activities will be implemented using plant approved programs and procedures that have received appropriate reviews (e.g., ISR).

200.3.3 General Decontamination and Dismantlement Considerations The following are general decontamination and dismantlement considerations that will be used during the decommissioning of systems, components and structures at YNPS. Specific decontamination and dismantlement considerations are presented in Sections 201 through 256 for applicable systems, structures and components.

a.

Caution must be used when working in areas which contain systems or structures that support operation of the YNPS ISFSI. The systems and structures cannot be affected by removal activities. Work packages must include specific steps either to physically protect the systems and structures or to establish safe load paths and protective zones around the systems and structures.

b.

Hazardous materials and wastes must be processed in accordance with the YNPS Hazardous Waste Management Program or equivalent YAEC approved contractor program (Section 511), including the following:

Asbestos containing material (e.g., insulation) must be removed and processed in accordance with the YNPS Asbestos Control Program or equivalent YAEC approved contractor program (Section 511.4.6). This activity should be scheduled prior to initiating equipment removal in an area.

Structures and components containing lead based coatings must be processed in accordance with the guidance presented in Section 511.4.7.

Paint coatings containing solid Polychlorinated Biphenyls (PCBs) must be managed in accordance with all federal and state regulations.

c.

Decommissioning activities that use liquids must ensure that contaminated liquids will be processed in accordance with the YNPS ODCM and NPDES discharge permit.

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YNPS FSAR REV. 6/05

d.

The following considerations must be incorporated into tank and vessel sludge removal activities:

The method used must ensure that any liquid inadvertently discharged from the system is processed in accordance with the YNPS ODCM and NPDES discharge permit.

Sludge removed from the system must be stabilized prior to shipment.

Waste water must be processed and analyzed before discharging from the facility.

The use of a high pressure water rinse should be used before dismantlement, if necessary, to reduce internal contamination levels.

e.

Radioactive particulate emissions will be monitored to the maximum extent practicable. The following must be implemented in accordance with the requirements of the Off-Site Dose Calculation Manual (ODCM):

Local HEPA filtration systems must be used when activities could result in the release of significant radioactive particulates or hazardous materials. Monitoring of the HEPA Filtration System exhaust will be performed when required by the ODCM. Prior to initial deployment and periodically, each HEPA System is tested against Dioctyl Phthalate (DOP) with an acceptance criterion of 99.95% filter retention. This ensures that airborne particulate releases to the environment have been reduced to as low as is reasonably achievable.

f.

Electrical and pneumatic services must be isolated from the systems, components and structures prior to dismantlement.

The following should be implemented:

Pumps, fans, heaters, motor operated valves, motor operated dampers and instrumentation power sources should be isolated and disconnected from station electrical and control systems at the motor control centers, supply breakers, fuse blocks, and at the equipment.

Pneumatically operated components and instrumentation should be isolated from the Compressed Air System at the root and equipment isolation valves.

g.

Openings in components must be enclosed with a protective cover to confine internal contamination.

h.

Explosive methods will not be used during the YNPS decommissioning without completion of a separate safety review. The analysis must include the effects of the use of explosives and assess the impact on the YNPS ISFSI and associated support systems.

i.

Before removing contaminated systems, structures and components with significant external contamination, they should be wiped down to remove external contamination or painted with a coating to stabilize external contamination.

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YNPS FSAR REV. 6105

j.

Contaminated piping and tubing should be dismantled as follows:

Large bore piping (greater than 2-1/2 inch diameter) should be cut into manageable lengths.

Small bore piping (less than 2-1/2 inch diameter) and tubing should be cut using an appropriate method based on the radiological conditions.

Remote cutting systems should be used as appropriate to maintain worker exposure as low as is reasonably achievable.

Underground piping must be visually examined after it is excavated to ensure that it is physically sound prior to cutting and removal.

k.

Contaminated supports should be removed in conjunction with equipment removal activities.

1.

Systems and components should be removed from areas and buildings prior to the commencement of area structural decontamination activities.

m.

Centralized processing and cutting stations should be considered to facilitate packaging of components for shipment to an off-site processing facility or a low level radioactive waste disposal facility.

200.3.4 Decontamination and Dismantlement Process The decontamination and dismantlement of contaminated systems, structures and components may be accomplished by either decontamination in place, removal and decontamination, or removal and disposal. A combination of these methods may be utilized to reduce contamination levels and reduce worker radiation exposures.

In general, contaminated and potentially contaminated systems, structures and components will be removed as follows:

Decommissioning work packages with sufficient detail will be developed, reviewed and approved in accordance with project and plant programs and procedures (Section 200.3.2)

Plant tag-out procedures will be used to de-energize electrical and control equipment, isolate and drain fluid systems, and isolate and depressurize pneumatic systems. Radiation Protection procedures will be used to ensure radiological requirements for control of contamination, worker protection and ALARA program are satisfied. Occupational Safety standards will be observed.

Contaminated concrete and structural steel components with contamination above accepted site criteria will be decontaminated or removed when all contaminated and uncontaminated systems and equipment have been removed from the area or building. This contaminated concrete will then be removed and packaged into containers for shipment to a low level radioactive waste disposal facility. Contaminated structural steel components may be moved to a processing area for volume reduction and packaging into containers for shipment to a processing facility for decontamination or low level radioactive waste disposal facility.

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YNPS FSAR REV. 6/05

Buried contaminated components (e.g., piping, drains, conduit) with contamination levels above accepted site criteria will be excavated. After excavation, the components will be examined to ensure that they are physically sound prior to cutting and removal. Most buried contaminated piping is located in steel conduits (i.e., pipes enclosed in pipes). Contamination controls will be modified as necessary if the components are significantly degraded.

A summary of currently available methods for decontamination of plant equipment and structures is presented in Table 200.2. The methods presented in this table are the most practicable and widely utilized at the time that the Decommissioning Plan was generated. However, new decontamination technologies will be adopted, if appropriate.

A summary of currently available methods for cutting plant equipment and structures is presented in Tables 200.3 and 200.4. The methods presented in this table are the most practicable and widely utilized at the time that the Decommissioning Plan was generated. However, new dismantlement technologies will be adopted, if appropriate.

200.4 Deleted 200.5 Decontamination and Dismantlement Plan: Systems, Structures and Components 200.5.1 Overview As required, Sections 201 through 256 present an initial description of the decontamination and dismantlement activities that are necessary to decommission each potentially contaminated system, structure and component at YNPS. Detailed planning includes engineering design; ALARA planning; and cost, schedule, and resource refinement.

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YNPS FSAR REV. 6105 I

TABLE 200.2 Decontamination Methods Method Advantages Disadvantages Carbon Dioxide Blasting Low waste volume High operating costs HEPA filtration may be required to control airborne contamination High setup costs Abrasive Blasting Very effective for surface contamination Large waste volumes HEPA filtration may be required to control airborne contamination Hydro Blasting Remote operation possible Large waste volumes Very effective for surface contamination HEPA filtration may be required to Easy to use control airborne contamination Strippable coatings Easy to use Only effective on loose Good for fixing loose surface contamination contamination Scarifying Effective on coated and uncoated surfaces Cannot cut rebar Removes concrete to 1/4 inch deep Vacuum with HEPA may be used to control waste and dust Scabbling Removes concrete to 1/ inch deep per pass Cannot cut rebar Easy to control removal depth per pass Vacuum with HEPA may be used to control waste and dust Not as effective on coated surfaces Spalling Low airborne contamination Slow process Good for limited access and small areas Cannot cut rebar Vacuum Cleaning Easy to use Only effective on loose Fast removal times contamination HEPA filtration may be required to control airborne contamination 200-7

YNPS FSAR REV. 6105 TABLE 200.3 Mechanical Cutting/Removal Methods Method Applicability Advantages Disadvantages Machining Contaminated pipe Minimum airborne Unit is heavy in larger sizes cutting contaminatation Vessel segmentation Easy setup Remote operate Quick cutting times Abrasive Water-Tank segmentation Remote operation possible Large amounts of liquid waste Jet Concrete cutting HEPA filtration may be required to control airborne contamination Slow cutting times Abrasive Wheel Steel cutting Inexpensive HEPA filtration may be Easy to setup and operate required to control airborne Can cut rebar and imbedded contamination steel Slow cutting times Diamond Wire Concrete cutting Capable of cutting thick Slow cutting times concrete Large amounts of liquid waste Can cut rebar and imbedded steel Remote operation possible Metal Disintegration Machining (MDM)

Vessel segmentation 0

0 0

Capable of cutting thick steel High precision obtainable Remote operation possible S

S Slow cutting times Complex control system required Mechanical Tubing, cable and Quick cutting times Cannot be used on large times Shears pipe cutting Best for tubing and sheet metal Band and Contaminated pipe Easy to use Slow cutting times Reciprocating and steel cutting Cannot be used on large items Saws Impact Hammer, Concrete removal Inexpensive HEPA filtration may be Wrecking Ball, Fast removal rates required to control airborne Concrete contamination Crushing Method Creates rubble 200-8

YNPS FSAR REV. 6105 TABLE 200.4 Thermal Cutting/Removal Methods Method Applicability Advantages Disadvantages Plasma Arc Tank segmentation Fast cutting times HEPA filtration may be Uncontaminated pipe Remote operation possible required to control airborne cutting contamination from Structural steel cutting contaminated component Contaminated piping removal (with appropriate Large power supply required radiological controls)

Oxy-Fuel Tank segmentation Easy to use HEPA filtration may be Uncontaminated pipe Fast cutting times required to control airborne cutting Remove operation possible contamination from Structural steel cutting contaminated component Contaminated piping removal (with appropriate Ineffective on stainless steel radiological controls)

Uses inflammable gases Electric Vessel segmentation Capable of cutting thick steel Slow cutting times Discharge High precision obtainable Complex control system Machining Remote operate possible required (EDM)

Flame Cutting Concrete cutting Capable of cutting thick HEPA filtration may be concrete structures required to control airborne Low waste volumes contamination Difficult to use Arc Saw None at this time Capable of cutting thick steel HEPA filtration maybe Remote operation possible required to control airborne contamination Large power supply required Heavy weight of unit Limited experience with unit 200-9

YNPS FSAR REV. 6/05 229 ELECTRICAL SYSTEM 229.1 Description The on-site electrical system is powered by the 13.8 kV Massachusetts Electric Line. The system consists of a transformer, switchboard, motor control centers, distribution panels and the necessary associated instrumentation and controls (Drawing Numbers 9699-FE-lJA, -iP, -1Q).

The 13.8 kV Massachusetts Electric Line also powers the Furlon House, Training Center and Trash Compactor. The electrical system at YNPS provides power to equipment which must remain energized during the final phase of decommissioning. All on-site electrical equipment is powered from a 480 V AC switchboard via a 13.8 kV/480 V, 100 kVA transformer. The 1600 amp Secondary Side Switchboard powers equipment located on the Secondary side of the site including the Gate House and ISFSI. Back-up power for portions of the site electrical system is provided by a 175 kW Security Diesel Generator.

229.2 Decontamination and Dismantlement Considerations General decontamination and dismantlement considerations are presented in Section 200.3.

229-1

YNPSFSAR REV. 6/05 232 FIRE DETECTION SYSTEM 232.1 Description The Fire Detection System provides the equipment needed to detect fires that could occur in structures supporting the ISFSI. The system consists of detectors, and the necessary associated instrumentation.

The Fire Detection system provides coverage in the Gatehouse and within the Instrument Enclosure Building at the ISFSI. All other portions of the Fire Protection and Detection System have been removed from service and/or disposed of. Fire Protection Program requirements are described in Section 513.

232.2 Decontamination and Dismantlement Considerations General decontamiantion and dismantlement considerations are presented in Section 200.3. The following is a specific system consideration:

Modifications to the Fire Detection System require review and modifications, as necessary, to the YNPS Fire Protection Plan (Section 513).

232-1

YNPS FSAR REV. 6/05 255 MISCELLANEOUS TANKS 255.1 Description This section presents descriptions of plant tanks. Some of these tanks are potentially contaminated, or are needed to support decommissioning activities and/or the ISFSI.

Fire Water Storage Tank - The Fire Water Storage Tank (TK-55) provided a back-up source of water to the Fire Protection System. The tank is constructed of steel. The tank has been removed from service and is awaiting dismantlement.

Fuel Oil Storage Tanks - Description of above-ground fuel oil storage tanks is presented in Section 511.4.1.

255-1

I YNPS FSAR REV. 6/05 256 METEOROLOGICAL TOWER 256.1 Description A Meteorological Tower installed at the ISFSI pad provides real time capability to determine wind speed and direction for on-site emergency planning purposes.

256-1

YNPS FSAR REV. 6105 257 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) 257.1 Description The Independent Spent Fuel Storage Installation (ISFSI) provides dry cask storage of spent nuclear fuel and Greater-Than-Class-C (GTCC) waste generated at the YNPS. Dry cask storage of spent nuclear fuel is conducted under the general license provisions of 1 OCFR72, Subpart K and 1 OCFR30. The NAC International, Inc., Multi-Purpose Canister (MPC) system (Docket No. 72-1025), which received a certificate of compliance in April 2000 (Reference 257-1), is used at the ISFSI.

The ISFSI is designed to store a combination of up to 18 NAC-MPC dry fuel or GTCC waste storage casks. It is located in the southwest corner of the site (Figure 257-1) and consists of a concrete pad, access road, security fences, temperature monitoring instrumentation, lighting, security, ISFSI Instrumentation Enclosure.

The ISFSI facility has 16 Vertical Concrete Casks (VCCs). Each VCC contains a Transportable Storage Canister (TSC). The TSCs contain either spent fuel or GTCC waste.

The VCC is the storage overpack for the TSC and provides structural support, shielding, protection from environmental conditions, and natural convection cooling of the TSC during storage. The VCC is a reinforced concrete structure with a carbon steel inner liner. Each VCC is approximately 160 inches in height, has an outside diameter of approximately 128 inches and weighs approximately 155,000 lbs. Each VCC has 4 inlet vents evenly spaced around the bottom of the VCC. There are also 4 outlet vents spaced evenly around the top of the VCC, offset by 45 degrees in order to provide a non-planar path for the natural circulation of air around the TSC and to minimize radiation streaming.

The TSC is a stainless steel circular cylindrical shell with a welded bottom plate. It has an outside diameter of approximately 71 inches and is approximately 123 inches in height. The TSC can accommodate up to 36 Yankee Class spent fuel assemblies. The TSC may also contain one or more Reconfigured Fuel Assemblies (RFAs), which are designed to hold Yankee Class spent fuel rods as intact or damaged fuel or fuel debris. An RFA can accept up to an equivalent of 64 full length spent fuel rods in an 8 by 8 array. The stainless steel RFA container has the same external dimensions of a standard fuel assembly. The weight of a fully loaded TSC is approximately 55,000 lbs.

The VCC Temperature Monitoring System is designed to monitor the thermal performance of each concrete cask. The system is comprised of a network of thermocouples at the outlet vents of all the casks and at four different points for ambient temperature that feed into a panel within the ISFSI Instrument Enclosure where temperature readings can be obtained.

Radiation monitoring using Thermoluminesent Dosimeters (TLDs) is in place to demonstrate compliance with 1 OCFR72.104 and 40CFRI 90.

The ISFSI pad is within a protected area located to the south of the general site area of the YNPS.

The ISFSI pad is a reinforced structural concrete pad that was designed to store and support the loaded VCCs. The pad is approximately 48 feet wide by 160 feet long. The east end of the pad is 257-1

YNPS FSAR REV. 6/05 elevated approximately 30 inches above the grade of the access road to facilitate the loading/removal of the VCCs from the heavy-haul trailer. The VCCs are moved by using air pads and a specially rigged fork lift vehicle (cask primary mover).

The ISFSI Enclosure Building is a free standing building, measuring approximately 12 feet wide by 15 feet long by 10 feet in height, located within the protected area to the west of the ISFSI pad. The ISFSI Enclosure Building houses the temperature monitoring panel, I/O racks, transformers, Uninterruptible Power Supplies (UPS), communications, fire detection and other ISFSI and security related equipment.

The ISFSI protected area is fenced, monitored by security, and has alarms to prevent unauthorized access. Access to the ISFSI area is controlled by the YNPS Security Department.

257.2 Status The ISFSI construction is complete. Spent nuclear fuel and GTCC waste transfer is complete. The YNPS ISFSI contains 15 NAC-MPC Spent Fuel Storage Casks in addition to 1 NAC-MPC Cask that contains GTCC waste.

REFERENCES 257-1 Letter, NRC to NAC International, Inc., dated March 17, 2000, Certificate of Compliance for the NAC International, Inc., Multi-Purpose Canister (MPC) System (TAC No. L22907).

257-2

YNPS FSAR REV. 6/05 El

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YNPS FSAR REV. 6/05 300 ENVIRONMENTAL - SITE CHARACTERISTICS 300.1 Demography YNPS is located in the Berkshire Hills of Franklin County in Rowe, Massachusetts. The site is at the bottom of a deep valley along the Deerfield River on the southeast bank of Sherman Reservoir.

The area surrounding the site is mostly wooded with very steep gradients on both sides of the Deerfield River.

The YNPS site boundary and plant exclusion area is shown in Figure 300-1.

300.2 Geography and Land Use The population density in the rural area surrounding the YNPS site is low. Within one mile of the site, for example, the population is 48 (based on 1989 Massachusetts municipal census counts). The populations of the two closest towns, Rowe and Monroe, are 354 and 141, respectively. The nearest population center of 25,000 or more is Pittsfield, Massachusetts, located about 21 miles south-west of the site. The regional population is expected to remain virtually unchanged over the next decade.

Land use near the site is made up of a few farms and some commercial businesses. The centers of Rowe and Monroe have small, mixed clusters of local businesses, municipal buildings, and residences, with homes scattered throughout the area. There is no large industrial activity within five miles of the site; the only industry in the area is the YNPS and several hydroelectric facilities along the Deerfield River.

The nearest highway and railroad are each about five miles south of the site. The closest airport is in North Adams, Massachusetts, about ten miles west of the site.

300-1

YNPS FSAR REV. 6/05 FIGURE 300-1 Site Boundary and Plant Exclusion Area 300-2

YNPS FSAR RE V. 6/05 301 METEOROLOGY (GENERAL CLIMATE, SEVERE WEATHER)

The site lies in the prevailing westerlies, the belt of eastward moving air that is found in middle latitudes. Many storms pass over Massachusetts compared to other parts of the United States. This is a result of extensive air masses originating at higher and lower latitudes. The three major air mass types that affect the site are cold, dry, subartic air from Canada; warm, moist air from the Gulf of Mexico; and cool, damp air from the North Atlantic Ocean.

The hills on either side of the site rise about 1 000 feet above ground level within one mile and extend from 12 miles north to 8 miles south southeast of the site. This feature affects the winds.

There is, for instance, a high frequency of occurrence of "channel flow" up and down the Deerfield River valley. Also, night -time drainage flow down the east side of the river valley occurs frequently.

Normal daily temperatures range from 1 00F in January to 80'F in July. Recorded extreme temperatures for the site region are -250F and 980F. Thunderstorms occur about 28 days per year; the annual flash density of ground lightning strikes is four flashes per square kilometer.

The site design wind speed (defined as the "fastest-mile" wind speed at 30 feet above the ground with a 100 year return period) is 80 mph. Hail storms occur about two days annually and freezing rain about 12 days per year. The maximum radial ice thickness expected for the region is 1.25 inches. Mean annual snowfall at the site totals 100 inches, with maximum snow depth on the ground totaling about 40 inches.

301-1

YNPS FSAR REV. 6/05 302 HYDROLOGY The site is located on the east bank of the Deerfield River adjacent to Sherman Reservoir, which serves as the source of the cooling water for the plant. The drainage area upstream of the plant is characterized by a dendritic pattern, is 236 square miles, and has an average annual rainfall of between 40 and 50 inches.

The Deerfield River flow is highly regulated by two large, upstream hydroelectric reservoirs. The average long-term flow near the plant is 738 ft /sec.

Bedrock in the region is not a significant source of groundwater nor are there major bedrock aquifers within the site area. The direction of groundwater flow under the site is from the recharge areas on the slopes surrounding the plant toward the Deerfield River.

302-1

YNPS FSAR REV. 6/05 303 GEOLOGY AND SEISMOLOGY The regional bedrock geology is complex; bedrock age ranges from 100 million to over one billion years old and is comprised of mostly a mosaic of metamorphic and igneous rocks. The youngest deposits are glacial soils, 10 to 12 thousand years old. Most volcanic and sedimentary rock in the area are now metamorphosed.

The site is situated on dense Wisconsinian-aged glacial till. YNPS structures are founded on this till, which ranges from 0 to 140 feet thick across the site. Bedrock under the till is part of the lower Cambrian Hoosac formation and consists of quartz-albite-biotite gneiss and a rusty gneiss in adjacent areas. Underlying these are garnet schist and a layered gneiss with some dolomitic marble, with the latter units belonging to the lower Cambrian or older Cavendish formation. A south-plunging anti-cline, whose axis is just east of the site, defines local bedrock structure.

Site bedrock is hard, internally welded metamorphic rock, not subject to significant deterioration.

Bedrock fracturing is not a prominent structural feature of this bedrock; outcrops exhibit either no joints or minor, discontinuous joint surfaces. Fracture pattern analysis of site vicinity joints, joint sets, and faults show no anomalous trends for fractures. This suggests the absence of any through-going zones of post-metamorphic faulting or shear.

The site is in the Western New England Fold Belt province. It borders the Adirondack Uplift province to the west, the Valley and Ridge province to the southwest, and the New York Recess to the south. The Southeast New England Platform and Merrimack Synclinorium occur to the east and northeast, respectively.

Regional seismic events are very infrequent and do not cause surface faulting. Only two events in the province were greater than Intensity V (MM). These events were 125 miles and 210 miles from the site. The site seismic design level for new seismically qualified installations at the plant is a peak ground acceleration of 0.19 g. The return period is between 10,000 and 100,000 years.

303-1

1 YNPS FSAR REV. 6/05 304 PLANT WATER SUPPLY I Potable water is supplied by an on site bedrock well.

304-1

YNPS FSAR REV. 6/05 305 PLANT EFFLUENT The liquid effluent release path used during the final stages of decommissioning will be via the storm drain system.

All liquid discharges will be controlled in accordance with the National Pollutant Discharge Elimination System (NPDES) permit and the ODCM.

Portable water treatment equipment is used in accordance with the ODCM and NPDES permit in conjunction with temporary construction dewatering storage tanks.

Surface water use downstream from the site is mostly for recreation and sport fishing, with limited irrigation. Water supply for the municipalities within five miles downstream of the plant is from private wells. The closest public water supplies are well fields 20 to 25 miles downstream of the plant.

305-1

YNPS FSAR REV. 6105 306 ENVIRONMENTAL RADIOLOGICAL STATUS 306.1 Facility Operating History Yankee Nuclear Power Station (YNPS) achieved initial criticality in 1960 and began commercial operation in 1961. On October 1, 1991, YNPS shut down after operating 31 years. On February 26, 1992, the YAEC Board of Directors decided to cease power operation permanently. YNPS operated with an average lifetime capacity factor of about 74%.

There were occurrences, during the operation of YNPS, that resulted in contamination of structures and components inside buildings located in the Radiation Control Area. In addition, several occurrences resulted in the contamination of the grounds outside of the buildings but inside the Radiation Control Area. Most of these events were minor, resulting in minimal contamination.

Following detection of an event, actions were taken to remove and control contamination and to institute corrective actions to preclude future occurrences.

Table 306.1 presents a summary of significant radiological contamination occurrences at YNPS.

The primary sources used to compile and review events were: Control Room Logs, Abnormal Occurrence Reports, Licensee Event Reports, Plant Information Reports and reports to the Atomic Energy Commission, Nuclear Regulatory Commission and the Commonwealth of Massachusetts.

Interviews were also conducted with present employees, recently discharged employees and retired long-term employees. The purpose of the interviews was to gain additional information concerning plant occurrences and operations that may have resulted in residual contamination.

The most significant contamination event at YNPS was leakage from the Ion Exchange Pit.

Significant leakage was first identified in May 1964. An unsuccessful attempt to isolate the leak was made in July 1964 by installing and sealing a concrete plug in the pit sump. Following that attempt, a decision was made to empty the resin capsules and drain the pit. This activity was completed in April 1965. In May 1965, a crack in a vertical joint at the northwest corner of the Ion Exchange Pit was found after draining the pit. The crack was repaired and the Ion Exchange Pit floor and walls were sealed to prevent further leakage.

In 1965, tritium was detected in Sherman Spring. The presence of tritium was attributed to migration of tritium from the Ion Exchange Pit into the groundwater. At the present time, the tritium concentration is significantly below the Environmental Protection Agency community water system limit (Section 306.3 and Reference 306-1 1). In addition, the water from Sherman Spring, as well as the Deerfield River, into which it flows, is not used for human consumption.

Systems in contact with the main coolant have been contaminated with activated corrosion products and fuel residue. In 1977, YNPS began converting from stainless steel clad fuel to zircaloy clad fuel. Following conversion, fuel clad failures began occurring. Most of the failures were minor failures of clad integrity, releasing iodines and other fission gasses into the main coolant. However, after Cycle 13, a number of peripheral fuel rods failed as a result of damage caused by water jetting from the core baffle spacer plates. These fuel failures were sufficient to cause contamination of the Reactor Vessel and Main Coolant System with fuel residue. Baffle jetting damage was eliminated by the addition of spacer plugs at the bottom of the core baffle spacer plates and modifying fuel assembly design. No significant fuel failures occurred after the modifications were implemented.

306-1

YNPS FSAR REV. 6105 Primary system integrity was very good during plant operations. Minor leakage from Main Coolant System flanges and valve stems to the Vapor Container occurred during plant operations. However, leakage rates were maintained below Technical Specification limits. The Vapor Container was decontaminated during refueling outages to remove contamination that may have accumulated during previous operating cycles. The YNPS steam generators performed well during the 31 year operating life. There was no significant leakage between the primary and secondary sides of the generators.

The results of the historical review were incorporated into the design and implementation of the radiological scoping survey and sampling program.

306.2 Deleted 306.3 Radiological Environmental Monitoring Program The Yankee Decommissioning Quality Assurance Program establishes the Radiological Environmental Monitoring Program. The purpose of the program is to monitor the radiation and radionuclides in the environs of the plant. The program includes the following elements:

Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters presented in the YNPS Off-Site Dose Calculation Manual.

Performance of a land use census to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the monitoring program are made if required.

Participation in an interlaboratory comparison program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the Quality Assurance Program for environmental monitoring.

The Yankee Decommissioning Quality Assurance Program (YDQAP) requires the submittal of an annual report to the NRC. The report shall include summaries, interpretations and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period.

The Radiological Environmental Monitoring Program was initiated in 1958, about two years before YNPS began commercial operation. The program has been executed continuously since its inception. In recent years, the program has incorporated continuous monitoring of air and automatic composite sampling of river water, as well as sampling of groundwater, river sediment, fish, cow milk, goat milk, vegetables, and maple syrup. Direct radiation exposure measurements, both in the immediate plant vicinity and at more distant locations, have been made continuously through the use of a network of thermoluminescent dosimeters (TLDs).

The Radiological Environmental Monitoring Program will continue to be updated and modified to reflect changes to the potential source term, the plant environs, and the surrounding land use during YNPS decommissioning.

306-2

YNPS FSAR REV. 6/05 REFERENCES 306-1 NUREG/CR-5849 (Draft), "Manual For Conducting Radiological Surveys in Support of License Termination," June 1992.

306-2 YRC-1024, "Basis for the Radiological Status of Plant Systems and Structures,"

P. Hollenbeck, September 1993.

306-3 REG 90/93, "Dose Conversion Factors For Yankee Rowe Piping," Y. J. Yu to R. A. Mellor, April 2,1993.

306-4 YRC-1031, "Yankee Rowe Component Activation Analysis," K. J. Morrissey, September 1993.

306-5 NUREG/CR-3474, "Long-Lived Activation Products in Reactor Materials," August 1984.

306-6 REG 147/93, "Dose Rate Estimate of the YNPS Reactor Vessel," Y. J. Yu to R. A. Mellor, June 4, 1993.

306-7 "YNPS Outdoor Site Scoping Data," F. X. Bellini, E. R. Cumming to P. Hollenbeck, September 24, 1993.

306-8 YNPS Off-Site Dose Calculation Manual.

306-9 NUREG/CR-5512, Vol. 1, "Residual Radioactive Contamination from Decommissioning

- Technical Basis for Translating Contamination Levels to Annual Total Effective Dose Equivalent," Final Report, October 1992.

306-10 1992 Annual Radiological Environmental Operating Report.

306-11 BYR 94-047, RAI on YNPS Decommissioning Plan and Decommissioning Environmental Report.

306-12 NUREG-1575, "Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM)," December 1997.

306-13 Letter, D. K. Davis (YAEC) to NRC, "Withdrawal of Proposed License Amendment to Approve Yankee Nuclear Power Station's License Termination Plan," May 25, 1999.

306-3

YNPS FSAR REV. 6/05 TABLE 306.1 Summary of Significant Radiological Contamination Events Date Description of Occurrence Radiological Consequences 9/20/61 A main coolant sample container was About 35 [iCi of radioactivity was released Incident dropped between the Primary Auxiliary as a result of the spill on the asphalt Report Building and the Waste Disposal Building.

surface. Fixed contamination remaining No. 61-15 The contents of the container spilled onto following decontamination resulted in a the asphalt surface.

contact dose of about 0.05 mR/hr.

9/18/63 A sampling valve located over the Ion Radiation levels of the spill area were Incident Exchange Pit was inadvertently left open 70-100 mR/hr measured 1 inch from the Report when filling the Shield Tank Cavity from surface. Ion Exchange Pit cover No. 63-12 the Safety Injection Tank. Water from the contamination levels were 1,000,000-tank spilled onto the cover of the Ion 10,000,000 dpm measured over several Exchange Pit. A portion of the water square inches. Asphalt contamination drained from the cover onto the asphalt on levels were 20,000-60,000 dpm/ft2. Both the west side of the Ion Exchange Pit.

the Ion Exchange Pit cover and the asphalt were decontaminated.

10/8/63 Several small holes were detected on the Water with a radioactivity level of about Incident bottom of a leakage collection drum located 6E-5 gCi/ml leaked from the drum into the Report in a storm drain catch basin outside of the catch basin. The basin was pumped to the No. 63-17 Spent Fuel Building. The drum was used to Waste Disposal System and flushed with collect leakage from the Fuel Transfer Service Water.

Chute Dewatering Pump.

306-4

YNPS FSAR REV. 6105 Date Description of Occurrence Radiological Consequences 9/3/64 A spill of main coolant occurred while About 270 ptCi of radioactivity was Incident filling and pressurizing the shutdown released as a result of the spill. Several Report No.

cooling pump gland seal water tank. Main samples were taken:

64-8 coolant flowed out the vent connection into a relief valve discharge header and onto the Seal tank drain 2E-3 pCi/ml Primary Auxiliary Building roof.

PAB roof 1E-3 pCi/ml Storm drainl E-6 pCi/ml Predominant isotopes were Co-58, Co-60 and Mn-54. The area was decontaminated.

Several years after the event, a new roof was installed and the old roofing material was disposed of as radioactive material.

10/3/64 A valve was inadvertently left open after Leakage from the Ion Exchange Pit AOR adding water to the Ion Exchange Pit.

contaminated the ground and asphalt on the No. 64-13 Water continued to flow into the Ion west side of the Ion Exchange Pit. The Exchange Pit from the Primary Water water had a specific activity of 8E-8 Storage Tank. Water leaked from the Ion pCi/ml and was below the maximum Exchange Pit into the ground resulting in permissible concentration. The area was water coming up through the asphalt on the flushed with service water to dilute the west side of the pit.

contamination as it entered the storm sewer.

2/17/65 A routine chemistry sample of Ion Analysis indicated that the activity was AOR Exchange Pit water indicated an activity comprised primarily of F-18 (half life = 1.8 No. 65-6 level of 1.2E-5 pCi/ml. This was about 100 hr) and an unknown isotope with a gamma times above the normal level. The cause energy of 0.14 MeV (half life = 40 hr).

was identified as a leaking outlet connection Based on an existing leak from the Ion on a cation bed that was in main coolant Exchange Pit to the soil, about 3,400 pCi in service. The activity decreased by about 74,000 gallons of water was released to the 50% within a day after isolating the leak.

area beneath the Ion Exchange Pit over the The connection was installed 19 days before 19-day period.

the leak was identified.

306-5

I YNPS FSAR REV. 6105 Date Description of Occurrence Radiological Consequences 3/25/65 A sample of Ion Exchange Pit water Main coolant leaked into the Ion Exchange AOR indicated a high activity level, 2E-4 piCi/ml.

Pit, increasing the radioactivity level. After No. 65-8 The cause was identified as a leak from an the leak was isolated, a feed and bleed anion exchange capsule and filter that were dilution of the Ion Exchange Pit was installed the previous day. Main Coolant established. About 88,000 gallons of System purification flow was stopped and demineralized water was added to the Ion all capsules were isolated.

Exchange Pit.

5/64 - 5/65 Significant Ion Exchange Pit leakage was Water from the Ion Exchange Pit leaked Operation identified in May 1964. An unsuccessful into the soil below the pit during the period Reports attempt to isolate the leak was made in July between May 1964 and April 1965. The Nos. 42-53 1964 by installing and sealing a concrete average activity during this period was plug in the pit sump. Following that 7E-6 gCi/ml resulting in a total release of attempt, a decision was made to empty the about 36,000 ptCi of identifiable activity.

resin capsules and to drain the pit. This Greater than 95% of this activity was due activity wvas completed in April 1965. In to Cr-51, Mn-54, Fe-59, Co-58, Co-60, May 1965, a crack in a vertical joint at the Ag-11 Om, and Hf-181. Based on the northwest corner of the Ion Exchange Pit average activity and the lowest maximum was found after draining the pit. The crack permissible concentration, the release from was repaired and the Ion Exchange Pit floor the Ion Exchange Pit was less than 12% of and walls were sealed to prevent further the maximum permissible concentration for leakage. In 1965, tritium was detected in unrestricted areas. Based on soil Sherman Spring. The presence of tritium characteristics it is unlikely that this was attributed to migration of tritium from activity migrated to an unrestricted area.

the Ion Exchange Pit into the groundwater.

The amount of tritium attributed to this Section 3.1.5 presents additional leakage has been estimated to be less than information regarding the activity in 200 Ci (Reference 306-1 1).

Sherman Spring.

4/8/66 Main coolant was inadvertently aspirated Main coolant was spilled with a AOR into the Primary Vent Stack fan from the radioactivity level of 2.26 pCi/ml tritium No. 66-3 Low Pressure Vent Header during outage and 3.4E-2 piCi/ml gross beta-gamma.

related Reactor Vessel venting operations.

Smears of the affected area indicated Main coolant leaked from the riveted joints 100,000-1,000,000 dpm/ft2. A survey of in the fan duct work onto the Primary the vent line and the duct work indicated Auxiliary Building Mechanical Equipment contact radiation levels of 10-70 mR/hr on Room floor.

horizontal runs and 10-25 mR/hr on vertical runs. The area was decontaminated.

306-6

I YNPS FSAR REV. 6/05 Date Dcscription of Occurrence Radiological Consequences 9/27/66 The Spent Fuel Pit overflowed when a A sample of the water that overflowed AOR Primary Water Storage Tank valve was from the Spent Fuel Pit indicated 3.2E-5 No. 66-7 inadvertently left open. Most of the water pLCi/ml gross activity and 5.4E-3 plCi/ml flowed from the building, down the north tritium. This occurrence resulted in a total exterior wall, over a small section of asphalt release of 4 PCi gross beta-gamma activity paving and into a storm drain. A few and 670 pCi of tritium. Affected areas gallons leaked into the New Fuel Vault.

were decontaminated, the storm drain was flushed with about 75,000 gallons of service water.

9/27/66 A sample of the west side storm drain Samples collected from the west side AOR culverts indicated elevated activity levels.

culvert indicated 6.7E-7 pCi/ml average No. 66-8 This culvert was not affected by the Spent gross beta-gamma activity. The Safety Fuel Pit spill. The source of activity was Injection Tank water analysis indicated traced to a leak from the Safety Injection 3E-5 pCi/ml gross beta-gamma activity and Tank heating system safety valve located in l.lE-1 giCi/ml tritium. About 0.8,Ci the Primary Auxiliary Building. Floor gross beta-gamma activity and drains in this section of the building 3.32 millicuries of tritium were released discharge to a storm drain on the south side into the west culvert. Affected surface of the building which discharge into the areas were decontaminated.

west culvert.

11/1/66 A temporary hose failed during a routine The spilled liquid had an activity of 3E-3 AOR drainage operation on the Fuel Transfer pCi/ml, resulting in a release of about No. 66-9 Chute pump discharge line. The spilled 113 jiCi into the storm drain. The storm water drained into a storm drain served by drain was flushed with about 250 gallons of the east culvert.

water.

2/18/72 Moisture was detected under insulation on The leakage from the tank had a AOR the Test Tank level indicator. A leak radioactivity content of 0.0004 pCi No. 72-3 developed from the indicator to the ground beta-gamma and 2,017 pCi tritium. About at the base of the tank coincident with the 4 ft3 of crushed stone and soil at the base of discovery. The leak was contained within the tank was removed for disposal.

15 minutes of discovery.

Analysis of the soil following removal indicated no radioactivity above background levels.

306-7

I YNPS FSAR REV. 6/05 Date Description of Occurrence Radiological Consequences 7/16/75 While walking near the Ion Exchange Pit a Several sources of contamination were PIR technician received an alarm on a portable identified: 1) single specks of radioactive No. 75-7 radiation detection monitor. Upon material, 2) soil contaminated with investigation, an area of bare soil with relatively low activity to a depth of several radioactive contamination of about inches, and 3) pieces of contaminated 500,000 dpm was detected. Contamination materials (e.g., wood, concrete, appeared to be limited to several square feet polyethylene) on top of the ground. Low of soil with most of the activity limited to levels of radioactivity were also detected in an area of about 1 ft2.

the normal rain water runoff paths.

A complete surface survey within the plant Analyses of several soil samples indicated protected area fence was initiated using that the radionuclides consisted of nuclides RM-14 radiation detectors with HP-210 with half-lives greater than one year probes held approximately I inch off the (predominately Co-60). The contamination ground. Fourteen additional areas of most likely was deposited several years contamination were identified. Ten of the prior to detection.

fourteen areas were located on the nonradiological side of the plant.

Contaminated soil and debris were removed, including soil at the bottom of the storm drain catch basins. Any remaining soil contaminated with residual, low level radioactivity was sprayed with an asphalt sealer and covered with a thin layer of soil. Paved areas were swept and sealed with an asphalt sealer.

8/10/77 Routine analysis of the laboratory Contaminated water was detected in the PIR demineralized water supply indicated the Demineralized Water Storage Tank and the No. 77-10 presence of tritium. The cause was inferred Auxiliary Boiler Condensate Tank. The to be a valve line-up error that allowed main estimated activity released to the secondary coolant to flow into the demineralized water plant was 4,000 jiCi tritium and 75 tiCi header.

nonvolatile activity. The Demineralized Water Storage Tank was discharged as a permitted release and refilled with clean water. Normal boiler drum blowdown reduced the radioactivity to undetectable levels.

306-8

I YNPS FSAR REV. 6105 Date Description of Occurrence Radiological Consequences 12/21/77 A pipe from the Service Building sump to The contaminated water released the PIR the Gravity Drain Tank was severed during following radioactivity into the soil:

No. 77-16 core boring activities. The sump tanks receive water from the Radiation Control 1-131 16.50 pCi Area sinks, chemistry laboratory, and I-133 2.76 pCi machine shop drains. A mechanical Cs-134 0.34 pCi float-switch assembly starts and stops the Cs-137 0.50 pCi pump which transfers liquid from the sump Co-60 0.58 pCi tank to the gravity drain tank. Water was released from the system before the pump The soil at the location of the rupture was could be secured.

removed and disposed of.

10/24/79 Routine analysis of the auxiliary plant The radioactivity level of water in the PIR systems indicated the presence of tritium.

demineralized water header was very low:

No. 79-4 The cause was inferred to be valve leakage 5E 3E-4 pCi/ml tritium and 1E 1E-7 between the Test Tank transfer pumps and pCi/ml Cs-134, Cs-137, Co-60, Mn-54 the Primary Water Storage Tank. Water activity.

was transferred to the Auxiliary Boiler Condensate Tank, and the Demineralized Water Storage Tank during normal make-up operations.

8/6/80 Several gallons of contaminated water and Resin contact radiation levels were less PIR about one quart of resin were released than 1 mR/hr and spilled liquid readings No. 80-9 through a pinhole leak in a hose while were about 300,000 dpm/100 cm2. The transferring spent resin from the Ion spill area was decontaminated to remove Exchange Pit into a shipping cask.

contamination, including excavation of some asphalt.

5/15/81 The reactor vessel head was bumped against General contamination levels on the asphalt PIR the side of the equipment hatch during below the equipment hatch were No. 81-9 removal from the Vapor Container.

1,000-500,000 dpm/100 cm2. An area Contaminated material from the underside about 30 ft by 50 ft was contaminated with of the head was released and fell to the a total activity of about 250 pCi. About asphalt below the equipment hatch.

10 pCi was discharged to Sherman Pond when rain washed the radioactive material into the east storm drain before the area could be decontaminated.

306-9

I YNPS FSAR REV. 6/05 Date Description of Occurrence Radiological Consequences 4

4 9/10/84 PIR No. 84-16 A failed PVC drain line from the Waste Disposal Building was discovered during soil excavation activities. The drain line had six pipe joints, each of which leaked apparently due to failure of the solvent welds.

Soil below one of the pipe joints was significantly contaminated. Analysis indicated 50,000 dpm over the area of maximum contamination. One "hot spot" contained 29,300 pCi/g Co-60. Average Co-60 activity, measured 2 feet below the pipe joint was about 2,100 pCi/g. Average Cs-137 activity at this location was about 17 times less than the average Co-60 activity.

Soil under all pipe joints was removed to a depth of 5-7 feet below plant grade. About 420 cubic feet of soil and rock were removed and disposed of as radioactive waste. All areas above the excavation were sealed under a concrete cap.

12/14/91 Analysis of the laboratory demineralized About 840,uCi of tritium was leaked into PIR water supply, the Demineralized Water the demineralized water supply. Water was No. 91-7 Storage Tank, and the Auxiliary Boiler discharged from the Demineralized Water Condensate Tank indicated the presence of Storage Tank as a permitted release and the tritium and boron. No activity was found in demineralized water header was flushed.

the Primary Water Storage Tank. The root cause of the event could not be determined.

306-10

I YNPS FSAR REV. 6/05 307 FINAL RADIOLOGICAL SURVEY - SITE RELEASE CRITERIA This information has been extracted, updated and relocated to the PSDAR within the FSAR.

307-1

YNPS FSAR REV. 6/05 400 TRANSIENTS 400.1 General Overview This section presents an accident analysis that assesses the impact of decommissioning on both occupational and public health and safety. A structured, comprehensive process was used to identify and evaluate events that could occur during the period from approval of the Decommissioning Plan through completion of the final radiation surveys (Reference 400-1). The accident analysis considered: (1) decommissioning activity events, (2) loss of support system events, (3) fire events, (4) explosion events, (5) external events and (6) spent fuel storage events. Analysis of decommissioning events included all phases of decommissioning activities: decontamination, dismantlement, packaging, storage and radioactive materials handling.

The risk of accidents resulting in a significant radiological release during decommissioning activities is considerably less than that during plant operations. YAEC evaluated all of the Final Safety Analysis Report Section 400 safety analyses for applicability to a permanently defueled condition (Reference 400-2). With the completion of fuel transfer from the SFP to the YNPS ISFSI, the SFP fuel handling accident is no longer a credible event. The remaining events which could occur in a permanently defueled condition and impact the health and safety of the public are related to the release of airborne radioactive materials during decommissioning activities. Most of the demolition activities are now complete and not a concern as well. For historic purposes, the remaining events have been retained in the YNPS FSAR until all of the decommissioning activities are complete later in 2005. The retention of these events is conservative. After the decommissioning and demolition is complete, at the next update, a revision will remove the remaining accidents and transient conditions.

400.1.1 Radionuclide Release Limits Prior to the decision to permanently cease power operations, radiological releases resulting from design basis accidents postulated in evaluations and the Final Safety Analysis Report were evaluated using dose reference values from 1 OCFR Part 100. The I OCFR Part 100 reference values limit dose to an individual at the Exclusion Area Boundary during the first two hours following the onset of a postulated fission product release. The dose limits are less than 25 rem total dose to the whole body and less than 300 rem thyroid dose from radioactive iodine.

Since the decision to permanently cease power operations, YAEC requested and received an exemption from the emergency preparedness requirements of IOCFR50.54(q) (Reference 400-3).

This exemption allowed YAEC to discontinue off-site emergency response activities and to refocus the scope of on-site response capability. NRC approval of the exemption was predicated on the absence of accidents at a level of severity where the off-site dose could exceed the Environmental Protection Agency (EPA) Protective Action Guides (PAGs). Off-site protective actions are not warranted if the off-site dose following a postulated accident is less than the EPA PAGs.

400-1

YNPS FSAR REV. 6/05 The EPA PAGs are limiting values based on the sum of the effective dose equivalent resulting from exposure to external sources and from the committed effective dose equivalent incurred from all significant inhalation pathways during the early phase of an event (Reference 400-4):

EPA PAGs, rem Whole Body 1

Thyroid 5

Skin 50 Releases resulting from accidents postulated in the decommissioning accident analysis were evaluated using the EPA PAGs as an upper limit. This ensures that the current ISFSI Emergency Plan remains adequate for decommissioning and eliminates the need to reinstitute an off-site emergency response capability. Use of the EPA PAGs as an administrative limit also ensures that postulated accident off-site doses are significantly less than the 1 OCFR Part 100 reference values.

400.1.2 Assumptions The following assumptions have been incorporated into the accident analysis:

Special Nuclear Material used as reactor fuel will not be moved into the Reactor Vessel. This a condition of the YNPS possession only license.

The airborne pathway is the dominant radioactivity release pathway. Activities that could result in release of radioactive liquids will be designed to contain the releases within the liquid waste processing system using existing or supplemental barriers.

Airborne releases are assumed to occur at ground level with a conservative dispersion factor of 2.84E-04 sec/M3 (Reference 400-5).

Direct failures and consequences of initiating events were considered in the consequence analyses. Separate, coincident, random failures were not considered.

Decommissioning activities are independent from each other. There are no credible common cause mechanisms that could result in the simultaneous release of radioactivity from multiple activities that would exceed the equivalent release of the radioactive contents of the single, bounding container or component that results in the highest off-site dose. Interactions between systems during radioactive materials handling activities will be precluded by the maintenance of safe load paths; protective zones around limiting systems, structures and components; and single handling criteria for higher contamination items. The consequences of fire and explosions may impact several activities simultaneously. These events are considered separately in Sections 405 and 406.

Decommissioning activities may be performed in the Vapor Container without isolating the Vapor Container from the environment. The release of all of the surface contamination remaining in the Vapor Container will not exceed the doses predicted for the Material Handling Event.

400-2

YNPS FSAR RE, V. 6/05 400.2 Event Identification Process A structured, comprehensive process was used to identify accident initiating events which could lead to radionuclide releases during the YNPS decommissioning (Reference 400-1). The process included development of a logic diagram to evaluate all phases of decontamination, dismantlement and fuel management activities, as well as to identify nonradiological events.

Accident initiating events were grouped by structures, systems and components within a plant area.

The accident initiating events were compared using previous YNPS accident analyses, as well as current evaluations and calculations to identify the dominant accident initiating events within each plant area and then among plant areas. Accident scenarios were developed for these dominant accident initiating events. The scenarios formed the bases and inputs for radiological dose calculations to determine the impact on the health and safety of the public. Most of the systems and components that were used in the accident scenarios have now been removed from the plant. The event descriptions remain in the FSAR as bounding accidents for the ongoing decommissioning.

The following events were considered in the accident analysis and are presented below:

Events affecting occupational health and safety, including radiological and nonradiological events.

Off-site events affecting public health and safety.

Nonradiological events affecting public health and safety.

Radiological events affecting public health and safety, including the following:

Decommissioning activity events, including decontamination, dismantlement, packaging, storage and materials handling.

Loss of support system events, including loss of off-site power, cooling water and compressed air.

Fire and explosion events.

External events.

Spent fuel storage events, including fuel handling event, loss of spent fuel cooling capability, and interactions between spent fuel and decommissioning activities.

400.3 Summary The accident analysis assessed the impact of decommissioning on both occupational and public health and safety by considering decommissioning and fuel storage events. The evaluation of events that could affect occupational health and safety indicated that implementation of the Radiation Protection, Occupational Safety Programs and ISFSI Emergency Plan ensures that these events are sufficiently minimized. Analysis of events that could affect public health and safety indicated that there were no events that could significantly affect public health and safety.

400-3

YNPS FSAR REGV. 6/05 Table 400.1 provides dose conversion factors for several radionuclide mixes (Reference 400-8). The dose conversion factors are in units of rem per curie released to the atmosphere. They apply at the Exclusion Area Boundary and were determined as of January 1994. Table 400.2 provides a summary of the radioactivity content of several systems as of the same January, 1994 date.

Due to radioactive decay, the dose conversion factors for representative radionuclide mixes will change with time. A bounding set of dose conversion factors had been calculated that was applicable until the date 1/1/2005 (Reference 400-18). These were determined by decaying several radionuclide mixes until that date and using the most limiting result. The bounding set of dose conversion factors in Table 400.1 has been updated and is now applicable until the date 1/1/2008.

The update was based on applying the results of a conservative calculation which decay corrected the most limiting radionuclide mix to maintain a TEDE of 320 mrem at the Exclusion Area Boundary (Reference 400-25). The ratio of the previous to the currently calculated limiting radionuclide release limit (4.35 Ci / 3.84 Ci) was used to conservatively increase the boundary set of dose conversion factors in Table 400.1.

The analyses identified requirements that must be implemented during decommissioning to ensure that the accident analysis basis is maintained. Table 400.3 summarizes these items. The analyses also present planning considerations that reduce the probability of occurrence or consequences of the events that were evaluated. Table 400.4 also summarizes these items.

400-4

YNPS FSAR REV. 6/05 REFERENCES 400-1 YRP 437/93, Decommissioning Safety Analysis, September 28, 1993.

400-2 BYR 92-057, Request for Exemption From Annual and Biennial Emergency Preparedness Exercise in 1992, S. P. Schultz to M. B. Fairtile (USNRC), May 22, 1992.

400-3 NYR 92-178, Exemption From The Emergency Preparedness Rule 10CFR50.54(q) and Approval of The Defueled Emergency Plan at the Yankee Nuclear Power Station (TAC No. M83991), M. B. Fairtile (USNRC) to J. M. Grant, October 30, 1992.

400-4 EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents, US Environmental Protection Agency, October 1991.

400-5 YRC-1 82, Yankee Rowe Accident Atmospheric Diffusion Factors, January 5, 1982.

400-6 YRC-1024, Basis for the Radiological Status of Plant Systems and Structures, September 1993.

400-7 ELISA - Technical Description, A Computer Code for the Radiological Evaluation of Licensing and Severe Accidents at Light-Water Nuclear Power Plants, J. N. Hamawi, April 1991.

400-8 YRC-1014 Revision 1, Decommissioning Accident Source Terms.

400-9 Deleted.

400-10 ESG 90/93, Flammable Gas Explosion Evaluation For Decommissioning Safety Analysis, G. A. Harper to R. A. Mellor, August 20, 1993.

400-11 NUREG/CR-2300, PRA Procedures Guide, January 1983.

400-12 NYR 92-144, Exemption From 10CFR Part 50 - Appendix E - Emergency Preparedness Training Exercises at the Yankee Nuclear Power Station (TAC No. M83415),

M. B. Fairtile (USNRC) to J. M. Grant, July 24,1992.

400-13 Deleted.

400-14 NUREG/CR-130, Technology, Safety, and Costs of Decommissioning a Reference Pressurized Water Reactor Power Station, Battelle Pacific Northwest Laboratory, June 1978.

400-15 YAEC-171 1, Yankee Nuclear Power Station Severe Accident Closure Submittal, December21, 1989.

400-5

YNPS FSAR REV. 6/05 400-16 NUREG-0825, Integrated Plant Safety Assessment Systematic Evaluation Program, Yankee Nuclear Power Station, June 1983.YRC-1 125, Propane Evaluation for Temporary Waste Disposal System, May 16, 1996.

400-17 EDC 96-305, "Yard Area Crane Upgrades."

400-18 YRC-1238, DBA Curie Limits, November 17, 2000 400-19 CH-03-0014, Estimate of Total Radioactivity Remaining in the Vapor Container, and the Off-Site Dose Impact Assuming all of this Activity is Released to the Exclusion Area Boundary, D.A. Montt to G. Babineau, May 15, 2003.

400-20 EDC 99-302, Independent Spent Fuel Storage Installation (ISFSI) Facility Design and Construction.

400-21 NUREG 1536, Standard Review Plan for Dry Cask Storage Systems, January 1997.

400-22 NAC-MPC, Final Safety Analysis Report, Multi-Purpose Canister System, Docket No.

72-1025.

400-23 EDC 01-001, Implementation of YR Dry Fuel Storage Using the NAC-MPC System.

l 400-24 ENG 05-11, Thermal Desorption System - Hypothetical Propane Explosion Evaluation, G.H. Philley/B.W. Holmgren to F.N. Williams, May 26, 2005.

400-25 YA-CALC-00-002-05, FSAR Calculation, June 2005 400-6

YNPS FSAR REV. 6/05 TABLE 400.1 Exclusion Area Boundary Dose per Curie Conversion Factors Dose Conversion Factors, rem/Ci TEDE Thyroid Skin Contamination:

MCS and Bleed Line in VC 3.9E-02 1.613-03 9.7E-05 Remaining Systems and Structures 2.713-02 8.813-04 5.613-05 Activation:

Bioshield Concrete 2.213-03 4.613-04 2.813-05 Vessel Cladding 1.11E-02 3.2E-03 1.8E-04 Vessel Wall 2.413-03 7.0E-04 3.9E-05 Bounding Dose Conversion Factors 8.3313-02 1.21E-03 6.5 1 E-05 NOTES:

1.

TEDE is total effective dose equivalent.

2.

Activation results are based on the average value in the core region.

3.

The Main Coolant System (MCS), Bleed Line in the Vapor Container (VC), Reactor Vessel, and the Bioshield Concrete have been permanently removed.

4.

As the radioactive source term on site decays, there are changes in the radionuclide mix available for potential release during an event. As a result of this changing mix, the dose conversion factors must be periodically updated. The Bounding Dose Conversion Factors reported in the table are the limiting values through 1/1/08.

400-7

YNPS FSAR REV. 6/05 TABLE 400.2 Summary of Materials Released in Postulated Events Radioactivity Source of Item Content Radioactivity Main Coolant System:

Piping (1 Loop) 11 Ci Internal Contamination Valves (1 Loop) 1.2 Ci Internal Contamination Pumps (1 Pump) 3.0 Ci Internal Contamination Charging & Volume Control System:

Bleed Line Piping (Total in VC) 4.6 Ci Internal Contamination Feed & Bleed Heat Exchanger (1 Shell) 9.5 Ci Internal Contamination Reactor Vessel:

Intact Removal (Internal Contamination) 23 Ci Internal Contamination Segmented Removal (Cutting Debris) 120 Ci Activated Metal Radioactive Materials Container:

Main Coolant System Piping Container 2.9 Ci Internal Contamination Reactor Vessel Segment Cask 2.3 Ci Internal Contamination Sea-Land Container with Combustible Material 2.9 Ci Internal Contamination NOTES:

1.

Radioactivity content is based on January 1994.

2.

The Main Coolant System, Charging & Volume Control System and Reactor Vessel have been permanently removed.

400-8

YNPS FSAR REV. 6/05 TABLE 400.3 Summary of Safety Analysis Requirements Accident Analysis Section Analysis Requirement 400.1.2 Special Nuclear Material used as reactor fuel will not be moved into the Reactor Vessel.

Refer to Note 1.

400.1.2 Deleted 403.1 Detailed planning of activities that use liquids will ensure liquids will be processed by a liquid waste processing system when required by the ODCM. Existing or supplemental barriers will be used to ensure that inadvertent spills are contained within the liquid waste processing system.

403.1 The following will be performed if the Reactor Vessel is removed as a single 403.3 component:

407.7 Internal Dross will be consolidated to stabilize contamination.

Limitations will be placed on the Reactor Vessel cask transporter operation to ensure that Spent Fuel Pit structural integrity is not adversely affected.

Refer to Note 1.

403.4 Radioactive materials storage areas will be located and arranged such that multiple 406.2 containers or components will not be significantly affected by a fire event causing a 407.7 release of airborne radioactivity exceeding the bounding materials handling event.

403.5 The following components will be handled and transported on-site as single 407.7 containers or components to reduce the consequences of a materials handling event:

Reactor Vessel Casks; Main Coolant System Piping Containers, Valve Containers, and Pumps; Feed and Bleed Heat Exchanger Shells; Bleed Line Piping Containers.

Refer to Note 1.

406 Explosives will not be used at YNPS without completion of a separate safety analysis. The analysis must include the effects of the use of explosives both on the Spent Fuel Pit Building and Fuel Transfer Chute structural integrity and on the l _potential release of airborne radioactivity. Refer to Note 1.

406.1 The quantity of inflammable gases stored on-site will be limited as follows:

The Vapor Container and Potentially Contaminated Area Warehouse will not be used as a general storage location for inflammable gas cylinders. Only cylinders in use or required in the near term should be located inside the structures at any given time. Refer to Note 1.

400-9

YNPS FSAR REV. 6/05 TABLE 400.3 (Continued)

Summary of Safety Analysis Requirements Analysis Requirement All fuel has been transferred to the ISFSI. The NAC-MPC FSAR contains the design basis safety analysis for the NAC-MPC system components in use at the Yankee-Rowe ISFSI.

Deleted Deleted Deleted NOTE:

I1.

The Main Coolant System, Charging & Volume Control System, Reactor Vessel, Spent Fuel Pit Building, Fuel Transfer Chute, and Vapor Container have been permanently removed.

This safety analysis requirement is no longer applicable to the YNPS.

400-10

YNPS FSAR REV. 6/05 TABLE 400.4 Summary of Decommissioning Planning Considerations from the Safety Analysis Accident Analysis Section Decommissioning Planning Considerations 401.1 The Radiation Protection Program will be applied to all activities performed on site involving radioactive materials:

Activities will be managed by qualified individuals who will implement program requirements in accordance with established procedures.

Radiation exposures and the release of radioactive materials to unrestricted areas will be maintained as far below specified limits as is reasonably achievable.

Radiation protection training will be provided to all occupationally exposed individuals.

401.1 Project management will ensure that work specifications, designs, and work packages involving radiation exposure or radioactive materials incorporate effective radiological controls.

401.2 The Occupational Safety Program will be implemented during YNPS decommissioning.

402.2 Hazardous materials handling will be controlled through the Nonradioactive Hazardous Materials Program and the Chemical Control Program.

402.2 Safe storage and use of inflammable gases and any other inflammable materials will be controlled through the Occupational Safety Program and the Fire Protection Program.

403.1 The following will be performed if the Reactor Vessel is removed as a single 403.3 component:

407.7

A local HEPA filtration unit will be used as necessary to remove airborne radioactivity.

A lifting fixture/cover and nozzle covers will be installed to provide additional barriers to the release of radioactivity.

Refer to Note 1.

403.2 Detailed planning will ensure that the following systems with high internal contamination will be dismantled using mechanical methods: Main Coolant System, Feed and Bleed Heat Exchanger, Bleed Line Piping.

Refer to Note 1.

400-11

YNPS FSAR REV. 6/05 Accident Analysis Section Decommissioning Planning Considerations 403.2 The following will be performed if the Reactor Vessel is segmented:

Mechanical cutting methods will be implemented to cut the Reactor Vessel.

Cutting of the Reactor Vessel beltline area will be conducted under water.

A local HEPA filtration unit will be used to remove airborne radioactivity.

Refer to Note 1.

403.5 Openings in components will be covered and sealed to minimize the spread of contamination after dismantlement and before on-site transportation.

404.1 Deleted 404.2 If service water is not available to the Vapor Container fire hose reels, an alternate source of water will be re-established in accordance with the Fire Protection Plan.

Refer to Note 1.

404.3 Deleted 405 The following fire protection features will be maintained through implementation of the Fire Protection Program:

Fire detection equipment and systems

Fire barrier maintenance and control

Personnel training and qualification programs Fire Protection Program procedures

Control of transient combustible materials and ignition sources 407.2 The impact of seismic events will be evaluated during planning of dismantlement activities. The evaluation will consider the possibility of the spread of contamination.

NOTE:

1.

The Main Coolant System, Charging & Volume Control System, Reactor Vessel, and Vapor I

Container have been permanently removed. This decommissioning planning consideration is no longer applicable to the YNPS.

400-12

YNPS FSAR REV. 6/05 401 EVENTS AFFECTING OCCUPATIONAL HEALTH AND SAFETY 401.1 Radiological Events Radiological events could occur which result in increased exposure of decommissioning workers to radiation. However, the occurrences of these events are minimized or the consequences are mitigated through the implementation of the Radiation Protection Program (Section 507) and the Defueled Emergency Plan (Section 515).

The Radiation Protection Program is applied to all activities performed on site involving radioactive materials. The primary objective of the Radiation Protection Program is to protect workers and visitors to the site from radiological hazards that have the potential to develop during decommissioning. The program requires YNPS and its contractors to provide sufficient qualified staff, facilities and equipment to perform decommissioning in a radiologically safe manner.

Activities conducted during decommissioning that have the potential for exposure of personnel to either radiation or radioactive materials will be managed by qualified individuals who will implement program requirements in accordance with established procedures. Radiological hazards will be monitored and evaluated on a routine basis to maintain radiation exposures and the release of radioactive materials to unrestricted areas as far below specified limits as is reasonably achievable.

Radiation protection training will be provided to all occupationally exposed individuals to ensure that they understand and accept the responsibility to follow procedures and to maintain their individual radiation dose as low as is reasonably achievable.

Project management will ensure that work specifications, designs and work packages involving potential radiation exposure or handling of radioactive materials incorporate effective radiological controls. Task planning will include consideration of the potential adverse events. The objective of this planning is to ensure that protective measures and contingency plans are developed to address the potential occurrence of these events and to minimize their impact on the workers as well as the public health and safety.

l The ISFSI Emergency Plan retains an on-site emergency response capability. This capability includes removal of personnel from an affected area, including site evacuation, if necessary. The plan is implemented by the control room personnel.

Implementation of these programs ensures that potential radiological events affecting occupational health and safety will be sufficiently minimized and mitigated to not warrant further consideration in this analysis.

401.2 Nonradiological Events Decommissioning YNPS may require different work activities than were typically conducted during normal plant operations. Effective implementation of the Occupational Safety Program (Section 510) to decommissioning activities will ensure worker safety. The goal of the Occupational Safety Program is to provide a hazard-free environment for employees. The program incorporates safety into every phase of decommissioning from early design through implementation.

401-1

I YNPS FSAR REV. 6/05 Implementation of the Occupational Safety Program will ensure that industrial safety events are eliminated to the maximum extent possible.

401-2

YNPS FSAR REV. 6/05 402 EVENTS AFFECTING PUBLIC HEALTH AND SAFETY 402.1 Off-Site Radiological Events The Decommissioning Environmental Report that was prepared to support the YNPS Decommissioning Plan presented the impact of decommissioning on the public and the environs.

Off-site events related to decommissioning activities are limited to those associated with the shipment of radioactive materials. Radioactive shipments will be made in accordance with all applicable requirements (e.g., NRC, Department of Transportation). The Radioactive Waste Management Program (Section 508) and the Decommissioning Quality Assurance Program (Section 514) assure compliance with these requirements.

Compliance with these requirements ensures that both the probability of occurrence and the consequences of an off-site event do not significantly affect the public health and safety.

402.2 Nonradiological Events There are no decommissioning events that can be initiated from nonradiological sources that could significantly impact public health and safety.

Hazardous materials handling will be controlled through the Nonradioactive Hazardous Materials Program and the Chemical Control Program (Section 511). There are no chemicals stored on-site that, after release, could significantly threaten public health and safety. Inflammable gases stored on-site include combustible gases used for cutting and welding and liquid propane gas (LPG) used for operation of forklift trucks and evaporator. Safe storage and use of these gases and any other inflammable materials is controlled through the Occupational Safety Program and Fire Protection Program (Section 510 and 513).

The programs described above are implemented through procedures that control material identification, inventory, handling, storage, use and disposal, minimizing the probability of on-site nonradiological events. In addition, procedures present mitigative measures that would be implemented if an event occurred.

Implementation of these programs ensures that the probability of occurrence and consequence of on-site nonradiological events do not significantly affect public health and safety.

402.3 Radiological Events There are no radiological events that could affect the health and safety of the public such that it results in a release of radioactive materials exceeding the EPA PAGs. Radiological events are divided into several categories: (1) decommissioning activity events, (2) loss of support system events, (3) fire events, (4) explosion events, (5) external events and (6) spent fuel storage events.

Sections 403 through 408 present the dominant radiological events.

402.4 Radiological Analysis Basis The consequences of postulated decommissioning accident scenarios on the health and safety of the public were determined by calculating the potential dose at the Exclusion Area Boundary. The 402-1

YNPS FSAR REV.6/05 location of the Exclusion Area Boundary is presented in Figure 300-1 as a point 3100 ft from the center of the Vapor Container. The airborne release is assumed to result from release of the entire potential airborne radioactivity in a container or component. Releases to the environment are assumed to be at ground level. Activities that could result in release of radioactive liquids will be designed to contain the releases within the liquid waste processing system using existing or supplemental barriers.

An atmospheric dispersion factor of 2.84E-04 sec/m3 (Reference 400-5) was used to estimate the two hour dose at the Exclusion Area Boundary resulting from a ground level release of radioactivity.

The radionuclide distributions used to evaluate postulated releases are estimated from the radiological scoping survey data. Radionuclide distributions, contamination levels, and radioactivity contents are calculated as of January 1, 1994. An updated calculation has been performed due to radioactive decay and changes in radionuclide distribution over time. This update is discussed in Section 400.

The ELISA computer program (Reference 400-7) was used to calculate the Exclusion Area Boundary doses relative to the radioactivity released (Reference 400-8). The following doses were calculated: total effective dose equivalent, thyroid dose, and skin dose. In each case, the total effective dose equivalent was the limiting value. Table 400.1 presents the dose to radioactivity conversion factors used in the analysis.

The off-site dose was determined based on the airborne radioactivity released in each accident scenario. Most of the calculations for the dominant scenarios used the highest dose to radioactivity conversion factors, which were based on the Main Coolant System and Bleed Line radionuclide distribution. Doses estimated for activated components were based on a combined release of loose activated base material (e.g., fine cutting debris, concrete dust) and surface contamination. Table 400.2 presents a summary of the materials that were assumed to be released.

The radiological analysis results in conservative, bounding estimates of the radiological consequences of the events considered in this analysis based on the following:

The atmospheric dispersion factor was based on conservative meteorology. Realistic meteorology would increase dispersion, decreasing the dose by about a factor of 3 (Reference 400-7).

The radioactivity release estimates assume that all of the radioactivity released to the environment is incorporated into a plume and is transported to the Exclusion Area Boundary.

Only a fraction of the radioactivity released will form a plume and a portion of the plume will drop out prior to reaching the Exclusion Area Boundary, decreasing the dose.

In most cases, the release fractions from the systems, structures and components were assumed to be 100% of the surface contamination and loose activated base metal. This could only result as a nonmechanistic release of radioactivity following a significant force being applied to the container or component. A significant fraction of the contamination that could be released is tightly bound to the surface. Radiological scoping analysis indicated that this fraction was between 50% to 80% (Reference 400-6). Only a fraction of the total radioactivity would be 402-2

I YNPS FSAR RE V.6/05 released as a result of the energy imparted during an impact on the component or container, if the component or container is breached. More realistic release fractions of 1% to 10% are most likely justifiable. These fractions are consistent with drop, fire, and explosion scenarios in comparable evaluations (Reference 400-14).

Releases resulting from accidents postulated in the decommissioning accident analysis were evaluated against the EPA PAGs. Events producing off-site doses less than the guides were classified as not having a significant effect on the public health and safety.

402-3

YNPS FSAR REV. 6105 403 DECOMMISSIONING ACTIVITY EVENTS Decommissioning activities were identified on a location-by-location basis. Plant areas containing radiologically contaminated and activated systems, structures and components were identified.

Dominant systems, structures and components were identified based on the amount of potential airborne radioactivity that could be released during decommissioning activities.

For each of the dominant systems, structures and components in each plant area, five decommissioning process steps were considered: decontamination, dismantlement, packaging, storage and materials handling.

All of the systems that contain significant amounts of surface contamination, and that were evaluated below as decommissioning activity events, have been removed from the site. Therefore, the evaluations below are historical, however, they provide the limiting dose criteria for these categories of events.

403.1 Decontamination Events Selected systems, structures and components will be decontaminated during decommissioning to remove radioactivity from or stabilize radioactivity on external and internal surfaces. General decontamination methods that may be applied during YNPS decommissioning are presented in Section 200. External contamination levels are significantly lower than the contamination levels on internal surfaces of systems and components. Therefore, the bounding decontamination event is based on decontamination of internal surfaces.

Internal decontamination methods typically use liquids to remove radioactivity from the surface (e.g., chemical decontamination, high pressure water washing). Detailed planning of decommissioning activities that use liquids will ensure that contaminated liquids will be processed by the liquid waste processing system. Additionally, existing or supplemental barriers will be used to ensure that inadvertent spills from these activities will be contained within the liquid waste processing system or an approved alternate. These precautions prevent an unmonitored release of radioactive liquids to the environment.

Based on the evaluation of a gas bottle explosion event in the Vapor Container presented in Section 406, the amount of radioactivity that would be transported from the Vapor Container to the environment following the event was about 10%. The motive forces associated with the release of contamination during a decontamination event are significantly less than those during an explosion.

However, an instantaneous release of 10% of the potential airborne radioactivity from the component to the environment was used as a bounding value for the purposes of the calculation.

The estimated radioactivity content of the Reactor Vessel internal surface contamination was about 8 Ci. A release of this material to the environment would result in an off-site dose at the Exclusion Area Boundary of about 0.28 rem. This is significantly less than the EPA PAGs. In addition, radiological scoping analyses indicated that between 50% and 80% of the internal surface contamination is tightly bound. Incorporation of this effect would reduce the dose to less than 0.14 rem.

403-1

YNPSFSAR REV. 6/05 The radiological consequences of the bounding decontamination event result in a dose at the Exclusion Area Boundary significantly less than the EPA PAGs. Therefore, the public health and safety are not significantly affected by the potential decontamination events that could occur during decommissioning.

403.2 Dismantlement Events Systems, structures and components will be dismantled during decommissioning to remove radioactive materials from the site. General dismantlement methods that may be applied during YNPS decommissioning are presented in Section 200. Detailed planning will ensure that systems with high internal contamination are dismantled using mechanical methods (e.g., split frame machining). This restriction limits the amount of airborne contamination generated during dismantlement activities. Thermal dismantlement methods (e.g., plasma arc, oxy-fuel) may be used on systems with lower radioactivity levels.

A bounding analysis was completed for the purposes of this accident analysis to estimate the consequences of generating airborne radioactivity during dismantlement activities. The dominant system, structure or component that could have caused the highest off-site dose as a result of a release of airborne radioactivity during dismantlement was one Main Coolant System Loop. The bounding analysis conservatively assumed that all of the radioactivity on the internal surface of one Main Coolant System Loop was nonmechanistically released to the Vapor Container atmosphere.

This event is bounding because underwater cutting of the highest radioactivity regions of the Reactor Vessel reduces the potential airborne contamination generation from that source.

Based on the evaluation of a gas bottle explosion event in the Vapor Container presented in Section 406, the amount of radioactivity that would be transported from the Vapor Container to the environment following the event was about 10%. The motive forces associated with the release of contamination during a dismantlement event are significantly less than those during an explosion.

However, an instantaneous release of 10% of the potential airborne radioactivity from the Vapor Container to the environment was used as a bounding value for the purposes of the calculation.

The estimated radioactivity content of a Main Coolant System Loop internal surface contamination was about 15 Ci. A release of 10% of this material to the environment would result in an off-site dose at the Exclusion Area Boundary of about 0.051 rem. This is significantly less than the EPA PAGs. In addition, radiological scoping analyses indicated that between 50% and 80% of the internal surface contamination is tightly bound. Incorporation of this effect would reduce the dose to less than 0.026 rem. A nonmechanistic release of the cuttings from segmentation of the upper portion of the Reactor Vessel segmentation would have resulted in an off-site dose of less than 0.013 rem.

The radiological consequences of the bounding dismantlement event result in a dose at the Exclusion Area Boundary significantly less than the EPA PAGs. Therefore, the public health and safety are not significantly affected by the potential dismantlement events that could occur during decommissioning.

403-2

YNPS FSAR REV. 6/05 403.3 Packaging Events Radioactive materials are packaged prior to shipment from YNPS to either a disposal facility or an off-site processing facility. Intermediate packaging may also be used prior to transporting radioactive materials from their removal area to a final packaging area. The materials handling event presented in Section 403.5 nonmechanistically assumes that the potential airborne radioactivity in a bounding container or component is released directly to the environment. This event also bounds any packaging event.

The radiological consequences of a packaging event are bounded by the consequences of a materials handling event. Therefore, the public health and safety are not significantly affected by the potential packaging events that could occur during decommissioning.

403.4 Storage Events Containers and components will be stored on-site prior to shipment to either a disposal facility or an off-site processing facility. Intermediate storage locations may also be used before the radioactive materials are moved to a packaging area. Several evaluations are presented in the accident analysis regarding the storage of radioactive materials:

Fire Events (Section 405)

Explosion Events (Section 406)

External Events (Section 407)

Each of these sections presents restrictions to ensure that adequate separation by barrier, distance or radioactivity content is employed to preclude an event causing a release that exceeds the bounding materials handling event presented in Section 403.5. The materials handling event presented in Section 403.5 nonmechanistically assumes the breach of a single container or component, which releases the total potential airborne radioactivity directly to the environment. Storage areas will be located and arranged such that multiple containers or components could not be affected by a single event causing a release of airborne radioactivity that exceeds the bounding materials handling event.

The radiological consequences of a storage event are bounded by the consequences of a materials handling event. Therefore, the public health and safety are not significantly affected by the potential storage events that could occur during decommissioning.

403.5 Materials Handling Events Materials handling events encompass those events that could potentially occur during movement of radioactive materials from their removal location to a staging location outside of the structure containing the materials. Subsequent handling of these materials is considered by the on-site transportation external event presented in Section 407.7.

After removal, all openings in components will be covered to minimize the spread of contamination.

Components will then either be placed in containers for on-site transportation or be transported individually. The following components containing high potential airborne radioactivity were handled as single containers or components to reduce the consequences of a materials handling 403-3

YNPS FSAR REV. 6/05 event: Reactor Vessel Casks; Main Coolant System Piping Containers, Valve Containers, and Pumps; Feed and Bleed Heat Exchanger Shells; Vapor Container Bleed Line Piping Containers.

The non-mechanistic release of the contents of the feed and bleed heat exchanger described below (for historical purposes), establishes a bounding dose of 320 mrem TEDE for a materials handling event. The feed and bleed heat exchanger has now been removed from the site. New dose conversion factors have been determined based on limiting radionuclide mixes and are presented in Table 400.1. These bounding dose conversion factors may be used to determine the concentration of radioactivity in a single container that would have the potential to produce 320 mrem at the Exclusion Area Boundary from an accidental release.

A bounding analysis was completed for the purposes of this accident analysis to estimate the consequences of generating airborne radioactivity resulting from a materials handling event. The dominant system, structure or component that could have caused the highest off-site dose as a result of a release of airborne radioactivity during handling was one of the four Feed and Bleed Heat Exchanger shells. The bounding analysis conservatively assumed that all of the radioactivity on the internal surfaces of a Feed and Bleed Heat Exchanger shell was nonmechanistically released to the environment. If the Reactor Vessel is removed as a single component, internal dross will be consolidated to stabilize the contamination. This would significantly reduce the amount of radioactivity that could be released during handling.

The estimated radioactivity content of one Feed and Bleed Heat Exchanger shell internal surface contamination was about 9.5 Ci. A release of this material to the environment would result in an off-site dose at the Exclusion Area Boundary of about 0.320 rem. This is significantly less than the EPA PAGs. In addition, radiological scoping analyses indicated that between 50% and 80% of the internal surface contamination was tightly bound. Incorporation of this effect would have reduced the dose to less than 0.160 rem. The nonmechanistic release assumed that the heat exchanger shell failed catastrophically, releasing radioactivity from all surfaces to the environment. Realistically, the shell is structurally stable and total failure is highly unlikely, further reducing the release of radioactivity.

The following are the potential off-site doses at the Exclusion Area Boundary for the nonmechanistic release of the radioactivity from high radioactivity containers presented above:

Vapor Container Bleed Line Piping 0.160 rem Main Coolant Pump (1) 0.100 rem Main Coolant Pipe Container 0.100 rem Reactor Vessel Segment Cask 0.078 rem Main Coolant Valve Container 0.041 rem The radiological consequences of the bounding materials handling event result in a dose at the Exclusion Area Boundary significantly less than the EPA PAGs. Therefore, the public health and safety are not significantly affected by the potential materials handling events that could occur during decommissioning.

403-4

YNPS FSAR RE V. 6/05 404 LOSS OF SUPPORT SYSTEM EVENTS The electric power, cooling water and compressed air systems provide support to ISFSI operations and decommissioning activities. Loss of these systems could potentially affect many systems and plant areas simultaneously. However, none of these support systems are safety related.

404.1 Loss of Off-Site Power Off-site power is used to energize tools, cranes, lighting and air filtering equipment used during decommissioning activities. The following results from a loss of off-site power:

Decommissioning tools, lighting and air filtering equipment are de-energized. All decommissioning activities will be terminated.

Cranes are de-energized and lifting operations are terminated. Cranes fail in a safe condition when they are de-energized.

Back-up power sources such as UPS and Security Diesel Generator backup power capability will be maintained to support Gatehouse and ISFSI operations, however, back-up power is not needed to support decommissioning activities.

Loss of off-site power will not result in the failure of containment systems designed to preclude the spread of contamination (e.g., local contamination control envelopes, HEPA filters). Although the HEPA filter fans will stop, the filter will remain intact and the contamination control envelope will not be breached, preventing unfiltered releases. A breach of the contamination envelope is an independent event and is not considered in this evaluation. Any significant breach of the contamination envelope would be detected and closed independent of a loss of off-site power.

A loss of off-site power does not result directly in a release of radioactive material to the environment during decommissioning activities. Therefore, public health and safety are not adversely affected by a loss of off-site power event.

404.2 Loss of Cooling Water Cooling water is supplied by the Auxiliary Service Water System to a heat exchanger which cools the Spent Fuel Pit Cooling System. As a result of the transfer of all spent fuel to the YNPS ISFSI in 2003, spent fuel cooling capability is no longer necessary.

A loss of cooling water does not result directly in a release of radioactive material to the environment during decommissioning activities. Section 408 demonstrates that the public health and safety would not be adversely affected from a loss of spent fuel cooling capability. Therefore, public health and safety are not adversely affected by a loss of cooling water event.

404-1

YNPS FSAR REV. 6/05 404.3 Loss of Compressed Air

Compressed air is supplied by various air compressors to operate and power pneumatic tools.

Following a loss of compressed air, decommissioning pneumatic tools shut down. This stops any potential releases from these activities.

The radiological consequences of a loss of compressed air event are small and are bounded by the materials handling event presented in Section 403.5. Therefore, public health and safety are not adversely affected by a loss of compressed air event.

404-2

YNPS FSAR REV. 6/05 405 FIRE EVENTS A fire event could affect several plant systems, structures and components simultaneously.

Combustible materials can be ignited by either external ignition sources (e.g., oxyacetylene torches) or internal ignition sources (e.g., spontaneous combustion). Adequate levels of the following fire protection features will be maintained through implementation of the Fire Protection Program (Section 513) minimizing the potential of occurrence of a fire:

Fire detection equipment and systems.

Personnel training and qualification programs.

Fire Protection Program procedures.

Control of transient combustible materials and ignition sources.

In addition, if a fire occurs, the following fire protection features will be employed to limit the consequences to those of the decommissioning materials handling event presented in Section 403.5:

Maintain sufficient fire detection, response and suppression capability.

Separate containers, as necessary, by distance, barrier or radioactivity content.

Higher radioactivity containers are unlikely to fail and cause a significant release due to a fire as these containers are designed for greater levels of integrity. Spontaneous combustion inside a container is highly unlikely as most containers are filled with noncombustible materials. Sea-land containers may be used to ship combustible radioactive materials. The estimated radioactivity level of a sea-land container filled with combustible radioactive material is about 2.9 Ci. Release of all of the radioactivity as a result of a fire would result in an off-site dose of 0.100 rem at the Exclusion Area Boundary. The assumption of a total release of radioactivity is very conservative. Reference 400-14 presents a release fraction of 0.00015 for a similar event. Incorporation of this assumption would significantly reduce the release.

Implementation of the Fire Protection Program minimizes the probability of occurrence of a fire.

Implementation of the restrictions presented above limits radiological consequences at the Exclusion Area Boundary to a value significantly less than the EPA PAGs. Therefore, public health and safety are not adversely affected by a fire event.

405-1

YNPS FSAR REV. 6/05 406 EXPLOSION EVENTS An explosion event could affect several plant systems, structures and components simultaneously.

Explosions are possible from the following sources:

Ion Exchange Resin Offgases

Explosives

Inflammable Gas Storage Bottles and Tanks Processing and limitation of ion exchange resin offgases will continue to be controlled by procedures that have successfully precluded this type of accident throughout the greater than 30-year operation of YNPS. Explosives will not be used at YNPS without completion of a separate safety analysis.

The analysis must include the effects of the use of explosives and assess the impact on the YNPS ISFSI and associated support systems.

Inflammable gases (e.g., acetylene, LPG) may be used during decommissioning for thermal cutting or to power material handling and processing equipment. Inflammable gas cylinders will be used, located and stored in quantities such that the possibility of explosion is minimized. Two 1,000 gallon propane tanks, were previously located approximately 650 feet from the ISFSI to supply the Temporary Waste Water Processing Island System's gas-fired boiler. Currently six 1,000 gallon propane tanks, located approximately 1180 feet from the ISFSI, supply a thermal desorption system in use to process PCB paint contaminated soils. The quantities of the inflammable gas cylinders and the location of the propane tanks will be such that the radiological consequences of an explosion are bounded by the events described below.

406.1 Explosion Events: Vapor Container Inflammable gases may be used in the Vapor Container during thermal cutting activities. The Decommissioning Plan recommends mechanical cutting of significantly contaminated systems, structures and components to minimize the generation of airborne radioactivity. However, thermal cutting methods utilizing inflammable gases could be used for dismantlement activities on components with lower contamination levels.

The release of radioactive material from an explosion event in the Vapor Container is discussed below for historical purposes. It established a limiting dose of 0.440 rem for this type of event. All of the contaminated systems have subsequently been removed from the Vapor Container. The total surface contamination remaining is currently estimated to be 0.05 Ci (Reference 400-19). Using the bounding dose conversion factor from Table 400.1, the dose at the Exclusion Area Boundary would be 3.68 mrem if 100% of this radioactivity were to be released to the environment. Since this dose is less than the limiting value established below, it is no longer necessary to require Vapor Container isolation or limit the quantity of explosive gases used in the Vapor Container.

An engineering evaluation was completed to determine the physical consequences of an explosion in the Vapor Container (Reference 400-10). If the Vapor Container is closed, the peak pressure resulting from an explosion of a standard cylinder of acetylene is less than 5% of the Vapor Container design pressure. If the Vapor Container is not isolated (e.g., Equipment Hatch open),

about 10% of the Vapor Container air mass could be released through existing openings before the 406-1

YNPS FSAR REV. 6/05 pressure inside the Vapor Container equalizes with the outside atmospheric pressure. It is assumed that radiological contamination controls are re-established after the explosion by closing Vapor Container openings to terminate the release to the environment.

The explosion evaluation assumes that all of the energy from the gas explosion is converted into a pressure increase. This is a highly conservative assumption; most of the energy would be lost to other effects (e.g., heat generation, mechanical deformation). Less than 5% of the energy most likely would be converted into the pressure increase.

If an explosion nonmechanistically releases all of the internal surface contamination contained in systems inside the Vapor Container to the Vapor Container atmosphere, the release to the environment would not exceed the EPA PAGs at the Exclusion Area Boundary. The radioactivity content of the internal surface contamination in systems inside the Vapor Container is less than 130 Ci. Release of 10% of this material to the environment would result in a dose at the Exclusion Area Boundary of about 0.440 rem.

It is highly unlikely that all systems either will be opened or will be in the Vapor Container such that an explosion could remove the internal radioactivity. In addition, a significant fraction of the internal surface contamination is tightly bound. Radiological scoping analyses indicated that this fraction varied between 50% and 80%. The combination of these effects would reduce the release by 70% to 90%, resulting in an off-site dose of less than 0.100 rem.

In order to minimize the overall risk of an inflammable gas-air explosion in the Vapor Container, the Vapor Container should not be used as a general storage location for inflammable gas cylinders.

Only cylinders in use or required in the near term should be located inside the structure at any given time. Additionally, thermal cutting methods using inflammable gases were not used to dismantle the following components with high contamination levels: Reactor Vessel, Main Coolant System, Feed and Bleed Heat Exchanger, Vapor Container Bleed Line Piping.

The consequences of an explosion in the Vapor Container are significantly less than the EPA PAGs.

Therefore, there is no significant impact on pubic health and safety.

406.2 Explosion Events: Potentially Contaminated Area Warehouse Inflammable gases may be used in the Potentially Contaminated Area Warehouse to power forklift trucks and to support thermal cutting activities. In the unlikely event that a single inflammable gas cylinder explodes in the warehouse, the pressure would be in excess of the capacity of the warehouse if the structure is sealed. Doors, portions of the roof or portions of the walls would likely fail in order to relieve excess pressure. Explosion released overpressure would not cause the containers to explode since the explosion is external to the containers. Some damage of the containers would be likely, but a release of significant portions of the contents of the containers as airborne radioactivity is not likely. In addition, the segregation and separation practices implemented for fire protection purposes would also limit the effects on storage containers.

In order to minimize the overall risk of an inflammable gas-air explosion in the Potentially Contaminated Area Warehouse, the warehouse should not be used as a general storage location for 406-2

YNPS FSAR REV. 6/05 inflammable gas cylinders. Only cylinders in use or required in the near term should be located inside the structure at any given time.

The consequences of an explosion in the Potentially Contaminated Area Warehouse are bounded by those of the decommissioning materials handling events presented in Section 403.5. Therefore, there is no significant impact on pubic health and safety.

406.3 Explosion Events: ISFSI In EDCR 01-001 (Reference 400-23), an evaluation concluded that the existing NAC analysis for overpressure bounded an explosion of the two 1000 gallon propane tanks and the resulting effects on the casks within the FTE. Since the FTE is closer than the ISFSI to the propane tanks, the effects of an explosion of these tanks on the ISFSI is also bounded by the existing NAC-MPC analysis. An evaluation (Reference 400-24) concluded that the existing NAC analysis for overpressure also bounded an explosion of the six 1000 gallon propane tanks and there would be no adverse impact on the spent fuel canisters stored at the ISFSI.

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YNPS FSAR REV. 6/05 407 EXTERNAL EVENTS A systematic assessment of external events was made to evaluate the effects of natural and manmade events on decommissioning activities. The hazards associated with these events are assumed to be consistent with those that could have occurred while YNPS was in operation. Seven external events were identified as having potential applicability to the YNPS decommissioning based on a review of the natural and manmade external events presented in NUREG/CR-2300 (Reference 400-11).

All YNPS spent fuel is located at the ISFSI and is stored within fifteen NAC Multi-Purpose Canisters (MPCs). Chapter 11 of the NAC-MPC FSAR addresses the various off-normal and accident events which were postulated in support of the licensing and certification of the system for General License use (Reference 400-22). In each case, there were no radiological consequences as a result of the postulated event.

407.1 Aircraft Impact YAEC evaluated the significance of a potential aircraft hazard on YNPS in response to Systematic Evaluation Program Topic III-4 (Reference 400-16). The analysis concluded that the annual probability of an aircraft impact was very low. Further consideration of the interaction between an aircraft impact and decommissioning is not warranted.

407.2 Earthquake A seismic event during decommissioning could initiate a materials handling event similar to those described in Section 403.5. The analysis in Section 403.5 concludes that the bounding material handling event results in an off-site dose that is significantly less than the EPA PAGs. In addition, detailed planning of dismantlement activities will consider the impact of seismic events on components that are affected by removal activities. These components will be evaluated and physically supported, as appropriate, to limit the off-site dose resulting from a release of radioactivity to less than the EPA PAGs.

The ISFSI pad was designed and constructed in accordance with EDCR 99-302 (Reference 400-20).

The ISFSI pad was qualified to the YNPS site design basis earthquake as developed during the Systematic Evaluation Program (SEP) with the NRC. NUREG-1536 (Reference 400-21) specifically identifies that, for nuclear plant sites licensed before the implementation of 10 CFR 100, the ISFSI Design Earthquake must be equal to or greater than the design earthquake developed for the SEP plants. The site design basis earthquake peak ground acceleration is O.19g horizontal, 0.13g vertical whereas the NAC-MPC system was designed for 0.25g horizontal, 0.17g vertical. Since the site design basis earthquake is less than the NAC-MPC design basis earthquake, a VCC tip-over will not occur.

The potential for radiological consequences from a seismic event resulting in a dose at the Exclusion Area Boundary greater than the EPA PAGs is extremely low. In the unlikely event that a materials handling event is initiated, the consequences would be significantly less than the EPA PAGs.

Therefore, the public health and safety are not significantly affected by a potential seismic event during decommissioning.

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YNPS FSAR REV. 6/05 407.3 External Flooding A flooding event at YNPS typically would be preceded by a sufficient warning period to prepare the site for the event by securing decommissioning activities. Most of the potentially removable radioactivity at YNPS is located in the Vapor Container, well above the potential flood height. Most of the balance of contaminated systems, structures and components would either be packaged for shipment or secured inside buildings. Containers that hold high radioactivity materials are designed for greater levels of structural integrity, providing additional protection. In the unlikely event that a lower radioactivity container or component is exposed to flood waters and radioactive material is dispersed, the flooding dilution effect results in a radiological consequence significantly less than an airborne release of a similar amount of radioactivity.

Flooding could initiate a loss of off-site power event. The analysis in Section 404.1 concludes that public health and safety are not adversely affected from a loss of off-site power event.

Flooding events at YNPS do not result in a significant radiological release, therefore, public health and safety are not adversely affected.

407.4 Tornadoes and Extreme Winds The annual strike probability of a tornado that could cause a significant release of radioactivity from a container or component is very low. In addition, most components and containers that would be vulnerable to a tornado will be packaged awaiting shipment. The integrity of these containers would limit the probability and consequences of a significant release of radioactivity. Further consideration of the interaction between a tornado and decommissioning is not warranted.

An extreme winds event (i.e., hurricanes above 74 mph, thunderstorms above 100 mph) at YNPS typically would be preceded by a sufficient warning period to prepare the site for the event by securing decommissioning activities. Most of the potentially airborne radioactivity at YNPS is located in the Vapor Container, which protects the components from the effects of extreme winds and associated missiles. Most of the balance of contaminated systems, structures and components would either be packaged for shipment or secured inside buildings with system integrity, providing protection from the extreme winds and associated missiles.

Containers that hold higher radioactivity materials are designed for greater levels of structural integrity, providing additional protection. In the unlikely event that a lower radioactivity container or component is unprotected and is exposed to extreme winds and radioactive material is dispersed, the combination of low radioactivity content and significant dispersion by the wind would result in an off-site dose that is bounded by the limiting release of the materials handling event presented in Section 403.5.

Tornadoes or extreme winds could initiate a loss of off-site power event. The analysis in Section 404.1 concludes that public health and safety are not adversely affected from a loss of off-site power event.

Extreme tornadoes and winds events at YNPS will not result in a significant radiological release, therefore, public health and safety are not adversely affected.

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YNPS FSAR REV. 6/05 407.5 Forest Fire A forest fire event at YNPS typically would be preceded by a sufficient warning period to prepare the site for the event by securing decommissioning activities if necessary. In addition, YNPS is protected from the effects of a forest fire by a buffer zone surrounding the facility. The probability of a forest fire of sufficient intensity to bridge the zone is very low (Reference 400-15).

In the unlikely event that a forest fire causes an on-site fire, the fire protection features described in Section 405 would be sufficient to mitigate the consequences of the fire. A forest fire could also initiate a loss of off-site power event. The analysis in Section 404.1 concludes that public health and safety are not adversely affected from a loss of off-site power event.

Forest fire events at YNPS will not result in a significant radiological release, therefore, public health and safety are not adversely affected.

407.6 Lightning Event The lightning strike annual probability for a decommissioning activity or single exposed container or component is very low. In addition, YNPS structures provide protection against lightning.

Although the effects of lightning generally are localized, a lightning strike could initiate a loss of off-site power event or a fire. The analyses in Sections 404.1 and 405 conclude that public health and safety are not adversely affected from a loss of off-site power event or fire event. Further consideration of the interaction between decommissioning and a lightning event are not warranted.

407.7 On-Site Transportation Accidents On-site transportation accident events consist of those events occurring after removal of a container or components from a structure, but before final packaging for shipment. These events could occur during transportation of a container or component to a packaging area.

Detailed planning will ensure that only one container or component with high potential airborne radioactivity is transported simultaneously by the same vehicle until they are prepared for final shipment. This constraint ensures that the off-site dose resulting from an on-site transportation event is bounded by the materials handling event presented in Section 403.5. Additionally, detailed planning will limit the maximum number of containers transported on-site to ensure that they are also bounded by this analysis.

Containers and components stored in materials storage areas awaiting final packaging will be located to preclude impact from a runaway vehicle. The location selection will consider separation by distance, barrier or radioactivity to ensure that the off-site dose resulting from an impact is bounded by the dose resulting from the materials handling event presented in Section 403.5.

The radiological consequences of an on-site transportation event are bounded by a decommissioning materials handling event. The physical consequences of an on-site transportation event does not l adversely affect the dry cask storage system at the ISFSI. Therefore, public health and safety are not adversely affected by an on-site transportation event.

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YNPS FSAR REV. 6/05 408 SPENT FUEL STORAGE EVENTS The initial decommissioning accident analysis evaluated three potential events related to spent fuel storage in the Spent Fuel Pit. These events were a fuel handling accident, a loss of spent fuel pool cooling capability, and interactions between fuel in the Spent Fuel Pit and decommissioning activities. The results of this evaluation were reported to be potential doses that were well below the EPA PAGs. The spent reactor fuel has now been removed from the Spent Fuel Pit and transferred to an on-site ISFSI. The evaluation of spent fuel storage events in the Spent Fuel Pit is, therefore, no longer necessary.

408.1 Events All YNPS spent fuel is located at the ISFSI and is stored within fifteen NAC Multi-Purpose Canisters (MPCs). Chapter 11 of the NAC-MPC FSAR addresses the various off-normal and accident events which were postulated in support of the licensing and certification of the system for General License use (Reference 400-22). In each case, there were no radiological consequences as a result of the postulated event.

The YNPS developed and maintains a 10CFR72.212 evaluation report, which concludes that all site-specific parameters found at the YNPS site are bounded by the NAC-MPC design basis analyses performed for the NAC-MPC system.

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YNPS FSAR REV. 6/05 500 ADMINISTRATION OF THE DECOMMISSIONING PLAN 500.1 Regulatory Basis for Administration of the Decommissioning Plan Regulatory Guide 1.185, "Standard Format and Content for Post-Shutdown Decommissioning Activities Report)" (Reference 500-6), encourages licensees with an approved Decommissioning Plan to "extract the pertinent detail from the Decommissioning Plan and submit a PSDAR update in the format and content specified by the regulatory guide." As a result, information from the approved YNPS Decommissioning Plan that was incorporated into the FSAR has since been extracted, updated and relocated to the PSDAR within the FSAR.

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YNPS FSAR REV. 6/05 REFERENCES 500-1 Letter, YAEC to USNRC, dated December 20, 1993.

500-2 Draft Regulatory Guide DG-1005, "Standard Format and Content For Decommissioning Plans for Nuclear Reactors."

500-3 Letter, USNRC to YAEC, dated February 14, 1995.

500-4 Letter, YAEC to USNRC, dated August 22, 1995.

500-5 NUREG-0586, Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities, August 1988.

500-6 Regulatory Guide 1.185, "Standard Format and Content for Post-Shutdown Decommissioning Activities Report, dated July 2000."

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YNPS FSAR RE V. 6/05 501 DECOMMISSIONING COST ESTIMATE AND FUNDING PLAN 501.1 Decommissioning Cost Estimate This information has been extracted, updated and relocated to the PSDAR within the FSAR.

501.2 Decommissioning Funding This information has been extracted, updated and relocated to the PSDAR within the FSAR.

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YNPS FSAR REV. 6/05 502 DECOMMISSIONING ORGANIZATION AND RESPONSIBILITIES 502.1 YAEC Commitment Yankee Atomic Electric Company (YAEC) is committed fully to compliance with the existing license and applicable regulatory requirements during all phases of YNPS decommissioning.

YAEC's commitment to the safe decommissioning of the facility will be accomplished with diligence and quality. Corporate principles, policies, and goals will be followed to ensure performance excellence, management competence, and high standards in every facet of the decommissioning.

502.1.1 Goals The primary goals of the YNPS decommissioning are 1) to safely remove the nuclear facilities from service and to reduce residual radioactivity to a level that permits release of the property for unrestricted use and termination of the license and 2) to safely store spent nuclear fuel until it can be removed from the site. While achieving these primary goals, YAEC will conduct all decommissioning operations consistent with applicable regulations and a focus on the following considerations:

Maintain radiation exposure to the public and on-site personnel as low as is reasonably achievable.

Ensure occupational safety for all on-site personnel.

Minimize environmental impact.

Minimize radioactive waste generation.

Ensure prudent expenditure of decommissioning funds.

502.1.2 Organizational Strategy The organizational strategy must ensure that adequate numbers of experienced and knowledgeable personnel are available to perform the technical and administrative tasks required to decommission YNPS. The YNPS decommissioning organization is currently staffed with experienced individuals who have worked at or are very knowledgeable of the plant. YAEC believes that retaining personnel with intimate knowledge of YNPS is important to the success of decommissioning.

YAEC is responsible for YNPS decommissioning. In this position YAEC has direct control and oversight over all decommissioning activities. This role is similar to that taken by YAEC during the 31 year operation of YNPS. In that role YAEC provided operational, technical, licensing, and project management support of YNPS. YAEC will contract services to supplement its capabilities as necessary.

Organizational impacts associated with the transition from plant closure to dismantlement were evaluated periodically to assure that adequate and appropriate staffing levels and capabilities were 502-1

YNPS FSAR REV. 6/05 preserved. YAEC will maintain staffing, to the extent possible, to optimize resource utilization and to capitalize on positive project momentum. Both of these are characteristics of successful project teams.

The Yankee Decommissioning Quality Assurance Program (YDQAP) requires indoctrination and training of staff personnel involved with YDQAP activities.

502.2 YAEC Organization and Functions The YAEC organizational structure that will be implemented during decommissioning is presented in Section 1 of the YDQAP. Decommissioning functions are described in site administrative procedures.

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YNPS FSAR RE4V. 6/05 503 TRAINING PROGRAM YAEC will maintain a training program commensurate with the needs of the various phases of decommissioning and the Training Rule I OCFR50.120. The training program, in conjunction with other administrative programs and controls, will ensure that qualified individuals are available to operate and maintain the facility in a safe manner. The training programs will be based on a systematic analysis ofjob performance requirements. The analysis will ensure that personnel will have qualifications commensurate with the performance requirements of their jobs.

503.1 General Employee Training General Employee Training will be provided to all personnel who have unescorted access to the YNPS site. Initial and annual requalification training will include the following topics commensurate with the stages of decommissioning which typically will consist of information on the following topics:

Plant Organization and Administration

Plant Description

Occupational Safety

Quality Assurance

Fire Protection

Emergency Response

Radiation Protection

Security 503.2 Radiation Worker Training Initial and annual requalification radiation worker training will be provided to personnel who require entry into the Radiologically Controlled Area (RCA). Training topics and scope will be commensurate with the stages of decommissioning, which typically will consist of information on the following topics:

Fundamentals of radiation.

Radiation and contamination measurement and control.

Maintaining radiation dose as low as is reasonably achievable.

Radioactive waste minimization.

Radiation work permits.

Radiation protection issues associated with decommissioning.

Donning and removing protective clothing.

503.3 Certified Fuel Handler YNPS had previously implemented, with NRC review and approval, a Certified Fuel Handler training program which replaced the 1 OCFR Part 55 NRC licensed operator training program 503-1

YNPS FSAR REV. 6/05 (Reference 503-2). Following the transfer of all spent fuel to the YNPS ISFSI, there is no longer a need to maintain Certified Fuel Handlers or the training program. NRC issued Amendment 157 which, in part, eliminated the requirement for the CFH training program (Reference 503-4).

503.4 Specific Job Training YNPS training programs will assure the following:

Personnel responsible for performing activities are instructed as to the purpose, scope, and implementation of applicable controlling procedures.

Personnel performing activities are trained as appropriate, in the principles and techniques of the activity being performed.

The methods of implementing the training programs are documented.

503.5 Nonradiation Worker Indoctrination Personnel who are not qualified radiation workers may be granted escorted access to the Radiologically Controlled Area (RCA). All personnel without radiation worker training require escorts that are qualified radiation workers.

503.6 Training Records Training records will be maintained and retained in accordance with administrative plant procedures.

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YNPS FSAR REV. 6/05 REFERENCES 503-1 BYR 93-055, Implementation of the Training Rule, IOCFR50.120, J. K. Thayerto M. B. Fairtile (USNRC), July 28, 1993.

503-2 NYR 92-122, NRC Approval of Certified Fuel Handler Program and Termination of Operator Licenses (TAC No. M83384), M. B. Fairtile (USNRC) to J. M. Grant, June 26, 1992.

503-3 NYR 93-149, NRC Partial Exemption from the Training Rule, 10CFR50.120 (TAC No. M87168), M. B. Fairtile (USNRC) to J. M. Grant (YAEC), November 19, 1993.

503-4 NYR 03-027, J. Hickman (USNRC) to J. Kay (YAEC), Issuance of Amendment #157 RE:

Deletion of Operational and Administrative Requirements Following Fuel Transfer to ISFSI, April 18, 2003.

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YNPS FSAR REV. 6/05 504 DEFUELED TECHNICAL SPECIFICATIONS 504.1 Description After the decision to permanently cease YNPS reactor operations, YAEC submitted a proposed change in March 1992 to the NRC modifying the plant full power operating license to a possession only license status (Reference 504-1). The proposed change removed authority of YAEC to operate the reactor and move fuel back into either the Reactor Vessel or Vapor Container. The facility license was amended in August 1992 to a possession only status (Reference 504-2).

Coincident with the possession only license submittal, YAEC reviewed the plant licensing basis to determine the applicability of existing Technical Specifications to a permanently defueled condition.

The Technical Specifications which were applicable to a defined mode associated with fuel in the reactor were determined to be not applicable to the permanently defueled condition. Technical Specifications that were determined to be applicable at all times were reviewed to assess their relevance to the permanently defueled condition. Individual Technical Specification changes were proposed for the higher priority items identified by the review.

On December 23, 1992, a Technical Specification change was proposed to complete the transition of the Technical Specifications to be applicable to the permanently defueled condition (Reference 504-3). The proposed change eliminated specifications that were not applicable and reformatted the specifications that remained applicable. The NRC approved the license amendment on June 11, 1993 (Reference 504-4).

In January 2003, YAEC submitted a proposed license and Technical Specification change to delete operational and administrative requirements that were no longer applicable following transfer of all spent fuel to the YNPS ISFSI (Reference 504-5). All remaining operational and administrative requirements were relocated to the YDQAP. The NRC approved the proposed changes on April 18, 2003 (Reference 504-6).

Limiting conditions for operation and surveillance requirements were reduced to the following items which have been relocated to the Yankee Decommissioning Quality Assurance Program (YDQAP):

Liquid Hold-Up Tanks - This specification presents radioactive material curie limits for tanks that are not protected by engineered features that would contain inadvertent release of their contents.

Sealed Source Contamination - This specification presents maximum contamination limits for sealed sources to ensure that allowable intake limits are not exceeded.

The design features and administrative controls were modified to delete portions that were not applicable to a permanently defueled condition. With all spent fuel located at the YNPS ISFSI, all remaining administrative controls have been relocated to the YDQAP.

In November 2003, YAEC submitted the YNPS License Termination Plan (LTP) and a proposed amendment to the YNPS Possession Only License (Reference 504-7). The License Termination 504-1

YNPS FSAR REV. 6/05 Plan demonstrates that the remaining decommissioning activities will be performed in accordance with the requirements of Title 10 of the Code of Federal Regulations, will not be inimical to the common defense and security or to the health and safety of the public, and will not have a significant effect on the quality of the environment. The proposed license amendment adds a license condition that provides criteria to determine the need for NRC approval of changes to the approved License Termination Plan. In November 2004, YAEC submitted Revision 1 to the YNPS LTP (Reference 504-8). This revision incorporated YAEC responses to NRC Request for Additional Information (RAIs). In April 2005, YAEC submitted a supplement to the proposed amendment to the YNPS Possession Only License (Reference 504-9). This supplement added a new proposed license condition related to groundwater radionuclide concentrations. NRC approval of these submittals is pending. Table 504.1 summarizes the Technical Specification changes approved by the NRC.

504-2

YNPS FSAR REV. 6/05 REFERENCES 504-1 BYR 92-037, Request For Modification of Yankee Nuclear Power Station's Operating License To Remove Authorization For Reactor Power Operation - Possession Only License, J. K. Thayer to T. Murley (USNRC), March 27, 1992.

504-2 NYR 92-148, Issuance of Amendment No. 142 To Facility License No. DPR Yankee Nuclear Power Station (TAC No. M83024), M. B. Fairtile (USNRC) to J. M. Grant, August 8, 1992.

504-3 BYR 92-112, Permanently Defueled Technical Specifications, J. K. Thayer to M. Fairtile (USNRC), December 21, 1992.

504-4 NYR 93-062, Issuance of Amendment No. 148 To Facility Possession Only License No. DPR Yankee Nuclear Power Station (TAC No. M85244), M. B. Fairtile (USNRC) to J. M. Grant, June 11, 1993.

504-5 BYR 03-001, Deletion of Operational and Administrative Requirements from the Yankee Nuclear Power Station Defueled Technical Specifications, K.J. Heider to NRC, January 14, 2003.

504-6 NYR 03-027, J. Hickman (USNRC) to J. Kay (YAEC), Issuance of Amendment 157 RE:

Deletion of Operational and Administrative Requirements Following Fuel Transfer to ISFSI, April 18, 2003.

504-7 BYR 03-80, B.C. Wood (YAEC) to USNRC, Submittal of Yankee Nuclear Power Station's License Termination Plan and Proposed Revision to Possession Only License, November 24, 2003 504-8 BYR 04-133, J.A. Kay (YAEC) to USNRC, submittal of Revision I to the Yankee Nuclear Power Station's License Termination Plan, November 19, 2004 504-9 BYR 05-28, J.D. Rollins (YAEC) to USNRC, Supplement to Proposed Amendment to YNPS Possession Only License, April 7, 2005 504-3

YNPS FSAR REV. 6/05 TABLE 504.1 Technical Specification Changes for Transition to Permanently Defucled Condition Amendment Submittal Approval Date Number Reference Minimum Shift July 22, 1992 141 NYR 92-143, Issuance of Staffing Amendment No. 141 To Facility Operating License No. DPR Yankee Nuclear Power Station (TAC No. M83383).

Possession Only August 8, 1992 142 NYR 92-148, Issuance of License Amendment No. 142 To Facility License No. DPR Yankee Nuclear Power Station (TAC No. M83024).

Transfer of Fire August 20, 1992 144 NYR 92-156, Issuance of Protection Amendment No. 144 To Facility Specifications to Possession Only License No.

Manual DPR Yankee Nuclear Power Station (TAC No. M83746).

PORC Administration September 4, 1992 145 NYR 92-182, Issuance of Change Amendment No. 145 To Facility Possession Only License No. DPR Yankee Nuclear Power Station (TAC No. M84005).

RETS/REMP Transfer November 5, 1992 146 NYR 92-183, Issuance of Amendment No. 146 To Facility Possession Only License No.

DPR Yankee Nuclear Power Station (TAC No. M84372).

Permanently Defueled June 11, 1993 148 NYR 93-062, Issuance of Technical Amendment No. 148 To Facility Specifications Possession Only License No. DPR Yankee Nuclear Power Station (TAC No. M85244).

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YNPS FSAR REV. 6/05 Submittal Approval Date Amendment Reference Number Spent Fuel Pit - Crane June 17, 1998 149 NYR 98-029, Issuance of Travel and Water Level Amendment No. 149 to Possession Changes Only License No. DPR Yankee Nuclear Power Station (TAC No.

M99527).

Deletion of Site Boundary February 3, 1999 150 NYR 99-012, Issuance of Definition and Relocation of Amendment No. 150 to Possession Tech Spec Figure 5.1-1 Only License No. DPR Yankee Nuclear Power Station (TAC No.

MA3414).

Change Submittal Interval of March 5, 1999 151 NYR 99-021, Issuance of RETS from Semi-annual to Amendment No. 151 to Possession Annual Only License No. DPR Yankee Nuclear Power Station (TAC No.

MA3894).

Eliminate Fitness-for-Duty August 27, 1999 152 NYR 99-071, Issuance of Requirement from License Amendment No. 152 to Possession Only License No. DPR Yankee Nuclear Power Station (TAC No.

MA5218).

Delete Overtime Restrictions October 8, 1999 153 NYR 99-079, Issuance of from Tech Specs Amendment No. 153 to Possession Only License No. DPR Yankee Nuclear Power Station (TAC No.

MA5030).

Consolidate Management June 20, 2000 154 NYR 00-044, Issuance of Positions and Transfer Amendment No. 154 to Possession Review and Audit Functions Only License No. DPR Yankee to the YDQAP Nuclear Power Station (TAC No.

MA5032).

Transfer Certain May 15, 2001 155 NYR 01-043, Issuance of Administrative Requirements Amendment No. 155 to Possession to the YDQAP Only License No. DPR Yankee Nuclear Power Station (TAC No.

MA0929).

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YNPS FSAR RE V. 6/05 Submittal Approval Date Amendment Reference Number Revise Security Plan to March 13, 2002 156 NYR 02-023, Issuance of address Independent Spent Amendment No. 156 to Possession Fuel Storage Installation Only License No. DPR Yankee Nuclear Power Station (TAC No.

MB0209).

Bases Change for Crane August 7, 2002 NYR 02-061, Bases Change for Travel Over the Spent Fuel Crane Travel Over the Spent Fuel Pit Pit - Yankee Nuclear Power Station (TAC No. MB5634).

Delete Operational and April 18, 2003 157 NYR 03-027, Issuance of Administrative Requirements Amendment No. 157 to Possession from Tech Specs. Relocate Only License No. DPR Yankee Remaining Operational and Nuclear Power Station (TAC No.

Administrative Requirements L52086).

to the YDQAP 504-6

YNPS FSAR REV. 6/05 505 PROCEDURES 505.1 Normal Defueled Operating Procedures The normal Operating Procedures used in the defueled condition have been developed in accordance with the Yankee Decommissioning Quality Assurance Program. The controlled procedures, utilized by plant department(s) personnel for the conduct of plant operations, are contained in the Plant Procedures Manual and are available for inspection.

505.2 Emergency Procedures The Emergency procedures are comprised of the Emergency Operating Procedures and the Emergency Plan Implementing Procedures. These procedures are revised, reviewed and controlled in accordance with the Yankee Decommissioning Quality Assurance Program.

The Emergency Operating Procedures provide instructions to the Operations personnel. These procedures identify various plant system(s) and equipment failure/malfunction(s) that may occur, describe the conditions that enable operator recognition of such an event, and provide instructions that assist in mitigating an event and its consequence(s) to the plant, plant personnel, the public, and the environment.

The Emergency Plan Implementing Procedures (EPIPs) provide instruction to Emergency Response Organization (ERO) personnel. These procedures describe the responsibilities and provide for:

Classifying emergencies; notifying plant personnel and off-site agencies; activating the Emergency Response Facilities (ERFs); controlling on-site radiation exposures; responding to medical emergencies, performing security activities; evaluating radiological data; formulating protective action recommendations; releasing public information; operating emergency computer equipment; and preparing for recovery of the plant. Other EPIPs provide for maintaining the ERFs, maintaining emergency response equipment, and training the ERO.

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YNPS FSAR REV. 6/05 506 MAINTENANCE PROGRAM The Maintenance Program is designed to ensure continued system reliability for those systems used to support dry cask storage at the YNPS ISFSI. The program is implemented through administrative and maintenance procedures. These procedures have been developed in accordance with the Yankee Decommissioning Quality Assurance Program. The controlled procedures are utilized for the conduct of maintenance activities and are available for inspection.

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YNPS FSAR REV. 6/05 507 RADIATION PROTECTION PROGRAM 507.1 Introduction The Code of Federal Regulations defines decommissioning as the activities necessary "to remove (as a facility) safely from service and reduce residual radioactivity to a level that permits release of the property for unrestricted use and termination of license" (1 OCFR50.2). A comprehensive Radiation Protection Program is needed to meet the objectives presented in the definition. YNPS intends to maintain essential elements of the Radiation Protection Program that were implemented successfully during its 31 year operating life to meet this goal. Changes to the program will be made as necessary to meet the needs of decommissioning.

The Radiation Protection Program has undergone and continues to undergo inspections and audits from both the NRC and the YAEC Quality Assurance Department. The purpose of these reviews is to ensure that the program complies with the Code of Federal Regulations, applicable regulatory guidance documents, and industry standards. YAEC is committed to maintaining a high level of performance and to enhancing the quality of the Radiation Protection Program throughout the YNPS decommissioning.

The YNPS Radiation Protection Program for decommissioning will continue to be implemented through existing YNPS administrative procedures. These procedures constitute the highest tier documentation of the Radiation Protection Program and define the radiation protection organization, responsibilities, authorities, administrative policies, program objectives, and standards to implement the Radiation Protection Program.

This section of the FSAR presents an overview of the Radiation Protection Program administrative and implementing procedures that will be used during decommissioning.

507.2 Management Policies 507.2.1 Management Policy Statement YAEC is committed to the safe decommissioning of YNPS. The primary objective of the Radiation Protection Program is to minimize the actual and potential exposure of workers, visitors, and general public to radiation. YAEC and its contractors will provide sufficient qualified staff, facilities, and equipment to conduct a radiologically safe decommissioning. YAEC will continue to comply with regulatory requirements, radiation exposure limits, and radioactive material release limits. In addition, YAEC will make every effort to maintain radiation exposures and releases of radioactive materials in effluents to unrestricted areas as low as is reasonably achievable (ALARA).

The ALARA philosophy will be incorporated into all decommissioning activities and will have full management support.

YAEC requires good radiation work practices as a condition of employment. Each radiation worker is responsible for performing work in a radiologically safe manner, consistent with the standards of conduct described in the Radiation Protection Program procedures.

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YNPS FSAR REV. 6/05 This management policy will continue to be communicated to all radiation workers through General Employee Training and will continue to be incorporated into all applicable procedures.

507.2.2 Administrative Policy YAEC will ensure that activities conducted during decommissioning will be managed by qualified individuals who will perform program operations in accordance with established procedures.

Radiological hazards will be monitored and evaluated on a routine basis to maintain radiation exposures and the release of radioactive materials to unrestricted areas as far below specified limits as is reasonably achievable. Each element of the Radiation Protection Program will be defined and implemented using written procedures. Radiation protection training will be provided to all occupationally exposed individuals to ensure that they understand and accept their responsibility to follow procedures and to maintain their individual radiation dose as low as is reasonably achievable.

YAEC project management will ensure that work specifications, designs, and work packages involving potential radiation exposure or handling of radioactive materials incorporate effective radiological controls. Project supervisors will incorporate radiation protection considerations in the work activities under their control.

Radiation protection records will be prepared and maintained using high standards of accuracy, traceability, and legibility to meet the requirements of regulatory agencies and company procedures.

507.2.3 ALARA Policy YAEC is committed to maintain an ALARA Program that is implemented based on guidance provided in Regulatory Guides 8.8 and 8.10 (References 507-2 and 507-3). All activities at YNPS involving radiation and radioactive materials will be conducted such that exposure of employees, contractors, and the general public to radiation is maintained as low as is reasonably achievable.

This determination will consider the current state of technology and the economics of improvements in relation to their benefit (i.e., reduction of dose).

Appropriate ALARA considerations will be incorporated into decommissioning activity planning and design activities at an early stage to allow full consideration of reasonable alternatives. Final plant modifications also will be reviewed to ensure that ALARA was incorporated into the activities.

YNPS management will establish and monitor progress towards specific goals and objectives for the YNPS decommissioning ALARA Program.

507.2.4 Regulatory Compliance Policy YAEC is committed to maintain the Radiation Protection Program in compliance with the requirements of the Code of Federal Regulations and, to the extent practical, information contained in industry standards, regulatory guides, and other guidance documents. The Radiation Protection Program is assessed against all new regulatory guidance and modified as necessary. YAEC implemented the revised 1 OCFR Part 20 on January 1, 1994.

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YNPS FSAR REV. 6105 507.2.5 Waste Minimization and Disposal Policy YAEC will ensure appropriate processing, packaging and monitoring of solid, liquid and gaseous wastes during decommissioning by continuing to implement the Process Control Program, the Radiological Effluent Control Program and the Radiological Environmental Monitoring Program.

These programs will be maintained in strict compliance with the YDQAP and Off-Site Dose Calculation Manual requirements to meet the requirements of 1 OCFR Parts 20, 50, 61, and 71; 49CFR; state regulations; disposal site requirements; and any other applicable requirements.

Implementing procedures will be maintained for the classification, treatment, packaging, and shipment of radioactive material.

YNPS will continue to implement and enforce a Radioactive Waste Reduction Program to minimize the generation of radioactive wastes. YNPS management will monitor waste minimization efforts for decommissioning. All decommissioning personnel will receive training in the applicable procedures and practices to minimize the generation of radioactive waste.

507.2.6 Respiratory Protection Policy YAEC is committed to minimizing the inhalation of air contaminated with dusts, mists, fumes, gases, vapors, and radionuclides at YNPS. The primary means of achieving this goal is to prevent or mitigate the hazardous condition at the source. Every reasonable effort will be made to achieve this objective by using engineering controls, including process modification, containment, and ventilation techniques. Respiratory protection equipment usage will be considered after engineering controls have been evaluated. Use of respiratory protection equipment for radionuclide inhalation reduction will be consistent with the goal of maintaining the total effective dose equivalent to personnel as low as is reasonably achievable.

The existing respiratory protection program will continue to be implemented and maintained in accordance with 1 OCFR Part 20 and other applicable regulatory guidance.

507.3 Decommissioning Exposure Projections This information has been extracted, updated and relocated to the PSDAR within the FSAR.

507.4 ALARA Program 507.4.1 General Program Description All activities at YNPS involving radiation and radioactive material will be conducted such that the radiation doses received by employees, contractors, and the general public are maintained as low as is reasonably achievable. This determination will consider the current state of technology and the economics of improvements in relation to the benefit (i.e., reduction of dose). The YNPS ALARA Program is implemented in an operating plant procedure, and is based on guidance provided in Regulatory Guides 8.8 and 8.10.

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YNPS FSAR REV. 6/05 The following criteria will continue to be used to determine when an activity requires a specific ALARA review:

The dose for the total completion of the activity exceeds 1 person-rem.

The whole body dose rate field exceeds 5 R/hr for an activity other than a surveillance or inspection.

The loose surface contamination exceeds 500,000 dpm/l OOcm2 (beta/gamma) for an activity other than surveillance or inspection.

The airborne radioactivity concentration exceeds 40 DAC as a result of an activity or in the area of an activity.

The work involves a planned special exposure.

The Radiation protection staff or ALARA Coordinator request an ALARA review.

In addition to these criteria, ALARA considerations will be incorporated into decommissioning activity planning and design activities at an early stage to allow full consideration of reasonable alternatives. Plant modifications also will be reviewed to ensure that ALARA was incorporated into the activities. YNPS management will establish and monitor progress towards specific goals and objectives for the YNPS decommissioning ALARA Program.

507.4.2 ALARA Program Organization and Responsibilities The Radiation Protection Manager coordinates the ALARA Program scope and implementation.

The ALARA Coordinator is responsible for completing the ALARA reviews. However, the actual implementation of specific ALARA actions, as incorporated into daily work activities, is the responsibility of each individual manager, supervisor, and worker.

An administrative plant procedure defines the responsibilities and authorities of the ALARA Committee. The primary responsibility of the ALARA Committee is to advise the Vice President of Decommissioning on matters related to exposure and contamination reduction. The committee will review the following:

All plant decommissioning activities, maintenance activities, and modifications that have an estimated dose expenditure in excess of 10 person-rem.

All ALARA post-job reviews for activities exceeding 10 person-rem.

Any individual's dose which exceeds the quarterly or annual administrative limits established for the individual.

Any exposure or contamination issue of concern requested by committee members.

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YNPS FSAR REV. 6/05 The ALARA Committee is chaired by the Radiation Protection Manager. Other members of the committee include other designated managers and supervisors involved in decommissioning activities. Members are given the appropriate authority and responsibility necessary to implement an effective ALARA Program.

507.4.3 ALARA Training and Instruction Commitment to the principles of the ALARA Program will be reflected in all radiation protection l training. Training courses will be evaluated by the Radiation Protection Manager to ensure that ALARA principles are incorporated into lesson plans.

507.5 Administrative Dose Control Administrative radiation dose controls will continue to be implemented during decommissioning.

Dose controls ensure the following:

Personnel do not exceed regulatory dose limits.

Equitable distribution of dose among available qualified workers.

Collective dose to workers is as low as is reasonably achievable.

Administrative plant procedures implement the program to control and limit external and internal radiation exposure. These procedures include the following elements:

A summary of administrative and regulatory dose limits.

A description of Radiation Control Area Postings and Controls.

A description of radiological survey data available on-site.

Instructions on the use and care of dosimetry.

Instructions on the conduct of work in the Radiation Control Area.

Instructions on personal monitoring for contamination.

Personnel dose reports are prepared weekly with more frequent reports during periods of high work activity. The reports are distributed to each plant department and are posted in the Radiation Control Area Control Point. Decommissioning supervisors are responsible for reviewing the dose reports and planning high dose activities such that the dose is distributed as evenly as possible among available qualified personnel.

507.6 Radiation Work Permits Radiation Work Permits will continue to be used to administratively control personnel entering or working in areas that have, or potentially have, radiological hazards present. The primary function of the Radiation Work Permit is to allow authorized activities to be conducted in radiologically controlled areas using safe and radiologically sound practices. The permit documents the work description, the worker names, the radiological conditions, and the radiological precautions and requirements. The permit also is an element of the ALARA program where it is used to screen activities to determine if a specific ALARA review is necessary and to track personnel and job exposure data.

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YNPS FSAR REV. 6/05 An administrative plant procedure presents the requirements for requesting, using and terminating a Radiation Work Permit. An additional operating plant procedure presents the process used by the radiation protection staff to prepare, issue and monitor a Radiation Work Permit.

Radiation Work Permits are required for the following activities:

Entry into a high or very high radiation area, an airborne radioactivity area, or any area posted with a sign stating that a Radiation Work Permit is required.

Maintenance on or inspections of equipment with loose surface contamination levels in excess of 10,000 dpm/l 00 cm2 (beta-gamma).

When prudent radiation protection practices warrant the use of a Radiation Work Permit, as determined by the Radiation Protection Manager.

507.7 Area Definitions and Postings An operating plant procedure describes the requirements for radiological postings at the entrance and boundaries of radiologically controlled areas. The purpose of the postings is to advise workers of radiological hazards that may be encountered in the areas. Informational postings may also be used to provide additional radiological instructions to workers. Each worker is responsible for the observance of the area postings and compliance with the indicated requirements.

507.8 External Dosimetry 507.8.1 General Considerations External radiation dose will be monitored through the use of thermoluminescent dosimeters (TLD),

direct reading dosimeters, and digital alarming dosimeters. The official record of external dose from beta and gamma radiation normally will be obtained from the TLD readings. Direct reading or digital alarming dosimeters will be used as a means for tracking dose between TLD processing and may also be used as a back-up to the TLDs. TLDs will be processed at a frequency that ensures personnel dose limits are not exceeded.

507.8.2 Monitoring Whole Body Dose All decommissioning workers are required to wear external radiation monitoring devices whenever they enter the Radiation Control Area. Radiation workers are instructed to read the direct reading and digital alarming dosimeters prior to use and periodically during the work activity. The TLD and the direct reading or digital alarming dosimeters are worn typically in close proximity to each other on the trunk of the body between the neck and waist. Under certain conditions, where the chest or trunk may not be the location of highest whole body dose, dosimetry devices may be relocated.

Multiple whole body dosimetry may be used if work is to be performed in a nonuniform radiation field in which the dose to a portion of the body that is exposed to the highest dose source cannot easily be determined. In these cases, multiple sets of dosimeters will be worn on those portions of the body expected to receive the highest dose. Guidance for conducting the evaluation and criteria 507-6

YNPS FSAR REV. 6/05 for determining when multiple dosimetry is required is provided in an operating and departmental plant procedure.

Dosimetry requirements are specified on the Radiation Work Permit.

507.8.3 Dosimetry Quality Control Periodic quality assurance checks of dosimetry will be conducted by exposing whole body, extremity and environmental dosimetry to known radiation doses and sending them to a qualified Environmental Laboratory for processing. A departmental plant procedure implements the quality control program for plant dosimetry. Discrepancies between the expected exposure and the laboratory results will be reconciled and documented. The Environmental Laboratory is accredited by the National Institute of Standards and Technology (NIST) under the National Voluntary Laboratory Accreditation Program (NVLAP) for dosimetry.

507.9 Internal Dosimetry Control and Monitoring 507.9.1 General Considerations Internal radiation dose inherently is more difficult to measure than external radiation dose, but it is generally much easier to prevent. Therefore, major emphasis is placed on preventing internal radiation dose, as long as it is consistent with the goal of keeping total effective dose as low as is reasonably achievable.

The primary methods for controlling the intake of radioactive material into the body is identifying and minimizing the sources of airborne radioactivity and applying engineering controls to reduce airborne radioactivity concentrations. The use of respiratory protection will be used after the primary methods have been implemented to the extent practicable.

An administrative plant procedure describes the program that is implemented to monitor potential internal radiation exposures.

507.9.2 Bioassay Program Whole body counting (in vivo) is the primary method that is used to determine the identity and quantity of gamma emitting radionuclides present in the body. Personnel will receive a whole body count after a suspected intake of radioactive materials. Radiation protection implementing procedures provide guidance on whole body counter operation, calibration, and quality control.

Indirect bioassay (in vitro) measurements will be made, as necessary, to monitor for alpha and beta emitting radionuclides and to provide data for calculation/determination of internal dose. This method of bioassay will typically be used only for radionuclides which cannot be determined by whole body counting or when additional information on an intake is required. Radiation protection implementing procedures include criteria for indirect bioassay and methods for data analysis and interpretation.

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YNPS FSAR REV. 6/05 An administrative plant procedure is used to implement the bioassay program at YNPS.

507.10 Respiratory Protection Program A Respiratory Protection Program will continue to be maintained in accordance with 1 OCFR Part 20 and other applicable regulatory guidance. The primary means of providing respiratory protection is to prevent or mitigate the hazardous condition at the source. Every reasonable effort will be made to achieve this objective by using engineering controls, including process modification, containment, and ventilation techniques. Respiratory protection equipment usage will be considered after engineering controls have been evaluated. Use of respiratory protection equipment will be consistent with the goal of maintaining the total effective dose to personnel as low as is reasonably achievable.

An administrative plant procedure is used to implement the Radiological Respiratory Protection Program at YNPS.

507.11 Radioactive Material Controls The Radiation Protection Program establishes radioactive material controls that ensure the following:

Prevention of inadvertent radioactive material release to uncontrolled areas.

Assurance that personnel are not exposed inadvertently to radiation from radioactive materials.

Minimization of the amount of radioactive waste material generated during decommissioning.

Radioactive material is defined as any of the following: 1) material activated by YNPS reactor operation, 2) material contaminated from the operation or decommissioning of YNPS, or 3) licensed material procured and used to support the operation and decommissioning of YNPS.

All materials leaving the Radiation Control Area, and subsequently the YNPS site, will be surveyed to ensure that radioactive materials are not inadvertently discharged from the facility. An administrative plant procedure will be used to ensure that all potentially radioactive or contaminated items removed from the Radiation Control Area or the YNPS site are surveyed. This procedure was written to incorporate the guidance presented in NRC Circular No. 81-07 and NRC Information Notice No. 85-92 (References 507-5 and 507-6). The following survey methods (instrumentation is typical) will be used:

Materials and Equipment - Direct frisking with a portable Geiger-Mueller, gas proportional and scintillation detectors (as appropriate). Large quantities of materials with a low probability of being contaminated (e.g., secondary side plant components) may be released through the analysis of representative samples (as appropriate for the materials being monitored) combined with aggregated waste surveys.

Smear Samples - Analysis with a Geiger-Mueller, gas proportional, scintillation, and solid state detector (as appropriate).

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YNPS FSAR REV. 6/05 Bulk Liquids or Solids - Analysis with high resolution gamma spectrometry system to the environmental lower limit of detection. Large quantities of materials (e.g., excavated soils) may be released through the analysis of representative samples combined with surveys of the aggregated wastes (e.g., dumpsters).

Materials will be released if no discernable plant-related activity is detected within the capability of the survey methods presented above. Any radioactive material that is shipped from the site is handled in accordance with an operating plant procedure which ensures compliance with NRC and Department of Transportation requirements.

An administrative plant procedure provides instructions regarding the proper handling and storage of contaminated tools and equipment. This procedure ensures that tools and equipment are decontaminated promptly. Tools and equipment that are not fully decontaminated are stored in designated radioactive material storage areas. An operating plant procedure ensures that the areas where radioactive materials are stored are posted clearly.

507.12 Surveillance Routine radiological surveillances will continue to be conducted during decommissioning to monitor radiation sources, to determine radiological conditions, and to comply with the requirements of 1 OCFR Part 20. Surveys also will be performed to evaluate radiological conditions in support of decommissioning work activities. Plant procedures will be used to implement the surveys. These procedures specify the types of instrumentation, survey methods, and review requirements for each survey performed.

The final radiation survey that will be completed following decontamination and dismantlement activities is described in the PSDAR.

507.13 Instrumentation A sufficient inventory and variety of operable and calibrated portable, semi-portable, and fixed radiological instrumentation will be maintained on site to allow for effective measurement and control of radiation exposure and radioactive material and to provide back-up capability for inoperable equipment. Equipment will be capable of measuring the range of gamma, beta, alpha and neutron dose rates and radioactivity concentrations expected. Instrumentation will be calibrated at prescribed intervals or prior to use against certified equipment having known valid relationships to nationally recognized standards. An administrative plant procedure will be used to control the use of radiation protection instrumentation.

507.14 Review and Audit To ensure the Radiation Protection Program is effectively implemented and maintained, an organized system of reviews and audits will continue to be implemented during decommissioning in accordance with the Quality Assurance Program presented in Section 514.

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YNPS FSAR REV. 6/05 507.15 Radiation Protection Program Performance Analysis An administrative plant procedure will be used during decommissioning to evaluate the causes of unacceptable performance, to initiate corrective actions, and to trend overall performance. This process will be used to address the following types of deficiencies:

Work activities generating unnecessary radiation exposure or contamination.

Procedural actions resulting in unacceptable radiological performance.

Unacceptable radiological work practices resulting in personnel contamination, spread of contamination, or unnecessary radiation exposure.

Activities resulting in unnecessary generation of liquid or solid radioactive waste.

Activities violating Radiation Work Permit instructions, postings, signs, and radiation protection implementing procedures.

Incidents that result in more serious radiological events will be reported using the same administrative plant procedure. These events include overexposures, large intakes of radioactive material, unplanned radioactive releases and significant radioactive spills. This procedure ensures that immediate and written notifications are made in accordance with regulatory requirements.

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YNPS FSAR REV. 6/05 REFERENCES 507-1 NYR 91-097, Systematic Assessment of Licensee Performance (SALP) Final Report for Yankee Nuclear Power Station for the Period August 1, 1989 to January 15, 1991 (50-29/89-99), T. T. Martin (USNRC) to A. C. Kadak, May 20, 1991.

507-2 Regulatory Guide 8.8, Information Relevant to Ensuring That Occupational Radiation Exposure at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable.

507-3 Regulatory Guide 8.10, Operating Philosophy For Maintaining Occupational Radiation Exposure As Low As Is Reasonably Achievable.

507-4 Regulatory Guide 1.8, Personnel Selection and Training.

507-5 NRC IE Circular No. 81-07: "Control of Radioactively Contaminated Material,"

May 14,1981.

507-6 NRC IE Information Notice No. 85-92: "Surveys of Wastes Before Disposal From Nuclear Reactor Facilities," December 2, 1985.

507-7 Off-Site Dose Calculation Manual.

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YNPS FSAR REV. 6/05 508 RADIOACTIVE WASTE MANAGEMENT YNPS decommissioning requires handling of a large volume of radioactive materials to reduce residual radioactivity to a level permitting release of the site for unrestricted use and termination of license. Materials that are not decontaminated and released will be processed as radioactive waste.

This section of the FSAR presents the programs used to control the processing of solid, liquid and gaseous radioactive waste.

YAEC will continue to ensure appropriate processing, packaging and monitoring of solid, liquid and gaseous wastes during decommissioning by implementing the Radiation Protection procedures, the Process Control Program, the Radioactive Effluent Controls Program and the Radiological Environmental Monitoring Program. These programs will be maintained in compliance with YDQAP requirements to meet federal and state regulations, disposal site requirements, and any other applicable requirements. The YNPS radioactive waste management program is implemented through the Radiation Protection Program (Section 507). Implementing procedures will be used to control the classification, treatment, packaging and shipment of radioactive material.

508.1 Solid Radioactive Waste Processing Solid radioactive waste handling at YNPS is divided into three phases: packaging, on-site storage awaiting shipment and shipment. Each of these phases will be implemented in strict compliance with applicable federal, state, disposal site and offsite processing facility requirements. In addition, all waste processing activities will be completed in accordance with the requirements of the Yankee Decommissioning Quality Assurance Program (Section 514), PCP and ODCM.

YNPS decommissioning solid waste is comprised of both high and low level radioactive waste. A portion of the reactor vessel internal components (e.g., core baffle) have radionuclide concentrations in excess of the 10CFR Part 61 Class C limits. These materials are not generally acceptable for near-surface disposal and have been classified as high level radioactive waste. High level radioactive waste will be stored with the fuel and will be shipped to an appropriate disposal facility after a location becomes available.

508.1.1 Solid Radioactive Waste Packaging Radioactive waste packaging at YNPS will be performed in areas that minimize radiation exposure to personnel, control the spread of contamination, and are adequate for packaging activities.

Temporary facilities, that meet the requirements above, may be constructed for waste packaging.

Radioactive waste packaging operations will be implemented through plant procedures that ensure the following:

A Radiation Work Permit has been issued for handling radioactive materials.

Specific packaging requirements are identified.

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YNPS FSAR REV. 6105

Containers are surveyed for external contamination.

Personnel will conduct inspection activities in accordance with the requirements of the Yankee Decommissioning Quality Assurance Program (Section 514).

Waste packages will meet the requirements for transportation and disposal or processing for each decommissioning waste stream. Examples of the waste containers that may be used are drums, boxes, liners, high integrity containers, sea-land containers, shielded casks, and other specialty containers such as intermodals. Waste container selection will be determined by the size, weight, classification, and activity level of the material to be packaged. In all cases, packaging will comply with requirements specified by 49CFR, 1 OCFR Part 71, and the disposal or processing facility site criteria, as applicable.

A plant procedure provides instructions for determining the 10CFR Part 61 classification of low level radioactive waste. The procedure is used to determine the radionuclide content of a container through a combination of direct measurements and/or radiation shielding calculations.

Spent resins, filter media and other wetted wastes requiring stabilization will be processed in accordance with the Process Control Program (Reference 508-4), which is implemented through plant procedures. Whenever possible, stabilization will be completed inside the disposal package or liner to minimize additional waste handling prior to disposal.

508.1.2 Solid Radioactive Waste Storage Awaiting Shipment Solid radioactive waste awaiting shipment to a facility for disposal or processing normally will be stored in accordance with applicable regulations.

Supplemental shielding may be used to shield waste containers to assure that exposure rates are maintained as low as is reasonably achievable.

Stored waste will be inspected to assure that container integrity is maintained. The inspections will include visual inspections of a representative sample of containers for container integrity.

Radioactive materials will be stored such that in the event of a fire or an explosion, sufficient fire detection, response, and suppression capability in conjunction with spatial separation will ensure that any radiological release would be bounded by the accident analyses presented in Section 400.

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YNPS FSAR RE, V. 6/05 Large components awaiting shipment may be stored in the Yard Area prior to shipment. Precautions will be taken to ensure that the components are within barriers, as necessary, and adequately protected from on-site hazards (e.g., heavy load movement).

508.1.3 Solid Radioactive Waste Shipment Solid radioactive wastes will be shipped in compliance with applicable federal and state regulations.

Operating plant procedures present the requirements for radioactive materials shipment. These procedures ensure the following:

Appropriate labeling and documentation of shipping containers.

Quality assurance oversight of shipment preparation.

Verification of acceptable package physical condition and contamination levels.

Appropriate permits and licenses for waste shipment.

Notification of appropriate governmental agencies prior to shipment.

Most of the radioactive material and waste shipments will be completed over roads. Rail transportation may be used for shipments of demolition rubble. Material will be transferred by highway to an acceptable rail shipping facility. The routing of shipments may vary with weather and highway conditions. Additionally, local and state restrictions pertaining to radioactive material transport may affect some route selections. The carrier is responsible for selecting the appropriate route, which must conform to applicable federal, state, and local shipping requirements and be in accordance with Department of Transportation and NRC regulations.

508.2 Liquid Radioactive Waste Processing Contaminated water will be generated during YNPS decommissioning as a result of draining, decontamination and cutting processes. The contaminated liquids will be processed in a temporary system (e.g., ion exchange and filtration system, solidification system). All liquid radioactive waste will be processed in accordance with the Process Control Program, the Off-Site Dose Calculation Manual (ODCM), Yankee Decommissioning Quality Assurance Program (YDQAP) and plant procedures.

The Process Control Program presents the administrative and technical controls for the liquid radioactive waste system to assure that waste meets shipment and disposal facility requirements.

Liquid waste processing is monitored to assure safe operation, storage and disposal of waste to approved waste disposal sites. Liquids released from the site are monitored and controlled to ensure all releases of radioactivity to the environment are as low as is reasonably achievable. The Process Control Program is maintained in accordance with the YDQAP.

The YDQAP establishes two programs affecting radioactive liquids processing: Radioactive Effluent Controls Program, Radiological Environmental Monitoring Program. The Radioactive Effluent Controls Program conforms with 10CFR50.36a requirements to control radioactive effluents and to maintain dose to members of the public from radioactive effluents as low as is reasonably achievable. This program is presented in the ODCM (Reference 508-5) and implemented through several plant procedures. This program complies with the requirements of the YDQAP.

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YNPS FSAR REV. 6/05 The ODCM contains methodologies and parameters used in the following:

Calculation of off-site doses resulting from radioactive gaseous and liquid effluents.

Calculation of gaseous and liquid effluent monitoring alarm and trip setpoints.

Conduct of the Radiological Environmental Monitoring Program.

The ODCM forms the basis of plant procedures which document the off-site doses due to plant operation. The off-site dose calculations demonstrate compliance with the numerical guides for design controls of 10CFR Part 50, Appendix I. Several plant procedures implement the ODCM requirements.

508.3 Airborne Radioactive Waste Processing Dismantlement activities will be designed to ensure that airborne releases are monitored to the maximum extent practicable by implementing the following considerations during detailed planning of decommissioning activities in accordance with the requirements of the ODCM:

Local HEPA filtration systems will be used when activities could result in the release of significant radioactive particulates. Monitoring of the HEPA Filtration System exhaust will be performed when required by the ODCM. Prior to initial deployment and periodically, each HEPA System is tested against Dioctyl Phthalate (DOP) with an acceptance criterion of 99.95%

filter retention. This ensures that airborne particulate releases to the environment have been reduced to as low as is reasonably achievable.

Local supplemental air monitoring will be performed as necessary to support decommissioning activities.

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YNPS FSAR RE V. 6105 508.4 Mixed Waste YAEC has a nonradioactive waste management program to ensure compliance with all the federal and state hazardous waste regulatory requirements. The use of hazardous materials and the generation of hazardous wastes are controlled through the nonradioactive hazardous waste management program. This program is presented in Section 511.

YNPS currently has a quantity of mixed waste, including Polychlorinated Biphenyls (PCBs) contaminated material. There are varying quantities of oil and paints.

No chemicals or other substances are anticipated to be used during decommissioning operations that could become mixed waste. If mixed wastes are generated, they will be managed according to Subtitle C of the Resource Conservation and Recovery Act (RCRA) to the extent it is not inconsistent with NRC handling, storage, and transportation regulations.

Mixed wastes from the YNPS will be transported only by authorized and licensed transporters and shipped only to authorized and licensed facilities. If technology, resources, and approved processes are available, processes will be evaluated to render the mixed waste nonhazardous.

YAEC has identified the presence of solid PCBs in some paint coatings, primarily in the Radiation Control Area. As in the cases of radiologically contaminated lead paint, asbestos, and other hazardous materials, contaminated paint that contains PCBs will be managed according to all applicable federal and state regulations.

508.5 Radioactive Waste Minimization 508.5.1 Radioactive Waste Reduction Program YNPS will continue to implement and enforce a Radioactive Waste Reduction Program through implementation of an administrative plant procedure. YNPS management will monitor waste minimization practices for decommissioning. All decommissioning personnel will receive training in the applicable procedures and practices to minimize the generation of radioactive waste.

All radiation workers entering the Radiation Control Area will receive radiation worker training.

This training will include a review of work techniques that prevent unnecessary contamination of areas and equipment, practices for reuse of materials, and policies to prevent the unnecessary generation of mixed or radioactive wastes.

The following radioactive waste volume reduction methods will be incorporated into YNPS decommissioning activities:

Prevention of Waste - Unnecessary generation of radioactive wastes will be controlled by procedures established to prevent unnecessary packaging, tools, and equipment from entering Radiation Control Area.

Decontamination and Re-Use of Materials -Materials will be reused to the maximum extent practicable. Typical materials reused during the decommissioning include contaminated tools, 508-5

YNPS FSAR REV. 6105 equipment, and clothing. Contaminated tools and equipment storage areas will be maintained.

Protective clothing will be laundered offsite and made available for reuse.

Voids in disposal containers will be filled with other contaminated material to reduce the total volume of waste for disposal to the maximum extent practicable. This produces a better waste form, maximizing burial efficiency, and minimizing project cost, disposal site usage, and transportation risk.

508.5.2 On-Site Decontamination Methods On-site decontamination techniques will be used for processing and volume reduction of radioactive materials. The following are examples of decontamination methods that may be used during decommissioning:

Strippable Coatings - Strippable coatings may be used to lift radionuclides from contaminated surfaces. A strippable coating is typically applied in a manner similar to spray painting a surface. Additives in the coating are designed to attract and to combine chemically with radioactive contaminants. Once the coating is dry, the contaminant is locked in the coating. The dried coating is easily removed from the surface, stripping the film containing the contamination.

The stripped film is packaged and processed as a solid waste. Strippable coating may also be used to protect surfaces from becoming contaminated.

Chemical or Solvent Decontamination - Chemical and solvent decontamination methods remove contamination by creating a solution of the radionuclides and the solvent used. This type of decontamination may be difficult to control because of the aggressiveness of the chemicals and solvents. However, the systems decontaminated at YNPS will not be returned to service after decontamination, therefore, excessive metal wastage is not significant. Chemicals used for decontamination will be evaluated for hazardous constituents. If the chemical could become a listed or characteristic hazardous mixed waste it will not be used.

Dry Abrasive Impingement - Dry abrasive impingement is effective for removing heavy or tightly adhering oxide films. Examples of this technology are sandblasting and dry ice blasting.

Wipe down with Simple Green for Rad/PCB surfaces.

Concrete scabbling techniques.

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YNPS FSAR REV. 6105

Water Washing - High pressure water washing is effective for removing surface contamination and for sluicing sludge fromn tanks. Barriers must be established to ensure that wash water is collected and processed in the site liquid waste processing system.

Vacuum Cleaning - HEPA filtered vacuum cleaners may be used in areas of high loose surface contamination.

508.5.3 Off-Site Radioactive Materials Processing Several off-site radioactive materials processing options are currently available. The current decommissioning cost estimate (See PSDAR Section) assumes that the availability of processing alternatives is limited and that all significantly contaminated and activated materials are sent to a low level radioactive waste disposal facility. However, all processing alternatives will be evaluated during decommissioning to determine the most effective processing of radioactive materials.

The following are examples of off-site processing alternatives for radioactive materials removed during YNPS decommissioning:

Decontamination - Decontamination facilities provide a wide range of decontamination technologies at centralized locations. The variety allows selection of appropriate technologies for each component of the decommissioning waste stream.

Volume Reduction - Volume reduction facilities provide various processes (e.g., sorting, super-compaction) to reduce the volume of material that is sent to the disposal facility. This processing alternative is attractive for asbestos compaction which requires specialized containment during processing.

Incineration - Incineration facilities safely incinerate materials resulting in very high volume reduction rates. Appropriate materials may include paper, certain plastics, lubricating oils, and solvents.

Metal Melting - Metal melting materials process low specific activity metals. The processed metal is recycled to the nuclear industry as shielding and potentially in the future as cask liners and fuel canisters.

l Waste packages will be transported to off-site facilities primarily in intermodals, rolloffs, and sea-land containers selected to meet transportation and receipt requirements of the off-site processing facility. Voids in transport containers are not a significant concern. However, efficient management of transportation resources is an important consideration to minimize the total number of shipments and decommissioning costs.

Radioactive material control and accountability procedures will be implemented to track material originating from YNPS during receipt, sorting, processing, and packaging for disposal. Off-site processing facilities will be selected that provide adequate radioactive material control and accountability procedures.

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YNPS FSAR REV. 6/05 508.6 Decommissioning Radioactive Waste Projections This information has been extracted, updated and relocated to the PSDAR within the FSAR.

508-8

YNPS FSAR REV. 6105 REFERENCES 508-1 58-FR-34947, Notification of Spent Fuel Management and Funding Plans By Licensees of Prematurely Shut Down Power Reactors, June 30, 1993.

508-2 YRP 435/92, Spent Nuclear Fuel Storage Study Report and Recommendations, B. W. Holmgren, J. M. Buchheit, R. A. Mellor to J. K. Thayer, October 9, 1992.

508-3 YRP 303/93, Impact of Wet Spent Fuel Storage on Decommissioning, P. A. Rainey to R. A. Mellor, July 15,1993.

508-4 Yankee Nuclear Power Station Process Control Program.

508-5 Yankee Nuclear Power Station Off-site Dose Calculation Manual.

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YNPS FSAR REV. 6105 509 TESTS I Tests are performed to ensure the continuous, safe, and efficient operation of equipment necessary to support the defueled condition. The various types of tests are those conducted in accordance with either YDQAP mandated requirements, governing applicable engineering practices and standards (ASME, IEEE, ISA, etc.), or those based on good engineering judgement and operational experience.

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YNPS FSAR RE V. 6/05 510 OCCUPATIONAL SAFETY PROGRAM 510.1 Introduction This section provides an overview of the YAEC Occupational Safety Program as provided in YNPS Health and Safety Manual, applicable plant procedures, and approved contractor programs.

510.2 Management Policy Statement YAEC and its management are committed to the safe decommissioning of YNPS. The primary objective of the Occupational Safety Program is to protect workers and visitors from industrial hazards that have the potential of developing during decommissioning activities. YAEC and its contractors will provide sufficient qualified staff, facilities, and equipment to perform decommissioning in a safe and effective manner. YAEC is committed to compliance with federal and state requirements and to the guidance provided through industry standards and good work practices. The effective implementation of the Occupational Safety Program is the responsibility of all decommissioning personnel.

510.3 Deleted 510.4 YAEC Occupational Safety Program The YAEC Occupational Safety Program was developed to establish and maintain a safe work place for YAEC workers, contractors, and visitors. The program provides guidelines and procedures to be used to reduce industrial hazards and risks.

The YNPS Health and Safety Manual provides guidelines and requirements which will be incorporated into the detailed decommissioning planning process.

The following areas are discussed in the manual:

General Information

Personnel Protection and Safety Equipment

Work Access and Working Surfaces

Material Handling and Storage

Hazardous Substances and Materials

Hand and Portable Powered Tools and Equipment

Welding, Cutting, and Brazing

Electrical Safety

Fire Prevention

Motor Vehicles

Excavation Additional safety guidelines and instructions are included in procedures.

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YNPS FSAR REV. 6/05 510.5 Safety Training and Meetings Safety training is conducted as part of the General Employee Training process, during routine safety meetings, and for job-specific purposes. The safety meetings focus on current safety issues and events as well as providing a forum for workers to ask questions and provide feedback.

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YNPS FSAR RE V. 6105 511 NONRADIOACTIVE WASTE MANAGEMENT 511.1 Introduction This section provides an overview of the YAEC Non-radioactive Waste Management Program and applicable plant procedures.

Site materials that are routinely handled as hazardous waste requiring disposal include solvents, oils, and absorbent materials used with these items. Non-routine wastes such as paint containing Polychlorinated Biphenyls (PCBs), mercury, paint, and battery acid are also handled and disposed of through this program. In addition, wastes handled for recycling include fluorescent light bulbs, incandescent light bulbs and batteries which are collected and transported offsite.

511.2 Management Policy Statement YAEC and its management are committed to the safe decommissioning of YNPS. The primary objective of the YAEC Nonradioactive Waste Management Program is to protect site workers, visitors, and the environment from the potential effects of hazardous materials and waste. YAEC is committed to strict compliance with all federal and state hazardous waste handling and disposal requirements.

511.3 Hazardous Material Management YAEC is required by the OSHA Hazard Communication Standard (29CFR1910.1200) to provide information to its employees and contractors concerning the hazardous substances to which they may be exposed. Administrative plant procedures were implemented to meet these requirements.

General Employee Training was revised to apprise employees and contractors of hazardous materials used at YNPS.

511.4 Hazardous Waste Management The YAEC Non-radioactive Waste Management Program was established to assure compliance with all the federal and state hazardous waste regulatory requirements. Non-radioactive hazardous wastes from YNPS are transported only by authorized and licensed transporters and shipped only to authorized and licensed disposal facilities.

The hazardous waste management program is implemented through an administrative plant procedure. This procedure provides direction for the handling, temporary storage and preparation for shipment of nonradioactive hazardous waste. Routine preventive and emergency response procedures have been developed for precluding and containing hazardous material incidents.

511.4.1 Above-Ground Fuel Oil Storage Tanks There are two 275 gallon above-ground oil storage tanks at YNPS: the two Security Diesel Generator fuel oil tanks (TK-76-1 and TK-76-2). In addition, there are several portable mobile and stationary fuel tanks used by the demolition contractor.

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YNPS FSAR RE V. 6105 These portable tanks will remain in service throughout the dismantlement phase. When no longer required, the tanks will be removed from the site by an authorized and licensed contractor.

511.4.2 Deleted 511.4.3 Waste Oil Uncontaminated waste oil is and will continue to be stored in drums in the Hazardous Waste Storage Area until disposal.

511.4.4 Deleted 511.4.5 Deleted 511.4.6 Asbestos Containing Materials Asbestos Containing Material (ACM) has been identified on many plant systems and in most areas and buildings. Historically, asbestos insulating material was replaced with non-asbestos material as a part of maintenance activities that required insulation removal and was labeled accordingly.

Decommissioning activities to date have removed the majority of ACM; however, additional decommissioning will remove the remaining systems originally covered with asbestos insulating material. Minor quantities of non-insulation asbestos containing materials, in the form of gaskets and packing, remain present in numerous systems at the plant. A licensed contractor has performed an asbestos survey of all remaining and accessible building materials.

Asbestos material will be removed by a licensed contractor prior to the start of dismantlement activities. Insulating material will be considered to be asbestos material unless marked "NON-ASBESTOS" or subsequent testing determines it not to be asbestos. All asbestos containing materials will be removed and processed in accordance with an administrative plant procedure to ensure compliance with federal and state regulations.

Radiologically contaminated asbestos containing material and non-asbestos material will be disposed of in accordance with the requirements given in Section 508.

511.4.7 Lead-Based and PCB-Containing Paints Lead-based and PCB-containing paints were used at YNPS to coat many steel components, concrete structures, and carbon steel piping. During the operating life of the paint, some of these paints have been covered with several coats of non-lead or non-PCB based paint.

Controls of the lead-based paint identification and removal process have been implemented to control workers exposure and ensure proper handling of lead materials. Lead-based paints on non-recyclable components (e.g., concrete) are removed, processed, and disposed of by qualified personnel.

Controls of the PCB-containing paint identification, removal, and disposal process have also been implemented. The controls are part of plant programs to manage workers exposure and ensure proper handling of PCB waste. An Alternative Method of Disposal Approval (AMDA) issued by the 511-2

YNPS FSAR REV. 6/05 U.S. Environmental Protection Agency authorizes management of a specific category of PCB wastes designated as "PCB Bulk Product" wastes.

Work associated with both lead and PCB paints are controlled in accordance with administrative plant procedures.

511.5 Sampling and Remedial Actions Site structures and environs are evaluated for non-radioactive hazardous materials by appropriate sampling and analytical protocols. Remedial actions are implemented as necessary to meet all federal, state, and local environmental quality requirements.

511.6 Industrial Waste Management The final dismantlement of YNPS will require the handling and disposal of system and building wastes. These wastes will include non-hazardous materials that were never radiologically contaminated, have been decontaminated to meet release criteria, or material that contains asbestos or other hazardous materials. Non-radioactive non-hazardous wastes are expected to include the following:

System Piping and Components (e.g., pumps, valves, tanks, non-asbestos insulation, heat exchanges, and supports).

Duct-Work and Associated Equipment (e.g., ducts, fans, filters, and supports).

Electrical Systems and Equipment (e.g., cables and trays, conduit, motor control centers, generators, motors, and panels).

Buildings and Structures (e.g., concrete, structural steel, roofing materials, siding, doors, and windows).

The materials presented above will be processed in accordance with the rules and regulations governing the disposal of non-radioactive, non-hazardous wastes.

511.7 Training All personnel involved in the handling of hazardous materials and wastes receive initial and annual training. This training includes information on the types of hazardous materials and wastes, handling precautions, temporary storage locations, and emergency response procedures.

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YNPS FSAR RtEV. 6/05 512 SECURITY PLAN 512.1 General After the decision to permanently cease YNPS power operation, YAEC completed a comprehensive review of the security requirements for YNPS. The review incorporated the following:

Re-evaluation of the design basis threat matrix to determine security needs for a permanently defueled condition.

Evaluation of the Vital and Access Control Areas to determine the appropriate level of protection for the permanently defueled condition. This analysis reviewed existing safety analyses, the potential consequences of radiological sabotage events and the potential consequences of fuel storage events. The evaluation indicated that nuclear security could be focussed on spent fuel storage.

Physical survey of the site to determine physical modifications needed to implement a reduced protected zone. Implementation of the recommendations of this survey allowed reduction of the protected area and the implementation of an industrial security zone for most of the plant site.

Based on this evaluation, the security plan was modified creating a YNPS Defueled Security Plan (References 512-1 and 512-2). The NRC approved the plan and on December 18, 1992, the plan was implemented by YNPS (Reference 512-3). The Defueled Security Plan reduced the protected area boundary to the Spent Fuel Pit Building outside wall. The area that was within the original protected area boundary (with the exception of the spent fuel complex) was reclassified as an industrial security area. An administrative plant procedure identifies individuals responsible for the management of the security program and for the supervision of the security force.

The NRC has inspected the implementation of the Defueled Security Plan to ensure that the changes had been implemented satisfactorily (Reference 512-4). The inspection concluded that the plan, "as implemented, was directed toward the protection of public health and safety."

To complete the plant decommissioning process, the spent nuclear fuel was removed from the Spent Fuel Pit. YAEC constructed an Independent Spent Fuel Storage Installation (ISFSI) onsite and transferred the spent fuel to this location for interim storage until the Department of Energy is ready to accept this fuel. Since YAEC is storing spent fuel under the general license provisions of IOCFR Part 72, Subpart K, YNPS developed an ISFSI Security Plan based on the requirements of 1 OCFR73.55 with exemptions as appropriate. The YNPS Security Plan amendment submittal (Reference 512-5) requested NRC approval of the ISFSI Security Plan for YNPS and was approved on March 13, 2002 (Reference 5 12-6). With the transfer of all spent fuel to the YNPS ISFSI, the protected area associated with the Spent Fuel Pit Building has been eliminated.

512.2 Site Access Control YNPS access control requirements are presented in administrative plant procedures. These procedures present requirements for general site access and access to the protected area (ISFSI).

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YNPS FSAR REV. 6105 YNPS will maintain a Fitness For Duty program for personnel granted unescorted access to either the general site area or the protected area. The objective of the program is to provide a drug and alcohol free work environment. The program is essential to maintaining the health and safety of I those working on site.

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YNPS FSAR REV. 6/05 REFERENCES 512-1 BYR 92-077, Defueled Security and Training and Qualification Plans, J. K. Thayer to M. B. Fairtile (USNRC), August 11, 1992.

512-2 BYR 92-102, Defueled Security and Training and Qualification Plans, J. K. Thayer to M. B. Fairtile (USNRC), October 22, 1992.

512-3 NYR 92-194, Exemptions From Certain Requirements of 10CFR73.55 For The Yankee Nuclear Power Station (YNPS) (TAC No. M84267), M. B. Fairtile to J. M. Grant, November 24, 1992.

512-4 NYR 93-027, NRC Inspection No. 50-29/93-03, J. H. Joyner (USNRC) to J. K. Thayer, March26, 1993.

512-5 BYR 2000-068, Proposed Amendement to YNPS Security Plan, B. Wood (YAEC) to NRC, dated October 12, 2000.

512-6 NYR 2002-023, YNPS - Issuance of Amendment and Exemption from Requirements of 1 OCFR73.55 (TAC No. MB0209), J. Hickman (NRC) to J. Kay (YAEC), dated March 13, 2002.

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YNPS FSAR REV. 6105 513 FIRE PROTECTION 513.1 Program Description On August 20, 1992, the NRC approved a Technical Specification change to remove the fire protection Technical Specifications (Reference 513-1). The fire protection Technical Specifications were replaced with a set of administrative controls. The Fire Protection Technical Requirements Manual (Reference 513-2) was developed to set forth the operational and surveillance requirements of the Fire Protection Plan as approved by NRC Safety Evaluation Reports dated March 15, 1979, and as supplemented October 1, 1980 and August 27, 1986.

The Fire Protection Plan is based on defense in depth. The plan will maintain the following features, as appropriate, during decommissioning:

Fire detection equipment and systems.

Personnel training and qualification program.

Fire Protection Program procedures.

Control of transient combustible materials and ignition sources, including limitations on inflammable gases as described in Section 406.

On April 18, 2003, the NRC approved a change to the Fire Protection License condition to reflect that fuel is no longer stored in the Spent Fuel Pit. The condition now allows changes to the Fire Protection Program without NRC approval if the program's effectiveness, taking into account the decommissioning plant conditions and activities, is not reduced (Reference 513-3).

Decommissioning activities will be reviewed during the safe storage and dismantlement periods to ensure that the appropriate level of fire protection is being implemented. In addition, fire protection requirements will be evaluated during detailed planning of decommissioning activities.

Administrative plant procedures implement the Fire Protection Plan:

NOTE:

On July 29,2004 the remaining requirements of the Fire Protection Technical Requirements Manual were incorporated into Site Administrative procedures and the manual cancelled.

Defines the Fire Protection Plan for the YNPS, describes the organization and structure of the fire protection program, and defines the administrative and functional responsibilities of assigned personnel.

Establishes the administrative controls for materials and events that could create potential fire hazards. Materials and processes, which represent a potential fire hazard, are controlled to minimize the possibility of a fire and the effect of a fire on the following:

The release of radioactive material to the environment.

Plant personnel safety.

Balance of the facility.

Establishes housekeeping requirements, provides guidelines for hot work, designates fire areas and defines the process for control of combustibles.

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YNPS FSAR REV. 6105 513.2 Fire Protection Program Implementation Continued operability of required fire detection systems ensures that adequate warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early stages. On July 6, 2004 the last of the active fire protection systems (ie. fire pumps, hydrant loop, etc.) were removed from service. Based on the advanced state of decommissioning activities, the potential for a serious radiological release that could affect the public or the environment as result of a fire had been eliminated. The onsite incipient fire brigade (utilizing fire extinguishers) is adequate to protect the facility until off-site assistance provided by the Town of Rowe arrives. The Rowe Fire Department is available within a 30-minute response time.

Fire detection capabilities will remain in place to protect those facilities associated with the ISFSI.

In addition, spatial separation, and the grouping by activity level will be used to minimize the radiological consequences of a fire in radioactive materials awaiting shipment from the site.

Supplemental fire detection and suppression measures may be employed as required.

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YNPS FSAR REV. 6105 REFERENCES 513-1 NYR 92-156, Issuance of Amendment No. 142 To Facility Possession Only License No.

DPR Yankee Nuclear Power Station (TAC No. M83746), M. B. Fairtile (USNRC) to J. M. Grant, August 20, 1992.

513-2 Fire Protection Technical Requirements Manual.

513-3 NYR 03-027, J. Hickman (USNRC) to J. Kay (YAEC), Issuance of Amendment 157 RE:

Deletion of Operational and Administrative Requirements Following Fuel Transfer to ISFSI, April 18, 2003.

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YNPS FSAR RE V. 6/05 514 DECOMMISSIONING QUALITY ASSURANCE PROGRAM 5 14.1 Description Yankee Atomic Electric Company (YAEC) has developed and implemented 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 Yankee Decommissioning Quality Assurance Program (YDQAP) assure that the storage of spent fuel and the decommissioning of the Yankee Nuclear Power Station are performed in a safe and effective manner. All remaining applicable operational and administrative requirements that were contained in the YNPS Defueled Technical Specifications have been relocated to the YDQAP (Reference 514-5).

The YDQAP complies with the requirements set forth in Appendix B of IOCFR 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.

The YDQAP is also established, maintained and executed to comply with the requirements of 1 OCFR71, Subpart H, and 1 OCFR72, Subpart G for the storage and transportation of spent nuclear fuel and high level waste under the provisions of a General License contained in these parts.

The YDQAP is submitted periodically to the NRC in accordance with 1 OCFR50.54(a).

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YNPS FSAR REV. 6105 REFERENCES 514-1 License No. DPR Yankee Nuclear Power Station 514-2 YDQAP; Yankee Decommissioning Quality Assurance Program 514-3 1 OCFR Part 50 Appendix B; Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants 514-4 NRC Inspection and Enforcement Manual, Chapter 2561; Reactor Inspection Program -

Post-Operational Phase 514-5 NYR 03-027, J. Hickman (USNRC) to J. Kay (YAEC), Issuance of Amendment #157 RE: Deletion of Operational and Administrative Requirements Following Fuel Transfer to ISFSI, April 18, 2003.

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YNPS FSAR REV. 6105 515 EMERGENCY PLAN The Defueled Emergency Plan for YNPS was approved by the NRC and issued on October 30, 1992.

With the completion of the campaign to remove all spent nuclear fuel and GTCC waste from the Spent Fuel Pool and transfer to the Independent Spent Fuel Storage Installation (ISFSI) and due to the advanced state of decommissioning (demolition of plant SSCs, reduction of source term), there are no longer any events (radiological or non-radiological) that could occur at the site outside of the ISFSI that could cause activation of the Defueled Emergency Plan. Therefore on March 3, 2005 a revision to the Emergency Plan was issued which eliminated all facets of the Plan not related to the ISFSI. The Plan was re-titled ISFSI Emergency Plan.

The ISFSI Emergency Plan is implemented through the use of Emergency Implementing Procedures which are referenced in Appendix B of the Plan. The ISFSI Emergency Plan is formally audited as part of the Quality Assurance Program.

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YNPS FSAR REV. 6/05 APPENDIX A Plant Drawings A-1

YNPS FSAR REV. 6/05 Drawing No.*

9699-FE-IP 9699-FE-I Q 9699-FY-6A APPENDIX A Plant Drawings Title One Line Diagram Mass Electric Line One Line Diagram Distribution Panel D-1 Plot Plan Page A-4 A-5 A-15

Note date of last revision; subsequent revisions may have occurred since printing of FSAR.

A-2

ML052150047

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