ML20141F639

License Termination Plan for Yankee Nuclear Power Station

ML20141F639
Person / Time
Site:	Yankee Rowe
Issue date:	05/15/1997
From:	YANKEE ATOMIC ELECTRIC CO.
To:
Shared Package
ML20141F623	List: ML20141F618 ML20141F639
References
PROC-970515, NUDOCS 9705210393
Download: ML20141F639 (100)

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YANKEE NUCLEAR POWER STATION l

LICENSE TERMINATION PLAN J

Revision 0 lO May 1997 J

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Prepared By:

Yankee Atomic Electric Company 580 Main Street Bolton, Massachusetts 01740

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YNPS LICENSE TERMINATION PLAN REVISION 0 TABLE OF CONTENTS l

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1. OVERVIEW OF THE LICENSE TERMINATION PLAN........................ 1-1 3

l.1 Executive S ummary............................................... 1 -1 1.2 Historical Background............................................. 1 -2 j

1.3 Regulatory Basis for Administration of the License Termination Plan........ 1-4 1.4 Major Tasks, Schedules, and Activities................................ 1-4 i

1.5 Re ferences....................................................... 1 -6

2. SITE CHARACTERIZATION............................................ 2-1 2.1 I ntrod uction......................................................

2-1 2.2 Site Characterization Procedures..................................... 2-1 2.3 Structural and System Surveys....................................... 2-1 l

2.3.1 Radionuclide Distribution................................... 2-2 2.3.2 Contamination Levels in Systems and Structures................. 2-2 i

2.3.3 Activation Analyses........................................ 2-3 I-2.4 Environmental Surveys.............................................. 2-3 2.4.1 Determination of Background Radiation Values.................. 2-3 2.4.2 Surface Soil and Asphalt Surveys............................. 2-4 2.4.3 Subsurface Soil Sampling and Gamma Logging in Open Land Areas.............................................. 2 -4 2.4.4 Subfloor Soil Sampling..................................... 2-5 2.4.5 Ground water............................................. 2-6 2.4.6 Deerfield River and Sherman Pond Sediment........,........... 2-7 l

2.4.7 Leach Fields.............................................. 2-7 2.4.8 Southeast Construction Fill Area.............................. 2-8

2. 5 Re ferences.......................................................

2-8 2

2.6 Procedure References.............................................. 2-9

3. REMAINING DISMANTLEMENT ACTIVITIES.............................. 3-1 3.1 Introd uctio n..................................................... 3 - 1 3.2 Remaining Systems, Structures, and Components not Required for Spent Fuel Storage (Phase 1)........................................ 3-1 3.3 Systems, Structures, and Components Associated With Storage of Spent Fuel in the Spent Fuel Pool (Phase 2),................................. 3-2 3.4 License Termination (Phase 3)....................................... 3-3 3.5 Re ferences...................................................... 3 -4 4

4. SITE REMEDI ATION....................................................

4-1 4.1 Introd uc tion......................................................

4-1 4.2 Remediation Methods..............................................

4-1 4.2.1 Building and Structure Surfaces.............................. 4-1 L

4.2.2 Surface Soils and Asphalt................................... 4-2

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, g'N 4.2.3 Sed iment................................................ 4-2 1

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YNPS LICENSE TERMINATION PLAN REVISION 0 TABLE OF CONTENTS 4.2.4 S ubfloo r Soil............................................. 4-3 4.3 Areas to be Remediated............................................ 4-3 4.4 AL ARA Analysis................................................. 4-4 4.4.1 A pproach................................................ 4-5 4.4.2 Assumptions............................................. 4-5 4.4.3 Soil Remediation.......................................... 4-6 4.4.4 Structural Surface Remediation............................... 4-7 4.4.5 Conclusions.............................................. 4-9 4.5 Re feren ce s....................................................... 4 -9

5. END U S E OF TH E SITE.................................................

5-1 5.1 Disc us si o n.......................................................

5-1 5.2 Re feren ces....................................................... 5 - I

6. ESTIMATE OF REMAINING DECOMMISSIONING COSTS................... 6-1 6.1 Introduction and Background..........................

..... 6-1 6.2 Remaining Decommissioning Cost Estimate............................ 6-1 6.3 Decommissioning Funding.......................................... 6-2 6.4 Re ferences....................................................... 6-3

7. ENVIRONMENT.AL CHANGE ASSOCIATED WITH LICENSE TERMINATION ACTIVITIES..........................................

7-1 7.1 Dis c ussi o n.......................................................

7-1 7.2 Re fe ren ce s....................................................... 7-2 APPENDIX A - FINAL STATUS SURVEY PLAN O

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YNPS LICENSE TERMINATION PLAN REVISION 0

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TABLE OF CONTENTS LIST OF TABLES t

TABLE I[RS 2-1 Summary of Surface Soil and Asphalt Samples 2-2 Summary of Radiological Subsurface Soil Analyses 2-3 Summary of Subfloor Soil Sample Analyses l

2-4 Summary of Radiological Ground Water Analyses 2-5 Summary of Radiological Analyses of Sediment Samples 3-1 May 1997 Status of Systems, Structures, and Components Described in FSAR 2

4-1 Cost to Remediate Soilin 100 m Reference Area 4-2 YNPS Surface Soil Remediation: Cost per Person-Rem at Incremental Dose Levels 6-1 1995 Cost Study Summary and FERC Settlement 62 Current Decommissioning Estimate including 1995 and 1996 Actual Costs and 1997 Budget f

LIST OF FIGURES FIGURE TITLE l-1 YNPS Decommissioning Schedule 2-1 Surface Soil Sampling Areas Shown on Table 2-1 2-2 Subsurface Soil and Groundwater Sampling Locations 2-3 Plan of Subfloor Samples 2-4 River Sediment Samples - General Locations 2-5 Sherman Pond and Deerfield River Sediment Sampling Study 6-1 Projection of Decommissioning Funds Availeble iii

1 YNPS LICENSE TERMINATION PLAN REVISION 0 TABLE OF CONTENTS ABBREVIATIONS. TERMS. AND UNITS ADD Activity Decay and Dilution (Tank)

Ag-108m Silver-108m ALARA As Low As is Reasonably Achievable CFR Code of Federal Regulations Ci Curie; unit of radioactivity = 3.7 x 105 disintegrations per second Co-60 Cobalt-60 Cs-137 Cesium-137 DOE Department of Energy EPA Environmental Protection Agency Fe-55 Iron-55 FERC Federal Energy Regulatory Commission FR Federal Register FSAR Final Safety Analysis Report GEIS Generic Environmental Impact Statement GLV Guideline Value GTCC Greater Than-Class-C (waste)

H-3 Tritium IIEPA liigh Efficiency Particulate Air (filter)

LLW Low Level Waste Liter MDA Minimum Detectable Activity MDC Minimum Detectable Concentration mrem Millirem; unit of dose equivalent,0.001 rem Mwe Megawatts Electric Mwt Megawatts Thermal ND No (activity) Detected Ni-63 Nickel-63 NRC Nuclear Regulatory Commission PCA Potentially Contaminated Area pCi Picoeurie, unit of radioactivity; 1 x 10 2 Ci Person-rem Collective radiation dose to a population PSDAR Post Shutdown Decommissioning Activities Report RCA Radiation Control Area Rem Unit of dose equivalent REMP Radiological Environmental Monitoring Program RESRAD Computer code used to medel dose from residual radioactivy RPV Reactor Pressure Vessel SFP Spent Fuel Pool SI Safety Injection TEDE Total Effective Dose Equivalent iv

YNPS LICENSE TERMINATION PLAN REVISION 0 -

TABLE OF CONTENTS ABBREVIATIONS. TERMS. AND UNITS TLG TLG Services,Inc.

VC Vapor Container YAEC Yankee Atomic Electric Company YNPS Yankee Nuclear Power Station i

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YNPS LICENSE TERMINATION PLAN REVISION 0 n

SECTION 1 OVERVIEW'OF THE LICENSE TERMINATION PLAN 1.1 EXECUTWE

SUMMARY

This License Termination Plan confonns to 10 CFR 50.82(a)(9) and supports the application by the Yankee Atomic Electric Company (YAEC) for termination of the license for the Yankee Nuclear Power Station (YNPS). The objective of this document is to demonstrate that:

(1) there are adequate funds to complete decommissioning and release the site for unrestricted use, (2) the site release criteria ensure that exposure to residual levels of radiation is kept As Low As is Reasonably Achievable (ALARA), and (3) the final status survey program is adequate to verify that the release criteria have been met.

O This plan dcacribes the process by which decommissioning will be completed and the site

.V released for unrestricted use. The following is a brief summary of the sections presented m the License Termination Plan:

S-etion 1: Overview of the License Termination Plan 'ihis.section preserts the objectives of the plan and summarizes the information contained within it. This section discusses long term spent fuel storage under the existing 10 CFR Part 50 license and describes how decommissioning ared site restoration activities will be completed in three phases. Some historical background, the regulatory basis for administration of the plan, and a decommissioning sche.lule are also provided.

Section 2: Site Characterization - This section describes the radiological surveys that have been conducted to characterize the extent and magnitude of contamination at YNPS.

General findings are presented and areas of the site that are likely to require remediation are identified. Plans for future surveys are also described.

Section " Identification of Remaining _ Dismantlement Activities - This section presents the sequence of dismantlement and decontamination activities for the remaining systems, structures, and components at YNPS.

Section 4: Plans for Site Remediation - This section describes the areas on site that may be subject to remediation and identifies the methods that may be used. Additionally, an

!q analysis in this section demonstrates that the site release criterion of 15 mrem / year is j

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YNPS LICENSE TERMINATION PLAN REVISION 0 adequate to ensure that residual levels of radioactivity at YNPS will be ALARA. For areas at YNPS with initial dose levels ofless than 15 mrem / year, the benefit of further remediation is not justified by the associated cost.

Section 5: End Use of the Site - This section presents the goal of decommissioning:

release of the site for unrestricted use and return of the site to a " green field" condition.

Section 6: Estimate of Remaining Decommissionine Costs - This section presents an estimate of remaining costs that is based on the 1995 cost study submitted to the Federal Energy Regulatory Commission. Including fuel storage and site restoration costs, total decommissioning costs remaining as of January 1,1997 are $235.0 million in constant value 1995 dollars. This section also discusses the funding assurance provided by YAEC's power contracts, which obligate the purchasers for the full cost of decommissioning.

Section 7: Environmental Statement - This section demonstrates that decommissioning activities will continue to be accomplished with no significant adverse envirorunental impacts. Decommissioning and license termination activities remain bounded by previously issued environmental impact statements.

Aopendix A: Final Status Survey Plan - The Final Status Survey Plan presents the site g

release criteria and describes the methods that will be used by YAEC to demonstrate that W

radiation and radioactive :antamination levels at YNPS have been reduced to levels below the site release criteria. The Total Effective Dose Equivalent to the average member of the critical population group from residual contamination on site must be less than 15 mrem / year. The methods governing the conduct of final status surveys are derived from regulatory guidance, specifically Regulatory Guide 1.86, Draft NUREG/CR-5849, and Draft NUREG-1500. The Final Status Survey Plan describes the division of the site into survey areas, the classification of those survey areas, and the requirement that all survey areas meet the release criteria with a 95% confidence level.

Management controls over all aspects of the project are discussed in detail, including quality assurance, bakground level determinations, data processing, and final status survey reports.

1.2 HISTORICAL BACKGROUND Yankee Nuclear Power Station achieved initial criticality in 1960 and began commercial operations in 1961. The Nuclear Steam Supply System was a fcur loop pressurized water reactor de signed by Westinghouse Electric Corporation. The original thennal power design hmit of 485 Mwt was upgraded to 600 MWt in 1963. The Turbine Generator, also designed by Westinghouse, was rated to produce 185 MWe.

On Februaq 26,1992, the YAEC Board of Directors decided to cease power operations 1-2

YNPS LICENSE TEIGilNATION PLAN REVISION 0 p

V permanently at YNPS. This decision was based on the following factors:

Economic analyses indicated that shutdown of the plant before expiration ofits Nuclear Regulatory Commission (NRC) operating license in July 2000 could produce a substantial savings to the electricity purchasers.

Significant regulatory uncertainty existed concerning the timing and cost of the completion of the NRC's review of the integrity of the YNPS Reactor Pressure Vessel.

On August 5,1992, the NRC amended the YNPS Facility Operating License (DPR-3) to a possession only status. Combined with other amendments and program changes, this possession only status formed the basis of the Decommissioning Plan. In December 1993, the Plan (Reference 1-1) was submitted by YAEC in accordance with the requirements of the recently superseded 10 CFR 50.82(a), which required the submittal of a proposed decommi"ioning plan within two years of the permanent cessation of operations. The Decommissioning Plan also included spent fuel management plans currently required by 10 CFR 50.54(bb). The Decommissioning Plan was approved on February 14,1995 (Reference 1-2). Due to subsequent litigation, the NRC suspended its approval of the Plan in October 1995; the Plan was re-approved on October 28,1996 p

(Reference 1-3).

V As reqAed by a commitment to the NRC, the Decommissioning Plan was incorporated into the.iune 1995 revision of the YNPS Final Safety Analysis Report (FSAR). Under the current 10 CFR 50.82 (Reference 1-4), an approved decommissioning plan is considered to be the licensee's post-shutdown decommissioning activities report (PSDAR). For YNPS, the tenns " Decommissioning Plan" and "PSDAR" are used interchangeably.

As of May 1997, the majority of the systems and components not required to support the storage of spent fuel have been dismantled and disposed of off-site in accordance with the YNPS Decommissioning Plan and FSAR. The Main Coolant System has been removed, including the Steam Generators, Pressurizer, and large diameter piping. The Reactor Pressure Vessel has also been removed from containment and shipped to the Barnwell low level waste facility in an NRC-approved transportation cask. Structural decontamination work has begun in the Ion Exchange Pit and Waste Disposal Building followmg the clean-out of systems and components from these areas. The Spent Fuel Pool (Spent Fuel Pit) and other systems associated with fuel storage have been electrically and mechanically isolated to create a Spent Fuel Pool " island" that will not be adversely impacted by ongoing decommissioning activities. Radiological characterization of YNPS has been ongoing since 1991 and involves surveys of systems, structures, and the environment.

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YNPS LICENSE TERMINATION PLAN REVISION 0 1.3 REGULATORY BASIS FOR ADMINISTRATION OF TIIE LICENSE TERMINATION PLAN The YNPS License Termination Plan has been prepared in accordance with the requirements of 10 CFR 50.82," Termination of License." The License Termination Plan provides documentation on the remaining activities necessary to terminate the facility license. The plan will be a supplement to the FSAR in accordance with 10 CFR 50.82(a)(9)(i), and as such may be amended without prior NRC approval provided the changes do not:

Involve an unreviewed safety question as defined in 10 CFR 50.59, Foreclose release of the site for possible unrestricted use, Result in significant environmental impacts not previously reviewed, or Result in there no longer being reasonable assurance that adequate funds will be available for decommissioning.

The following locuments will constitute the decommissioning licensing basis of YNPS following NRC approval of the License Termination Plan:

g FSAR including the Decommissioning Plan (PSDAR) and License Ter r. nation a

Plan NRC Possession Only License, and YNPS Technical Specifications.

1.4 MAJOR TA.W S, SCIIEDULES, AND ACTIVITIES Decommissioning will be completed in three phases. The current phase consists of the decontamination and dismantlement of remaining systems and components which do not s spport fuel storage. After the removal of all spent fuel from the Spent Fuel Pool (SFP),

the second phase of decommissioning will involve 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 will occur after all spent fuel has been taken off site. If a dry cask storage facility is constructed, license termination will take place after the decommissioning of that facility. All decommissioning activities will be accomplished with no significant adverse emironmental impacts.

In the current phase, remaining systems and components not required for fuel storage are l-4

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being removed from plant structures. After the removal of all contaminated systems and components from a structure, the structure is decontaminated. As decommissioning 1

progresses, site characterization activities will continue to evaluate soil, asphalt, sediment, and structural surfaces for remediation. All structures and environmental media which do not meet the site release criteria contained in Appendix A will be remediated. Final status surveys will be conducted according to Appendix A to verify that the release criteria are met. Independent verification of the results by the NRC will allow for the release of the infividual surveyed structures and open land areas as non-radiologically controlled mater:al available for ceaventional demolition and disposal. In order to facilitate remediation, stoctures may be released and demolished prior to remediating subfloor soils beneath the structures. Measures will be implemented as described in Section 4.1.7 of Appendix A to prevent recontamination of surveyeJ areas prior to finallicense termination.

There are currently 533 fuel assemblies, a small amount of reconfigured fuel, and 21 canisters of Greater-Than-Class-C (GTCC) waste stored in the SFP. A final decision has not been made on the long-term storage method for spent fuel at YNPS, however, design and certification of a dry cask storage facility suitable for Yankee-class fuel is currently underway. For planning purposes, the current decommissioning cost estimate I

assumes that a dry cask storage facility for spent fuel will be constructed on site. The i

O facility will utilize a design approved by the NRC and consist of a concrete pad, fuel storage canisters, and concrete overpacks. Yankee will operate the facility using its existing programs under its 10 CFR Part 50 license. All spent fuel will be transferred

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from the SFP when the dry cask storage facility is complete. Some or all of the GTCC waste may be taken by the Department of Energy (DOE) as part of a pilot project. Any remaining GTCC waste will be stored in the dry cask storage facility until final disposition by the DOE.

The second phase of decommissioning will commence once the spent fuel and GTCC waste have been transferred to the dry cask storage facility or taken off site. The SFP and the systems, structures, and components that support it will be dismantled and decontaminated. Soil remediation will also be performed to meet the site release criteria.

Structures will be demolished after final status surveys and independent verification by the NRC. Site restoration activities will result in the removal of all released structures to three feet below final site grade and the deposition of sufficient topsoil to permit the growth ofvegetation.

For planning purposes, the final phase of decommissioning is assumed to take place in 2018 after all spent fuel and GTCC waste has been shipped to the DOE. At that time, the dry cask storage facility will be dismantled and decontaminated. This will consist primarily of the disposal of the concrete canister overpacks. It is anticipated that only a small amount of radioactive waste will be generated. The YNPS license will be terminated after a final stage of final status surveys and independent NRC verification.

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YNPS IJCENSE TERMINATION PLAN REVISION 0 The site will then be returned completely to a " green field" condition.

Figure 1-1 shows the current schedule for the activities described above. The current phase of dismantlement and decontamination and final status surveys will be complete in June 1998. It is estimated that the design and construction of the dry cask storage facility will be complete by June 1999. The transfer of spent fuel and GTCC waste from the SFP will last approximately six months. Decommissioning of the SFP island will last approximately one year, including final status surveys. Site restoration activities will take place from January 1999 to December 2001. The dry cask storage facility will be operated from 1999 to 2018, when the last fuel assembly is assumed to be taken off-site.

Using this assumption, the YNPS license will be terminated in December 2018 after the dry cask storage facility is decommissioned.

Planning sequences and dates are based on current knowledge and could change in the future. Several factors influence the choice of the spent fuel storage method at YNPS, including low-level radioactive waste site availability, the impact of the chosen method on decommissioning the balance of the site, the timing of spent fuel acceptance by the J DOE, and the relative economics of dry versus wet storage. Yankee may store spent fuel in the SFP until all fuel is taken off-site. Both wet and dry storage options are addressed in the spent fuel management plans contained in the FSAR. Yankee will continue to inform the NRC of all major changes to planned decommissioning activities.

1.5 REFERENCES

1-1 Yankee Nuclear Power Station Decommissioning Plan, Revision 0.0.

l-2 Letter, M. B. Fairtile (USNRC) to J. A. Kay (YAEC), Order Approving the Decommissioning Plan and Authorizing Decommissioning of the Yankee Nuclear Power Station, February 14,1995.

1-3 Letter, M. B. Fairtile (USNRC) to J. A. Kay (YAEC), Completion of Hearing Process Regardiaa Approval of Decommissioning Plan for the Yankee Nuclear Power Station, Oc.ober 28,1996.

1-4 61-FR-39278, Decommissioning of Nuclear Power Reactors, July 29,1996.

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YNPS DECOMMISSIONING SCHEDULE Tssk Name

' 1997 l 1998 l 1999 l 2000 l 2001 l 2002 2003 l 2004 ' 2006 l2006 l 20071 2008l2009l 2010i! 2011 l 2012 l 20 iJ l 2014 l 2016 l 2016 l 2017 l 2018 l 2019 Dismantiement/ Site Restoration Dismande/Decon Batance of P! art Final Status Survey Phase 1

'gi NRC Verircation of FSS Phase 1 g

Dismantie/Decon SFP island Final Status Survey Phase 2 NRC Verification of FSS Phase 2 Sie Restoration gggggg,j Spent Fuel Storage p

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, gggggg Design and Construct Dry Storage FacR!!y j Transfer Fuel from SFP zw_

%r Dry Storage ggygggggggigggggggggggggggggy gggggggggggggggggggggg;g,,,

g Decommission Dry Storage Facitlty g

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Terminate License 4

FIGURE 1-1

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YNPS LICENSE TERMINATION PLAN REVISION 0

!V SECTION 2 t

SITE CHARACTERIZATION

2.1 INTRODUCTION

The purpose of this section is to describe the radiological surveys that have been conducted to characterize the extent and magnitude of contamination at YNPS. Geneml findings are presented and plans for future surveys are described.

Surveys performed at YNPS during decommissioning have been conducted using guidance from NUREG/CR-5849 (Reference 2-1). Scoping surveys, whose purpose it is to identify the potential radionuclide contaminants at the site, the relative ratios of these radionuclides, and the general extent of contamination, were performed at YNPS during 1992 and 1993. The results of these surveys are summarized in Section 3.1 of the Decommissioning Plan (Reference 2-2). Site characterization surveys, which more precisely define the extent and magnitude of contamination, began in 1994 and are currently ongoing.

The results of the scoping and characterization surveys are used to identify areas of the site that will require remediation. Contamination levels can be compared against

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guideline values to determine if the resultant dose under a given scenario would exceed the site release criterion of 15 mrem / year. As defined in the Final Status Survey Plan (Appendix A), a guideline value is the level of residual radioactivity that equates to the site release criteria for a particular pathway or measurement. Guideline values for individual radionuclides are generally taken from the building occupancy or residential scenarios in Table B-2 of NUREG-1500 (Reference 2-3). Site characterization survey results have also been used in accordance with NUREG/CR-5849 to classify final status

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survey areas as "affected" and " unaffected." This area classification, shown in Figure 2.3 of Appendix A, will dictate the survey measurement frequency for the final status surveys.

2.2 SITE CHARACTERIZATION PROCEDURES 4

In addition to the general procedures used as a standard part of the radiation protection and environmental laboratory programs, procedures specific to site characterization are used to accurately implement and document site characterization surveys. These procedures are listed as procedure references at the end of this section.

2.3 STRUCTURAL AND SYSTEM SURVEYS Surveys performed on plant systems and structures consisted of area and contact dose rate qV measurements, samples of plant piping, smears of structures and components, and 2-1

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samples of paint and concrete. The results from these surveys were used to determine i

radionuclide distributiors and contamination levels. Additionally, an analytical j

activation analysis of neutron iiradiated components was performed.

2.3.1 Radionuclide Distributics i

During operation, many samples were taken from process and waste streams to determine radionuclide distributions. Prior to and during the scoping and characterization surveys, additional samples of process pipes and smears were obtained to determine the contamination radionuclide distributions. The data indicate that Co-60 is the most significant contributor to ganuna activity (approximately 75 percent of total gamma activity), and that Fe-55, Ni-63 and Co-60 comprise approximately 98 percent of the total activity in structures and systems. Contamination radionuclide distributions for the Main Coolant System and the balance of the plant are found in Table 3.1-2 of the Decommissioning Plan.

2.3.2 Contamination Levels in Systems and Structures Internal contamination levels were calculated for several systems and components gl based on extemal dose rates. Dose rate conversion factors were calculated based on the assumption that Co-60 is the primary contributor to the dose rates. The total internal contamination level in several plant systems were determined by scaling individual radionuclides to Co-60. Table 3.1-4 of the Deconunissioning Plan shows the average contamination levels that existed in all major plant systems as of January 1994.

Contamination levels for plant structures were estimated from concrete core bores, smears, and paint and concrete scrapings. Contamination levels from these samples are presented in Table 3.1-6 of the Decommissioning Plan. To determine

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the depth of contamination in concrete floors and walls,75 mm long core bore samples were obtained from various locations. These data are summarized in Table 3.1-5 of the Decommissioning Plan. With the exception of areas which were exposed to contaminated liquids for significant periods of time, most contamination was found to be limited to a depth of 7 mm. Recently, additional concrete core bore samples have also been collected from several locations in the Bioshield wall which surrounded the Neutron Shield Tank and the Reactor Pressure Vessel. Radiological analyses of these samples are currently being performed. When completed, these analyses will determine the depth of activated and contaminated concrete that will need to be removed in order to meet the site release criteria.

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YNPS LICENSE TERMINATION PLAN REVISION 0 O

Samples of roofing material were collected from most buildings on site during 1996. These samples were analyzed for gamma emitting radionuclides. While 5 of the 20 samples contained low but detectable levels of radioactivity, only two samples from the east end of the Primary Auxiliary Building contained radioactivity in excess of the guideline values.

1 2.3.3 Activation Analyses I

Neutron activation of the Reactor Pressure Vessel (RPV), the Neutron Shield Tank, and the Bioshield wall was estimated analytically. Gamma ray transport analyses were performed that calculated neutron activation by isotope from i

neutron flux data, reactor operating history, and material properties of the individual components. The results of the analyses are shown in Tables 3.1-7 through 3.1-9 of the Decommissioning Plan. The analyses showed that the water in the Neutron Shield Tank provided adequate shielding to reduce the activation of the outer Neutron Shield Tank wall and the Bioshield wall by five to six orders of magnitude. With the RPV and Neutron Shield Tank removed, concrete core i

samples recently taken from the Bioshield wall will provide further information i

on the depth of neutron activation.

2.4 ENVIRONMENTAL SURVEYS Environmental radiological investigations have included surveys for contamination in soil, asphalt, sediment, and groundwater, as well as the storm drain and leach field systems. Systematic sampling and monitoring (primarily of soil and asphalt) was performed during the scoping survey. Targeted sampling was then done as part of characterization surveys for all environmental media. This sampling was based on input from previous surveys, review of historical plant records (summarized in Appendix B of 1-the Decommissioning Plan), interviews with plant personnel, Radiological Environmental l

Monitoring Program (REMP) data, and aerial photos. These targeted surveys were expanded as necessary in an iterative fashion to determine the horizontal and vertical 4

extent of any significant contamination that was found. Contamination levels in environmental media were compared to the guideline values for the 15 mrem / year residential scenario from Table B-2 of NUREG-1500. If a radionuclide was present on site and was not listed in Table B-2, guideline values were determined by the use of the RESRAD computer code (Reference 2-4). Environmental surveys are ongoing.

2.4.1 Determination of Background Radiation Values As a result of global fallout from atmospheric nuclear weapons tests, Cs-137 is a significant part of the ambient background in the northeastern United States, particularly in forest soils and richly organic sediment. In order to assign i

O background values to various parts of the YNPS site for this fallout-related 2-3 l

YNPS LICENSE TERMINATION PLAN REVISION 0 Cs-137, in situ gamma spectroscopy measurements have been made and samples collected from reference locations. Analyses currently underway will determine the appropriate background values to be subtracted from final status survey results.

2.4.2 Surface Soil and Asphalt Surveys Surface contaminatum in surface soil and asphalt has been evaluated by the collection of soil grab samples, soil core samples, asphalt grab samples, and in situ measurements. Surface soil samples include soil, sod, stream sediment, and humus that is within the top 300 mm from the surface. Soil immediately under asphalt is considered surface soil. As part of the scoping and site characterization surveys, over 400 soil grab samples and over 110 asphalt grab samples have been collected. In situ gamma spectroscopy measurements have also been made at 140 locations with a high purity germanium detector (coilimated to reduce on-site background radiation levels). Based on infonnation collected as part of the preparation for final status surveys, a limited number of additional surface soil and asphalt samples, as well as in situ measurements, are planned for 1997.

Table 2-1 gives a summary of the characterization surveys done to date in each of the site areas shown on Figure 2-1. The primary radionuclides detected in soil on the YNPS site have been Co-60, Cs-137, and in limited areas, Ag-108m.

Although site characterization surveys are ongoing, results of surface sampling to date indicate that most contamination is limited to approximately the top 150 mm of soil. Only the Outdoor Radioactive Materials Storage Area does not currently meet the surface soil guideline values for the conservative residential scenario of NUREG-1500.

2.4.3 Subsurface Soil Sampling and Gamma Logging in Open Land Areas Subsurface soil studies in open land areas were begun in 1993, documented in Section 3.1.4.2 of the Decommissioning Plan, and currently include data from 11 soil borings ranging in depth from 2.4 to 7.6. These consist of 75 mm diameter continuous borings that were divided into multiple samples from 300 to 600 mm in length. Several of the borings were also used for the installation of groundy'ater observation wells, as described in Section 2.4.5. In addition to soil borings, several test pits were excavated with a backhoe to procure subsurface samples up to 2.1 feet deep. Borings and test pits were intentionally located at sites where contamination is known or suspected to be present below the surface. The locations of these wells, borings, and test pits are shown on Figure 2-2.

Results of analyses of soil samples are summarized in Table 2-2. For reference, 2-4

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the table also includes the 15 mrem / year guideline values for Co-60 and Cs-137.

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No contamination has been found beneath the top (surface) sample in each boring.

This is consistent with the results of the surface sampling discussed in the J

previous section. Only boring CB-9, located in the area beneath the Vapor Container, contained activity that exceeds the guideline values.- The soil in this area has since been removed to an approximate depth of 450 mm as part of the Reactor Pressure Vessel removal project. Further sampling will be done to determine if there is any remaining contamination.

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4 In addition to the laboratory analyses of soil boring samples, those borings in which observation wells were constructed were subjected to in situ gamma logging measurements. Logging was done at depth intervals of about three feet in those portions of the wells above the groundwater table. Only boring CB-8 in the i

Outdoor Radioactive Material Storage Area yielded gamma logging results that indicated elevated plant-related activity. This measurement was taken near the surface and is consistent with surface soil sample data from this general area.

1 Subsurface soil data examined as part of site characterization studies also includes j

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samples taken in 1992 from an excavation around the west and north sides of the Spent Fuel Pool Building. Out of 80 soil samples, eight contained activity in r

excess of the NUREG-1500 guideline values established for a 15 mrem / year dose

t.

limit. This area will be the subject of further characterization and is likely to require some soil remediation.

2.4.4 Subfloor Soil Sampling As shown in Figure 2-3, subfloor soil samples have been taken from 12 locations beneath structures on site. Subfloor soil samples are generally obtained by i

drilling through a floor to obtain shallow grab samples from the soil immediately beneath the concrete slab. Existing floor penetrations and soil borings have also been used to collect subfloor grab samples. Soil samples have generally consisted of constmetion fill sands and gravel.

The results of sampling performed to date are shown in Table 2-3. Radionuclide concentrations close to or in excess of the concentrations which correspond to 15 mrem / year have been identified in subfloor soil at six locations. These areas will be subject to further investigation or remediation:

Waste Disposal Building Evaporator Pit

. Waste Disposal Building Ash Drumming Pit Old PCA Building Decon Tub Area South Decon Pad

_ Ion Exchange Pit PCA Warehouse 2-5

YNPS LICENSE TERMINATION PLAN REVISION 0 The subfloor samples that have been taken to date were from targeted locations where contandnation was likely to be found. These sampling locations are typically in low areas of buildings with unlined concrete floors where either periodic or continuous water storage has occurred. They are not believed to be representative of typical subfloor soils at YNPS. More subfloor sampling is planned to investigate the potential for plant-related activity in other areas beneath stmetmes.

2.4.5 Groundwater Groundwater is located on site at depths ranging from 0.9 to 6.1 m. Relatively impctmeable soil appears to perch groundwater close to the surface beneath most plant buildings. Observation wells are used to obtain samples of groundwater and to assess its depth and flow pattern. A total of 17 observation wells on site range in depth from approximately 6.1 to 9.1 m. One deeper well,in place for a prior construction project, is also available for monitoring. Well locations are shown in Figure 2-2. Many wells have been constructed in existing soil borings where known sources of activity were present and where there may have been releases to the soil. In addition, samples from Sherman Spring are used to provide groundwater data near the Deerfield River as part of the YNPS REMP.

Preliminary results of radiological groundwater analyses are shown in Table 2-4.

Samples from 7 t ells have contained low concentrations of tritium. The highest tritium concentration measured was about 8,000 pCi/1, which is well below the Environmental Protection Agency limit of 20,000 pCi/l for man-made radionuclides in community water systems (40 CFR 141.16). The source of the tritium was most likely a leak from the Ion Exchange Pit that occurred in the 1960's and is described in Section 3.1.1 of the Decommissioning Plan Tritium concentrations in the observation wells vary according to their distance from the Ion Exchange Pit. At the Sherman Spring and the Deerfield River bank, tritium concentrations are less than 300 pCi/1.

Other plant radionuclides have been found in three locations. As shown on Table 2-4, an observation well located a short oistance down grade from the Spent Fuel Pool and the Ion Exchange Pit has produced a maximum Co-60 concentration of 23 pCi/l. This concentration yields a dose that is well below the EPA drinking water limit of 4 mrem / year. Water bailed from subfloor sampling holes in the Ion Exchange Pit and the Primary Auxiliary Building have also given positive indications of plant-related activity, but further investigation is needed to quantitatively determine the levels of contamination Periodic sampling of groundwater will continue in the future. Remediation of 2-6

-)

l YNPS LICENSE TERMINATION PLAN' REVISION 0 '

n

-Q j

contaminated soil that does not meet the site release criteria may partially remove j

the source of the activity detected in the groundwater.

l 2.4.6 Deerfield River and Sherman Pond Sediment' Sherman Pond was the cooling water source for the Circulating Water System and

]

is still the licensed discharge pathway for plant releases to the Deerfield River.

L

- Low levels of radioactivity were also released to the East and West Storm drains during plant operations as a result of several events described in Appendix B of l

the Decommissioning Plan. Site characterization has, therefore, included studies of the sediment in the Deerfield River, Sherman Pond, and the plant's storm drain outfalls.

4 Sediment sample locations are shown on Figures 2-4 and 2-5, and the analytical F

results are summarized in Table 2-5. Also included in the table are the guideline values for the 15 mrem / year residential scenario presented in Table B-2 of NUREG-1500.' This is an extremely conservative scenario since the actual

?

exposure pathways for contamination in sediment are much fewer than in the 4

residential scenario. The only plant-related radionuclides detected in sediment were Co-60 and Cs-137. None of the samples from Sherman Pond exceeded the 9

guideline values. In addition, the activity found in sediment from the East Storm (V

Drain Outfall was also below the guideline values. Sediment from the West Storm Drain Outfall contained even less activity, which was well below the guideline values. Additional sampling of the West Storm Drain Outfall is planned i

to confirm these findings.

2.4.7 Leach Fields i

The plant has used two septic leaching fields in the course ofits operational history. The approximate locations of these fields are shown Figure 2 2. Some septage solids from the systems serving these fields have been documented to contain very low levels of radioactivity.

i The original septic field has been investigated by means of a soil boring and a grot:ndwater observation well constructed in that boring. Since this boring and i

ob 2 tion wd is located down grade from the new septic field, it can also be l'

used to characterize the new field. Data from the water and soil samples taken at this location are summarized in Sections 2.4.3 and 2.4.5 of this report. All soil and groundwater samples from the original septic field have contained no I

detectable activity. Further investigation of the new septic field will be performed

. in the future.

O 2-7

{

i YNPS LICENSE TERMINATION PLAN REVISION 0 0!

2.4.8 Southeast Construction Fill Area The Southeast Construction Fill Area is located at the southeast margin of the site between the Pole Barn and the steep wooded hills that form the Deerfield River valley (Figure 2-2). This area is about 2 acres in size and was used to deposit excess soil from the plant's onginal construction. Subsequent plant construction projects also used this area for placement of excess construction soils. Although no specific historical documentation addresses this material, at least a portion of the fill may have been excavated from the plant Radiation Control Area.

Soil samples from this area have been obtained from surface sampling, test pits, a soil boring, and a groundwater observation well. Also, in-situ gamma measurements have been taken in 15 locations within this area. The only radionuclide that has been detected is Cs-137. Occurrence of this nuclide may be due exclusively to fallout. Further sampling will be performed to assess whether any plant-related radionuclides are detectable in this area.

2.5 REFERENCES

2-1 NUREG/CR-5849, Manual for Conducting Radiological Surveys In Support of License Termination (Draft Report for Comment), May 1992.

2-2 Yankee Nuclear Power Station Decommissioning Plan, Revision 0.0.

2-3 NUREG-1500, Working Draft Regulatory Guide on Release Criteria for Decommissioning, August 1994.

2-4 Yu, C. F. et al., Manual for Implementing Residual Radioactivity Materials Guidelines Using RESRAD, Environmental Assessment Division, Argonne National Laboratory.

2.6 PROCEDURE REFERENCES AP-0831," Administrative Program for Radiological Characterization Surveys to Support YNPS Deconunissioning Planning."

DP-8120," Collection of Surface Soil, Asphalt and Concrete Samples."

DP-8121,"In-Plant Radiological Surveys to Support the Radiological Characterization Program."

O 2-8

YNPS LICENSE TERMINATION PLAN REVISION 0 DP-8122, " Subsurface Soil Sampling and Monitoring Well Installation."

DP-8123," Sample Chain of Custody."

DP-8124," Collection of Pond Sediment Samples for Site Characterization."

DP-9745," Groundwater Level Measurement and Sample Collection in Observation Wells."

OP-8122, " Subsurface Soil Sampling and Monitoring Well Installation."

O f

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TABLE 2-1

SUMMARY

OF SURFACE SOIL AND ASPHALT SAMPLES 1993 -1996 Outdoor Rad Under vapor Balance of Industrial Area Balance of Owner Outside Guideline Mat. Storage Container RCA Controlled Area Owner Values Area (Note 1)

(Note 2)

(Note 3)

Controlled (pCUg)

Fence (Note 4)

Area Number on Figure 2-1 1

2 3

4 5

6 Approx. Area (m')

460 1,900 7,400 13,000 37,000 IN SITU EXPOSURE RATE MEASUREMENTS No.of Measurements 3

2 65 29 21 20 No. of Measurements >GLV (Note 5) 1 0

1 0

0 0

Range of Co-60 (pCVg) 0.20 -1 5 0.23- 0.42 ND - 3.9 NO - 0 090 ND - 0.15 ND - 0.097 2.97 Range of Cs-137 (pCVg)

ND - 10 ND - 0.54 ND - 2.4 ND - 0.28 ND - 0.45 ND - 0.84 10.7 Range of Ag-108m (pCVg)

ND ND ND ND ND ND 3.78 SURFACE SOIL GRAB SAMPLES (Note 6)

No. of Samples 54 44 79 54 46 13S No. of Samp6es >GLV (Ncte 5) 20 C

1 0

0 0

Range of Co-60 (pCVg)

ND - 1008 ND - 2.8 ND - 2.6 ND-1.7 ND ND - 1.4 2.97 Range of Cs-137 (pCVg)

ND - 61 ND - 1.5 ND - 3.1 ND - 1.0 ND - 2.9 ND - 2.0 10.7 Range of Ag-108m (pCVg)

ND - 100 ND - 0.10 ND-0.58 ND ND ND 3.78 ASPHALT GRAB SAMPLES No. of Samples 27 29 43 16 0

0 No. of Samples >GLV(Note 5) 4 2

1 0

NA NA Range of Co 60 (pCVg)

ND - 4.1 ND-7.2 ND - 4.2 ND NA NA 2.97 Range of Cs-137 (pCVg)

ND - 2.8 ND - 1.4 ND - 2.0 ND - 0.12 NA NA 10.7 l

Range of Ag-108m (pCVg)

ND - 6.0 ND - 0.17 ND ND NA NA 3.78 NOTES: (1) Some of this soil was removed during the Reactor Pressure Vessel removal project.

(2) The Radiation Control Area (RCA), excluding the Radioactive Materials Storage Area (Area 1 on Figure 2-1) and the area under the Vapor Container (Area 2 on Figure 2-1).

l (3) Not including the Radiation Control Area or the Industrial Area.

(4) Near-site samples only.

r i

(5) When multiple radionuclides are present, the guideline value (GLV) for the distribution is considered to be exceeded when the sum of the ratios of the i

individual radionuclide distributions to their respective GLVs is greater than 1. Individual GLVs are taken from Table B-2 of NUREG-1500 (15 i

mrem 9 ear residential scenario) or RESRAD code data for the YNPS site.

(6) Surface soil grab samples include soil (generally within the top 300 mm), sod, humus and stream sediment. Soil immediately under asphalt is l

considered surface soil.

l (7) ND = No activity detected.

l l

,~

c V-Table 2-2 Summary Of Radiological Subsurface Soil Analyses

~

No. of No. of Sample Depth of No.of Soil Analyses Analyses "W

Location Samples Containing Containing Analyzed Detectable Detectable Fig. 2-2)

(meters)

Co-60 Cs-137 oCi/g pCi/g CB-1 Down gradient of SFP Building 7.8 20 1

1 0.09 0.08 2

CB-2 Roadway down-gradient of Site 6.6 20 0

0 ND ND CB-3 Adjacent to New Fire Tank 4.5 15 0

2 ND 022 CB-4 in Old Septic Leaching Field 5.8 19 0

0 ND ND CB-5 Southeast Construction Fill Area 9.4 17 0

1 ND 0.07 CB-8 Outdoor Rad Waste Storage Area 3.8 8

2 2

0.26 ND CB-9 Under Vapor Container 7.3 13 2

2 2.51 5.7 SB-1 Between Waste Bldg and PAB 2.4 4

0 0

ND ND SB-1A Down gradient of Waste Bidg.

6.0 11 1

1 0.80 0.48 SB-2 Adjacent to Waste Bldg.

7.6 16 0

0 ND ND SB-3 Adjacent to Waste Bldg.

7.4 11 1

1 0.13 0.1 TP-1 Southeast Construction Fill Area 1.8 3

0 1

ND 0.12 TP-2 Southeast Construction Fill Area 1.8 4

0 0

ND ND TP-3 Southeast Construction Fill Area 1.2 3

0 0

ND ND TP-4 Southeast Construction FCI Area 2.1 4

0 2

ND 0.09 i

l Guideline Values' l

l l

2.97 l

10.7 i

'NUREG-1500 soil guideline values (15 mrem /yr residential scenario) 2ND = No activity detected l

l l

l

i l-n Table 2-3 l]

Summary of Subfloor Sol Sample Analyses Location Depth Detected Detected Detected Detected Detected Hole No.

(See Fig 2-3 for Location (mm) below Co40 Co-137 Co-144 Ag-108m Cs-134 Plan) top of floor pCVg pCVg pCVg pCVg pCilg 3

SF-1 Evaporskr PR, SE comer 356410 5.25 5.15 ND ND 0.63 Weste Depoesi Side.

610 686 6.26 6.54 ND ND 0.73 356435 4.55 3.20 ND ND 0.39 356 435 10.29 3.22 ND ND 0.38 SF-1 A Evaporator Pt 445-691 0.63 2.86 ND ND 0.25 Weste Deposal Bido.

SF-2A Ash Drumerung Pit 457-660 7.33 3.91 ND ND 3.33 Weste Deposal Bido.

SF-3 Near NW Comer of Tub 289-432 0.15 ND ND ND ND PCA 81 (Old) 610486 0.25 ND ND ND ND 610413 0.55 ND ND ND ND SF-3A Tub Dreen Velve Pt 229-533 695.39 20.67 3.54 ND ND PCA 81 (Old) 533-737 917.53 67.97 ND ND NO SF-4 Near NW comer 333-432

'O NC ND NO ND New FuelVault 432-564 ND 0.05 ND ND ND 584486 0.11 0.15 ND ND ND SF-5A NW comer sleeve 432-737 271.60 19 40 ND ND ND South Decon Pod SF-5H SE comer sleeve 432435 5700.00 91.00 ND ND ND South Decon Pad SF 8A W ond IX Pt Pipe Tunnel 229-559 1.30 ND ND ND ND SF 10 NW comer IX PA 610-914 19.18 30.59 ND 8.96 ND CB-SF-12A PAB PipeTrenchfloor 305-1219 0.10 ND ND ND ND SF-13 Center of foundation 305-508 ND 0 38 ND ND ND ADD Tk foundation CB-7 PCA #1 Tub Drain Line ln 203-317 312 0.49 ND ND ND Red Weste We Wmune 2

Soil Guideline'/elues' 2.97 10.70 152.01 3 90 l

4.90

'Bened NUREG-1500 guideline'waes for 15 mrlyr, reesdental sceneno

% for A0-100m from RESRAD date for sNe 440 = No actMty detected n

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t Table 2 Summary Of Radiolo0ical Ground Water Analyses Max.

Avg.

Min.

No. of Max.

g,,

No. of Detectable Detectable Detectable Analyses Detectable Well Location Analyses Tritium Tritium Tritium Detecting Co40 ng To-Date Concentra-Concentra-Concentra-

Co40, Concentra-tion, pCM tion, pCM tion, pCM pCM tion' pCM CB-1 Down gradient of SFP Building 17 17 8045 6404 3830 7

22.7 CB-2 Roadway down-gradient of Site 13 10 3038 1651 1460 0

ND 3

CB-3 Adjacent to New Fire Tank 12 0

ND ND ND 0

ND CB-4 in Old Septic Leaching Field 13 0

ND ND ND 0

ND CB-5 Southeast Construction Fill Area 9

0 ND ND ND 0

ND I

CB4 River bank down gradient of Site 10 9

1140 591 220 0

ND CB-7 PCA Warehouse 1'

O ND ND ND 0

ND CB4 Outdoor Rad Waste Storage Area 10 0

ND ND ND' O

ND CB-9 Under Vapor Container 9

9 7500 6406 4930 0

ND CW-1 Down gradient of Diesel Generator Bldg.

13 0

ND ND ND 0

ND CW-2 Down Gradient of Si Tanks and Waste Bldg 13 0

ND ND ND 0

ND CW-3 Adjacent to Ion Exchange Pit 13 0

ND ND ND 0

ND CW-4 Between PCA Storage Bldgs.

13 0

ND ND ND 0

ND CW-5 Adjacent to Service Bldg.

13 0

ND ND ND 0

ND CW4 Down gradient of Front Office 13 0

ND ND ND 0

ND CW-7 Down gradient of Service Bldg.

10 4

780 530 400 0

ND CW-8 Turtnne Bld0 10 3

630 560 490 0

ND B-1 Down gradient of lon Exchange Pit 13 12 5335 3832 2840 0

ND 2

Guideline Values 20000 i

Approx. 200

'CB-7 installed January,1997 2Applicable U.S. EPA Drinking Water Limit (40 CFR141) for Tritium and approximate concentration equivalent for 4 mrem /yr dose of Co40 per USEPA Drinking Water Standards

'ND = No activity detected l

[

OO Table 2-5 Summary of Radiologlu:1 Analyses of Sediment Samples No. of No. of Location:

Maximum Maximum Sampes Detectable 88*d*8 Detectable General Plan of No. of h

h Sampling Locations in Samples CM0 Cs-137 NO Figures 2-4 and 2-5 PU9 Cs-137 Co-60 Sherman Pond at CW 8

2 0.96 8

1.93 Discharge Area Sherman Pond 14 3

0.76 9

1.58 Wheeler Bk Cove' West Storm Drain 6

1 0.1 6

0.22 Outfall Stream Bed West Storm Drain 9

3 0.32 8

0.52 Outfall Delta l

Sherman Pond at south 20 1

0.09 20 3.03 end Sherman Pond Central 3

10 0

ND 10 2.19 Area Sherman Pond

/

Shoreline in Wheeler 3

0 ND 2

0.21

(

Bk Cove Sherman Pond 5

0 ND 3

0.12 Shoreline Central Area No. 5 Dam Pond and 9

0 ND 7

0.28 Shoreline No. 5 Dam, 5

0 ND 1

0.46 Downstream Canals Bear Swamp Lower 10 0

ND 9

0.59 Reservoir i

River Shoreline Down-14 0

ND 4

0.46 stream of YNPS No.4 Dam Reservoir 6

0 ND 6

0.03 Guideline values 2.97 10.7 2

' Includes area adjacent to East Storm Drain outfall 2NUREG-1500 guideline values for soil, given a 15 mrem /yr residential scenario 8ND = No activity detected i

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2) Under Vapor Container

3) Balance of RCA scum

4) Industrial Area

5) Balance of Owner Controlled Area O

Figure 2-1

6) Outside Owner Controlled Fence Surface Soil Sampling Areas Shown onhnble 2-1

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Yankee Nuclear Power Station i

i LEGEND i

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SF-2, 2A Drum Drying Pit i

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Figure 2-4

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i l

YNPS LICENSE TERMINATION PLAN REVISION 0 O

SECTION 3 IDENTIFICATION OF REMAINING DISMANTIEMENT ACTIVITIES

3.1 INTRODUCTION

As of April 1997, the majority of systems and components not required to support the storage of spent fuel have been dismantled and disposed ofin accordance with the Yankee Nuclear Power Station Decommissioning Plan and Final Safety Analysis Report (References 3-1 and 3-2). In addition, the Spent Fuel Pool and other systems associated l

with fuel storage have been electrically and mechanically isolated to create a Spent Fuel Pool " island" that will not be adversely impacted by ongoing decommissioning activities.

The current status of the systems, structures, and components described in the Final Safety Analysis Report is shown in Table 3-1. The purpose of this section is to identify the dismantlement and decontamination activities associated with the remaining systems, structures, and components at YNPS.

Decommissioning will be completed in three phases. The current phase consists of the removal of any remaining systems and components which do not support fuel storage or l

subsequent decontamination activities. After the removal of all spent fuel and Greater-C Than-Class-C (GTCC) waste from the Spent Fuel Pool (SFP), the second phase of decommissioning will involve the dismantlement and decontamination of the SFP and its supporting systems, structures, and components. In the final phase of decommissioning, 1,

the possession only license will be terminated.

3.1 REMAINING SYSTEMS, STRUCTURES, AND COMPONENTS NOT NECESSARY FOR SPENT FUEL STORAGE (PHASE 1) 1 The remaining systems and components listed below are not necessary to support spent fuel storage and will be dismantled and decontaminated in the current phase of decommissioning. Dismantlement and decontamination activities will be performed in r.ccordance with the general considerations of Section 2.3 of the YNPS Decommissioning Plan and Section 200 of the FSAR. In many cases, only a part of the system will be removed, with the remaining portion maintained in service to support spent fuel storage.

Dismantlement activities will continue to be planned to ensure that there will be no adverse impact on spent fuel storage operations. Any applicable sections of the Decommissioning Plan and FSAR that are specific to a system or component are shown respectively in parentheses:

' Portions of the Radiation Monitoring System (Section 2.3.5.15; Section 215)

Vapor Container Ventilation and Purge System (Section 2.3.5.16; Section 216)

Portions of the Fuel IIandling System (Section 2.3.5.20, Section 220)

' ^

Original Plant Septic Tank and Piping

<i 3-1

'~

v

~

-=

_m,

YNPS LICENSE TERMINATION PIAN REVISION 0 Waste IIold-Up Tank (Section 2.3.5.55; Section 225)

Portions of the Fire Protection and Detection System (Section 2.3.5.32; Section 232)

Vapor Container Polar Crane (Section 2.3.5.38; Section 238)

Yard Area Piping (burica piping from several different systems)

After me removal of systems and components from an area or building, contaminated concrete, steel, and other building materials will be decontaminated or removed in a manner consistent with Section 't.3 of the YNPS Decommissioning Plan. The structures listed below are not required to store spent fuel and will be decontaminated in the current stage of decommissioning. Any applicable sections of the Decommissioning Plan and FSAR that are specific to a structure are shown respectively in parentheses:

Vapor Container (Section 2.3.5.36; Sec tion 236)

Reactor Support Structure (Section 2.3.137; Section 237)

Upper and Lower Pipe Chases (Section 2.3.5.41; Section 241)

Fuel Transfer Chute (Section 2.3.5.42; Section 242)

South, East, and West Walls ofIon Exchange Pit (Section 2.3.5.44; Section 244)

Primary Auxiliary Building (Section 2.3.5.48; Section 248)

Diesel Generator Building (Section 2.3.5.49; Section 249)

Waste Disposal Building (Section 2.3.5.50; Section 250) g Safe Shutdown Bui! ding (Section 2.3.5.51; Section 251)

Potentially Contaminated Area (PCA) Buildings 1 and 2 (Section 2.3.5.52; Section 252)

Compactor Building (Section 2.3.5.53; Section 253)

Final status surveys will be conducted according to Appendix A to verify that structures or open land areas not required for spent fuel storage meet the release criteria.

Independent verification of the results by the NRC will allow for the release of the individual surveyed structures and open land areas as non-radiologically controlled material available for conventional demolition and disposal. In order to facilitate rem:diation, stnictures may be released and demolished prior to remediating subfloor soils beneath the structures. Measures will be implemented as described in Section 4.1.7 of Appendix A to prevent recontamination of surveyed areas prior to final license termination.

3.3 SYSTEMS, STRUCTURES, AND COMPONENTS ASS,0CIATED WITII STORAGE OF SPENT FUEL IN TIIE SPENT FUEL POOL (PIIASE 2)

Afler the spent fuel and GTCC waste have been removed from the Spent Fuel Pool, t'te remaining components of the systems listed below will be dismantled and decontaniinated in acmrdwce with Section 2.3 of the Decommissioning Plan and Section 3-2

v YNPS LICENSE TERMINATION PLAN REVISION 0 4

200 of the FSAR. Any applicable sections of the Decommissioning Plan and FSAR that are specific to a system or component are shown respectively in parentheses:

.j Radiation Monitoring System (Section 2.3.5.15; Section 215) -

Fuel Handling System (Section 2.3.5.20; Section 220)

SFP Cooling and Purification System (Section 221 of FSAR)

Temporary Liquid Waste Processing System (Section 255 of FSAR)

Auxiliary Service Water System (Section 2.3.5.27; Section 226)

=

Demineralized Water System (Section 2.3.5.27; Section 227)

Electrical System (Section 2.3.5.29; Section 229)

Ventilation System (Section 2.3.5.31; Section 231) l Fire Protection aad Detection System (Section 2.3.5.32; Section 232) l After the removal of systems and components from an area or building associated with the Spent Fuel Pool island, contaminated structural concrete, steel and other building materials will be decontaminated or removed in a manner consistent with Section 2.3 of the Decommissioning Plan. 'Ihe structures listed below will be decontaminated during decommissioning of the Spent Fuel Poolisland. The applicable sections of the Decommissioning Plan and FSAR that are specific to a structure are shown respectively in parentheses:

y t

Yard Area Crane and Support Structure (Section 2.2.5.43; Section 244)

North Wall ofIon Exchange Pit (Section 2.3.5.44; Section 244)

]

4 Primary Vent Stack (Section 2.3.5.45; Section 245)

Spent Fuel Pool and SFP Building (Section 2.3.5.46; Section 246) j New Fuel Vault (Section 2.3.5.47; Section 247)

Potentially Contaminated Area Warehouse (Section 2.3.5.52; Section 252)

Service Building (Section 2.3.5.54; Section 254)

Meteorological Tower (Section 2.3.5.56; Section 256)

=

Final status surveys will be conducted according to Appendix A to verify that structures t

or open land areas associated with the SFP meet the release criteria. Independent verification of the results by the NRC will allow for the release of the individual surveyed l

structures and open land areas as non-radiologically controlled material available for i

conventional demolition and disposal. In order to facilitate remediation, structures may be released and demolished prior to remediating subfloor soils beneath the structures.

Measures will be implemented as described in Section 4.1.7 of Appendix A to prevent recontamination of surveyed areas prior to final license termination.

l

- 3.4 LICENSE TERMINATION (PHASE 3) i

. The final phase of decommissioning will take place after all spent fuel and GTCC waste (q

is taken ofTsite. In the interim, spent fuel and GTCC waste may be stored either in the

/

I i

3-3 i

+

m

YNPS LICENSE TERMINATION PLAN REVISION 0 Spent Fuel Pool or in an on site dry cask storage facility. Due to uncertainty regarding the Department of Energy's acceptance of spent fuel, a final decision has not been made on the long-term storage method for spent fuel at YNPS. If the spent fuel and GTCC waste is kept in the SFP until it is taken off site, license termination will take place immediately after decontamination and dismantlement of the SFP island. If a dry cask storage facility is constructed and operated, the final phase of deconunissioning will follow the dismantlement and decontraination of this facility.

Decommissioning of the dry cask storage facility would consist primarily of the disposal of the concrete canister ovegacks if they ve not shipped with the spent fuel casks. The overpacks would be designed to minimac neutron activation, and it is anticipated that only a small amount of radioactive waste would be generated. This waste should be able to be disposed of at a very low level radioactive waste disposal site.

Regardless of the spent fuel storage option chosen, the 10 CFR Part 50 license will be terminated after a last stage of final status surveys and independent NRC verification.

Site restoration activities will then be conducted to return the remaining small portion of the site to a " green field" condition.

3.5 REFERENCES

3-1 Yankee Nuclear Power Station Decommissioning Plan, Revision 0.0.

3-2 Yankee Nuclear Power Station Final Safety Analysis Report, June 1995.

O 3-4

L i.

y

. I

'e TABLE 3-1

)

i i

MAY 1997 STATUS OF SYSTEMS, STRUCTURES, AND COMPONENTS DESCRIBED IN

]

i FSAR System, Structure, or Component FSAR Needed for Status Section Wet Fuel j

Storage l

Reactor Vessel 201 No Removed.

Steam Generators 202 No Removed.

Main Coolant System 203 No Removed.

Pressure Control and Relief System 204 No Removed.

Charging and Volume Control System 205 No Removed.

i ChemicalShutdown System 206 No Removed.

Purification System 207 No Removed.

Component Cooling System 208 No Removed.

Primary Plant Corrosion Control System 209 No Removed.

Primary Plant Sample System 210 No Removed.

Waste Disposal System 211 Yes Ori;;inalsystem removed, replaced with Temporary Liquid Waste gystem. [ Note]

Shutdown Cooling System 212 No

' Removed.

Primary Plant Vent and Drain System 213 No Removed.

Emergency Core Cooling System 214 No Removed.

i Radiation Monitoring System 215 Yes Partially removed, in service. [ Note]

VC Ventilation and Purge System 216 No Partially removed, in service.

VC Heating and Cooling System 217 No Removed.

Post Accident Hydrogen Control System 218 No Removed.

Containment isolation System 219 No Removed.

i Fuel Handling Equipment System -

220 Yes Partially removed, in service. [ Note]

SFP Cooling and Purification System 221 Yes Modified for SFPisland. [ Note]

Note:

After all spent fuel and Greater ~llian-Class-C waste has been removed from the Spent Fuel Pool, the p

remaining portions of this system or structure that are in service to support SFP operations will be j (

decontaminated and dismantled in accordance with Section 2.1 of the Decommissioning Plan and Section 200 of the FSAR.

E

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- +,

-n-

-,m w

i.

l t

I TABLE 3-1 (Continued)

O MAY 1997 STATUS OF SYSTEMS, STRUCTURES, AND COMPONENTS DE5 CRIBED IN FSAR

)

Systess, Structure, or Cos.poneet FSAR Needed for Status Section Wet Fuel j

Storage l

Main Steam System 222 No Removed.

4 Feedwater System 223 No Removed.

i Steam Generator Blowdown System 224 No Removed i

j Emergency Feedwater System 225 No Removed.

Service Water System 226 Yes Partially removed; Auxiliary Service j

Water System installed for SFP island. [ Note]

^

j Dcmineralized Water Sy:: tem 227 Yes Partially rem aved, in service. [ Note)

Compressed AirSystem 228 Yes Original system removed, temporary system in service for SFP. [ Note]

ElectricalSystem 229 Yes Partially removed, in service. [ Note]

)

, Heating System 230 No Panially removed.

f Ventilation System 231 Yes Partially removed, in service. [ Note]

Fire Protection and Detection System 232 Yes Partially removed, in service. [Notej Primary Pump Seal Water System 233 No Removed.

Safe Shutdown System 234 No Removed.

j Water Cleanup System 235 No Removed.

I Vapor Container 236 No To be decontaminated in 1997.

Reactor Support Structure 237 No To be decontaminated in 1997.

Vapor Container Polar Crane 238 No To be removed in 1997.

Radiation Shielding 249 No Partially removed /decontaminatti Neutron Shield Tank 240 No Removed.

j Pipe Chases 241 No To be decontaminated in 1997.

Note:

Afler all spent fuel and Greater Than-Class-C waste has been removed from the Spent Fuel Pool, the remaining portions of this system or structure that are in service to support SFT operations will be

- O decontaminated t.nd dismantled in accordance with Section 2.3 of the Decommissioning Plan and Section 200 of the FSAR.

i I

. m.

.- m

~ _..

,,m

i l,

i Q.

TABLE 3-1 (Continued) b' l

MAY 1997 STATUS OF SYSTEMS, STRUCTURES, AND COMPONENTS DESCRIBED IN i

FSAR i

System Siructure,or Component FSAR Needed for Status Section Wet Fuel l

Storage FuciTransfer Chute 242 No To be decontaminated in 1997.

l Yard Area Crane and Support Structure 243 Yes in service. [ Note]

lon Exchange Pit 244 Yes Partial decontamination in 1997, full i

decon after fuel removed from SFP.'

North wall required structurally for i

SFP. [ Note]

l Primary Vent Stack 245 Yes in service. [ Note]

Spent Fuel Pit and Spent Fuel Pit 246 Yes In service. [ Note]

+

Building New Fuel Vault 247 Yes To be decontaminated after fuel removed from SFP. West wall i

required structurally for SFP. (Note]

Primary Auxiliary Building 248 No To be decontaminated in 1997.

i Diesel Generator Building 249 No To be decontaminated in 1997.

7 l

Waste Disposal Building 250 No To be decontaminated in 1997.

Safe Shutdown System Building 251 No To be decontaminated in 1997.

{

Potentially Contaminated Area Storage 252 Yes PCA Bldgs. No. I and 2 to be Buildings 1 and 2 and Warehouse decontaminated in 1998. Warehouse may be used during wet or dry fuel storage.

Compactor Building 253 No To be decontaminated in 1998.

Service Buildbg 254 Yes Portions of building within the Radiation Control Area are in service.

(Note]

1 Miscellaneous Tanks 255 Yes Most removed; one ank to remain for i

storage of demineraded water for SFP. [ Note]

MeteorologicalTower 256' Yes In service. Function will be retained during wet or dry storage.

Note:

After all spent fuel and Greater 'ihan-Class-C waste has been removed from the Spent Fuel Pool, the Os remaining portions of this system or structure that are in service to support SFP operations will be decontaminated and dismantled in accordance with Section 2.3 of the Decommissioning Plan and Section j

200 of the FSAR.

i I

l YNPS LICENSE TERMINATION PLAN REVISION 0 i

t j

V SECTION 4 PLANS FOR SITE REMEDIATION

4.1 INTRODUCTION

i The ultimate goal of decommissioning at the Yankee Nuclear Power Station is to release the site for unrestricted use and retum it to a " green field" condition. Release of the site for unrestricted use requires assurance that any future use of the sis will not result in i

y j

exposing individuals to unacceptable levels of radiation. The site release criteria contained within the Final Status Survey Plan (Appendix A) require that the Total s

Effective Dose Equivalent (TEDE) to the average member of the critical population group from residual contamination be maintained less than 15 mrem / year. In addition, plant related contamination in groundwater and surface water must not exceed the Environmental Protection Agency National Primary Drinking Water Standards for radioactivity. Remediation of some areas of the site will be necessary in order to meet the release criteria. The administrative action levels presented in Section 5.2 of Appendix A determine whether further investigation or remediation is required to ensure that an area meets the site release criteria. The purpose of this section is to identify the remediation methods that may be used, describe the areas on site that may be subject to (q

remediation, and demonstrate that the site release criterion of 15 mrem / year is adequate to

'v ensure that residual levels of radioactivity at YNPS will be As Low As is Reasonably Achievable (ALARA).

4.2 REMEDIATION METHODS Remediation will be performed in accordance with the general and specific decontamination and dismantlement considerations of Section 2.3 of the YNPS Final Safety Analysis Report (Reference 4-1). All areas of the site that do not meet the release criteria will be remediated. In keeping with the principles of ALARA, residual Itvels of radioactivity will be reduced commensurate with the minimization of total risk. For example, because of the effectiveness of some remediation methods, many areas that initially exceed the site release criteria will be remediated to residual dose levels well j

below 15 mrem / year at little or no additional cost compared with the cost to remediate to the 15 mrem / year dose level.

4.2.1 Building and %ructure Surfaces Many decontamination methods exist to remediate the surthces of buildings and structures. Several factors determine the choice of decontamination method for a given application, including the extent of the contaminated area, surface material,,

depth of contamination, and access considerations. Table 200.3 of the FSAR q(j describes the decontamination methods that may be used at YNPS. In addition, 4-1

YNPS LICENSE TERMINATION PLAN REVISION 0 several additional decontamination technologies may be used as described below.

All methods will be utilized with proper regard to the potential for airborne contamination and will be used in conjunction with containment enclosures or HEPA units as necessary in order to minimize airborne activity.

Abrasive Vacuum Blasting with Reevelable Blast Media - This method is highly effective for surface contamination up to a depth of 6 mm. Unlike conventional abrasive blasting, this rcethod uses a recyclable blast media such as steel shot to minimize the volume of waste that is generated.

Concrete Sawing - This method can be used to remove large pieces of

~

contaminated concrete where other methods would be too slow. Cutting may be performed with an abrasive blade or a diamond wire.

Needle Gunning - This method is useful for removing surface contamination in areas that are too small or inaccessible for other methods to be used.

Manual Removal of Building Material - Contaminated building material such as roofing will be manually removed and disposed of at an appropriate low level radioactive waste facility.

4.2.2 Surface Soils and Asphalt As discussed in Section 2, site characterization has shown that radioactive soil contamination in surface soils and asphalt is usually found no deeper than 15 cm from the surface. Soil remediation will involve the removal of all soil necessary to meet the site release criteria. Asphalt will be treated in the same way as soil.

Both soil and asphalt will be removed with excavation equipment, and due care will be taken to prevent the spread of contamination during excavation and handling. As described in Section 4.43 of the YNPS Deconunissioning Environmental Report (Reference 4-2), controls will be in place to minimize the creation of fugitive dust. All contaminated soil will be disposed of at a low-level waste facility, 4.23 Sediment To date, no sediment contamination has been found that exceeds the release criteria, including sediment impacted by licensed discharges monitored as part of the YNPS Radiological Environmental Monitoring Program (REMP). Sediment impacted by licensed discharges is considered acceptable and will not be remediated. Other sediment found to be contaminated from plant operations will be remediated if the levels of contamination exceed the release criteria of 4-2

_~

i s

l 1

YNPS LICENSE TERMINATION PLAN REVISION O

.l

' C

~\\

~ Appendix A. Contaminated sediment may be removed by dredging or by other excavation techniques using a coffer dam or berm.

1 4.2.4 Subfloar Soil l

Where radionuclide concentrations that exceed the action level for remediation are j

found in subfloor soil beneath structures, the concrete floors will be cut or broken up and the underlying soil excavated and removed along with any contaminated i

concrete. Because of the difficulty in excavating beneath an existing structure, remediation of subfloor soil may take place after the structure has been demolished. Remediation may also be performed with the structure intact.

Several factors influence the approach that is used, including structural stability l

issues, the extent and depth of the contamination, the accessability of the work area to excavation equipment, and the need to protect the work area from the j

elements. Ifit is necessary to demolish a structure, the part that is above grade and the accessible parts below grade (interior floors, walls, ceilings) will be

]

decontaminated first. A final status survey will then be performed on these areas to verify that the release criteria of Appendix A have been met. NRC confirmation of the results will release the surveyed parts of the structure for conventional demolition and disposal without the need for additional surveys.

[

After the debris is removed, radiation area controls will be re-established and the

(

radiologically contaminated subfloor soil will be excavated.

4.3

- AREAS TO BE REMEDIATED Site characterization to determine remaining areas that need remediation is ongoing and will continue throughout decommissioning. The site release criteria and the principles of ALARA will be applied on a case by case basis to identify the extent of the remediation.

The following areas either have already been identified for remediation or will undergo further analysis:

Outdoor Radioactive Material Storage Area - This area is located in the yard area to the east of the Potentially Contaminated Area (PCA) Building No. I and is shown in Figure 2-1. This area was used for the storage of radioactive materials i

during plant operation and decommissioning and will be remediated prior to site i

2 release. The area encompasses approximately 460 m and contamination is present to an approximate depth of 15 cm.

Soil and Amhnit Under the Vanor Container - Contaminated soil and asphalt have already been removed from underneath the Equipment Hatch as part of the j

Reactor Pressure Vessel removal project. Additional soil and asphalt will be remediated as necessary if future site characterization finds contamination in excess of the site release criteria.

4-3 1

1 i

l YNPS LICENSE TERMINATION PLAN REVISION 0 01 Soil on North Side of the_Soent Fuel Pit Building - Soil from this area was

=

excavated in 1992 for the installation of a security wall and was found to be contaminated in excess of the site release criteria. Further characterization of this area will be conducted to determine if additional soil requires remediation.

Structural Decontaminations mBuildings and structures which have been identified for remediation include the Vapor Container, Diesel Generator Building, Primary Auxiliary Building, Ion Exchange Pit, Compactor Building, Waste Disposal Building, PCA Warehouse, PCA Storage Building No.1. and the Safe Shutdown System Building. If additional structures are found to be contaminated in excess of the site release criteria, they will be decontaminated when they are no longer required to support fuel storage operations.

Subfloor Soil - Contaminated subfloor soil has been found beneath several buildings in the Radiation Control Area. Soil samples containing radioactivity in excess of the site release criteria have been taken from beneath the Ion Exchange Pit, South Decontamination Pad in the Service Building, and the Decontamination Tub in PCA Building No.1. Subfloor contamination is the subject of ongoing site characterization sampling in the Waste Disposal Building and other structures, and remediation will be undertaken where necessary to meet the site g

release criteria.

W Storm Drain Outfalls - Table 306.2 in the FSAR de Ges various contamination events in the Radiation Control Area over the cour: ;f the plant's life that have resulted in d.ischarges oflow levels of radioactivity to the east and west storm drains. The east storm drain empties into Sherman Pond east of the Screenwell House, and the west storm drain empties into an unnamed tributary of the Deerfield River below the Sherman Dam. Based on the site characterization work to date, the storm drain outfall areas are classified as "Affected" in accordance with the Final Status Survey Plan. However, neither area is expected to require remediation because the concentration of radionuclides is low in comparison with the guideline values that equate to the site release criteria.

4.4 ALARA ANALYSIS The YNPS FSAR requires that an optimization process, based on ALARA, be used to reduce the levels of radioactivity on site commensurate with the minimization of total risk.

The objective of this analysis is to determine if the site release criterion of 15 mrem / year is adequate to ensure that residual levels of radioactivity at YNPS will be ALARA. This analysis will show that, in areas with dose levels already lower than 15 mrem / year for an average member of the critical population group, the benefits of further remediation are not proportionate to the associated costs.

4-4

YNPS LICENSE TERMINATION PLAN REVISION 0 D.Y 4.4.1 Approach-

. This analysis follows the general approach of NUREG-1500 (Reference 4-3),

which describes the ALARA method as a three-step process involving:

4

]

~

(1) characterization of contamination at the facility.

(2) an estimate of the impacts and decommissioning costs at residual dose levels, and (3) an assessment of ALARA based on a comparison of the incremental i

reductions in impacts to the cost of achieving reductions in those impacts.

l 1

The first step involves an accurate determination of the radioactive contamination 4

remaining on site. Site characterization at YNPS is being performed using the j

guidance of NUREG/CR-5849 (Reference 4-4), and results are presented in j

Section 2. The second step in the ALARA approach involves estimating the

. radiological and non-radiological impacts on site workers and the public from the j

decommissioning process and from residual radiation remaining on the site after i-its release. The second step also involves estimating the cost to remediate areas of j

the site. This cost includes the cost of removing, packaging, shipping, and disposing of contaminated material. In this analysis, the impacts and costs of remediating soils and structural surfaces at YNPS are examined in detail. The third step in the ALARA method is to determine if the costs incurred are l

commensurate with the reduction in impacts. For a reduction in radiological impacts, the associated costs and benefits may be compared on a dollar per person-rem basis. In NUREG-1530 (Reference 4-5), the NRC adopts a $2,000 per person-rem conversion factor as a reasonable cost for reducing collective dose i

to the public.

i 4.4.2 Assumptions i

l Several assumptions are made to estimate the impacts and costs of remediation.

These assumptions are conservative in that they increase the benefit assumed from remediation relative to its cost.

First, it is assumed that remediation is 100% effective and removes all contamination from an area. Second, conservatively high population densities are assumed for people living and working on the site afler its release. Third, no l

l cmdit is taken for dilution of the contamination due to weathering of the soil or i

demolition of the buildings. The effect of these three assumptions is to maximize the collective dose assumed to be received by people occupying the site after its release if no remediation were performed. This dose is assumed to be completely O-avoided if remediation is performed.

[

4-5

YNPS LICENSE TERMINATION PLAN REVISION 0 Another conservative assumpti~ n is made by excluding non-radiological impacts o

of remediation, such as the risk to workers and the public from non-radiological accidents during decontamination and transportation activities. Examples of these are conventional industrial accidents and traffic accidents. By considering only radiological effects, the impacts of remediation can be expressed in terms of dose received or avoided by workers and the public. It is also assumed that workers receive no radiation exposure during decontamination activities and that workers and the public receive no exposure during waste transportation. The effect of these assumptions is to consider only the positive impacts of remediation, ignoring any associated risks or other detrimental effects, and thereby weighting the cost / benefit evaluation in favor of remediation.

4.4.3 Soil Remediation For the purpose of estimating soil remediation costs, an unpaved open area of 100 m with radioactive contamination to a depth of 15 cm is considered as a 2

reference area. The 15 m$ volume of contaminated soil is assumed to contain a homogenous concentration of radionuclides which yields a dose of 15 mrem / year using the residential scenario of Reference 4-3. This is a conservative, least costly example that bounds the more expensive cases for sediment, subsurface soil, and subfloor soil. The size of the reference area is suggested by Reference 4-4, which 2

nilows soil contamination to be averaged over an area up to 100 m for the purpose of determining if remediation is required. The depth of contamination is suggested by the site characterization data presented in Section 2.

The cost to remediate the reference area is estimated using unit cost factors based on current cost data at YNPS. Each component of the total cost is calculated by multiplying the volume of contaminated soil by the appropriate unit cost factor.

As shown in Table 4-1, the total remediation cost is the sum of the component costs. For the 100 m reference area, the cost of remediation is $61,400 (COST).

2 Note that because the remediation techniques proposed at YNPS remove layers of the bulk material containing the contamination, the total cost is independent of the initial level of contamination present.

J l

The benefit gained from remediation of the reference area can be quantified in terms of the collective dose avoided by people occupying the site after its release.

If no remediation were performed, the dose that this population would receive from plant-related activity is calculated as:

< Population) x (Initial Dose Rate) x (Mean Life) = Total Collective Dose Since remediation is assumed to be 100% effective, this total collective dose represents the maximum dose savings achieved by remediation. While Yankee 4-6

YNPS LICENSE TF.RMINATION PLAN

- REVISION 0 s

currently has no plans to develop the site for future use, it is conservatively i-assumed that the future population density on site is 20 persons per acre. This 2

equates to 0.5 persons per 100 m,

i Radiological decay is taken into account through the use of a mean life for a given contamination radionuclide distribution. 'Ihe mean life represents the period of exposure at the initial dose rate that will account for all potential exposure until 3

the radioactivity has been eliminated by decay. With a mean life of i

approximately 8 years, Co-60 is responsible for approximately 75% of the dose at YNPS. However, in order to calculate a mean life for this ALARA analysis, it is conservative to use a radionuclide distribution which assumes that 50% of the i

dose results from Co-60 and 50% results from the longer-lived Cs-137 radionuclide. This hypothetical distribution has a mean life of 26 years and bounds the actual radionuclide distributions at YNPS, which decay at faster rates 2

and have shorter mean lives.

2 Therefore, the total collective dose benefit gained by remediating the 100 m reference area from 15 mrem / year to 0 mrem / year is:

p (0.5 persons) x (0.015 rem / year) x (26 years) = 0.20 person-rem (BENEFIT)

Dividing the remediation cost by the dose benefit, a ratio of $310,000 per person-rem is obtained. Despite conservative assumptions, this ratio is much higher than the $2,000 per person-rem value accepted by the NRC. Therefore, remediation of areas with initial dose levels below 15 mrem / year does not achieve a i

commensurate reduction of dose in comparison to the associated cost. At successively lower initial dose levels, the cost of remediation remains constant but i

the dose benefit decreases. In fact, using the cost assumptions made in this section with actual dose level data from site characterization, Table 4-2 illustrates j

that the ratio of dollars to person-rem continues to increase at levels below 15 mrem / year. For example, remediation of all areas on site with dose levels i

+

F between 10 and 15 mrem / year would require the expenditure of $390,000 per person rem of dose benefit. In conclusion, remediation of contaminated soil that is below the site release criterion of 15 mrem / year is not justified by ALARA i

considerations.

3 4.4.4 Structural Surface Remediation For planning purposes,24,155 m (260,000 ft2) of structural surfaces is estimated 2

to be contaminated at YNPS For the purpose of this ALARA analysis, the cost to remediate these surfaces is estimated to be $5.46 million, including labor, 4

equipment, support services, and waste disposal. This figure is based in part on information fro n decontamination equipment vendors and is conservatively low 4-7 r

YNPS LICENSE TERMINATION PLAN REVISION 0 in comparison with the $9.5 million estimate for the remediation of site tures in the 1995 Decommissic,ning Cost Estimate (Reference 4-6). Dividing 2

2

$5.46 million by 24,155 m, the average cost of remediation at YNPS is $226/m,

In order to compare the costs and benefits of remediation, a hypothetical room measuring 10 m by 10 m is considered as an example. The walls are assumed to be contaminated up to a height of 2 m and the floor is assumed to be entirely 2

contaminated. The total contaminated surface area is therefore 180 m. As with soil contamination, the cost to remediate structural surfaces does not vary with the initial dose level, therefore the total cost to remediate the reference contaminated room is:

2 2

($226/m ) x (180 m ) = $40,680 (COST)

Based on NUREG-1496 (Reference 4-7), it is assumed that 225 workers occupy the site for a repowering project. This scenario is chosen because it is more conservative than a scenario in which the plant structures are demolished and the rubble left on site. For this example, it is conservatively assumed that 5 workers are concentrated in the reference contaminated room on a full-time basis. The initial dose rate to an average member of the critical population group is assumed

]

to be 15 mrem / year. Again, a conservative mean life of 26 years is used for the gl contamination radionuclide distribution. Assuming that remediation removes all W

contamination from the room, the dose level is reduced from 15 mrem / year to 0 mrem / year and the total collective dose benefit to the workers is:

(5 persons) x (.015 rem / year) x (26 years) = 2.0 person-rem (BENEFIT) j A ratio of $20,000 per person-rem is obtained by dividing the cost of remediation by its benefit. This ratio is greater than the $2,000 per person-rem conversion factor accepted by the NRC, and remediation of structural surfaces with initial dose levels below 15 mrem / year is notjustified by ALARA. As with soil contamination, the cost to remediate structural surfaces is assumed to be independent of the dose level. However, because the dose benefit decreases at successively lower dose levels, the incremental cost of remediation increases on a per person-rem basis.

It should be noted that the surface contamination limits that will be used at YNPS are conservative in that they equate to doses ofless than 15 mrem / year using the methodology of NUREG-1500. These limits are presented in Table 3-1 of the Final Status Survey Plan.

O 4-8

YNPS LICENSE TERMINATION PLAN REVISION 0 f.

4.4.5. Conclusions Using highly conservative assumptions which increase the assumed benefit of.~

remediation relative to its cost, it has been shown that the benefits of further j

remediation are not proportionate to its costs for soil areas and structural surfaces with initial dose levels less than 15 mrem / year. The soil analysis also bounds cases of contamination in subfloor soil, subsurface soil, and sediment. In all cases, at initial dose levels of 15 mrem / year or less, the ratio of dollars spent to person-rem saved is greater than the NRC-accepted ratio. Yankee will remediate all areas of YNPS necessary to meet the site release criteria. The site release criterion of 15 mrem / year will ensure that residual levels of radioactivity will be 4

reduced on site commensurate with the minimization of total risk. Although conservative surface contamination limits and effective remediation methods will reduce dose levels to below 15 mrem / year in areas which do not initially meet the site release criteria, further remediation of areas with initial dose levels already below 15 mrem / year is notjustified by ALARA considerations.

i 4

4.5 REFERENCES

4-1 Yankee Nuclear Power Station Final Safety Analysis Report, June 1995.

4-2 Yankee Nuclear Power Station Decommissioning Environmental Report, December 1993.

4-3 NUREG-1500, Working Draft Regulatory Guide on Release Criteria for Decommissioning, August 1994.

4-4 NUREG/CR-5849 Manual for Conducting Radiological Surveys In Support of License Termination (Draft Report for Comment), May 1992.

4-5 NUREG-1530, Reassessment of NRC's Dollar Per Person-Rem Conversion Factor Policy.

4-6 Federal Energy Regulatory Commission Docket No. ER95-835-000 Supplemental Testimony of K.J. Heider, August 15,1995.

4 4-7 NUREG-1496, Generic Environmental Impact Statement in Support of Rulemaking on Radiological Criteria for Decommissioning of NRC-Licensed Nuclear Facilities (Draft Report for Comment),' August 1994.

4-9 i

i TABLE 4-1 2

COST TO REMEDIATE SOIL IN 100 m REFERENCE AREA i

Cost Unit Cost Volume Subtotal Factor 2

$500 Soil Removal (labor and equipment)

$33/m3 15 m 2

Packaging

$390/m' 15 m

$5,850 l

Containers

$250/m' 15 m'

$3,750 Transportation

$390/m' 15 m'

$5,850 Bulk Waste Disposal

$3,030/m' 15 m'

$45,450 l

Soil Remediation Total for i

2 100 m Reference Area (COST)

$61,400 5

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1 TABLE 4-2 YNPS SURFACE SOIL REMEDIATION COST PER PERSON-REM AT INCREMENTAL DOSE LEVELS Dose Rate of Remediated Person-Rem Cost

• Dollars per Person-Remediated Areas' Area (m )

Saved 2 Rem 2

2 (mrem / year)

>15 485

>0.95

$300,000

<$310,000 t

c 14.9-10.0 190 0.31

$120,000

$390,000 9.9-5.0 710

'O.69

$440,000

$640,000 4-4.9-3.0 270 0.14

$170,000

$1,200,000 i

i 2.9-1.0 1170 0.30

$720,000

$2,400,000 1

Notes:

l. Dose data from surface soil samples and direct exposure measurements taken as part of site characterization program.

2. Estimated based on site characterization results.

3. Person-rem savings based on a dose rate at the midpoint of the indicated range.

4. Cost data based on $61,400 cost to remediate 100 m area with contamination to depth of 15 2

cm.

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YNPS LICENSE TERMINATION PLAN REVISION 0 SECTION 5 END USE OF THE SITE 5.1 DISCUSSION i

The objective of decommissioning at YNPS is to remove the nuclear theility safely from service and to reduce residual radioactivity to a level that permits release of the site for unrestricted use. Decommissioning will be completed in phases as areas that are no longer necessary to support remaining operations become available for decontamination and dismantlement. After final status surveys and NRC verification, individual surveyed structures and open land areas will be released as non-radiologically controlled material for conventional demolition and disposal. YAEC will maintain control over the entire site until termination of the 10 CFR Part 50 license.

Site restoration activities will take place as structures and open land areas of the site are released as non-radiologically controlled material. Site restoration will be completed following license termination. As described in Section 1.2 of the Decommissioning Plan (Reference 5-1), all building foundations will be back filled with concrete rubble and i

structural fill. Site areas will be graded and landscaped as necessary to restore the site to l

a " green field" condition.

i

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5.2 REFERE.NCES 5-1 Yankee Nuclear Power Station Decommissioning Plan, Revision 0.0 i

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5-1 1

I YNPS LICENSE TERMINATION PLAN REVISION O I

SECTION 6 ESTIMATE OF REMAINING DECOMMISSIONING COSTS l

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j 6.1-INTRODUCTION AND BACKGROUND The purpose of this section is to present an estimate of remaining costs to retum the YNPS site to a " green field" that is based on the 1995 cost study submitted to the Federal Energy Regulatory Commission (FERC). Total decommissioning costs remaining as of l

January 1,1997 are $235.0 million in constant value 1995 dollars. This section also l

discusses the funding assurance provided by Yankee's power contracts, which obligate the purchasers for the full cost of decommissioning.

e e

3 The YNPS Decommissioning Plan (Reference 6-1) was submitted to the NRC in December 1993 and included a cost study for operating the facility through a safe storage 3

4 period, decommissioning the facility, restoring the site to a " green field," and storing i

3 spent fuel until its transfer to the Department of Energy. In October 1994, Yankee i

completed a revised cost study (Reference 6-2) to assist the NRC in its review of the Decommissioning Plan and to fulfill a commitment to FERC. This 1994 cost study was based on the assumption that dismantlement activities would not begin until a low-level i

radioactive waste disposal site became available to Yankee in 2003.

I In June 1995, the State of South Carolina re-opened the low-level waste facility in i

Bamwell, South Carolina to radioactive waste generators throughout the United States.

In response, Yankee updated 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 (Reference 6-3). In addition to the earlier availability of a low-level radioactive waste facility, the overall decommissioning cost estimate was affected by lower unit costs for radioactive waste disposal, earlier availability of a dry

~

cask storage facility, and revised operating costs. The fundamental assumptions of the 1995 Cost Study remain valid, and the study is appropriate to use for an estimate of e

remaining decommissioning costs from 1997 until the end of fuel storage on site.

6.2 REM AINING DECOMMISSIONING COST ESTIMATE f

The 1995 Cost Study is a site-specific cost study that was' developed by adjusting 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 is $303.2 million. In addition, as part of the final December 1995 FERC settlement (Reference 6-4), 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 J

i 6-1 l

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YNPS LICENSE TERMIFATION PLAN REVISION 0 (1995 dollars). A summary of the 1995 Cost Study and FERC settlement is presented in Table 6-1.

As of 1997, the major assumptions of the 1995 Cost Study remain valid and the study continues to be representative of total remaining decommissioning costs. Like the 1995 Cost Study, the scope of the current decommissioning effort encompasses the dismantlement and decontamination of plant systems and structures, spent fuel storage, and site restoration. Barnwell continues to be available to Yankee for the disposal of low-level radioactive waste. Although minor changes have been made to the decommissioning schedule, the sequence of major decommissioning phases in the 1995 Cost Study is also in agreement with current plans.

Due to the adjudicatory hold on major decommissioning activities throughout much of 1996 and the slow progress of the Department of Energy towards acceptance of spent fuel, the schedules of some decommissioning activities have been affected. However, while the timing of actual activities and expenditures may be different from that of the 1995 Cost Study, the total cost of remaining activities is not expected to differ significantly from the cost study. Importantly,2018 is still the assumed end date for spent fuel storage and dry cask storage facility decommissioning. The current schedule is shown in Figure 1-1.

Table 6-2 updates the 1995 Cost Study to include the actual costs incurred in 1995 and 1996 and the current budget for 1997. The main difference between the 1995 Cost Study and the current estimate is the timing of expenditures within the period from 1995 to 1998. Additionally, $18.7 million, which is not anticipated to be spent before the end of 1997, is assumed to be spent during 1998-2018. This deferred expenditure is 6.1% of the total cost estimate and represents unspent contingency that may be used to defray unexpected future costs. Contingency applied to various decommissioning costs in the 1995 Cost Study ranges from 5% to 50%, and the overall decommissioning project contingency is approximately 12%. Figure 6-1 shows a projection of funds available for each remaining year of decommissioning if the deferred contingency is distributed among the years 1998-2018. Escalation rates contained in Attachment B of the 1995 FERC settlement are used to project cash flows through time.

63 DECOMMISSIONING FUNDING Power from YNPS was sold at wholesale to the ten New England utilities that own Yankee Atomic Electric Company under Power Contracts filed as rate schedules with FERC. Under these Power Contracts, each of the purchasers agreed to purchase a percentage of the capacity and output of YNPS and to pay a like percentage of Yankee's costs and expenses, including decommissioning costs. In March 1995, Yankee submitted a wholesale rate application to FERC based on the 1994 Cost Study. After the Bamwell low-level radioactive waste disposal facility became available to Yankee, supplemental 6-2

YNPS LICENSE TERMINATION PLAN REVISION 0

.O testimony was filed in August 1995 that included the 1995 Cost Study. This testimony V

indicated an estimated cost of $303.2 million in 1995 dollars to complete the remaining decommissioning activities at YNPS. The final FERC settlement approved collections 1

between November 1995 and June 2000 to fund the amount of $306.4 million, including a $3.2 million adjustment for adjudicatory delays during re-approval of the Decommissioning Plan.

)

)

As described in the Decommissioning Plan, Yankee is currently collecting decommissioning funds through its Power Contracts. The collections are 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 10 CFR 50.75(e)(1)(ii) and a copy of the tmst document has been provided to the NRC (Reference 6-5). ' Die Power Contracts obligate the purchasers for i

the full costs of decommissioning YNPS including spent fuel. The FERC orders received 1

a 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. A new submittal to j

FERC is currently required by January 1,2000 (Reference 6-4).

j i

6.4 REFERENCES

6-1 Yankee Nuclear Power Station Decommissioning Plan, Revision 0.0.

i 6-2 BYR 94-064, Updated Decommissioning Study, H. T. Tracy (YAEC) to M. B.

Fainile (USNRC), dated October 26,1994.

6-3 Federal Energy Regulatory Commission Docket No. ER95-835-000 Supplemental Testimony of K. J. Heider, August 15,1995.

6-4 Federal Energy Regulatory Commission Docket No. ER95-835-000 Offer of Settlement, December 29,1995.

l 6-5 BYR 90-102, Decommissioning Funding Assurance Report and Certification, H.

T. Tracy to Document Control Desk (NRC), July 25,1990.

9 6-3

-[

TABLE 6-1 1995 COST STUDY

SUMMARY

AND FERC SETTLEMENT -

i i

i Period Period Description Dates Cost in 1995 Designation

$ (000's) 1A 1995 Dismantlement and Decontamination 1/95 - 12/95

$41,638

[

j 1B 1996 Dismantlement and Decontamination 1/96 - 12/96

$66,888 i

IC 1997 Dismantlement and Decontamination 1/97 - 12/97

$47,487 2

Load Dry Cask Storage Facility 1/98 - 12/98

$21,432

[

3 Dismantle and Decontaminate SFP Island 1/99 - 9/99

$19,825 4

Release of SFP Island Structures 10/99 -3/00

$5,315 l

5 Site Restoration 4/00 - 3/01

$34,969 6

Dry Cask Storage Facility Operation 4/01 - 12/18

$65,615 FERC Adjustment for Adjudicatory Delays

$3,182 TOTAL COST TO DECOMMISSION YNPS 1/95 - 12/18

$306,352 FROM JAN.1995 TO DEC. 2018 4

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TABLE 6-2 CURRENT DECOMMISSIONING ESTIMATE INCLUDING 1995 AND 1996 ACTUAL COSTS AND 1997 BUDGET Current Estimate 1995 $'s (000's)

T 1995 Activities Completed

$30,543 1996 Activities Completed

$40,839 1997 Budget

$62.564 1995-1997 Subtotal

$133,946 1998 Estimate

$28,002 Deferred Contingency

$18,679 1999-2018 Estimate

$125,724 i

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Total Decommissioning Costs 1995-2018

$306,352 O

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YNPS LICENSE TERMINATION PLAN REVISION 0 j

i b(3 SECTION 7 ENVIRONMENTAL CHANGE ASSOCIATED WITH LICENSE TERMINATION ACTIVITIES 7.1 DISCUSSION The Yankee Nuclear Power Station Decommissioning Environmental Report (Reference 7-1), prepared and submitted in conjunction with the YNPS Decommissioning Plan and approved by NRC, concluded that decommissioning activities wuild be accomplished with no significant adverse environmental impacts. This conclusion is demonstrated in that:

No site specific factors pertaining to YNPS have altered the conclusions of the Generic Environmental Impact Statement (GEIS) in NUREG-0586 (Reference 7-2).

Radiation dose to the public has been minimal.

Average annual radiation dose to decommissioning workers has been less than that during operations.

v-Decommissioning has not been a health or safety problem and continues to have a positive environmental impact.

These conditions will remain valid through the period oflicense termination.

Furthermore, non-radiological environmental impacts will continue to be temporary and not significant.

The total radiation exposure impact for decommissioning was estimated in the Decommissioning Plan (Reference 7-3) to be approximately 744 person-rem. This estimate was re-evaluated in 1996, resulting in a lower value of 580 person-rem (Reference 7-4). To date, the actual exposure (through 12/31/96) for decommissioning activities is 457 person-rem. For the remaining decommissioning and fuel storage activities, the estimated exposure is approximately 115 person-rem. These values are much less than the exposure estimate of the GEIS for decommissioning a pressurized water reactor.

The public exposure due to radiological efiluents will continue to remain well below the limits of 10 CFR Part 20 and the ALARA dose objectives of 10 CFR Part 50, Appendix I.

This conclusion is supported by the data submitted to the NRC in the YNPS Semi-Ammal Efiluent Release Reports in which individual doses to members of the public are (p) calculated for station liquid and gaseous efIluents.

v 7-3

YNPS LICENSE TERMINATION PLAN REVISION 0 Finally, the impact due to disposal of the remaining YNPS low level radioactive waste (LLW) is unchanged. The Decommissioning Plan contains an estimate of 3,738 m' (132,000 fl') of LLW. As of December 1996, an approximate volume equivalent to the burial of 2,265 m' (80,000 ft') of LLW had been shipped in 189 shipments.

Approximately one-half of this material was shipped to reprocessing facilities. The waste volume that remains at YNPS is well bounded by the original estimate and by the GEIS.

Since approval of the Decommissioning Plan and the Decommissioning Environmental Report, Yankee has identified the presence of solid Polychlorinated Biphenyls (PCB's) 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 PCB's will be managed according to all applicable federal and state regulations.

7.2 REFERENCES

7-1 Yankee Nuclear Power Station Decommissioning Environmental Report, December 1993.

7-2 NUREG-0586, Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities, August 1988.

O1 7-3 Yankee Nuclear Power Station Decommissioning Plan, Revision 0.0.

7-4 USNRC Atomic Safety and Licensing Board Docket No. 50-029-DCOM, Supplemental Affidavit of Russell A. Mellor, September 3,1996.

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APPENDIX A FINAL STATUS SURVEY PLAN

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'O FINAL STATUS SURVEY PLAN FOR SITE RELEASE ANKEE Yankee Nuclear Power Station i

.O 1

Revision 0 May 1997 Prepared By:

Yankee Atomic Electric Company 580 Main Street Bolton, Massachusetts 01740

YNPS FINAL STATUS SURVEY PLAN REVISION 0 O

TABLE OF CONTENTS Section Title Page EXECUTIVE

SUMMARY

................................... A-vii 1.0 HISTORICAL BACKGROUND................................ A-1 2.0 SITE INFORM ATION..................................... A-3 2.1 Site Description....................

. A-3 2.2 Site Conditions at Time of Final Status Survey.................. A-3 3.0 FIN AL STATUS SURVEY OVERVIEW......................... A-7 3.1 Survey Objectives..................................... A-7 3.2 Scope of the Final Status Survey A-7 3.3 Identity of Contaminants................................ A-8 3.3.1 Building Surfaces and Systems........................ A-8 3.3.2 Soil and Groundwater.............................. A-9 3.4 Determination of Guideline Values......................... A-9 3.4.1' Surface Contamination Guideline Values

.................A-9 3.4.2 Soil, Sediment and Bulk Material Activity Guideline Values.... A-11 3.5 Organization and Responsibilities.........................

A-11 3.6 Quality Assurance..............................

A-11 3.6.1 Instrumentation Selection, Calibration and Operation A-12 3.6.2 Survey Documentation.............

A-12 3.6.3 Quality Control - Replicate Surveys................... A-12 3.6.4 Written Procedures............

..... A-13 3.6.5 Chain of Custody

...... A-13 3.6.6 Records Management............................ A-13 3.6.7 Independent Review of Survey Results................. A-13 3.6.8 Access Control of Surveyed Areas and Systems..........

A-13 3.6.9 Control of Vendor-Supplied Services.................,. A-13 3.7 Training......................................... A-14

...... A-14 3.8 Laboratory Services 3.9 Sched ule................................

....... A-14 3.10 Final Status Survey Reporting........................... A-15 4.0 SURVEY PLAN AND PROCEDURES A-17 4.1 General Survey Plan................................. A-17 4.1.1 History File Preparation.

A-17 4.1.2 Initial Area Classification.....

.... A-18 4.1.3 Walkdown...........

A-20 4.1.4 Turnover for Final Status Survey..

.... A-20 A-ii

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YNPS FINAL STATUS SURVEY PLAN REVISION O

~

TABIR OF CONTENTS -

~

(Continued) l t

Section Title Eage 5

4.1.5 Investigation.................................. A-21 4.1.6 Reclassification................................. A-21 4.1.7 Access Control Measures.......................... A-21

^

4.2 Initial Area Classification of the YNPS Site................... A-22 4.3 Measurement Frequency

...............................A-23 4.3.1 Structures.................................... A-23 4.3.2. Plant Systems.................................. A-25

'4.3.3 Open land Areas................................ A-26 4.4 Survey Methods.................................... A-29 4.4.1 Scan Surveys.................................. A-29 4.4.2 Surface Contamination Measurements.................. A-29 t'

4.4.3 Exposure Rate Measurements........................ A-29 4.4.4 Soil and Groundwater Sampling...................... A-30 4.4.5 Special Measurements and Samples.................... A-30 l

4.4.6 Sampling for Hard to Detect Nuclides (HTDN)............ A-32 4.5 Background Level Determinations......................... A.

(

4.5.1 Total Surface Beta-Gamma Measurements............... A-3 3 1

4.5.2 Removable Surface Beta-Gamma Measurements............ A-33 4.5.3 Soil and Sediment Radionuclide Measurements............ A-3 3

{

4.5.4 Gamma Exposure Rate Measurements.................. A-33 4.5.5 Specialized Measurements.......................... A-34 l

4.5.6 Verification of Background Measurement Population........ A-34 4.6 Instramentation

.....................................A-34 4.6.1 Minimum Detectable Concentration Calculation............ A-35 4.6.2 Detector Sensitivity.............................. A-35 4.6.3 Calibration and Maintenance........................ A-36 l

4.7 Survey Areas, Units, Locations and Points.................... A-36 l

4.7.1 Survey Areas and Survey Units...................... A-36 4.7.2 Survey Locations and Survey Points................... A-37 L

5.0 DATA PROCESSING AND INTERPRETATION.................... A-41 l

5.1 Conversion of Measurements to Reporting Units................ A-41 5.1.1 - Total Surface Contamination......................., A-41 5.1.2 Removable Surface Contamination.................... A-41 51.3 Exposure Rate

...............................A-42 5.1.4 Soil and other Bulk Samples........................ A-42 5.2

' Comparison with Administrative Action Levels................. A-42 e

(]m 5.2.1 Mean and Standard Deviation....................... A-4 3 5.2.2-Elevated Areas of Contamination..................... A-44 A-lii I

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YNPS FINAL STATUS SURVEY PLAN RE',ISION 0 TABLE OF CONTENTS (Continued)

Section Title Eage 5.2.3 Calculation of the 95% Cordidence Interval of the Mean.....

A-45 5.2.4 Processing ofInvestigation Survey Data..

A-46

............ A-47 5.3 ALARA Analysis....................

6.0 FINAL STATUS SURVEY REPORTS.......................

A-50 6.1 Topical Outline A-50 A-50 6.2 Reporting of Survey Findings 6.2.1 Detail Data Reporting for Each Survey Area....

...... A-50

.... A-50 6.2.2 Summary Data Reporting for Each Survey Area..

6.3 Final TEDE Evaluation................................ A-51

7.0 REFERENCES

..... A-52 7.1 Survey Plan References...

....... A-52 7.2 Procedure References A-53 8.0 G LO SS ARY.....................................

A-54 1

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A-iv

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-YNPS FINAL STATUS SURVEY PLAN REVISION 0 LIST OF TABLES Table Title Page 3.1 Surface Contamination Limits............................. A-16 1

4.1 Final Status Survey Instrument Sununary....................... A-38 4.2 Nominal Measurement Detection Sensitivities.................... A-39

]

5.1 Summary of Action Levels for Reclassification and Remediation........ A-48 i

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1 YNPS FINAL STATUS SURVEY PLAN REVISION 0 0:

LIST OF FIGURES Egute Title lhe 2.1 Facility Layout......................................... A-4 2.2 Radiation Control Area (RCA)..

................ A -5 2.3 Initial Classification of YNPS Site

. A-6 4.1 Survey Area Classification Process......................

A-40 5.1 Interpreting and Comparing Survey Data with Guideline Values A-49 9

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A-vi

4 YNPS FINAL STATUS SURVEY PLAN REVISION 0 O

EXECUTIVE

SUMMARY

This document describes the methods to be used by the Yankee Atomic Electric Company (YAEC) to demonstrate that radiation and radioactive contamination levels at the Yankee Nuclear Power Station (YNPS) have been reduced to levels below criteria established for unrestricted use. This plan has been developed in accordance with the YNPS Final Safety Analysis Report (FSAR) which incorporates the YNPS Decommissioning Plan.

This plan has been developed as administrative guidance and is intended to provide the basis for the implementing procedures governing the conduct of the final status survey.

The methods described have been derived from regulatory guidance, specifically Regulatory Guide 1.86 " Termination of Operating Licenses for Nuclear Reactors" (Reference 1), Draft NUREG/CR-5849 " Manual for Conducting Radiological Surveys in Support of License Termination" (Reference 2) and Draft NUREG-1500, " Working Draft Regulatory Guide on Release Criteria for Decommissioning: NRC Staff's Draft for Comment" (Reference 3). Recent reactor faci 2y decommissioning experience at the Fort St. Vrain Nuclear Station and the Shoreham Nuclear Power Station were also incorporated as appropriate.

The site will be divided into one of three types of survey areas: structures, systems or open land areas. These areas will then be classified as Affected or Unaffected based upon the area's radiological history. Survey packages will then be generated that define the survey methodology and intensity, based upon its classification. After collection, radiological data in a Avey area will be analyzed and the results compared to the release criteria. These s.trvey areas are the basic entities for data analysis.

Every survey area on site must meet the release criteria with a 95% confidence level.

This plan also describes the methods used to determine a survey instrument's response to unaffected building materials and soils for the purposes of subtracting background.

It also describes the management controls over all aspects of the project which will ensure that the Plan is implemented properly.

The YNPS Final Status Survey Plan is part of the License Termination Plan which is considered a supplement to the FSAR in accordance with 10CFR50.82(a)(9)(i).

Therefore, revisions to this plan may be implemented without prior NRC approval provided the changes do not:

violate the criteria of Section 1.3 of the License Termination Plan, a

reduce the required st'rvey frequency for a specific classification of survey area, increase the action levels for reclassification or remediation, or

=

A-vii

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YNPS FINAL STATUS SURVEY PLAN REVISION 0 i

affect the methodology for statistical treatment of survey data in a manner which reduces une confidence that the site meets the criteria for unrestricted use.

For the definitions of terminology used in the final ratus survey plan, please refer to Section 8.0, " Glossary."

i O

A-viii

YNPS FINAL STATUS oURVEY PLAN REVISION 0 0

1.0 HISTORICAL BACKGROUND Yankee Nuclear Power Station (YNPS) achieved initial criticality in 1960 and began commercial operations in 1961. The Nuclear Steam Supply System was 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.

On February 26,1992, the Yankee Atomic Electric Company (YAEC) Board of Directors de !d:d to cease poser operations permanently at YNPS. This decision was based on the following factors:

Economic analyses indicated that shutdown of the plant before expiration of its e

Nuclear Regulatory Commission (NRC) operating license in July 2000 could produce a substantial savings to the electricity purchasers.

Significant regulatory uncertainty associated with the timing and cost of the completion of the NRC's review of the integrity of the YNPS Reactor Pressure Vessel.

p

(

On August 5,1992, the NRC amended the YNPS Facility Operating License (DPR-3) to a possession only status. Combined with other amendments and program changes, this formed the basis of the Decommissioning Plan. In December 1993, the plan was j

submitted by YAEC in accordance with the requirements of 10 CFR 50.82(a), which required the submittsi of a proposed decommissioning plan within two years of the permanent cessation of operations. The Decommissioning Plan also incorporated the Spent Fuel Management P!rn required by 10 CFR 50.54(bb). The Decommissioning Plan was approved on February 14, 1995. Due to subsequent litigation, the NRC suspended its approval of the Plan in October 1995; the Plan was finally re-approved on October 28,1996, i

As required by a commitment to the NRC, the Decommissioning Plan was incorporated into the June 1995 revision of the YNPS Final Safety Analysis Report (FSAR). Under

]

the most recent revision of 10 CFR 50.82, an approved decommissioning plan is 1

considered to be the licensee's post-shutdown decommissioning activities report (PSDAR). For YNPS, the terms " Decommissioning Plan" and "PSDAR" may be used interchangeably.

As of May 1997, the majority of the systems and components not required to support the storage of spent fuel have been dismantled and disposed of off-site in accordance A-1

YNPS FINAL STATUS SURVEY PLAN REVISION 0 O

with the YNPS Decommissioning Plan and FSAR. The Main Coolant System has been removed, including the Steam Generators, Pressurizer, and large diameter piping.

The Reactor Pressure Vessel has been removed from containment and shipped to the Barnwell low level waste facility in an NRC-approved transportation cask. Suuctural decontamination work has begun in the Ion Exchange Pit and Waste Disposal Building following the cleanmut of systems and components from these areas. The Spent Fuel Pool and other systems associated with fuel storage have been electrically and mechanically isolated to create a Spent Fuel Pool " island" that will not be adversely impacted by ongoing decommissioning activities. Radiological characterization of YNPS has been ongoing since 1992 and involves survey of systems, structures, and the environment.

Title 10CFR50.82(a)(9)(ii) requires that detailed plans for a final status survey be submitted as part of a License Teimination Plan. This survey will be performed to ensure that the site, including any remaining structures, meets the unrestricted release criteria in order to support termination of the Part 50 License. This Final Status Survey Plan describes the method for conducting the final status survey. Independent verification of survey results will be impleraented through the NRC.

O i

A-2

YNPS FINAL STATUS SURVEY PLAN REVISION 0 i

J l

2.0.

SITE INFORMATION 4

2.1 Site Description YNPS is located in the Berkshire Hills of Franklin County on a 2200 acre site along i

the Deerfield River in the towns of Rowe and Monroe Bridge, Massachusetts. About 10 acres of the site were developed for plant use. The site is at the bottom of a decp valley along the river at the southeast corner of Sherman Reservoir. The area sur-

~ !

rounding the site is mostly wooded with very steep slopes on both sides of the

. Deerfield River. The hills on either side of the site rise about 1000 feet above the river level within one mile and extend from 12 miles north to 8 miles south southeast of the i

site. ~ Sherman Reservoir served as the source of cooling water for the plant.

Data on the levels of radiation and radioactive material in the environs have been j

collected as part of the annual Radiological Environmental Monitoring Program (REMP). The REMP will be continued throughout decommissioning.

2 l

Figures 2.1 and 2.2 show the layout of the overall facility and the Radiation Control l

Area (RCA), respectively.

l 2.2 Site Conditions at Time of Final Status Survey j.

At the time of final status survey, the structures will be largely intact. To date, essentially all contaminated systems and components not supporting the storage of spent fuel have been removed from the site. The remaining contaminated systems and i

structures will be dismantled or decontaminated prior to initiation of the survey in those areas. With spent fuel still in the Spent Fuel Pool, and contingent on having a low enough background radiation level, a final status survey will be completed on structures that do not support Spent Fuel Pool operation. A final status survey will be completed on the remaining structures after fuel is removed from the Spent Fuel Pool.

If the fuel is transferred to a dry cask storage facility, a final status survey will be performed on that facility after the fuel is transferred offsite.

i t

)

It A-3'

YNPS FINAL STATUS SURVEY PLAN REVISION 0 O

Kev to Features in Ficure 2.1 (Facility Layout Man):

01. Vapor Container

11. Primary Auxihary Building

02. Turbine Buildmg

12. Fuel Storage

03. Service Buildmg

13. lon Exchange Pit w

04 Warehouse

14. Compactor Buildmg

05. Screenwell House

15. Waste Das sal Building u

N 01 Discharge Structure

16. PCA use

07. Gate House

17. PCA Storage Buildin No.1

08. Switchyard

18. PCA Storage Buildm No. 2

09. Trainmg Center

19. Pole Barn

10. Diebel Generator Building

20. Safe Shutdown System Buildmg

21. East Stctm Drain Outfall M

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FIGURE 2.1 Eacility La.ynut A-4

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YNPS FINAL STATUS SURVEY PLAN REVISION 0 m.

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Radiation Control Area (RCA) 9)

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A-5

i i

YNPS FINAL STATUS SURVEY PLAN REVISION 0 U

u i

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Visitors Center m snown:

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Other Parking Lots m chown) l UNAFFECTED AREA West Storm Drain OutfaR m sno-as l

AFFECTED AREA j

Remainder of Yankee property m snown:

j la classified as NON-lMPACTED 1

l l

FIGURE 2.3 Initial Classification of YNPS Site l

A-6 1

4 REVISION 0 YNPS FINAL STATUS SURVEY PLAN

...p 3.0 FINAL STATUS SURVEY OVERVIE"I j

3.1 Survey Objectives The final status survey is designed to demonstrate that licensed radioactive materials have been removed from Yankee Nuclear Power Station (YNPS) facilities and prcperty to the extent that residual levels of radioactive contamination are below the release

}

criteria approved by the NRC. The site release criteria are as follows:

i Surface contamination must not exceed the values presented in Table 1 of e

Regulatory Guide 1.86 for average, maximum and removable contamination.

These limits have been adjusted, as described in this Plan, for hard to detect radioauclides.

Exposure rates from gamma emitting radionuclides, created as a result of e-reactor operation and located in soil and activated materials including concrete, components, and structures must not exceed an average 5 R per hour above i

I natural background measured one meter from the surface with a maximum of 10 R per hour at any point of measurement.

The Total Effective Dose Equivalent (TEDE) to the average member of the critical population group from residual contamination must be maintained below 15 mrem per year. ALARA principles will be used to reduce the residual levels of radioactivity on-site commensurate with the minimization of total risk. (See Section 4 of the License Termination Plan)

The contamination of groundwater and surface water as the result of site work e

must not exceed the Environmental Protection Agency regulation 40CFR Part 141, National Primary Drinking Water Standards for radioactivity.

L 3.2 Scope of the Final Status Survey The final status survey will include all structures, surfaces and systems identified as contaminated or potentially contaminated. The survey will include:

Surveys within the Radiation Control Area (RCA), including pavement, soil, j

groundwater, and structures such as the Vapor Container, PCA Warehouse, and i

Primary Auxiliary, Spent Fuel Storage, Waste Disposal, and Compactor 7

Buildings.

i nU P

A-7 1

YNPS FINAL STATUS SURVEY PLAN REVISION 0 0

Areas outside the RCA but within the Owner Controlled Fence including pavement, soil, groundwater, and secondary side structures such as the Turbine and Service Buildings and office areas.

Structures, pavement, soil, groundwater, and surface water outside the Owner Controlled Fence, not to include those areas identified as "non-impacted" (see Section 4.0).

The final status survey will not include sediment that was impacted by monitored and licensed liquid effluent discharges, and which has been and continues to be monitored as part of the Radiological Environmental Monitoring Program (REMP). Any other areas of sediment found to be contaminated from plant operations will be evaluated against the site release criteria, and surveyed in accordance with this Plan.

3.3 Identity of Contaminants 3.3.1 Building Surfaces and Systems Residual contamination on plant piping and building surfaces is the result of the deposition of fission and activation products. Radioactivity found in the bioshield concrete is the result of contamination and neutron activation. During plant operation, many samples were taken from process and waste streams to determine radionuclide distributions. Prior to and during the scoping and characterization surveys, additional samples of process pipes and smears were obtained to determine the contamination radionuclide distributions. In addition, an analysis was performed to estimate radioactivity levels and radionuclide distributions in the reactor vessel, neutron shield tank and bioshield concrete.

A summary of the analyses of the contaminated and activated materials was presented in the YNPS Decommissioning Plan. (Reference 4, Section 3.1)

Three radionuclides, Fe-55, Ni-63 and Co-60, comprise approximately 98 percent of the radioactivity in the structures and systems. Although the Decommissioning Plan provided radionuclide distributions for the reactor vessel and neutron shield tank, they will have been removed at the time of final status survey. Only a portion of the bioshield concrete wall will remain. Based upon the neutron activation analyses, four radionuclides, H-3, Fe-55, Co-60 and Eu-152, comprised approximately 97 percent of the radioactivity remaining in the activated bioshield concrete (as of January 1994). As the plant dismantlement and decontamination continues, additional samples will be analyzed to verify and correct, if necessary, the radionuclide distributions in the remaining structures and systems.

i A-8

i YNPS FINAL STATUS SURVEY PLAN REVISION 0 O

L 3.3.2 Soil and Groundwater,

Surface soil has been sampled and analyzed for radionuclides. The predominant plant-related, gamma emitting radionuclide is Co-60. This radionuclide has been sporadically detected in the top six inches (15 cm) of on-site soil. Other radionuclides that have been detected are Cs-137, Ag-108m and to a lesser degree, Cs-134 and Mn-54.

Groundwater samples have been collected and analyzed for beta and gamma niiliting radionuclides. The primary contaminant is H-3, which has been found in monitoring wells located between the Ion Exchange Pit and Sherman Spring.

The maximum H-3 radioactivity concentration detected during decommissioning was 8,000 pCi/1. This value is well below the 40CFR141.16 limit of 20,000 pCi/1. H-3 concentrations in groundwater will be monitored throughout the dismantlement and decontamination process.

i 3.4 Determination of Guideline Values The contamination levels of all structures, systems and open land areas remaining on site at the time of the final status survey will be compared to guideline values (GLVs).

i V

Guideline values are those residual radioactivity levels that equate to the site release criteria for that particular pathway or measurement. The values to be used at YNPS 3

are described in the following sections.

t 3.4.1 Surface Contamination Guideline Values The surface contamination limits to be used for beta and gamma emitters (excluding H-3, Fe-55 and Ni-63), as specified in Table 1 of Regulatory Guide 2

1.86 and Table 3.1 of this plan are 5,000 dpm/100 cm average,15,000 2

2 dpm/100 cm maximum, and 1,000 dpm/100 cm removable. In addition to these, YAEC will use the following for H-3, Fe-55 and Ni-63: 200,000 2

2 dpm/100 cm average,600,000 dpm/100 cm maximum, and 40,000 dpm/100 2

cm removable (see Table 3.1).

The surface contamination limits for H-3 and Fe-55 were applied at the decommissioning projects at Fort St. Vrain Nuclear Station (FSV) and Shoreham Nuclear Power Station, and their justification was provided in SECY-94-145_(Reference 5). YAEC provided the average and maximum limits in response (Reference 6) to an NRC request for additional information following the submittal of the YNPS Decommissioning Plan. The increase in the total surface contamination limits, as desenbed in SECY-94-145, would make the tisk from the H-3 and Fe-55 similar to the risk from other radionuclides in Table 1 of Regulatory Guide 1.86.

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YNPS FINAL STATUS SURVEY PLAN REVISION 0 0

The removable contamination limit for H-3 and Fe-55 has been set by YAEC to 2

20% of the average contamination limit, or 40,000 dpm/100 cm. This limit is consistent with the other radionuclides in Table 1 of R.G.1.86 and is based on the NRC's Safety Evaluation (Reference 7) of FSV's request to raise the removable contamination level to this limit.

Since Ni-63 comprises a significant fraction of the radionuclide inventory at YNPS (15 to 50%), and because its risk is comparable to that of Fe-55, YAEC will use the same average, maximum and removable surface contamination limits as for Fe-55 and H-3.

The site specific GLVs for removable and average surface contamination will be calculated using Equation 3.1 (derived from Equation A-2 in Draft NUREG/CR-5849, Appendix A):

(3.1)

GLV =

f f

f.

1 i

2

+

+...

1 2

aj 4

where:

j Guideline Value, i.e., the acceptable level of residual surface GLV

=

contamination for radionuclides nonnally detected during field measurement Fraction of the total activity contributed by each radionuclide f,

=

Surface contamination limit for each radionuclide G,

=

Detectable radionuclide fraction.

F

=

Affected Area Guideline Values: For affected survey areas, GLVs will be based upon previously determined radionuclide distributions for specific areas.

For all other areas, where the radionuclide distribution has not been determined, the most conservative distribution, resulting in the lowest GLV of those j

specified areas, will be used. The distributions will be based on all of the radionuclides identified in composite samples from the specific areas, collected both during power operation and during decommissioning. If new radionuclide distribution data is obtained during the completion of decommissioning, and A-10

l j

l YNPS FINAL STATUS SURVEY PLAN REVISION 0 (3

Q determined to be more appropriate for use, the GLVs may be re-evaluated and altered during the course of the final status survey.

Total surface and removable contamination limits ("G" in Equation 3.1) will be taken from the " average" and " removable" columns of Table 3.1, respectively.

The maximum surface contamination limit is set at three times the average surface contamination GLV.

Unnffected Area Guideline Values: Plant-related radionuclides are not expected to be present in unaffected areas. However, if any non-background j

radioactivity is found in these areas, it will be assumed that its source was plant-related. Therefore, the GLV for unaffected areas will also be based on the most conservative distribution found in affected areas and will be the same as the l

GLV for non-specific affected areas.

3.4.2 Soil, Sediment and Bulk Material Activity Guideline Values GLVs for surface, subsurface and subfloor soil, sediment and bulk materials will be the concentration per unit mass of radionuclides that would lead to a TEDE of 15 mrem during a period of one year. These GLVs will be taken c

from Table B-2 (Residential Scenario) of NUREG-1500 (Reference 3). If a radionuclide is present on site and not listed in Table B-2, GLVs may be determined by use of computer codes such as RESRAD (Reference 8). Average concentrations above background for radionuclides in these materials may be compared against these GLVs, as described in Section 5.0. Where more than one radionuclide is present, the sum of the ratios of each radionuclide concentration to its respective GLV should not exceed unity.

3.5 Organization and Responsibilities Final status survey plan development and implementation will be performed by Yankee personnel and supplemented by contractor personnel. All survey personnel will be qualified and trained in accordance with approved YNPS procedures. Key position descriptions delineating responsibilities and interfaces have been developed and described by administrative procedure AP-8800. (Procedure Reference 1) 3.6 Quality Assurance i

The Decommissioning Quality Assurance Plan is based on the requirements of 10CFR50, Appendix B as they apply to decommissioning activities, and is discussed in Section $14 of the YNPS FSAR. This plan will be applied to the activities of the final O

status survey. For all radiological analyses performed by the Yankee Atomic d

A-11

YNPS FINAL STATUS SURVEY PLAN REVISION 0 O

Environmental Laboratory, the Laboratory's Quality Assurance Plan will be applicable.

Examples of the QA program application are described in the subsections that follow.

3.6.1 Instrumentation Selection, Calibration and Operation Selection and use of instrumentation will ensure that sensitivities are sufficient to detect radionuclides at the minimum detection requirements specified in Section 4.6, and will assure the validity of the survey data. Instmment calibration will be performed either under approved written procedures using calibration sources traceable to the National Institute of Standards and Technology (NIST), or by qualified vendors providing results traceable to NIST. Issuance, control, and operation of all survey instrumentation will be established by procedures.

3.6.2 Survey Documentation Each final status survey measurement will be identified by date, instrument, location, type of measurement, and mode of instmment operation.

3.6.3 Quality Control - Replicate Surveys O

For structures and systems, QC replicate measurements will consist of resurveys of 5% of randomly selected survey areas.

QC surface contamination and exposure rate measurements will be compared to the original measurements and if the same conclusions are reached without any j

additional exceptions the QC survey will be accepted as satisfactory. (Procedure References 9 and 11)

For open land areas, replicate QC surface contamination and exposure rate measurements will be treated the same as described above for structures and systems. For soil, water and sediment samples, QC sampling will consist of field blanks, blind duplicate samples and third party analyses from each open land survey area. Field blanks and/or blind duplicate samples are not applicable to certain sample types, such as asphalt or mbble.

If QC replicate measurements or sample analyses fall outside of their acceptance criteria, a documented investigation is required which may result in a resurvey or resampling.

O.

A-12

- =.

i YNPS FINAL STATUS SURVEY PLAN REVISION 0 V)

/"

3.6.4 Written Procedures All final status survey tasks which are essential to survey data quality will be implemented and controlled by procedure. As appropriate, dry runs and mockups will be performed to test principle procedures and methods prior to implementation in the field.

3.6.5 Chain of Custody YNPS procedures will establish responsibility for custody of samples between the point of measurement or collection until final results are obtained. When custody is transferred (e.g., when samples are sent off-site to another lab for analysis), a chain of custody form will accompany the sample for tracking purposes.

3.6.6 Records Management Generation, handling and storage of the original final status survey design and data packages will be controlled by approved procedures.

4 3.6.7 Independent Review of Survey Results Randomly selected survey packages from survey areas will be independently reviewed by Yankee Atomic's Quality Services Group to ensure that the survey results are documented in accordance with approved procedures.

3.6.8 Access Control of Surveyed Areas and Systems t

Administrative and physical controls for access to surveyed areas will be established to preclude the possibility of contamination subsequent to completion of the final status survey. See Section 4.1.7 for additional information.

3.6.9 Control of Vendor-Supplied Services Quality-related services, such as instrument calibration and laboratory sample analysis, will be procured only from qualified vendors whose internal quality assurance activities are subject to audit in accordance with the Decommissioning Quality Assurance Plan.

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YNPS FINAL STATUS SURVEY PLAN REVISION 0 0

3.7 Training Qualification and training requirements have been established within the framework of the YNPS FSAR. Before assignment to the final status survey, technicians will receive training in the following:

Overview and objectives of the final status survey plan, Procedures governing the conduct of the final status survey, Operation of the appropriate field and laboratory instrumentation used in the final status survey, and Collection of final status survey measurements and samples.

The extent of training and qualifications will be commensurate with the education, experience and proficiency of the individual and the scope, complexity and nature of the activity. Records of training, including testing to demonstrate qualification, will be maintained in accordance with established procedures.

3.8 Laboratory Services The YNPS radioanalytical programs are administered by the Radiation Protection and Chemistry Department in accordance with written procedures and an approved quality assurance program. On-site laboratory radioanalytical capabilities will be utilized to support the final status survey. Current on-site capabilities include gamma spectroscopy (HPGe), liquid scintillation, and gas proportional counting. Off-site laboratory sample analysis will be provided by the Yankee Atomic Envircnmental Laboratory or procured from other qualified vendors.

3.9 Schedule The final status survey will be carried out in several phases. The first phase will occur while spent fuel is stored in the Spent Fuel Pool. It will encompass the entire site, excluding the Spent Fuel Pool, its supporting systems, structures and components, and a suitable buffer area. Throughout all pha'ses, all completed areas will be controlled to prevent recontamination from decommissioning activities or the subsequent operation or decommissioning of the Spent Fuel Pool. The second phase of the final status survey will proceed after fuel is removed from the Spent Fuel Pool. The second phase of the final status survey will encompass the Spent Fuel Pool, its supporting systems, stmetures and components, and the previously-defined buffer area. If the fuel is transferred to an on-site dry cask storage facility, the final status survey following A-14

i YNP.3 FINAL STATUS SURVEY PLAN REVISION 0

'O decommissioning of the storage facility will be covered via a supplement to the Final Status Survey Report. (See Figure 1-1 of the License Termination Plan)

' Prior to completion of each phase, individual buildings or survey areas will be made available for independent verification of their radiological status. Upon completion of each major phase, records will be compiled, and an interim report will be prepared.

Upon official acceptance of the radiological status by the NRC, the surveyed stnictures and open land areas will be released and available for conventional demolition and disposal as non-radiological material.

3.10 Final Status Survey Reporting Upon completion of each survey phase, an interim report may be submitted to NRC.

At the completion of the entire decommissioning project, a final report will be i

prepared. All reports will be written to meet the intent of Regulatory Guide 1.86 for final status survey reporting, and will follow the guidance of Draft NUREG/CR-5849 regarding content and format. The interim and final reports are described in Section 6.0, " Final Status Survey Reports."

O f

I a

1 A-15

REVISION 0 YNPS FINAL STATUS SURVEY PLAN 0

TABLE 3.1 Surface Contamination Limits 6

REMOVABLE' NUCLIDE AVERAGE' MAXIMUM 2

2 2

U-nat, U-235,U-238 5,000 dpm a /100 cm 15,000 dpm a /100 cm 1,000 dpm a /100 cm associated decay products 2

2 2

Transuranics, Ra-226, Ra-228, 100 dpm /100 cm 300 dpm /100 cm 20 dpm /100 cm Th-230, D-228, Pa-231, Ac-227,1-125, I-129 2

2 2

Th-nat, R-232, Sr-90, 1,000 dpm /100 cm 3,000 dpm /100 cm 200 dpm /100 cm Ra-223, Ra-224, U-232, 1-126,1-?31,1-133 2

2 2

11-3, Fe-55, Ni-63 200,000 dpm /100 cm 600,000 dpm /100 cm 40,000 dpm /100 cm 2

2 2

Beta-gamma emitters 5,000 dpm /100 cm 15,000 dpm /100 cm 1,000 dpm /100 cm (nuclides with decay modes other than alpha emission or spontaneous fission) except Sr-90 and others noted above.

Notes Measurements of average contamination levels should not be averaged over more than I square a.

meter. For objects of less surface area, the average should be derived for each object.

2 b.

The maximum contamination level applies to an area of not more than 100 cm.

When removable contamination on objects of less surface area is determined, the pertinent levels c.

should be reduced proportionally ano the entire surface should be wiped.

O A-16

4

[

YNPS FINAL STATUS SURVEY PLAN REVISION 0 1

4.0 SURVEY PLAN AND PROCEDURES j

4.1 General Survey Plan The design approach of the YNPS final status survey is affected by the fm' al configuration of the facility, i.e., with most systems removed and structures largely

- intact. The majority of the survey effort will occur in areas where radioactive materials were used or handled, such as the Vapor Container, Spent Fuel Building, Waste Disposal Building, Primary Auxiliary Building, and Yard Area.

Three categories or types of survey areas have been established, including:

e.

Structures, to include building interiors and exteriors, and the exterior surfaces of plant systems, Plant systems, to include the interior surfaces of process piping and ventilation e

duct work, and a

Open land areas, to include soil, subfloor soil, sediment and groundwater.

e

'('

For a discussion of survey areas, see Section 4.7.

Due to the large scope of the final status survey and the requirement that some survey activities be conducted in parallel with decommissioning work, a systematic approach is necessary. The final status survey planning and implementation process for each survey area will therefore involve, in addition to the actual survey activities discussed in the plan, each of the following activities, as applicable:

e History file preparation Initial classification e

Walkdown e

Turnover for final status survey (affected survey areas only) e Investigation, if required Reclassificatico, if required e

e Access control measures j

These steps are described in the following sections.

4.1.1 History File Preparation The history file will be a compilation of relevant operational and decommissioning data in a standardized format. The purpose of this process is A-17 3

~

w

REVISION 0 YNPS FIN AL STATUS SURVEY PLAN Oi to provide a substantive basis for the survey area classification, and hence the level of intensity of the f' mal status survey. The history file will contain:

The survey area description Operating history which could affect radiological status Summarized scoping and site characterization data A summary of radiological surveys performed during decommissioning Other relevant information 1

4.1.2 Initial Area Classification 1

All areas of the YNPS site (including structures, plant systems and open land areas) will not have the same potential for the presence of residual contamination and therefore will not require the same level of survey coverage to achieve an acceptable level of confidence that an area satisfies the established release limits. Therefore, to provide an overall planning basis for the final status survey, the site has been initially classified into affected and unaffected j

areas, using the criteria given below and in Figure 4.1.

Initial classification of site areas was based on data provided in the YNPS Decommissioning Plan, subsequent site characterization data, history of g ',

radioactive materials involvement or potential for contamination of the survey area, reconunendations by YNPS personnel knowledgeable of the facility conditions, and any applicable survey data. Data from operational surveys done in support of decommissioning, routine surveillances, and any other applicable survey data may be used to change the initial classification of any individual survey area, if necessary, prior to the commencement of the final status survey in that survey area. The site breakdown and classifications are controlled by administrative procedure AP-8801 (Procedure Reference 2).

Areas are classified as follows:

Affected Areas: Affected Areas are those areas that have potential radioactive contamination based on plant operating history. They are also those areas, and where appropriate, the areas immediately surrounding or adjacent to them, that

1. ave known radioactive contamination (based on past or preliminary radiological surveys) due to one or more of the following activities in the areas:

1) the use or storage of radioactive materials; 2) spills or other unusual occurrences that could have resulted in the spread of contamination; or 3) the burial of radioactive material. Any area that has been previously remediated will also be classified as an Affected Area.

O A-18

YNPS ITNAL STATUS SURVEY PLAN REVISION O

.A;U Areas which are expected to have radioactive contamination existing at average levels in excess of 25% of the average total surface contamination, or soil guideline values, or greater than the Minimum Detectable Concentration (MDC) for the equipment used to perform the survey, whichever is larger, will be classified as Affected Areas.

To allow a more concentrated survey effort in those areas most likely to be i

contaminated, Affected Areas within a room or building will be further subdivided into Suspect Affected and Non-Suspect Affected Survey Units, as -

follows:

e Susnect Affected Survey Units: The designation assigned to floor and wall surfaces below two meters within an Affected Area. Also, the designation assigned to all additional surfaces if there is reason to i

believe. that there was licensed material present, unless it meets the criteria for a Non-Suspect Affected Survey Unit.

o Non-Susnect Affected Survey Units: The designation assigned to ceiling and wall surfaces above two meters within an Affected Area, where it is not expected that radioactive materials exist at average levels

,f in excess of 25% of the average total surface contamination guideline values, or the MDC for the equipment used to perform the survey, whichever is larger. This designation is based on engineering 1

judgement, in consideration of operational history, characterization data, operational surveys performed in support of decomtnissioning, and routine surveillances.

Unnifected Areas: These areas have a low potential for radioactive contamination, based on a knowledge of site history and previous survey information. Average measurements for average total surface contamination, and for average concentrations of radionuclides in soil or asphalt are expected to be less than 25% of the applicable guideline value or less than the MDC for the equipment used to perform the survey, whichever is larger. Previous remediation precludes a survey area from being initially classified as unaffected.

Non-Imnacted Areas: Those areas of YNPS property that are outside of unaffected areas will be classified as non-impacted and will not be surveyed.

These areas have an insignificant potential for residual contamination.

(Reference 9)

I V

A-19 m

=

e.'

YNPS FINAL STATUS SURVEY PLAN REVISION 0 4.1.3 Walkdown The principal objective of the walkdown is to assess the physical scope of the survey area. For systems, it will include a review of system flow diagrams and piping drawings, and a physical walkdown of the system. Stnictures and open land areas will also be walked down. The walkdown is best completed when the final configuration is known, usually near or after completion of deconunissioning activities.

Specific requirements will be identified for accessing the survey area and obtaining support functions necessary to conduct the final status surveys, such as scaffolding, interference removal, and electrical tagout. Safety concerns such as access to confined spaces, high walls, and ceilings will be identified.

4.1.4 Turnover foi Final Status Survey Prior to acceptance of a survey area for the final status survey, a number of conditions must be satisfied. These include:

Decommissioning activities having the potential to contaminate the survey area must be completed.

All tools and equipment not required to perform the survey must be e

removed, and housekeeping and area cleanup must be completed.

e Decontamination activities in the survey area must be completed.

Final operational radiological surveys must be completed in affected e

areas to ensure that no additional remediation is necessary and will consist of:

1.

Scan surveys to ensure that surface contamination is within the FSS total surface contamination limits.

2.

Smear surveys to ensure that the removable surface contamination is within the FSS removable surface contamination

limits, e

Access control or other measures to prevent recontamination must be implemented.

G

~

A-20

i i

l YNPS FINAL STA'IUS SURVEY PLAN REVISION 0 O

4.1.5 Investigation When an administrative action level is reached, an investigation is performed to confirm the initial final status survey measurement, and if verified to exceed action levels, to define the bounds of the elevated activity. Depending on the results of the investigation, the survey unit may require no action, remediation, and/or reclassification and resurvey. The administrative action levels for such action may be found in Table 5.1. Specific instructions for the processing of data following an investigation survey are included in Section 5.2.4.

4.1.6 Reclassification Based on information obtained during final status survey activities, it may be necessary to reclassify certain survey units to ensure that the measurement frequency is appropriate for the contamination potentizl. Following reclassification, the newly classified survey unit will be re-surveyed under the appropriate survey requirements for that area. The existing final status survey measurements may be included with the additional survey data. Administrative action levels for reclassification are found in Table 5.1.

Q 4.1.7 Access Control Measures

'V Since all decommissioning activities will not be completed prior to the start of final status survey, measures will be implemented to protect survey areas from contamination during and subsequent to the final status survey. In all cases, decommissioning activities creating a potential for the spread of contamination must be completed within each survey area prior to the final status survey.

Additionally, decommissioning activities which create a potential for the spread of contamination to adjacent survey areas will be evaluated and controlled.

Upon commencement of the final status survey for survey areas where there is a potential for contamination, implementation of one or more of the following control measures will be required:

o Personnel training, e

Installation of barriers to control access to surveyed areas, Installation of postings requiring personnel to perform contamination e

monitoring prior to surveyed area access, Locking of entrances to surveyed areas of the facility, e

A-21

f YNPS FINAL STATUS SURVEY PLAN REVISION 0 Installation of tamper-evident labels.

Controls will be implemented through procedure AP-8810. (Procedure Reference 10) 4.2 Initial Area Classification of the YNPS Site The following areas on site have been initially classified as Affected Areas:

The interior surfaces of any remaining structures within the currently defined Radiological Control Area (RCA) (see Figure 2.2), including the primary side instrumentation lab, counting rooms, primary side chemistry lab, contaminated sink and shower areas, north and south decontamination rooms, RCA machine shop, PC change room and the controlled side of the RP control point. The exterior surfaces of any systems in this area will also be classified as affected.

The interior surfaces of the Turbine Building, Front Office Building, original Service Building and Service Building Addition.

The PAB roof, all exterior surfaces of the Spent Fuel Pool Building, and the Vapor Container exterior surface below the equipment hatch.

A small activated area on the top center of the exterior surface of the Vapor Container.

All open land areas within the RCA.

1 A 10-meter wide buffer zone around the chain link fence surrounding the outdoor portions of the RCA.

The small section of landscaped area bordered on the east by the west wall of the turbine building, on the south by the RCA fenceline, on the west by pavement, and on the north by the driveway leading to the overhead door on the west wall of the turbine building.

A 10-meter wide area from the east storm drain outfall to the Sherman Pond shoreline. This area will also include the sediment in a 10 meter radius semi-circle directly below the east storm drain outfall.

The stream-bed sediment from the point of the West Storm Drain Outfall to the Deerfield River.

Both the operable and inoperable plant septic systems.

A-22

YNPS FINAL STATUS SURVEY PLAN REVISION 0 V

The following areas have been initially classified as Unaffected Areas:

e Any areas inside the Owner Controlled Area Fence not classified as Affected, The exterior surfaces of structures not previously classified as Affected.

e e

A 20-meter wide buffer zone outside the Owner Controlled Area fence, and a I

20 meter wide buffer zone outside of Affected Areas which are beyond the fence, where possible, e

The interior surfaces of the Visitor Center (Furlon House), and Training

Center, o

Parking lots i

.These large areas will be subdivided into survey areas. Ifjustified by survey data or other information, the initial classification of an individual survey area may be changed in accordance with AP-8801 (Procedure Reference 2) prior to the commencement of the final status survey in that survey area.

Prior to the commencement of the final status survey, the interior surfaces of any O5 remaining systems will be subdivided into survey areas as necessary, and will be classified at that time.

All areas of the YNPS site that have not been classified as Affected or Unaffected have been classified as Non-Impacted. These areas will not be surveyed.

4.3 Measurement Frequency Measurement frequencies, or the physical spacing of samples and measurements, are selected to allow for a concentrated survey effort in those areas most likely to be contaminated or activated, taking into account the type and size of the survey unit. The measurement frequency for each area type is described below.

4.3.1 Structures Suenact Affected Survey Units: The final status survey of Suspect Affected Survey Units will include:

e A scan survey of 100% of the accessible surface area.

NOTE-The scan survey may be performed at the time of survey O

unit turnover.

A-23

_ _ _ =

YNPS FINAL STATUS SURVEY PLAN REVISION 0 Total surface contamination (fixed point) measurements, as follows:

For areas s 30 square meters, a minimum of 30 measurement locations; For areas > 30 square meters, a minimum of one measurement location for each one square meter.

Removable contamination (smears) will be obtained at locations w'here the total surface contamination measurement MDC or result is greater than the removable contamination limit.

Exposure rate measurements will be taken only on activated concrete, stmetures or components at a frequency of at least one measurement per 4 square meters.

In general, these areas will be gridded into one square meter grids, and the measurements will be taken at the center of grid blocks.

1 Non-Suspect Affected Survey Units: The final status survey of Non-Suspect Affected Survey Units will include:

O1 A scan survey at least 10% of the accessible surface area of the survey unit.

1 NOTE:

The scan survey may be performed at the time of survey unit turnover.

Total surface contamination (fixed point) measurements, as follows:

For areas s 600 square meters, a minimum of 30 measurement locations; For areas > 600 square meters, a minimum of one measurement location for each 20 square meters.

Removable contamination (smears) will be obtained at locations where the total surface contambation measurement MDC or result is greater than the removable contamination limit.

In general, these areas will not be gridded. Measurement locations will be distributed throughout the survey unit. In addition, some measurement locations will be selected from locations most likely to accumulate contamination.

A-24

YNPS FINAL STATUS SURVEY PLAN REVISION 0 O

Unntracted Survey Units: The final status survey of Unaffected Survey Units will include:

e-A scan survey of an area around each survey location so that the total area scanned is at least 10% of the accessible surface area of the survey unit.

e Total surface contamination (fixed point) measurements on floors and walls below two meters, as follows:

For areas s 1500 square meters, a minimum of 30 measurement locations; For areas > 1500 square meters, a minimum of one measurement location for each 50 square meters.

Removable contamination (smears) will be obtained at locations where the total surface contamination measurement MDC or result is grester than the removable contamination limit.

In general, these areas will not be gridded. Measurement locations will be distributed throughout the survey unit and will be documented.

1 The final status survey of all building exteriors not previously classified as affected will be done in accordance with the protocol established for unaffected building surfaces.

4.3.2 Plant Systems Affected Plant Systems: The final status survey of the interior surfaces of affected plant systems will include:

e A scan survey of the accessible surfaces at each survey location.

e A minimum of 30 survey locations, as allowed by system size, to include a minimum of 30 measurements of total surface contamination

.I within each survey area.

l e

. Removable contamination (smears) will be obtained at locations where the total surface contamination measurement MDC or result is greater than the removable contamination limit.

O A-25

YNPS FINAL STATUS SURVEY l'LAN REVISION u Measurement locations will be selected from locations wheie contamination is most likely to accumulate. If appropriate, sampling and analysis of scale and tediment will be performed.

. Unaffected Plant SySicms: The final status survey of the interior surfaces of unaffected plant systems will include:

A scan arvey in the immediate vicinity of the total surface e

contamination location.

A minimum of 10 survey locations, as allowed by system size, to include a minimum of 30 measurements of total surface contamination within each survey area.

e Removable contamination (smears) will be obtained at locations where the total surface contamination measurement MDC or result is greater than the removable contamination limit.

Measurement locations will be selected from locations where contamination is most likely to accumulate. If appropriate, sampling and analysis of scale and sediment will be performed.

J 4.3.3 Open land er tas Afrected Open land Areas Without Paved Surfaces: Th.: final status survey of affected open land areas without paved surfaces will include:

e A scan survey of 100% of the acassible surface area.

Surface soil samples (15 centimeters deep) as follows:

e For areas s 750 square meters, a minimum of 30 survey points; For areas > 750 square meters, a minimum of 4 sorvey points for each 100 square meters.

Exposure rate measurements taken at one meter from the surface at the e

location of each soil sample.

These areas will normally be gridded into 10 x 10 meter grids. Samples and measurements will be taken systematically within those grids.

O Ada

YNPS FINAL STATUS SURVEY PLAN REVISION 0 Affected Open Land Areas With Paved Surfaces: The final status survey of affected open land areas with paved surfaces will include:

1 A scan survey of 100% of the accessible surface area.

Total surface contamination (fixed point) measurements, as follows:

For areas s 30 square meters, a minimum of 30 survey points; For areas > 30 square meters, a minimum of one survey point for each one square meter.

In general, these areas will be gridded into one square meter grids, and the measurements will be taken at the ceater of grid blocks.

Pavement samples and surface soil samples (15 centimeters deep, e

Mneath the pavement) as follows:

For areas s 375 square meters, a minimum of 30 survey points;

! O For areas > 375 square meters, a minimum of 8 survey points for each V

100 square meters.

Exposure rate measurements taken at one meter from the surface at the e

location of each soil / pavement sample.

Samples and measurements will be taken systematically within those grids.

Unnrfected Open land Areas: Tbn final status survey of both paved and unpaved surfaces in Unaffected Amu willinclude:

A scan survey of an area around eacle survey location so that the total e

area scanned is at least 10% of the accessible surface area of the survey unit.

e Surface soil samples (15 centimeters deep) and exposure rate measurements, and fc aved surfaces, pavement samples and total surface contamination (fixed point) measurements, as follows:

For areas s 37,500 square meters, a minimum of 30 survey points;

. L]

f(

For areas > 37,500 square meters, a minimum of one survey point for cach 1,250 square meters.

A-27 O

YNPS FINAL STATUS SURVEY PLAN REVISION 0 In general these areas will not be gridded, although survey points will be documented.

Affected Subiloor Areas With Floors Removed: The final status survey of affected subfloor areas with floors removed will include:

o A scan survey of 100% of the accessible surface area.

Soil samples (15 centimeters deep) as follows:

For areas s 750 square meters, a minimum of 30 survey points; For areas > 750 square meters, a minimum of 4 survey points for each 100 square meters.

Exposure rate measurements taken at one meter from the surface at the location of each soil sample.

These areas will normally be gridded. Samples and measurements will be taken systematically within those grids.

O Mfected Subfloor Areas With Floors Intact: The final status survey of affected subfloor areas with floors intact will include:

Soil samples (15 centimeters deep) as follows:

For arc:. s 375 square meters, a minimum of 30 survey points; For areas > 375 square meters, a minimum of 8 survey points for each 100 square meters.

In general, these areas will be gridded, and the samples will be taken systematically within those grids.

i Groundwater and Surface Water: A groundwater sample will be collected as i

part of the final status survey in any survey unit where an operable groundwater observation well is located. A representative surface water grab sample will be collected as part of the final status survey in any survey unit where a permanent surface water source exists.

A-28

l i

YNPS FINAL STATUS SURVEY PLAN REVISION 0..

..AQ

- 4.4 Survey' Methods i

4.4.1 Scan Surveys Scan surveys will be performed to screen surface areas to identify the presence of any locations of elevated contamination in order to search for areas above the release limits and to detect localized areas above the maximum release limit.

l The scanning methods (instmments and survey technique) for surface contamination measurements will normally be designed to detect less than 75 %

of the average total surface contamination release limits. If an area of elevated contamination is identified during the scan of a survey Unit, the location will be marked and included as a part of the total surface contamination measurements.

4.4.2 Surface Contamination Measurements Surface contamination measurements will be taken at discrete measurement

. locations and at frequencies which are based on the classification of the survey unit. Typically only a total surface contamination measurement will be taken at each location.

i O

Before final status surveys are initiated, turnover surveys are performed by operational Radiation Protection Department personnel to verify that the measurement results are within the final status survey limits for removable and total surface contamination. Final status survey total surface contamination i

measurements will detect both removable and fixed contamination. If the total surface contamination results are less than the removable contamination limit, additional removable contamination measurements (smears) will not be made.

If the total surface contamination results are greater than the removable surface contamination limits or the MDC of the measurement is greater than the removable contamination limit, additional removable surface contamination measurements will be taken.

4.4.3 Exposure Rate Measurements Exposure rate measurements will be obtained at discrete locations which are based on the classification of the survey unit. Exposure rate measurements will only be taken on activated components and structures and in open land areas.

Measurements e be taken at one meter from surfaces.

j Exposure rates from infinite plane sources can be derived from data in NUREG-1500. The derived exposure rate from Co-60 in the NUREG was j

Q benchmarked with a commercially available computer code and found to be i

.k) :

comparable. Various realistic geometries such as equipment rooms and large i

1 A-29 1

YNPS FINAL STATUS SURVEY PLAN REVISION 0 areas like the Turbine Deck at YNPS were evaluated to determine exposure rates from surface contamination. Assuming uniform contamination on the room surfaces and using the expected radionuclide distribution as input, the calculated exposure rate was much less than 5 R/h. Non uniform contamination was also evaluated and found to be much less than 5 R/h.

Therefore, exposure rates are not necessary where the only source of residual contamination is surface contamination.

4.4.4 Soil and Groundwater Sampling Surface soil grab samples will generally be collected to a depth of 15 centimeters at discrete locations, and at frequencies which are based on the classification of the survey unit. Sampling at greater depths will be done in areas where site characterization or other information indicates potential contamination at depths greater than 15 centimeters. Groundwater grab samples will be collected from existing observation wells and surface water grab samples will be collected from existing permanent sources of surface water.

4.4.5 Special Measurements and Samples I

Sampling of Miscellaneous Loose Material 1

Samples of loose paint, tank sediment, sewage sludge, roofing material, j

concrete and other bulk materials will be collected for laboratory analysis as part of biased sampling and measurements. Such samples may be collected in drain receptacles, sumps, and other catchments in affected areas. Selected storm drain catchments will be sampled in accessible locations on the site.

These samples will be quantitatively analyzed by gamma spectroscopy for principal gamma emitting radionuclides and the results compared to the applicable guideline value.

Ein!Ledded Piplig Surveys Measurements will be made to demonstrate that normally inaccessible piping remaining at the time of the final status survey (e.g., embedded piping) is below the guideline values for surface contamination. This will include but not be limited to the use of calibrated detectors extended into piping runs in a controlled manner and the use of data acquisition equipment to document the measurements.

The detectors that may be used for direct measurement of total surface contamination within embedded piping include, but are not limited to:

g

1) cylindrical gas-flow proportional detectors ranging in diameters from 0.5" to A-30

w

\\

YNPS FINAL STNIUS SURVEY PLAN REVISION 0 O

1.5"; 2) 125 cm gas-flow proportional detectors; 3) 15.5 cm Ge$ger-Mueller 2

2 a

detectors; and 4) other detector types and sizes as situations dictate. The detectors may be fitted with a delivery device to provide a consistent and reproducible calibration / measurement geometry and to protect the detector from damage or potential contamination.

For embedded piping, total and removable surface contamination will be measured. The method used to assess removable contamination within

}

embedded piping will consist of one of the following: the measurement of total surface contamination, smear media positioned with a remote handling tool, or smears taken at accessible openings.

[

To support the conclusion that sections of embedded piping meet the criteria of j

release for unrestricted use, totri surface contamination measurements and l

removable contamination measurements will typically be performed prior to,

_i' during, and after decontamination processes to evaluate the effectiveness of the decontamination technique (s). The decontamination processes will be controlled to ensure that all portions receive similar treatment by the decontamination process. Based upon the effectiveness of the decontamination process, as defined by the measurement results obtained, the radiological condition of any inaccessible portion of the embedded piping may be inferred.

Pond Sediment t

Areas of pond sediment that are included in the final status survey will be surveyed at a frequency of 30 samples per survey unit. Samples will consist of j

bottom and/or shoreline grab samples.

l In h Gamma Snectrascany Measurements In certain areas of the site, large variations in background exposure rates may result in investigative surveys. Investigative surveys may be performed with in situ gamma spectroscopy to not only identify the radionoclides and their concentrations, but also to quantify the exposure rate.

The Environmental Measurement Laboratory (EML) at the US Department of Energy has performed many detailed evaluations of portable gamma spectrometry systems. These evaluations have been instrumental to the routine use of gamma spectrometry in radiological environmental monitoring programs to evaluate soil activity concentrations and exposure rates. In situ gamma spectrometry systems have also been used to evaluate radioactive waste p

containers. In addition, Fort St. Vrain was approved to use in situ gamma

,i spectrometry to measure exposure rates during their final status survey.

l t

A-31 1

a

YNPS FLNAL STATUS SURVEY PLAN REVISION 0 Therefore, the technology to convert gamma ray spectra from building surf aces and soil into activity concentrations and exposure rates may be used at YNPS during the course of the final status survey.

4.4.6 Sampling for Hard to Detect Nuclides (HTDN)

Samples will be taken and analyzed from random survey locations in affected areen of the facility. The ratios of IITDN, such as H-3, Fe-55 and Ni-63, to readily detected radionuclides, previously used to establish site specific GLVs, will be revised as appropriate.

Both II-3 and Ni-63 emit low energy beta particles during radioactive decay.

Iron-55 emits low energy X-rnys during radioactive decay. These decay products are difficult to detect with standard field instrumentation. The risk from these radionuclides is much lower than the risk from the easily detected Co-60 as shown in the surface concentration values of Table B-2 of NUREG-1500 which are between 500 to 5000 times the limit for Co-60. Although difficult to detect, the guideline value equation (Equation 3.1) incorporates the presence of these radionuclides at YNPS.

4.5 Background Level Determinations g

Background levels will be established for each type of instrument to be used for total surface contamination measurements, removable contamination measurements, and gamma exposure rate measurements. In addition, backgrounds will be determined for specialized detectors and detector systems. These will include large area detectors for floor monitoring and detectors for surveying piping interiors (e.g., mdople GM detectors). For soil and sediment samples, background levels will be determined for those man-made radionuclides not resulting from plant activities.

The objectives of background determinations for YNPS final status survey measurements will be to:

Establish the reference background values for each type of detector used in the survey.

Assess the variability in background responses for principal detectors under different applications and conditions of use.

Determine the need for correction factors or special measurements to establish the background for final status survey measurements in specific locations.

Account for man-made radioactivity not resulting from plant operations.

A-32

YNPS ITNAL STATUS SURVEY PLAN REVISION 0 l'

t Background determinations will be performed in accordance with approved procedures.

A Values will be re-evaluated as necessary. The methods for background determinations for each type of final status survey measurement are summarized below.

4.5.1 Total Surface Beta-Gamma Measurements Eackground measurements will be performed on surfaces unaffected by licensed activities. The principal criteria for selection of measurement locations for building surfaces will be the location's similarity to YNPS facility construction and freedom from radioactivity attributable to YNPS operations.

4.5.2 Removable Surface Beta-Gamma Measurements Background determinations of beta-gamma smear counters will be made by taking a count of a blank smear or carrier, usually on each day of operation.

4.5.3 Soil and Sediment Radionuclide Measurements Soil and sediment samples will be collected from areas unaffected by licensed operations in order to establish the background levels of man-made O

radionuclides not resulting from plant operations. The background samples will be collected at locations similar to their respective on-site samp.'ing locations.

4.5.4 Gamma Exposure Rate Measurements A pressurized ion chamber will be used to establish the background gamma exposure rate at the site, which will be used as the baseline for demontrating that average residual gamma exposure rate measurements are below 5 pR/hr above background (measured at one meter). The pressurized ion chamber will be used as the reference instrument for establishing the gamma exposure rate background and the pR/hr equivalent response of portable survey meters.

The latter will be used for the bulk of the final status survey gamma exposure rate measurements.

This background value will be used for most survey units. Where actual background conditions are shown to vary significantly, area-specific background values will be used.

Alternatively, in situ gamma spectroscopy may be used to identify those components of the gamma exposure rate that are due to plant contamination, effectively precluding the need for background comparisons.

I v

A-33

J YNPS 11NAL STATUS SURVEY PLAN REVISION 0 4.5.5 Specialized Measurements It has been observed that detector background for direct beta-gamma measurements is affected by conditions in the immediate vicinity of the detector.

Significant variations from background reference values have been observed.

These variations are caused by the natural radioactivity composition of materials and by shielding effects in some cases. As a result, background measurements 2

for special condi. sns will be compiled for use in calculating net dpm/100 cm values to reduce the bias in survey unit population statistics. Special condition backgrounds may be compiled for ceramic and clay tile materials, porcelain fixtures, poured concrete materials, roofing materials, large bore piping, embedded piping and other materials.

4.5.6 Verification of Background Measurement Population Each population of background measurements will be analyzed by using the following equation, taken from Equation 8-22 of Draft NUREG/CR-5849, to ensure that the number of measurements in the data set is adequate to characterize the background mean value to within 20% at the 95%

confidence level:

t S

,8 95 s. af x

(4.1) 0.2 f,

number of background measurements required, n,

=

E mean of background measurements,

=

s standard deviation of initial background measurements, Sx

=

and t statistic for 95% confidence at n-1 degrees of freedom,

!95 %,dr

=

where n is the number of initial background data points.

4.6 Instrumentation nadiation detection and measurement instrumentation for the firal status survey will be selected to provide both reliable operation and adequate sensitivity to demonstrate attainment of the release criteria. Detector selection will be based upon detection sensitivity, operating characteristics and expected performance in the field under expected conditions of use.

A-34

1 l

YNPS FINAL STATUS SURVEY PLAN REVISION 0 t

The prmcipal instruments being evaluated for final status survey measurements are identified in Table 4.1. The detectors used for total surface contamination monitoring -

will typically be operated with data logging survey meters.

4.6.1 Minimum Detectable Concentration Calculation t

Minimum Detectable Concentration (MDC) values for field and laboratory counting instrumentation will be determined using the following equation, taken from Equation 3-10 in Draft NUREG-1507 (Reference 10):

t 3 + 3.29 R*I (1 + '-)

t, (4.2)

MD C =

E (A/100) t,

where, MDC =

the minimum activity concentration on a surface or material volume that can be statistically detected above background with a 95% probability, sample counting time interval, tg

=

l background counting rate, R

=

b background Counting time interval, t

=

b l

counting efficiency, and j

E

=

l I

area of the detector, or the area sampled for smear samples.

A

=

4.6.2 Detector Sensitivity The detection sensitivity of some of the detectors being selected for surface contamination measurements has been estimated. These results are summarized in Table 4.2. The results are shown for the principal instruments that are expected to be used for field and laboratory beta-gamma total surface and removable contamination measurements.

The MDC is dependent on several factors, such as sample count time, l

background count time, background count rate and detector efficiency. The A-35 I

1

~

l

r YNPS FINAL STATUS SURVEY PLAN REVISION 0 MDC for scan survey measurements will be calculated using Equation 4.2, with the scanning efficiency substituted for the efficiency term.

Count times will be selected to ensure that the measurements will be sufficiently sensitive with respect to applicable GLVs. The count times associated with measurements for total surface contamination and gamma spectral analysis (soil and water) will normally be set to ensure an MDC of <25% of the applicable guideline value.

]

It should be noted that it may not be possible to scan or take exposure rate measurements at 25 % of the applicable GLV, based on significant background interference and the relative instrument sensitivities, as shown in Table 4.2.

4.6.3 Calibration and Maintenance Instruments and detectors used in the final status survey will be calibrated and maintained at YNPS according to approved procedures. Detectors for beta or gamma measurements will be calibrated using NIST traceable sources. The pressurized ion chamber will be calibrated by Yankee Atomic Environmental Laboratory (YAEL) or a 1 approved vendor, and will also be NIST traceable.

O 4.7 Survey Areas, Units, Locations and Points 4.7.1 Survey Areas and Survey Units A survey area is usually a contiguous area with similar characteristics and contamination potential. A survey area will not include both affected and unaffected areas. For plant systems, a survey area may be defined to include an entire system. Individual survey areas will typically not include more than one system.

Engineering judgement will be used to divide a survey area into survey units to support reproducible confirmatory measurements. This will be accomplished by selecting existing facility characteristics such as horizontal and vertical structural support beams, concrete pour seams or piping runs to define the boundaries of survey units within survey areas. Distance measurements will be included as necessary to support reproducibility.

In addition, the use of photographs, drawings or maps may be included to support survey unit identification. This method for area designation will also A-36

YNPS FINAL STATUS SURVEY PLAN REVISION 0 serve to help survey personnel to define areas of elevated 1ctivity requiring V

3 addition investigational surveys and/or remediation.

4.7.2 Survey Locations and Survey Points Survey locations will be clearly identified to allow for reproducibility of measurements. Whenever it is appropriate and cost effective, gridding will be used. However, the use of appropriate surface markings may be substituted for a physical grid layout.

Survey points will be identified by a unique reference location ID code or number. The numbering convention will allow the survey data to be easily referenced to survey points identified on maps or drawings.

Y l

(%

\\

)

'~/

A-37

YNPS FINAL STATUS SURVEY PLAN REVISION 0 O

TABLE 4.1 Final Status Survey Instrument Summary Typical Measurement Type Detector Manufacturer Detector Type Area-Density

& Model #

Units Surface Gas 100 cm2 Eberline epm Alpha / Beta-Gamma Proportional 1 mg/cm HP-100 CGS 2

2 Ludlum cpm Surface Gas Flow 425 cm 2

Beta-Ganuna Proportional 0.8 mg/cm 43-37 2

LND/TGM cpm Surface Geiger-Mueller 15.5 cm 2

Beta-Gamma 2 mg/cm Gamma Scan NaI(TI) 2" x 2" Eberline SPA-3 cpm Removable Surface Gas Flow 24.2 cm2 Protean cpm 2

Alpha / Beta-Gamma Proportional 80 pg/cm WPC 9550 Gamma Exposure Pressurized Ion 8 L Sphere Reuter-Stokes R/hr Rate Chamber (PIC)

RSS-112 Gamma Exposure Compensated 1" x 6" Eberline pR/hr Rate Geiger-Mueller HP-270 Gamma Exposure Compensated 2" x 13" Eberline pR/hr Rate Geiger-Mueller HP-300 Ganuna High-Purity 2" x 2" EG&G Geometry Spectroscopy Germanium Ortec/ Canberra Dependent Background and High-Purity 2" x 2" EG&G Geometry Special Germanium Ortec/ Canberra Dependent Measurements A-38

i YNPS FINAL STATUS SURVEY PLAN REVISION 0 '

[

I l&

.U e

a TABLE 4.2 i-Nomimi Measurement Detection Semitivities Efficiency Detection Type of 1

ector Bac 5 0und

- Measurement (c/d)-

Sensitivity _

Betu-Gamma Pancake GM 150 cpm 0.08 3650 f

Surface Scan-(HP-210/HP-260) dpm/100 cm j

2 Beta-Gamma 100 cm Gas Proportional 550 cpm 0.10 850 2

2 Surface Scan.

(HP-100 CGS) dpm/100 cm 425 cm Gas Flow 2000 cpm 0.07 500 2

Beta-Gamma 1

2 Surface Scan (Ludlum 43-37)-

dpm/100 cm Beta-Gamma Pancake GM 50 cpm 0.10 1800 2

Direct (HP-210/HP-260) dpm/100 cm Beta-Gamma 100 cm Gas Proportional 300 cpm 0.14 450 2

2 Direct (HP-100 CGS) dpm'm)cm Alpha Direct 100 cm Gas Proportional 1 cpm 0.25 20 2

2 (HP-100 CGS) dpm/100 cm Gamma Scan NaI(TI) 12,000 cpm 0.012 32,000 2

(SPA-3) dpm/100 cm Gamma Pressurized Ion Chamber N/A N/A 1 - 2 pR/h Exposure Rate (Reuter-Stokes)

Gamma Compensated GM 15 pR/h 1.46 5 R/h j

Exposure Rate (HP-270)

Gamma Compensated GM 12 pR/h 1.22 3 R/h Exposure Rate (HP-300)

]

Beta-Gamma Gas Flow Proportional I cpm 0.22 70 1

2 Removable (Protean) dpm/100 cm 1

Alpha Removable Gas Flow Proportional 0.1 cpm 0.26 40 1

2 (Protean) dpm/100 cm A-39 i

- ^ - - - - -

-.m

_,l

YNPS FINAL STATUS SURVEY PLAN REVISION 0 O

Define tfm boundaries of the survey area.

Perforrn evaluation including review of operafon and decommissioning history.

Potential or average measure Classify survey area NO activity >25% of GLV, as Unaffected.

or > MDC, whicheve is larger.

YES Classify survey area as Affected.

1P GLV - Guideline Value MDC Minimum Detectable Concentration Op "

nd Classify as Affected Open Land YES Area?

NO For rfaces abov 2 meters Average YES of the measurements expected to be >25% of the GLV, or >MDC whchever is larger?

NO i

Classify surfaces above 2 Classify all surfaces as meters as Non-Suspect Affected.

l Suspect Affected, Classify surfaces below 2 meters as Suspect Affected.

l FIGURE 4.1 l

Survey Atea Classification Process O

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YNPS FINAL STATUS SURVEY PLAN REVISION 0 p

\\

5.0 DATA PROCESSING AND INTERPRETATION 5.1 Conversion of Measurements to Reporting Units With respect to the Final Status Survey, removable and total surface contamination will be reported in units of dpm/100 cm, exposure rate will be reported in units of R/hr, 2

6 and volumetric material will be reported in units of pCi/g (solids) or pCi/l (liquids).

Prior to comparison with any action level or release criteria, all radioactivity or exposure rate measurements will have the appropriate background subtracted.

Corrections for detector size (total surface activity measurements) and instrument response (exposure rate measurements) may also be made, as discussed below.

5.1.1 Total Surface Contamination 2

Total surface contamination measurements will normally be taken with 100 cm detectors (if the detector is smaller, the data will be area corrected to reflect a 2

100 square w,atimeter area) and compared to the 100 cm average and maximum surface contamination limits.

If large area detectors are used, the observed contamination will be limited to Q

that contaminadon which would be acceptable when confined to an area of

,b 100 cm. S nce it cannot be discerned that the observed contamination is 2

uniformly distributed, it will be assumed that it could be attributed to an area of 100 cm or less. In the event that contamination in excess of what would be 2

acceptable for an area of 100 cm is observed when using a large area detector, 2

an investigation survey will be performed to ascertain compliance with the average guideline values.

Total surface contamination measurement results will be reviewed to ensure that 1

the applied background values are appropriate. If any background values are determined to be inappropriate they will be adjusted as necessary. All adjustments, and their justifications, will be documented to ensure traceability.

5.1.2 Removable Surface Contamination Measurements of removable surface contamination are converted from a gross count rate to units of net dpm/100 cm by 1,ubtracting the background count rate 2

of the smear counter and dividing the net count rate by the detector efficiency and correcting for the smear area (when necessary). Removable contamination results are then compared to the applicable guideline values.

O V

A-41

YNPS FINAL STATUS SURVEY PLAN REVISION 0 5.1.3 Exposure Rate Exposme rates will normally be measured with a compensated GM tube with a data logging survey meter. Results from these meters will be compared to the response of a Reuter Stokes pressurized ion chamber (PIC). A correction coefficient will be calculated and applied to each exposure rate measurement to correct the reading to the PIC value. (The PIC is an industry recognized, standard quality device which measures exposure rate at environmental levels with negligible energy dependence.) Measurements may be taken directly with the PIC. The appropriate background exposure rate will be subtracted from all measurements.

Alternatively, exposure rates may be determined through in situ gamma spectroscopy. This process identifies those components of the gamma exposure that are due to plant contamination, eliminating the need to subtract background.

5.1.4 Soil and other Bulk Samples Soil and other bulk samples are analyzed on a gamma spectrometry system.

This system identifies both plant-related and naturally occurring radionuclides along with their concentrations and MDCs. Background levels for man-made radionuclides are subtracted, as appropriate.

For each sample, those radionuclides that are identified as plant-produced are reported and the sum of the ratios (SOR) of their activity to their respective r;uideline values (GLVs) are compared to unity.

5.2 Comparison with Administrative Action Levels Two categories of administrative action levels have been established. The first is to initiate reclassification (see Section 4.1.6) and the re-surveying of an area when survey results indicate that inore extensive surveying is needed. The second is to either directly initiate remediation, or where more data are needed to ascertain whether or not the survey unit meets the release criteria, to initiate an investigation. Each of these is specific to the type of survey (removable contamination, total surface contamination, soil and bulk material, or exposure rate) and to the classification of the area being surveyed (affected, suspect affected, non-suspect affected or unaffected). These action levels are shown in Table 5.1. Specific requirements for the handling of data revolving around investigation surveys are described in Section 5.2.4.

O A-42

YNPS FINAL STA'I11S SURVEY PLAN REVISION 0

/~~\\

Q ).

Most action levels are based on a direct comparison of a single measurement or the i

. mean of a group of measuremen's with the applicable guideline value. Two other comparisons are also described in this'section. The first involves the calculation of a -

weighted mean for an elevated activity area, and the comparison of that weighted mean with the guideline value. The second comparison is done to ensure that the mean -

activity level in a sun'ey unit meets the release criteria at the 95% confidence level.

5.2.1 Mean and Standard Deviation In order to compare survey measurement results to admin.istrative action levels, a mean and standard deviation must be calculated for each type of measurement

.for each survey unit.

I Mean The mean is calculated as follows:

x = 1 f x, (5.1)

D

g. 3

where,

)

= sample mean, x

xi = individual measurement values, n = total number of measurements in the survey unit.

4 Standard Deviation f

The standard deviation is calculated as follows:

1

• 8~*

(5.2) 8 =

n-1 1

i

where, i

OQ s, = sample standard deviation,~

A-43

YNPS FtNAL STATUS SURVEY PLAN REVISION 0 xi = individual measurement values, i = sample mean, n = number of measurements in the survey unit.

5.2.2 Elevated Areas of Contamination Levels of residual contamination that exceed a guideline value will be evaluated against an elevated area criterion, as defined in NUREG/CR-5849. A weighted mean is calculated for a defined area, and that average is compared against the guideline value. The weighted mean for total surface contamination measurements is based on the average surface contamination withirc a 2

2 contiguous 1 m area. For soil, the average is based on an area of 100 m. The equations and their use are described below.

For exposure rates, the mean (equation 5.1) is used instead of the weighted 2

mean. Exposure rates on structures can be averaged over 10 m and cutdoor 2

areas can be averaged over 100 m,

Surface Contamination When total surface contamination is between one and three times the guideline value, additional measurements will be performed to further characterize the level and areal extent of the contamination. The weighted mean is then calculated to determine whether the average contamination level within a contiguous 1 m area containing the elevated area is less than the guideline 2

value. Where the weighted mean exceeds the guideline value, or where an individual measurement exceeds three times the guideline value, the area will be remediated. Equation 5.3 is used for the calculation of the weighted mean.

Soil nnd Sediment When a radionuclide concentration is between one and three times the guideline value, additional measurements will le performed to determine the activity level and its areal extent. If the concentratwa in the elevated area exceeds the guideline value by a factor of (100/A)U2, where A is the area of the elevated activity in nf, die area must be remediated. If the concentration in the elevated area does not exceed the guideline value by a factor of (100/A)u2, the weighted mean activity in the 100 m contiguous area is then determined and compared 2

against the guideline value. If the weighted mean exceeds the guideline value, or an individual measurement exceeds three times the guideline value, the g

A-44

1 l

YNPS HNAL STATUS SURVEY PLAN REVISION 0 l

b elevated area will be remediated.- Equation 5.3 is used for the calculation of the weighted mean.

The weighted mean is calculated as follows:

E, = 1 x, 1-A, y,A,

(5.3)

+

n, i.

t.

g.:

where, i,

weighted mean including elevated area (s),

=

number of systematic and random measurements, n,

=

xi

= systematic and random measurements at point i, number of elevated areas, n,

=

2 2

fraction of 1 m (100 m for soil) occupied by elevated area k, and A,

=

O y,

= elevated area activity in area k.

N/

Note:

For plant systems and embedded piping surface contamination measurements, equation 5.3 will not be used. la these cases, equation 5.1 (a non-weighted mean) will be used to determine average contamination levels.

5.2.3 Calculation of the 95% Confidence Interval of the Mean After the mean levels are shown to satisfy the guideline values and conditions,

^

the results will be further evaluated to determine whether the data for each survey unit provides a 95 % confidence level that the true mean activity level meets the guidelines; The confidence interval is calculated using normal statistics (one-tailed test) at the 95% confidence level, as follows:

S g,= i+t,,,,

(5.4) 8

where, upper confidence limit of sample mean,

=

-O i

sample mean, S

=

g A-45

+

YNPS FINAL STATUS SURVEY PLAN REVISION 0 O

student t statistic for the degree of confidence and degrees of t.or

=

i freedom; df (degrees of freedom) is equal to n - 1, and a is 0.05 for this test, sample standard deviation, which includes the weighted means s,

=

for any elevated areas, and number of measurements in the survey unit.

n

=

The value p, will be compared with the guideline value. If p, is less than the guideline, the area being tested meets the guideline at a 95% confidence level.

This means that the probability is less than 5% that p, will pass the test when the true mean activity level exceeds the guideline value.

Where, is greater than the guideline value and the mean is less than the guideline value (i.e., the test of confidence is inconclusive), either (1) remediation will be conducted, followed by a re-survey, or (2) additional measurements / sampling will be performed.

5.2.4 Processing of Investigation Survey Data 9'

If during an investigation survey the initial measurement results cannot be confirmed, the initial measurement results will be replaced by the investigative data. The investigation measurement results will be considered as the fimal status survey of record and will be included in the calculation of the mean and standard deviation for the survey unit to determine compliance with the guideline values.

If during an investigation survey the initial measurement results are confirmed, and remediation and/or reclassification is required, the initial measurement results will be archived as characterization data in the final status survey database. The results of the post-remediation final status survey, or the final status survey subsequent to reclassification, will be considered as the final status survey of record and will be included in the calculation of the mean and standard deviation for the survey unit to determine compliance with the guideline values.

If during an investigation survey the initial measurement results are confirmed, but remediation and/or reclassification is not required, the initial measurement results and the investigation survey results will be evaluated by the weighted mean equation. The weighted mean value will replace the initial value and be considered as the final status survey result for the location of record. The new A-16

YNPS FINAL STATUS SURVEY PLAN REVISION 0

.1 O.

value will then be included in the calculation of the mean and standard deviation for the survey unit to determine compliance with de guideline values.

5.3 ALARA Analysis In the YNPS Final Safety Analysis Report, Yankee committed to a site release criterion of 15 mrem / year Total Effective Dose Equivalent (TEDE). Yankee also committed to an optimization process based on ALARA principles to reduce the residual levels of radioactivity on the site commensurate with the minimization of total risk. The ALARA analysis will take into account the following:

Radiation doses and environmental impacts from the decommissioning process 4

and from the residual radiation remaining on the site after completion of decommissioning.

Other costs and risks associated with the decontamination and decommissioning

{

of the site.

d See Section 4.4 of the License Termination Plan for more details.

O d

d

. m A-47 m

l YNPS FINAL STATUS SURVEY PLAN REVISION 0 TABLE 5.1 Summarv of Action Levels for Reclassification and Remediation*

SURVEY TYPE REMOVABLE TOTAL SURFACE SOIL and BULK MATERIAL CONTAMINATION CONTAMINATION CONCENTRATION EXPOSURE RATE Reclassification Reclassify if mean > 25% of a) Reclassify if single measurement Action Izvels**

NA GLV.

> 75% of SOR.

NA b) Reclassify if mean > 25% of SOR.

Remediation a) Remediate if single a) Remediate if single a) Remediate if single a) Remediate if single Action levels"*

measurement > GLV; measurement > 3 x GLV; measurement > 3 x SOR.;

measurement

> 10 R/hr; b) Remediate or take more b) Remediate if mean > GLV; b) Remediate if mean > SOR; samples if mean < GLV and b) Remediate if mean p > GLV.

c) Remediate if single c) Remediate if single

> 5 pR/hr; measurement > GLV, and rreasurement > SOR and mean 2

weighted mean over 1 m is of elevated area is c) Remediate or take

> GLV;

> (100/A)n: x SOR; more samples if mean

< 5 pR/hr and d) Remediate er take more samples d) Remediate or take more samples p, > 5 pR/hr.

if mean < GLV and p. > GLV.

if mean < SOR and p. > SOR.

The table is a summary. Refer to Section 5.2 for complete information on the above action levels. All measurements assume the appropriate background has been subtracted.

Reclassification applies only to unaffected or non-suspect affected areas. Unaffected Areas will be reclassified as Affected. Non-Suspect Affected Areas will be reclassified as Suspect Affected.

Reclassification will be required following remediation for Unaffected and Non-Suspect Affected Areas. As described in the text, some of the remediation action levels in this table allcw for the collection of additional samples as a possible alternative to remediation.

O W

G

YNPS FINAL STATUS SURVE'.' PLAN REVISION 0 7

.y I

)

' x)

pgo, Measuremerits and Sunpling 4

Co.wert Data to Standard Units 4

Calculate Survey Unit Mean and Standard Deviation 4

Compare with Reciassification Reclassify Crtteria Exceeds YES Reclassification CrPeria?

NO or N/A Remediate or Take More Compare with Remedation Samples Action Levels (O

'N-)

Exceeds YES Remediation Action Levels NO Calculate for 95%

Confidenca Level Collect Additional Data Compare With GLV YES*

Exceeds YES*

GLV7 GLV. Guideline Value MDC. Minimum Detectable Concentration

,NO

. Implies Choice of Actions Prepare Final Report FIGURE 5.1

,s r

\\

i

~'

Interoreting and Comnaring Survey Data with Guideline Valup A-49

YNPS FINAL STATUS SURVEY PLAN REVISION 0 6.0 FINAL STATUS SURVEY REPORTS Upon completion of each major final status survey phase an interim report will be prepared for review by the Nuclear Regulatory Commission. Upon completion of all final status survey phases, a final report will be prepared. These reports will meet the intent of draft NU. REG /CR-5849 and Regulatory Guide 1.86 for final status survey reporting.

6.1 Topical Outline The following outline illustrates a general format for the final report for the final status survey. The outline below may be adjusted to provide a clearer presentation of the information. The level of detail will be sufficient to clearly describe the final status survey program and certify the results.

1.0 Background Information 2.0 Site Description 3.0 Operating History 4.0 Decommissioning Activities 5.0 Final Status Survey Methodology 6.0 Final Status Survey Results g

7.0 Summary W

6.2 Reporting of Survey Findings 6.2.1 Detail Data Reporting for Each Survey Area A computer generated report for each survey area will be generated which will show all measurement results, both positive and negative. In addition, the reports will show the conversion factors, backgrounds and the MDCs to allow independent verification of the results. These measurement results may also be presented graphically. If requested by the NRC, these data may be provided in electronic form.

6.2.2 Summary Data Reporting for Each Survey Area A computer generated summary of the measurement results for each survey area will be provided. The number of measurements, the average and maximum values, and the upper limit of the confidence interval about the mean for each survey unit in a survey area will be reported for comparison to the release criteria. If requested by the NRC, these data may be provided in electronic form.

A-50

~. _ -

~.

YNPS FINAL STATUS SURVEY PLAN REVISION 0

'O 6.3 Final TEDE Evaluation 4

Following completion of the final status survey, the survey data will be used to produce l

an annual TEDE to an average member of the critical population group. An assessment tool, such as the computer code RESRAD or D&D Screen, will be utilized in making this dose analysis. All realistic pathways for exposure, including direct exposure, drinking water, and agriculture, will be included. The objective of this analysis will be to demonstrate that the annual dose to any real individual will be well below the 15 mrem.

t e

I i

.m c

A-51

-+

i YNPS HNAL STATUS SURVEY PLAN REVISION 0 Ol

7.0 REFERENCES

7.1 Survey Plan References 1.

U.S. Atomic Energy Commission, Regulatory Guide 1.86, " Termination of Operating Licenses for Nuclear Reactors," dated June 1974.

2.

Draft NUREG/CR-5849, Manual for Conducting Radiobgical Surveys in Support of License Termination, June 1992 (Draft).

3.

Draft NUREG-1500, Working Draft Regulatory Guide on Release Criteria for Decommissioning: NRC Staff's Draft for Comment 4.

YNPS Decommissioning Plan, Revision 0.0 i

5.

SECY-94-145, " Increase of Tritium and Iron-55 Unrestricted Use Limits for Surface Contamination at Shoreham and Fort St. Vrain," dated May 27,1994.

6.

YAEC letter, J.K. Thayer to M.B. Fairtile, dated August 5,1994, " Request for Additional Information on Yankee Nuclear Power Station's Decommissioning Plan and Decommissioning Environmental Report."

g 7.

NRC Letter, C.L. Pittiglie to A.C. Crawford (Public Service of Colorado),

" Response to Proposed Modification of Removable Surface Contamination Release Criteria of Removable Surface Contamination for Tritium and Iron-55 at Fort St. Vrain Nuclear Generating Station," dated January 18,1995.

8.

Yu, C.F. et al., Manual for Implementing Residual Radioactivity Materials Guidelines Using RESRAD, Environmental Assessment Division, Argonne National Laboratory.

9.

Draft NUREG-1505, A Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys, dated August 1995.

10.

Draft NUREG-1507, Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, dated August 1995.

i O

A-52

f i

YNPS FINAL STATUS SURVEY PLAN REVISION 0 7.2 Procedure References 5

1.

AP-8800, -

" Final Status Survey Organization, Training and Technical Basis Documents" -

2.

AP-8801, "FSS Survey Area Classification and Description" l

i 3.

DP-8802, "FSS Survey Location Designation" 4.

OP-8803, "FSS Package Preparation, Issuance and Closure" 5.

OP-8804, "FSS for Systems and Structures" 6.

OP-8805, "FSS for Open land Areas" 7.

OP-8807, "FSS Background Assessment" 8.

OP-8808, "FSS Data Receipt and Management" i

h 9.

OP-8809, "FSS Data Processing" Ob 10.

AP-8810, "FSS Survey Area Turnover and Control" j

11.

AP-8812, "FSS Quality Control" k

d 3

O A-53 9,

YNPS FINAL STATUS SURVEY PLAN REVISION 0 8.0 GLOSSARY 2

Accessible Surface Area - An area accessible to a 100 cm proportional detector or a 2

15.5 cm GM detector for direct or scanning measurements.

Administrative Action Level - A contamination level used as the investigation threshold in survey units to evaluate the need for additional investigation, reclassification or remediation.

Backgrour;d Radiation - Naturally occurring radiation which may include cosmic, terrestrial (radiation from the naturally radioactive elements) and man-made radiation from global fallout.

Characterization Survev - A radiological survey and its supporting evaluations performed to establish the YNPS radiological condition for planning decommissioning i

activities.

Confidence Interval-A range of values derived from a sample such that there is a l

probability that a population parameter being estimated, e.g., a mean value, lies within the range.

9I Confidence Level - The probability associated with a confidence interval which expresses the probability that the confidence interval contains the population parameter value being estimated.

Elevated Area - Areas of residual contamination exceeding the guideline value.

Final Status Survey - Radiological measurements, evaluations and supporting activities undertaken to demonstrate that the YNPS facility satisfies the criteria for unrestricted use.

Guideline Value - (GLV), residual radioactivity levels that equate to the site release criteria for that particular pathway or measurement.

Hard-to-Detect Nuclide - (HTDN), A 1. aionuclide emitting radiation (s) that are difficult to detect with field instrumentation.

History File - A compilation of information prepared for use in planning the final status survey of a survey area. It summarizes the operational history, characterization survey data, operational surveys and other information to help establish the basis for the design of the final status survey.

O A-54

YNPS FINAL STATUS SURVEY PLAN REVISION 0 O

V Mimmum Detectable concentration - The minimum activity concentration on a surface or material volume that can be statistically detected above background. This is usually set at the 95% confidence level.

Operational Radiological Survev - A radi0 logical survey performed in accordance with YNPS procedures. Operational surveys are distinct from, and usually performed prior to, final status surveys.

Release Criteria - A term used to identify the radiological requirements for release of the YNPS facility for unrestricted use.

Renlicate Survey - A survey which consists of repeat measurements on a specified fraction of survey areas. The survey areas are usually selected at random to provide an independent check of final status survey measurements.

Scan Survey - A qualitative radiological monitoring technique which is performed by moving a detector over a surface at a specified constant speed to detect elevated activity levels. Also called Surface Scan.

Structures - All YNPS site buildings and their surfaces. In addition, platforms,

(

restraints and supports, and external surfaces of piping systems, heating and ventilation C

systems, tanks, stacks, etc., are also treated as structures in the final status survey.

Sum of Ratios - Where more than one radionuclide is present, the sum of the ratios of each radionuclide concentration to its respective GLV should not exceed unity.

Surface Contamination - The total of both fixed and removable contamination. For the purposes of this Plan, this would also include any remaining neutron-activated material.

Also called total surface contamination.

Sunn Area - The basic entity for the management of the Final Status Survey. It is comprised of one or more survey units, the bounds of which are defined by existing facility physical features, such as a room, intersection of walls, column-and-row layout of a floor elevation, or structural I-beams.

Survey Location - In a structural or open land survey area, a survey location is usually represented by a single grid block. In a system survey area, a specified length of piping or a component such as a valve or tank is referred to as a survey location. A survey location can contain one or more survey points. Also referred to as a Measurement Location.

_ r]

Survcv Package - A collection of information in a standardized format for controlling V

and documenting field measurements Mken for the Final Status Survey. A survey package is prepared for each survey area. The survey package may include the survey A-55

YNPS FINAL STATUS SURVEY PLAN REVISION 0 instructions, a control form, grid map (s), survey measurement data sheets and survey maps.

Survey Point - A smaller subdivision within a survey location (grid block, system, component) where local measurements are taken. For stmetures and systems, a survey 2

point generally refers to an area covered by a detector, or an area of 100 cm when a smear is taken. For open land areas, a survey point refers to the area covered by a detector (for paved surfaces), the point at which an exposure rate measurement is taken, or the point at which a soil or pavement sample is collected.

Survey Unit - A contiguous area (usually) with similar characteristics and contamination potential. Survey units are established to facilitate the survey process and aid in the statistical evaluation of the survey data. Because survey units are generally designed to be contiguous areas with similar characteristics and potential for contamination, the actual size of a survey unit is not deemed to be critical, provided each survey unit contains a sufficient number of measurement locations.

Total Effective Dose Eauivalent (TEDE) - The sum of the deep dose equivalent (for external exposures) and the committed effective dose equivalent (for internal i

exposures).

O' O;

A-56